POSIVA 2012-12
Safety Case for the Disposal of
Spent Nuclear Fuel at Olkiluoto
- Synhesis 2012
Posiva Oy
December 2012
POSIVA OY
Olkiluoto
FIN-27160 EURAJOKI, FINLAND
Phone (02) 8372 31 (nat.), (+358-2-) 8372 31 (int.)
Fax (02) 8372 3809 (nat.), (+358-2-) 8372 3809 (int.)
POSIVA 2012-12
Safety Case for the Disposal of
Spent Nuclear Fuel at Olkiluoto
- Synthesis 2012
Posiva Oy
December 2012
POSIVA OY
Olkiluoto
FI-27160 EURAJOKI, FINLAND
Phone (02) 8372 31 (nat.), (+358-2-) 8372 31 (int.)
Fax (02) 8372 3809 (nat.), (+358-2-) 8372 3809 (int.)
ISBN 978-951-652-193-3
ISSN 1239-3096
Posiva-raportti – Posiva Report
Raportin tunnus – Report code
POSIVA 2012-12
Posiva Oy
Olkiluoto
FI-27160 EURAJOKI, FINLAND
Puh. 02-8372 (31) – Int. Tel. +358 2 8372 (31)
Julkaisuaika – Date
December 2012
Tekijä(t) – Author(s)
Toimeksiantaja(t) – Commissioned by
Posiva Oy
Posiva Oy
Nimeke – Title
SAFETY CASE FOR THE DISPOSAL OF SPENT NUCLEAR FUEL AT OLKILUOTO –
SYNTHESIS 2012
Tiivistelmä – Abstract
TURVA-2012 is Posiva’s safety case in support of the Preliminary Safety Analysis Report (PSAR
2012) and application for a construction licence for a spent nuclear fuel repository. . Consistent
with the Government Decisions-in- Principle, this foresees a repository developed in bedrock at
the Olkiluoto site according to the KBS-3 method, designed to accept spent nuclear fuel from the
lifetime operations of the Olkiluoto and Loviisa reactors.
Synthesis 2012 presents a synthesis of Posiva Oy’s Safety Case “TURVA-2012” portfolio. It
summarises the design basis for the repository at the Olkiluoto site, the assessment methodology
and key results of performance and safety assessments. It brings together all the lines of argument
for safety, evaluation of compliance with the regulatory requirements, and statement of
confidence in long-term safety and Posiva’s safety analyses.
The TURVA-2012 safety case demonstrates that the proposed repository design provides a safe
solution for the disposal of spent nuclear fuel, and that the performance and safety assessments
are fully consistent with all the legal and regulatory requirements related to long-term safety as set
out in Government Decree 736/2008 and in guidance from the nuclear regulator - the STUK.
Moreover, Posiva considers that the level of confidence in the demonstration of safety is
appropriate and sufficient to submit the construction licence application to the authorities. The
assessment of long-term safety includes uncertainties, but these do not affect the basic
conclusions on the long-term safety of the repository.
Avainsanat - Keywords
Safety case, safety assessment, synthesis, KBS-3V, Olkiluoto
ISBN
ISSN
ISBN 978-951-652-193-3
Sivumäärä – Number of pages
277
ISSN 1239-3096
Kieli – Language
English
Posiva-raportti – Posiva Report
Raportin tunnus – Report code
POSIVA 2012-12
Posiva Oy
Olkiluoto
FI-27160 EURAJOKI, FINLAND
Puh. 02-8372 (31) – Int. Tel. +358 2 8372 (31)
Julkaisuaika – Date
Joulukuu 2012
Tekijä(t) – Author(s)
Toimeksiantaja(t) – Commissioned by
Posiva Oy
Posiva Oy
Nimeke – Title
TURVALLISUUSPERUSTELU KÄYTETYN YDINPOLTTOAINEEN LOPPUSIJOITUKSELLE OLKILUODOSSA – SYNTEESIRAPORTTI 2012
Tiivistelmä – Abstract
Posiva Oy on laatinut käytetyn ydinpolttoaineen loppusijoituslaitoksen pitkäaikaisturvallisuutta
käsittelevän turvallisuusperustelun TURVA-2012 täydentämään loppusijoituslaitoksen alustavaa
turvallisuusselostetta (PSAR2012) ja rakentamislupahakemusta. Eduskunnan vahvistamien
periaatepäätösten mukaan Olkiluo¬don ja Loviisan voimalaitoksissa syntyvä käytetty
ydinpolttoaine loppusijoitetaan Olkiluodon kallioperään rakennettavaan KBS-3 menetelmän
mukaiseen loppusijoituslaitokseen.
Synthesis 2012 -raportti on synteesi TURVA-2012 turvallisuusperustelun muodostavista
raporteista. Raportissa on esitetty yhteenveto Olkiluotoon rakennettavan loppusijoituslaitoksen
suunnitteluperusteista, turvallisuusperustelun metodologiasta sekä toimintakykyanalyysin ja
turvallisuusanalyysin keskeisimmistä tuloksista. Raportissa on niin ikään esitetty yhteenveto
turvallisuutta tukevista perusteluista, arvio pitkäaikaisturvallisuuteen ja turvallisuusperustelua
koskevien viranomaisvaatimusten täyttymisestä sekä arvio pitkäaikaisturvallisuuden ja Posiva
Oy:n turvallisuusanalyysien luotettavuudesta.
Turvallisuusperustelu TURVA-2012 osoittaa, että käytetyn ydinpolttoaineen loppusijoitus
suunnitellulla tavalla on turvallista ja että toimintakyky- ja turvallisuusanalyysit vastaavat
valtioneuvoston asetuksessa 736/2008 ja Säteilyturvakeskuksen YVL-ohjeissa esitettyjä
pitkäaikaisturvallisuutta koskevia vaatimuksia. Lisäksi, Posiva Oy:n näkemyksen mukaan
loppusijoituksen turvallisuus on osoitettu riittävän luotettavasti rakentamislupahakemusta varten.
Loppusijoituksen pitkäaikaisturvallisuuden arviointiin liittyy epävarmuuksia, mutta näillä ei ole
vaikutusta
johtopäätöksiin
käytetyn
ydinpolttoaineen
loppusijoituslaitoksen
pitkäaikaisturvallisuudesta.
Avainsanat - Keywords
Turvallisuusperustelu, turvallisuusanalyysi, KBS-3V, Olkiluoto
ISBN
ISSN
ISBN 978-951-652-193-3
Sivumäärä – Number of pages
277
ISSN 1239-3096
Kieli – Language
Englanti
1
EXECUTIVE SUMMARY
TABLE OF CONTENTS
Spent nuclear fuel............................................................................................................2
22The TURVA-2012 safety case.....................................................................................3
Objective, audience and scope..............................................................................3
Quality assurance..................................................................................................4
The KBS-3 method and the Olkiluoto site.......................................................................4
The KBS-3 method.................................................................................................4
The Olkiluoto site....................................................................................................5
Legal and regulatory requirements..................................................................................6
Design methodology........................................................................................................8
Requirements management...................................................................................8
Safety principles, safety concept and safety functions...........................................9
Design basis and specifications...........................................................................11
Assessment methodology.............................................................................................12
System description...............................................................................................12
Features, events and processes..........................................................................13
Future lines of evolution.......................................................................................14
Performance assessment....................................................................................14
Formulation of release scenarios.........................................................................15
Safety assessment...............................................................................................16
Results of performance assessment.............................................................................18
Excavation and operation up to closure of the disposal facility...................... .....18
Post-closure period during the next 10,000 years................................................19
Evolution during repeated glacial cycles..............................................................21
Uncertainties in performance assessment...........................................................24
Assessment of radionuclide release scenarios.......................................................... ..25
Scenarios and cases............................................................................................24
Analysis of the base scenario Reference Case...................................................25
Analysis of variant scenarios................................................................................27
Analysis of disturbance scenarios........................................................................30
Complementary analysis......................................................................................31
Summary of results and uncertainties..................................................................32
Complementary considerations.....................................................................................34
Choice of geological disposal...............................................................................34
Support for the robustness of the KBS-3 method.................................................34
Support for the suitability of the Olkiluoto site......................................................35
Conclusions...................................................................................................................36
The main research and development needs during the coming years................36
Conclusions on compliance.................................................................................36
2
Spent nuclear fuel
The spent nuclear fuel that arises from the generation of electricity at the Loviisa and
Olkiluoto nuclear power plants is classified as nuclear waste. According to the Nuclear
Energy Act, including amendments, nuclear waste generated in Finland must be
processed, stored and disposed of in Finland.
Posiva Oy (Posiva) was established by Imatran Voima Oy (later Fortum Power and
Heat Oy) and Teollisuuden Voima Oy in 1995. Its mission is to implement the disposal
programme for spent nuclear fuel from the Loviisa and Olkiluoto nuclear power plants,
and to carry out related research, technical design and development.
In 2001, the Parliament endorsed a Decision-in-Principle (DiP) whereby the spent
nuclear fuel produced by the operating nuclear reactors at Olkiluoto and Loviisa will be
disposed of in a geological repository at Olkiluoto. Subsequently, additional DiPs were
issued allowing extension of the repository to accommodate spent nuclear fuel from the
operation of additional reactors that are under construction or are planned at Olkiluoto.
The safety case supporting the construction licence application to be submitted by the
end of 2012 foresees the disposal of spent nuclear fuel produced by these reactors
during their operating lifetime, in total 9000 tU.
Figure 1 provides a timeline for nuclear waste management for Olkiluoto and Loviisa
reactors in which the aim is to start the disposal of spent fuel around 2020.
Figure 1. Timeline for nuclear waste management relating to the Loviisa and Olkiluoto
reactors until 2020. The target is to begin disposal of spent nuclear fuel around 2020.
3
The TURVA-2012 safety case
Objective, audience and scope
TURVA-2012 is Posiva’s safety case in support of the Preliminary Safety Analysis
Report (PSAR 2012) and application for a construction licence for a disposal facility for
spent nuclear fuel at the Olkiluoto site.
It is addressed to the nuclear regulator, STUK, and other national stakeholders as well
as the international scientific and technical communities engaged in the discussion on
nuclear waste disposal. STUK will review the safety case and related topical reports as
part of its evaluation of construction licence application and the PSAR and give a
statement on the construction licence application, which will form a basis for the
Government judgement on issuance of the construction licence.
The TURVA-2012 safety case presents the arguments for the long-term radiological
safety of the planned disposal system. It includes:

a description of the spent nuclear fuel to be disposed of in the geological repository;

a description of the natural and engineered barrier system that the repository system
provides, a definition of the safety functions and targets set for these, and a
description of the present understanding of the processes that may affect the
evolution and performance of the spent nuclear fuel, engineered barriers, host rock
and the surface environment;

a performance assessment systematically analysing the ability of the repository
system to provide containment and isolation of the spent nuclear fuel for as long as
it remains hazardous;

a definition of the lines of evolution that may lead to failure of the canisters
containing the spent nuclear fuel and to the releases of radionuclides (scenarios);

analyses of the potential rates of release of radionuclides from the failed canisters,
the retention, transport and distribution of radionuclides within the repository
system and surface environment and the potential radiation doses to humans, plants
and animals including the associated uncertainties and an evaluation of their
impacts;

the models and data used in the description of the evolution of the repository system
and the development of the surface environment and for the analysis of activity
releases and dose assessment;

a range of qualitative evidence and arguments that complement and support the
reliability of the results of the quantitative analyses; and

a comparison of the outcome of the analyses with safety requirements.
Aspects of safety related to the period of operations are dealt with in other parts of the
PSAR.
The TURVA-2012 safety case is presented in a portfolio of safety case reports and
supporting documents, and a synthesis of these that brings together all the lines of
4
arguments for safety, including the main starting points, methodology, results and
conclusions.
Quality assurance
The quality of the TURVA-2012 safety case has been assured through documented
procedures in accord with Posiva’s quality management principle, which is based on the
ISO 9001:2008 standard. A graded approach is applied whereby the primary emphasis
is on quality control of those activities that have a direct bearing on safety.
The overall plan, goals and constraints for the TURVA-2012 safety case production
process are presented in Posiva’s Safety Case Plan 2008. The organisation of the
TURVA-2012 safety case production process is referred to as SAFCA. The details of
how the Safety Case Plan is being implemented are described in the SAFCA project
plan. The work is managed and coordinated by a SAFCA project group and supervised
by a steering group.
A SAFCA quality co-ordinator has been designated for activities related to quality
assurance measures applied to the production of the safety case. Improvements are
made to the process as deemed useful or necessary. The quality co-ordinator is also
responsible for the coordination of the expert reviews, maintenance of schedules, and
review and approval of the reports.
Posiva’s quality manager undertakes regular auditing of the safety case production
process.
The KBS-3 method and the Olkiluoto site
The 2001 DiP states that disposal of spent nuclear fuel shall take place in a geological
repository at the Olkiluoto site, developed according to the KBS-3 method.
The KBS-3 method
The KBS-3 method was conceived as a solution for the disposal of spent nuclear fuel in
Sweden in the early 1980s. Since then, the method has been developed and its key
elements tested by SKB in Sweden and Posiva in Finland, and in joint projects. The
method envisages disposal of spent nuclear fuel within a system of multiple barriers,
which consists of engineered barriers and the natural barrier provided by the host rock.
Posiva’s reference design is based on the emplacement of canisters containing the spent
nuclear fuel in vertical deposition holes (KBS-3V). Posiva is jointly with SKB
developing a potential alternative design where multiple canisters are emplaced
horizontally in deposition drifts (KBS-3H). The present safety case is based on the
reference design.
The repository is constructed on a single level with the floor of the deposition tunnels at
a depth of between 400 and 450 m below the ground surface in the Olkiluoto bedrock
(Figure 2).
5
Figure 2. Schematic illustration of the KBS-3V design.
In the reference design, the spent nuclear fuel assemblies are placed into copper
canisters with cast iron load-bearing inserts, and the canisters are emplaced vertically in
individual deposition holes bored in the floors of the deposition tunnels. The canisters
are surrounded by a swelling clay buffer material that separates them from the bedrock.
The deposition tunnels, central tunnels, access tunnel and other underground openings
are backfilled with materials that help to restore the natural conditions in the bedrock
after operations.
The Olkiluoto site
The Olkiluoto site, located on the coast of south-western Finland, has been investigated
as a potential site for geological disposal of spent nuclear fuel for over 25 years. This
has included the construction of an underground rock characterisation facility − the
ONKALO. Olkiluoto Island has an area of about 10 km2; the surface facilities including
the encapsulation plant will occupy about 0.1 km2; according to the current design and
required capacity, the deposition tunnels and other tunnels will occupy about 2 km2.
The characterisation of the Olkiluoto site is focused on a volume of bedrock situated
between 400 and 500 metres below ground. At this depth, favourable and predictable
bedrock and groundwater conditions are found. In addition, the likelihood of inadvertent
human intrusion is low.
6
Key features of the Olkiluoto site with respect to its suitability for geological disposal of
radioactive waste include:

a stable tectonic situation within the Fennoscandian Shield, away from active plate
margins;

good quality crystalline bedrock suitable for the excavation of self-supporting
tunnels and other underground openings, such as deposition holes, technical rooms
and shafts;

reducing conditions at disposal depth and also otherwise favourable geochemical
characteristics of the groundwater; and

low groundwater flow at depth occurs currently, as it has occurred over a long
period in the past, and is expected to persist for a long period into the future.
The conditions in the Olkiluoto bedrock provide favourable conditions for longevity and
reliable functioning of the engineered barrier system (EBS). In addition, the low
groundwater flows, and physical and chemical retardation processes, limit the
movement of radionuclides.
Key features and processes that provide constraints on the layout of the repository and
other underground openings, or that must be taken into account in the assessments of
long-term performance and safety, include:

presence of deformation and fractured zones, displaying more mixed geotechnical
properties and in some cases increased hydraulic activity;

higher rock stress at depth which may cause disturbance to the rock, making
underground openings less stable;

temperature and thermal conductivity of rock and residual heat output of the spent
nuclear fuel;

high salinity of groundwater at depth, which may affect the performance of the
engineered barriers;

continuing post-glacial crustal uplift and, in the longer term, climatic cooling and
glaciation leading to changes in rock stress and potential changes in groundwater
flow and hydrochemistry, e.g. influx of dilute glacial melt waters into the host rock.
Legal and regulatory requirements
The basis for the use of nuclear energy in Finland is given in the Nuclear Energy Act
(YEL 990/1987) and Nuclear Energy Decree (YEA 161/1987), which came into effect
in 1988. According to the Nuclear Energy Act:
Nuclear waste shall be managed so that after disposal of the waste no radiation
exposure is caused, which would exceed the level considered acceptable at the time the
final disposal is implemented.
The safe management of nuclear waste is the responsibility of the utilities that generate
the waste. The responsible parties must submit reports to the Ministry of Trade and
Employment every three years. These reports include a description of the measures
7
taken towards implementation of nuclear waste management during the last three-year
period, as well as an outline of the plans for the next three years. The most recent report
was submitted in September 2012.
The schedule for the disposal of spent nuclear fuel was first defined by the Government
in 1983 and slightly modified by the Ministry of Trade and Industry (KTM)1 in 2003.
According to the Ministry decision, the parties under the nuclear waste management
obligation shall, separately, together or through Posiva Oy, present all reports and plans
required to obtain a construction licence for a disposal facility for spent nuclear fuel by
the end of 2012. The disposal facility is expected to become operational around 2020.
Government Decree 736/2008 sets the legal requirements regarding the safety of
disposal of spent nuclear fuel. The Radiation and Nuclear Safety Authority (STUK)
issues guidance documents on the fulfilment of the requirements set in the Government
Decree. These guides also set out STUK’s expectation for the content, quality and
criteria to be met by a safety case submission for disposal of nuclear waste. A total of
five Guides apply to the disposal of spent nuclear fuel. The most relevant here is Guide
YVL D.3, which provides guidance on the handling, storage and encapsulation of spent
nuclear fuel and YVL D.52, which provides guidance on the planning of the disposal
method, design and operation of the disposal facility, safety requirements and
demonstration of compliance with safety requirements, regulatory control and on the
compilation of a safety case.
Guide YVL D.5 applies to disposal of all types of nuclear waste and provides guidance
related to operational and long-term safety. Key requirements, stemming from
GD 736/2008 and set out in Guide YVL D.5, are summarised in Table 1. The Guide
provides substantial additional information on the meaning of these requirements and
the evidence needed to show compliance.
The Guide YVL D.5 does not specify the precise time frames over which assessments
are needed. Posiva consider, however, that radiation doses can be assessed, assuming
human habits, nutritional needs and metabolism remain unchanged, with sufficient
reliability over a period of up to 10,000 years, and that the fulfilment of the safety
functions of the repository system and the release of radionuclide to the surface
environment can be reasonably assessed up to one million years after repository closure.
Table 1. Synthesis of key requirements for long-term safety from STUK’s Guide
YVL D.5. Please refer to the Guide for actual wording and context.
Related to long-term radiological impacts


For expected evolution scenarios, and in the period during which the radiation exposure can be
assessed with sufficient reliability (at least over several millennia):

the annual dose to the most exposed people shall remain below the value of 0.1 mSv;

the average annual doses to other people shall remain insignificantly low.
In the longer term, the radiation impacts arising from disposal can at a maximum be equivalent to
those arising from natural radioactive substances in Earth’s crust, and on a large scale should remain
1
Now Ministry of Trade and Emplyoment
2
The Guides YVL D.3 and YVL D.5 is are available in draft form. STUK has agreed that the licence application can be based on
draft version 4 of both Guides (version 17.3.2011 has been used).
8



insignificantly low. The nuclide-specific constraints on releases to the environment (average release
of radioactive substances per annum) are specified in YVL D.5.
For the activity releases that arise from the expected evolution scenarios, the sum of the ratios
between the nuclide-specific activity release rates and the respective constraints given in YVL D.5
shall be less than one (this is evaluated through the release rates for radionuclides from the
geosphere to the biosphere),
The importance of unlikely events impairing long-term safety shall be assessed, and whenever
practicable, the radiation impacts caused shall be assessed quantitatively. The resulting annual
radiation dose or activity release shall be calculated and multiplied by its estimated probability of
occurrence. The obtained expectation value shall be below the dose constraint (see above) or release
constraints given in Table 2-4.
The assessed radiation exposures to fauna and flora shall remain clearly below the levels that could
cause decline in biodiversity or other significant detriment to any living population.
Related to providing long-term safety





Disposal shall be implemented in stages, with particular attention paid to aspects affecting long-term
safety.
The long-term safety of disposal shall be based on safety functions achieved through mutually
complementary barriers so that a deficiency of an individual safety function or a predictable geological
change will not jeopardise the long-term safety.
Targets shall be specified for the performance of each safety function based on high quality scientific
knowledge and expert judgement.
For spent fuel, the safety functions provided by the engineered barriers shall limit effectively the
release of radioactive substances into bedrock for at least 10,000 years.
The characteristics of the host rock shall be favourable for the long-term performance of engineered
barriers and with respect to the groundwater flow regime at the disposal site.
Design methodology
Requirements management
Posiva has developed a robust design for geological disposal of spent nuclear fuel at
Olkiluoto through a formal requirements management system (VAHA). This provides a
rigorous, traceable method of translating safety principles and the safety concept to a set
of safety functions, performance requirements, design requirements and design
specifications for the various barriers, i.e. a specification for enactment of the disposal
concept at the Olkiluoto site. The VAHA sets out:

At Level 1, stakeholder requirements that come from laws, decisions-in-principle,
regulatory requirements, and other stakeholder requirements;

At Level 2, the long-term safety principles, which lead to the definition of the safety
concept and safety functions;

At Level 3, the performance requirements, consisting of performance targets for the
engineered barriers and target properties for the host rock, such that the safety
functions are fulfilled;

At Level 4, the design requirements for the engineered barriers and the underground
openings including rock suitability classification criteria (RSC criteria), such that
the performance requirements will be met;

At Level 5, the design specifications, which are the detailed specifications to be
used in the design, construction and manufacturing.
9
Safety principles, safety concept and safety functions
The long-term safety principles set out for the KBS-3 method are based on the use of a
multi-barrier disposal system consisting of engineered barriers and host rock. The
engineered barrier system consists of the canister, buffer, backfill of the deposition
tunnel and closure. The role of the engineered barriers is to provide the primary
containment against the release of radionuclides. The host rock should provide
favourable conditions for the long-term performance of the engineered barriers, but also
limit or retard the transport of radionuclides. The multi-barrier system as a whole should
be able to protect the living environment even if one of the barriers turns out to be
deficient.
The safety concept (Figure 3) is a conceptual description of how these principles are
applied to achieve safe disposal of spent nuclear fuel in the present-day and future
conditions of the Olkiluoto site.
Containment of the radionuclide inventory associated with the spent nuclear fuel is
provided first and foremost by encapsulating the fuel in sealed (gas-tight and watertight) copper-iron canisters. The other EBS components (buffer, backfill and closure)
provide favourable near-field conditions for the canisters to remain intact and, in the
event of canister failure, slow down and limit releases of radionuclides from the
canister. The containment of radionuclides is ensured by the proven technical quality of
the EBS. Other elements of the safety concept include sufficient depth for the repository,
favourable and predictable bedrock and groundwater conditions and well-characterised
material properties of both the bedrock and the EBS. A robust system design ensures
that single deficiencies in the design or implementation of the design, or uncertainties in
future conditions, do not lead to significant weakening of the overall safety of the
repository system.
Safety functions are assigned to the components of the engineered barrier system (EBS)
and the host rock as shown in Table 2.
Most of the activity in the spent nuclear fuel is contained in a ceramic matrix (UO2) that
is resistant to dissolution in the expected repository conditions. The slow release of
radionuclides from the spent fuel matrix in the event of canister failure is part of
Posiva’s safety concept. However, no safety functions or performance requirements (see
below) are assigned to spent nuclear fuel; rather, the properties of the spent fuel are used
as a starting point in the design of the disposal system. Posiva is responsible for the
disposal of all spent nuclear fuel from Olkiluoto and Loviisa power plants, and if the
current design of the disposal system would not provide of a sufficient level of safety
for disposal of a possible new specific fuel type, the design will be modified to meet the
safety requirements.
10
SAFE DISPOSAL
FAVOURABLE, PREDICTABLE BEDROCK
AND GROUNDWATER CONDITIONS
PROVEN TECHNICAL QUALITY
OF THE EBS
Slow diffusive
transport in the buffer
Slow release from the
spent fuel matrix
Retention and retardation of
radionuclides
Slow transport in the
geosphere
FAVOURABLE NEAR-FIELD
CONDITIONS FOR THE
CANISTER
LONG-TERM ISOLATION AND CONTAINMENT
WELL-CHARACTERISED MATERIAL
PROPERTIES
SUFFICIENT DEPTH
ROBUST SYSTEM DESIGN
Figure 3. Outline of the safety concept. Orange blocks indicate the primary safety
features and properties of the disposal system. Green blocks indicate secondary safety
features that become important in the event of a radionuclide release from a canister.
Table 2. Safety functions assigned to the barriers (EBS components and host rock) in
Posiva’s KBS-3V repository.
Barrier
Safety functions
Canister

Ensure a prolonged period of containment of the spent nuclear fuel. This safety
function rests first and foremost on the mechanical strength of the canister’s cast
iron insert and the corrosion resistance of the copper surrounding it.
Buffer

Contribute to mechanical, geochemical and hydrogeological conditions that are
predictable and favourable to the canister.
Protect canisters from external processes that could compromise the safety function
of complete containment of the spent nuclear fuel and associated radionuclides
Limit and retard radionuclide releases in the event of canister failure.


Deposition
tunnel backfill



Host rock



Closure



Contribute to favourable and predictable mechanical, geochemical and
hydrogeological conditions for the buffer and canisters.
Limit and retard radionuclide releases in the possible event of canister failure.
Contribute to the mechanical stability of the rock adjacent to the deposition tunnels.
Isolate the spent nuclear fuel repository from the surface environment and normal
habitats for humans, plants and animals and limit the possibility of human intrusion,
and isolate the repository from changing conditions at the ground surface.
Provide favourable and predictable mechanical, geochemical and hydrogeological
conditions for the engineered barriers.
Limit the transport and retard the migration of harmful substances that could be
released from the repository.
Prevent the underground openings from compromising the long-term isolation of the
repository from the surface environment and normal habitats for humans, plants and
animals.
Contribute to favourable and predictable geochemical and hydrogeological
conditions for the other engineered barriers by preventing the formation of
significant water conductive flow paths through the openings.
Limit and retard inflow to and release of harmful substances from the repository.
11
Design basis and specifications
The definition of the performance targets for the safety functions of the engineered
barriers and the target properties for the safety functions of the host rock requires the
identification of the different loads and interactions that may act on the repository
system at the time of canister emplacement and in the long term. To achieve this, the
potential future conditions have to be described as alternative lines of evolution, and
their likelihoods are assessed based on present-day understanding and the findings of
earlier assessments. All the lines of evolution and expected loads that are judged
reasonably likely to occur (based on this understanding and previous findings) are taken
into account and, hence, included in the design basis. Thus, by definition, when the
performance targets and target properties are met and the future follows the reasonably
likely lines of evolution (design basis scenarios), the safety functions are fulfilled.
From the performance targets and target properties (VAHA level 3) the design
requirements are derived (VAHA level 4). Then, design specifications are worked out
such that the fulfilment of these requirements can be verified at implementation (VAHA
level 5). Performance assessment shows that the system, as designed and built according
to the design requirements and specifications, will meet the performance targets and
target properties and thus that the safety functions will be fulfilled for an envelope of
future conditions that includes all reasonably likely lines of evolution.
In defining the performance targets for the engineered barriers, implementation aspects
have to be considered. The performance targets have to be set considering, on the one
hand, the long-term safety aspects and, on the other hand, that the design and
implementation must be robust, as that is the foundation of the safety concept.
For the rock barrier, the target properties set the starting point for the definition of the
Rock Suitability Classification system (RSC) developed by Posiva. The classification
system includes both the updated rock suitability criteria as well as the procedure for the
suitability classification during the construction of the repository. The RSC is used to
identify suitable rock volumes for repository panels and to assess the suitability of
deposition tunnels for locating deposition holes and to accept deposition holes for
disposal.
The performance targets and target properties, together with the derived design
requirements and the underlying design basis scenarios, form the design basis of the
repository. The background and premises for the design basis are presented in Design
Basis.
A repository system designed and built according to the design basis is expected to
comply with the regulatory safety requirements. However, the safety case also includes
a discussion of situations in which the system does not meet all the requirements, where
there are uncertainties in whether they are met, or if the system follows an unlikely line
of evolution.
12
Assessment methodology
Figure 4 outlines the approach to the development of the safety case, whereby the
design basis is developed, the performance of the repository system assessed, and
scenarios leading to radionuclide release are formulated and assessed. The design basis
and definition of performance targets and target properties are developed iteratively
between performance assessment, formulation and assessment of radionuclide release
scenarios and presentation of the safety case. Available scientific understanding,
including the results from earlier assessments, is used in the definition of the
performance targets, target properties for the host rock, design requirements and criteria
for rock classification. These will be updated as scientific understanding is further
developed, taking into account the results of the performance assessment and
assessment of radionuclide release scenarios of the current safety case (the two-way
arrows in Figure 4).
System description
An accurate and reliable description of the disposal system is the foundation both for the
development of a robust design (previous section), and for an understanding of the
possible lines of evolution of the disposal system, assessments of performance and
safety, and complementary considerations that comprise the safety case. The results of
the safety case are used to specify the further development of the disposal facility, if
needed.
Characterisation studies of the Olkiluoto site have been made for over 25 years. This
has lead to a detailed description and understanding of the site in respect of all
characteristics relevant to the construction of a repository for spent nuclear fuel and to
its long-term evolution. Studies of the surface environment of the site form the basis for
a description of the biosphere sufficient to characterise the environment to be protected
and its potential future use and occupation by humans, plants and animals. Descriptions
of the site and surface environment are provided in Site Description and Biosphere
Description.
The KBS-3 method and the KBS-3V design have been developed over more than 30
years. The specific realisation of the design as planned for implementation of a
repository at the Olkiluoto site is the result of thorough analyses of the functional
requirements of the engineered barriers and host rock and of the overall safety of the
repository system. Detailed descriptions of the components of the repository system and
evidence concerning their practical realisation and feasibility have been compiled in
production line reports. The main characteristics, initial state including uncertainties of
the repository system components (spent nuclear fuel, EBS and host rock) and of the
surface environment to be used as input to the safety assessment have been compiled in
Description of the Disposal System.
13
Figure 4. Approach to the development of the safety case (FEP= Features, Events and
Processes, PSAR= Preliminary Safety Analysis Report, FSAR = Final Safety Analysis
Report).
Features, events and processes
Identifying and describing the features, events and processes (FEPs) that are relevant to
the evolution of the disposal system, or to its potential performance and safety, is an
essential step towards ensuring comprehensiveness of the assessments and safety case.
For the TURVA-2012 safety case, the identification and screening of FEPs has been
carried out by a team of scientific subject and assessment experts, based on a review of
the FEPs considered in Posiva’s previous assessments, the NEA FEP-database and the
FEPs considered in safety cases in other nuclear waste programmes, as well as an
examination of the specific characteristics of the disposal system and the Olkiluoto site.
A FEP database has been developed providing a structured classification of relevant
FEPs and couplings between these. The FEPs are presented in Features, Events and
Processes, including a description of each FEP and the fundamental uncertainties based
on current scientific understanding. The relevance of each FEP for the long-term safety
of the disposal facility to be constructed at Olkiluoto has been evaluated based on the
situations in which the FEP could occur at the Olkiluoto site and its couplings to other
FEPs.
Spent nuclear fuel must be kept isolated for as long as it could cause significant harm to
the normal habitats for humans, animals and plants. In TURVA-2012 an assessment
time frame of up to one million years into the future is considered. This is consistent
with other assessments of spent nuclear fuel disposal internationally. At one million
14
years, the activity of the spent nuclear fuel is similar to that of the original uranium ore
from which the fuel was fabricated.
Future lines of evolution
The understanding of FEPs is used to develop descriptions of future lines of evolution
of the repository system (the engineered barriers and host rock) and of the surface
environment. This provides the framework for estimating the thermal, hydraulic,
mechanical and chemical (THMC) loads that will be placed on the system.
During the construction and operation of the repository up to its closure, the main
changes are related to excavation effects and draining of water from the underground
openings, plus introduction of radiogenic heat from the spent nuclear fuel. Some limited
mechanical damage immediately around the openings, as well as an increase of
groundwater flow into the repository volume and changes in hydrochemistry are
expected. After closure, groundwater flow will return towards preconstruction
conditions, although modified by radiogenic heat from the spent fuel for a time. Salinity
will be reduced in the longer term due to infiltration of meteoric waters.
In the longer term, the main driver for change is climate evolution, where the expected
case is a continuation of glacial-interglacial cycling as experienced over the last one
million years of the Quaternary. However, best scientific understanding indicates the
past and continuing anthropogenic emissions of CO2 and other greenhouse gases will
lead to increased global temperatures over a period of many thousands of years,
delaying the onset of cooler climate conditions. Thus, over the next 50,000 years,
conditions are expected to remain essentially as today, i.e. a temperate climate with a
boreal ecosystem. A first cold period is not expected until about 50,000 years after
present (50 ka AP) with temperature and precipitation changes leading to permafrost
development and, later on, to ice-sheet development. For the assessment, from 50 to 170
ka AP, a repetition of the sequence of events during the last glacial cycle is assumed.
After 170 ka AP, seven repetitions of the cycle from 50 ka to 170 ka AP are assumed.
Thus, a total of eight glacial cycles are accounted for in the assessment time frame (up
to one million years after present). Variations in the duration and intensity of individual
glacial cycles are not expected to have a significant impact on repository safety.
Performance assessment
Performance assessment shows that the system, designed and built according to the
design requirements and specifications, is compliant with the performance targets and
target properties initially and in the long term and that the safety functions will be
fulfilled (Table 2), which, being so, will lead to isolation of the spent nuclear fuel, and
complete containment over hundreds of thousands of years and even for the one million
year time frame.
The performance of the repository system is analysed and the fulfilment of performance
requirements is evaluated taking into account the expected thermal, hydraulic,
mechanical and chemical evolution of the repository system, and uncertainties in the
expected lines of evolution. Less expected lines of evolution, including the possibility of
disruptive events, are also identified. Account is taken of the natural evolution of the
environment, chiefly driven by climatic evolution, which imposes external loads on the
15
repository system, and also internal loads, chiefly from the effects of excavation and
emplacement of the spent nuclear fuel and the engineered barriers.
The performance is considered below in three time windows: (1) during the excavation
and operational period up to closure; (2) up until 10,000 years after closure; (3) beyond
10,000 years over repeated glacial cycles.
The fulfilment of performance targets and target properties in each time window is
assessed considering time-dependent and space-dependent loads on the engineered
barriers and host rock. Quantitative assessments are made whenever possible, e.g. to
calculate safety margins and demonstrate the robustness of the design. Uncertainties are
highlighted, conditions that could lead to deviations from performance targets and target
properties are identified, and the likelihood and effects of the deviations estimated
whenever possible. In particular, conditions and events (incidental deviations) that could
lead to the release of radionuclides are identified; these are taken forward to the
formulation of radionuclide release scenarios and to the radiological impact assessment
of radionuclide releases.
Formulation of release scenarios
Consistent with the regulatory guidance (Table 1 and Guide YVL D.5), Posiva
distinguishes between the expected evolution of the disposal system and unlikely
evolutions and events.
The repository system is designed in a way that for the expected lines of evolution of
the system, each component of the EBS meets the performance targets assigned to it,
and the host rock conforms to its target properties. In this case the copper canisters (with
iron inserts) remain intact for the whole assessment time frame and there is no release of
radionuclides. This is confirmed in performance assessment.
Performance assessment shows, however, that there are some plausible conditions and
(incidental deviations) that could lead to the reduction of one or more safety functions,
and thus may give rise to radionuclide releases. In addition, there are some unlikely
events and processes that could disrupt the repository, e.g. related to human intrusion
and rock shear. These incidental deviations and unlikely events are systematically
examined to define a set of scenarios that encompass the combinations of initial
conditions, evolution and disruptive events.
In the current and past assessments by Posiva, the scenario of a canister with an initial
penetrating defect has been considered to test the radiological performance of the other
engineered barriers and host rock. This defect is most likely in the weld. Although the
likelihood that a canister with an initial undetected penetrating defect will be emplaced
is low, this is a useful base scenario for safety assessment (radionuclide release
calculations) against which the efficiency of the other engineered barriers and the host
rock to limit the radionuclide releases can be tested and that also complies with
GD 736/2008.
The classification of scenarios in TURVA-2012 is illustrated in Figure 5. The base
scenario addresses the most likely lines of evolution (in which the performance targets
and safety functions are met), but takes into account the possibility of the emplacement
16
Figure 5. Classification of scenarios in TURVA-2012, which is consistent with STUK’s
Guide YVL D.5.
of one or a few canisters with initial undetected penetrating defects. The variant
scenarios address situations that are considered reasonably likely and in which there
may be reduced performance of one or more safety functions of the barriers.
Disturbance scenarios address the lines of evolution that are considered unlikely but
cannot be completely eliminated.
Safety assessment
The aim of safety assessment is to analyse the radionuclide release, transport and
radiological impacts of the identified scenarios, and scenario combinations, and to
compare the calculated impacts with regulatory criteria in order to judge their
acceptability. In the evaluation of the releases due to unlikely scenarios, the likelihood
of the assumed conditions or events is considered.
The main safety indicators calculated in TURVA-2012 are the following.
1. The radioactive releases from the bedrock to the biosphere (surface environment),
which are calculated for all release scenarios and assessed against the nuclidespecific constraints for the radioactive releases to the environment (average annual
release rates of radioactive substances) defined in YVL D.5.
2. Annual doses3 to humans. Consistent with regulatory guidance (Table 1) these are
calculated for scenarios that give rise to releases to the surface environment in the
first 10,000 years.
3. Absorbed dose rates to plants and animals calculated for releases to the surface
environment in the first 10,000 years.
The repository system is analysed using models that represent:

release from the spent nuclear fuel (taking account of the locations of radionuclides
in the fuel, its cladding and other parts of the fuel element);
3
In this report, annual dose refers to the sum of the effective dose arising from external radiation within the period of one year, and
the committed effective dose from the intake of radioactive substances within the same year (GD 736/2008). Furthermore “dose”
refers to effective dose, unless otherwise explicitly stated.
17

release, retention and transport in the near field (release from the canister, migration
through the buffer, migration by alternative routes to water-conducting fractures in
the host rock); and

retention and transport in the geosphere (through water-conducting fractures taking
account of variability in flow paths).
This yields radioactive releases from the geosphere to the biosphere, which are used as
input to biosphere models. Since the repository system models are run independently of
the biosphere models, the output from a single repository system calculation can be
input to alternative biosphere models so as to represent alternative surface environment
conditions at the time of release.
Modelling for biosphere assessment includes, first, a screening process to identify those
radionuclides that could make significant contributions to the total radiological impact.
These radionuclides are carried forward to detailed biosphere modelling, based on a
model of the future landscape and ecosystem development in the Olkiluoto area over the
next 10,000 years. This provides the framework for modelling of radionuclide
movements within compartments of the future surface environment and calculation of
the radiation doses to humans, plants and animals, inhabiting or making use of the
various areas and resources that may become contaminated.
Calculation cases analyse the radiological impacts and illustrate the impact of specific
uncertainties or combinations of uncertainties. These are uncertainties related to the
scenario definitions, alternative model representations and data used in the models. Four
types of calculation cases are distinguished:

A Reference Case is one model realisation of the base scenario. Models and data for
the Reference Case are, in most instances, selected to be either realistic or
moderately cautious, i.e. radiological impacts are not to be underestimated nor
excessively overestimated.

Sensitivity cases represent alternate models and/or data to those of the Reference
Case, but remain within the scope of the base scenario and/or variant scenarios.
Analyses of the sensitivity cases illustrate the effect of model and data uncertainties.

What-if cases are mainly model representations of disturbance scenarios. Models
and data for these what-if cases are selected to represent unlikely events and
processes.

Complementary cases are designed to develop a better understanding of the
modelled system or subsystems and to test robustness.
All of the above cases are analysed deterministically, i.e. calculations are carried out for
a specific set of input parameter values. In addition, the disposal system behaviour is
explored by Monte Carlo simulations and probabilistic sensitivity analysis (PSA). This
involves performing a set of calculations in which the values of input parameters are
selected randomly from specified probability density functions (PDFs) that represent the
uncertainty in each parameter. The outcomes of multiple simulations are assembled into
a probability distribution and the sensitivity of outcome to inputs is investigated by
statistical techniques.
18
Results of performance assessment
Excavation and operation up to closure of the disposal facility
In the period of excavation and operation, groundwater flow modelling indicates an
increase of approximately two orders of magnitude in flow rates in the rock volume
surrounding the repository from a pre-construction baseline. Following installation of
tunnel and shaft backfill and seals (i.e. after closure), modelled flow rates return to near
pre-excavation rates; however, a few deposition holes with flow rates and transport
resistances outside the range defined by the target values may remain. During the ‘open’
period, the average salinity around the repository remains similar to the pre-construction
phase, but increased groundwater flow into the repository volume may lead to mixing of
water and either more dilute or more saline conditions locally at the repository depth.
The disturbed conditions are related to the main hydrogeological zones and the
ONKALO facility, not necessarily to the repository panels themselves. Moreover, the
disturbed conditions are likely to last a limited time; in the order of tens of years, and
thus the impact on the performance of the buffer and backfill is limited. The changes in
groundwater salinity remain consistent with target properties, meaning that the buffer
and backfill functions are preserved in most deposition holes.
Calculations of temperature evolution show a maximum temperature at the canister
surface of 95 °C assuming an unsaturated buffer and 75 °C for a saturated buffer. The
maximum rock temperature at the deposition hole wall is about 65 °C at 40 years after
emplacement. Thus, temperatures will remain within the performance targets.
Excavation will cause a damaged zone (EDZ) to form, especially below the tunnel
floors, although the damage is probably not continuous. In addition, excavation and the
heat produced by the spent nuclear fuel may cause spalling or other types of stressinduced damage around the excavated openings. The uncertainties concerning the
properties of the EDZ and the rock damage around the deposition holes are taken
account of in groundwater flow modelling.
Before full saturation, some buffer and backfill material may be lost through piping and
erosion. Based on calculated inflows to deposition holes, some limited buffer loss is
expected in roughly one third of the positions, but in all cases the average buffer density
remains consistent with the performance target, so that the necessary low hydraulic
conductivity and sufficient swelling pressure will be achieved as the buffer saturates. It
is estimated that 13,000 kg of the backfill material could at most be lost locally by
piping and erosion, and redistributed within the deposition tunnel. This is rather small
compared with the total mass of backfill material in the tunnel (more than 8000 tonnes
in a 300 m long deposition tunnel). The effect on the backfill performance depends on
how the mass loss is distributed in the backfill. Backfill loss will be largest in the
vicinity of fractures with a high enough inflow to transport the mass further down the
tunnel. However, this type of erosion would not be detrimental for the EBS as no
deposition holes would be located near such a fracture. Thus, the buffer and backfill will
remain consistent with their performance targets even considering the process of piping
and erosion.
19
For both unsaturated and saturated conditions, the consumption of oxygen in the
backfill and buffer will be relatively rapid, due to its reaction with pyrite and other
accessory minerals. Thus, anoxic, reducing conditions will be quickly established
around the emplaced canisters and throughout the buffer and backfill. Cementitious
leachates from grouting of fractures, from grout used to stabilise rock bolts and from the
plug in the deposition tunnel may locally affect the backfill. However, no cement is in
direct contact with the buffer and the flux of cementitious leachates reaching the buffer
is estimated to be of little significance.
The maximum corrosion depth of the copper canisters from the atmospheric and
initially trapped oxygen is expected to be less than 0.5 mm.
A probabilistic analysis has been carried out to assess the potential number of defective
canisters that could be placed in the repository. The conclusion was that the data
currently available are not enough for a statistical evaluation of the probability that the
penetrating defects are detected before emplacement. Consequently, the probability of
detection can only be based on expert judgement, taking account of the results from
both the non-destructive testing and destructive testing of the weld. With more data
becoming available in the future, it is likely that it will be possible to demonstrate that
the probability of emplacing more than one canister with an initial undetected
penetrating defect is less than one per cent. In summary, the properties of the EBS and host rock will conform to the performance
targets and target properties at the end of the operational period, with some possibility
of incidental deviations: an undetected penetrating defect in one or a few canisters,
higher flow rate or lower transport resistance than the target values for a few deposition
holes and groundwater composition outside the target range for a short time during
operation and soon after closure for a few deposition holes.
Post-closure period during the next 10,000 years
Over the next 10,000 years, the climate is expected to remain essentially as today, i.e. a
temperate climate with a boreal ecosystem. Groundwater flow and chemistry will
recover from the disturbances caused by the excavation, and will slowly evolve as a
response to naturally occurring gradients. Key processes during this period will be water
uptake, swelling and homogenisation of the clays in the buffer, backfill and seals, and
the decline of the residual heat from the spent nuclear fuel.
Crustal uplift will continue but at gradually lower rates, and higher hydraulic gradients
will develop close to the shoreline. At 1000 to 2000 years after present, the shoreline
will have retreated far enough that further changes will not affect the flow rates in the
repository volume.
The heat from the spent nuclear fuel increases the flow rates at the repository depth by a
factor of 2 to 3 compared with the natural state during the first hundreds of years, and
enhances temporarily and locally upward flows. The heat tends to result in an upward
driving force for the water, but when combined with the stronger natural downward
forces, the flow remains mainly directed downwards. Heat production declines to very
low levels after the first few thousands of years, and the flow returns to its natural state.
20
Groundwater flow modelling based on a discrete fracture network approach provides
information about the migration paths and flows around the deposition holes. This
information is used to assess whether the target properties are met as well as to provide
input to the radionuclide release and transport analysis. The modelling studies quantify
the effect of the tunnel EDZ and the rock damage around the deposition holes (including
thermally induced damage) on local flow rates and other flow-related transport
parameters. The presence of the damaged zone increases the connectivity of fractures
and flow around the deposition hole, but the effects on the natural fractures are limited,
and flow rates in natural fractures and the transport resistances in the vicinity of the
deposition holes are consistent with target properties for most deposition holes.
Groundwater flow modelling and subsequent reactive transport modelling show that the
salinity field at the repository depth recovers from repository excavation, but at a much
slower rate than the flow field. The natural salinity state is reached within hundreds of
years. As the disturbances caused by repository construction cease, the groundwater
composition stabilises and the variation seen during the operational period diminished.
The few local values that were outside the target value range return within the range in a
relatively short time. At repository depth, the pH remains close to 7.5 and reducing
conditions prevail. In the longer term, salinity, chloride concentration and total charge
equivalent of cations all decrease very slowly, due to the infiltration of meteoric water,
but the concentrations remain consistent with the target values over the time window in
question.
Groundwater flowing into the repository leads to saturation and swelling of the buffer
and backfill. Initial differences in the density and swelling pressure will be evened out
(homogenisation), although some heterogeneity will remain. The time to reach full
saturation in the buffer is calculated as a few tens to several thousands of years,
depending on the local hydraulic conditions. Calculations show that expansion of the
buffer into the backfill and the changes in the density of the buffer will not be sufficient
to threaten the performance targets for the buffer and backfill (i.e. a sufficiently high
density will be maintained).
While heat is generated by the spent nuclear fuel, the thermo-hydro-mechanicalchemical evolution will lead to geochemical changes in the buffer, but these will be
limited. After saturation and development of the full swelling capacity, the changes will
be even lower, constrained by diffusive processes. In particular, no or only minor clay
mineralogical changes will occur. The production of sulphide via microbial processes in
the buffer will be minor. Further, the already minor impact of cementitious leachates on
the buffer is estimated to diminish.
The evolution of porewater chemistry in the backfill will be similar to that in the buffer,
but will be less affected by the heat from the spent fuel. Thermally-induced clay
alteration and cementation will be negligible. Disturbances due to leachates from
cement materials will diminish in general and also locally, due to the low concentrations
of alkalis in the leachates. Production of sulphide via microbial reduction of sulphate
cannot be ruled out in localised zones of low backfill density; this is accounted for in the
analysis of canister corrosion.
21
There are no major uncertainties in the evolution of the closure components during the
first 10,000 years after closure. Even if the hydraulic plugs degrade, no preferential
paths will form. At depth, transport through closure components will still be dominated
by diffusion.
Sulphide is the main agent for canister corrosion. Microbially produced sulphide in the
buffer is negligible in this period; sulphide supply from the backfill is limited by the
precipitation of iron sulphide and losses to the rock mass. Moreover, the sulphide has to
diffuse through a thick layer of bentonite to reach the canister. Corrosion calculations
coupled with groundwater flow modelling, and taking account of the possibility of early
buffer erosion, show that the total corrosion depth will be negligible during the first
10,000 years. The initially intact canisters will remain intact for all conceivable loads
that could occur during the first 10,000 years and thus the spent nuclear fuel remains
contained within the canister.
In summary, the properties of the EBS and host rock will conform to the performance
targets and target properties over the period up to 10,000 years, with some possibility of
incidental deviations: an undetected penetrating defect in one or a few canisters; higher
flow rate or lower transport resistance than the target values for a few deposition holes
and groundwater composition outside the target range for a short time during repository
operation and soon after closure for a few deposition holes; and local lower density
areas in the backfill where there is the possibility that sulphate reduction may occur.
Evolution during repeated glacial cycles
In the longer term, major climatic changes are expected, as described under ‘Future
lines of evolution’ above. Effects include permafrost, glaciation and associated sea-level
changes. These changes affect the isostatic load, rock stresses, and groundwater flow
and composition, as well as the mechanical and thermal evolution of the EBS and host
rock.
During the continued temperate climate up to 50,000 years AP, there is a slight increase
in the groundwater flow rates in the upper part of the bedrock, due to surface
environment changes. The flow rates at repository depth are not significantly affected.
The continuing infiltration of meteoric water results in slowly decreasing salinity so
that, towards the end of this period, a few canister positions may experience dilute
conditions.
Groundwater flow and salinity have been modelled for two representative periods of
permafrost development, during which permafrost reaches depths of about 80 m and
300 m. The effects of an ice sheet have also been modelled considering an immobile ice
sheet over the whole of Olkiluoto Island for 1000 years, and a retreating ice sheet.
Under permafrost conditions, the hydraulic conductivity in the rock is reduced by
several orders of magnitude and the infiltration is very low. As a result, the groundwater
salinities remain at the level prevailing before the onset of the permafrost.
During ice-sheet retreat, the flow rates through the repository volume depend on the
location of the ice margin with respect to the repository. While the repository is still
below the ice sheet but the ice margin is close, the flow rates are significantly increased
22
(by a factor of 4 to 7) and directed downwards. As the ice passes the site, the main flow
direction is upwards and flow rates reduce as the distance to the ice margin increases.
Some canister locations might then experience higher flow rates and lower transport
resistances than the target properties. This has been taken into account in the assessment
of the canister corrosion rates and in the formulation of release scenarios. Nevertheless,
for most of the deposition holes, the host rock target properties related to groundwater
flow are fulfilled during ice-sheet retreat.
Although there is no evidence that fresh meltwater ever reached repository depth at
Olkiluoto during the last glacial cycle or in the previous ones, dilute conditions around
some of the deposition holes during a future ice-sheet retreat phase might be possible
depending on the duration of meltwater infiltration, which could give rise to chemical
erosion of the buffer and backfill (see below).
Other geochemical properties (pH, redox conditions, chloride concentration, total
charge equivalent of cations, sulphur and iron species) are all expected to remain
consistent with the target properties throughout the period, including during ice-sheet
retreat and melting. Oxygen will be consumed within short distances along the flow
path and thus not reach the repository level.
Although groundwater data clearly indicate sulphide values below 1 mg/L, a pessimistic
upper bound of 3 mg/L is adopted in corrosion calculations described below; this
accounts for possible solubility control by the more soluble amorphous iron sulphide in
combination with kinetically constrained limited availability of iron and the
uncertainties related to the microbial activity and availability of nutrients and energy
sources.
The possibility of a large earthquake leading to secondary shear movements on fractures
intersecting deposition holes and to canister failure, especially at a time of glacial
retreat, cannot totally be excluded. The risk of canister failures due to secondary shear
movements in the event of a large earthquake can be reduced by locating the deposition
holes away from large deformation zones and by avoiding large fracture intersections in
deposition holes, but it is estimated that few tens of canisters may still be in positions
such that they could potentially fail in such an event over a one million year time frame.
On the other hand, the average annual probability of an earthquake large enough
potentially to lead to canister failure due to secondary movements on fractures is,
estimated to be low, in the order of 10-7. This is based on the frequency of occurrence of
earthquakes in the Olkiluoto area and the fact that there are around five fault zones
within and around the area of the repository that could host such an earthquake. Thus,
during the first glacial cycle, there is little likelihood of canister failure due to rock
shear, although the possibility of such failures cannot be discounted over a one million
year time frame. Freezing of the buffer or the deposition tunnel backfill is not an issue because, based on
evidence from the past, permafrost will not reach the repository depth. In any case, the
buffer and backfill would withstand the freeze/thaw cycles without damage to their
safety functions.
23
The evolution of porewater salinities in the buffer and backfill will follow those in the
surrounding groundwaters, which will remain within the required target ranges, except
perhaps for short times during ice-sheet retreat and melting period. Under these
conditions, dilute groundwater conditions might cause some chemical erosion of buffer
and backfill. With the reference assumptions on groundwater flow (a selected realisation
of the DFN flow model) and evolving groundwater composition, one canister position is
calculated to undergo buffer erosion during the first glacial cycle to an extent that
advective conditions arise. This calculation should be seen as illustrative, being based
on only a single realisation of the DFN groundwater flow model. An analysis of
statistical distributions of flow-related parameter values between canister positions
shows that measures such as the mean and 90th percentile vary little between DFN
realisations. However, the number of canister positions experiencing advective
conditions is determined by the tails of these distributions, and is therefore subject to
more uncertainty. Taking a more cautious view on this and other uncertainties, buffer
erosion might result in advective conditions in a few canister positions.
The backfill in parts of the central tunnels may lose clay components due to chemical
erosion, but this will not jeopardise the overall safety functions of closure. Degradation
of closure plugs is uncertain, but the swelling clays used in the lower parts of the
tunnels and shafts will ensure sufficient isolation capacity of the sealing structures.
As at earlier times, sulphide is the main agent for corrosion of the copper canisters.
Calculations of the corrosion depth in one million years have been made assuming that
the buffer performs as designed, a pessimistic sulphide concentration of 3 mg/L and a
range of flow conditions. The results show that the overall corrosion depth will not
exceed few tenths of a millimetre even over one million years. Thus, if the buffer
performs as designed, no canister failures due to corrosion are expected even with high
sulphide concentrations. Furthermore, even if the buffer is affected by chemical erosion,
few if any canister failures due to corrosion are expected during the first glacial cycle,
as long as conditions otherwise correspond to the expected evolution (i.e. performance
targets and target properties are met). The calculated rate of corrosion and the calculated
number of canister failures in these circumstances depends on the assumptions made
about groundwater flow and composition, corrosion area, fracture apertures, the rate of
buffer erosion and the possibility of locally thinner parts of the copper overpack.
Cautiously assuming a sulphide concentration of 3 mg/L in the groundwater, but with
realistic assumptions concerning these other factors, chemical erosion of the buffer and
subsequent corrosion by sulphide is calculated to lead to no canister failures within the
first glacial cycle, and 4−5 failures in the million year time frame. Based on more
cautious assumptions, around 3 canister failures are calculated to occur within the first
glacial cycle, and a few tens of failures in the million year time frame.
In summary, after the first glacial cycle, i.e. more than 100,000 years after repository
closure, the properties of the EBS and host rock will still conform to the performance
targets and target properties, with some incidental deviations: an undetected penetrating
defect in one or a few canisters, higher flow rate or lower transport resistance than the
target values for a few deposition holes, erosion of buffer in some deposition holes due
to long-term infiltration of meteoric water or dilute glacial meltwater and canister
failure by corrosion due to unfavourable groundwater conditions and buffer erosion and
canister failure due to shear displacements in fractures during ice-sheet retreat.
24
Successive glacial cycles will impose similar loads as considered during the first glacial
cycle. Thus, over the one million year assessment time frame:

the potential for buffer erosion increases for deposition holes that experience dilute
groundwater conditions during ice-sheet retreat;

the number of deposition holes that suffer a shear displacement sufficient to cause
canister failure could increase;

the extent of canister corrosion in deposition holes that suffer buffer erosion could
increase.
Over the one million year time frame, the properties of the EBS and host rock will still
conform to the performance targets and target properties except for the incidental
deviations listed above.
Uncertainties in performance assessment
If the engineered barriers and the rock for the whole repository system fulfil the set
performance requirements, no releases are expected during more than 100,000 years
after closure. However, deviations from performance targets and target properties
(Table 3) may lead to radionuclide releases. The importance of these releases has been
assessed considering different radionuclide release scenarios.
Table 3. Summary of deviations from performance targets and target properties as may
occur and are relevant in each time window.
Up to
closure
of the
disposal
facility
Up to
10,000
years
During
repeated
glacial
cycles
Possibility of an initial penetrating defect in one or a few canisters.



Higher flow rate or lower transport resistance than the target
values for a few deposition holes.



Groundwater composition outside the target range for a short time
during operation and soon after closure for a few deposition holes.


–
Low density areas in the backfill where sulphate reduction to
sulphide cannot be ruled out.
–


Erosion of buffer in some deposition holes due to long-term
infiltration of meteoric water or dilute glacial meltwater.
–
–

Canister failure by corrosion due to unfavourable groundwater
conditions and buffer erosion.
–
–

Canister failure due to shear displacements in fractures during
ice-sheet retreat.
–
–

Deviations
25
Assessment of radionuclide release scenarios
Scenarios and cases
In the Reference Case realisation (BS-RC) of the base scenario (BS) for radionuclide
release, an incidental deviation is assumed whereby one canister with an initial
penetrating defect of 1.0 mm diameter is emplaced in the repository. All other canisters
are assumed to comply with quality requirements. The single defective canister is
cautiously assumed to be located in a deposition hole with relatively unfavourable
hydrogeological characteristics. Except for the single defective canister, all other EBS
performance requirements are assumed to be met and upheld during the evolution.
Other cases within the base scenario consider alternative, cautiously selected positions
of defective canisters within the repository and consequent different flow path
characteristics, alternative near-field and geosphere speciation of radionuclides, and
delayed establishment of the transport path through the canister defect.
Two variant scenarios are identified that are considered plausible: an enlarging defect
and partial degradation of buffer (VS1); and failure of an initially intact canister by
corrosion following buffer erosion (VS2). Three disturbance scenarios are identified
that are considered very unlikely: accelerated insert corrosion rate (AIC); rock shear
(RS); rock shear followed by buffer erosion (RS-DIL). Various calculation cases for
these scenarios are considered to illustrate parameter and model uncertainties.
Combinations of these scenarios are also considered, both with each other and with the
base scenario.
Finally, complementary deterministic and probabilistic sensitivity analyses and Monte
Carlo simulations are carried out to investigate further uncertainties, for example,
considering the possibility of multiple canisters with initial penetrating defects,
alternative groundwater types, and alternative assumptions and parameter values for
radionuclide transport.
Analysis of the base scenario Reference Case
Analysis of the Reference Case shows that the highest rate of radionuclide release is of
C-14, which peaks at around 4500 years and then declines due to radioactive decay.
Other, longer-lived, radionuclides Cl-36, I-129 and Cs-135 contribute at early times and
dominate beyond a few ten thousand years. The dominant migration path is from the
buffer into fractures intersecting the deposition hole; migration paths in the EDZ of the
deposition tunnel or in the tunnel backfill are less important.
Figure 6 shows the near-field release and geosphere release rates for the base scenario
Reference Case, normalised with respect to the radionuclide-specific constraint for the
radioactive releases to the environment defined in STUK Guide YVL D.5. The figure
indicates that during the dose criteria time window (up to 10,000 years) the normalised
activity release is almost four orders of magnitude below the criterion of one as also
given in Para 313 of YVL D.5; beyond a few tens of thousands of years the normalised
activity release rate decreases to between five and six orders of magnitude below one.
The limited role of the geosphere in attenuating the peak release rate is related to the
cautious assumption that the defective canister in the Reference Case is located in a
26
Figure 6. Evolution of the near-field and geosphere release rates for the base scenario
Reference Case, with the release rate for each radionuclide normalised with respect to
the regulatory nuclide-specific constraints for radioactive releases to the environment.
Regulatory geo-bio flux constraint denotes the constraint of 1 for the sum of the ratios
between the nuclide specific activity releases and the respective constraints given in
YVL D.5.
deposition hole with relatively unfavourable hydrogeological characteristics. The
complementary analyses show that, for most locations, the assumed canister defect
would result in much lower C-14 peak release as most C-14 would decay during the
transport in the geosphere. Cases investigating emplacement of a canister with an initial
defect at other cautiously selected locations in the repository show little variation from
the Reference Case results; this is to be expected as the near-field flows and geosphere
transport parameters are similar also in these cases. Results from cases considering
alternative speciation, allowing isotopes of silver, molybdenum and niobium to migrate
in anionic form are also virtually indistinguishable from those of the Reference Case.
The key results in the biosphere modelling are the projections of the development of the
surface environment during the first 10,000 years and the potential radiological impacts
on humans, plants and animals living in that environment. The present assessment
calculates pathway-, radionuclide- and biosphere object-specific annual doses to a
person and combines them into landscape dose. The annual landscape doses to each
exposed individual in the population form the dose distribution, from which the annual
dose to a representative person for the most exposed people and the average annual dose
to other people are identified. The potential radiological impacts on plants and animals
are estimated by calculation of absorbed dose rates. The results from analysing the
Reference Case (BSA-RC) are briefly summarised below and in more detail in
Biosphere Assessment.
27
Surface environment development
The projection of the development of the terrain and ecosystems in the surface
environment for the next 10,000 years is presented in Terrain and Ecosystems
Development Modelling. Some results for the Reference Case are presented in Figure 7.
Doses to humans
The screening analysis performed on the geosphere releases in the repository Reference
Case BS-RC results in five radionuclides being propagated all the way through the
biosphere modelling chain in calculation case BSA-RC. These radionuclides are C-14,
Cl-36, Mo-93, Ag-108m and I-129.
The calculated annual doses to a representative person within the most exposed group
(Emost_exp) and for other people (Eother) are presented in Figure 8. The structure of the
time profile of the activity releases in Figure 6 and the resulting annual doses are rather
similar. The more irregular shapes of the dose curves are mainly caused by the
development of the surface environment, i.e., changes in the sizes of contaminated
biosphere objects and their ecosystem types. The calculated dose maximum for the most
exposed group is 2.0·10-7 mSv and occurs at about 3,000 years after closure; the
corresponding dose maximum for other exposed people is 1.3· 10-9 mSv and about
1,000 years earlier. These results are about 6 to 7 orders of magnitudes below the
regulatory radiation dose constraints. The results also show that the contribution from
C-14 dominates the annual doses.
Doses to plants and animals
The absorbed dose rates for plants and animals for the calculation case (BSA-RC) are
all below 3·10-7 mikroGy/h; the highest dose rates are observed for Pike and Beaver in
the freshwater environment. All results for calculated absorbed dose rates to plants and
animals are reported in Biosphere Assessment.
Analysis of variant scenarios
In Variant Scenario 1 (VS1) it is assumed that processes occurring at the buffer/rock
interface lead to degradation of the outer part of the buffer and partial loss of its
radionuclide retention capacity. Furthermore, there is an initial penetrating defect in one
of the canisters. Enhanced transport of corrosive agents, such as sulphide, from the rock
to the canister when the buffer is degraded may accelerate corrosion of the insert of this
defective canister, as well as the overpack. It is assumed that the defect thus becomes
enlarged over time due, for example, to volume expansion of the insert as it corrodes or
to corrosion of the copper overpack.
Results from cases designed to represent VS1 show that peak normalised release rates
are about one order of magnitude higher than in the Reference Case, i.e. still almost
three orders of magnitude below the regulatory requirement on the activity releases
from the geosphere to the surface environment (regulatory geo-bio flux constraint). The
peak is again dominated by C-14 but occurs later, at about 20,000 years, reflecting the
influence of the progressively increasing diameter of the penetrating hole.
28
Figure 7. Surface environment projections for three time steps in the Reference Case.
The approximate location of central part the repository is indicated with a red circle
and the discharge locations with a green circle.
29
1.E-06
E_most_exp
E_other
Annual dose [mSv]
1.E-08
1.E-10
1.E-12
1.E-14
BSA‐RC
1.E-16
2020
4020
6020
8020
10020
12020
Year
Figure 8. The annual doses to a representative person within the most exposed group
(E_most_exp) and for other exposed people (E_other) for the calculation case BSA-RC.
In Variant Scenario 2 (VS2), chemical erosion of the buffer takes place associated with
ice-sheet retreat. Significant buffer erosion is considered unlikely, but cannot currently
be excluded in at least some of the deposition holes. Eventually, advective conditions
are established around the canisters in these deposition holes, leading to enhanced
corrosion of the canister by sulphide, and eventually to canister failure (no initial
penetrating defect is assumed but a thinner canister wall 35 mm is adopted, which is the
minimum thickness according to the design specifications). Taking account of results of
the modelling of buffer erosion and sulphide corrosion from the performance
assessment, canister failure is not expected to occur for at least several hundred
thousand years. At these long times, the geosphere release rate is dominated by the nonsorbing and long-lived radionuclides, namely I-129 and Cl-36. Modelled geosphere
release rates also show periodic maxima, due to relatively rapid flushing of these
radionuclides from the geosphere during periods of high flow associated with ice-sheet
retreat. In the case representing the least favourable deposition position (VS2-H1), the
peak normalised geosphere release rate is more than three orders of magnitude below
the geo-bio flux constraint. The low peak normalised release rates calculated for a single
failed canister in scenario VS2 indicate that VS2 indicate that the few canister failures
that could potentially occur in the more likely lines of evolution (or even the few tens of
canister failures calculated to occur based on highly pessimistic assumptions) could
easily be tolerated without exceeding the regulatory constraint.
The results of the calculation cases for the biosphere variant scenarios are reported in
Biosphere Assessment. They show that the annual doses will not exceed the dose
constraints set by the regulations.
30
Analysis of disturbance scenarios
The rock shear (RS) scenario considers canister failure due to shear movements on
fractures intersecting the deposition holes in the event of a large earthquake. Two cases
have been analysed: RS1 and RS2, in which rock shear and canister failure are assumed
to occur respectively at 40,000, i.e. during the present, temperate period, and at 155,000
years, during a period of ice-sheet retreat. The highest peak normalised release rates
from the geosphere are, in both cases, more than two orders of magnitude below the
regulatory geo-bio flux constraint. This implies that more than one hundred canisters
would have to fail simultaneously before the regulatory geo-bio flux constraint would
be exceeded, even without taking into account the low probability that this event would
actually happen. This exceeds the few tens of canisters estimated to be in critical
positions that are vulnerable to failure in the event of a large earthquake.
In the scenario of rock shear followed by buffer erosion (RS-DIL), the buffer undergoes
either immediate damage or longer-term erosion following the rock shear, due to the
penetration of low-ionic strength water to repository depth. The peak release rates for
RS-DIL cases are higher than for RS cases, but, nevertheless, the peak expectation value
of the normalised release rate in the RS-DIL scenario is still around an order of
magnitude below the regulatory limit.
The accelerated insert corrosion scenario (AIC) considers the possibility that an initial
penetrating defect in a canister becomes enlarged over time due to faster than expected
corrosion of the insert whereas the performance targets are fulfilled for all the other
engineered barriers and the host rock is expected to meet the target properties during the
evolution for the whole time window. More pessimistically than in VS1, the
enlargement of the defect is assumed to occur instantaneously at 15,000 years leading to
complete loss of transport resistance of the defect. The analysis of this scenario focuses
on the significance of whether or not a transport path between the canister interior and
the buffer exists prior to defect enlargement. Two cases have been considered: AIC-TI
assumes no path exists before enlargement and AIC-LI includes such a path. In both
cases, release rates increase rapidly at 15,000 years to peak shortly thereafter. The peak
is somewhat lower in AIC-TI compared with AIC-LI. The largest normalised releases
from the geosphere are in both cases at least one order of magnitude below the
regulatory constraint.
Scenarios for inadvertent human intrusion caused by borehole drilling have been
formulated and analysed. Expectation values of effective doses to drilling technicians
and site geologists have been derived based on a stylised approach to the dose
calculations and estimation of indicative annual probabilities of an intrusion event. The
peak expectation value of the dose in the calculation case, where drilling affects the
canister (DS(F)-HI-CANISTER) is around an order of magnitude below the regulatory
radiation dose constraint for the most exposed people. The peak expectation value of the
dose in the calculation cases where drilling affects contaminated buffer and backfill
(DS(F)-HI-BUFFER and DS(F)-HI-BACKFILL) is several orders of magnitude below
the regulatory radiation dose constraint for the most exposed people.
Possible binary combinations of scenarios have also been considered. Many can be
excluded from detailed analysis on qualitative grounds. Where it is appropriate to sum
31
the releases of two different scenarios, the combined normalised releases still do not
exceed the regulatory constraint.
The results of the calculation cases for other biosphere disturbance scenarios are
reported in Biosphere Assessment. They show that the annual doses will not exceed the
dose constraints set by the regulations.
Complementary analysis
Monte Carlo simulations and probabilistic sensitivity analysis (PSA) have been carried
out for two model cases (Figure 9):
1. the “hole forever” case, where the initial penetrating defect in the canister overpack
remains unchanged in size over time; and
2. the “growing hole” case, where the defect becomes instantaneously enlarged at a
randomly sampled time.
The PSA provides a rich source of understanding of the sensitivity of model outputs to
variations in input parameter values, allowing the most important parameters and
parameter combinations to be determined. It has been shown that C-14, Cl-36 and I-129
control the normalised release rates for both cases and that the peak of the mean release
rates in the growing-hole case is about two orders of magnitude greater than in the holeforever case. The 10 % of the realisations with the greatest peak activity release rates to
the surface environment are controlled by C-14 for both cases. The time to loss of hole
resistance in the growing hole case, varied between 5000 and 50,000 years, is important
in determining the peak release rate of C-14 (half-life 5700 years) whereas, for the
longer-lived radionuclides, its influence is much smaller or even negligible. For longerlived, sorbing radionuclides, uncertainty in the diffusion and retention properties of the
buffer are important in determining peak release rates.
32
1.E+00
Normalised release rate to the biosphere
1.E-01
Total
95 percentile
1.E-02
1.E-03
Regulatory geo-bio
flux constraint
th
Reference Case
Hole forever
1.E-04
th
1.E-05
1.E-06
99 percentile
Mean
Maximum
value
Median
50th percentile
1.E-07
1.E-08
th
5 percentile
1.E-09
1.E-10
1.E+01
st
1 percentile
1.E+02
1.E+03
1.E+04
Minimum
value
1.E+05
1.E+06
Time (years)
1.E+00
Normalised release rate to the biosphere
1.E-01
Total
95 percentile
1.E-02
1.E-03
Regulatory geo-bio
flux constraint
th
Growing hole
1.E-04
th
1.E-05
1.E-06
99 percentile
Maximum
value
Mean
1.E-07
Median
50th percentile
1.E-08
th
Minimum
value
5 percentile
1.E-09
1.E-10
1.E+01
1st percentile
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
Time (years)
Figure 9. Total normalised release rates to the environment in two Monte Carlo cases
with 10,000 realisations (hole-forever and growing-hole cases).
Summary of results and uncertainties
The scenarios analysed address uncertainties in the evolution of the disposal system. A
range of calculation cases has been analysed for each scenario. Case assumptions have
been applied within each scenario that include cautious views on the severity of
initiating events, subsequent degradation of engineered barriers and migration paths
from defective or damaged canisters. Combinations of scenarios have also been
33
analysed. Parameter uncertainties have been investigated most thoroughly for the case
of canister failure due to an initial penetrating defect; in this case deterministic analyses
were complemented by Monte Carlo simulations and PSA. The model results are found
to be consistent with the current understanding and show significant attenuation and
delay of releases in the near field and geosphere.
Figure 10 shows the peak normalised activity release rates from the geosphere to the
surface environment for all calculation cases within the base, variant and disturbance
scenarios. For cases RS1, RS2, RS1-DIL and RS2-DIL, 1000-year moving averaging
has been applied before calculating the peaks, which is consistent with STUK Guide
YVL D.5.
The lowest peak normalised releases are for the Reference Case (BS-RC) and sensitivity
cases within the base scenario. In all cases, peak normalised release rates to the surface
environment are below the regulatory geo-bio flux constraint by around an order of
magnitude or more.
For the Reference Case the maximum of annual dose to a representative person within
the most exposed group is about four to five orders of magnitude below the regulatory
radiation dose constraint.
Figure 10. Peak normalised geosphere release rates for all calculation cases within the
base, variant and disturbance scenarios, each assuming the failure of a single canister.
Colours are used to group cases by scenario. * indicates that 1000 year averaging is
applied, in these cases. The right hand subfigure shows ranges of values for the peak
probability-weighted normalised release rates in the RS and RS-DIL scenarios. These
ranges arise due to uncertainties in the numbers of canisters failing due to rock shear,
as well as the timing of failure.
34
Complementary considerations
Complementary considerations provide additional evidence for the long-term safety of
disposal according the KBS-3 method at the Olkiluoto site.
Choice of geological disposal
The choice of geological disposal as a concept for disposal of radioactive waste is
backed by technical experience and international consensus.
An appropriately chosen geological formation provides an environment that is stable
over many millions of years – geological timescales – and the nature of changes that can
occur is predictable from the geological sciences. A repository concept is developed that
is consistent with the chosen geological formation, taking advantage of the benign or
beneficial qualities and designed to withstand expected and unlikely events and
processes that could affect the geological formation in the long term. The depth below
ground provides buffering of the repository system from processes occurring in the
surface environment and protection from unauthorised or inadvertent human intrusion.
Support for the robustness of the KBS-3 method
The KBS-3 method uses a few simple, common materials – copper and iron for the
canister, natural swelling clay for the buffer and backfill. This reduces the number of
materials whose properties need to be understood and the number of interactions
between the materials.
The shorter-term properties of the materials are well known from their long use in
engineering and industrial production in the case of copper and iron, and from their use
in ground and underground engineering in the case of swelling clays. There is no
difficulty in manufacturing or refining of the materials to the grade required by the
KBS-3 design. Their longer-term properties are indicated from a range of natural
analogues.
Copper is one of the few metallic elements to occur in elemental form as a natural
mineral – native copper. There is evidence from a range of occurrences of native copper
for very low corrosion rates of copper for millions of years under reducing conditions.
Archaeological artefacts, while representing more variable and often more severe
conditions, suggest low copper corrosion rates and also that localised corrosion is low
compared with general corrosion.
The bentonite buffer needs to maintain its low permeability and plasticity, and limit
microbial activity. Studies of naturally-occurring bentonite deposits show that mineral
alteration processes that are detrimental to the properties of low permeability and
plasticity only occur significantly above about 150–200 °C even over geological
timescales. Significant changes also depend on a supply of potassium, which will be
limited in the buffer (due to a favourable groundwater composition and limited amount
of potassium introduced to the system as foreign materials). Thus the bentonite buffer
will remain stable during the repository thermal period, in which a maximum buffer
temperature of around 90–100 °C is foreseen. There are several excellent examples of
35
bentonite and other clays functioning as a hydraulic barrier to preserve wood and human
cadavers; these analogues also indicate that microbial activity was significantly reduced.
Furthermore, radionuclide concentrations in the buffer and backfill have been shown to
be comparable to examples of naturally-occurring radioactive material (NORM), and
radionuclide release rates to be be generally comparable to, or less than, naturally
occurring activity fluxes in groundwater at the site.
Support for the suitability of the Olkiluoto site
In Finland, there are limited choices of deep geological settings, leaving fractured
crystalline basement as the only realistic choice of repository host rock. Such rocks have
also been considered as suitable for locating a deep geological repository in many other
countries, including several in which alternative rock formations are available.
Key features of the Olkiluoto site include the stable tectonic situation, presence of
suitable volumes of good quality rock suitable for repository construction, low
groundwater flows, reducing conditions and also otherwise favourable groundwater
conditions at repository depth. No natural resources have been found at Olkiluoto or
nearby, reducing the risk of human intrusion.
The Olkiluoto site is situated within the Fennoscandian Shield, away from active plate
margins. In general, the frequency and magnitude of earthquakes in Finland is very low;
earthquake magnitudes have never exceeded 5 (M=~5) since records began in the
1880s. Further, according to the data from historical earthquakes, the Olkiluoto area is
located within a zone of lower seismicity, between two seismically active belts. There
have been only nine recorded earthquakes within 100 km, with the nearest event
(M=3.1) at 35−40 km from Olkiluoto in 1926.
An important consideration is to find sufficient volume of rock, with generally low and
minor fracturing, to accommodate the spatial extent of the repository. Deposition
tunnels must be placed avoiding shear zones or heavily fractured zones, although the
access tunnel or shafts may cross such zones. At Olkiluoto, several options have been
considered and suitable volumes of rock have been defined such that the deposition
tunnels can be placed on a single level at a depth of 400−450 m.
The rock at Olkiluoto is geotechnically suitable for the construction of generally selfsupporting tunnels requiring only light rock support. Water inflows at depth are low and
zones of inflow can be treated by local grouting. Significant local experience exists
from construction of the ONKALO at Olkiluoto.
Evidence from boreholes and the ONKALO show that, in the natural situation,
groundwaters at repository depth are reducing and also have otherwise favourable
hydrochemistry for the longevity and function of the canister, buffer and backfill.
Levels of sulphide, which is expected to be the main agent for canister corrosion, and
sulphate that might be reduced to sulphide at the repository depth, are both suitably low.
There is no evidence of that dilute meltwater ever reached repository depth during past
glacial cycles.
36
Conclusions
The main research and development needs during the coming years
The TURVA-2012 safety case assesses the performance and long-term safety of a
KBS-3 type spent nuclear fuel disposal facility at Olkiluoto. The safety case also
addresses the known uncertainties that may have an impact on the performance of the
facility. The TURVA-2012 safety case forms the basis for the construction licence
application in which Posiva proposes that the construction of the repository can be
started. Some uncertainties still remain, but these do not affect the conclusions on longterm safety. Additional research and development will, however, help increase the
reliability of the safety case to be compiled for the operational license application. The
focus of the research and development in the coming years are on the:

better understanding of the processes affecting canister corrosion and erosion of
buffer and backfill;

rock conditions in potential volumes of rock for the repository and the application of
RSC criteria for the selection of repository panels, tunnels and deposition holes;

demonstration of the implementation of the components of the repository system at
full scale according to the technical design and quality performance requirements.
Further investigations of the properties of the rock in the repository area will reduce the
probability of locating the canisters in unfavourable positions with respect to future
loads. The processes affecting the performance of the engineered barriers will continue
to be experimentally studied. Technical tests will be applied to demonstrate that the
repository can be implemented according to the assumptions made in the safety case.
Conclusions on compliance
The TURVA-2012 safety case has been compiled according to the regulatory
requirements. It demonstrates that Posiva’s repository design and analyses of
performance and safety are fully consistent with all the legal and regulatory
requirements related to long-term safety as set out in Government Decree 736/2008 and
STUK YVL Guides. A detailed trace showing that each of legal and regulatory
requirements is fulfilled is contained within the body of the TURVA-2012 portfolio.
Key features of the demonstration are summarised below.
The Posiva repository design is based on a robust system of multiple barriers. For the
expected evolution of conditions in the Olkiluoto bedrock and engineered barriers, the
copper canisters, in which the spent nuclear fuel is contained, are expected to contain all
radionuclides for over one million years. The location of the repository, at a depth of
about 400−450 m below ground, will provide isolation from the surface environment
and protection against inadvertent human intrusion.
The mutually complementary barriers provide well-defined safety functions and the
barriers are arranged so that the detrimental impact of a deficiency in any individual
barrier on its safety functions will be compensated for by other safety functions.
Similarly, the system of complementary barriers and safety functions provides
robustness with respect to external events and processes, including geological and
climatic changes. The requirements for the reliable operation of each safety function are
37
expressed in terms of performance targets for the engineered barriers and target
properties for the host rock. These lead to design requirements for the engineered
barriers and definition of a Rock Suitability Classification system (RSC), by which the
local suitability of the rock for development of underground openings and deposition of
spent nuclear fuel can be assessed.
A comprehensive examination has been made of the features, events and processes that
could affect the evolution of the disposal system (repository system plus surface
environment), or the performance of individual barriers or fulfilment of their safety
functions. Understanding of the changes due to construction and operation of the
repository, and understanding of the longer-term natural processes (mainly related to
climate changes) that will control the evolution of the natural setting of the repository,
leads to the definition of future lines of evolution of the repository and its setting.
The performance of the repository system has been systematically analysed in different
time windows. The analyses take account of the uncertainties in the initial state and
expected thermal, hydraulic, mechanical and chemical evolution of the repository
system, and uncertainties in the expected future lines of evolution, and also the
occurrence of unexpected or disruptive events. The analyses show that, under most
conditions and lines of evolution of the host rock and engineered barriers, all
performance requirements will be met. In this case, the copper canisters will remain
intact and no releases of radionuclides will occur over at least one million years. Up to
50,000 years, the only plausible cause of release of radionuclides is that a canister with
an initial penetrating defect has escaped detection and is emplaced in the repository. In
the longer term, glacial episodes at the site may cause hydrogeological and
hydrochemical changes leading to buffer erosion and increased canister corrosion and
seismic disturbances leading to shear movements on fractures intersecting the
deposition holes. These changes and disturbances, if they were to occur, could
potentially lead to the failure of up to a few tens of canisters and to the release of
radionuclides in less favourable locations within the repository.
Although releases of radionuclides to the environment are not expected, the safety
analyses focus on the cases in which releases of radionuclides could occur. It is shown
that even accounting for unlikely combinations of emplacement of a canister with an
initial penetrating defect in less favourable local rock conditions, peak normalised
radionuclide release rates to the surface environment are orders of magnitude below the
radionuclide-specific constraints specified in the STUK YVL Guide D.5. In the long
term (approximately 100,000 years or more), calculated radionuclide release rates
remain below the regulatory constraint for the radioactive release to the environment,
even for pessimistic and unlikely combinations of damage to canisters by rock shear
events and erosion of buffer material due to dilute groundwater conditions.
Overall, it is concluded that the TURVA-2012 safety case demonstrates compliance
with the legal and regulatory requirements for the planned and designed disposal facility
for spent nuclear fuel at Olkiluoto. Some uncertainties still remain in the data and
models, and some of these are unlikely to be eliminated. However, the analyses
performed have shown that the repository system is robust against these uncertainties,
and that the conclusions drawn about the compliance with the safety requirements hold
even when these uncertainties are taken into account.
38
1
TABLE OF CONTENTS
ABSTRACT
TIIVISTELMÄ
EXECUTIVE SUMMARY
ABBREVIATIONS AND DEFINITIONS .......................................................................... 5 FOREWORD ................................................................................................................ 13 1 INTRODUCTION ................................................................................................. 15 1.1 Spent nuclear fuel management in Finland ................................................ 15 1.2 Nature and evolution of the hazards presented by spent nuclear fuel ....... 16 1.2.1 Radiation risks .............................................................................. 16 1.2.2 Hazards and protection ................................................................ 16 1.2.3 Evolution of the hazard and implications for design ..................... 17 1.3 Posiva’s programme for spent nuclear fuel disposal .................................. 18 1.4 The TURVA-2012 safety case ................................................................... 20 1.4.1 What is a safety case? ................................................................. 20 1.4.2 The TURVA-2012 safety case...................................................... 20 1.4.3 The TURVA-2012 portfolio ........................................................... 21 1.4.4 Quality assurance ........................................................................ 23 1.5 Legal and regulatory context for the management of spent fuel ................ 23 1.5.1 International treaties and agreements .......................................... 23 1.5.2 Legal requirements....................................................................... 24 1.5.3 Regulatory guidance .................................................................... 25 1.5.4 Safety submissions in support of construction and operating
licences ........................................................................................ 26 1.6 Feedback from STUK on the Interim Summary Report 2009 .................... 27 1.7 Structure of this report................................................................................ 27 2 METHODOLOGY ................................................................................................ 31 2.1 The KBS-3 method and the Olkiluoto site .................................................. 31 2.1.1 The KBS-3 disposal method......................................................... 31 2.1.2 The Olkiluoto site.......................................................................... 32 2.2 Design methodology .................................................................................. 34 2.2.1 Safety principles, safety concept and safety functions ................. 35 2.2.2 Performance targets and target properties................................... 39 2.2.3 Design requirements, rock suitability classification and
design specifications .................................................................... 43 2.3 Assessment methodology .......................................................................... 45 2.3.1 Iterative approach ........................................................................ 45 2.3.2 Description of the disposal system ............................................... 47 2.3.3 Features, events and processes .................................................. 48 2.3.4 Models and data and their use ..................................................... 48 2.3.5 Assessment of performance of the repository system under
the most likely lines of evolution ................................................... 50 2.3.6 Scenario formulation .................................................................... 52 2.3.7 Approach to the analysis of radionuclide releases, transport
and radiological impact................................................................. 53 2.3.8 Treatment of uncertainty .............................................................. 56 2.3.9 Complementary considerations and supporting evidence............ 58 2.4 Uncertainty management ........................................................................... 59 2.5 Quality management .................................................................................. 60 2
2.5.1 2.5.2 2.5.3 2.5.4 2.5.5 Goals and principles ..................................................................... 60 Application to TURVA-2012 safety case production .................... 61 Model qualification and code verification...................................... 62 Data clearance ............................................................................. 63 Report and product review and approval process ........................ 65 3 DESCRIPTION OF THE DISPOSAL SYSTEM ................................................... 67 3.1 Host rock .................................................................................................... 67 3.2 Surface environment .................................................................................. 71 3.3 Underground openings and repository layout ............................................ 74 3.4 Spent nuclear fuel ...................................................................................... 77 3.5 Canister ...................................................................................................... 78 3.6 Buffer.......................................................................................................... 80 3.7 Backfill and plug ......................................................................................... 81 3.7.1 Deposition tunnel backfill.............................................................. 81 3.7.2 Deposition tunnel plug .................................................................. 82 3.8 Closure ....................................................................................................... 82 4 FEATURES, EVENTS AND PROCESSES ......................................................... 85 4.1 Identification and screening of FEPs .......................................................... 85 4.1.1 Identification of potentially relevant FEPs .................................... 85 4.1.2 Screening for relevance to TURVA-2012 ..................................... 86 4.1.3 Organisation of the FEPs ............................................................. 87 4.2 Development of the FEP descriptions ........................................................ 89 4.2.1 FEP descriptions .......................................................................... 89 4.2.2 Coupling between FEPs and aggregation/disaggregation ........... 90 4.3 Onward use of the FEP descriptions .......................................................... 91 4.4 Future lines of evolution ............................................................................. 91 5 MODELS AND DATA .......................................................................................... 95 5.1 Models and data for climate evolution and climate-driven processes ........ 95 5.2 Key models and data for performance assessment and for
formulation of radionuclide release scenarios ............................................ 99 5.2.1 Models and data for geosphere evolution .................................. 101 5.2.2 Models and data for engineered barrier system performance.... 104 5.3 Models and data for the analysis of radionuclide release scenarios ........ 113 5.4 Models and data for the biosphere assessment ...................................... 119 5.4.1 Development of surface environment......................................... 121 5.4.2 Screening analysis ..................................................................... 122 5.4.3 Landscape modelling ................................................................. 122 5.4.4 Radiological impact analysis ...................................................... 124 6 PERFORMANCE ASSESSMENT OF THE REPOSITORY SYSTEM .............. 129 6.1 Excavation and operation up to closure of the disposal facility ................ 129 6.1.1 Repository system evolution and performance .......................... 129 6.1.2 Fulfilment of performance targets and target properties ............ 136 6.2 Post-closure evolution during the next 10,000 years ............................... 137 6.2.1 Repository system evolution and performance .......................... 137 6.2.2 Fulfilment of performance targets and target properties ............ 145 6.3 Beyond 10,000 years during repeated glacial cycles ............................... 146 6.3.1 Repository system evolution and evaluation of performance ..... 146 6.3.2 Fulfilment of performance targets and target properties ............ 155 6.4 Summary statement of performance and uncertainties ........................... 156 3
7 FORMULATION OF RADIONUCLIDE RELEASE SCENARIOS AND
CALCULATION CASES .................................................................................... 159 7.1 Lines of evolution framing the scenarios and scenario formulation ......... 159 7.1.1 The link to scenario hierarchy .................................................... 160 7.1.2 Process for identification of scenarios and cases ...................... 161 7.2 Repository system scenarios ................................................................... 163 7.2.1 Base scenario for the repository system .................................... 163 7.2.2 Variant scenarios for the repository system ............................... 166 7.2.3 Disturbance scenarios for the repository system ....................... 167 7.2.4 Radionuclide release scenarios and cases ................................ 169 7.3 Surface environment scenarios ................................................................ 170 7.3.1 Base scenario for the surface environment ................................ 170 7.3.2 Variant scenarios for the surface environment ........................... 172 7.3.3 Disturbance scenarios for the surface environment ................... 173 7.3.4 Calculation cases ....................................................................... 174 7.4 Summary and discussion on comprehensiveness ................................... 180 7.4.1 Demonstrating that the set of scenarios is comprehensive ........ 180 7.4.2 Combinations of repository system scenarios ............................ 182 8 ASSESSMENT OF RADIONUCLIDE RELEASE SCENARIOS ........................ 185 8.1 Analysis of the Reference Case in the base scenario .............................. 185 8.1.1 Results for the repository system ............................................... 185 8.1.2 Results for the surface environment........................................... 188 8.2 Analysis of other cases in the base scenario ........................................... 193 8.2.1 Alternative canister positions BS-LOC1 and BS-LOC2 .............. 193 8.2.2 Alternative speciation BS-ANNFF / BSA-ANNFF ....................... 194 8.2.3 Delayed establishment of the transport path BS-TIME / BSATIME ........................................................................................... 194 8.3 Analysis of the variant scenarios in the repository system ....................... 196 8.3.1 Cases in Variant Scenario 1 (VS1)............................................. 196 8.3.2 Cases in Variant Scenario 2 (VS2)............................................. 199 8.4 Analysis of the variant scenarios in the surface environment .................. 201 8.5 Analysis of the disturbance scenarios in the repository system ............... 202 8.5.1 Cases in the accelerated iron insert corrosion AIC scenario...... 202 8.5.2 Cases in the rock shear RS scenario ......................................... 204 8.5.3 Case for the rock shear followed by buffer erosion in the RSDIL scenario ............................................................................... 206 8.6 Analysis of the disturbance scenarios in the surface environment .......... 206 8.6.1 Cases in the human intrusion scenario (DS(F)-HI) .................... 208 8.7 Complementary analyses ......................................................................... 209 8.7.1 More than one defective canister in the repository..................... 209 8.7.2 Monte Carlo analyses and probabilistic sensitivity analysis ....... 210 8.8 Combinations of repository radionuclide release scenarios ..................... 216 8.9 Summary of safety assessment results and uncertainties ....................... 216 8.9.1 Geosphere release rates ............................................................ 217 8.9.2 Doses to humans, animals and plants ....................................... 219 9 COMPLEMENTARY CONSIDERATIONS AND SUPPORTING EVIDENCE.... 221 9.1 Choice of geological disposal ................................................................... 221 9.2 Support for the robustness of the KBS-3 method .................................... 221 9.3 Support for the suitability of geological disposal at the Olkiluoto site ....... 223 9.4 Safety and complementary indicators ...................................................... 225 4
10 COMPLIANCE WITH LEGAL REQUIREMENTS AND REGULATIONS AND
ASSOCIATED UNCERTAINTIES ..................................................................... 229 10.1 Compliance with legal and regulatory requirements ................................ 229 10.2 The main research and development needs during the coming years .... 230 11 STATEMENT OF CONFIDENCE ...................................................................... 233 REFERENCES ........................................................................................................... 237 APPENDIX 1: GOVERNMENT DECREE (736/2008) ................................................ 251 APPENDIX 2: AUDIT OF LEGAL AND REGULATORY REQUIREMENTS RELATED
TO THE LONG-TERM SAFETY CASE ............................................................. 257 APPENDIX 3: REPOSITORY SYSTEM COMPONENTS FEPS AND SCENARIOS . 277 5
ABBREVIATIONS AND DEFINITIONS
3DEC
A rock mechanics code, a three-dimensional numerical
program based on distinct element method for
discontinuum modelling.
ABAQUS
The finite element code used for calculating the glacially
induced stresses, which combined with a synthetic
regional background stress model, is used for assessment
of the fault stability.
AD
Anno Domini.
AIC
Scenario abbreviation: Accelerated Insert Corrosion.
ALARA
As Low As Reasonably Achievable.
ANNFF
Scenario abbreviation: ANions in the Near and Far Field.
AP
After Present.
BBM
Barcelona Basic Model, a critical state model that
reproduces the mechanical behaviour of unsaturated soils
under different boundary conditions.
BFZ
Brittle Fault Zone.
BIOPROTA
Project which was set up to address the key uncertainties
in long term assessments of contaminant releases into the
environment arising from radioactive waste disposal
(www.bioprota.org).
BP
Before Present.
BS
Base Scenario.
BSA
Biosphere Assessment.
BWR
Boiling Water Reactor (Olkiluoto 1&2).
CDF
Cumulative Distribution Function.
CFM
Colloid Formation and Migration test at Grimsel.
CLIMBER-2
CLIMate-BiosphERe model, an Earth System Model of
Intermediate Complexity (EMIC) used for simulating the
climate evolution.
CLIMBER-2-SICOPOLIS
CLIMate and BiosphERe and SImulation COde for
POLythermal Ice Sheets, models used in the future climate
modelling.
CODE BRIGHT
COupled DEformation BRIne, Gas and Heat Transport,
the finite element code used to model the thermohydraulic behaviour of clay.
ConnectFlow
The suite of groundwater modelling software that includes
the NAMMU continuum porous medium (CPM) module
as well as the NAPSAC discrete fracture network (DFN)
6
module, which is used to develop DFN-based models
groundwater flow and transport at the Olkiluoto site.
CPM
Continuous Porous Medium.
DDM
Displacement Discontinuity Method.
DFN
Discrete Fracture Network (an approach used in
groundwater flow modelling).
DIL
Scenario abbreviation: DILute groundwater conditions, for
example in RS-DIL, rock shear followed by buffer
erosion.
DiP
(Government) Decision-in-Principle.
Disposal facility
All underground tunnels, shafts, service areas and
deposition panels (tunnels and holes), plus above-ground
buildings that service the underground facility, but
excluding the encapsulation plant. In this report, the
above-ground parts are not discussed, as they are assumed
to be dismantled upon closure and thereby have no effect
on the long-term safety.
Disposal system
Repository system + surface environment.
DP
Dual Porosity groundwater modelling approach.
DSn
Scenario abbreviation: Disturbance Scenario.
DZ-path
Release path with exit from a deposition hole to the
escavation damaged zone (EDZ) below the tunnel floor.
EB
Electron beam (weld/welding).
EBS
Engineered Barrier System, see Table 2-1.
EBW
Electron Beam Welding.
Ecolego
Simulation software tool used for creating dynamic
models and performing deterministic and probabilistic
simulations.
ECPM
Equivalent Continuous Porous Medium.
EDZ
Excavation Damaged Zone; zone of the rock that is
irreversibly damaged by the excavation of the tunnel.
Eh
Redox potential.
EMCL
Environmental Media Concentration Limit.
EMIC
Earth system Model of Intermediate Complexity.
EPA
U.S. Environmental Protection Agency.
EPR
European Pressurised Water Reactor, trade name for the
PWR reactor type for OL3.
7
ERICA
Project under the EC 6th Framework Programme aimed at
providing an integrated approach to scientific, managerial
and societal issues concerning the environmental effects of
contaminants emitting ionising radiation, with emphasis
on biota and ecosystems.
FASTREACT
FrAmework for Stochastic REACtive Transport, code
used in hydrogeochemical modelling.
FEFTRA
The finite-element program package for groundwater flow
modelling which applies the ECPM approach to model
transient and density-driven flow and heat transfer by
conduction and the DP approach for modelling salt
transport.
FEM
Finite Element Method.
FEP
Feature, Event or Process (or as plural FEPs: Features,
Events and Processes).
F-path
Release path with exit from a deposition hole to a hostrock fracture intersecting the deposition hole.
FPI
Full Perimeter Intersection, used to describe fracture
extent in underground openings.
Fracod2D
A fracture mechanics code based on the Displacement
Discontinuity Method (DDM) that has been used for
predicting potential for spalling.
FSAR
Final Safety Analysis Report – needed in support of an
application for an operating licence.
FTRANS
Code used in previous safety assessments for the analysis
of the radionuclide release, retention and transport in the
geosphere.
GAM
Generalised Additive Model used to downscale nearsurface air temperature and precipitation from the
CLIMBER-2-SICOPOLIS results.
GD
Government Decree.
GIS
Geographical Information System.
GoldSim
Code used for the analysis of the radionuclide release in
the near field, and for probabilistic assessment in the near
field and far field.
GW
Groundwater.
HE
HEterogeneous.
HI
Human Intrusion.
HIPH
Calculation case: HIghly alkaline (pH) water in geosphere
and near field.
8
HIPH-NF
Calculation case: HIghly alkaline (pH) water in the near
field.
Hn
Canister position n in variant scenario V2.
HZ
Hydrogeological Zone.
IAEA
International Atomic Energy Agency.
ICRP
International Commission on Radiological Protection.
IRF
Instant Release Fraction.
KBS
(Kärnbränslesäkerhet). The method for deep geological
disposal of spent nuclear fuel based on multiple barriers.
KBS-3H
(Kärnbränslesäkerhet 3-Horisontell). Design alternative of
the KBS-3 method in which several spent nuclear fuel
canisters are emplaced horizontally in each deposition
drift.
KBS-3V
(Kärnbränslesäkerhet 3-Vertikal). The reference design
alternative of the KBS-3 method, in which the spent
nuclear fuel canisters are emplaced in individual vertical
deposition holes.
Kd
Distribution coefficient.
L/ILW
Low and Intermediate Level (radioactive) Waste.
LDF
Layout Determining Feature.
LI
Calculation case: Leaky Insert, in AIC-LI calculation case.
LO1−2
Loviisa reactors 1 and 2.
LOC
Calculation case: Canister location in calculation case BSLOC.
LOT
The long-term test of buffer material (at Äspö).
M
Magnitude.
MARFA
Migration Analysis of Radionuclides in the FAr field:
code, used to model radionuclide transport in geosphere.
MATLAB
MATrix LABoratory (a numerical computing environment
and programming language).
MPI/UW
Earth system model of Max Planck Institute used for the
estimation of the climate evolution on a time scale of
10,000 years.
MX-80
Commercial name of the reference buffer bentonite. A
high grade sodium bentonite from Wyoming, U.S., with a
montmorillonite content of 75−90 % (properties as
specified in this report and references herein).
NDT
Non-Destructive Testing.
9
NEA
Nuclear Energy Agency.
NORM
Naturally-Occurring Radioactive Material.
OL1−2
Olkiluoto 1 and 2 reactors.
OL3
Olkiluoto 3 reactor.
OL4
Olkiluoto 4 reactor to be constructed at Olkiluoto.
Expected to be similar to OL3 in TURVA-2012 safety
case.
ONKALO
Underground research facility constructed at Olkiluoto.
Pandora
Code used for radionuclide transport modelling in the
biosphere.
PDF
Probability Density Function.
PHREEQC
Reactive transport modelling code used for assessing the
evolution of the groundwater chemistry.
POTTI
Database at Posiva.
PSA
Probabilistic Sensitivity Analysis.
PSAR
Preliminary Safety Analysis Report – a part of the
construction licence application.
PWR
Pressurised Water Reactor.
QA
Quality Assurance.
QC
Quality Coordinator.
QDZ
Flow rate in the DZ-path.
QF
Flow rate in the F-path.
QM
Quality Manager.
QTDZ
Flow rate in the TDZ path.
RC
Calculation case: Reference Case in the Base Scenario
(BS).
RCC
Rank Correlation Coefficient.
REPCOM
Code used in previous safety assessments for the analysis
of the radionuclide retention and transport in the near
field.
Repository
Deposition tunnels + deposition holes.
Repository system
Spent nuclear fuel, canister, buffer, backfill (deposition
tunnel backfill + deposition tunnel end plug), closure
components and host rock. Excludes the surface
environment.
RIA
Radiological Impact Assessment.
RNT
RadioNuclide release and Transport (Model/ling).
10
RQ
Risk Quotient.
RS
Scenario abbreviation: Rock Shear (caused by an
earthquake).
RSC
Rock Suitability Classification system.
RTD
Research, Technical Development and Design, see also
TKS.
RQ
Risk Quotient.
SAFCA
The organisation of the TURVA-2012 safety case
production process.
SCC
Stress corrosion cracking.
SFR
Sparsely Fractured Rock.
SH
Semi Homogeneous.
SHYD
Surface and near surface HYDrogeological model(ling).
SICOPOLIS
Ice-sheet model describing the evolution of the Northern
Hemisphere ice sheets, their thickness and areal extent,
basal temperature and bedrock elevation.
SKB
Swedish Nuclear Waste Management Company.
SRB
Sulphate Reducing Bacteria.
SRRC
Standardised Rank Regression Coefficient.
SR-Site
Safety assessment for a repository in Forsmark by SKB.
STUK
Finnish Radiation and Nuclear Safety Authority.
TDS
Total Dissolved Solids.
TDZ-path
Release path with exit from the deposition hole to the
tunnel backfill above the deposition hole.
TEM
Ministry of Employment and the Economy, previously
Ministry of Trade and Industry (KTM).
TESM
Terrain and EcoSystem development Model(ling).
THM
Thermal, Hydrological, Mechanical.
THMC
Thermal, Hydrological, Mechanical, Chemical.
TI
Calculation case: Tight Insert in calculation case AIC-TI.
TILA-96
Name of Posiva’s safety assessment 1996.
TILA-99
Name of Posiva’s safety assessment 1999.
TKS
Finnish equivalent for RTD (see RTD).
TOUGHREACT
An integral finite difference code used in thermo-hydrogeochemical modelling.
TURVA-2012
Name of Posiva’s safety case 2012, TURVA means safety.
11
TVO
Teollisuuden Voima Oyj. Owner of the Olkiluoto power
plants and co-owner of Posiva Oy.
UNTAMO
A GIS toolbox customised for Posiva for TESM.
URL
Underground Research Laboratory.
UVic
Earth system model of the University of Victoria used for
the estimation of the climate evolution on a time scale of
10,000 years.
VAHA
Requirements management system at Posiva.
VSn
Variant Scenario n.
VTT
VTT Technical Research Centre of Finland.
VVER-440
Pressurised water reactor type at Loviisa.
YEA
Finnish abbreviation for Nuclear Energy Decree.
YEL
Finnish abbreviation for Nuclear Energy Act.
YJH
Finnish abbreviation for Nuclear Waste Management.
YVL
STUK’s (see STUK) regulatory guide series for nuclear
facilities.
12
13
FOREWORD
This report has been compiled and edited by Trevor Sumerling (Safety Assessment
Management Ltd, UK) and Margit Snellman (Saanio & Riekkola Oy). Other
contributors were Heini Laine, Nuria Marcos, Thomas Hjerpe, Pirjo Hellä and Annika
Hagros (all Saanio & Riekkola Oy) and Paul Smith (SAM Switzerland GmbH). .
The progress of the report was supervised by the SAFCA project group consisting of
Ari Ikonen and Marja Vuorio (Posiva Oy), Pirjo Hellä, Thomas Hjerpe, Heini Laine,
Nuria Marcos, Barbara Pastina and Margit Snellman (all Saanio & Riekkola Oy), and
Paul Smith (SAM Switzerland GmbH).
The report was reviewed at different stages by Juhani Vira (Posiva Oy), as well as the
SAFCA project group and the various report contributors mentioned above.
The final report review was carried out by the following individuals: Mike Thorne
(Mike Thorne and Associates Limited, UK), Lawrence Johnson (Nagra, Switzerland),
Ivars Neretnieks (KTH, Sweden), Allan Hedin (Executive Summary only) and Johan
Andersson (SKB, Sweden), and Paul Degnan (Catalyst Geoscience - Geological &
hydrogeological consultancy service, AUS). Their comments on the report are
appreciated
14
15
1
INTRODUCTION
This chapter introduces spent nuclear fuel management in Finland, the nature and
evolution of the hazards presented by spent nuclear fuel, Posiva’s programme for
implementing its disposal at the Olkiluoto site, and the TURVA-2012 safety case that is
required in support of Posiva’s application for a licence to construct the disposal
facility. The legal and regulatory context for the project, regulatory feedback from the
previous safety submission, and the structure of this report are also set out.
1.1
Spent nuclear fuel management in Finland
The spent nuclear fuel that arises from the generation of electricity at the Loviisa and
Olkiluoto nuclear power plants is classified as nuclear waste. The safety and security of
nuclear power plants, and the management of the wastes and spent nuclear fuel that
result, are closely controlled within a framework of international treaties and
agreements, and under national laws and regulations (see Section 1.5).
In Finland, according to the Nuclear Energy Act of 1987 and including amendments
made up to Act 410/2012:

nuclear waste – including spent nuclear fuel – generated in Finland must be
processed, stored and disposed of in Finland, and

all practical and financial measures to ensure the safe and secure management and
disposal are the responsibility of the nuclear power companies that produce the
waste.
This is consistent with both international treaties and conventions, and the ethical
consensus for the management of such waste.
In 1995, the two Finnish nuclear power companies, Teollisuuden Voima Oy (TVO) and
Imatran Voima Oy (later Fortum Power and Heat Oy (Fortum)) established Posiva Oy
(Posiva) to implement the final disposal programme for spent nuclear fuel and to carry
out related research, technical design and development (RTD or TKS in Finnish). Other
nuclear wastes are managed and disposed of by the power companies themselves.
On assignment by its owners, Fortum and TVO, Posiva will take care of the disposal of
spent fuel from the nuclear power plants at Loviisa and Olkiluoto. At Loviisa, two
Russian-designed pressurised water reactors (VVER-440) are in operation; at Olkiluoto,
two boiling water reactors (BWR) are operating and one pressurised water reactor
(PWR) is under construction. Plans exist for a fourth nuclear power unit at Olkiluoto. At
both sites there are facilities for interim storage of the spent fuel before disposal.
In 2001, the Parliament of Finland endorsed a Decision-in-Principle (DiP) whereby the
spent nuclear fuel produced by the operating Loviisa and Olkiluoto reactors will be
disposed of in a geological repository at Olkiluoto. This first DiP allowed for the
disposal of a maximum amount of spent nuclear fuel corresponding to 6500 tonnes of
uranium (tU) initially loaded into the reactors. Subsequently, additional DiPs were
issued in 2002 and 2010 allowing extension of the repository (up to 9000 tU) to
accommodate spent fuel from the operations of the OL3 reactor and the planned OL4
reactor.
16
1.2
Nature and evolution of the hazards presented by spent nuclear
fuel
1.2.1
Radiation risks
Exposure to radiation at high levels, or radioactive materials at high concentrations, can
lead to detrimental health effects. Humans have been exposed to naturally occurring
radiation and radioactive materials throughout their evolution, however, and it is
established that at low levels of radiation exposure the risks are very small, such that the
likelihood of any health detriment can only be estimated statistically, e.g. see NEA
(1997).
Two types of radiation health effect are distinguished: deterministic effects and
stochastic effects.

Deterministic effects are radiation effects the severity of which depends on the dose
received and the effects are regarded as detrimental above some dose threshold;
deterministic effects include tissue reactions, e.g. erythema, or organ damage.

Stochastic effects are effects that are not certain to occur and the probability (but not
the severity) of the effect is related to the radiation dose received and organs
exposed4; cancer is the primary concern.
Systems of radiological protection, for workers and the public, aim to manage radiation
exposures such that exposures are as low as reasonably achievable (ALARA) and in
any case remain below defined constraints such that detrimental deterministic effects
will not occur and that the probability of any stochastic health effect in any individual is
very low. For example, the Government Decree 736/2008 requirement that “the annual
dose5 to the most exposed people shall remain below the value of 0.1 mSv” (see
Appendix 1) implies that the annual increment to the lifetime risk of death to the most
exposed individual must be less than 10-5, i.e. less than one chance in 100,000.
1.2.2
Hazards and protection
Both humans and the environment have to be protected from the hazards presented by
spent nuclear fuel. The two main radiological hazards are6:

direct external radiation from the concentrated source as an intact unit, and

external irradiation and also internal irradiation due to ingestion or inhalation of
small amounts of the radioactive material (radionuclides) if it should become
dispersed in the environment.
4
The exact relationship between dose and effect is uncertain especially at low doses. For radiation protection purposes, it is
cautiously assumed that there is no dose threshold and the probability of a stochastic effect is proportional to the dose.
5
In this report, annual dose refers to the sum of the effective dose arising from external radiation within the period of one year, and
the committed effective dose from the intake of radioactive substances within the same year (GD 736/2008). Furthermore “dose”
refers to effective dose, unless otherwise explicitly stated. The effective dose is the tissue-weighted sum of the equivalent doses in
all specified tissues and organs of the body, where the tissue weighting factor represents the relative contribution of that tissue or
organ to the total health detriment resulting from uniform irradiation of the body. The equivalent doses are mean absorbed doses
in each tissue or organ, weighted by a factor that depends on the radiation type (ICRP 2007). Thus, effective dose is a quantity
designed to reflect the amount of health detriment likely to result from the dose, based on current radiobiological,
epidemiological and medical knowledge).
6
A third hazard for spent nuclear fuel is the potential for nuclear criticality, which is prevented by design, see Description of the
Disposal System, Section 6.3.6.
17
Protection against these hazards is provided by isolation and containment. That is the
spent nuclear fuel is placed:

in a secure or inaccessible location away from humans and the environment, such
that it cannot be reached except by deliberate actions, i.e. limiting accessibility;

within a system of robust engineered and natural barriers such as to prevent any
release of radionuclides, or to limit and attenuate any releases that might occur, i.e.
preventing or limiting releases.
1.2.3
Evolution of the hazard and implications for design
On its discharge from a nuclear reactor, spent nuclear fuel is highly radioactive, much
more radioactive than any material found on Earth at the present-day. Simple point
source calculations show that the external irradiation dose at 1 metre from 1 tonne of
spent fuel at one year after discharge from the reactor is sufficient to deliver a fatal
radiation dose within less than a minute (Hedin 1997). For this and other reasons spent
nuclear fuel is very carefully managed such that access to the fuel is prevented by both
physical barriers and procedural controls.
At discharge from the reactor, spent fuel assemblies are placed in cooling ponds at the
nuclear power plant sites. After a period of about 30 to 50 years, depending on the fuel
type and its irradiation history, the radioactive heat output has reduced from an initial
value of around 100 kW/tU shortly after at discharge to about 1 kW/tU. This level of
heat output is low enough for encapsulation and disposal to proceed.
The hundred-fold decrease in heat output results from the hundred-fold decrease in
radioactivity taking place in this time. Figure 1-1 shows the activity of Finnish OL3
spent fuel relative to the activity of the uranium ore needed for its manufacture.
The figure illustrates the relatively rapid decay of fission products and slower long-term
decay of actinides such that the rate of decay declines after about 30 to 50 years. This
indicates an appropriate time to move from storage to disposal, since the rate of decline
of heat output and activity is slow by this time.
Figure 1-1 also illustrates that after a few hundred thousand years the radioactivity of
the spent fuel is similar to that of the uranium ore from which it was manufactured. This
provides the basis for the design life-time of the canister in Posiva’s design, which is set
at hundreds of thousands of years. Design requirements for the other engineered barriers
are set so as to support the canister life-time requirement, as discussed in Section 2.2.1.
18
Figure 1-1. The total activity of one tonne of Finnish OL3 (denoted as EPR in the
figure) spent nuclear fuel with a burn-up of 60 MWd/kgU is shown relative to the
activity of 8 tonnes of natural uranium needed for its manufacture. The spent nuclear
fuel activity is shown by type of radionuclide.
1.3
Posiva’s programme for spent nuclear fuel disposal
TVO began investigation work for the disposal of spent nuclear fuel in the late 1970s.
Imatran Voima Oy (later Fortum) joined this work in the mid-1990s following the
prohibition of the export of spent nuclear fuel from Finland.
The programmes for location and development of a site for the disposal of spent nuclear
fuel were united under Posiva Oy on its formation in 1995. Following extensive site
investigations including national and regional surveys, and detailed investigations of
five sites, Posiva selected Olkiluoto as the preferred site in 1999. A Government DiP
endorsed by Parliament in 2001 affirmed that construction of a single disposal facility at
Olkiluoto, serving the disposal needs of the four nuclear power units owned by TVO
and Fortum, is in the overall best interests of society.
Figure 1-2 provides a timeline for nuclear waste management for the Olkiluoto and
Loviisa reactors in which the aim is to start the disposal of spent fuel around 2020.
19
Figure 1-2. Timeline for nuclear waste management relating to the Loviisa and Olkiluoto
reactors until 2020. The target is to begin disposal of spent nuclear fuel around 2020.
During the past few years, key activities in the programme have been related to:

completion of the investigations for site confirmation at Olkiluoto both through
analysis of data from surface-drilled characterisation holes and surveys and studies
carried out in the ONKALO underground research facility,

the design of the required surface and sub-surface facilities,

the development of the selected disposal technology to the level required for the
construction licence application, and

demonstration of the long-term safety of the disposal of spent nuclear fuel including
the preparation of a safety case presented as several separate reports, including the
present report.
Posiva’s RTD (research, development and technical design) phase for the years
2010−2012 was introduced in the TKS-2009 report (Posiva 2009a), which also provides
insight into developments from previous RTD phases. The programme for 2013−2015
(YJH-2012) was published earlier this year.
20
1.4
The TURVA-2012 safety case
1.4.1
What is a safety case?
Internationally, a safety case has been defined as a synthesis of evidence, analyses and
arguments that quantify and substantiate the safety, and the level of expert confidence in
the safety, of a geological disposal facility for radioactive waste (IAEA 2006, NEA
2004, 2012). The safety case is a key input to decision-making at several steps in the
repository planning and implementation process. It becomes more comprehensive and
rigorous as the programme progresses.
1.4.2
The TURVA-2012 safety case
TURVA-20127 is Posiva’s safety case in support of the Preliminary Safety Analysis
Report (PSAR 2012) and application for a construction licence for a repository for
disposal of spent nuclear fuel at the Olkiluoto site. It presents the long-term radiological
safety case, which is concerned with the evolution, performance and safety of the
disposal system following emplacement of the spent nuclear fuel. Other aspects of
safety are dealt with in other parts of the PSAR.
The TURVA-2012 safety case builds on an ‘Interim Summary Report of the Safety
Case 2009’ that was published in 2010 (Posiva 2010a). The direction for further work to
bring the safety case to maturity for submission in support of an application for a
construction licence was laid out within Posiva’s ‘Review of Current Status and Future
Plans for 2010-2012’ (TKS-2009, Posiva 2009a).
TURVA-2012 is addressed to the nuclear regulator, STUK, and other national
stakeholders as well as the international scientific and technical communities engaged in
the discussion on nuclear waste disposal. STUK will review the safety case and related
topical reports as part of its evaluation of the construction licence application and the
PSAR. STUK will then give a statement on the construction licence application, which
will form the basis for the Government judgement on issuance of the construction
licence.
The TURVA-2012 safety case presents the arguments for the long-term radiological
safety of the planned disposal system. It includes:




7
a description of the spent nuclear fuel to be disposed of in the geological repository
a description of the natural and engineered barriers that the repository system
provide, a definition of the safety functions and targets set for these and a
description of the present understanding of the processes that may affect the
evolution and performance of the repository system and the surface environment;
a performance assessment systematically analysing the ability of the repository
system to provide containment and isolation of the spent nuclear fuel for as long as
it remains hazardous;
a definition of the lines of evolution that may lead to failure of the canisters
containing the spent nuclear fuel and to the releases of radionuclides (scenarios);
TURVA means safety in Finnish.
21




analyses of the potential rates of release of radionuclides from the failed canisters,
the retention, transport and distribution of radionuclides within the repository
system and surface environment and the potential radiation doses to humans, plants
and animals including the associated uncertainties, and an evaluation of their
impacts;
the models and data used in the description of the evolution of the repository system
and the development of the surface environment and for the analysis of activity
releases and dose assessment;
a range of qualitative evidence and arguments that complement and support the
reliability of the results of the quantitative analyses; and
a comparison of the outcome of the analyses with safety requirements.
Aspects of safety related to the period of operations are dealt with in other parts of the
PSAR.
The safety case and supporting analyses will be further developed towards a Final
Safety Analysis Report (FSAR) that will be submitted at the time of the operational
licence application.
1.4.3
The TURVA-2012 portfolio
The TURVA-2012 safety case is presented in a portfolio of safety case reports and
supporting documents (Figure 1-3), and a synthesis that brings together all the lines of
arguments for safety, including the main starting points, methodology, results and
conclusions. The report names and brief descriptions of their contents are given in the
figure. In this report, all TURVA-2012 portfolio reports are referenced using the report
title (as in Figure 1-3) in italics. The full titles and report numbers are listed at the
beginning of the reference list.
The safety case portfolio has been developed based on the plan published in 2008
(Posiva 2008), which updated an earlier plan published in 2005 (Vieno & Ikonen 2005).
Since 2008, the safety case and its presentation in the portfolio have been developed
based on the feedback received from STUK. The updated safety case plan, which has
directed the development of the TURVA-2012 safety case, places emphasis on quality
assurance and control procedures and their documentation as well as on consistent
handling of different types of uncertainties.
Each safety case report has been subject to review in two steps: first, an internal review
by safety case experts and other subject-matter experts within Posiva’s RTD programme
and, second, a review by external experts. Records of the external experts’ comments
and Posiva’s responses are documented and archived.
22
TURVA-2012
Synthesis
Description of the overall methodology of analysis, bringing together all the lines of arguments for
safety, and the statement of confidence and the evaluation of compliance with long-term safety
constraints
Site Description
Biosphere Description
Understanding of the present state and past
evolution of the host rock
Understanding of the present state and evolution of the
surface environment
Design Basis
Performance targets and target properties for the repository system
Production Lines
Design, production and initial state of the EBS and the underground openings
Description of the Disposal System
Summary of the initial state of the repository system and present state of the surface environment
Features, Events and Processes
General description of features, events and processes affecting the disposal system
Performance Assessment
Analysis of the performance of the repository system and evaluation of the fulfillment of performance
targets and target properties
Formulation of Radionuclide Release Scenarios
Description of climate evolution and definition of release scenarios
Models and Data for the
Repository System
Biosphere Data Basis
Models and data used in the performance
assessment and in the analysis of the
radionuclide release scenarios
Data used in the biosphere assessment and summary
of models
Biosphere Assessment: Modelling reports
Description of the models and detailed modelling of surface environment
Assessment of Radionuclide
Release Scenarios for the
Repository System
Biosphere Assessment
Analysis of releases and calculation of doses and activity fluxes.
Complementary Considerations
Supporting evidence incl. natural and anthropogenic analogues
Main reports
Main supporting documents
Figure 1-3. The TURVA-2012 safety case portfolio. The portfolio consists of safety case
reports (green boxes) and supporting reports (blue boxes); brief descriptions of the
contents are given (white boxes). Disposal system = repository system + surface
environment.
23
1.4.4
Quality assurance
The quality of the TURVA-2012 safety case has been assured through documented
procedures in accordance with Posiva’s quality management principle, which is based
on the ISO 9001:2008 standard. A graded approach is applied whereby the primary
emphasis is on quality control of those activities that have a direct bearing on long-term
safety.
Posiva’s general quality guidelines are applied to the composition and quality
management of portfolio reports and to the appointment of expert reviewers. Special
attention is paid to the management of the processes that are applied to produce the
safety case and its foundations. The purpose of this enhanced process control is to
provide full traceability and transparency of the data, assumptions, models, calculations
and results. The regulatory requirements on quality assurance are also followed.
The overall plan, goals and constraints for the TURVA-2012 safety case production
process are presented in Posiva’s Safety Case Plan 2008. The organisation of the
TURVA-2012 safety case production process is referred to as SAFCA. The details of
how the Safety Case Plan is being implemented are described in the SAFCA project
plan. The work is managed and coordinated by a SAFCA project group and supervised
by a steering group.
A SAFCA quality co-ordinator (QC) has been designated for activities related to quality
assurance measures applied to the production of the safety case. Improvements are
made to the process as deemed useful or necessary. The QC is also responsible for the
coordination of the expert reviews, maintenance of schedules, and review and approval
of the reports.
Posiva’s quality manager (QM) undertakes regular auditing of the safety case
production process.
Further details of the quality management system and its application to the production
of the TURVA-2012 safety case are given in Section 2.5.
1.5
Legal and regulatory context for the management of spent fuel
1.5.1
International treaties and agreements
Finland is signatory to international conventions and treaties that define national
obligations, standards of practice and protection, and reporting requirements for dealing
with spent nuclear fuel. These include the:

Joint Convention on the Safety of Spent Fuel Management and on the Safety of
Radioactive Waste Management;

Treaty on the Non-Proliferation of Nuclear Weapons;

Convention on the Physical Protection of Nuclear Material;

Treaties and Directives of the European Union.
24
Finland’s international obligations for safe and secure management of nuclear materials
and wastes, including spent nuclear fuel, on Finnish territory are translated into national
laws. In particular the decision to seek permanent disposal of spent nuclear fuel in
Finland is fully in accord with all relevant international treaties and agreements. The
framework of legal and regulatory requirements to ensure safe and secure final disposal
are as described in the following sub-sections.
1.5.2
Legal requirements
The basis for the use of nuclear energy in Finland is given in the Nuclear Energy Act
(YEL 990/1987) and Nuclear Energy Decree (YEA 161/1987), which came into effect
in 1988. According to the Nuclear Energy Act:
Nuclear waste shall be managed so that after disposal of the waste no radiation
exposure is caused, which would exceed the level considered acceptable at the time the
final disposal is implemented.
and
The disposal of nuclear waste in a manner intended as permanent shall be planned
giving priority to safety and so that ensuring long-term safety does not require the
surveillance of the final disposal site.
According to the law, the Ministry of Employment and the Economy (TEM; previously
the Ministry of Trade and Industry, KTM) decides on the principles to be followed in
waste management of spent fuel and other nuclear waste.
The safe management of nuclear waste is the responsibility of the utilities that generate
the waste. The law also stipulates that the parties under the nuclear waste obligation
must regularly submit to the Ministry a report setting out the responsible parties’ plans
concerning the implementation of the measures associated with nuclear waste
management and the preparation of these measures. Following the entry into force of
the amendment to the Nuclear Energy Act in 2009, the reports must now be submitted
every three years and include a description of the measures taken during the last threeyear period, as well as an outline of the plans for the next three years. The most recent
report was submitted in 2012.
The schedule for the disposal of spent nuclear fuel was first defined by the Government
in 1983 and slightly modified by the Ministry of Trade and Industry (KTM) in 2003
(9/815/2003). According to the Ministry decision, the parties under the nuclear waste
management obligation shall, separately, together or through Posiva Oy, present all
reports and plans required to obtain a construction licence for a disposal facility for
spent nuclear fuel by the end of 2012. The disposal facility is expected to become
operational around 2020.
The legislation concerning nuclear energy was updated in 2008. As part of the
legislative reform, a number of the relevant Government Decisions were replaced with
Government Decrees (GD). The Decrees entered into force on 1st December 2008. The
Government Decision (478/1999) regarding the safety of disposal of spent nuclear fuel,
which particularly applied to the disposal facility, was replaced by Government Decree
25
736/2008, issued 27 November 2008. Government Decree 736/2008 sets the legal
requirements regarding the safety of disposal of spent nuclear fuel.
The scope and contents of Government Decree 736/2008 are summarised in
Appendix 1.
1.5.3
Regulatory guidance
The Radiation and Nuclear Safety Authority (STUK) issues guidance documents on the
practical fulfilment of the legal requirements set out in Government Decree 736/2008.
These guides also set out STUK’s expectation for the content, quality and radiological
criteria to be met by any safety case submission for disposal of nuclear waste in Finland.
A total of five Guides apply to the disposal of spent nuclear fuel. The most relevant here
are Guide YVL D.3, which provides guidance on the handling, storage and
encapsulation of spent nuclear fuel and Guide YVL D.58, which provides guidance on
the planning of the disposal method, design and operation of the disposal facility, safety
requirements and demonstration of compliance with safety requirements, regulatory
control and on the compilation of a safety case. Other Guides deal with nuclear nonproliferation control (Guide YVL D.1), transport of nuclear material and nuclear waste
(Guide YVL D.2), and nuclear waste management and decommissioning activities
(Guide YVL D.4).
Guide YVL D.5 applies to disposal of all types of nuclear waste and provides guidance
related to both operational and long-term safety. Key requirements, stemming from
GD 736/2008 and set out in Guide YVL D.5, are summarised in Table 1-1. The Guide
provides substantial additional information on the meaning of, and evidence needed to
show compliance with, these requirements.
The Guide YVL D.5 does not specify the precise time frames over which assessments
are needed. Posiva consider, however, that radiation doses can be assessed, assuming
human habits, nutritional needs and metabolism remain unchanged, with sufficient
reliability over a period of up to 10,000 years, and that the fulfilment of the safety
functions of the repository system and the release of radionuclide to the surface
environment can be reasonably assessed up to one million years after repository closure.
8
The Guides YVL D.3 and YVL D.5 are available in draft form. STUK has agreed that the licence application can be based on
version 4 of both Guides (version 17.3.2011 has been used).
26
Table 1-1. Synthesis of key requirements for long-term safety from STUK´s Guide YVL
D.5. Refer to the Guide for actual wording and context.
Related to long-term radiological impacts





For expected evolution scenarios, and in the period during which the radiation exposure can be
assessed with sufficient reliability (at least over several millennia):
 the annual dose to the most exposed people shall remain below the value of 0.1 mSv;
 the average annual doses to other people shall remain insignificantly low.
In the longer term, the radiation impacts arising from disposal can at a maximum be equivalent to
those arising from natural radioactive substances in Earth’s crust, and on a large scale should remain
insignificantly low. The nuclide-specific constraints on releases to the environment (average release of
radioactive substances per annum) are specified in YVL D.5.
For the activity releases that arise from the expected evolution scenarios, the sum of the ratios
between the nuclide-specific activity release rates and the respective constraints given in YVL D.5
shall be less than one (evaluated for the release rates for radionuclides from the geosphere to the
biosphere),
The importance of unlikely events impairing long-term safety shall be assessed, and whenever
practicable, the radiation impacts caused shall be assessed quantitatively. The resulting annual
effective dose or activity release shall be calculated and multiplied by its estimated probability of
occurrence. The obtained expectation value shall be below the dose constraint referred to above or
release constraints given inTable 2-4.
The assessed radiation exposures to fauna and flora shall remain clearly below the levels that could
cause decline in biodiversity or other significant detriment to any living population.
Related to providing long-term safety





Disposal shall be implemented in stages, with particular attention paid to aspects affecting long-term
safety.
The long-term safety of disposal shall be based on safety functions achieved through mutually
complementary barriers so that a deficiency of an individual safety function or a predictable geological
change will not jeopardise the long-term safety.
Targets shall be specified for the performance of each safety function based on high quality scientific
knowledge and expert judgement.
For spent fuel, the safety functions provided by the engineered barriers shall limit effectively the
release of radioactive substances into the bedrock for at least 10,000 years.
The characteristics of the host rock shall be favourable for the long-term performance of engineered
barriers and with respect to the groundwater flow regime at the disposal site.
1.5.4
Safety submissions in support of construction and operating licences
According to the decision of the Ministry of Trade and Industry (KTM) in 2003 (see
above), Posiva is to submit an application for a construction licence for a disposal
facility at Olkiluoto by the end of 2012. This will be followed by an application to be
made in 2018 for a licence to begin disposal operations. These applications will be
accompanied by, respectively:

a Preliminary Safety Analysis Report (PSAR) in support of the construction licence
application and

a Final Safety Analysis Report (FSAR) in support of the operating licence
application.
The PSAR and FSAR are to be prepared according to structures that are specified by
STUK based on legislation and regulations. The PSAR and the FSAR will include the
case for the operation of the disposal facility as a major nuclear facility and will cover
both conventional and radiological safety during construction, operation, closure and in
27
the long term following completion of disposal operations and closure of the disposal
facility. In particular, STUK specifies that the safety of the disposal facility in the long
term should be assured through a ‘safety case’ for disposal.
1.6
Feedback from STUK on the Interim Summary Report 2009
Following Posiva’s submission of the Interim Summary Report of the Safety Case 2009
(Posiva 2010), the nuclear regulator, STUK, evaluated Posiva’s preparedness to
demonstrate long-term safety and operational safety, and the fulfilment of the safety
requirements for nuclear waste disposal against the GD 736/2008.
STUK noted that the material that Posiva submitted for the evaluation in 2009 covered
the required elements at least on the design level, and that Posiva had done a
considerable amount of research, development and design work preparing for nuclear
waste disposal, and presented the acquired results in its reports.
As a general observation, STUK noted that although Posiva presented a considerable
amount of information in several safety-related areas, the information is not always
consistently presented and sometimes it is difficult to trace. Therefore, in 2010, STUK
was not able to reach a conclusion on the completeness of the material that Posiva
proposed for the construction licence application. In addition, STUK considered that the
material submitted had some shortcomings in demonstrating the fulfilment of the
requirements and substantiating the conclusions drawn, because of limitations in the
reasoning and analysis.
STUK’s safety evaluation report (STUK 2011) provided feedback and advice that has
been taken into account as key issues that have been prioritised within Posiva’s RTD
programme and in the development of the TURVA-2012 safety case. The present safety
case shows that the remaining shortcomings and uncertainties have an insignificant
impact on long-term safety.
The feedback has also been taken into account in the systematic structuring of the safety
case and the reports included in the portfolio. The formulation of radionuclide release
scenarios (see Formulation of Radionuclide Release Scenarios) follows a systematic
approach taking into account the safety functions of the barriers of the repository system
and the uncertainties in the features, events, and processes (see Features, Events and
Processes) that may affect the disposal system from the emplacement of the first
canister until the far future. Compliance with the performance targets and target
properties (see Design Basis), which assures that the safety functions of the engineered
and natural barriers will be achieved, is shown in Performance Assessment. This takes
account of uncertainties in the initial state of the barriers and in the evolution of the
repository system.
1.7
Structure of this report
The present report is the TURVA-2012 Synthesis. It provides a description of the
overall methodology of safety case, bringing together all the lines of argument, the
evaluation of compliance with long-term safety constraints, and a statement of
confidence in the TURVA-2012 safety case.
28
The structure of the report is as follows:

Chapter 1 introduces spent nuclear fuel management in Finland, the nature and
evolution of the hazards presented by spent nuclear fuel, Posiva’s programme for
implementing disposal at the Olkiluoto site and the TURVA-2012 safety case. The
legal and regulatory context for the project, regulatory feedback from the previous
safety submission, and the structure of this report are also set out.

Chapter 2 introduces the KBS-3 disposal method and key features of the Olkiluoto
site that bear on the development of the disposal concept and safety case for disposal
of spent nuclear fuel at the site. The design methodology and assessment
methodology are then set out. The Posiva quality management system, its
application to production of the TURVA-2012 safety case, and model and data
quality processes are also outlined.

Chapter 3 presents a summary description of the disposal system in its initial state;
that is, descriptions of the host rock and surface environment, and of the spent
nuclear fuel and engineered barriers (canister, buffer, backfill and closure). These
descriptions provide the basis for assessing the performance of the repository system
and the safety of the disposal system.

Chapter 4 describes the identification and screening of features, events and
processes (FEPs) and development of a database of FEPs relevant to the
performance assessment and analysis of potential radionuclide releases and
radiological impacts. It also describes the onward use of the FEP descriptions in
performance assessment and radiological modelling and analyses, and outlines the
potential future lines of evolution of the repository system and surface environment.

Chapter 5 describes the various models and data needed for the analyses supporting
the safety case. The models are of four types: models describing the climate
evolution and climate-driven processes; models to represent the FEPs that determine
the evolution of the disposal system and that have to be taken into account to assess
the performance of the engineered barriers and conditions in the host rock; models
to analyse radionuclide release and transport from the near field through the
geosphere to the surface environment; and models for biosphere assessment
including models representing landscape development, radionuclide transport in the
surface environment and potential doses or dose rates to humans, plants and
animals.

Chapter 6 summarises the performance of the repository system and demonstrates
fulfilment of the performance targets and target properties for the engineered
barriers and host rock. The performance assessment takes account of the expected
thermal, hydraulic, mechanical and chemical (THMC) evolution of the repository
system, and uncertainties in the expected evolution. The performance and fulfilment
of performance requirements are considered for three periods: during excavation and
operation up to closure; in the post-closure period during the next 10,000 years;
beyond 10,000 years over repeated glacial cycles up to one million years.

Chapter 7 summarises the formulation of radionuclide release scenarios and
calculation cases. These focus on deviations in conditions and uncertainties in
evolution, and unexpected events that could lead to the release of radionuclides.
Scenarios are defined as a base scenario, variant scenarios and disturbance
29
scenarios; in relation to these a set of calculation cases (reference case, sensitivity
cases, “what if” cases) is defined.

Chapter 8 summarises the assessment of the radionuclide release scenarios and cases
as defined in Chapter 7. This includes analysis of radionuclide release and transport
in the repository system, simulation of the surface environment and analysis of
potential radiological impacts on humans, plants and animals. Base, variant and
disturbance scenarios are analysed, and uncertainties within these scenarios are
investigated using a range of deterministic calculation cases as well as Monte Carlo
simulations. Probabilistic sensitivity analyses have been carried out to assess
sensitivities to parameter values and to explore the consequences of alternative
model assumptions.

Chapter 9 outlines complementary considerations that provide additional evidence
for the long-term safety of disposal. Complementary considerations and additional
evidence related to the choice of the geological disposal concept, the robustness of
the KBS-3 method and the suitability of the Olkiluoto site. Selected results from
evaluations of a range of complementary indicators for the repository system are
also presented.

Chapter 10 confirms the compliance with legal and regulatory requirements based
on findings and results presented in the preceding chapters. It also outlines the main
research and development needs during the coming years.

Chapter 11 provides a statement of confidence, confirming that the TURVA-2012
safety case shows, at a level of detail appropriate to the repository construction
licence application, that the safe disposal of spent nuclear fuel can be implemented
through the KBS-3 method at the Olkiluoto site.
30
31
2
METHODOLOGY
This chapter introduces first the KBS-3 disposal method and key features of the
Olkiluoto site that bear on the development of the disposal concept and safety case for
disposal of spent nuclear fuel at the site. The design and assessment methodology are
then outlined.
In the design methodology, the safety concept and safety functions are defined based on
long-term safety principles. This leads to the development of the design basis, which
includes the performance targets for the engineered barriers, target properties for host
rock and design requirements for the repository system. The definition of the
performance targets for the engineered barriers and target properties for the host rock,
take into account the different loads and interactions, and features, events and processes
(FEPs) that may act on the repository system at the time of canister emplacement and in
the long-term. From the performance targets and target properties the design
requirements are derived. Then, design specifications are worked out such that the
fulfilment of these requirements can be verified during implementation.
The performance of the repository system and its components is analysed taking into
account the expected lines of evolution and the uncertainties involved. Conditions and
events that could give rise to the release of radionuclides are identified, which provide
the basis for the formulation of radionuclide release scenarios. These scenarios provide
the basis for analyses of radionuclide release and transport, and of radiological impacts
to humans, plants and animals. Complementary considerations and supporting evidence
are also assembled in support of the performance and radiological safety analyses.
2.1
The KBS-3 method and the Olkiluoto site
The 2001 DiP states that disposal of spent nuclear fuel shall take place in a geological
repository at the Olkiluoto site, developed according to the KBS-3 method. The KBS-3
method and key features of the Olkiluoto site are outlined in the following subsections.
Further information on the KBS-3V design and its technical realisation at the Olkiluoto
site is given in Chapter 3.
2.1.1
The KBS-3 disposal method
The KBS-3 method was conceived as a solution for the disposal of spent nuclear fuel in
Sweden in the early 1980s. Since then, the method has been developed and its key
elements tested by SKB in Sweden and Posiva in Finland, and in joint projects. The
method envisages the disposal of spent nuclear fuel within a system of multiple barriers,
which consists of engineered barriers and the natural barrier provided by the host rock.
Posiva’s reference design in the construction licence application is based on vertical
emplacement of the spent nuclear fuel canisters individually in deposition holes (KBS3V)9.
9
A potential alternative design of horizontal emplacement of multiple canisters in deposition drifts (KBS-3H) is being jointly
developed by the Swedish Nuclear Fuel and Waste Management Company (SKB) and Posiva. The present safety case is based on
the KBS-3V reference design.
32
In the reference design the repository is constructed on a single level with the floor of
the deposition tunnels at a depth of between 400 and 450 m below the ground surface in
the Olkiluoto bedrock (Figure 2-1).
The spent nuclear fuel elements are placed into copper canisters with cast iron loadbearing inserts, and the canisters are emplaced vertically in individual deposition holes
bored in the floors of the deposition tunnels. The canisters are surrounded by a swelling
clay buffer material that separates them from the bedrock. The deposition tunnels,
central tunnels, access tunnel, shafts and the other underground openings are backfilled
with materials that help to restore the natural conditions in the bedrock after closure.
2.1.2
The Olkiluoto site
The Olkiluoto site, located on the coast of south-western Finland (Figure 2-2), has been
investigated as a potential site for disposal of spent nuclear fuel for over 25 years. This
has included the construction of an underground rock characterisation facility − the
ONKALO. Olkiluoto Island has an area of about 10 km2; the surface facilities including
the encapsulation plant will occupy about 0.1 km2; according to the current design and
required capacity, the deposition tunnels and other tunnels will occupy about 2 km2.
The Olkiluoto site, as seen today, is the consequence of events and processes that have
taken place over billions of years, from those reflected in the geological properties of
the rocks forming the geosphere, to the much shorter-term changes related to more
recent climate-driven processes: mainly changes in groundwater flow and groundwater
composition and the geomechanical response to crustal movements related to glacial
loading and unloading. A detailed description of the Olkiluoto site is given in Site
Description, which describes the host rock, and in Biosphere Description, which
describes the surface environment.
Key features of the Olkiluoto site with respect to its suitability for the deep geological
disposal of radioactive waste include:

a stable tectonic situation within the Fennoscandian Shield, away from active plate
margins;

good quality crystalline bedrock suitable for the excavation of self-supporting
tunnels and other underground openings, such as deposition holes, technical rooms
and shafts;

reducing conditions at disposal depths and also otherwise favourable geochemical
characteristics of the groundwater, and

low groundwater flow at depth occurs currently, as it has occurred over a long
period in the past and it is expected to persist for a long period into the future.
33
Figure 2-1. Schematic illustration of the KBS-3V design at the Olkiluoto site.
Figure 2-2. Olkiluoto Island is situated on the coast of the Baltic Sea in south-western
Finland. Photograph by Helifoto Oy.
34
The conditions in the Olkiluoto bedrock provide favourable conditions for longevity and
reliable functioning of the engineered barrier system (EBS). In addition, the low
groundwater flows, and physical and chemical retardation processes, limit the
movement of radionuclides.
Key features and processes that provide constraints on the layout of the repository and
other underground openings, or that must be taken into account in the assessment of
long-term performance and safety include:

presence of deformation and fractured zones, displaying more mixed geotechnical
properties and, in some cases, increased hydraulic activity;

higher rock stress at depth, which may cause disturbance to the rock making
underground openings less stable;

temperature and thermal conductivity of rock and residual heat output of the spent
nuclear fuel;

high salinity of groundwater at depth which may affect the performance of the
engineered barrier system;

continuing post-glacial crustal uplift and, in the longer term, climatic cooling and
glaciation, leading to changes in rock stress and potential changes in groundwater
flow and hydrochemistry, e.g. influx of dilute glacial melt waters into the host rock.
Post-glacial crustal uplift and eustatic sea-level changes lead to relative changes in local
sea level. Thus, a transition from a coastal to an inland environment is expected over the
next millennia, which must be taken into account in the assessments of the potential
impact of releases of radionuclides to the biosphere.
2.2
Design methodology
Requirements management
Posiva has developed a robust system design for geological disposal of spent nuclear
fuel at Olkiluoto through a formal requirements management system (VAHA). This
provides a rigorous, traceable method of translating the safety principles and the safety
concept to a set of safety functions, performance requirements, design requirements and
design specifications for the various barriers, i.e. a specification for realisation of the
concept at the Olkiluoto site. The VAHA sets out:

At Level 1, the stakeholder requirements that come from laws, decisions-inprinciple, regulatory requirements and other stakeholder requirements;

At Level 2, the long-term safety principles, which lead to a definition of the safety
concept and safety functions;

At Level 3, the performance requirements consisting of performance targets for the
engineered barriers, and target properties for the host rock, such that the required
safety functions are fulfilled;

At Level 4, the design requirements for the engineered barriers, and the
underground openings, including rock suitability classification criteria (RSC
criteria), such that the performance requirements will be met;
35

At Level 5, the design specifications, which are the detailed specifications to be
used in the design, construction and manufacturing.
2.2.1
Safety principles, safety concept and safety functions
Long-term safety principles
The long-term safety principles set out for the KBS-3 method are based on the use of a
multi-barrier disposal system consisting of engineered barriers and host rock. The role
of the engineered barriers is to provide the primary containment against the release of
radionuclides. The host rock should provide favourable conditions for the long-term
performance of the engineered barriers, but also limit and retard the transport of
radionuclides. The multi-barrier system as a whole should be able to protect the living
environment even if one of the barriers turns out to be deficient.
The long-term safety principles are described at Level 2 of the VAHA as follows.
1. The spent fuel elements are disposed of in a repository located deep in the Olkiluoto
bedrock. The release of radionuclides is prevented with a multi-barrier disposal
system consisting of a system of engineered barriers (EBS) and host rock such that
the system effectively isolates the radionuclides from the living environment.
2. The engineered barrier system consists of
a) canister to contain the radionuclides as long as these could cause significant
harm to the environment
b) buffer between the canisters and the host rock to protect the canisters as long as
containment of radionuclides is needed
c) deposition tunnel backfill and plugs to keep the buffer in place and help restore
the natural conditions in the host rock
d) the closure, i.e. the backfill and sealing structures to decouple the repository
from the surface environment.
3. The host rock and depth of the repository are selected in such a way as to make it
possible for the EBS to fulfill the functions of containment and isolation described
above.
4. Should any of the canisters start to leak, the repository system as a whole will
hinder or retard releases of radionuclides to the surface environment to the level
required by the long-term safety criteria.
The safety concept
The safety concept (Figure 2-3) is a conceptual description of how these principles are
applied together to achieve safe disposal of spent nuclear fuel in the present-day and
future conditions of the Olkiluoto site.
Containment of radionuclide inventory associated with the spent nuclear fuel is
provided first and foremost by encapsulating the fuel in sealed (gas-tight and watertight) copper-iron canisters. The other EBS components (buffer, backfill and closure)
provide favourable near-field conditions for the canisters to remain intact and, in the
event of canister failure, slow down or limit releases of radionuclides from the canister.
The containment of radionuclides is ensured by the proven technical quality of the EBS.
36
Other elements of the safety concept include sufficient depth for the repository,
favourable and predictable bedrock and groundwater conditions and well-characterised
material properties of both the bedrock and the EBS (the key safety features of the
system in Figure 2-3). A robust system design ensures that single deficiencies in the
design or implementation of the design, or uncertainties in future conditions, do not lead
to significant weakening of the overall safe functioning of the repository system.
Safety functions are assigned to the components of the engineered barrier system (EBS)
and the host rock as shown in Table 2-1.
The purpose of the multiple and complementary barriers, as described above, are to
ensure that any single detrimental phenomenon or uncertainty cannot undermine the
safety of the whole system, as required by Government Decree 736/2008:
“The long-term safety of disposal shall be based on redundant barriers so that
deficiency in one of the barriers or a predictable geological change does not jeopardise
long-term safety.”
SAFE DISPOSAL
FAVOURABLE, PREDICTABLE BEDROCK
AND GROUNDWATER CONDITIONS
PROVEN TECHNICAL QUALITY
OF THE EBS
Slow diffusive
transport in the buffer
Slow release from the
spent fuel matrix
Retention and retardation of
radionuclides
Slow transport in the
geosphere
FAVOURABLE NEAR-FIELD
CONDITIONS FOR THE
CANISTER
LONG-TERM ISOLATION AND CONTAINMENT
WELL-CHARACTERISED MATERIAL
PROPERTIES
SUFFICIENT DEPTH
ROBUST SYSTEM DESIGN
Figure 2-3. Outline of the safety concept for a KBS-3 type repository for spent fuel in a
crystalline bedrock (adapted from Posiva 2003). The safety concept is based on a
robust system design. Orange pillars and blocks indicate the primary safety features
and properties of the disposal system. Green pillars and blocks indicate the secondary
safety features that may become important in the event of a radionuclide release from a
canister.
37
Thus, for example, features and processes associated with the green columns in Figure
2-3 are, at least partly, independent of each other.
The characterisation of the Olkiluoto site for the repository design is focused on a
volume of bedrock situated between 400 and 500 metres below the ground surface. At
such depths, the likelihood of inadvertent human intrusion is low, and favourable and
predictable bedrock and groundwater conditions, such as reducing conditions, low
frequency of water-conducting fractures and slow movement of groundwater, are found.
The depth range is consistent with guidance in YVL Guide D.5 according to which the
repository should be located:
“...at the depth of several hundreds of metres in order to mitigate adequately the
impacts from aboveground natural phenomena, such as glaciation, and human
actions.” (YVL D.5, paragraph 412)
Should any initially defective canisters be present or subsequent breaches in the
canisters occur, the consequences of radionuclide releases for humans and other biota
inhabiting the surface environment will be mitigated by the slow release from the spent
nuclear fuel matrix, slow diffusive transport in the buffer and backfill, and slow
radionuclide transport in the geosphere. Together, the engineered barriers and the rock
provide for retention and retardation of radionuclides. Radioactive decay during
transport also decreases activity releases into the environment. These are depicted in
Figure 2-3 as secondary features of the safety concept (green blocks and pillars) since
they become important only in the event of canister failure.
Safety functions
The long-term safety of disposal is based on a system of natural and engineered barriers
which all have their roles in establishing the required long-term safety of the repository
system. These roles constitute the safety functions of the barriers. According to YVL
D.5, paragraph 405:
“Engineered barriers and their safety functions may consist of waste matrix, in which
radioactive substances are incorporated; hermetic, corrosion resistant and
mechanically strong container, in which the waste is enclosed; chemical environment
around waste packages, which limits the dissolution and migration of radioactive
substances; material around waste canisters (the buffer), which provides containment
and yields to minor rock movements; other containment structures in the emplacement
rooms; backfilling materials and sealing structures, which limit transport of radioactive
substances through excavated rooms.”
Posiva’s definition of safety functions follows this guidance with respect to engineered
barriers (canister, buffer, backfill and closure).
Most of the activity in the spent nuclear fuel is contained in a ceramic matrix (UO2) that
is resistant to dissolution in the expected repository conditions. The slow release of
radionuclides from the spent fuel matrix in the event of canister failure is part of
Posiva’s safety concept. However, no safety functions or performance requirements are
assigned to spent nuclear fuel; rather, the properties of the spent fuel are used as starting
point in the design of the disposal system.
38
According to YVL Guide D.5, paragraph 406, the natural barriers and their safety
functions may consist of

“stable and intact rock with low groundwater flow rate around disposal canisters

rock around waste emplacement rooms where low groundwater flow, reducing and
also otherwise favourable groundwater chemistry and retardation of dissolved
substances in rock limit the mobility of radionuclides

protection provided by the host rock against natural phenomena and human
actions.”
In Posiva’s repository concept, the natural barriers consist of the host rock, which
carries the safety functions mentioned in the YVL Guide D.5, paragraph 406.
The surface environment does not provide any safety functions; rather it is considered
the object to be protected by the repository system.
Thus, in summary, the disposal system consists of the surface environment, which is the
object to be protected, plus the repository system, which consists of the spent nuclear
fuel (the source of hazard) and the engineered and natural barriers that provide safety
functions and thus protection from the hazard (Figure 2-4)
The safety functions of the components of the engineered barrier system and host rock
(barriers) as considered by Posiva are summarised in Table 2-1.
Figure 2-4. The components of the disposal system. Safety functions are assigned to the
natural barrier (host rock) and to components of the engineered barrier system
(closure, backfill, buffer and canister).
39
Table 2-1. Safety functions assigned to the barriers (EBS components and host rock) in
Posiva’s KBS-3V repository.
Barrier
Safety functions
Canister
Ensure a prolonged period of containment of the spent fuel. This safety function rests
first and foremost on the mechanical strength of the canister’s cast iron insert and the
corrosion resistance of the copper surrounding it.
Buffer
Contribute to mechanical, geochemical and hydrogeological conditions that are
predictable and favourable to the canister.
Protect canisters from external processes that could compromise the safety function
of complete containment of the spent nuclear fuel and associated radionuclides.
Limit and retard radionuclide releases in the event of canister failure.
Deposition tunnel
backfill
Contribute to favourable and predictable mechanical, geochemical and
hydrogeological conditions for the buffer and canisters.
Limit and retard radionuclide releases in the possible event of canister failure.
Contribute to the mechanical stability of the rock adjacent to the deposition tunnels.
Host rock
Isolate the spent nuclear fuel repository from the surface environment and normal
habitats for humans, plants and animals and limit the possibility of human intrusion,
and isolate the repository from changing conditions at the ground surface.
Provide favourable and predictable mechanical, geochemical and hydrogeological
conditions for the engineered barriers.
Limit the transport and retard the migration of harmful substances that could be
released from the repository.
Closure
Prevent the underground openings from compromising the long-term isolation of the
repository from the surface environment and normal habitats for humans, plants and
animals.
Contribute to favourable and predictable geochemical and hydrogeological conditions
for the other engineered barriers by preventing the formation of significant water
conductive flow paths through the openings.
Limit and retard inflow to and release of harmful substances from the repository.
2.2.2
Performance targets and target properties
The safety functions described above are implemented in the proposed design through a
set of technical design requirements, based on performance requirements that are
defined for each barrier of the repository system. The performance requirements are
expressed as performance targets (engineered barriers) and target properties (host rock)
that the system should meet in the long-term to provide the required level of safety.
In defining the performance targets for the engineered barriers, implementation aspects
also have to be considered: the performance targets have to be set considering, on the
one hand, the long-term safety aspects and, on the other hand, the need for the design
and implementation to be robust as that is fundamental to the safety concept.
The definition of the performance targets and target properties requires the
identification of the different loads and interactions that may act on the repository
system at the time of canister emplacement and in the long-term. To achieve this, the
potential future conditions are described as alternative lines of evolution, and their
likelihoods are assessed on the basis of present-day knowledge and the findings of
earlier assessments. In the definition of the performance targets and target properties, all
the lines of evolution and expected loads that are judged reasonably likely to occur
(based on current understanding and previous findings) are taken into account and,
40
hence, included in the design basis. Performance assessment is used to show that the
system, designed and built according to the design requirements and specifications, will
meet the performance targets and target properties and thus that the safety functions will
be fulfilled for all the reasonably likely lines of evolution. In this case there will be no
radionuclide releases within the one million year time frame. Performance assessment
also identifies any shortcomings that might occur in unlikely lines of evolution and
these are carried forward to safety assessment wherein it is assessed whether the
repository still will provide the protection level required by the regulations.
This is consistent with STUK YVL Guide D.5, which states:
“Targets based on high quality scientific knowledge and expert judgement shall be
specified for the performance of each safety function. In doing so, the potential changes
and events affecting the disposal conditions during each assessment period shall be
taken into account. In an assessment period extending up to several thousands of years,
one can assume that the bedrock of the site remains in its current state, taking however
account of the changes due to predictable processes, such as land uplift and those due
to excavations and disposed waste”. (STUK-YVL D.5, paragraph 408).
In addition, the definition of the performance targets takes into account the requirement:
“The safety approach for disposal of spent fuel shall be that the safety functions
provided by the engineered barriers will limit effectively the release of radioactive
substances into bedrock for at least 10 000 years.”
The derivation of the performance targets and target properties from the safety functions
is described in Design Basis. Table 2-2 and Table 2-3, respectively, list the performance
targets for the engineered components and target properties for the host rock, as
catalogued in VAHA Level 3. These are derived based on the safety functions and
expected evolution of the site.
Table 2-2. Performance targets for (a) the canister, (b) the buffer, (c) the deposition
tunnel backfill and plugs, and (d) closure. The performance targets, their rationale and
the related design requirements are discussed in detail in the Design Basis.
VAHA ID
Performance targets
a) Performance targets for the canister
L3-CAN-4
The canister shall initially be intact when leaving the encapsulation plant for disposal
except for incidental deviations.
L3-CAN-5
In the expected repository conditions the canister shall remain intact for hundreds of
thousands of years except for incidental deviations.
L3-CAN-7
The canister shall withstand corrosion in the expected repository conditions.
L3-CAN-9
The canister shall withstand the expected mechanical loads in the repository.
L3-CAN-11
The canister shall not impair the safety functions of other barriers.
L3-CAN-14
The canister shall be subcritical in all postulated operational and repository conditions
including intrusion of water through a damaged canister wall.
L3-CAN-16
The canisters shall be stored, transferred and emplaced in such a way that the copper
shell is not damaged.
L3-CAN-18
The design of the canister shall facilitate the retrievability of spent fuel assemblies from
the repository.
VAHA ID
b) Performance targets for the buffer
41
VAHA ID
Performance targets
L3-BUF-4
Unless otherwise stated, the buffer shall fulfill the requirements listed below over
hundreds of thousands of years in the expected repository conditions except for
incidental deviations.
L3-BUF-10
The buffer shall mitigate the impact of rock shear on the canister.
L3-BUF-8
The buffer shall limit microbial activity.
L3-BUF-12
The buffer shall be impermeable enough to limit the transport of radionuclides from the
canisters into the bedrock.
L3-BUF-13
The buffer shall be impermeable enough to limit the transport of corroding substances
from the rock onto the canister surface.
L3-BUF-14
The buffer shall limit the transport of radiocolloids to the rock.
L3-BUF-16
The buffer shall provide support to the deposition hole walls to mitigate potential effects
of rock damage.
L3-BUF-17
The buffer shall be able to keep the canister in the correct position (to prevent sinking
and tilting).
L3-BUF-6
The buffer shall transfer the heat from the canister efficiently enough to keep the buffer
temperature < 100oC.
L3-BUF-19
The buffer shall allow gases to pass through it without causing damage to the repository
system.
L3-BUF-21
The amount of substances in the buffer that could adversely affect the canister, backfill
or rock shall be limited.
VAHA ID
c) Performance targets for the deposition tunnel backfill and plugs
L3-BAC-5
Unless otherwise stated, the backfill and plugs shall fulfill the performance targets listed
below over hundreds of thousands of years in the expected repository conditions except
for incidental deviations.
L3-BAC-8
The backfill shall limit advective flow along the deposition tunnels.
L3-BAC-9
The plugs shall isolate the deposition tunnels hydraulically during the operational phase
of the repository.
L3-BAC-13
The chemical composition of the backfill and plugs shall not jeopardise the performance
of the buffer, canister or bedrock.
L3-BAC-16
The backfill shall keep the buffer in place.
L3-BAC-17
The backfill shall contribute to the mechanical stability of the deposition tunnels.
L3-BAC-18
The plugs shall keep the backfill in place during the operational phase.
L3-BAC-19
The backfill shall contribute to prevent uplifting of the canister in the deposition hole.
VAHA ID
d) Performance targets for closure
L3-CLO-13
Unless otherwise stated, the closure materials and structures shall fulfill the
performance targets listed below over hundreds of thousands of years in the expected
repository conditions except for incidental deviations.
L3-CLO-5
Closure shall complete the isolation of the spent nuclear fuel by reducing the likelihood
of unintentional human intrusion through the closed volumes.
L3-CLO-6
Closure shall restore the favourable, natural conditions of the bedrock as well as
possible.
L3-CLO-7
Closure shall prevent the formation of preferential flow paths and transport routes
between the ground surface and deposition tunnels/deposition holes.
L3-CLO-8
Closure shall not endanger the favourable conditions for the other parts of the EBS and
the host rock.
L3-CLO-11
Retrieval of the spent nuclear fuel canisters shall be technically feasible in spite of
repository tunnel and closure structures.
42
Table 2-3. Target properties for the host rock. The target properties and their rationale
are discussed in detail in the Design Basis.
VAHA ID
Target properties for the host rock
L3-ROC-3
Host rock shall, with the exception of incidental deviations, retain its favourable
properties over hundreds of thousands of years.
L3-ROC-5
The repository shall be located at minimum depth of 400 m.
L3-ROC-10
To avoid canister corrosion, groundwater at the repository level shall be anoxic except
during the initial period until the time when the oxygen entrapped in the near-field has
been consumed.
Therefore, no dissolved oxygen shall be present after the initially entrapped oxygen in
the near-field has been consumed.
L3-ROC-11
Groundwater at the repository level shall a have high enough pH and a low enough
chloride concentration to avoid chloride corrosion of the canisters.
Therefore, pH shall be higher than 4 and chloride concentration [Cl-] < 2M.
L3-ROC-12
Concentration of canister-corroding agents (HS-, NO2-, NO3- and NH4+, acetate) shall be
limited in the groundwater at the repository level.
L3-ROC-13
Groundwater at the repository level shall have low organic matter, H2 and Stot and
methane contents to limit microbial activity, especially that of sulphate reducing
bacteria.
L3-ROC-14
Groundwater at the repository level shall initially have sufficiently high ionic strength to
reduce the likelihood of chemical erosion of the buffer or backfill. Therefore, total charge
equivalent of cations Σq[Mq+]*, shall initially be higher than 4 mM.
* [Mq+] = molar concentration of cations, q = charge number of ion.
L3-ROC-15
Groundwater at the repository level shall have limited salinity so that the buffer and
backfill will maintain a high enough swelling pressure.
Therefore, in the future expected conditions the groundwater salinity (TDS, total
dissolved solids) at the repository level shall be less than 35 g/L TDS. During the initial
transient caused by the construction activities salinities up to 70 g/L TDS can be
accepted.
L3-ROC-16
The pH of the groundwater at the repository level shall be within a range where the
buffer and backfill remain stable (no montmorillonite dissolution).
Therefore, the pH shall be in the range of 5 −10, but initially a higher pH (up to 11) is
allowed locally. The acceptable level also depends on silica and calcium
concentrations.
L3-ROC-17
Concentration of solutes that can have a detrimental effect on the stability of buffer and
backfill (K+, Fetot) shall be limited in the groundwater at the repository level.
L3-ROC-29
Groundwater conditions shall be reducing in order to have a stable fuel matrix and low
solubility of the radionuclides.
L3-ROC-31
In the vicinity of the deposition holes, natural groundwater shall have a low colloid and
organic content to limit radionuclide transport.
L3-ROC-19
Under saturated conditions the groundwater flow in any fracture in the vicinity of a
deposition hole shall be low to limit mass transfer to and from EBS.
Therefore, the flow rate in such a fracture shall be in the order of one litre of flow per
one metre of intercepting fracture width in a year (L/(m*year)) at the most. In case of
more than one fracture, the sum of flow rates is applied.
L3-ROC-20
Flow conditions in the host rock shall contribute to high transport resistance.
Therefore, migration paths in the vicinity of the deposition hole, shall have a transport
resistance (WL/Q) higher than 10,000 years/m for most of the deposition holes and at
least a few thousand years/m.
L3-ROC-21
Inflow of groundwater to deposition tunnels shall be limited to ensure the performance
of the backfill.
L3-ROC-33
The properties of the host rock shall be favourable for matrix diffusion and sorption.
L3-ROC-23
The location of the deposition holes shall be selected so as to minimise the likelihood of
the rock shear movements large enough to break the canister.
43
VAHA ID
Target properties for the host rock
Therefore, the likelihood of a shear displacement exceeding 5 cm shall be low.
L3-ROC-30
To ascertain the data for sorption parameters, the pH shall be in the range of 6−10 after
the initial period when a higher pH of up to 11 is allowed.
2.2.3
Design requirements, rock suitability classification and design
specifications
Design requirements and design specifications
From the performance targets and target properties, the design requirements are derived.
Further, the repository system is defined (by specifications) such that the fulfilment of
the requirements can be verified at implementation.
The long-term performance targets and target properties are used to derive the design
requirements that the repository system must meet. For the engineered barriers these
define the requirements that the barriers must meet in order to withstand the future
expected loads. Design requirements form level 4 in VAHA.
The performance targets, target properties and design requirements of each EBS
component, and the underground openings and host rock are discussed in the Design
Basis.
Design specifications are the detailed specifications to be used in the design,
construction and manufacturing that have been derived from the more general design
requirements. They are defined so that the safety functions and performance targets are
achieved initially and will be fulfilled in the expected conditions during the time that the
spent nuclear fuel presents a significant hazard. Design specifications form level 5 in
VAHA.
The design specifications are discussed in the Canister-, Buffer-, Backfill, Closure- and
Underground Openings Production Line reports for each component and summarised in
Description of the Disposal System.
Rock suitability classification
For the rock barrier, the target properties set the starting point for the definition of the
Rock Suitability Classification system (RSC) developed by Posiva. The Classification
system includes both the updated rock suitability criteria as well the procedure for the
suitability classification during the construction of the repository (McEwen et al. 2013).
The RSC is used to identify suitable rock volumes for repository panels and to assess
the suitability of deposition tunnels for locating deposition holes and to accept
deposition holes for disposal. The aim is to avoid features of the host rock that may be
detrimental to favourable conditions for safety either initially or in long term. The target
properties presented in Table 2-3 outline the conditions that are considered to be
favourable.
The criteria developed for use in the classification system need to be based on
observable and measurable properties of the host rock. These rock suitability
44
classification criteria (RSC criteria) constrain the rock properties around the repository.
Based on interpretation, modelling and general understanding of the site properties, it is
shown that the target properties for the host rock (Table 2-3) are fulfilled when the RSC
criteria are met.
Classification of the host rock according to RSC is carried out at different scales,
including repository, panel, tunnel and deposition hole, and applied at different stages of
the repository design and construction, proceeding from the layout design of the whole
repository to the more detailed design and construction of panels, tunnels and, finally,
deposition holes. The aims associated with the classifications at the different scales are
as follows.
Classification at the repository scale aims to define the rock volumes to be used for
repository layout planning. Layout determining features (LDFs) are identified, as well
as their respect volumes, which are to be avoided when locating deposition tunnels and
holes. LDFs are either large fault zones that are potentially mechanically unstable in the
current or future stress field, or they are main groundwater flow routes important or
potentially important in the future for transport of solutes and chemical stability at the
site.

Classification at the panel scale aims to define suitable areas for the tunnels within a
specific panel and to assess the degree of utilisation10 of the panel area for the
detailed design of the panel. The panel consists of a central tunnel and a number of
deposition tunnels that will be excavated and used en bloc. The classification is done
based on the more detailed data on deformation zones and hydraulically conductive
zones that will become available during the construction of the central tunnels for
the panel.

The tunnel scale classification aims at defining suitable tunnel sections for the
deposition holes, so that the LDFs and smaller, local deformation zones and their
respect volumes, large fractures and high inflows to the deposition holes are
avoided.

At the deposition hole scale, the fulfilment of the rock suitability criteria is checked
as part of the acceptance procedure for the individual deposition hole.
The target properties and the rock suitability criteria are discussed further in Design
Basis and in the Rock Suitability Classification report (McEwen et al. 2013). McEwen
et al. (2013) also addresses the overall suitability and adequacy of the site as a natural
barrier (YVL Guide D.5, paragraph 406), including checking of properties that would
indicate unsuitability of the site, e.g. proximity of exploitable natural resources,
abnormally high rock stresses with regards to the strength of rock (YVL Guide D.5,
paragraph 410).
10
The degree of utilisation is determined by the number of suitable deposition holes with respect to the theoretical maximum
number and is related to whether the volume of rock is being used in an economical and effective manner.
45
2.3
Assessment methodology
2.3.1
Iterative approach
The design basis (Section 2.2) has been developed, and performance and safety
assessments (Section 2.3) have been carried out, in an iterative fashion.
Figure 2-5 outlines the approach to the development of the safety case, whereby the
design basis is developed, the performance of the repository system assessed, and
scenarios leading to radionuclide release are formulated and assessed. The design basis
and definition of performance targets and target properties are developed iteratively
between performance assessment, formulation and assessment of radionuclide release
scenarios and presentation of the safety case. Available scientific understanding,
including the results from earlier assessments, is used in the definition of the
performance targets, target properties for the host rock, design requirements and criteria
for rock classification. These will be updated as scientific understanding is further
developed, taking into account the results of the performance assessment and
assessment of radionuclide release scenarios of the current safety case (the two-way
arrows in Figure 2-5).
The potential future conditions that are taken into account in the design process are
described through a set of design basis scenarios. As required by regulation, the
likelihood of different scenarios is assessed and those that are judged reasonably likely
are included in the design basis scenarios. The performance targets and target
properties, together with the derived design requirements and the underlying design
basis scenarios, form the design basis of the repository. The design basis refers to the
current and future environmentally induced loads and interactions that are taken into
account in the design of the disposal system, and, ultimately, to the requirements that
the planned disposal system must fulfil in order to achieve the objectives set for safety.
A repository system designed and built according to the specified technical
requirements will be compliant with the regulatory safety requirements. The situations
in which the system does not fulfil the requirements, or there are significant
uncertainties, or the evolution in the future is not to the design basis scenarios, are taken
into account in the performance assessment and analysed in the safety assessment.
The formulation and assessment of scenarios leading to radionuclide release is
collectively termed safety assessment. In general, performance assessment and safety
assessment provide feedback and guidance to the system design concerning:

indications of the need for improved engineered solutions to increase robustness and
confidence in the safety case; and

specifications of the uncertainties and deviations that can be tolerated such that a
performance target/target property is still achieved.
The iteration between the design, performance assessment and safety assessment
ensures, as far as possible:

mutual compatibility of the engineered barriers with each other and with the
bedrock, taking into account their respective safety functions;
46

resistance of the engineered barriers to the main thermal, hydraulic, mechanical and
chemical loads to which they will be subjected during evolution of the system; and

robustness with respect to slow processes and unlikely events that may occur over
the regulatory compliance period, and

a safety case that properly takes into account uncertainties in the implementation of
the design (i.e. initial state uncertainties).
Figure 2-5 outlines the approach to the development of the safety case, whereby the
design basis is developed, the performance of the repository system is assessed, and
scenarios leading to radionuclide release are formulated and assessed.
Figure 2-5. Approach to the development of the safety case.
47
2.3.2
Description of the disposal system
Posiva defines the disposal system as comprising the repository system (spent nuclear
fuel, engineered barriers and host rock) and the surface environment (Figure 2-4).
An accurate and reliable description of the disposal system is the foundation for
development of understanding of possible lines of evolution, for assessments of
performance and safety, and for development of complementary considerations that
together comprise the safety case.
Further, the STUK Guide YVL D.5 (draft) states:
The preliminary and final safety analysis reports for a disposal facility shall include at
least …
 detailed description of the disposal site and description of its bedrock based on the
investigations made so far

description of the wastes to be disposed of, including their conditioning and
packaging method and any materials to be installed around the disposed waste
packages

description of the disposal facility (excavations, engineered structures, and systems)
and the way of its implementation (construction, operation and closure) …
Characterisation studies of the Olkiluoto site and bedrock have been made over a period
of 25 years. This has lead to a detailed description and understanding of the site in
respect of all characteristics relevant to its use for construction of a repository for spent
nuclear fuel and in relation to its long-term evolution. Studies of the surface
environment of the site form the basis for a description of the biosphere sufficient to
characterise the environment to be protected and its potential future use and occupation
by humans, plants and animals. Descriptions of the site and surface environment are
provided in Site Description and Biosphere Description.
The KBS-3 method and the KBS-3V design have been developed over more than 30
years. The specific realisation of the concept as planned for implementation of a
repository for spent nuclear fuel at the Olkiluoto site is the result of thorough analysis of
the functional requirements for the engineered barriers and host rock and of the overall
safety of the repository system (as described in Design Basis). Detailed descriptions of
the individual elements of the repository system and evidence concerning their practical
realisation and feasibility are presented in Production Line reports (Canister, Buffer,
Backfill, Closure and Underground Openings Production Line reports).
Studies of spent nuclear fuel and its characteristics relevant to disposal have been
carried out in many countries, and also in Finland with respect to the fuel types that will
arise from the Olkiluoto and Loviisa reactors.
The main characteristics and initial state, including uncertainties, of the repository
system components (spent nuclear fuel, EBS and host rock) and of the surface
environment to be used as an input to the safety assessment have been compiled in
Description of the Disposal System.
A summary description of the disposal system is given in Chapter 3 of this report.
48
2.3.3
Features, events and processes
Identifying and describing the features, events and processes (FEPs) that are relevant to
the evolution of the disposal system, or to its potential performance and safety, is an
essential step towards ensuring completeness of the assessments and the safety case.
For the TURVA-2012 safety case, the identification and description of FEPs has been
carried out by a team of scientific subject and assessment experts, based on a review of
the FEPs considered in previous assessments, the NEA FEP Database and FEPs
considered in safety cases in other nuclear waste programme as well as examination of
the specific characteristics of the Posiva disposal system and the Olkiluoto site.
Following identification and description of FEPs, a systematic screening process is
applied to rule out those FEPs that are not be relevant in the context of the repository
system as proposed at the Olkiluoto site.
The process took advantage of and drew on the experience from previous Posiva
studies, as well as from the development of FEP lists in support of the assessment of the
KBS-3V disposal method in Sweden.
A FEP database has been developed providing a structured classification of relevant
FEPs and couplings between them.
In Features, Events and Processes relevant FEPs for the safety case are presented
including a description of each FEP. Each description includes the current scientific
understanding and relevance in the context of the Posiva disposal system at the
Olkiluoto site, plus a note of any fundamental uncertainties in the scientific
understanding.
The FEP descriptions are organised according to the main components of the disposal
system: Spent nuclear fuel; Canister; Buffer; Backfill; Auxiliary components11;
Geosphere; Surface environment; along with relevant External features, events and
processes are discussed. For each component, FEPs affecting the physical state of the
disposal system (evolution-related FEPs) and FEPs that mostly affect the transport of
radionuclides (migration-related FEPs) are also distinguished.
The process of developing the FEP database and the complete set of descriptions,
references and other information is presented in Features, Events and Processes. The
process is summarised and a list of retained FEPs is given in Chapter 4 of this report.
2.3.4
Models and data and their use
Performance assessment and safety assessment require a range of models and input
data. In some cases the models may be relatively simple, e.g. to determine the rate at
which a single process may proceed for given conditions; in other cases, they may be
complex models of coupled processes, that may vary in space and time, and/or take
11
Auxiliary components refer to backfilling of central tunnels, service areas, access tunnel and shafts, and seals and plugs that are
installed both at the mouths of the deposition tunnels and as part of closure.
49
account of changing boundary conditions. In general, two types of model are
distinguished:

detailed models that aim at a realistic description of specific processes − sometimes
termed “process models”;

more simplified models used for scoping the impact of key processes and for
analysing radionuclide release, retention and transport in a cautious manner.
The first class of models are used to gain a ‘realistic’ understanding of the possible
evolution of aspects of the disposal system, e.g. the response of the groundwater flows
and salinity to excavation of the repository or impacts of a glacial episode on the
evolution of temperature and rock stress around the repository. These define the range
of THMC conditions and loads under which the repository components must maintain
their safety functions. More cautious, simple calculations may then suffice to show that
safety functions are preserved. For example, even for pessimistic assumptions of
groundwater composition and supply of oxidants, it is possible to use a simple
calculation (see Appendix B in Performance Assessment) to show that the rate of
corrosion of the copper canisters is so slow that canisters will not fail by corrosion of
the copper canisters is so slow that the number of canisters failing by corrosion within
the one million year assessment time frame remains limited.
Another example of detailed modelling is geochemical modelling to support the
evaluation of radionuclide speciation and retention parameters for use in assessments of
radionuclide release cases. Input from hydrogeological and geochemical modelling may
provide direct input to radionuclide release, retention and transport modelling, or, less
directly, support judgements regarding input parameter value selection. In the case of
biosphere modelling, detailed models of the evolution of the future landscape at the site
provide the template for defining future ecological conditions and human uses of the
environment that are taken into account in radionuclide transport modelling in the
biosphere and in the radiological impact modelling.
Radionuclide release and transport models, and also radiological impact models are
defined and used in accord with STUK’s Guide YVL D.5, which states:
“Simplifications of the models and the determination of the required input data shall be
based on the principle that the performance of a safety function will not be
overestimated while neither overly underestimated”. (A06)
and
“Selection of the computational methods, performance targets and input data shall be
based on the principle that the actual radiation exposures or quantities of released
radioactive substances shall with high degree of certainty be lower than those obtained
through safety analyses. The uncertainties included in the safety analysis shall be
assessed by means of appropriate methods, e.g. by sensitivity analyses or probabilistic
methods”. (A08)
Model simplifications are applied where processes or data are uncertain, but must
always be implemented in such a way that they can be seen to be cautious, e.g. by
50
reference to alternative models, by omission of processes that can only be beneficial to
performance or safety, or by selecting input data such that impacts cannot be underestimated. Such simplifications may reduce data requirements, which is advantageous as
it allows the effort of data gathering and quality assurance to focus on key parameters.
Key input data for the performance assessment is the initial state of the repository
system and the barriers provided by the Production Line reports summarised in
Description of the Disposal System. Other key data reports are the canister design report
(Raiko 2012), thermal dimensioning report (Ikonen & Raiko 2013) as well as the design
of the disposal facility report (Saanio et al. 2013) presenting the repository layout. Site
Description provides the description of the bedrock and the groundwater system, and
the interacting processes and mechanisms. Biosphere Description provides the
description of the present state and evolution of the surface environment. Several
studies based on testing and modelling have been carried out to describe the evolution
of the disposal system − the evolving conditions at the disposal site and performance of
the EBS under different conditions, e.g. groundwater flow, terrain and ecosystem
development, buffer and backfill saturation, and erosion and canister corrosion. A
summary of the results of these studies for the repository system is given in
Performance Assessment and for the biosphere in Biosphere Assessment. These studies
form the basis for the selection of the parameter values to be used in the safety
assessment.
For the repository system, the models together with the data forming the basis for the
selection of the specific parameter values are described in Models and Data for the
Repository System. As to the surface environment, the data used in the biosphere
assessment are summarised in Biosphere Data Basis, and the models are discussed in
Terrain and Ecosystems Development Modelling, Surface and Near-Surface
Hydrological Modelling, Biosphere Radionuclide Transport and Dose Assessment and
Dose Assessment for Plants and Animals.
An overview of models and data is provided in Chapter 5 in this report.
2.3.5
Assessment of performance of the repository system under the most
likely lines of evolution
Posiva’s safety concept is based on long-term isolation and containment, which is
achieved through robust engineered barrier system design and favourable geological
conditions at the repository site, as discussed in Section 2.1.
The Government Decree 736/2008 (Section 11) states:
“The long-term safety of disposal shall be based on safety functions achieved through
mutually complementary barriers so that a deficiency of an individual safety function or
a predictable geological change will not jeopardise the long-term safety.
Safety functions shall effectively prevent releases of disposed radioactive materials into
the bedrock for a certain period, the length of which depends on the duration of the
radioactivity in waste. For short-lived waste, this period shall be at least several
hundreds of years, and for long-lived waste, at least several thousands of years.”
51
The Guide YVL D.5 (draft 4, 17.3.2011 in Finnish only) further advises:
“Performance targets for the safety functions of engineered barriers shall be specified
taking account of the activity level of waste and the half-lives of dominating
radionuclides. The safety approach for disposal of spent fuel shall be that the safety
functions provided by the engineered barriers will limit effectively the release of
radioactive substances into bedrock for at least 10 000 years.” (Para 408)
The aim of performance assessment is to present the evidence that the safety functions
(as set out in Table 2-1) will be fulfilled, which being so, lead to isolation of the spent
nuclear fuel, and complete containment over hundreds of thousands of years.
Performance is assessed for the design basis scenarios, which define the envelope of
future conditions taking into account the reasonably expected lines of evolution. These
are also taken into account in the definition of the performance targets and target
properties for the repository components.
The performance of the repository system is analysed and the fulfilment of the
performance targets and target properties (as set out in Table 2-2 and Table 2-3) is
evaluated taking into account the expected thermal, hydraulic, mechanical and chemical
(THMC) evolution of the repository system and the uncertainties in the expected lines
of evolution. The possibilities for occurrence of less expected or disruptive events and
processes, that could lead to reduction of one or more safety functions and, potentially,
give rise to radionuclide releases, are also identified. Account is taken of the natural
evolution of the environment, chiefly driven by climatic evolution, which imposes
external loads on the repository system, and also internal loads, chiefly from the effects
of excavation and emplacement of the spent nuclear fuel and the engineered barriers.
Performance Assessment covers the performance of the system for the entire assessment
time frame of one million years with a special focus on the containment safety function
of the canister and isolating safety function of other EBS components and the geosphere
in the first 10,000 years (as required by YVL D.5 paragraph 408). The performance is
considered in three time windows: (1) during the excavation and operational period up
to closure; (2) up until 10,000 years after closure; (3) beyond 10,000 years over repeated
glacial cycles. The fulfilment of performance targets and target properties in each time
window is assessed considering time-dependent and space-dependent loads on the
engineered barriers and host rock.
The assessment first considers the initial state of the repository system as defined in the
production line reports (including deviations and uncertainties) constrained by the
requirements and implemented according to design specifications. The performance
targets and target properties (as defined in VAHA L3) are checked against evolutionrelated FEPs to ensure that the relevant processes and factors that could pose a threat to
the performance of the barriers have been identified. The performance of the repository
system and analyses of the response of the barriers to the evolution is then considered.
The discussion is based on existing data and knowledge available in published reports
and literature. Whenever possible, quantitative arguments are used, for example to
provide estimates of safety margins and evidence of robustness of design.
52
Performance in each time window is summarised by a discussion on the state of the
barriers with respect to the performance targets and target properties at the end of each
of the periods, uncertainties are highlighted and the likelihood and effects of the
deviations estimated. Special attention is given to identifying conditions and events
(incidental deviations) that could lead to the release of radionuclides; these are taken
forward to the formulation of radionuclide release scenarios, see Section 2.3.6.
The process is fully documented in Performance Assessment. A summary description of
the main findings is given in Chapter 6 of this report.
2.3.6
Scenario formulation
Uncertainties in the evolution and performance of the repository system and the surface
environment mean that, in safety assessment, a range of scenarios, each representing
one or more possible time histories of conditions, or “lines of evolution”, must be
formulated and analysed.
Consistent with the regulatory and international guidance (Guide YVL D.5; and IAEA
2009, 2011, 2012), Posiva distinguishes between the expected evolution of the disposal
system and unlikely events and processes; account is also taken of the time window (or
windows) in which releases of radionuclides might occur.
Guide YVL D.5 states:
“Compliance with the requirements concerning long-term radiation safety, and the
suitability of the disposal method and disposal site, shall be proven through a safety
case that must analyse both expected evolution scenarios and unlikely events impairing
long-term safety.”
The Guide goes on to define three types of scenarios:

Base scenario: The base scenario shall assume the performance targets for each
safety function, taking account of incidental deviations from the target values.

Variant scenarios: The influence of declined performance of a single safety function
or, in case of coupling between safety functions, the combined effects of declined
performance of more than one function shall be analysed by means of variant
scenarios.

Disturbance scenarios: Disturbance scenarios shall be constructed for the analysis of
unlikely events impairing long-term safety.
The classification of scenarios is illustrated in Figure 2-6.
The repository system is designed in a way that, for the design basis scenarios, except
for incidental deviations, each component of the EBS meets the performance targets,
assigned to it, and the host rock conforms to its target properties. In this case, the
copper-iron canisters remain intact for the whole assessment time frame and there is no
release of radionuclides. This is confirmed by the performance assessment.
The performance assessment shows, however, that there are some plausible conditions
and events (incidental deviations) that could lead to reduction of one or more safety
53
Figure 2-6. Classification of scenarios in TURVA-2012, which is consistent with
STUK’s Guide YVL D.5.
functions, and thus may give rise to radionuclide releases. In addition, there are some
very unlikely events and processes that could disrupt the repository, e.g. related to
human intrusion and rock shear. These incidental deviations and unlikely events are
systematically examined to define a set of scenarios that encompass the important
combinations of initial conditions, natural evolution and disruptive events.
In the current and past assessments by Posiva, the scenario of a canister with an initial
penetrating defect has been considered in order to test the radiological performance of
the other engineered barriers and host rock. This defect is most likely in the weld.
Although the likelihood that a canister with an initial undetected penetrating defect will
be emplaced is low, this is a useful base scenario for safety assessment (radionuclide
release calculations) against which the efficiency of the other technical barriers and the
host rock to limit the radionuclide releases can be tested and that also complies with the
GD 736/2008.
Thus, as indicated in Figure 2-6, the base scenario addresses the most likely lines of
evolution (in which the performance targets and safety functions are met), but takes into
account the possibility of one or a few canisters with initial undetected penetrating
defects. Emplacement of a canister with an initial penetrating defect is not expected, but
is an incidental deviation that cannot be ruled out. The variant scenarios address
situations that are considered reasonably likely and in which there may be reduced
performance of one or more safety functions of the barriers. Disturbance scenarios
address the lines of evolution that are considered unlikely but cannot be completely
eliminated.
The process is documented in Formulation of Radionuclide Release Scenarios. A
summary description of the scenarios is given in Chapter 7 of this report.
2.3.7
Approach to the analysis of radionuclide releases, transport and
radiological impact
The approach to analysis of radionuclide releases, transport and raiological impact, and
in particular the endpoints to be calculated, are specified in legal and regulatory
requirements. As discussed in Section 1.5, the Government Decree on the safety of
disposal of nuclear waste (GD 736/2008) and the STUK Guide YVL D.5 set out the
54
requirements relating to radiological protection and provide guidance on the analysis of
scenarios.
Two “time windows” are distinguished in these documents, in which different
protection criteria apply. For the earlier time window, which shall extend at a minimum
over several millennia, Section 4 of GD 736/2008 states:
“The annual dose to the most exposed people shall remain below the value of 0.1 mSv”,
and that
“The average annual doses to people shall remain insignificantly low”.
In the longer term, the quantitative regulatory criteria relate directly to radioactive
releases to the environment. Guide YVL D.5 states:
“The sum of the ratios between the nuclide specific activity releases and the respective
constraints shall be less than one” and
“These activity releases can be averaged over 1000 years at the most”.
The nuclide-specific constraints referred to above are set out in Guide YVL D.5,
paragraph 312. They form the basis for a sum-of-fractions approach to limiting
radiological impact in the biosphere taking account of the relative potential for
radiological impact of different groups of radionuclides.
The probability of unlikely events giving rise to radionuclide releases may be taken into
account when assessing compliance. Guide YVL D.5 states:
“The importance to safety of such unlikely events shall be assessed and whenever
practicable, the resulting annual radiation dose or activity release shall be calculated
and multiplied by its expected probability of occurrence. The obtained expectation value
shall be below the radiation dose constraint … or activity release constraint …”.
Requirements for protection of species of fauna and flora are also set out, thus:
“Disposal shall not affect detrimentally to species of fauna and flora. This shall be
demonstrated by assessing the typical radiation exposures of terrestrial and aquatic
populations in the disposal site environment, assuming the present kind of living populations. The assessed exposures shall remain clearly below the levels which, on the basis of the best available scientific knowledge, would cause decline in biodiversity or
other significant detriment to any living population.” (Para 316)
Consistent with regulatory guidance, the main safety indicators calculated and assessed
in TURVA-2012 are:

The radioactive releases from the bedrock to the biosphere (surface environment),
which are calculated for all release scenarios and assessed against the nuclidespecific constraints for radioactive releases to the environment (average annual
release rates of radioactive substances) defined in YVL D.5 (see Table 2-4);

Annual doses to humans. Consistent with regulatory guidance these are calculated
for scenarios that give rise to releases to the surface environment in the first 10,000
years. They are assessed against the requirements that:
55


“The annual dose to the most exposed people shall remain below the value of
0.1 mSv”

and “The average annual doses to people shall remain insignificantly low”.
Absorbed dose rates to plants and animals. These are also only calculated for
releases to the surface environment in the first 10,000 years. They are assessed
against the requirement that the assessed exposures shall remain clearly below the
levels that would cause decline in biodiversity or other significant detriment to any
living population.
The repository system is analysed using models that represent:



release from the spent nuclear fuel (taking account of the locations of radionuclides
in the fuel, its cladding and other parts of the fuel element);
release, retention and transport in the near field (release from the canister, migration
through the buffer, migration by alternative routes to water-conducting fractures in
the host rock); and
retention and transport in the geosphere (through water-conducting fractures taking
account of variability in flow paths).
This yields radioactive releases from the geosphere to the biosphere, which are used as
input to biosphere models. Since the repository system models are run independently of,
the biosphere models, the output from a single repository system calculation can be
input to alternative biosphere models so as to represent alternative surface environment
conditions at the time of release.
In addition, complementary indicators are evaluated to broaden the understanding of the
repository system performance with respect to retention and release of radionuclides.
These indicators include total amounts of activity and activity concentrations in model
compartments, activity fluxes between compartments and delay times (see Chapter 9).
Table 2-4. Radionuclide-specific constraints for radioactive releases to the
environment, as set out in STUK Guide YVL D.5. In a given row, the constraint applies
to each individual radionuclide. The sum of the ratios between the nuclide specific
activity releases and the respective constraints shall be less than one. These activity
releases can be averaged over 1000 years at the most. The probability of unlikely events
giving rise to activity releases may be taken into account.
*
Radionuclides
Constraints
[GBq/a ]
Long-lived alpha-emitting Ra, Th, Pa, Pu, Am and Cm isotopes
0.03
Se-79; Nb-94; I-129; Np-237
0.1
C-14; Cl-36; Cs-135; long-lived uranium isotopes
0.3
Sn-126
1
Tc-99; Mo-93*
3
Zr-93
10
Ni-59
30
Pd-107
100
Mo-93 is not mentioned in YVL D.5. However, based on a preliminary evaluation, STUK has
recommended that the same nuclide-specific constraint as for Tc-99 is used.
56
Modelling for biosphere assessment includes, first, a screening process to identify those
radionuclides that could make significant contributions to total radiological impacts.
These radionuclides are carried forward to detailed biosphere modelling, based on a
model of the future landscape and ecosystem development in the Olkiluoto region over
the next 10,000 years. This provides the framework for modelling of radionuclide
movements within compartments of the future surface environment and calculation of
the radiation doses to humans, and to plants and animals, inhabiting or making use of
the various areas and resources that may become contaminated.
In TURVA-2012, to assess the performance of the repository system, the quantity
presented is the normalised activity release rate of a given radionuclide, or the sum of
normalised activity release rates. The normalised activity release rate is a dimensionless
quantity defined as the activity release rate divided by the respective radionuclidespecific constraint, as given by Table 2-4. To satisfy the regulatory constraint on the
release rates of activity from the geosphere to the biosphere (geo-bio fluxes), the
normalised activity release rate, summed over all radionuclides, must be less than one.
2.3.8
Treatment of uncertainty
Guide YVL D.5 (draft) requires:
The uncertainties included in the safety analysis shall be assessed by means of
appropriate methods, e.g., sensitivity analysis or probabilistic methods. The safety case
shall include an assessment of the confidence level with regard to compliance with the
safety criteria and of uncertainties with most contribution to the confidence level.
Consistent with international best practice in safety assessments, uncertainties are
analysed by a number of complementary methods, which include consideration of a
range of calculation cases representing alternative future evolutions of the disposal
system and the potential occurrence of unlikely events, alternative models of key
processes, and uncertainties in data values.
The derivation of scenarios that represent alternative future evolutions of the disposal
system and the occurrence of unlikely events has been discussed in Section 2.3.6. The
remainder of this section considers the development of calculation cases to represent
alternative scenarios, models and data.
Calculation cases
The radionuclide release scenarios (Section 2.3.6) form the framework for the definition
of calculation cases that explore the uncertainties within each scenario.
Calculation cases are defined to evaluate compliance of the repository with regulatory
requirements on radiological protection, as well as to illustrate the impact of specific
uncertainties or combinations of uncertainties on the calculated results. Each case
illustrates different possibilities for how the repository might evolve and perform over
time, taking into account uncertainties in the models and parameters used to represent
radionuclide release, retention and transport, and radiological impact.
While all uncertainties and combinations of uncertainties need to be considered in
formulating the calculation cases, some combinations of uncertain model assumptions
57
and parameter values can be excluded on the grounds that they represent very unlikely
or implausible outcomes. Thus a reduced set of calculation cases can be carried forward
that, nevertheless, spans the domain of plausible outcomes. Uncertain model
assumptions or parameter values can also be excluded if they can be argued, or shown,
to yield lower consequences than existing calculation cases incorporating more
cautiously chosen model assumptions or parameter values.
Four types of calculation cases are distinguished:

A Reference Case is one model realisation of the base scenario. Models and data for
the Reference Case are, in most instances, selected to be either realistic or
moderately cautious, i.e. radiological impacts are not to be underestimated nor
excessively overestimated.

Sensitivity cases represent alternate models and/or data to those of the Reference
Case, but that remain within the scope of the base scenario and/or variant
scenarios. Analyses of the sensitivity cases illustrate the effect of model and data
uncertainties.

What-if cases are mainly model representations of disturbance scenarios. Models
and data for these what-if cases are selected to represent unlikely events and
processes.

Complementary cases are designed to develop a better understanding of the
modelled system or subsystems.
The number of calculation cases requiring biosphere modelling is reduced by
considerations of conservatism and of likelihood, as discussed above. In particular,
combinations of highly conservative/unlikely cases of the repository system model with
highly conservative/what-if cases of the biosphere are avoided.
All of the above cases are analysed deterministically, i.e. each calculation is carried out
for a defined set of input parameter values. In addition, the disposal system behaviour is
explored by Monte Carlo simulations and probabilistic sensitivity analysis (see below).
Probabilistic sensitivity analysis (PSA)
Many of the parameters used in the radionuclide release and transport calculations are
affected by significant uncertainties, due to spatial variability, time evolution of the
environmental conditions and statistical and systematic uncertainties in the parameter
values. It is not practical, using individual, deterministically specified calculation cases,
to explore the consequences of every possible combination of parameter values.
Probabilistic sensitivity analysis, which complements the deterministic evaluation of
calculation cases, is used to overcome this problem.
The probabilistic sensitivity analyses (PSAs) carried out for TURVA-2012 were based
on the technique of Monte Carlo simulation. In Monte Carlo simulation, performance
measures, such as radionuclide release rates from the geosphere to the biosphere, are
calculated a large number of times. Each calculation represents a different realisation of
the modelled system, in which all uncertain input parameters are sampled randomly
from probability density functions (PDFs). This means that single random values are
selected from specified PDFs describing each parameter. As a result of this random
58
sampling from PDFs, each realisation is considered to be equally probable. For each
realisation, all uncertain input parameters are sampled, meaning that a single random
value is selected from a specified distribution describing each parameter (probability
density function, or PDF). The results of the multiple realisations are assembled into
probability distributions of possible outcomes. A statistical analysis is then carried out
to identify, for example, the input parameters, or combinations of parameters (lumped
parameters), for which variance in the input most affects the variance in the calculated
performance measures. Different statistical analysis methods can be used. If the
outcome is a similar ranking of important parameters, this shows the robustness of the
sensitivity analysis.
In TURVA-2012, the PSA has been carried out for the base scenario of the repository
system, in which it is assumed that a canister with an initial penetrating defect is
emplaced in the repository. The defective canister may be present at any location in the
repository.
2.3.9
Complementary considerations and supporting evidence
Complementary considerations provide additional evidence for the long-term safety of
disposal according to the KBS-3 method at the Olkiluoto site.
STUK’s Guide YVL D.5 states:
“The importance to safety of such scenarios that cannot reasonably be assessed by
means of quantitative safety analyses, shall be examined by means of complementary
considerations. They may include e.g. analyses by simplified methods, comparisons with
natural analogues or observations of the geological history of the disposal site. The
significance of such considerations grows as the assessment period increases, and
safety evaluations extending beyond a time horizon of one million years can mainly be
based on the complementary considerations. Complementary considerations shall also
be applied parallel to the actual safety assessment in order to enhance the confidence in
results of the analysis or certain part of it.” (A09)
Complementary considerations and supporting evidence in support of the TURVA-2012
safety case have been assembled related to:

the choice of geological disposal as a concept for disposal of radioactive waste,
which is a choice backed by international accord;

support for the robustness of the KBS-3 method, which makes use of a few simple
materials with well-known properties established by engineering practice and by
evidence from natural analogues;

the favourable features of the Olkiluoto site, which include the stable tectonic
situation, the presence of suitable volumes of good quality rock appropriate for
repository construction, and low groundwater flows, reducing conditions and
otherwise favourable groundwater conditions at repository depth.
Arguments and evidence related to each of these points is compiled in Complementary
Considerations; the arguments and evidence are summarised in Chapter 9 of this report.
In addition, Chapter 9 summarises results from calculations of alternative indicators that
59
provide additional perspective on the level of protection provided by Posiva’s KBS-3V
repository system.
2.4
Uncertainty management
Posiva has developed a systematic approach to the management of uncertainties in the
safety case based on an iterative process where research, development, technical design
and assessment results play an essential role (Figure 2-7).
The overall strategy can be summarised as: identify, avoid, reduce and assess.
Uncertainties need to be identified, i.e. described and quantified, and their relevance to
safety needs to be considered. The safety case is largely composed of identification of
uncertainties and assessment of their relevance. Uncertainties in the theoretical and
conceptual understanding of FEPs are identified in Features, Events and Processes. The
uncertainties in the model and data that describe the FEPs are reported in Models and
Data for the Repository System and Biosphere Data Basis. Discussion of the
representation and impact of uncertainties are discussed in Performance Assessment,
Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere
Assessment, with respect to the uncertainties in each of these assessments.
The development of the repository system is based on the idea of robustness, which
means, where practicable, avoiding concepts and components the behaviour of which
would be difficult to understand and predict. By means of research, it is possible to gain
new knowledge and understanding of the system behaviour and, thereby, to reduce the
uncertainties. The impact of uncertainties can also be reduced by technical design, for
example by introducing safety margins. A practical example of reducing uncertainties is
the development of a Rock Suitability Classification system (RSC) that can be used to
reduce the uncertainties in conditions to which canisters are exposed. A robust disposal
system is usually also based on a design that is fairly simple and that works in a
transparent and predictable fashion.
Some uncertainties will always remain and have to be assessed in terms of their
relevance to the final conclusions on safety. The treatment of uncertainty in TURVA2012 assessments is outlined in Section 2.3.8 and is further illustrated in Chapters 6, 7
and 8. The combination of deterministic and probabilistic approaches allows the impact
of uncertainties on system performance and safety to be determined individually and
jointly. This allows uncertainties that could potentially weaken the safety case to be
identified, and avoided or reduced by further research and technical development. One
of the outputs of TURVA-2012 is the identification of key safety related issues to be
addressed in future RTD studies (Section 10.2).
60
Figure 2-7. Iterative approach to the management of uncertainties.
2.5
Quality management
2.5.1
Goals and principles
Posiva applies a quality management system that complies with the ISO 9001:2008
standard for all activities including the production of the safety case reports, and
requires the pursuit of the same quality assurance principles from all its contractors and
suppliers. The system was first launched in 1997 and has since been subject to
continuous maintenance, updating and several internal and external audits.
The purpose of Posiva’s management system is to ensure, in a documented and
traceable way, that Posiva's products – whether in the form of abstract knowledge and
information, published reports or physical objects – fulfil the requirements set for them.
61
The general quality objectives, requirements and instructions defined in Posiva’s
management system will also form the foundation for the quality management of safety
case activities carried out in the future. However, special attention is paid to the
management of the processes that are applied to produce the safety case and its basis.
The purpose of this enhanced process control is to offer full traceability and
transparency of the data, assumptions, modelling and calculations.
While the approach is based on the ISO 9001:2008-standard, which means management
through processes, the principle of a graded approach, as proposed in the safety guides
for nuclear facilities, is pursued in safety case production. The graded approach means
that the primary emphasis in the quality control and assurance of safety case activities is
placed on those parts of the assessment that have a direct bearing on the arguments and
conclusions on the long-term safety of disposal, whereas standard quality measures are
applied in the supporting work.
2.5.2
Application to TURVA-2012 safety case production
The overall plan, main goals and constraints for the TURVA-2012 safety case
production process are presented in the Safety Case Plan (Posiva 2008). The details of
how the Safety Case Plan 2008 is being implemented are described in the SAFCA
project plan. The work is managed and coordinated by a SAFCA core group and
supervised by a steering group.
A SAFCA quality co-ordinator (QC) has been designated for the activities related to the
quality assurance measures applied to the production of the safety case contents. The
QC is responsible for checking that the instructions and guidelines are followed and
improvements are made in the process as deemed useful or necessary. The QC is also
responsible for coordination of the external expert reviews, maintenance of schedules,
review and approval of products, and management of the expert elicitation process. The
QC also leads the quality review of models and data used in the Data Handling and
Modelling subprocess. Regular auditing of the safety case production process is done as
part of Posiva’s internal audit programme.
The production of the safety case is divided into four main subprocesses:
Conceptualisation and Methodology, Data Handling and Modelling, Safety Assessment,
and Evaluation of Compliance and Confidence.

The Conceptualisation and Methodology subprocess frames the assessment
providing the description of the disposal system, FEP analysis and the formulation
of scenarios, including system evolution. It guides the definition of the performance
targets for the EBS and the bedrock, which form the core of the requirements
management system (VAHA). An approach to evaluating the suitability of the rock
at various scales has been developed through application of the rock suitability
classification (RSC) system.

The Data Handling and Modelling subprocess identifies the lines of evolution that
could lead to the release of radionuclides and formulates the scenarios that are
analysed first to quantify the releases from the repository system to the surface
environment and then to quantify the radiological impact of those releases to
humans, plants and animals.
62

The Safety Assessment subprocess identifies the lines of evolution that could lead to
the release of radionuclides and formulates the scenarios that are analysed first to
quantify the releases from the repository system to the surface environment and then
to quantify the radiological impact of those releases to humans, plants and animals.

The Evaluation of Compliance and Confidence subprocess is responsible for the
final evaluation of compliance of the assessment results with the regulatory criteria
and the overall confidence in the safety case, taking into account the completeness
of the scenarios considered, uncertainties within the assessment and complementary
considerations regarding the long-term safety of geological disposal.
It is essential that the information and data passed between subprocesses is quality
assured. Models and Data for the Repository System and Biosphere Data Basis and
Biosphere Assessment act as the quality assured interface between the safety case
activities and the research and technical activities: they include all the essential EBS and
site information and data needed for the performance and safety assessment
calculations, while more details can be found in the supporting background reports, such
as Site Description and various Production Lines reports. The quality of Site
Description is mainly ensured by the application of scientific principles, while the
methods of quality control for the design and implementation depend on the nature of
the materials and technology in question.
2.5.3
Model qualification and code verification
A range of quality control and assurance measures has been adopted to promote
confidence in the models and codes and hence to promote confidence in the analysis of
the calculation cases. According to Posiva’s Safety Case Plan (Posiva 2008), the quality
control and assurance measures comprise:
1. validation of input data for the scenarios and models considered; the limits of
applicability of the input data are checked against the assumptions related to the
scenarios and models,
2. validation of the models used to analyse the scenarios,
3. verification of assessment codes,
4. validation of the assessment codes for the intended application,
5. documentation of input for the production runs,
6. application of a procedure to ensure codes are correctly applied,
7. documentation of the code versions used, and
8. reporting of non-conformities.
Measures 1 and 2 relate to the quality of models and to the selection and checking of
data that are implemented in the codes. Actions undertaken to validate and promote
confidence in the models and data used in TURVA-2012 are described in Models and
Data for the Repository System and for the surface environment in Terrain and
Ecosystems Development Modelling, Surface and Near-Surface Hydrological
Modelling, Biosphere Radionuclide Transport and Dose Assessment and Dose
Assessment for Plants and Animals.
63
At a more general level, Features, Events and Processes and Complementary
Considerations summarise the understanding of processes relevant to repository
performance and safety that can be gained from observations at the site, including its
regional geological environment, and from natural and anthropogenic analogues for the
repository and its components.
Measures 3 to 8 relate to the selection, testing and application of computer codes used
for the radionuclide release and transport calculations as well as dose assessment and to
the documentation of results. Actions undertaken to verify and promote confidence in
the computer codes and their application are described in Assessment of Radionuclide
Release Scenarios for the Repository System and Biosphere Assessment.
Verification measures, including benchmarking exercises that address specific functions
of GoldSim and MARFA, have been carried out during the development of these codes.
In addition, benchmarking exercises have been carried out in which results generated by
these codes were compared with those generated by REPCOM and FTRANS, which are
the codes that were used in previous Posiva safety analyses and have been shown to
handle the main features, events and processes of relevance. The benchmarking
exercises used test cases that are representative of the types of calculations for which
GoldSim and MARFA are used in the TURVA-2012, and so contribute to validation as
well as verification. Finally, an external review of MARFA has been carried out and
deficiencies identified in the review were addressed before the calculations for
TURVA-2012 were undertaken. Based on all these measures, it is concluded that
GoldSim and MARFA have been verified and validated to the extent required for use in
TURVA-2012. Regarding code application, numerical parameters, such as the size of
the time steps used by GoldSim and the number of particles calculated by MARFA, are
carefully selected to ensure that the model results are sufficiently accurate. A version
management system (VMS) has been used to keep track of any changes in input files
and thus maintain the reproducibility of calculation results. An assessment database has
been set up for the storage, checking and exchange of input data, intermediate results
and final results. Finally, an electronic system docgen12 has been developed to keep
track of, and to archive, the results of safety assessment calculations as they are
produced. Results of model calculations and their associated input files are downloaded
to docgen automatically from the assessment database. In this way, it has been possible
to follow the progress of the calculations and carry out quality assurance and
plausibility checks in a timely manner.
2.5.4
Data clearance
A wide variety of data have been used for the compilation of the safety case. An
important activity for ensuring the quality, transparency and consistency of the data
used in the safety case is data clearance. Data clearance is the formal procedure to
approve the data to be used as input to the models used in the analyses reported in the
safety case, such as the assessment of the performance of the repository system, the
analysis of the release scenarios and the analysis of radiological consequences.
12 The docgen system was originally developed for Nagra, the Swiss National Cooperative for the Disposal of Radioactive Waste.
The version used in TURVA-2012 has been extended and adapted for Posiva.
64
The data produced, e.g. by site investigations or laboratory tests, are usually not directly
suitable for the models used in the safety case, and further interpretation and modelling
are needed. Sometimes there are no site-specific data available, thus literature data and
data from other sources, e.g. data produced by other nuclear waste management
organisations, need to be used. The applicability of the data for the specific purpose and
conditions analysed in the safety case is assessed as part of the data clearance process,
and potential sensitivity cases to be addressed by modelling are suggested. The data
may be in the form of single parameter values, a range of parameter values or a
probability distribution function. In some cases, different data apply to specific model
variants or versions (e.g. applying to a specific hydrogeological model or repository
layout).
The data clearance procedure consists of the following steps: (i) identification of the
data needs, (ii) collection of suitable data, (iii) documentation of the suggested data,
their intended use and justification for their selection and (iv) data approval. Separate
reports on various categories of data collection have been produced, e.g. regarding
climate evolution, solubility and sorption data for the near field and far field, and
earthquake frequency. Further review of the data by subject matter experts and safety
analysts has been carried out and, in some specific cases a formal expert elicitation has
been applied. Purpose-specific databases have been applied to manage the data
clearance procedure in a structured way and to ensure the controlled use and traceability
of input data used as input to safety related assessment calculations.
The expert elicitation process has been applied to a specific case (solubility and sorption
data) to identify the main sources of uncertainty and determine whether different views
may have to be propagated through the safety assessment. This expert elicitation
process has been initiated, recruited, documented and managed by the SAFCA Quality
Co-ordinator.
The clearance procedure is documented in Models and Data for the Repository System,
Biosphere Data Basis and Biosphere Assessment. These reports give an overview of the
modelling carried out within the safety case and how the different models link to each
other. They also present the key models and data used in the safety case. For each
model, the conceptual model, the numerical model and the codes used are described.
This description covers the key assumptions and simplification, e.g. omission of certain
processes. The uncertainties related to modelling and their impact on the results is
presented. Also, possible alternative models are discussed and the basis for selection of
the specific model is given. Discussion of the data describes the production,
qualification and uncertainties related to the data as well as potential alternative data.
In order to assess confidence in the models and data, the applicability of the models and
data to the conditions at Olkiluoto and to the safety case purposes as well as the applied
quality measures are discussed. Further, the impact of the uncertainties in the models
and data on the modelling outcome is assessed and needs for model and data
improvements are identified, if necessary.
The clearance process has been implemented according to guidelines that address the
documentation of data sources and quality aspects. Single items of data and databases
are approved through a clearance procedure supervised by the SAFCA Quality Co-
65
ordinator. Process owners check and approve the data and while the Quality Coordinator checks and approves the procedure. Data used may also be approved using
other Posiva databases like VAHA or POTTI and the respective approval processes. A
clearance procedure has been applied to all key data used in the performance assessment
(i.e. showing compliance with performance targets and target properties), and in safety
assessment (i.e. radionuclide transport analysis and dose calculations).
2.5.5
Report and product review and approval process
The review and approval of the safety case products (i.e. main portfolio reports) has
been done in a fully traceable manner. This has included, first, an internal review by
safety case experts and subject-matter experts within Posiva’s RTD programme and,
second, a review by external experts. A group of external experts covering the essential
areas of knowledge and expertise needed in safety case production has been set up. The
review comments are managed via review templates, which record the review
comments and how each comment has been addressed. Upon completion, this template
is checked and approved according to the quality guidelines of Posiva.
Quality assurance and quality control measures related to the production and operation
of the repository are discussed in detail in Production Line reports (Canister, Buffer,
Backfill, Closure and Underground Openings Production Line reports).
66
67
3
DESCRIPTION OF THE DISPOSAL SYSTEM
This chapter presents a summary description of the disposal system in its initial state;
that is, descriptions of the host rock and surface environment, and of the spent nuclear
fuel and engineered barriers (canister, buffer, backfill and closure).
The disposal system (Figure 2-4) is composed of:

the spent nuclear fuel, which is the source of hazard to be isolated and contained;

the barriers, which are the engineered barriers that are designed, and host rock that
is chosen, to isolate and contain the waste, and that are subject to performance
requirements (see Section 2.2);

the surface environment, which is the environment to be protected.
In the TURVA-2012 safety case, the disposal system is described first in its initial state.
This is presented in Description of the Disposal System, wherein the initial state13 is
defined as:
“the state when the direct control over that specific part of the system ceases and only
limited information can be made available on the subsequent development of conditions
in that part of the system or its near-field”.
Description of the Disposal System provides a compilation of information presented in
more detail in various background reports, the most important being the Canister,
Buffer, Backfill, Closure and Underground Openings Production Line reports, in respect
of the engineered system, and in Site Description and Biosphere Description, in respect
of the natural system.
These descriptions provide the basis for considering the performance of the repository
system within Performance Assessment (Chapter 6 in this report) and the release of
radionuclides and calculation of doses in Assessment of Radionuclide Release Scenarios
for the Repository System and Biosphere Assessment (Chapter 8 in this report).
3.1
Host rock
The initial state for the host rock is defined to be the baseline conditions prior to starting
the construction of the ONKALO. The Olkiluoto site, as seen today, is the consequence
of events and processes that have taken place over billions of years, from those reflected
in the geological properties of the rocks forming the geosphere, to the shorter-term
changes related to climate-driven processes that mainly cause changes in groundwater
flow and groundwater composition and the geomechanical response to crustal
movements related to glacial loading and unloading. The present state and the past
history of the geology, rock mechanics and thermal properties, hydrogeology and
hydrogeochemistry of Olkiluoto are discussed in detail in Site Description.
The repository is to be excavated on Olkiluoto Island in south-western Finland (Figure
2-2). The crystalline bedrock of Finland is a part of the Precambrian Fennoscandian
13
Note that the definition is somewhat different for the surface environmnet and the host rock (see text).
68
Shield, which, in south-western Finland belongs to the Svecofennian domain, which
developed between 1930 Ma and 1800 Ma ago. The rocks of Olkiluoto consist of two
major classes: high-grade metamorphic rocks including gneisses with varying degree of
migmatisation, and igneous rocks including pegmatitic granites and diabase dykes
(Figure 3-1). The bedrock has been affected by five stages of ductile deformation
resulting in lithological layering, foliation, and strong migmatisation and folding.
Extensive hydrothermal alteration has also affected the properties of fractures and
certain rock volumes, the main alteration minerals being illite, kaolinite, sulphides and
calcite. As a result, the rock properties at Olkiluoto are heterogeneous, which is
reflected also in the variation of the thermal and rock mechanics properties and seen for
example in the anisotropic thermal properties due to foliation and gneissic banding.
The fault zones at Olkiluoto are mainly SE-dipping thrust faults formed during the latest
stages of the Svecofennian orogeny, approximately 1800 Ma ago, and were reactivated
in several deformation phases (see Figure 3-1 and Figure 3-2). In addition, NE-SW
striking strike-slip faults are also common. The occurrence of fracturing varies between
different rock domains, but the following three fracture sets are typical for the site: (i)
east-west striking fractures with generally subvertical dips to both the north and south,
(ii) north-south striking fractures with generally subvertical dips to both the east and the
west and (iii) moderately-dipping to gently-dipping fractures with strikes that are
generally sub-parallel to the aggregate foliation directions in a particular fracture
domain.
In Fennoscandia, the orientation of the major principal stress is attributed to an E-W
compression from the mid-Atlantic ridge push and a N-S compression from the Alpine
margin, resulting in a roughly NW-SE orientation of major principal stress (Heidbach et
al. 2008). This is also supported by the regional in situ data from Olkiluoto and other
Finnish sites studied during the site selection programme. Changes in isostatic load due
to glaciations and related isostatic adjustment and the existence of brittle deformation
zones change the stress regime at the site. Currently, a thrust faulting stress regime is
present, i.e. the horizontal stresses are larger than the vertical stress, H>h>v and the
principal stresses are approximately oriented horizontally and vertically, respectively.
The orientation of H at the site is found to vary slightly with depth and at the repository
depth is in the range NW-SE to E-W. The vertical stress is generally close to that
expected due to the weight of the overlying rock.
Located away from active plate margins, Fennoscandia, and Olkiluoto in particular is
known as a seismically quiescent region. Increased seismicity in Fennoscandia is
possible in connection to the most recent glaciation and post-glacial faults have been
discovered in northern Fennoscandia (e.g. Kuivamäki et al. 1998). There are no direct
signs of post-glacial faulting in the vicinity of Olkiluoto (e.g. Lindberg 2007) although
disturbances of the sea-bottom sediments have been suggested to be related to postglacial faulting (Hutri et al. 2007). According to the data from historical earthquakes,
the Olkiluoto area is located within a zone of lower seismicity, the Southern Finland
Quiet Zone (SFQZ), between two seismically active belts, Åland–Paldis–Pskov (Å-P-P)
69
Figure 3-1. A geological map of Olkiluoto Island showing the lithology and the brittle
fault zones (BFZ) defined as layout determining features, i.e. the ones that restrict the
repository layout.
Figure 3-2. Three-dimensional representation of the main hydrogeological zones at
Olkiluoto (HZ in blue) and their correlation with the fault zones (BFZ, in red) (outline
of the island is shown in the figure, oblique view towards the northeast).
70
and Bothnian Bay–Ladoga (B-L) (see Figure 2-2 in Saari 2008). These seismically
active zones seem to be essential elements when the driving mechanisms of the
seismicity of southern Finland are regarded. The zones are distinguished from their
surroundings particularly by the occurrence of relatively large (M  3.5) earthquakes.
In the crystalline bedrock at Olkiluoto, groundwater flow takes place in hydraulically
active deformation zones (hydrogeological zones) and fractures. The larger-scale
hydrogeological zones, which are related to brittle fault zones, carry most of the
volumetric water flow in the deep bedrock. There is a general decrease in the frequency
of transmissive fractures, and of transmissivity of both fractures and the
hydrogeological zones with depth. Under natural conditions, groundwater flow at
Olkiluoto occurs mainly as a response to freshwater infiltration dependent on the
topography, although salinity (density) variation driven flow also takes place to a lesser
extent. The porewater within the rock matrix is stagnant but exchanges solutes by
diffusion with the flowing groundwater in the fractures.
The distribution of the groundwater types is the result of progressive mixing of
groundwaters and the slow interaction between the groundwater, porewater and the
minerals of the rocks (see Figure 3-3 and Site Description). The groundwater
composition is also affected by microbial activity. Water-rock interactions, such as
carbon and sulphur cycling and silicate reactions, buffer the pH and redox conditions
and stabilise the groundwater chemistry.
Weathering processes during infiltration play a major role in determining the shallow
groundwater composition. Pyrite and other iron sulphides are common in waterconducting fractures throughout the investigated depth zone indicating a strong
lithological buffer against oxic waters over geological time scales. Groundwaters, in the
range down to 300 m depth show indications of having been affected by infiltrating
waters of glacial, marine and meteoric origin during the alternating periods of
glaciations and interglacials during the Quaternary. On the other hand, these indications
are absent in fracture groundwaters below 300 m, implying that these groundwaters are
older.
The current fracture groundwater is characterised by a significant, variation of salinity
with depth (see Figure 3-3). Fresh waters (<1 g/L) rich in dissolved carbonate are found
at shallow depths, in the uppermost tens of metres. Brackish groundwater, with salinity
up to 10 g/L dominates at depths between 30 m and about 400 m. Sulphate-rich waters
are common in the depth layer 100−300 m, whereas brackish chloride water, poor in
sulphate dominates at depths of 300−400 m. Saline groundwaters (salinity >10 g/L)
dominate at still greater depths. The matrix porewaters seem to be in equilibrium with
the fracture groundwaters in the upper part of the bedrock (0−150 m), suggesting
similar origin and strong interaction between groundwater in fractures and matrix at
these depths. At deeper levels (150−500 m), the matrix porewater is less saline and
increasingly enriched in δ18O; this has been interpreted to represent fresh water
conditions during a warm climate, probably during the preglacial Tertiary period, anion
exclusion being another possible explanation (Posiva 2009b).
71
Figure 3-3. Illustrative hydrogeochemical site model of baseline groundwater
conditions with the main water-rock interactions at Olkiluoto. Changes in colour
indicate alterations in water type. The hydrogeologically most significant zones are
represented. Blue arrows represent flow directions. Rounded rectangles contain the
main sources and sinks affecting pH and redox conditions. Enhanced chemical
reactions dominate the infiltration zone at shallow depths, and at the interface between
Na-Cl-SO4 and Na-Cl groundwater types. The illustration depicts hydrogeochemical
conditions in the water-conductive fracture system, not in the diffusion-dominated rock
matrix (Site Description).
3.2
Surface environment
The surface environment (including the overburden and the surface hydrogeology and
hydrogeochemistry) is described in detail in Biosphere Description. The initial state of
the surface environment corresponds to the description of the present-day conditions.
Topography in the Olkiluoto area, and in general in south-western Finland, is flat and
soil erosion rates are very low. Glacial erosion features such as glacially smoothed
72
bedrock outcrops and roches moutonnées14 are common. As a result of the last
glaciation, the bedrock depressions are filled with a thicker layer of overburden, mainly
sandy till and fine-grained till (e.g. Lahdenperä 2009 and Biosphere Description).
The sea around Olkiluoto Island is shallow, except for a few areas where water reaches
a depth of about 15 m. The seabed deposits in the surroundings of Olkiluoto are
heterogeneous and sediment thickness is variable. About 40−50 % of the offshore is
covered by till. Exposed bedrock and sedimentary rock form about 15−20 % of the area
of the seabed and various kinds of soft sediments cover about 20−30 % in the deeper
open sea area and in sheltered near-shore basins (Rantataro & Kaskela 2009, Ch. 5).
Littorina clays have been deposited in a marine environment, when environmental
conditions in the sea were favourable for a moderately abundant fauna. Because of the
amount of sedimented and intact organic material, these clays are gyttja clay in which
there has often been gas formation, inside the sediment, as a consequence of the
breakdown of organic material (Rantataro & Kaskela 2009, p. 13). This is of particular
interest because of the continued land uplift of the Olkiluoto area which will expose
areas of current seabed sediments in the next few thousand years – the time window for
which the radionuclide releases must be quantified in terms of annual doses to humans
and absorbed dose rates to other biota. In addition, the effects of land uplift are
accentuated by paludification e.g. reedbed growth in the coastal areas, especially in
shallow bays (Haapanen & Lahdenperä 2011).
The ecosystem succession during uplift, and the redistribution of sediments and
groundwater flow, will influence the areas of potential deep groundwater recharge and
discharge from the repository (Haapanen et al. 2007, 2009). The net changes in sea level
are a result of both crustal uplift and changes in sea level due to changes in climate
globally. The effect of uplift could be, at least to some extent, enhanced or reversed by
changes in sea level. This is taken into account in the climate scenarios in Biosphere
Assessment.
At present, freshwater (limnic) ecosystems are few in the Olkiluoto area and there are
no natural lakes on the island. The nearby lake basins were isolated during the various
stages of development of the Baltic Sea as a result of isostatic uplift and tilting of the
land. The closest rivers are the Eurajoki and Lapijoki, which discharge to the sea to the
north and east of Olkiluoto, increasing the concentration of nutrients and solids,
especially at the river mouths (Haapanen et al. 2009). The use of lakes and mires –
currently absent from the island – in the surrounding region as analogues for future
biosphere conditions at Olkiluoto is discussed by Haapanen et al. (2010). Figure 3-4
shows land uplift and an example of surface environment development through to
10,000 years from the present.
14
Outcrop of hard rock that has been shaped by the action of a glacial movement and erosion. The roche typically has a low,
smooth, rounded end pointing ’upstream’, with reference to the direction of ice movement, and a higher, rougher, ice-plucked,
downstream end. The surfaces may be marked by glacial striations.
73
Figure 3-4. Land uplift and an example of the biosphere development through to
10,000 a after the present. Map data: Topographic database by the National Land
Survey of Finland (permission 41/MYY/11) and Posiva Oy. Map layout by Jani
Helin/Posiva Oy. Note: dates are given as AD, i.e. 12020 is 10,000 years after the
reference date of 2020 AD.
74
3.3
Underground openings and repository layout
Underground openings of the disposal system include all spaces excavated
underground, including access tunnel, shafts, technical and demonstration rooms,
central tunnels, deposition tunnels and deposition holes. Drillholes within the area are
also included. Underground openings are constructed, utilised and backfilled in a stepwise manner during the course of the disposal operation. The initial state of the
underground openings is the state of rock at the time of emplacement of the material
(e.g. buffer, backfill) intended for a specific space. Rock Suitability Classification
criteria (RSC criteria) (McEwen et al. 2013) are applied during construction
(Underground Openings Production Line) to achieve conditions that meet the target
properties set for the host rock in Design Basis.
The drill and blast method is used in the construction of tunnels and raise boring in
shafts. To limit the extent and connectivity of the EDZ, the deposition holes will be
bored. To control groundwater inflow in the tunnels, grouting has been and will be used
when necessary during construction15 (for estimates of the grout quantities, see
Karvonen 2011), but grouting will not be allowed in the deposition holes.
At the repository level, central tunnels lead to deposition tunnels, in which the
deposition holes are located. Defining locations for these spaces follows the procedures
being set by the RSC system (McEwen et al. 2013, see also Section 2.2.3). The final
placing of the deposition tunnels and subsequently deposition holes is determined by
exploratory means before construction.
The layout of the underground openings is constrained by the layout-determining
features (LDFs), which are large lineaments, significant brittle fault zones (BFZ) or
hydrogeological zones (HZ) (McEwen et al. 2013). The layout used in the TURVA2012 safety case is presented in Figure 3-5 (Saanio et al. 2013). There is flexibility to
adapt the layout according to the additional and more detailed geological information to
be gained during construction. The basic statistics for the reference layout are
summarised in Table 3-1; deposition tunnel and hole dimensions depend on the fuel
type to be disposed. The layout provides for a total of 5400 positions for deposition
holes compared to 4500 actually required. This allows for rejection of some sections of
deposition tunnels and some deposition holes that do not meet the RSC criteria.
Several layout adaptations of the repository have been produced for a repository to host
either 5500 (Kirkkomäki 2009), 5440 or 9000 tU (Saanio et al. 2013). The final layout
will be adjusted in future taking into account the findings of the continued site
characterisation and possible other constraints (e.g. land use restrictions). The current
reference layout in the construction licence application is presented in Figure 3-6.
Provided that the deposition tunnels and holes are located so that major fault zones and
hydrogeological zones are avoided, the findings of the safety case are not sensitive to
details of the layout.
15
Low pH grouts are recommended to be used in ONKALO and are also required to be used as primary grouting material below the
depth of 300 m. Low pH grouts are used to avoid pH increase in groundwater. In addition, at repository level non-cementitious
grouts will be favoured where possible.
75
Foreign materials will be introduced during the construction and operation and,
although most will be removed and significant spills contained, inevitably some foreign
materials will be left in the repository. The amounts of foreign materials (e.g.
cementitious materials) introduced are being followed and controlled (including the
imposition of limitations) at the construction site and the total amounts are estimated
regularly (see Karvonen 2011 for the latest estimates).
Table 3-1. Layout statistics following the Description of the Disposal System report.
Common layout parameters
Total amount of the fuel to be disposed, tU
9000
Total number of canisters to be disposed
4500
Total number of canister locations in the layout
5400
Repository depth (metres below ground)
400 to 450
Orientation of the deposition tunnels
parallel to the main horizontal stress, which
is at repository level in the range NW-SE
and E-W.
Maximum length of deposition tunnel (metres)
350
Deposition hole diameter (mm)
1750
Fuel type specific parameters for:
Number of canisters
2
Nominal cross-section of deposition tunnels (m )
OL1−2
OL3−4
LO1−2
1400
2350
750
14.1
14.1 (OL3)
12.7
Deposition hole height (metres)
7.8
8.25
6.6
Distance between deposition holes (metres)
9.1
10.8
7.3
76
Figure 3-5. Layout adaptation for a repository hosting 9000 tU of spent nuclear fuel
used in the TURVA-2012 safety case, dark grey areas are not suitable for deposition
tunnels according to the RSC as they are intersected by LDFs and their respect
volumes. Red ovals denote respect distances to drillholes (Saanio et al. 2013).
Figure 3-6. The current reference layout (green). Grey areas are not suitable for
deposition tunnels based on Rock Suitability Classification (RSC). Red ovals denote
respect distances to drillholes. Red line surrounding the repository shows the area
reserved for the repository in urban planning.
77
3.4
Spent nuclear fuel
The spent nuclear fuel produced by the currently operating reactors, the OL3 unit under
construction and the planned OL4 unit are each different depending on the reactor type.
The OL1 and OL2 reactors at Olkiluoto are boiling water reactors (BWR), Loviisa LO1
and LO2 are VVER-440 type reactors and OL3, currently under construction, will be a
pressurised water reactor (PWR, trade name EPR). The design of the fuel assemblies
varies depending on the reactor type. The OL4 reactor type has not been decided yet and
in the TURVA-2012 safety case it is assumed to correspond to OL3. The initial state of
the spent fuel is described in Chapter 5 of Description of the Disposal System.
The spent nuclear fuel comes from storage at the nuclear power plants as assemblies
that contain the spent fuel pellets within alloy tubes and other metal parts comprising
the assembly. Figure 3-7 provides illustrations of different fuel assemblies. For each
fuel type, the canister design is adapted to accommodate the spent fuel to be disposed
(see Section 3.5 and Figure 3-8), in particular in terms of overall length and insert.
Figure 3-7. Representative illustrations (from left) LO1−2, OL1−2 and OL3 type fuel
assemblies. LO1−2 and OL1−2 fuel elements are partly cut open to show the internal
structures. The pictures are not to scale.
78
The spent fuel pellets from all the Finnish reactor types are, from a chemical point of
view, made of the same sintered material, UO2. However, the UO2 pellet geometry,
U-235 content and burnable poison (absorber) proportion, and cladding material, as well
as other components of the assembly, are different depending on the reactor type. The
average U-235 enrichment of a fuel assembly may vary roughly between 3–4 %, but
within a single fuel rod/pellet could be nearly 5 %. In the future, the assembly average
enrichment rate could be over 4 %, enabling higher burn-ups to be achieved. Fuel
cladding is made of various types of zirconium alloys because the zirconium crosssection for thermal neutrons is very small and because zirconium alloys typically have
great mechanical strength and good corrosion resistance. The other structural elements
of the fuel assemblies (i.e. upper and lower tie-plates, end plug, spacer grid, and channel
and nose piece) are fabricated from stainless steel, zirconium alloys or nickel-based
alloy). For more details, see Description of the Disposal System.
The most important properties that are considered are the material properties of the
assemblies, degree of burn-up, heat output levels and radionuclide inventory, which are
defined for the initial state (see Description of the Disposal System, Chapter 5). Some
basic characteristics of the fuel types are listed in Table 3-2.
3.5
Canister
The initial state of a single canister is the state when the canister filled with spent fuel
has been emplaced in a deposition hole, the surrounding buffer is present (Section 3.6)
and the deposition tunnel backfill has been emplaced on top of the deposition hole
(Section 3.7). The detailed design and initial state of the canister are described in
Description of the Disposal System and in Canister Production Line. Due to the several
spent fuel types (see Section 3.4 and Chapter 5 in Description of the Disposal System
for fuel details) there are geometrical differences between the canister types (Figure 38a).
The canister is composed of a cast iron insert and a copper overpack (Figure 3-8b). All
canister types have an external diameter of 1.05 m; heights vary between 3.5 and 5.2 m.
Table 3-2. Representative fuel characteristics for OL1−2, LO1−2 and OL3 fuels (per
assembly) (Canister Production Line report).
Fuel type
OL1−2
LO1−2
OL3
Mass of uranium (kg)
172−180
120−126
530−533
Anticipated maximum average burn-up of
50
a fuel assembly (MWd/kgU)
55
50
Estimated average burn-up of all the fuel
38−39
(MWd/kgU)
39−40
46−47
Typical enrichment U-235 (%)
3.3−4.4
3.6−4.4
3.6−4.2
Minimum cooling time of a single
assembly (years)
20
20
20
Minimum average cooling time with
average burn-up (years)
43.7
31.5
56.5
Allowable average decay heat at
disposal (full canisters) (W/tU)
806
950
862
79
The material for the copper components is phosphorus-alloyed oxygen-free copper with
the following requirements: O <5 ppm, P 30−100 ppm, H <0.6 ppm, S <8 ppm. Creep
testing of Cu-OF (oxygen-free copper) doped with 30 to 120 ppm phosphorus has
shown higher creep strength and much better creep ductility than copper without
phosphorus according to Andersson-Östling & Sandström (2009, Section 12). The cast
iron material composition of the insert is specified only with respect to an upper limit on
the content of copper to avoid the risk of radiation embrittlement. The content of copper
shall therefore not exceed 0.05 %. During the development of the casting process for the
nodular cast iron inserts, the standard requirements in EN 1563 grade EN-GJS-400-15U
have been used regarding mechanical properties (Raiko et al. 2010).
Dose rates outside the canister, canister temperatures, presence and composition of
water and gas in the canister, the type and probability of initial penetrating defects as
well as the residual welding stresses at the initial state are presented in Chapter 6 of the
Description of the Disposal System.
The design aim for the canisters is that all canisters are intact at emplacement and will
remain intact for hundreds thousands of years. It cannot be ruled out, however, that one
or few canisters with an initial penetrating defect may be emplaced (as discussed in the
Canister Production Line report). The cautious assumption that one or a few canisters
may have initial penetrating defects in the weld is based on expert judgement
concerning the canister welding method (electron beam welding − EBW) and nondestructive testing (NDT) capabilities. With more data becoming available in the future,
it is likely that it will be possible to demonstrate that the probability of emplacing more
than one canister with an initial undetected penetrating defect is less than one per cent.
At the moment, therefore, the number of defective canisters is assumed to be one
canister out of 4500 in the formulation of release scenarios; e.g. the reference case
realisation of the base scenario (see Chapter 7).
Figure 3-8. a) Canister geometries from left to right for spent fuels from LO1−2,
OL1−2 and OL3 (OL4). b) Canister overpack is made out of copper and the insert is
cast iron, spent fuel rods are placed in the channels in the insert.
80
3.6
Buffer
The reference buffer material is high grade Na-bentonite from Wyoming (MX-80).
Other similar materials can be considered in the future as long as they meet the design
requirements. The typical mineralogical composition for the buffer bentonite is given in
Table 7-3 in Description of the Disposal System.
The compacted rings and discs of buffer will be emplaced in the deposition hole that is
bored and accepted according to the RSC criteria. The canister will be emplaced within
the buffer as illustrated in Figure 3-9. The bottom of the deposition hole will be
smoothed to ensure a tight contact between the host rock and the lowermost buffer disk.
The inner gap between the buffer and the canister is unfilled and will present a small air
gap. The outer gap between the buffer blocks and rock will be filled with pellets of
buffer material. The buffer is in contact with the deposition tunnel floor backfill
material. Detailed dimensions of the buffer depend on the canister type to be emplaced
(see Figure 3-9) (also see Table 7-1 in Description of the Disposal System for buffer
dimensions). The initial water content in the buffer will be 17 %. The porosity of the
rings surrounding the canister is to be 36.0 % and the porosity of the discs on top and
below the canister is to be 38.2 %. Detailed buffer properties at the initial state
including the amounts of air, oxygen and water present, as well as impurities, are
presented in Description of the Disposal System (in Chapter 7, Table 7-2).
Figure 3-9. Illustration of the emplacement of the buffer and canister in the deposition
hole and placing the backfill on top (background figure).The buffer designs (three
designs depending on the spent fuel) are presented as schematic figures on left (from
left to right LO1−2, OL1−2 and OL3).
81
3.7
Backfill and plug
At the initial state of the deposition tunnel backfill both the clay backfill and the
deposition tunnel plug have been installed as illustrated in a schematic Figure 3-10. The
layout of the deposition tunnel backfill varies due to the different deposition tunnel sizes
within the repository, which are described in more detail in the Backfill Production Line
report and in Description of the Disposal System, Chapter 8.
3.7.1
Deposition tunnel backfill
The backfill components (blocks, foundation layer and pellets) will be emplaced in the
deposition tunnels as illustrated in Figure 3-10. The main backfill material in the
reference design is Friedland clay (blocks). Bentonite clay materials are used for pellet
fill surrounding the blocks (Cebogel pellets) and for the foundation layer to be emplaced
on the tunnel floor (Minelco granules). The foundation layer smoothes out the
unevenness of the drill and blast excavated floor in the tunnels allowing the
emplacement of the blocks.
Figure 3-10. Schematic figures showing (upper) the main backfill components in the
deposition tunnel and (lower) a cross-section of the OL1−3 tunnel with the schematic
presentation of theoretical excavation extent (min and max with dashed lines) and a
possible realisation of the backfilling components installed in the tunnel (Backfill
Production Line report).
82
Backfill has been designed taking into account the tolerances in excavation (Figure 3-10
lower figure). Prior to saturation, the average degree of saturation (ratio of volume of
water to volume of voids) is 55 % for the backfill. The rest of the void volume is filled
with air (45 %).
Detailed descriptions of the backfill materials, their composition and other properties at
initial state are given in Chapter 8 of Description of the Disposal System.
3.7.2
Deposition tunnel plug
The current deposition tunnel plug design is based on SKB’s plug design (see SKB
2010a) (see Figure 3-10 upper figure). It consists of a concrete dome; bentonite sealing
layer and a sand filter (see Description of the Disposal System for more details). The
combination structure ensures that the plug has sufficient hydraulic isolation capacity as
well as structural strength. However, the plug design is still at a conceptual level and
tests are needed to verify its hydraulic isolation capacity. In addition, the formulation of
the concrete mix is under development and may change in the future.
The concrete components (concrete plug and beams) in the deposition tunnel end plug
are made of low pH concrete (a concrete with a pH of the leachate < 10, with a short
period of initially higher pH of about 11), such as presented in the Backfill Production
Line report. For the formulation currently adopted, the water to cement ratio is 1.375
(kg/kg), water to binder ratio is 0.825 (kg/kg) and water to dry material ratio is 0.29
(kg/kg).
The installation of deposition tunnel plug allows the use of the central tunnels during the
operational phase. Thus, the hydraulic requirements set for the plug are only for the
operational period (see Table 2-2, L3-BAC-9 and L3-BAC-18).
The previous design by Haaramo & Lehtonen (2009) has been used as reference
geometry in the TURVA-2012 safety case (see Description of the Disposal System for
more details).
3.8
Closure
Closure will complete the isolation of the waste, restore and maintain favourable natural
conditions in the bedrock, and prevent the formation of preferential flow paths and
transport routes between ground surface and deposition tunnels and holes.
Closure of the disposal facility covers all backfilling and plugs outside the deposition
tunnels, including sealing of drillholes. The current reference design for the closure of
the disposal facility is presented in Figure 3-11. The detailed description of the
reference design, the backfill materials and methods, as well as the principles of the
hydraulic, mechanical and intrusion-obstructing plugs are given in Closure Production
Line report.
83
Figure 3-11. Current reference design for closure showing the access tunnel (1) and
shafts (2), technical rooms at the repository level (3) and the central tunnels (4) leading
from the technical rooms to deposition tunnels, location of the L/ILW repository is also
shown (5) (Closure Production Line report).
The reference design for closure will deploy a flexible tool-box of techniques to
accomplish closure to the required standards and performance requirements. The
available techniques provide alternative solutions throughout the closure process for the
emplacement of backfill and plugs in a manner that meets the requirements set. Natural
materials are utilised in backfill (such as clays, aggregates and mixtures of these). In
plugs, at least below structure HZ20 (a major hydrogeological zone lying above the
repository which forms a hydraulic discontinuity – see Figure 3-2), low pH concrete
will be used. See also Description of the Disposal System (Chapter 9) for the initial state
details.
84
85
4
FEATURES, EVENTS AND PROCESSES
This chapter describes the identification and screening of features, events and processes
(FEPs) and development of a database of FEPs relevant to the performance assessment
and analysis of potential radionuclide releases and radiological impacts. It also outlines
the onward use of the FEP descriptions in performance assessment and radiological
analyses and modelling, and outlines the future lines of evolution of the disposal system
and its environment.
4.1
Identification and screening of FEPs
4.1.1
Identification of potentially relevant FEPs
Identifying and describing the features, events and processes (FEPs) that are relevant to
understanding the evolution of the disposal system, or to its performance and safety, is
an essential step towards ensuring comprehensiveness of the assessments and safety
case.
For the TURVA-2012 safety case, the identification and description of FEPs and
couplings between these has been done based on scientific understanding without
considering the capabilities of the models that may be used to represent the processes or
disposal system components. This is to promote the elicitation of a complete set of
relevant FEPs, including FEPs that may not be included in the set of models that are
initially available. Following identification of a comprehensive list of FEPs a systematic
screening process is applied to rule out those FEPs that cannot be relevant in the context
of the repository system as proposed at the Olkiluoto site.
The identification and screening of FEPs was carried out by a team of scientific subject
and assessment experts. The process was conducted in a structured manner relying on
the collaborative (individual and joint) judgements of the experts.
This is the fourth iteration of Posiva’s Process Report, following from Vieno &
Nordman (1997) in support of the TILA-99 assessment, Rasilainen (2004) and Miller &
Marcos (2007) in support of the Interim Safety Case 2009 (Posiva 2010). The process
took advantage of and drew on the experience from previous Posiva studies, as well as
from the development of FEP lists in support of the assessment of the KBS-3V disposal
method in Sweden, notably Miller et al. (2002).
This is the first iteration also including FEPs for the surface environment. These have in
previous assessments mainly been addressed in the biosphere description reports
(Haapanen et al. 2007, 2009).
An initial long list of FEPs for identification and screening for the TURVA-2012 safety
case was derived from the previous Posiva Process Report (Miller & Marcos 2007), the
previous Bisophere Description Report (Haapanen et al. 2009), the NEA International
FEP list (NEA 1999) and its supporting project database (NEA 2006), together with
other relevant safety cases, notably SKB’s SR-Can assessment (SKB 2006)16.
16
SKB’s SR-Site assessment became available only after the initial FEP list for Features, Events and Processes was compiled.
86
4.1.2
Screening for relevance to TURVA-2012
The FEPs in this long list were then examined to determine their relevance and potential
significance against the following criteria:

relevance to the KBS-3V type repository design for spent nuclear fuel disposal;

relevance to the present-day Olkiluoto site characteristics and likely future site
characteristics evolving in response to climatic changes and other external factors;

relevance to the national regulatory requirements and guidelines;

previous experience in FEP screening and safety case development by Posiva (see
above);

knowledge and information gaps identified during the course of Posiva’s and SKB’s
ongoing RTD programmes;

the outcomes from previous safety cases for and performance assessments of the
KBS-3V type repository design;

expert knowledge and awareness of other developing national and international
RTD and safety case programmes, and

feedback from the regulatory agency (STUK) on previous safety case reports and
Posiva’s RTD programme.
Examples of FEPs that were excluded during the screening process included:

‘Spent fuel degradation due to high-pH waters (pH > 10)’ was excluded because
expert judgement indicates that it is unlikely that significant quantities of high-pH
water generated by the chemical degradation of cementitious materials can migrate
through the buffer and come into contact with the spent fuel. Reaction between
bentonite and backfill and high-pH waters is addressed, however, because of the
physical proximity of parts of the buffer [5.2.6] and backfill [6.2.5] to the
cementitious plugs at the ends of the deposition tunnels.

‘Deliberate human intrusion’ was excluded because it is assumed that, if an
intrusion is deliberate, appropriate measures would be taken to protect people and
the environment.
Most of the FEPs that were screened out in the process were also screened out in
previous FEP analyses for TILA-96 (Vieno & Nordman 1997) and TILA-99 (Vieno &
Nordman 1999). The process of FEP documentation therefore focused on ‘retained’
FEPs, thus Features, Events and Processes contains descriptions only of those FEPs
that passed screening and are considered potentially significant for the long-term safety
of the disposal facility.
87
4.1.3
Organisation of the FEPs
The FEPs in Features, Events and Processes are organised according to the main
components of the disposal system: Spent nuclear fuel; Canister; Buffer; Backfill;
Auxiliary components17; Geosphere; Surface environment; the external FEPs are also
discussed. For each component, FEPs affecting the evolution the disposal system
(termed evolution-related FEPs) and FEPs relevant to radionuclide transport (migrationrelated FEPs) are also distinguished. This leads to the list of retained FEPs as shown in
Table 4-118.
Table 4-1. List of retained FEPs organised according to disposal system component as
well as the external FEPs.
3.
Spent nuclear fuel
3.2.1
Radioactive decay (and in-growth)
3.2.9
Release of the labile fraction of the
inventory
3.2.2
Heat generation
3.2.10
Production of helium gas
3.2.3
Heat transfer
3.2.11
Criticality
3.2.4
Structural alteration of the fuel pellets
3.3.1
Aqueous solubility and speciation
3.2.5
Radiolysis of residual water (in an intact
canister)
3.3.2
Precipitation and co-precipitation
3.2.6
Radiolysis of the canister water
3.3.3
Sorption
3.2.7
Corrosion of cladding tubes and metallic
parts of the fuel assembly
3.3.4
Diffusion in fuel pellets
3.2.8
Alteration and dissolution of the fuel
matrix
4.
Canister
4.2.1
Radiation attenuation
4.3.1
Aqueous solubility and speciation
4.2.2
Heat transfer
4.3.2
Precipitation and co-precipitation
4.2.3
Deformation
4.3.3
Sorption
4.2.4
Thermal expansion of the canister
4.3.4
Diffusion
4.2.5
Corrosion of the copper overpack
4.3.5
Advection
4.2.6
Corrosion of the cast iron insert
4.3.6
Colloid transport
4.2.7
Stress corrosion cracking
4.3.7
Gas transport
5.
Buffer
5.2.1
Heat transfer
5.2.9
Freezing and thawing
5.2.2
Water uptake and swelling
5.3.1
Aqueous solubility and speciation
5.2.3
Piping and erosion
5.3.2
Precipitation and co-precipitation
5.2.4
Chemical erosion
5.3.3
Sorption
5.2.5
Radiolysis of porewater
5.3.4
Diffusion
5.2.6
Montmorillonite transformation
5.3.5
Advection
5.2.7
Alteration of accessory minerals
5.3.6
Colloid transport
5.2.8
Microbial activity
5.3.7
Gas transport
17
Auxiliary components refers to backfilling of central tunnels, service areas, access tunnel and shafts, and seals and plugs that are
installed both at the mouths of the deposition tunnels and as part of closure. In Features, Events and Processes, all backfill issues
are discussed in the chapter Tunnel backfill, so the chapter Auxiliary components focuses on the seals and plugs (both the
deposition tunnel plugs and the closure plugs).
18
The numbering of the FEPs is based on the chapters and sections in Features, Events and Process, where the discussion of FEPs
begins in Chapter 3 (Spent nuclear fuel).
88
6.
Backfill
6.2.1
Heat transfer
6.3.1
Aqueous solubility and speciation
6.2.2
Water uptake and swelling
6.3.2
Precipitation and co-precipitation
6.2.3
Piping and erosion
6.3.3
Sorption
6.2.4
Chemical erosion
6.3.4
Diffusion
6.2.5
Montmorillonite transformation
6.3.5
Advection
6.2.6
Alteration of accessory minerals
6.3.6
Colloid transport
6.2.7
Microbial activity
6.3.7
Gas transport
6.2.8
Freezing and thawing
7.
Auxiliary components
7.2.1
Chemical degradation
7.2.3
Freezing and thawing
7.2.2
Physical degradation
7.3.1
Transport through auxiliary components
8.
Geosphere
8.2.1
Heat transfer
8.2.10
Microbial activity
8.2.2
Stress redistribution
8.3.1
Aqueous solubility and speciation
8.2.3
Reactivation-displacements along
existing fractures
8.3.2
Precipitation and co-precipitation
8.2.4
Spalling
8.3.3
Sorption
8.2.5
Creep
8.3.4
Diffusion and matrix diffusion
8.2.6
Erosion and sedimentation in fractures
8.3.5
Groundwater flow and advective transport
8.2.7
Rock-water interaction
8.3.6
Colloid transport
8.2.8
Methane hydrate formation
8.3.7
Gas transport
8.2.9
Salt exclusion
9.
Surface environment
9.2.1
Erosion
9.2.22
Gas generation
9.2.2
Degradation
9.2.23
Ingestion of food
9.2.3
Podzolisation
9.2.24
Inhalation of air
9.2.4
Agriculture and aquaculture
9.2.25
Respiration
9.2.5
Forest and peatland management
9.2.26
External radiation from the ground
9.2.6
Infiltration
9.2.27
Exposure from radiation sources
9.2.7
Groundwater discharge and recharge
9.2.28
Topography
9.2.8
Runoff
9.2.29
Well
9.2.9
Drainage
9.2.30
Construction of a well
9.2.10
Capillary rise
9.2.31
Food source potential
9.2.11
Uptake
9.2.32
Dietary profile
9.2.12
Evapotranspiration
9.2.33
Demographics
9.2.13
Translocation
9.2.34
Exposed population
9.2.14
Litterfall
9.3.1
Terrestrialisation
9.2.15
Bioturbation
9.3.2
Advection
9.2.16
Migration of fauna
9.3.3
Dispersion
9.2.17
Senescence
9.3.4
Water exchange
9.2.18
Atmospheric deposition
9.3.5
Sedimentation and resuspension
9.2.19
Atmospheric resuspension
9.3.6
Ingestion of drinking water
9.2.20
Diffusion
9.3.7
Flooding
9.2.21
Sorption
9.3.8
Water source potential
89
10.
External FEPs
10.2.1
Climate evolution
10.2.4
Land uplift and depression
10.2.2
Glaciation
10.2.5
Inadvertent human intrusion
10.2.3
Permafrost formation
4.2
Development of the FEP descriptions
4.2.1
FEP descriptions
The aim of the FEP descriptions (presented in Features, Events and Processes) is to
provide a concise summary of the current understanding of each FEP and its associated
uncertainties, and to set out when and where it may occur in the disposal system, and
how it is coupled to other FEPs. Interactions between disposal system components are
also indicated. The descriptions are not intended to define how the FEP is treated in the
TURVA-2012 safety case, which is part of the assessment and modelling judgement
described in Performance Assessment, Formulation of Radionuclide Release Scenarios,
Models and Data for the Repository System, and for the surface environment in Terrain
and Ecosystems Development Modelling, Surface and Near-Surface Hydrological
Modelling, Biosphere Radionuclide Transport and Dose Assessment and Dose
Assessment for Plants and Animals.
A description of each of the retained FEPs (Table 4-1) was compiled by a subject matter
expert familiar with Posiva’s RTD and assessment programme in the relevant topic
area. Each description includes the current fundamental scientific understanding and
expected relevance in the context of the Posiva disposal system at the Olkiluoto site,
e.g. in terms of temporal and spatial occurrence, plus a note of any fundamental
uncertainties in scientific understanding. Extensive references are also included.
All FEPs are described using a common template with standard fields as defined in
Table 4-2.
90
Table 4-2. Common template for FEP descriptions.
Name and number: a unique name and number are given to each FEP; these are the same as those
used in Posiva’s developing FEP database.
Type: feature, event or process.
Class: whether the FEP is system evolution-related or migration-related, or both. (There is not always a
clear-cut distinction, but the distinction is useful to indicate whether the FEP is to be considered in
performance assessment, radiological assessment or both.)
General description: a concise description of the FEP and its consequences for system evolution and
safety, covering:
• current fundamental scientific understanding,
• any relevant properties, conditions and constraints that affect its operation,
• the likely temporal and spatial variability of its operation,
• the conditions under which it may be expected to occur at Olkiluoto, and
• when in the evolution of the disposal system it is expected to occur.
Reference is made to evidence for the FEP from RTD studies in the field or laboratory. Where possible,
quantities are given to illustrate the ‘magnitude’ of the FEP (e.g. process rates), but these are not intended
to define the actual parameter values or ranges used in the safety case; these are defined in Models and
Data for the Repository System, and for the surface environment, the data used in the biosphere
assessment are summarised in Biosphere Data Basis, and the models are discussed in Terrain and
Ecosystems Development Modelling, Surface and Near-Surface Hydrological Modelling, Biosphere
Radionuclide Transport and Dose Assessment and Dose Assessment for Plants and Animals.
Uncertainties in the understanding of the FEP: a short discussion about any uncertainties in conceptual
understanding and gaps in knowledge, but generally not about uncertainties in data or modelling
approaches; these are defined in Models and Data for the Repository System and Biosphere Data Basis.
Couplings to other FEPs: this section lists which FEPs affect or are affected by the FEP being described.
References: key references to the main scientific literature and relevant Posiva reports, particularly any
recent summary reports on the FEP. This is not intended to be a comprehensive bibliography, but is a
‘signpost’ to the relevant literature.
4.2.2
Coupling between FEPs and aggregation/disaggregation
Interaction matrices were developed for the FEPs of each main component of the
disposal system. These were used to check and identify which FEPs are coupled with
each other, and also to highlight which FEPs are most important to controlling
radionuclide transport between components of the disposal system.
When identifying the potentially significant FEPs, expert judgement was applied when
deciding how to address coupled processes. For example, ‘water uptake by the buffer’
and ‘swelling of the buffer’ are two separate but connected processes. In Features,
Events and Processes and in the retained FEP list (Table 4-1), these two processes are
addressed in a single FEP description [5.2.2 Water uptake and swelling] because they
are so closely coupled that they can be addressed as a single process.
In other cases, closely connected processes have been described separately because they
may need to be addressed differently in the development of scenarios and in the
performance assessment models. For example, ‘radioactive decay’ and ‘radiogenic heat
generation’ are very closely related but they each have their own FEP descriptions
[3.2.1 and 3.2.2].
In some cases, FEPs have been aggregated or disaggregated differently compared with
the previous version of the Process Report (Miller & Marcos 2007). This reflects
developing understanding of the roles and importance of different FEPs. The individual
91
FEP descriptions include references to features, events and processes in the previous
version of the report, for traceability.
The process of developing the FEP database and the complete set of descriptions,
references and other information is presented in Features, Events and Processes.
4.3
Onward use of the FEP descriptions
The initial state of the repository can be described by a set of features that depend on the
amount and characteristics of the spent nuclear fuel to be disposed of in the repository,
repository design and characteristics of the site. These features are discussed in
Description of the Disposal System.
As noted above, FEPs are categorised as evolution-related FEPs, which mostly affect
the physical state of the disposal system, or migration-related FEPs which mostly affect
the release, transport and accumulation of radionuclides (and other chemical species).
The synthesis of evolution-related FEPs provides a description of the ways in which the
disposal system, its components and its environment might evolve, termed future lines
of evolution. A summary of the future lines of evolution for the disposal system
environment and key components is presented in the next section; these future lines of
evolution is an important input to Performance Assessment and Formulation of
Radionuclide Release Scenarios.
The evolution-related FEPs for each component also provide a checklist of processes
that should be considered in evaluating the performance of the repository system and its
components; this is described in Performance Assessment. This consideration leads to
the confirmation that under expected initial conditions and likely future lines of
evolution there should be no release of radionuclides over several hundreds of
thousands of years, but also identifies the conditions and events that could lead to
radionuclide releases.
The migration-related FEPs for each component provide a checklist of processes that
should be considered in the models for radionuclide release, transport in the repository
system, and dispersion and impact in the biosphere, as described in Models and Data
for the Repository System and Biosphere Data Basis, Terrain and Ecosystems
Development Modelling, Surface and Near-Surface Hydrological Modelling, Biosphere
Radionuclide Transport and Dose Assessment and Dose Assessment for Plants and
Animals. Thus, those FEPs most important to radiological assessment are carried
forward to Assessment of Radionuclide Release Scenarios for the Repository System and
Biosphere Assessment.
4.4
Future lines of evolution
The understanding of FEPs is used to develop descriptions of the future lines of
evolution of the repository system itself (the engineered barriers and host rock) and of
the natural setting of the site. This provides the framework for estimating the thermal,
hydraulic, mechanical, and chemical loads that will be placed on the system.
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During the construction and operation of the repository up to its closure, the main
changes will be related to excavation effects and draining of water from the
underground openings, plus introduction of heat from the spent fuel. Rock stress
changes and some limited damage immediately around the openings are expected, as
well as an increase of the groundwater flow into the repository volume, and hence
changes in hydrogeology and hydrogeochemistry. After closure, the groundwater flow
regime will return towards preconstruction conditions, although modified by radiogenic
heat from the spent fuel for a time. Hydrogeochemical changes induced during the open
period may persist for longer because of the very low groundwater flows.
In the longer term, the main driver for changes will be climate evolution, where the
expected case is a reversion to glacial-interglacial cycling as experienced over the last
one million years of the Quaternary. However, best scientific understanding indicates
that past and continuing anthropogenic emissions of CO2 and other greenhouse gases
will lead to increased global temperatures over a period of many thousands of years,
delaying the onset of cooler climate conditions. Thus, over the next 50,000 years (50
ka), conditions are expected to remain essentially as today, i.e. a temperate climate with
the boreal ecosystem. Glaciation-related crustal uplift (isostatic rebound) will continue,
although at a decreasing rate, and will outpace global (eustatic) sea-level rise; this will
lead to local sea-level fall relative to the land, at least for the next several millennia.
Minor changes in hydraulic boundary conditions will occur due to relative sea-level fall
and build up of peat deposits in low-lying areas.
A first cold period is expected to occur at about 50 thousand years after present (ka AP)
with temperature and precipitation changes leading to permafrost development and,
later on, to ice-sheet development (see Ch. 4 in Formulation of Radionuclide Release
Scenarios based on Pimenoff et al. 2011). The general form and degree of changes
beyond this can be estimated based on reconstruction of past global climate changes,
although there is uncertainty in the timing of changes. For the assessment, a repetition
of the sequence of events during the last glacial cycle (the Eemian-Weichselian) is
assumed. Reliable proxy data exist for this cycle, which can be taken as representative.
In the assumed forward sequence, the first permafrost period is between 50 and 60
ka AP, with a second permafrost period between about 73 and 81 ka AP, before the
onset of three successive periods of ice sheet cover lasting until about 155 ka AP
(Figure 4-1).
After 170 ka AP, a repetition of the cycle from 50 ka to 170 ka AP is assumed. This is
as expected in the absence of anthropogenic effects and with a return to naturally-driven
climate cycling. In this case, seven glacial cycles including alternating temperate and
cold periods may be expected from 170 ka AP to 1000 ka AP. That is a total of eight
glacial cycles in the assessment time frame (up to one million years after present).
Realistically, variability between future glacial-interglacial cycles is to be expected (in
both duration and intensity), but that this would not substantially alter the results of the
performance or safety assessment.
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A 130
110 90 70 50
30
10 0
ka BP 50
70
90
110
130
150
170
ka AP
B Permafrost Ice Sheet Temperate Time window (ka AP) →
Climate type ↓
Temperate (T)
Cold /Permafrost (P)
Up
to 50
50–
60
T
60–
73
7381
T
P
8192
92106
106113
113132
131141
141156
T
P
Cold/Ice Sheet (IS)
P
IS
P
IS
IS
Temperate (T) = climate as today; Permafrost (P); Ice Sheet (IS)
Figure 4-1. Schematic representation of the occurrence of permafrost, ice sheets, and
temperate safety assessment climate types during (A) the last glacial cycle, (B) assumed
repetition of the past glacial cycle from 50 ka AP onwards, and the representation of
the future sequence in terms of the climate types in time windows.
94
95
5
MODELS AND DATA
This chapter describes the various models and data needed for the analyses supporting
the safety case. Four separate categories of model can be recognised, based on their
purpose:

models describing the climate evolution and climate-driven processes; these models
frame the analysis of the disposal system by giving information on the climate
conditions that occur within defined time windows (Section 5.1);

models needed to represent the features, events and processes that are the main
drivers of the repository system’s evolution and used to assess the performance of
the engineered barrier system and conditions in the repository host rock (Section
5.2);

models used for analysing radionuclide release and transport from the near field
through the geosphere to the surface environment (Section 5.3);

models used for the biosphere assessment including models describing the
development of the surface environment, models describing radionuclide transport
in the surface environment and models for assessing potential doses to humans,
plants and animals (Section 5.4).
Different models are used for each of the components, depending on the processes
described. These models are linked, however, such that the initial and boundary
conditions as well as the input data for the different models are selected in a consistent
way. Thus, the output from one model can be used as an input to the next in the
modelling chain. The first three items in the above list are described in full in Models
and Data for the Repository System; the last item is described in Biosphere Data Basis
and Biosphere Assessment. These reports include references to the studies that describe
the models and data in more detail.
5.1
Models and data for climate evolution and climate-driven
processes
Climate modelling has been carried out to support the definition of the climate scenarios
and define the time windows for temperate, permafrost and glacial climate conditions
(Pimenoff et al. 2011). The main factors affecting the climate evolution and the onset of
the next glaciation are the Earth’s orbital variations and associated variations in solar
insolation as well as the changing atmospheric CO2 concentration. Simulations of future
climate evolution were made using constant reasonable concentrations of CO2 and
emission scenarios for the current century and the consequent evolution of CO2
concentrations over the next several millennia. Changes in solar activity and in the
concentration of volcanic dusts in the atmosphere may also affect climate, but such
changes are uncertain and were not included in the models.
The estimation of the climate states on a time scale of 120,000 years is based on
analysis of climate simulations of an Earth-System model (CLIMBER-2) coupled with
an ice-sheet model (SICOPOLIS). The large-scale output of the climate was downscaled
using a regression model (GAM) so that the climate conditions relevant for the
Olkiluoto site could be extracted (see Figure 5-1). In addition, a more detailed study of
96
the climate evolution on a time scale of 10,000 years was carried out (Pimenoff et al.
2012). In this study, other types of Earth System models (MPI/UW, UVic) were used.
The climate modelling approach is summarised in Figure 5-1 and Table 5-1.
The modelling results have been used to define the climate scenarios, as well as
providing input to the permafrost modelling, groundwater flow modelling and the
development of the surface environment at the Olkiluoto site.
Permafrost occurs when cold and dry climate conditions prevail for extended periods
without ice sheet cover, and beneath cold-based ice sheets. The development of
permafrost and frozen ground depends on heat-exchange processes across the
atmosphere-ground boundary layers and on an almost time-independent geothermal heat
flow from the Earth’s interior. A main driver for permafrost development is the
evolution of the climate conditions at the site. The development of permafrost at the
Olkiluoto site has been modelled by Hartikainen (2013). The model describes the
freezing and thawing of groundwater-saturated bedrock either in 1D or 3D by
considering heat transfer and freezing accounting for groundwater salinity and hydraulic
pressure. The bedrock is considered as an elastic porous medium and the groundwater
as an ideal solution of water and ionic solvents. The model has been used to represent
the selected future evolution of the climate (see Section 4.4) with the climate conditions
based on Pimenoff et al. (2011, 2012). The crustal radiogenic heat production and the
heat generated by the spent fuel were taken into account. Key input for the permafrost
modelling is shown in Table 5-1. The results of the modelling have been used for the
formulation of the radionuclide release scenarios as well as providing input to
groundwater flow modelling.
The rock responds to the weight of the ice sheet by uplift in front of the advancing ice
margin, by depression during the advance of the ice sheet and by uplift during and after
the ice sheet retreat. The response of the rock to the ice load is governed by the
rheological parameters of the rock. The ice loading has an effect on the rock stresses,
hydrogeology and hydrogeochemistry of the site. Post-glacial crustal uplift together
with changes in the global sea level is an important factor affecting the development of
the surface environment. The post-glacial uplift considered in the biosphere assessment
and groundwater flow modelling has been modelled based on the semi-empirical model
presented by Påsse (2001), see Table 5-1. The model parameters are defined by fitting
the model to relevant shoreline displacement data and by using crustal thickness data.
For the current safety case, the parameters are defined (Chapter 9 of the Biosphere Data
Basis) taking into account a revision of Bothnian Sea shore-level data and using a
derivative method applying crustal thickness and current uplift maps as well as taking
into account complementary data e.g. archaeological observations.
97
Input to CLIMBER-2 – SICOPOLIS
Atmospheric CO2 concentration
Solar insolation
Atmospheric dust
Geothermal heat flux
CLIMBER-2-SICOPOLIS
Combination of Earth System Model of Intermediate Complexity
(EMIC, CLIMBER-2) used for simulating the climate evolution and an
Ice sheet model (SICOPOLIS) describing the evolution of the Northern
Hemisphere ice sheets
Output of CLIMBER-2
Output of SICOPOLIS
Mean near-surface air temperature
Mean precipitation
Solar flux at the surface
Global radiation at the surface
Vegetation
Evolution of NH ice sheets
Ice sheets basal temperature
Bedrock elevation
Distance to the ice sheet margin
Topography
Mean near-surface temperature
Mean precipitation
GAM
Downscaling the output of the large-scale global model to regional
scale using a regression model based on regional climate data.
Output of GAM
Mean near-surface temperature
Mean precipitation at Olkiluoto
Figure 5-1. Schematic presentation of the climate modelling approach using global
models and downscaling the results to the regional scale.
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Table 5-1. Main models and codes used for modelling climate evolution and climatedriven processes for the safety case TURVA-2012.
Models and codes
Purpose and scope
Climate evolution
Modelling of the climate evolution to support the selection of the climate scenarios and climatic conditions
during temperate, permafrost and glacial periods.
The key input for the model is shown in Figure 5-1.
Main references: Pimenoff et al. (2011, 2012)
CLIMBER-2
Earth System Model of Intermediate Complexity (EMIC) used for simulating
the climate evolution (near-surface air temperature, precipitation, solar flux at
the surface, global radiation at the surface and vegetation) on a time scale of
120,000 years.
SICOPOLIS
Ice-sheet model describing the evolution of the Northern Hemisphere ice
sheets, their thickness and areal extent, basal temperature and bedrock
elevation.
GAM
Generalised Additive Model used to downscale near-surface air temperature
and precipitation from the CLIMBER-2-SICOPOLIS results to the Olkiluoto
area.
MPI/UW
Earth System Model of Max Planck Institute used for the estimation of the
climate evolution (atmospheric CO2 concentration, near-surface air
temperature, precipitation, sea level, incoming shortwave radiation at the
surface) on a time scale of 10,000 years. The model enables a coupled
treatment of atmosphere, ocean, sea-ice, ocean carbon cycle and dynamic
vegetation.
UVic
Earth System Model of the University of Victoria used for the estimation of
the climate evolution (atmospheric CO2 concentration, near-surface air
temperature, sea level, incoming shortwave radiation at the surface) on a
time scale of 10,000 years. The model enables a coupled treatment of
atmosphere, ocean, sea-ice, ocean carbon cycle and land and terrestrial
vegetation carbon cycle and oxic-only sediment respiration.
Permafrost modelling
Modelling of permafrost development e.g. depth of the permafrost to support the formulation of the
radionuclide release scenarios as well as input to groundwater flow modelling.
Key inputs to the permafrost modelling are the climate conditions, e.g. air temperature and vegetation,
based on the climate model (see above), soil cover and water bodies and the properties of the rock mass
and groundwater based on Olkiluoto specific data and the heat generated by the spent fuel.
Main reference: Hartikainen (2013).
1D model
Freezing and thawing of groundwater saturated rock by applying a 1Dmodel.
3D model
Freezing and thawing of groundwater saturated rock by applying a 3Dmodel.
Crustal uplift
Crustal uplift for groundwater flow modelling and the development of the surface environment.
Key inputs for the model are shoreline displacement data and crustal thickness data.
Main references: Påsse (2001) and Chapter 9 of Biosphere Data Basis
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5.2
Key models and data for performance assessment and for
formulation of radionuclide release scenarios
The aim of performance assessment is to evaluate the behaviour of the engineered
barrier system and, in particular, to confirm that the performance targets will be met and
that the host rock conditions will remain consistent with the target properties under the
different conditions of the repository system evolution (operational period, thermal
period and repository saturation, temperate climate, permafrost and glaciation). The
main modelling activities (Figure 5-2) in support of the performance assessment are:

modelling of the hydraulic evolution, hydrogeochemical evolution and rock
mechanical evolution of the geosphere − giving estimates of flow rates, flow paths,
groundwater composition, stress state, rock damage around the underground
openings including deposition tunnels and deposition holes and shear displacements
along fractures;

description of the evolution of the closure, which considers degradation of the
closure with consequent release of alkaline leachates;

modelling of the mechanical and hydraulic evolution, geochemical evolution and
mechanical and chemical erosion of the buffer and backfill, which results in an
estimate of the mass loss, porewater composition and hydraulic conductivity of the
buffer;

modelling of the initial defects of the canister, the canister corrosion, mechanical
evolution of the canister and creep, which results in an estimate of the overall
canister performance;

thermal evolution is considered in the modelling of each of the barriers.
The models applied consider the main processes affecting the performance of a barrier.
There are couplings between the processes and across the barriers. These are not always
modelled explicitly, but are taken into account by selecting consistent boundary and
initial conditions. The models and data used for assessing performance of the disposal
facility are discussed in Models and Data for the Repository System and the results
summarised in Performance Assessment. Key input for the assessment is the initial state
of the repository system and the barriers provided by the Production Line reports and
summarised in Description of the Disposal System. A summary of the models and codes
used in support of the assessment of the geosphere evolution is presented in Table 5-2
and of the EBS in Table 5-3.
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Figure 5-2. Overview of the models used for performance assessment. Permafrost depth
and ice-sheet thickness from climate and climate-driven modelling (Section 5.1) and
results from surface environment modelling (Section 5.4) are applied to assessment of
the performance of the repository system. Green boxes present the main processes
modelled and yellow ovals present the main outcomes of the models.
101
5.2.1
Models and data for geosphere evolution
Modelling of geosphere evolution – hydraulic evolution, geochemical evolution and
rock mechanical evolution – aims to determine whether favourable conditions, such as
low flow rate, reducing and also otherwise favourable groundwater chemistry, and
mechanically stable characteristics prevail at the repository depth. The main modelling
activities and codes used for describing the geosphere evolution are presented in Table
5-2. The modelling assesses the impact of the excavation of the tunnels, presence of
open tunnels, heat generated by the spent fuel, as well as of the natural phenomena
related to the ongoing crustal uplift and to future glaciations. Site Description provides
the description of the bedrock and the groundwater system at the Olkiluoto site, and,
together with Features, Events and Processes, the interacting processes and
mechanisms. Site Description also provides input to modelling the geosphere evolution.
The modelling of geosphere evolution, in turn, provides input for assessing the
performance of the EBS.
Groundwater flow modelling has been carried out to represent the hydraulic evolution
and to assess the flow conditions including flow rates, flow paths both to and from the
repository, and the salinity evolution in the geosphere, especially around and through
the repository and underground facilities. Two types of flow models, equivalent
continuous porous medium (ECPM) combined with a dual porosity (DP) approach and
discrete fracture network (DFN) models, have been applied in the groundwater flow
modelling. The ECPM conceptualisation has been used mainly to simulate the evolution
of the groundwater flow at the site scale. The site-scale model modelling considers
density driven groundwater flow and thermal conduction in order to study the effect of
the heat generated by the spent fuel and the effect of variable salinity. In the DP
approach, advection and dispersion are the dominant processes within the water-bearing
fractures, whereas in the rock matrix, solutes are transported by diffusion. DFN models
have been used to describe the distribution of the groundwater flow on a detailed scale
in the vicinity of deposition tunnels and deposition holes. The approach is based on a
stochastic representation of the bedrock fractures.
The main processes considered in the geochemical evolution in the bedrock are mixing
of groundwaters, and water-rock interactions. The hydrogeochemical evolution at the
site under the temperate climate during the first 10,000 years and during ice-sheet
retreat has been studied by reactive transport modelling applying FASTREACT
(FrAmework for Stochastic REACtive Transport, Trinchero et al. 2010). The model
evaluates the mixing of the infiltrating waters with the initial waters, taking into account
the main reactions between these mixed waters and rock matrix and fracture minerals
along the streamlines. Microbial populations and processes also affect the groundwater
composition. The microbially mediated reactions have, so far, not been explicitly taken
into account in the reactive transport modelling. However, considering only inorganic
oxygen consumption in the reactive transport modelling is cautious as the microbial
reactions would increase oxygen consumption. The role of microbially mediated
reactions in sulphate reduction has been assessed based on the extensive groundwater
and microbial sampling data from the site. These reactions have an important role in
defining the redox conditions in the host rock. Leachates from the cement components
102
e.g. grouts and in plugs, will also affect the groundwater composition (see discussion in
Geochemical evolution of backfill and buffer in Section 5.2.2).
The main modelling activities related to rock mechanical evolution have focussed on the
disturbances caused by the excavation and the thermal load generated by the spent fuel,
stress evolution and rock stability during a glacial cycle, and the rock shear
displacements caused by future earthquakes. Different methods have been applied to
estimation of the spalling around the underground openings including deposition
tunnels and deposition holes; traditional continuum thermomechanics and fracture
mechanics approaches as well as analytical methods. Statistical methods have been used
for defining the input data. The evolution of the rock stresses during the different phases
of a glacial cycle has been described. Further, the interaction of the in situ stress state
and the fault zones, i.e. how the fault zones affect the stress magnitudes and
orientations, and the stability of the faults under different stress states have been
studied. Shear displacements in fractures induced by post-glacial seismic events in
nearby fault zones have also been studied. The modelling considered a selection of the
fault zones at Olkiluoto with varying orientation, extent and location with respect to the
repository.
Table 5-2. Main models and codes used for host rock evolution in TURVA-2012.
Models and
codes
Purpose and use
Hydrological evolution
Groundwater flow modelling has been carried out to describe the groundwater flow rates, flow paths and
salinity distribution in the geosphere and around the repository. Two conceptualisations of the flow have
been applied, equivalent continuous porous medium (ECPM) combined with a dual porosity (DP) approach
and discrete fracture network (DFN) models. The shore line evolution and the assumed climate conditions
are taken into account by initial and boundary conditions and the repository layout and the underground
openings and their properties are represented in the models. The results are applied to assessing the
geochemical evolution and performance of the EBS, and they also provide input to the radionuclide release
and transport analysis.
Main references: Löfman & Karvonen (2012), Hartley et al. (2013 a,b,c)
ConnectFlow
Software package for simulation of the groundwater flow at different scales using both
ECPM and DFN approaches and their combination. Enables detailed study of the flow paths
by the stochastic representation of individual fractures and representation of the
underground openings in the model. The hydrogeological DFN models and hydrogeological
structure model are key inputs to the model. Three different scales (see Figure 5-3);
regional, site and repository, with varying levels of detail of the representation of the
sparsely fractured rock, hydrogeological zones and underground openings (e.g. deposition
holes, deposition tunnels, other tunnels and shafts) are used. Main results used in the
safety case are the inflow to and flow rates around the deposition holes and deposition
tunnels, flow paths between the repository and the surface environment and flow and
transport properties along these paths, but the salinity evolution has also been modelled.
FEFTRA
The finite-element program package for groundwater flow modelling which applies the
ECPM approach to model transient and density-driven flow and heat transfer by conduction
and the DP approach for modelling salt transport. The rock is represented by two hydraulic
units, hydrogeological zones and sparsely fractured rock. Both of these units have averaged
hydraulic properties based on site-specific data. The tunnel system is considered by using
the appropriate boundary conditions to represent the hydraulic or thermal impact of tunnels.
Key inputs to the model include outputs from the hydrogeological structure model and the
hydraulic conductivity of the sparsely fractured rock, properties of the rock related to salt
transport and thermal properties of the rock. Main results used in the safety case are the
evolution of the flow conditions and groundwater salinity.
Geochemical evolution
103
Models and
codes
Purpose and use
Assessment of the geochemical evolution of the site is based on understanding of the past evolution of the
site as well as on reactive transport modelling. The main processes considered are mixing of groundwaters,
and water-rock interactions. The FASTREACT (FrAmework for Stochastic REACtive Transport, Trinchero et
al. 2010) approach has been used to combine the flow and the chemical reactions. In this approach, particle
tracking methods are used to define particle trajectories, streamlines, along which reactive transport
simulations are carried out. Expected and bounding groundwaters for different future conditions are defined
based on these assessments and are used for assessing the performance of the EBS as well as for
determining the solubility, speciation and retention parameters for the radionuclide release and transport
analysis.
Main references: Models and Data for the Repository System, Trinchero et al. (2013), Wersin et al. (2013c)
PHREEQC
Reactive transport modelling code used for assessing the evolution of the groundwater
chemistry at the site. The solute concentration in the whole set of particle trajectories is
reproduced using a set of PHREEQC one-dimensional reactive transport simulations where
the longitudinal coordinate (e.g. the distance from the infiltration location) along the
trajectory is interpreted in terms of travel time. The key inputs to the model are the velocity
field and the initial salinity field derived from the FEFTRA model for representative times,
which are used to define the streamlines. Further inputs are site-specific mineralogical data
and the composition of the initial and infiltrating waters. The main results are the
groundwater composition at the repository depth and evaluation of the buffering capacity of
the host rock against the infiltrating waters.
Rock mechanics evolution
Modelling has been carried out to estimate the disturbances caused by the excavation and thermal load
generated by the spent fuel, the stress evolution at site during a glacial cycle and the rock shear
displacements caused by earthquakes. The results have been taken into account in the groundwater flow
modelling and in estimation of the likelihood of canister failures due to shear displacements.
Main references: Site Description, Ch. 9.2, Valli et al. (2011), Lund & Schmidt (2011), Fälth & Hökmark
(2011, 2012), Hakala et al. (2008)
3DEC
A rock mechanics code, a three-dimensional numerical program based on distinct element
method for discontinuum modelling. 3DEC simulates the response of the (such as jointed
rock mass) subjected to either static or dynamic loading. Thermal loads can be taken into
account. The discontinuous medium is represented by an assemblage of discrete blocks
and the discontinuities are treated as boundary conditions between the blocks. The code
has been used for several studies including spalling predictions and assessment of the
effect of the faults on the in situ stress field and analysis of the shear displacements due to
end-glacial earthquakes. Key inputs for the analyses have been site-specific data on rock
mechanics properties of the rock mass, fractures and deformation zones and thermal
properties of the rock, rock stresses and the estimated effect of the glacial load.
Fracod2D
A fracture mechanics code based on the Displacement Discontinuity Method (DDM) that
has been used for predicting potential for spalling. Key input for the analysis have been sitespecific data on rock mechanics properties of the rock mass, fractures and deformation
zones and thermal properties of the rock and rock stresses.
ABAQUS
The finite element code used for calculating the glacially induced stresses, which combined
with a synthetic regional background stress model, is used for assessment of the fault
stability. In addition to the stress model, key inputs for the analysis include a model of the
Weichselian ice sheet, and variant Earth models for the lithosphere.
104
Figure 5-3. Groundwater flow models and data flows between them (after Hartley et al.
2013b).
5.2.2
Models and data for engineered barrier system performance
The models and data used for assessing performance of the EBS are discussed in
Models and Data for the Repository System and the results summarised in Performance
Assessment. This section summarises the main models, codes and data used (for a
summary, see Table 5-3).
Thermal evolution
The model that encompasses both the EBS and the geosphere is the thermal evolution
model. Decay heat from the spent nuclear fuel will increase temperatures within and
around the repository for up to 50,000 years. In developing the repository design and
layout, the principal constraint is to limit the maximum temperature experienced at the
buffer/canister interface to 100 °C; this is to preclude significant thermal alteration that
might degrade the desired swelling and hydraulic properties of the buffer. The
temperature evolution also affects groundwater flow and permafrost development.
Therefore, the thermal evolution needs to be modelled for the entire repository system.
Heat is transferred from the canister through the host rock mainly by conduction
through the different materials between the canister and the wall rock. The conductivity
of the copper, of the bentonite buffer (and backfill), the degree of saturation of the
buffer and the presence of air/water gaps at the interface with the canister and with the
rock also affect the near-field thermal evolution. Heat is transferred through the
105
geosphere by a combination of conduction in the solid rock mass and advection of
flowing groundwater through fractures. The thermal properties of the rock/groundwater
system, together with groundwater flow, will therefore govern the temperature evolution
throughout the geosphere. Diffusivity (of heat) is the most important thermal parameter,
which is a function of thermal conductivity, heat capacity and density of the rock mass.
Thermal dimensioning of the repository and the resulting temperatures are presented in
Ikonen & Raiko (2013). The model is described in Models and Data for the Repository
System.
Mechanical,
components
hydraulic
and
geochemical
evolution
of
closure
There is no specific performance assessment model for the mechanical, hydraulic and
geochemical evolution of closure components. Degradation of cement and the various
processes involved are qualitatively discussed in Section 6.7.3 of Performance
Assessment.
In the groundwater flow model, the decreased performance of the deposition tunnel and
closure backfill is taken into account (Hartley et al. 2013b) assuming that the hydraulic
conductivity of the tunnels is one to two orders of magnitude higher than the reference
assumptions.
Based on experimental data and monitoring of leachates of cement grouting of fractures
in ONKALO, the evolution of pH and alkali contents in cement leachates has been
evaluated. The output from this estimate is then used to assess the impact of cement
leachates on buffer and backfill in Performance Assessment.
Mechanical and hydraulic evolution of backfill and buffer
A model for the mechanical and hydraulic evolution of the backfill and buffer is needed
to evaluate the capability of the buffer and the backfill: to protect the canister from rock
movement, to provide a favourable hydraulic and mechanical environment for the
canister, to limit transport of harmful substances (e.g. sulphides) to the canister, and to
limit transport of radionuclides out of the canister in case of a release. The main relevant
properties of the buffer and the backfill are swelling pressure and hydraulic
conductivity. These properties depend, among other parameters, on the thermohydraulic behaviour of the clay.
The finite element code CODE BRIGHT is used to model the thermo-hydraulic
behaviour of clay. Although the code is able to represent the mechanical behaviour of a
porous medium in a coupled form, only the thermal and water flow capacities of the
code have been used. In CODE BRIGHT, equations for mass balance were established
following the compositional approach. That is, mass balances were performed for water
and air instead of using solid, liquid and gas phases. The equation for balance of energy
was established for the medium as a whole.
The final objective was to estimate the temperature, T, and liquid pressure, Pl, for the
thermo-hydraulic analysis from water mass and energy balance equations. Gas pressure
was assumed constant in these analyses.
106
The mechanical model reproduces the behaviour of the porous medium under
mechanical boundary conditions (displacements and forces or stresses). This is called a
constitutive model and for this application the model used is the BBM (Barcelona Basic
Model). The model is formulated within the framework of hardening plasticity using
two independent sets of variables: the excess of stress over air pressure and suction. The
BBM model is able to represent many of the fundamental features of the behaviour of
partially saturated soils. On reaching saturation, the model becomes a conventional
critical state model (Schofield & Wroth 1968).
Geochemical evolution of backfill and buffer
The following processes potentially affecting the main properties of the buffer, such as
its swelling pressure and its transport properties. These are assessed either through
modelling or by reference to experimental data.

Oxygen depletion and changes in pH: The end point is the time it takes for all the
originally trapped air to be consumed in the buffer and in the backfill, i.e. how long
it takes for the conditions to become anoxic in the near field. This is estimated using
a mass-balance approach or a semi-empirical approach that takes into account the
reaction rate of O2 with pyrite in saturated or unsaturated conditions (see Section
5.5.2 in the Performance Assessment for details).

Evolution of colloid population: The evolution of the population of colloids
introduced with the buffer and the backfill, especially those associated with the
degradation of repository materials, and their mobility and stability under changing
groundwater conditions is not yet well defined. Bentonite colloids in low ionic
strength groundwaters are currently being addressed in the CFM project in Grimsel
(Möri 2004) and in the Äspö Colloid Project (Laaksoharju & Wold 2005). There are
also some experimental data from URL studies to date – although in the Äspö
Colloid Project, bentonite colloids were shown to be unstable in the high ionic
strength deep Äspö and Olkiluoto groundwaters, and so the concentrations were as
low as those in the natural groundwater. In the Performance Assessment the initial
colloid concentration was neglected (Section 5.5.3 in the Performance Assessment)

Effect of cementitious leachates on the buffer and backfill: The impact of
cementitious leachates on the engineered barrier system and the rock is assessed
based on experimental data and monitoring of leachates from cement grouting of
fractures in ONKALO (Sections 5.5.4 and 6.5.8 in Performance Assessment). The
effects of leaching of other sealing materials (Silica sol) on bentonite are considered
through experimental data (Section 5.5.5 in Performance Assessment).

Geochemical evolution of unsaturated buffer porewater during the thermal stage:
This has been assessed by thermo-hydro-geochemical modeling, using the integral
finite-difference code TOUGHREACT. The objective of the modelling was to
identify changes in porewater chemistry and to assess the redistribution of salts and
minerals that could induce cementation (Section 6.5.2 in Performance Assessment).

Buffer and backfill porewater chemistry after saturation: The porewater chemistry
and its evolution have been modelled by coupled diffusion-reaction modelling,
accounting for the evolving groundwater composition. The porewater composition is
approximated assuming diffusive equilibration and chemical equilibrium between
107
the groundwater and the clay. Porewater composition has been modelled using the
code PHREEQC. This has been done in a thermodynamic model for bentonite,
which is based on the microstructure, the electrochemical properties of the clay and
the anion exclusion concept (Wersin et al. 2013a) (Section 6.5.5 in Performance
Assessment).

Microbial activity in the buffer and sulphide production at the buffer/rock interface:
The gypsum (sulphate) pool in the buffer will dissolve, and may be a source of
sulphide production at the buffer/rock interface in zones of larger porosity through
the activity of sulphate-reducing bacteria (SRB). This process was considered in a
bounding analysis, using both an analytical shrinking core model and a linear 1D
reactive transport model. From these considerations, maximum sulphide fluxes
towards the canister were estimated (Section 6.5.7 in Performance Assessment).

Microbial activity in the backfill and the effect of organic materials on sulphide
production in backfill: No process-based model has yet been developed. Microbially
induced sulphide production is evaluated through experimental data, a mass balance
approach and chemical kinetics considerations (Sections 6.6.3 and 7.4.7 in
Performance Assessment).

Microbial activity, organic carbon and sulphide production in backfill: Gypsum and
organic carbon in the backfill are potential sources for microbial activity and
sulphide production, which might occur in low porosity zones at the backfill/rock
boundary. This rather complex process was addressed with a step-wise approach.
First, a bounding analysis including an analytical shrinking core model and a 1D
reactive transport model was conducted to estimate maximum sulphate and sulphide
fluxes into the rock and the buffer (Wersin et al. 2013c). Then, iron sulphide
formation, an important sink for sulphide, was addressed by simple equilibrium
calculations and natural analogue data. Finally, a more advanced, but still
preliminary coupled reactive transport model with different boundary assumptions
at the backfill/rock interface was applied to verify the results obtained from the
simple model considerations, to assess both sulphur and iron fluxes and to evaluate
uncertainties (Sections 6.6.3, 6.5.7, and 7.4.7 in Performance Assessment).

Montmorillonite transformation: the processes of illitisation and cementation have
been assessed using semi-empirical approaches and data from natural analogues. For
illitisation, two approaches were used: kinetic rate equations as a function of the
temperature and mass transfer constraints. Natural analogues considerations were
used to assess the smectite-to-illite conversion rates for relevant repository
conditions (see Section 6.5.3 of Performance Assessment). Cementation was
assessed through experimental data from the LOT experiment (Sena et al. 2010,
Karnland et al. 2009) and reactive transport modelling to calculate the amount of
silica precipitated from dissolution-precipitation processes of accessory minerals
(see Section 6.5.4 of Performance Assessment for details). For the long-term
performance of bentonite, thermodynamic and kinetic considerations as well as
natural analogue evidence were used (Section 7.4.5 of Performance Assessment).
The effect of the canister corrosion products on buffer stability was assessed on the
basis of empirical data from the LOT experiment (Section 7.4.6 of Performance
Assessment). Iron-clay interactions in the backfill are discussed using a mass
balance approach taking into account the rock bolts and the remaining steel mesh
108
and metal debris which are in direct contact with the backfill. The results are
discussed in Section 6.6.5 of Performance Assessment.
Mechanical and chemical erosion of buffer and backfill
Mechanical and chemical erosion of the buffer and the backfill are processes that may
lead to mass loss both with consequent loss of swelling pressure and an increase of
hydraulic conductivity. Mass loss in the buffer can also affect the ability to filter
colloids in the buffer and to limit microbial activity. Therefore, the processes that may
lead to erosion have to be modelled to assess the effect on the transport properties of the
buffer and the backfill.
For mechanical erosion, the background data and basis for mass loss estimation are
presented in Section 5.4 of Performance Assessment. A simplified approach is used to
estimate buffer and backfill mass loss. This approach is based on the calculation of mass
loss as a function of the total volume of water flowing through a piping channel in the
buffer or backfill and the erosion rate. The volume of eroding water is based on the
distribution of the groundwater flow between the deposition holes and the deposition
tunnels and thus it depends on the groundwater flow model results. The erosion rate is
based on experimental tests done by Posiva and SKB in conditions relevant to those at
Olkiluoto in terms of salinity, flow and geometric properties of the fracture potentially
intersecting the deposition tunnel or deposition hole. The eroded mass loss is derived by
numerical integration of the measured erosion rate as a function of cumulative flow of
water. From the eroded mass loss, the impact on hydraulic conductivity and swelling
pressure can be estimated.
The chemical erosion model for predicting the rate of erosion of the bentonite buffer in
low ionic strength water used has been developed by Moreno et al. (2010). The model
details of chemical and surface chemical processes related to chemical erosion, the input
data and experimental results are presented in Section 7.5.4 of Performance Assessment.
The model uses the results of discrete fracture network (DFN) flow modelling to
estimate the chemical erosion rate. Three processes are modelled: 1) Transport of
sodium ions in montmorillonite pore water, 2) Expansion of montmorillonite in the
fracture and 3) Flow of montmorillonite gel and sol and of water. The same model has
been used to make equivalent calculations for the backfill by Sane et al. (2013). Schatz
et al. (2013) performed a series of small-scale, flow-through, artificial fracture
experiments in which swelling clay material could extrude/erode into a well-defined
system representing a fracture, in order to provide the basis for modelling the potential
extrusion/erosion behaviour of bentonite buffer material at a transmissive fracture
interface.
The chemical erosion model was applied to the repository at Olkiluoto, using the
following data:

the evolution of groundwater velocity with time in fractures intersecting the
deposition holes and tunnels;

the transport apertures of these fractures;
109

the periods during which low-ionic strength conditions can be expected to prevail
around each deposition hole.
A key input assumption is the amount of buffer and backfill mass loss that would lead
to advective conditions in the deposition holes. The data used are from Section 10.3.9 of
SR-Site (SKB 2011).
The end point of the model is the number of deposition holes in which advective
conditions are established as a function of time based on the mass of eroded buffer (or
backfill). The output of this model is then used as input for the canister corrosion
calculations by sulphide attack in the long term (see canister corrosion, below).
Freezing and thawing of buffer and backfill
Freezing of the buffer or backfill is not expected as permafrost is not expected to reach
repository depths. If it occurred, freezing of the bentonite porewater and/or backfill
porewater would be expected to affect swelling pressure. A theoretical description of
the temperature dependence of swelling pressure for the buffer was derived (Birgersson
et al. 2010). The pressure responses of fully saturated buffer and backfill material
samples have been observed down to -10 °C. The results are discussed in Section 7.3 of
the Performance Assessment.
Canister corrosion
The evolution of the copper overpack has been assessed. The following corrosion
processes are considered:

atmospheric corrosion,

localised corrosion,

generalised corrosion due to sulphide (including microbially induced corrosion) and
to oxygen, and

stress corrosion cracking as well as copper corrosion in oxygen-free water.
The corrosion loads were determined by assessing the total amount or the flow of a
given corroding agent toward the copper surface and the duration of a given corrosion
process. For each of the identified corrosion processes and corrosion loads, calculations
were carried out to estimate the corrosion depth that the processes can induce on the
copper wall.
The extent of damage on the canister surface due to localised corrosion, mostly from
chloride ions, has been assessed from experimental and natural analogue results rather
than by calculation. A given corrosion depth (rather than using a pitting factor, as was
done in the past) is then attributed to the effect of localised corrosion from chloride ions.
The generalised corrosion of the copper canister due to oxygen and sulphide has been
calculated based either on mass balance approach or on diffusive flow of the corroding
agent to the copper surface. The flow of corroding agent (e.g. oxygen or sulphide)
assumed to react with the canister is estimated using the results from the geochemical
models for groundwater and for the buffer and backfill (see above). The groundwater
flow conditions in each deposition hole (derived from the DFN model) are taken into
account to determine the flow of corroding agent to the canister.
110
The state of the buffer with respect of erosion is another important assumption affecting
the diffusion rate of corroding agents to the copper surface. The copper surface that is
exposed to the corroding agent (i.e. the entire surface of the canister or a limited area of
the canister) determines the corrosion depth and ultimately the durability of the canister.
In the case of stress corrosion cracking and other postulated corrosion processes (e.g.
copper corrosion in anoxic water) the conditions in which they occur are discussed in
Features, Events and Processes and reported in Performance Assessment. The corrosion
depths expected from the different corrosion processes during the near-field evolution
are then compared with the total corrosion depth and the number and timing of canister
failures by corrosion are thus estimated.
Copper creep and copper overpack lifetime
The stresses and strains developing in the copper canister during its first decades of
repository service are modelled using finite element analysis simulations optimised to
predict the EB-welded copper canister creep life. The creep model takes into account
the swelling of bentonite around the canister and the presence of initial residual stresses.
The model includes elastic-plastic as well as creep simulations. The model simulates the
mechanical properties of the base material and that of the EB weld and it includes the
base material creep rupture model, base material creep strain model, as well as the EB
weld strain model. The uniaxial creep model is translated to a multiaxial constitutive
equation form using a VTT in-house translation routine developed for the use of creep
models that are not directly supported by the computational tool ABAQUS (Holmström
et al. 2013).
Mechanical loads on the canister
The mechanical loads on the canister have been assessed for the design of the canister
so that the canister fulfils its containment safety function for several hundred thousand
years. The expected mechanical loads during canister evolution are due to the
following:

swelling of the bentonite (even or uneven),

isostatic load due to the presence of an ice sheet on the surface and

dynamic loads in the case of a rock shear movement that could happen in connection
with the advance or retreat of an ice sheet.
The mechanical response of the canister is analysed using 2D- or global 3D-finiteelement models, including large-deformation and non-linear material modelling as
described in Raiko (2012) and in Raiko et al. (2010). In some cases, the creep model
developed by SKB has been used, as described in Raiko et al. (2010). In the case of rock
shear movement, the mechanical response of the canister as well as that of the buffer
surrounding it is analysed.
In addition to the expected loads, a disturbance scenario of freezing of the bentonite
buffer down to -5 °C during permafrost has been analysed using the same model (Raiko
2012, Section 8.4.4). The canister mechanical integrity is assessed partly from the stress
and strain results obtained using global models and partly from fracture resistance
analyses using the sub-modelling technique (see below). The sub-model analyses utilise
the deformations from the global analyses as constraints on the sub-model boundaries
111
and more detailed finite-element meshes are defined with defects included in the
models, together with elastic-plastic material models. The canister mechanical analyses
are carried out in collaboration between SKB and Posiva (as described in Raiko et al.
2010). The SKB data are considered applicable to Olkiluoto because of the similarities
in future climatic conditions. In the case of the isostatic load from the ice sheet, the
isostatic load used by SKB is more severe than the one expected at the Olkiluoto site
since, according to the climate model (see Section 5.1). The maximum thickness of the
ice sheet expected at the Finnish site is 2 km whereas that expected at the Swedish site
is 3 km.
Number of initial penetrated canisters
The potential number of defective canisters that might be emplaced in the repository has
been analysed probabilistically by Holmberg & Kuusela (2011). The main objective of
this study was to estimate the reliability of the welding and of the non-destructive
testing (NDT) processes for detecting penetrating defects in the copper overpack. After
varying the assumptions used as input data, it was concluded that the currently available
data are insufficient to make a reasonable estimate of the probability of emplacing a
penetrated canister in the repository. A Bayesian approach was therefore used based on
the existing experience on welding and NDT method reliability. The outcome of the
model is a preliminary estimate of the number of initially defective canisters that may
be incidentally emplaced in the repository.
Table 5-3. Main models and codes used for EBS performance assessment in safety case
TURVA-2012.
Models and codes
Purpose and use
Thermal evolution
Modelling activity carried out to describe the evolution of the temperature in the near field (at the canister
surface in the buffer and at the buffer/rock interface) and in the far field. Analytical and numerical
approaches developed by VTT are used.
Main references: Ikonen & Raiko (2013), Raiko (2012).
Analytical and numerical
approaches
Analytical and numerical approaches are used to model the heat transfer
among the different barriers. The processes represented are conduction,
radiation and convection. The input data are principally the heat output of the
canister (which in turn is determined by the spent nuclear fuel loading and
the storage time of the spent nuclear fuel prior to emplacement in the
canister), the repository layout and the thermal properties of the canister,
bentonite buffer and backfill, and the surrounding rock.
Mechanical, hydraulic and geochemical evolution of closure components
No codes, empirical data only.
Main references: See Sections 5.6, 6.7 of Performance Assessment.
Mechanical and hydraulic evolution of backfill and buffer
Modelling activity carried out to describe the duration of saturation and piping and erosion issues.
Main references: Sections 5.4 and 6.4 of Performance Assessment.
CODE BRIGHT (Olivella et
al.1994, 1996)
The finite element code CODE BRIGHT is used to model the thermohydraulic behaviour of clay. Although the code allows study of the
mechanical behaviour of a porous medium in a coupled form, only the
thermal and water flow capacities of the code have been considered. Initially,
the code was developed for non-isothermal multiphase flow of brine and gas
through porous deformable saline media. Key assumptions used in the
material models include: Fourier’s law for heat transport, Fick’s law for non-
112
Models and codes
Purpose and use
advective liquid flow (diffusion) and Darcy´s law for advective liquid flow. The
heat flow from the canister and the temperature and liquid evolution at the
boundaries are the key boundary conditions. The initial conditions are the
initial temperature and the liquid pressure distribution, which is considered
hydrostatic in rock and negative in buffer and backfill (they are unsaturated at
the initial state).
BBM (Barcelona Basic
Model, Alonso et al. 1990)
The BBM is a critical state model that reproduces the mechanical behaviour
of unsaturated soils under different boundary conditions (displacements and
forces or stresses). The BBM is an extension of the modified Cam clay
model that has become popular in applications involving unsaturated soils
and, in particular, in simulations using the finite element method. Partially
saturated soils can be loaded in different ways, for instance, mechanically
and/or hydraulically. In addition, cycles of loading and unloading can be
applied. Key input data used in the model include: the initial porosity, initial
temperature and initial liquid pressure, which is negative because the
materials analysed are unsaturated at the initial state. The initial stresses are
used as well. The boundary conditions in the mechanical model are stresses
or constant rates of displacement (usually zero, i.e. fixed boundary).
Geochemical evolution of backfill and buffer
The following are the main codes used for performance assessment (processes assessed only through
data are discussed in the text):
PHREEQC
Code used to describe the buffer and backfill porewater composition based
on a thermodynamic equilibrium model (described in Wersin et al. 2013a,
2012b). The key input data for the buffer are the groundwater composition
and the mineralogical and pore water composition of the bentonite. The
dissolution and precipitation of accessory minerals as well as surface
reactions including cation exchange and protonation/deprotonation are taken
into account.
TOUGHREACT
Integral finite difference code used for thermo-hydro-geochemical modelling
(Xu et al. 2008, Idiart et al. 2013). The objective of the modelling is to identify
changes in porewater chemistry during the early evolution of the system. In
addition to mineral reactions, cation exchange was accounted for in the
geochemical model. Key input includes mineral composition of the buffer and
backfill, groundwater composition and groundwater flow rates in the
repository near field.
Mechanical loads on canister
2D- or global 3D-finite-element models are used including large-deformation and non-linear material
modelling and, in some cases, also creep. The models include material models for copper and bentonite
mechanical behaviour. All materials are modelled according to the elastic-plastic material model developed
by von Mises (von Mises 1913). The material models include strain hardening, in some cases also strain
rate hardening, swelling pressure dependency, and temperature dependency. In elevated temperature
analyses, creep models are also used for copper and in some cases also for iron. Codes applying the finite
element method (FEM) are also used to simulate material flaws or lack of material. In the case of cracks,
fracture parameters are calculated and the allowable crack sizes are determined. The bentonite material
model is described in Börgesson et al. (2010). The swelling pressure and the yield strength of the saturated
bentonite are strongly dependent on the density of the bentonite. Different strength and swelling pressure
estimates for Na and Ca bentonites are used. The rock shear analysis used a strain-rate dependent
material model, so the material stress-strain curve was presented for static and dynamic strain rates
(Dillström 2010, Hernelind 2010).
Main references: Raiko (2012) and Raiko et al. (2010).
ABAQUS (Hillbit et al.
1994) and ANSYS
(www.ansys.net)
Codes used in 3D FEM-analyses. ABAQUS is used to model the bentonite
material and the rock shear response of the buffer/canister/insert system.
ANSYS is used to simulate static (stationary), dynamic (moving) and heat
transfer (thermal) problems. Key inputs to the codes include the material
models described above as well the loading on the canister buffer system
e.g. isostatic load, the shear load in terms of velocity and forced displacement. The geometry of the canister and buffer is modelled in full 3D with solid
elements, including gaps and contacts, and fillets of structural corners.
113
Models and codes
Purpose and use
Copper creep and copper overpack lifetime
The creep model used for canister overpack creep simulation has been developed for Posiva by Holmström
& Auerkari (2006). The model is based on the logistic creep strain prediction model, which is a creep strain
prediction tool able to predict representative creep strain curves and strain rates in a large stress and
temperature range. The geometry used in the simulations is that of the OL1 and OL2 (BWR) spent fuel
canister.
The creep model developed by Sandström et al. (2009) has been used in the canister design strength
analyses carried out in collaboration with SKB.
Number of canisters with an initial penetrating defect
Probabilistic calculation to assess the probability to emplace one or more canisters with an initial
penetrating defect in the repository.
Main reference: Holmberg & Kuusela (2011)
Estimation of the number
of canisters with an initial
defect based on the
Bayesian method
5.3
Combines information about the reliability of the welding process (Ronneteg
et al. 2006, p. 105) and that of the NDT methods as well as human error
(Swain & Guttmann 1983) to assess the number of canisters with an initial
penetrating defect that could be accidentally emplaced in the repository. Key
input assumptions are the reliability of the sealing process (only one of 100
canisters might have a critical flaw) based on the opinion of manufacturing
experts and the probability for human error during the NDT process (0.003)
based on a general screening value for human errors in nuclear power plant
risk analysis.
Models and data for the analysis of radionuclide release scenarios
In the models used in the analysis of most release scenarios, radionuclides released from
a failed canister are dissolved in water and conveyed in solution through the repository
near field and through the geosphere towards the biosphere (gas- and colloid-mediated
transport are also considered in some calculation cases). It is assumed that radionuclides
migrate from the repository near field to the geosphere, but not vice versa. Similarly,
radionuclides may migrate only from the geosphere to the biosphere and not vice versa.
These model simplifications are cautious and consistent with the assumption that
radionuclide transport in the geosphere is dominated by advection (retarded by matrix
diffusion and sorption) from the near field to the biosphere.
The modelling of radionuclide release, retention and transport in the repository system
is carried out in two steps: near-field release, retention and transport modelling, and
geosphere retention and transport modelling. The near field comprises the spent fuel, the
deposition holes including the canisters and buffer, the backfilled deposition tunnels and
that part of the immediately surrounding host rock that is affected by the presence of the
repository (e.g. excavation-disturbed zones). The geosphere comprises the remainder of
the host rock. The models and flow of information in near-field and geosphere
modelling are shown in Figure 5-4 and the codes used are shown in Figure 5-5 and
summarised in Table 5-4. The models, flow of information and the codes used in
biosphere assessment including dose assessment are also shown in these figures, and are
described in more detail in Section 5.4.
To ensure the traceability of the modelling calculations, a procedure has been developed
for the management and documentation of the input and output of the various models.
This procedure has been applied to each calculation case; it includes the specification of
the models and parameter values to be used, checking of output files, some of which are
114
used as input files in subsequent steps in the calculation chain, and the storing of both
input and output in the assessment database.
Description of fuel
and engineered
barrier system
Bedrock description
Near-field release and
transport
Near-field flows
Geosphere
transport
resistances
Groundwater flow
modelling
Radionuclide transport in the
geosphere
Geosphere paths
and release
locations
Description of
surface
environment
Terrain and ecosystem
development model
Landscape
model set-up
Surface and
near-surface
hydrology
model
Landscape model
(radionuclide transport in
the biosphere)
Yes
Dose
calculations
needed?
Annual activity
release
Complementary
indicators
Annual doses to
humans
Dose models
Biosphere
assessment
Absorbed dose
rates to plants and
animals
Assessment
endpoints
Figure 5-4. Models and information flows. Radionuclide release and transport models
are shown in white boxes. System descriptions and understanding are shown in light
blue boxes, key supporting models in green boxes and their principal outputs in dark
blue ovals.
115
GoldSim
Near field
MARFA or
GoldSim (for PSA)
Geosphere
Annual activity
release
Dose assessment
needed?
No
Yes
Biosphere
UNTAMO, SHYD,
Pandora, Ecolego
Radiological
impact
MATLAB, ERICA
Assessment
endpoint
Annual activity
release
Radiation doses
Figure 5-5. Computer codes for analysis of radionuclide release scenarios. Codes used
for the analysis of repository system scenarios are shown in red and discussed in this
Section. Codes used for biosphere assessment are shown in blue and discussed in
Section 5.4.
For modelling purposes, the radionuclide inventory inside the spent nuclear fuel
canisters is assigned to three characteristic locations, namely (i) the fuel matrix, (ii) the
grain boundaries and gaps, and (iii) structural materials (zirconium alloys and other
metal parts). Following canister failure and contact of the fuel with water, there will be
a relatively rapid release to solution of the radionuclide inventory at grain boundaries, in
gaps and in corroding metals. This part of the radionuclide inventory is quantified by
multiplication of the total inventory by an instant release fraction, or IRF, so-called
because the release of this part of the inventory to solution is cautiously modelled as
instantaneous. The remainder of the inventory is assumed to be uniformly mixed within
the fuel matrix and structural materials, and to be released congruently with their
degradation over time.
The water inside the canister is groundwater conditioned by the surrounding buffer.
Different types of groundwater are considered (e.g. brackish, saline or glacial)
according to assumptions inherent in each scenario. In most of the cases, the
geochemical conditions and flow distributions are assumed to be time invariant.
However, this approach is complemented by calculation cases in which the evolving
geochemical conditions and flow distribution (in the variant and disturbance scenarios)
are modelled explicitly (albeit in a simplified manner). The explicit modelling of time
dependency of flow and geochemical conditions takes account of the adverse possibility
that radionuclides may accumulate within the repository system under one set of
116
conditions, and then be relatively rapidly transported to the biosphere when conditions
change, for example in connection with permafrost episodes.
Once released, radionuclides are assumed to dissolve in the water in a canister or to
precipitate if their respective solubility limits are reached. Concentrations of all isotopes
of the same element, including stable isotopes, are taken into account in determining
whether this is the case. The main retention and transport processes in the near field are
assumed to be solubility limitation, sorption, diffusion and, in the case of the deposition
tunnel and its EDZ, advection. In the majority of calculation cases, radionuclides diffuse
from the internal void space of the canister and through the surrounding buffer, retarded
by sorption, exiting the deposition hole by three possible paths (Figure 5-6):

the F-path, which leads from the canister, through the buffer to a host-rock fracture
intersecting the deposition hole;

the DZ-path, which leads from the canister to the deposition tunnel EDZ, either
directly through the buffer or via a damaged zone, which is assumed to surround the
deposition hole, and thence to a host-rock fracture intersecting the EDZ; and

the TDZ-path, which leads from the canister, through the buffer to the deposition
tunnel backfill, and thence to a host-rock fracture intersecting the deposition tunnel.
In cases in which the canister fails due to the presence of a penetrating defect, the
diffusive transport resistance of the defect is taken into account. In a few calculation
cases, the buffer is assumed to be eroded over time, in which case advection is assumed
to be the dominant transport process in the eroded part of the buffer.
Nearfield release, retention and transport modelling is performed using the transport
module of the GoldSim computer code. The calculated radionuclide release rates via
each the three paths listed above provide input to geosphere transport modelling.
117
Deposition tunnel
Deposition
Failure location
F-path
EDZ
Canister
hole
Water-conducting
fractures
Damaged zone
DZ-path
TDZ-path
Figure 5-6. Main features and physical dimensions of the near-field model in a vertical
section passing through the centre of a deposition hole parallel to the deposition tunnel
axis (blue arrows denote the water flow in the deposition tunnel backfill).
The main transport process in the geosphere is assumed to be advection along fractures,
retarded by matrix diffusion and sorption (Figure 5-7). The transport paths and their
properties − flow rate, the fraction of the flow path belonging to a specific fracture type
and discharge location are based on the groundwater flow modelling supporting the
analysis of the radionuclide releases scenarios. The fracture types are defined based on
fracture data from the Olkiluoto site and consist of:

clay (and possibly sulphide) coated fractures;

calcite (and possibly clay and sulphide) coated fractures;

slickensided19 fractures; and

other fractures.
19
Slickenside is a polished fault surface formed by frictional wear during sliding, but now used to denote any of several types of
lineated fault surfaces.
118
These fracture types differ in the characteristics (thickness, porosity and effective
diffusion coefficient) assumed for the fracture coatings and adjacent rock matrix layers,
and hence in the degree of retention by matrix diffusion and sorption experienced by
migrating radionuclides.
The definition of the sorption parameters is discussed in Models and Data for the
Repository System. The values used in the assessment are selected from the available
sorption data for Olkiluoto-specific rocks and groundwaters. The geosphere transport
modelling provides, as output, release locations and release rates to the biosphere, i.e.
the geo-bio fluxes, which are used as input in the biosphere assessment. Geosphere
transport modelling is carried out using either GoldSim (for stochastic simulations and
for simplified modelling carried out for the rock shear scenarios RS and RS-DIL) or the
MARFA code (for other deterministic calculation cases). Stochastic simulations have
been carried as part of the probabilistic sensitivity assessment to study the sensitivity of
the model output to variations in the input parameter values. The analysis considered the
total release rates from the near field to geosphere and from geosphere to biosphere
(summed over the three paths, see above) and applied Monte Carlo methods, graphical
methods and variance decomposition methods (see Cormenzana 2013a for details).
Figure 5-7. The main retention and transport processes in a water-conducting feature.
119
Table 5-4. Main models and modelling tools used for the analysis of radionuclide
release scenarios for the repository system.
Models and codes
Purpose and use
Models for the analysis of radionuclide release scenarios for the repository system
Models take into account the following processes: radioactive decay, radionuclide release, solubility
limitation, sorption, diffusion and advection (and dispersion in the geosphere, see Figure 5-6 and Figure
5-7). The codes applied are summarised below. The activity fluxes provided as output are used to address
compliance with the regulatory requirements or are used as input for the biosphere assessment.
Main reference: Assessment of Radionuclide Release Scenarios for the Repository System.
GoldSim
The transport module of GoldSim is used for analysis of the near-field release,
retention and transport, also for geosphere retention and transport modelling in
stochastic simulations and in the simplified geosphere modelling carried out for
the rock shear scenarios RS and RS-DIL. The GoldSim near-field model
addresses radionuclide release from a failed canister; and radionuclide transport
through the repository near field to the geosphere resulting in the release from
the near field to the geosphere The input to near field modelling includes
radionuclide inventory and half lives, amount of fuel in the failed canister, fuel
dissolution rate, corrosion rates of zirconium alloy and other metal parts, flow
properties around the deposition hole containing the failed canister, properties of
the canister defect (if any), density, porosity and and elemental solubilities,
diffusion coefficients and distribution coefficients in the buffer and backfill
dependent on the assumed groundwater composition and flow rates according to
the CONNECTFLOW model.
The geosphere transport model implemented in GoldSim addresses the
migration of radionuclides from each of the entry points (see Figure 5-6) through
the geosphere fracture network. Inputs to geosphere modelling include the
transport resistance of the geosphere (WL/Q) and the porosity, elemental
distribution coefficients and diffusion coefficients in the fracture coatings and
rock matrix.
Flow-related parameters (e.g. near-field flows and WL/Q) are, in general, based
on the results of groundwater flow modelling using CONNECTFLOW (see Table
5-2).
The main output from near-field modelling is the radionuclide release rates to the
geosphere along the F-, DZ- and TDZ-paths. The main output of geosphere
modelling is the spatially-integrated radionuclide release rates to the biosphere
(the geo-bio fluxes).
MARFA
MARFA is used for analysis of radionuclide retention and transport in the
geosphere in most of the deterministic calculation cases.
The geosphere transport model implemented in MARFA addresses the migration
of radionuclides from each of the entry points (see Figure 5-6) through the
geosphere fracture network. The variation of the geosphere transport resistance
(WL/Q) along the migration paths is read directly from the result of groundwater
flow modelling (CONNECTFLOW, see Table 5-2). The release rates from the
near field are based on GoldSim calculations. Other inputs are the same as
described above for geosphere retention and transport modelling using GoldSim
The main output of geosphere modelling is the spatially-integrated radionuclide
release rates to the biosphere (the geo-bio fluxes) and also the release locations
associated with the F-, DZ- and TDZ-paths.
5.4
Models and data for the biosphere assessment
An overview of the biosphere assessment process is given in Figure 5-8. The biosphere
description sub-process includes environmental studies and monitoring at the site,
compilation of a scientific synthesis of the current state of the surface environment at
the site (Biosphere Description) and production of site- and regional-specific data for
the safety analysis (Biosphere Data Basis). Formulation of scenarios defines the lines of
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Site
characterisation*
Formulation of
scenarios
Geosphere
modelling
Geosphere Geosphere Climatic
modelling
modelling conditions
Biosphere
description
Climatic
conditions
evolution and the calculation cases to be addressed. In the following, the modelling
carried out within the four remaining sub-processes, surface environment development,
screening analysis, landscape modelling and radiological impact analysis is discussed. A
summary of the models and codes used in the biosphere assessment is given in Table 55. The radionuclide release and transport models, the main supporting models and
information flows between these models are shown in Figure 5-4 and the codes applied
to the analysis of radionuclide releases and transport in Figure 5-5.
Surface
environment
development
Screening analysis
Landscape
modelling
Radiological
impact analysis
• Environmental studies and monitoring
• Past & future* development
• Current state of the surface environment
• FEPs* and narrative lines of evolution
• Site & regional data compilation
• Credible lines of evolution
• Scenario drivers *
• Formulation and classification of scenarios
• Definition of calculation cases
• Terrain evolution extrapolated from past development
• Typical succession lines and resource driven land use
• Conceptual models for elemental circulation and accumulation
• Projections of the surface environment development
• Surface and near-surface hydrological modelling
• Stylised representation of the surface environment
• Highly cautious transport and dose models*
• Highly cautious selection of parameter values
• Screening out radionuclides for further assessment
• Simplified representation of the surface environment
• Hydrologically connected biosphere objects
• Cautious transport models*
• Time-dependent radionuclide-specific activity distributions
• Maximum use of local resources
• Doses to each individual (full dose distributions)
• Typical absorbed dose rates to plants and animals
• Annual doses to most exposed and other people
* Experience and knowledge from previous iterations of the biosphere assessment important
Figure 5-8. The biosphere assessment process, its sub-processes (blue boxes) and key
features and products of each subprocess. The green boxes show main inputs from other
activities in the TURVA-2012 safety case.
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5.4.1
Development of surface environment
The objective in this sub-process is to project the future development of the surface
environment by extrapolating the past regional development, using knowledge of the
present conditions in the surface environment and the understanding of relevant FEPs.
The set of projections produced in the present assessment is intended to bound the
credible trajectories of the future, from a radiation exposure of humans, plants and
animals perspective.
The future development of the surface environment is assessed using terrain and
ecosystems development modelling (TESM, see Terrain and Ecosystem Development
Modelling for details) and surface and near-surface hydrological modelling (SHYD, see
Surface and Near-surface Hydrological Modelling for details) (see Figure 5-8 and Table
5-5). In the TESM, land-uplift-driven changes and other changes in the surface
environment are simulated, until and beyond the time when the potential releases would
reach it. The projections use typical succession lines for the development of sea bottom,
shoreline, forests, mires, lakes, small water bodies and, rivers. Humans are assuming to
use the landscape based on the resources it provides and on needs, typically taking the
most suitable (in terms of profit, accessibility, etc.) first resources into use. This gives a
link to the exposure pathways that are defined to represent present-day human habits.
Natural developments and changes in how humans use the land result in a projection
containing distinct biotopes, in which fauna find their habitats. The food web or the
structure of the biotic community can then be outlined and representative species and
their habits identified.
A sub-set of the projections produced by TESM is selected for propagation to landscape
modelling. The biosphere objects relevant to these projections are identified and
characterised. A biosphere object describes a continuous and reasonably homogeneous
segment of the modelled area into which radionuclides may be released. The
contamination can take place either by direct release of radionuclides from the
geosphere or by horizontal transport of radionuclides within the surface environment
during the dose assessment time window. Each biosphere object is characterised by one
or more biotopes and a set of object-specific parameters. The biotopes can be divided
into two main sub-groups: terrestrial and aquatic biotopes, refined further, e.g. to forest,
cropland with varying products, rivers and open sea.
The SHYD model is a tool that can be used to study the water balance components at
the Olkiluoto site. The model links the unsaturated and saturated soil water in the
overburden and groundwater in the bedrock as a continuous system. The fluxes for the
biosphere assessment are calculated in two steps: first steady-state recharge/discharge
to/from bedrock is computed for each time step; and in the second step vertical and
horizontal fluxes are computed for each delineated biosphere object. These fluxes are
averages for the specific biosphere objects from the results of the full 3D-model. A new
feature of the SHYD modelling compared with previous assessments is that (shallow)
wells (both those dug in the overburden and drilled in the bedrock) can be added as sink
points in the computational grid.
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5.4.2
Screening analysis
Experience from previous assessments shows that the calculated individual
radionuclide-specific releases from the geosphere span several orders of magnitude,
both in activity and radiotoxicity, and that, ultimately, the potential radiological impact
is dominated by a small number of radionuclides, as commonly identified through
cooperation in BIOPROTA20. The screening models are implemented with a high
degree of conservatism to ensure that the calculation results undoubtedly overestimate
any potential radiological impacts. The screening analysis applied is in line with a
graded approach for safety assessments (IAEA 2009, Requirement 1). The graded
approach is implemented by employing a three-tiered modelling approach for
radionuclide transport and dose calculations, where the complexity and realism are
greater for higher tiers than for lower tiers. Tier 1 and Tier 2 (described below) are
generic radionuclide screening analyses requiring a minimum of site-specific data. Tier
3 is based on landscape modelling (Section 5.4.3) and the radiological impact analysis
described in Section 5.4.4. The aim of the screening analysis is to identify radionuclides
that will – with high confidence − have insignificant radiological impact, and hence can
be screened out from analysis with the landscape model. The screening analysis is
described in more detail in Biosphere Radionuclide Transport and Dose Assessment
Modelling.
Tier 1. Radiotoxicity screening analysis
In Tier 1 it is assumed that a hypothetical person is exposed, via ingestion, to the total
activity released from the geosphere during the whole dose assessment time window.
Hence, this hypothetical person is exposed to the entire radiotoxicity of the released
activity, which is unarguably very cautious.
Tier 2. Generic biosphere model screening analysis
The model (Section 5.1.5 of Biosphere Assessment) used in Tier 2 analysis has a higher
degree of realism than the Tier 1 model, but is still sufficiently cautious for screening
purposes. It contains a set of mainly generic ecosystem-specific sub-models, similar to
the screening models proposed in IAEA (2001) that may be used to determine through
a simplified but cautious assessment whether a radionuclide can be neglected from
further consideration. The sub-models in Tier 2 comprises a water well, a lake, a forest,
an irrigated cropland and an irrigated pasture land. Exposure pathways considered are
ingestion of radionuclides in water, milk, crops, livestock meat, game, mushrooms and
berries, inhalation of radionuclides and external exposure from radionuclides in the
ground. The screening decision for a specific radionuclide is based on the sub-model
resulting in the highest calculated dose.
5.4.3
Landscape modelling
The landscape model is a state-of-the-art, time-dependent and site-specific radionuclide
transport model that takes the properties of the dynamic site into account.
20
International collaboration forum which seeks to address key uncertainties in the assessment of radiation doses in the long term
arising from releases of radionuclides as a result of radioactive waste management practices (www.bioprota.org).
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The landscape model is constructed from the biosphere objects delineated in the surface
environment development sub-process (Section 5.4.1). Defining the initial state for the
landscape model and how it develops with time is the landscape set-up activity, which is
the task interfacing the surface environment development and the landscape modelling
(see Figure 5-4). The connections between the biosphere objects are derived from
terrain projections for the period from the present (initial state) to the end of the dose
assessment time window. These connections determine how the biosphere objects are
hydrologically connected, i.e. if one object is upstream or downstream of another.
Each ecosystem type in each biosphere object is associated with a deterministic
radionuclide transport compartment model (see Figure 5-9, left). An important feature
of the landscape model is that it represents ecosystem succession, by allowing
transitions over time from one ecosystem type to another. The underlying radionuclide
transport models within biosphere objects must be consistent. Thus, it is important to
ensure: 1) that the water fluxes between biosphere objects can be handled in a
continuous manner, and 2) that the activity content of a specific compartment in a
specific biosphere object can be transferred to a corresponding compartment when an
ecosystem type develops into another type. Within the landscape modelling concept, the
biosphere objects form aggregates which are called ‘super-objects’ (Figure 5-9, right).
The ‘super-objects’ are within landscape zones delineated so that their area does not
change with the landscape development, i.e. the transport of radionuclides due to
change of the ecosystem types is confined within the given zone. This facilitates the
maintenance of mass balance in the model and prevents transitional effects occurring as
landscape changes.
Terrestrial blocks
Mire
Upland forest
Cropland
irrigated
Cropland
Non-irrigated
Aquatic blocks
Reed bed
Open water
Out
Figure 5-9. Left: the common structure for modelling any ecosystem type in a biosphere
object. Right: The general ecosystem type structure of ‘super-objects’.
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The main output of the landscape modelling sub-process is time-dependent
radionuclide-specific spatial activity distributions in all biosphere objects for the
analysed calculation cases. These are the key input to the radiological impact analysis.
5.4.4
Radiological impact analysis
In the radiological impact analysis, the spatially distributed, time-dependent
radionuclide-specific activity concentrations in environmental media, produced by
landscape modelling, are used to calculate the potential radiological impact, in terms of
annual doses to humans and absorbed dose rates to plants and animals. The calculated
annual doses and absorbed dose rates are assessed against the radiation protection
criteria defined by the STUK (see Table 1-1).
Assessing doses to humans
The regulations require assessments for two exposed groups: annual dose to the most
exposed people and average annual doses to other people. The dose assessment method
is based on the deterministic approach developed and applied in Broed et al. (2007) and
further refined in the interim safety case (Hjerpe et al. 2010). This concept has been
further developed for this assessment, to include consideration of shallow drinking
water wells and consumption of farm animal products. It is based on the guidance given
by the ICRP (2006) on assessing the dose to the ‘representative person’ (see Biosphere
Radionuclide Transport and Dose Assessment Modelling for details).
The key information needed for the dose assessments is radionuclide concentrations in
environmental media, such as soils, sediments, surface waters, well water and air. This
information is derived from landscape modelling. Production rates of individual food
products are also needed, to derive the total productivity, and concentration ratios of
radionuclides from environmental media to foodstuffs.
The exposure characteristics are mainly based on site-specific conditions, regional land
use, and present-day behaviour of the regional population. In long-term safety
assessments of geological disposal facilities it is common that a few exposure pathways
dominate the calculated doses. The following pathways are considered in the modelling:
ingestion of foodstuffs from aquatic, terrestrial and agricultural ecosystems, ingestion of
drinking water from freshwater surface waters and wells, inhalation of airborne
contaminants and external radiation from radionuclides in the ground.
The dose calculations for foodstuffs are based on values of food energy intake, for
individuals making the greatest reasonable use of the most contaminated local
resources. The dose coefficients for ingestion and inhalation are values recommended
for adults (ICRP 1996). Calculations of external radiation doses from contaminated soil
and sediments use dose coefficients for radionuclides uniformly distributed to an
infinite depth (based on Table III.7 in EPA 1993, extracted using the software
Radiological Toolbox21).
21
U.S. Nuclear Regulatory Commission Radiological Toolbox, (version 2.0.0, August 2006) (www.nrc.gov/aboutnrc/regulatory/research/radiological-toolbox.html)
125
The present assessment calculates landscape doses, which are pathway-, radionuclideand biosphere object-specific annual doses to a person. These are then combined to
obtain an annual landscape dose, by summing the dose maxima from each pathway.
Then the annual landscape dose is calculated for each exposed person to yield a
distribution of doses called the dose distribution. It is assumed that the exposed
population is no larger than the size of the present-day population of Eurajoki
municipally (about 6000 persons); unless the calculations result in that the production of
contaminated foodstuffs may support an even larger population, then the population size
is increased. Furthermore, it is assumed that the exposed population obtains all their
drinking water from contaminated water sources, consumes all contaminated foodstuffs
from the whole modelled area and inhabits all suitable residential areas. The dose
distribution is calculated for each generation using the contaminated area during the
assessment time window.
The dose distributions, derived for each generation using the contaminated area during
the dose assessment time window, are then divided into two groups: the most exposed
group and other people. The most exposed group is identified as a sub-group in the
exposed population that receives the highest doses; a size of 20 persons is considered
appropriate for this group. The group of other exposed people is then taken as the whole
exposed population, excluding the most exposed group. The average doses in these two
groups are then calculated and provide the annual doses to representative persons for the
most exposed people and for other people, as required by the STUK Guide YVL D.5.
Assessing doses to plants and animals
The first step in assessing dose to plant and animals is to identify representative species.
As new land areas will form over time as a result of continued post-glacial land uplift,
lake and inland mire biotopes not currently present at Olkiluoto will arise. Species
representative of such biotopes have been identified through Posiva’s survey and
monitoring program of reference lakes and mires (see Biosphere Description).
The process by which representative species have been selected for assessment and the
means by which they have been parameterised to allow dose calculations to be
performed are described in the following.
The diversity of plants and animals in terrestrial and aquatic ecosystems is such that it is
not possible to consider all species of plant and animal explicitly within an assessment
(Ulanovsky & Pröhl 2008). Simplification is required to allow dose implications from
the long-term releases from the proposed Olkiluoto repository to be evaluated. Such
simplifications are implicit within available assessment approaches and tools, including
the ERICA (ERICA 2007) and the ICRP Reference Animals and Plants (ICRP 2008)
approaches.
A sensitivity and knowledge quality assessment, as reported in Smith et al. (2010),
identified conceptual uncertainties associated with the application of the ERICA
assessment method to post-closure assessments. Based on the information in Smith et al
(2010), the following criteria were considered to be relevant to the selection of
representative biota for the Olkiluoto assessment:
 Species that are strongly identified with the reference area (or that can be reasonably
predicted to migrate into new biotopes as they are formed through land emergence
126
and natural succession – identified from the monitoring programme for reference
biotopes);

Species considered having a greater exposure potential, e.g., organisms with soildwelling habits, for which external exposure may be maximised; Organisms
occupying likely groundwater discharge areas, particularly water-body margins and
wetlands (mires), etc.

Representation of the different trophic levels as depicted in biotope-specific food
webs;

Species that are socioeconomically important and therefore of particular public
interest; and

Species for which there is reasonable knowledge relating to behaviour within
ecosystems and for which data on radionuclide transport were considered likely to
be available either from site-studies or the broader international literature (or for
which analogue data could reasonably be applied).
Each thus selected organism is represented by an ellipsoid of dimension and mass
commensurate with that organism. Ellipsoids are then used as a means of calculating
dose conversion coefficients (DCC) that enable absorbed dose rate to be calculated.
Two sets of DCC’s are applied:
 DCCext relates the activity concentration of a radionuclide in environmental media
(soil, sediment, water or air) to the external absorbed dose rate (µGy/h) received by
the organism in relation to its occupancy habits in the environmental media;

DCCint then provides the mechanism by which internal absorbed dose rate (µGy/h)
can be calculated in relation to radionuclides within the body of the organism.

The value of the DCC, both external and internal, is dependent upon the size and
geometry of the organism and its position relative to environmental media.
DCC’s for each representative organism were calculated using the ‘add reference
organism’ functionality within the ERICA assessment tool (see details in Dose
Assessment to Plants and Animals).
The approach to deriving typical dose rates for representative species is broadly
commensurate with that applied in deriving average annual dose to people. It uses the
DCC’s discussed above in conjunction with the same radionuclide concentrations in
environmental media as used in calculating doses to humans. Typical dose rates are
calculated by deriving an area-weighted dose rate for each representative species
exemplar throughout the contaminated biosphere objects in the modelled area. The area
weighted dose rate is computed for each 50 year assessment time-step (i.e. biotope
progression throughout the assessment timeframe is taken into account). For species
transient between different biotopes, the same approach is taken to calculate the area
weighted dose rate per biotope. The relative biotope occupancy is then applied to
calculate the dose rate according to time spent between the different biotopes.
127
Table 5-5. Main models and codes used for the biosphere assessment in safety case
TURVA-2012.
Models and codes
Purpose and use
Models and codes for the development of the surface environment
Projecting the development of the surface environment is implemented in two modelling activities: the
terrain and ecosystems development modelling (TESM) and the surface and near-surface hydrological
modelling (SHYD). In the TESM, land-uplift-driven changes and other changes in the surface environment
are simulated, until and beyond the time when the potential releases would reach it (see UNTAMO below).
The TESM and SHYD modelling are documented in detail in Terrain and Ecosystem Development
Modelling and Surface and Near-Surface Hydrological Modelling, respectively.
UNTAMO
UNTAMO is a multi-module geographical information system (GIS) toolbox
that has been developed for projecting the development of the terrain and
ecosystem during the dose assessment time window. Each module in the
toolbox addresses a certain aspect of the surface environment development:
land uplift and delineation of the sea area, surface water bodies, erosion and
deposition, terrestrial and aquatic vegetation and land use. The UNTAMO
toolbox is used together with the SHYD model. The terrain and ecosystems
are projected into future with UNTAMO and delivered as input data to the
SHYD model to simulate groundwater flow and water table characteristics in
detail; on the other hand, UNTAMO uses a simplified groundwater table
model derived from an earlier iteration with SHYD to identify areas of shallow
and deep groundwater table.
SHYD
The SHYD model estimates the movement and storage of water in the
radionuclide transport models of the surface environment, including
horizontal and vertical water fluxes in the overburden and at the ground
surface. The model uses a 3-D grid with various types of spatial and temporal
simplifications (conceptualisations) linking the unsaturated and saturated soil
water in the overburden and groundwater in the bedrock as a continuous
pressure system. Key data used are documented in Biosphere Data Basis
and summarised in Table 5-6 of Biosphere Assessment.
Models and codes for the screening analysis
In the screening analysis, a two-tiered (see text) approach is employed, in order to to identify radionuclides
that are highly confidently expected to have insignificant radiological impact, and hence can be screened
out from further analysis with the complex landscape model. The screening models are implemented with a
high degree of conservatism to ensure that the calculation results undoubtedly overestimate any potential
radiological impacts.
Ecolego
The screening analyses were carried out using the software package
Ecolego (www.ecolego.facilia.se), which is a simulation software tool used for
creating dynamic models and performing deterministic and probabilistic
simulations. The screening models are documented in detail in Biosphere
Radionuclide Transport and Dose Assessment.
Models and codes for the landscape modelling
The landscape model is a state-of-the-art, time-dependent and site-specific radionuclide transport (RNT)
model that takes the properties of the dynamic site into account. The main outputs of the landscape
modelling sub-process are time-dependent radionuclide-specific spatial activity distributions in all biosphere
objects for the analysed calculation cases. These are the key input to the next sub-process, the radiological
impact analysis. The landscape model is documented in detail in Biosphere Radionuclide Transport and
Dose Assessment.
Pandora
Pandora is a tool developed by Facilia AB for Posiva and SKB and used by
Posiva for radionuclide transport modelling in the surface environment.
Pandora is based on the Matlab/Simulink© environment
(www.mathworks.com). Pandora was developed to simplify development of
models resulting in large systems of differential equations where decay of
radionuclides is included in the model and to enhance the graphical user
interface to be more suitable for radioecological modelling.
Models and codes for radiological impact assessment
128
Models and codes
Purpose and use
The aim of the models for radiological impact assessment is to produce estimates for appropriate radiation
exposure quantities. Spatially distributed, time-dependent radionuclide-specific activity concentrations in
environmental media, produced by the landscape modelling, are used to assess the potential radiological
impact, in terms of annual doses to humans and absorbed dose rates to plants and animals, based on
present-day radiation protection criteria. The model to calculate annual doses to humans is documented in
detail in Biosphere Radionuclide Transport and Dose Assessment and the model to calculate absorbed
dose rates to plants and animals is documented in detail in Dose Assessment for Plants and Animals.
MATLAB
The dose calculations are performed using models implemented in Matlab
that utilise the environmental activity concentrations resulting from the
landscape modelling.
Annual doses to humans are derived for each exposed person individually
taking all relevant exposure pathways into account, such as: food ingestion
from crops, animal products, fish, game, mushrooms and berries, ingestion of
drinking water from wells and freshwater surface bodies, inhalation of
breathing air and external exposure from the ground.
Absorbed doses rates to plants and animals are derived for each
representative species exemplar occupying individual biosphere objects
throughout the modelled area.
ERICA
The ERICA Tool has a structure based upon the ERICA tiered approach to
assessing the radiological risk to other biota. The tool is here used to derive
dose conversion coefficients via external and internal radiation to reference
organisms characterised by default attributes relating to radioecology and
dosimetry. The key attributes are equilibrium concentration ratios, occupancy
factors, and ellipsoidal geometries.
129
6
PERFORMANCE ASSESSMENT OF THE REPOSITORY SYSTEM
This chapter summarises the performance of the repository system and the fulfilment of
performance targets and target properties for the engineered barriers and host rock as
identified in Table 2-2 and Table 2-3. The performance assessment takes account of the
expected thermal, hydraulic, mechanical and chemical (THMC) evolution of the
repository system, and uncertainties in the expected lines of evolution. Unlikely lines of
evolution, including the possibility of disruptive events, are also identified. Account is
taken of the natural evolution of the environment, chiefly driven by climatic evolution,
which imposes external loads on the repository system, and also internal loads, chiefly
from the effects of excavation and emplacement of the spent nuclear fuel and the
engineered barriers.
The assessment considers the different evolutionary processes that potentially can affect
the performance targets and target properties for the three time windows; the period of
excavation and operation up to closure, the post-closure period during the next 10,000
years and beyond 10,000 years over repeated glacial cycles up to million years.
The fulfilment of performance targets and target properties in each time window is
assessed considering time-dependent and space-dependent loads on the engineered
barriers and host rock. Quantitative assessments are made whenever possible, e.g. to
calculate safety margins and demonstrate the robustness of the design. Uncertainties are
highlighted, conditions that could lead to deviations from performance targets and target
properties are identified, and the likelihood and effects of the deviations estimated. In
particular, conditions and events (incidental deviations) that could lead to the release of
radionuclides are identified. These deviations from the desired initial state or expected
evolution are propagated to Formulation of Radionuclide Release Scenarios, which
defines the scenarios and the calculation cases for both the repository system and the
surface environment.
6.1
Excavation and operation up to closure of the disposal facility
The excavation and operational period of the underground disposal facility may
potentially affect the long-term repository performance, since the changes in the
thermal, mechanical, hydrological and chemical conditions induced by the excavation
and operational activities may affect the engineered barriers and the host rock in the
longer term. The duration of this period in various parts of the repository can be
assumed to be from several tens up to about one hundred years, depending on the
progress of the excavation and operational activities and the total number of canisters to
be disposed.
6.1.1
Repository system evolution and performance
Hydraulic and geochemical evolution of geosphere
Groundwater flow at Olkiluoto takes place mainly through a network of fractures and
deformation zones, within which channelling of the flow is likely. Thus, there will be
significant local variation of the flow conditions and possibly also of salinity and
groundwater composition near the deposition holes.
130
During the operational period, the flow rates are approximately two orders of magnitude
higher than before the repository construction (see Figure 6-1). Following installation of
tunnel and shaft backfill and seals (i.e. after closure), modelled flow rates return to near
pre-excavation values. During the operational period, the open tunnels draw water from
all directions so that the flow below the repository is upwards compared with the mainly
downward flow in the natural state. The upward flow is somewhat strengthened by the
heat produced by the spent nuclear fuel.
The correlation between the inflow and post-closure flow rate is not necessarily uniform
and may vary spatially. Thus it is possible that even if all the deposition holes with
inflow over the maximum inflow criterion to a deposition hole defined by RSC,
0.1 L/min, are discarded, it cannot be excluded that a few deposition holes will be
associated with a higher post-closure flow rate or a lower transport resistance than the
target values (see Table 2-3).
Figure 6-1. Flow rate into and out of the reference volume containing the repository
(Löfman & Karvonen 2012; 2009SH refers to a model variant with layout for 5500 tU
and the hydrogeological model described in the previous site description (Posiva
2009b) with semi-homogeneous (SH) hydraulic properties of the hydrogeological zones
(HZ) and sparsely fractured rock (SFR) i.e. for most of the HZs homogeneous
properties are used, whereas the SFR is divided into the depth intervals, in which either
depth-dependent or homogeneous values are used. 2011SH refers to a model with
layout for 9000 tU and the hydrogeological model according to Site Description with
the hydraulic properties of HZs and SFR described as in 2009SH, and 2011HE refers to
a model with layout for 9000 tU and the hydrogeological model according to Site
Description with the heterogeneous (HE) hydraulic properties of HZs and SFR
(Performance Assessment, Chapter 5).
131
Excavation will cause a damaged zone (EDZ) to form, especially below the tunnel
floors, although the damage is unlikely to form a continuous hydraulic pathway along
the tunnel length (Site Description, Section 11.2.4). In addition, excavation, and later
the heat produced by the spent nuclear fuel, may cause spalling or other types of stressinduced damage around the excavated rooms, deposition tunnels and deposition holes.
There are minor uncertainties related to the extent and properties of the EDZ and the
stress-induced rock damage. These uncertainties are taken into account in the
groundwater flow modelling. The potential presence of an EDZ affects inflows below
1 mL/min and means that all deposition holes have some, although in many cases very
limited inflow compared with 40 % of deposition holes with no inflow when there is no
EDZ present.
During the operational period, the average salinity around the repository remains similar
to the pre-construction phase and changes are rather moderate (Figure 6-2). However,
increased groundwater flow into the repository volume may lead to mixing of water and
to either more dilute or more saline conditions locally. However, even under pessimistic
assumptions, the maximum salinities in the reference volume are expected to remain
below 70 g/L and salinities over 35 g/L are not expected at repository depth. Most of the
modelling results suggest that the lowest salinities during the operational period will be
at least a few grams per litre in most parts of the repository. However, the possibility of
salinities close to 0.3 to 0.4 g/L (which corresponds roughly to the minimum acceptable
total charge equivalent of cations of 4 mM) cannot totally be excluded. Such low
salinities are obtained only under pessimistic assumptions with infiltrating water of zero
salinity and neglecting the geochemical and hydrochemical reactions in the overburden
and along the recharge paths.
The lowest and highest salinities discussed above are related to the main
hydrogeological zones in the ONKALO and would not necessarily occur in the
repository panels themselves, which avoid hydrogeologically active zones. Moreover,
the disturbed conditions of either low or extremely low salinities are likely to last a
limited time − in the order of a few tens of years, and thus the impact on the
performance of the buffer and backfill is limited. In summary, in spite of the rather large
variations in the flow conditions during the excavation and operational period, the
groundwater composition with respect to salinity, chloride content and total charge
equivalent of cations will remain within the target range meaning that the buffer and
backfill functions are fully preserved, except at a few canister positions.
Organics, pyrite and other sulphides in the overburden are able to consume oxygen in
infiltrating waters by microbially mediated reactions. Thus, oxygen is not expected to
penetrate more than a few tens of metres along fractures and is very unlikely to reach
repository depth. The pH in the infiltrating water is neutralised by calcite. The pH in the
natural groundwater at the repository depth is expected to be in the range of 7 to 9 and is
thus well within the range defined by the target properties.
132
Figure 6-2. Salinity (TDS) evolution during the excavation and operation period, and
after closure until 50,000 years after the start of construction of the ONKALO; the
maximum, minimum and average salinity in the reference volume for model variants
2009SH, 2011SH and 2011HE, see Figure 6-1 for explanations (from Performance
Assessment, Chapter 5).
The Olkiluoto groundwater has a naturally low colloid content. Colloids may be formed
by cement or silica sol degradation. However, as the groundwater at the repository
depth generally has such a high ionic strength (salinity), such colloid formation is
expected to be limited.
The sulphide concentration for the main water types in the natural (undisturbed) state is
well below 1 mg/L due to the control of sulphide concentration by iron in the
groundwater forming iron sulphides, which have low solubility. It has been observed
that site characterisation activities and ONKALO construction have caused artificially
disturbed transient conditions due to mixing of different groundwater types and
anomalous sulphide levels have been measured (max. 12 mg/L) at a depth of around
300 m. The high concentrations of sulphide are probably due to a delay in the
availability of iron; however, sulphide concentrations are still evidently controlled by
iron sulphide phases in the longer term. According to monitoring results, sulphide
concentrations decrease from the anomalously high values once the groundwater
conditions stabilise. Although the groundwater data clearly indicate sulphide values
well below 1 mg/L, a pessimistic upper value of 3 mg/L is used in Performance
Assessment, for the whole assessment period, which accounts for the possibility of
solubility control by the more soluble amorphous iron sulphide in combination with
kinetically-constrained availability of iron and the uncertainties related to microbial
activity.
133
Thermal evolution
The calculated temperature evolution at the canister surface and at the deposition hole
wall is illustrated in Figure 6-3 assuming either unsaturated or saturated buffer. The
maximum temperature at the canister surface is 95 °C at initial state, when there is a dry
air gap between the canister and the buffer and the buffer is unsaturated, and it is
achieved within about 20 years after canister emplacement. In the case of a saturated
buffer, the temperature at the canister surface is at most about 75 °C. The maximum
temperature in rock at the deposition hole wall is reached within about 40 years and is
about 65 °C. The results shown are for a canister located in the central part of the
repository, illustrating the maximum temperatures that will be reached. Thus, the
maximum temperatures after emplacement and during the operational period will
remain within the acceptable ranges.
Figure 6-3. Canister surface temperature estimates in the repository (central area) as a
function of time since emplacement assuming the two extreme saturation degrees for the
bentonite buffer, either unsaturated (buffer in initial condition) or saturated buffer. The
temperature evolution of rock at deposition hole wall (buffer-rock interface), which
does not depend on the degree of buffer saturation, is also shown. OL3 canister,
average burnup 50 MWd/kgU, separation between deposition holes and deposition
tunnels of 10.5 m and 25 m respectively, buffer conductivity is 1.0 W/m/K in the initial
condition and 1.3 W/m/K in saturated condition. In the initial condition, there is a
10 mm air gap between the canister and the buffer, in the saturated condition the gap is
closed. The outer 50 mm gap between buffer and rock is assumed to be filled with
bentonite pellets that have conductivity of 0.2 W/m/K in the initial condition and
0.6 W/m/K when saturated. Based on the results of Ikonen & Raiko (2013).
134
Rock mechanical evolution
Excavation and thermal load caused by the decay heat from the spent nuclear fuel will
cause some local damage, e.g. spalling, reactivation of fractures, within the near-field
rock. Although rock damage is not directly considered in the target properties, rock
damage around the excavated rooms, deposition tunnels and deposition holes may have
an impact on the hydraulic properties of the rock (as discussed above).
Excavation will cause a damaged zone (EDZ) to form, especially below the tunnel
floors, although the damage will probably not be continuous. Due to the uncertainties
related to the continuity and hydraulic properties of the EDZ and its impact on the
groundwater flow, variations in its characteristics have been studied using alternative
assumptions as to its properties in groundwater flow modelling. Although there is a
good understanding of the processes affecting the mechanical state of the rock in
general, there are still uncertainties related to the elastic and rock strength parameters of
the rock at Olkiluoto and especially concerning the in-situ stress state, which need
further study. The heterogeneity of the rock leads to spatial variations in rock properties,
which complicate the assessment of rock damage. To cope with the uncertainties as to
the existence of rock damage around deposition holes, two cases have been considered
in the groundwater flow modelling: i) a hydraulically significant rock damage zone
around the deposition hole and ii) no hydraulically significant rock damage around the
deposition hole. The hydraulic conductivity of the damage zone has also been varied.
Reactivation of fractures can change their hydraulic properties, but the changes in the
hydraulic properties are minor, especially when compared with changes caused by the
formation of the EDZ around the tunnels and rock damage around the deposition holes.
Mechanical and hydraulic evolution of buffer and backfill
Before full saturation, some buffer and backfill material may be lost through piping and
erosion. Based on the calculated inflow to deposition holes, roughly one third of the
deposition holes are such that some limited buffer mass loss is expected due to piping
and erosion. In a base case, considering the maximum inflow criterion of 0.1 L/min in a
deposition hole, the estimated mass loss is at most 185 kg (the total buffer mass in one
deposition hole is 20,300−24,300 kg). There are variant cases with larger losses, but the
average buffer density remains such that no significant changes are expected in the
hydraulic conductivity or the swelling pressure of the buffer and the necessary low
hydraulic conductivity and sufficient swelling pressure will be achieved as the buffer
saturates.
It is estimated that at most 13,000 kg of the backfill could be locally lost by piping and
erosion, but the eroded material would be redistributed within the deposition tunnel.
This is rather small compared with the total mass of backfill material in the tunnel
(more than 8000 tonnes in a 300 m long deposition tunnel). The effect on the backfill
performance depends on how the mass loss is distributed in the backfill. For example, if
all of the 13,000 kg were lost from a tunnel section of 1 m, the mass loss would have a
significant effect on the backfill density at this location. Such an event could perhaps be
possible in the vicinity of a fracture with a high enough inflow to transport all this mass
further down in the tunnel. It is also possible, after erosive mass redistribution is
completed, that homogenisation over time may mitigate the localised mass loss to some
135
extent. In any case such erosion would not be detrimental to the performance of the EBS
system, since no deposition hole would be located near such a fracture according to the
RSC. In conclusion, the buffer and backfill will maintain properties consistent with their
performance targets even considering the process of piping and erosion. The remaining
uncertainties of a situation with a significant loss of buffer in a deposition hole are
considered in the formulation of release scenarios.
Geochemical evolution of buffer and backfill
During saturation some limited dissolution and precipitation of salts in the buffer may
take place due to heat transfer from the canister. However, this is expected to have a
limited impact on the buffer.
Cementitious leachates from grouting of fractures, from grout used to stabilise rock
bolts and from the plug in the deposition tunnel may locally affect the backfill during
saturation due to degradation and leaching of cementitious materials. However, no
cement is in direct contact with the buffer and thus the impact on the buffer is expected
to be of little significance.
For both non-saturated and saturated conditions, the consumption of initially present
oxygen in the backfill and buffer will be relatively rapid (in the order of a few days to a
few years), due to its reaction with pyrite and other accessory minerals. Oxygen is also
partly consumed by the canister. Thus, anoxic reducing conditions will be quickly
established around the emplaced canisters and throughout the buffer and backfill.
Similarly to the natural groundwater colloids and cementitious colloids, the introduced
colloids through degradation of buffer and backfill materials are expected to be scarce
in the high ionic strength groundwaters during the operational phase and the postclosure period up until much later when the infiltration of dilute meltwater after an icesheet retreat has to be considered (see Section 6.3).
Mechanical, hydraulic and geochemical evolution of closure
There is a relatively short time (30−50 years) between the start of the emplacement of
the first closure components (by 2070) and the finalisation of the closure (by
2100−2120) and thus there should not be major uncertainties in their behaviour during
the operational period, provided that appropriate quality assurance measures are
followed.
Colloids generated by the degradation of closure materials (or other introduced
materials) will be unstable in the high ionic strength groundwaters at repository depth,
and hence of negligible impact.
Canister corrosion
At emplacement the canisters will be covered by a thin layer of corrosion products. The
presence of a few defective canisters in the repository cannot be ruled out at this stage
of technical development.
The maximum corrosion depth from the atmospheric and initially entrapped oxygen is
expected to be less than 0.5 mm, and will thus have a negligible impact on the minimum
136
thickness of the copper canister wall. Also, during the early aerobic phase in the
evolution of the repository environment, the period of susceptibility to stress corrosion
cracking is short and it is highly unlikely that the required conditions (redox potential,
porewater salinity, interfacial pH and SCC-inducing species) will be present
simultaneously. Residual stresses on the surface of the canister (at the weld location) are
currently being quantified.
Mechanical impacts on the canister
The impact from potential canister handling accidents is not a concern in the long-term
safety case, since if such accidents happen, the canister will be returned to the
encapsulation plant for examination and assessment. If necessary, the canister will be
opened and unloaded, and the fuel re-encapsulated in a new canister. To control the
canister surface condition, a final control point for surface damage will be established
just before the canister is lowered into the deposition hole.
Subcriticality
Criticality safety analyses are performed for transportation and encapsulation purposes
to ensure, with a high margin of safety, that the canister will be in a sub-critical state at
the time of emplacement. Criticality during the early evolution of the repository is not
possible because the spent nuclear fuel will remain in the same geometrical
configuration as in the initial state and no water (a neutron moderator) is present in the
canister. Criticality safety analyses show that, even if it is cautiously assumed that the
canister is filled with water, the spent nuclear fuel is expected to remain in sub-critical
condition (for OL3 canister, the use of burn-up credit is necessary). There are currently
no qualified methods for the use of burnup credit for geological disposal purposes but
work is ongoing at international level to address this issue.
6.1.2
Fulfilment of performance targets and target properties
As discussed above, for the expected evolution of conditions during the period of
operation and closure, the properties of the EBS and host rock will conform to the
performance targets and target properties at the end of the operational period with a high
degree of confidence.
There are some possibilities for incidental deviations, although none of these are
expected.

Even if all the deposition holes with inflow over the limit of 0.1 L/min are rejected,
it is not possible to exclude the possibility that a few deposition holes might
experience a higher post-closure flow rate or a lower transport resistance than the
target values. For a few canister positions, the groundwater composition with
respect to salinity, chloride content and total charge concentration of the cations
may, for a short time, be outside the target ranges. Any such deviations are expected
to disappear shortly after closure of the disposal facility and will have minimal
effects on the backfill, buffer or canister.

It is expected that all canisters leaving the encapsulation plant will be intact, but it
cannot be ruled out that a few canisters will have an initial penetrating defect that
escapes detection. As discussed in Section 3.5, at present, the probability of
137
detection can only be based on expert judgement, taking into account of the results
from both the non-destructive testing and destructive testing of the weld. With more
data becoming available in the future, it is likely that it will be possible to
demonstrate that the probability of emplacing more than one canister with an initial
undetected penetrating defect is less than one per cent.
Of these possible incidental deviations, only the last gives any possibility for release of
radionuclides from the EBS during the first several hundreds of thousands of years.
Therefore, the case of one or a few canisters with an initial penetrating defect is carried
forward to the formulation of release scenarios and assessment of radionuclide releases.
6.2
Post-closure evolution during the next 10,000 years
The climate during the next 10,000 years is expected to remain essentially as today, i.e.
a temperate climate associated with a boreal ecosystem. Crustal uplift will continue in
southern Finland, outpacing global eustatic sea-level rise. Thus, locally, relative sea
level will fall, and as a consequence, hydraulic gradients will increase during the first
few thousands of years after repository closure. At 1000 to 2000 years after present, the
shoreline will have retreated far enough that further changes will not affect the
hydraulic gradient nor alter the flow rates in the repository volume. Groundwater flow
and groundwater chemistry will recover from the disturbances caused by the excavation,
as discussed in Section 6.1, to return to conditions similar to those of the present day.
The groundwater flow is governed by the hydraulic gradients caused by the topography
and salinity field. The main impact on the groundwater composition will be due to
continued infiltration of meteoric waters. The main processes ongoing in the repository
during this stage will be water uptake, saturation, swelling and homogenisation of the
swelling clays in the buffer, backfill and closure and the gradual decline of the residual
heat in the spent nuclear fuel.
6.2.1
Repository system evolution and performance
Hydraulic and geochemical evolution of the geosphere
After closure of the disposal facility, the site will recover from the disturbances caused
by repository construction, operation and closure. Flow rates will reduce as the
drawdown caused by the repository decreases and saturation of buffer and backfill
occurs.
For the first hundreds of years after closure the heat produced by the spent nuclear fuel
increases the flow rates by a factor of 2 to 3 compared with the natural state and
enhances temporarily and locally upward flows. Generally the heat trends to result in an
upward driving force for the water, but when combined with the stronger natural
downward forces the flow remains still mainly directed downwards. The heat
production declines to very low levels after the first few thousands of years. Beyond
that time, the main factors affecting the hydrogeological and hydrogeochemical
evolution of the site will be the continued crustal uplift and the infiltration of meteoric
waters. After closure, the flow rates will reduce significantly due to reduced hydraulic
gradients. However, in deeper parts of the rock, including the rock around the
repository, the flow rates will remain somewhat higher than in the natural state before
138
construction, because the slow recovery of the salinity field will affect the flow field for
hundreds of years.
Discrete fracture network modelling carried out by Hartley et al. (2013b) provides
detailed information about the migration paths and flow around the deposition holes.
This information is used both to assess whether the target properties are met as well as
providing an input to the radionuclide release, retention and transport analysis. The
modelling studies quantify the impacts of the tunnel EDZ and the rock damage around
the deposition holes as key factors determining local flow rates and other flow-related
transport parameters. Figure 6-4 shows the variation of the cumulative distribution of
the flow rate (CDF) per unit width in the release location and flow-related transport
resistance for the three release path types (for release paths see Figure 5-6), based on
discrete fracture network modelling with particle tracking, with exit from the deposition
hole to:

a host-rock fracture intersecting the deposition hole (F-path),

the excavation damaged zone (EDZ) below the tunnel floor DZ-path) or

the tunnel backfill above the deposition hole (TDZ-path).
Different assumptions on the continuity of the EDZ and on the presence of rock damage
around deposition holes are considered, as well as the case of an open crown space in
the deposition tunnel.
The modelling of the impacts of the excavation damaged zone and the potential rock
damage around the deposition holes on the groundwater flow shows that the
connectivity of fractures and flow rate around deposition holes is indeed increased, but
the effect of the increased connectivity is limited to the deposition holes that are not
intersected by flowing fractures at all or are intersected by fractures with low flow rates
per unit width (less than 10-4 m3/(m·a)) and high transport resistance (higher than
500,000 years/m). The effects on the natural fractures are limited, however, and flow
rates in natural fractures and the transport resistances in the vicinity of the deposition
holes are consistent with target properties except for a few deposition holes.
The salinity evolution has been estimated based on the modelling by Löfman &
Karvonen (2012) and Trinchero et al. (2013).
The salinity evolution is shown in Figure 6-6 based on the two model variants out of the
three considered in the modelling by Löfman & Karvonen (2012). Depending on the
model variant, the modelling period covered either 10,000 years or 50,000 years, which
is the assumed duration of the temperate period before the next permafrost period. All
the three model variants resulted in a more or less similar behaviour of the salinity field.
Following the reduced flow in the repository volume and the recovery of the downward
flow direction, the salinity field recovers from the disturbance caused by the repository
excavation. The recovery of the salinity field is slower than that of the flow field. The
changes in salinity are faster within and close to the hydrogeological zones compared
with sparsely fractured rock, as can be seen from Figure 6-6. In the model variant with
heterogeneous properties of the sparsely fractured rock, there is more variation in
139
1.0
Base Case
0.9
No EDZ
fraction
0.8
Continuous EDZ
0.7
No Spalling
0.6
Crown Spacing
0.5
0.4
0.3
0.2
0.1
0.0
0
-1
-2
-3
-4
-5
-6
-4
-5
-6
3
log10(U)[m /m,a]
1.0
Base Case
fraction
0.9
0.8
Continuous EDZ
0.7
No Spalling
0.6
Crown Spacing
0.5
0.4
0.3
0.2
0.1
0.0
0
-1
-2
-3
3
log10(U)[m /m,a]
1.0
Base Case
fraction
0.9
No EDZ
0.8
Continuous EDZ
0.7
No Spalling
0.6
Crown Spacing
0.5
0.4
0.3
0.2
0.1
0.0
0
-1
-2
-3
-4
-5
-6
3
log10(U)[m /m,a]
Figure 6-4. CDF plots of initial flow-rate, U, for particles reaching the model top
boundary, with RSC inflow screening applied. Results are shown for QF (top), QDZ
(middle) and QTDZ (bottom) release paths for a number of variants with different
assumptions as to the EDZ and rock damage around the deposition hole as well as a
case assuming that the top part of the backfill has high conductivity (presence of a
crown space). In the base case, a discontinuous EDZ and rock damage around the
deposition hole are assumed. (Note: for the QDZ release path there is no CDF for the
No EDZ case.)
140
1.0
Base Case
0.9
No EDZ
fraction
0.8
Continuous EDZ
0.7
No Spalling
0.6
Crown Spacing
0.5
0.4
0.3
0.2
0.1
0.0
2
3
4
5
6
7
8
9
10
11
12
log10(Fr)[a/m]
1.0
0.9
0.8
fraction
0.7
0.6
0.5
0.4
Base Case
0.3
Continuous EDZ
0.2
No Spalling
0.1
Crown Spacing
0.0
2
3
4
5
6
7
8
9
10
11
12
log10(Fr)[a/m]
1.0
0.9
0.8
fraction
0.7
0.6
0.5
0.4
Base Case
0.3
No EDZ
Continuous EDZ
0.2
No Spalling
0.1
Crown Spacing
0.0
2
3
4
5
6
7
8
9
10
11
12
log10(Fr)[a/m]
Figure 6-5. CDF plots of flow-related transport resistance, Fr (=2WL/Q), for particles
reaching the model top boundary, with RSC inflow screening applied. Results are
shown for QF (top), QDZ (middle) and QTDZ (bottom) release paths for a number of
variants with different assumptions as to the EDZ and rock damage around the
deposition hole as well as a case assuming that the top part of the backfill has high
conductivity (presence of a crown space). In the base case a discontinuous EDZ and
rock damage around the deposition hole are assumed. (Note: for the QDZ release path
there is no CDF for the No EDZ case.)
141
Model variant 3 (2011HE)
Model variant 1 (2009SH)
Layout for 9000 tU and the hydrogeological
Layout for 5500 tU and the hydrogeological model
model 2011 with the heterogeneous (HE)
2009 with semi-homogeneous (SH) hydraulic
hydraulic properties of of the hydrogeological
properties of the hydrogeological zones (HZ) and
zones (HZ) and sparsely fractured rock (SFR)
sparsely fractured rock (SFR)
t = 1000 years
t = 10,000 years
t = 50,000 years
Figure 6-6. Distribution of the salinity at the repository level (Z = -410 m) at 1000,
10,000 and 50,000 years after the start of the disposal operations for model variants
2009SH and 2011HE (Löfman & Karvonen 2012 and Performance Assessment,
Chapters 6 and 7). Model variant 2011HE corresponds to the DFN models used in the
safety case.
salinity at the repository depth and, specifically, areas with a low salinity develop with
time. This model variant corresponds to the DFN models used in the safety case.
142
As the disturbances caused by repository construction cease, the groundwater
composition will stabilise and the variation seen during the operational period will
diminish. The few local values that were outside the target value range will return
within the range in a relatively short time. At repository depth, the pH will remain close
to 7.5 and reducing conditions prevail. During this time period, as a result of the
infiltration of meteoric water at a slow, nearly constant rate, a decreasing trend in
salinity, chloride and total charge equivalent of cations is expected. These values are,
however expected to stay within the limits of the target ranges. The groundwater flow
and transport modelling results indicate the possibility that a few canister positions may
experience dilute conditions immediately after closure. The sulphide concentrations in
the groundwaters after the post-closure period up to 10,000 years are expected to
recover towards the steady state conditions, and it is expected that the initially
controlling amorphous iron sulphide phases will successively evolve towards more
crystalline iron sulphide phases with a lower solubility. However, as noted in Section
6.2.1 a pessimistic sulphide concentration of 3 mg/L is adopted for use in subsequent
analyses of canister corrosion.
Thermal evolution of near field
The ability of the buffer to transfer decay heat from the canister to the rock will remain
sufficient to ensure the requirement of the maximum buffer temperature of 100 ºC is
respected regardless of the presence of air-filled gaps and uncertainties in the
mineralogical composition of the buffer. The better the buffer can transfer the decay
heat from the canister to the host rock, the lower the canister surface temperature will
remain.
Mechanical evolution of the rock
After the excavation and operational period and closure of the disposal facility, the rock
stresses in the near field will be affected by the swelling of the buffer and backfill and
by the thermal load from the spent nuclear fuel. There is a possibility of reactivation of
fractures and rock damage, most notably thermally induced spalling, which may change
the hydraulic properties of the near-field rock and thereby affect the target properties
concerning limited groundwater flow and high transport resistance in the vicinity of the
deposition holes. These factors are discussed above. Over time the thermal load will
decrease and stable conditions will be reached. No performance targets will be violated
due to the mechanical evolution of the rock after closure.
Mechanical and hydraulic evolution of the buffer and backfill
Groundwater flowing into the repository leads to saturation and swelling of the buffer
and backfill. The time to reach full saturation in the buffer is calculated as a few tens of
years to several thousands of years, depending on the local hydraulic conditions.
Initial differences in the density and swelling pressure in the buffer and backfill will be
evened out by homogenisation, although some heterogeneity will remain.
Homogenisation is a process that is not completely understood and the development of
numerical models will be continued. However, experiments and numerical assessments
(THM-modelling) show that homogenisation takes place in the buffer-pellet-rock
interface. Homogenisation has also been shown to take place in the backfill.
143
Initially, the low water content in the buffer will limit the microbial activity within it.
Later on, at full buffer saturation, the high density will limit microbial activity. At the
same time the buffer is impermeable enough to limit the mass flow to and from the
canister and thereby limit transport of harmful substances, radionuclides and colloids.
Calculations show that expansion of the buffer into the backfill and the changes in the
density of the buffer will not be sufficient to threaten the performance targets for the
buffer and backfill (i.e. a sufficiently high density will be maintained).
Geochemical evolution of the buffer
The complex thermo-hydro-mechanical-chemical (THMC) evolution during the thermal
period will lead to geochemical changes in the buffer, but these will have limited impact
on the performance targets. After saturation and development of the full swelling
capacity, the changes will be much more moderate and constrained by diffusive
processes.
The increased temperatures in the buffer will induce no or only minor montmorillonite
transformation (maximum 1 %) and very limited masses of newly cementing material
(< 2 % by volume).
The impact of cementitious leachates on montmorillonite transformation and porewater
chemistry during the temperate period has been assessed based on the amounts of
cementitious materials used in the repository. This effect has been estimated to be
negligible (Section 7.4.4 in Performance Assessment).
Initially, the production of sulphide via microbial processes in the buffer will be
inhibited by the low water content. After saturation, microbial activity will be restricted
by the high buffer density.
According to the results of groundwater flow modelling, the buffer at some deposition
holes could potentially be affected by dilute waters and chemical erosion for a short
period of time during the operational period and soon after closure. In these models, the
infiltrating water is assumed to have zero salinity and no reactions in the overburden or
along recharge path likely to increase the ionic strength of the infiltrating water are
taken into account. This case is assessed together with cases suggesting dilute
conditions and buffer erosion during a glacial cycle (Section 7.1 in Performance
Assessment).
Geochemical evolution of the backfill
The evolution of porewater chemistry in the backfill will be similar to that in the buffer,
but will be much less affected by the heat from the spent nuclear fuel (Section 7.4.4 in
Performance Assessment). Thus, any montmorillonite alteration and cementation due to
thermally-induced changes will be negligible in the backfill. With regard to
disturbances, the following conclusions can be drawn.
a) The degradation of cement materials in the deposition tunnel end plug contacting
the backfill will not affect the fulfilment of the performance targets of backfill
during the temperate period or afterwards. Disturbances due to leachates from
144
cementitious materials will diminish in general and also locally due to the lower
concentrations of the alkaline species in the leachates.
b) The corrosion of iron from construction materials will have an insignificant impact
on the performance targets of the backfill.
c) The large sulphate pool in the backfill is a potential source for microbial sulphide
production. In view of the large uncertainties related to backfill homogenisation and
microbial activity in the boundary areas, the sulphide fluxes that may affect the
canister can only be assessed by a bounding analysis. In the case of insufficient
homogenisation and areas of lower density (for example in the interface area with
the rock), sulphide may be produced by sulphate reducing bacteria. For the long
time perspectives considered, the sulphide formed is not expected to rise above the
sulphide levels calculated for mackinawite equilibrium (0.23−0.64 mg/L). For any
sulphide formed and even for the very pessimistic assumption that all of the
sulphate will eventually be reduced to sulphide, the main processes attenuating the
sulphide flux to the canister are slow diffusion transport, the precipitation of iron
sulphide and the advective loss to the rock mass.
d) In the case of good homogenisation, the high swelling pressure (high density) and
small pore size will effectively restrict microbial activity and the conditions in the
backfill will be similar to those in the (intact) buffer. If, however, low density areas
should persist, then significant sulphate reduction cannot be ruled out, and thus it is
considered in the canister corrosion analysis.
Mechanical, hydraulic and geochemical evolution of the closure
There are no major uncertainties in the evolution of the closure components during the
first 10,000 years after closure. Even if it is assumed that the hydraulic plugs will
become degraded, and some of the materials such as clays, aggregates and mixtures of
these may be eroded or suffer settlement, this is expected to be a limited effect and no
continuous preferential paths are expected to be formed. Therefore, at depth, transport
through closure components will be dominated by diffusion during the first 10,000
years.
Canister corrosion
Sulphide is the main copper corrosion agent after all oxygen has been consumed.
Microbially produced sulphide in the buffer is negligible in this period; sulphide supply
from the backfill is limited by the precipitation of iron sulphide and losses to the rock
mass, hence, the main source of sulphide is expected to be groundwater. Quantitative
corrosion calculations coupled with groundwater flow modelling have been carried out
(see Section 6.3 in this report and Section 7.7 Performance Assessment). These
calculations also take into account the possibility of early buffer erosion due to low
salinity, as mentioned above. Microbially produced sulphide in the buffer or in the
backfill is considered negligible during this phase. The calculations show that total
corrosion depth will be negligible during the first 10,000 years. The initially intact
canisters will remain intact for all conceivable loads that could occur during the first
10,000 years (see below) and thus the spent nuclear fuel will remain contained within
the canister.
145
Mechanical loading on canister
During the temperate period, the canister(s) will remain intact, i.e. meet all its
performance targets, for all conceivable loads (e.g. isostatic and uneven swelling
pressure and groundwater pressure) that could occur during this period.
6.2.2
Fulfilment of performance targets and target properties
As discussed above, for the expected evolution of conditions, during the post-closure
period up to 10,000 years after present, the properties of the EBS and host rock will
conform to the performance targets and target properties with a high degree of
confidence. The few possible exceptions on their own do not threaten the integrity of
the canisters.
All relevant FEPs and FEP interactions have been evaluated and considered in reaching
this conclusion. The key features and processes include:

residual heat transfers local to individual containers and throughout the repository;

heterogeneity of host rock and hence of groundwater flows and groundwater
composition;

long-term infiltration of meteoric waters;

continued buffer and backfill swelling and homogenisation, piping and erosion;

thermal and chemical reactions in the buffer and backfill (montmorillonite
transformation), alteration of accessory minerals;

localised and generalised corrosion of the copper overpack, due to sulphide.
Groundwater modelling provides a quantitative evaluation of coupled thermal, hydraulic
and chemical processes that accounts for the heterogeneity of host rock properties.
There are some possibilities for incidental deviations, although none of these are
expected. These incidental deviations are described below.

As found for the operational period, it cannot be excluded that a few deposition
holes might experience a higher post-closure flow rate or a lower transport
resistance than the target values.

Immediately after closure, modelling results indicate the possibility that a few
canister positions may experience dilute conditions such that chemical erosion of the
buffer could be possible. This result is considered to be due to simplified and
pessimistic model assumptions and does not reflect the overall understanding of the
likely future hydrogeochemical evolution of the site.

Homogenisation of the buffer and backfill should ensure a sufficiently high density
to restrict microbial activity; the conditions in the backfill will be similar to those in
the buffer. If, however, low-density areas should persist in the backfill, then sulphate
reduction to sulphide cannot be ruled out, and this is considered in the canister
corrosion analysis.
Of the above incidental deviations, none gives any possibility for release of
radionuclides on its own. Even combining the pessimistic assumption on sulphide
146
(3 mg/L) in groundwater, microbial reduction of sulphate to sulphide in the localised
parts of the backfill, and chemical erosion of the buffer in a few canister positions, the
calculated corrosion depth is not enough to lead to canister failure in the 10,000 year
time window.
The case of a canister with an initial penetrating defect can be further affected by some
of the above incidental deviations. These combinations are carried forward to be
considered in the formulation of scenarios and assessment of radionuclide releases.
6.3
Beyond 10,000 years during repeated glacial cycles
In the long term, i.e. over the next hundred thousand years or so, major climatic changes
are expected to occur, being part of a future glacial cycle. According to the climate
evolution described in Section 4.4, the current temperate period is assumed to last for
the next 50,000 years. This is followed by a new cycle with characteristics based on the
most recent glacial cycle. This implies alternating permafrost, warmer periods,
permafrost and glaciations up to about 170,000 years after present (170 ka AP), with
three glaciations, i.e. times during which the site is covered by an ice sheet between
about 90 and 155 ka AP.
Eight glacial cycles are postulated to occur in the next one million years. It is assumed
that the cycles will include similar varying conditions, and similar loads will be
imposed. After each glacial cycle a groundwater composition similar to that found at
present day is expected to re-establish, although at repository depth the changes will in
any case be minor and localised. The impact on the fulfilment of the performance
targets and target properties can thus be assessed by considering the consequences of
repeated loads similar to those assessed for the next glacial cycle.
Key effects are:

global sea-level fall, drier cold conditions and permafrost (ground freezing) and ice
sheet formation and retreat, leading to changes in groundwater flows and
composition; and

mechanical loading and unloading of the crust due to ice-sheet growth and retreat,
leading to rock stress changes and an increased likelihood of earthquakes during the
retreat phase.
These changes affect the mechanical and thermal evolution of the EBS and host rock.
6.3.1
Repository system evolution and evaluation of performance
Hydraulic and geochemical evolution of geosphere
Löfman & Karvonen (2012) discuss the groundwater flow and salinity evolution during
the continued temperate period until 50,000 years AP, under permafrost conditions and
during an ice-sheet retreat. The modelling has been carried out for selected time periods
from the reference climate evolution to represent permafrost and ice-sheet conditions.
Different model variants have also been used in the modelling. Further groundwater
flow modelling and reactive transport modelling during the ice-sheet retreat is presented
by Hartley et al. (2013b) and Trinchero et al. (2013).
147
During the continued temperate climate from 10,000 years AP until 50,000 years AP, a
slight increase is expected in the groundwater flow rates in the upper part of the bedrock
down to approximately 300 m depth. This increase is related to the increase of hydraulic
head due to the changes in the surface environment. Following the retreat of the
coastline, peatlands are expected to develop in the low-lying areas and the hydraulic
head in these areas will increase. The difference in hydraulic head between the area
above the repository and the discharge areas around the present island changes from
12 m (10,000 years AP) to 14 m (50,000 years AP). Flow rates at the repository depth
will, however, not be significantly affected.
The groundwater flow modelling under permafrost conditions has been carried out for
two representative distinct periods of the last glacial cycle (Figure 6-7). Both these
periods last about 10,000 years. During period 1, the permafrost reaches a depth of
approximately 80 m, and, during period 2, permafrost reaches around 300 metres depth.
The assumed permafrost depths are based on Hartikainen (2013). At the beginning of
the permafrost periods, the hydraulic conditions are taken as those at 10,000 years AP
due to the better reliability of the data and models for this period. Modelling cases with
and without taliks have been defined. It has been assumed that taliks may form in the
current sea area north, northwest and southwest of Olkiluoto Island (see Haapanen et al.
2010). Under permafrost conditions, the hydraulic conductivity of the rock is reduced
by several orders of magnitude (see Löfman & Karvonen 2012) and the infiltration is
very low. Consequently, the groundwater flow is significantly reduced (see Figure 6-7).
Modelling groundwater flow and salinity evolution during the retreat of the ice sheet
comprised three cases with an ice margin staying at Olkiluoto for 1000 years, but at
different locations with respect to the repository, and a case with a constantly retreating
ice sheet. In all simulations, the ice front was initially located at a position such that the
ice sheet overlay the whole of Olkiluoto Island. The retreating ice front then moved to
case-specific locations at a rate of 200 m/year. The salinity at the initial state for the
simulations of ice-sheet retreat period was assumed to be that at the end of permafrost
period 1 with taliks.
During the ice sheet retreat, the flow rates through the repository volume depend on the
location of the ice margin with respect to the repository (see Figure 6-8). While the
repository is still below the ice sheet (although the ice margin is close), the flow rates
are increased by a factor of 4 to 7 compared with the situation at the end of the
temperate climate or with the natural state before construction of the ONKALO.. In the
subsequent analyses, it has been assumed that the flow rates are 10 times higher during
the ice-sheet retreat than during the temperate time windows. The flow direction below
the ice sheet is mainly downwards and thus, during this time, the flow paths entering the
repository originate mainly from areas below the ice sheet. As the ice sheet passes the
site, the main flow direction is upwards and the flow rates reduce as the distance to the
ice margin increases. The site is likely to be submerged (below sea level) after the
retreat of an ice sheet. During the submerged period, when the ice margin is not in the
vicinity of the site, flow rates at the repository depth are lower than during the temperate
period as the only driving force is density variation.
148
Period 1
Period 2
Figure 6-7. Total flow rate to the reference volume under the two permafrost periods
considered (Löfman & Karvonen 2012). For explanation of 2009SH, 2011SH and
2011HE, see Figure 6-1. The permafrost development was from Hartikainen (2013), 1D
results were used in the groundwater flow modelling. Taken from Performance
Assessment (Chapter 7).
149
Figure 6-8. Total flow rate into the reference volume at the time of the ice-sheet retreat;
above mobile ice sheet retreating with a velocity of 200 m/year and below the immobile
ice sheet located northwest of the repository (Löfman & Karvonen 2012). 2009SH
model variant 1, 2011SH model variant 2, 2011HE model variant 3. Taken from
Performance Assessment (Chapter 7).
During the continued temperate period, the infiltration of meteoric water at a slow and
nearly constant rate results in a decreasing trend in salinity. The modelling results show
that, towards the end of this period, a few percent of the canister positions may
experience dilute conditions. As a result of the significantly reduced groundwater flow
during the permafrost periods, the groundwater salinities remain at the level prevailing
before the onset of the permafrost. Dilute conditions may also be experienced during the
ice-sheet retreat phase, but the estimate of the number of such positions is strongly
dependent on the duration of the melt water intrusion and especially on the modelling
concept of the interaction between the fracture water and the rock matrix. Set against
150
this, there is no evidence that fresh meltwater has reached repository depth at Olkiluoto
during the last glacial cycle or the previous ones. However, these uncertainties are
considered in scenario development, and lead to cases in which some deposition hole
locations are assumed to experience dilute conditions during the initial extended
temperate period and in association with glacial (ice-sheet margin) conditions. The
number of positions affected is influenced by the duration of the meltwater intrusion.
Similarly to the salinity results, other key geochemical properties (i.e. pH, Eh, Cl
concentration, total charge equivalent of cations, sulphur and iron species) are also
expected to stay within the limits established for the target properties during ice-sheet
retreat and melting for most of the deposition holes. Oxygen will be consumed within
short distances along the flow path and thus does not reach the repository level.
Freezing/thawing of buffer and backfill
The potential for freezing of the buffer or the deposition tunnel backfill is not an issue
since permafrost will not reach repository depth (Hartikainen 2006, 2013). Even if the
freezing front were to reach the repository depth, the materials, and design selected for
the buffer and backfill would withstand the freeze/thaw cycles without damage to their
safety functions (Schatz & Martikainen 2010, 2013).
The stability of clay materials against freeze/thaw cycles also implies that these
performance targets would also be upheld for more extreme climate evolutions.
Geochemical evolution of buffer and backfill
The evolution of porewater salinities in the buffer and backfill will follow those in the
surrounding groundwaters, which will remain within the required performance target
ranges, except perhaps during short times within the ice retreat and melting period.
Under these conditions, dilute groundwater conditions may cause some chemical
erosion of buffer and backfill and local lowering of density. In these deposition holes
and tunnel sections with lower density, sulphate reduction may occur. However, the
sulphide concentration in solution is still limited by the availability of iron in the
materials and the surrounding rock, which results in precipitation of iron sulphides,
keeping the sulphide concentration below 3 mg/L. Ongoing degradation of cementitious
materials will gradually release less aggressive leachates, whose effects on the clays in
the near field will generally be very limited. Changing hydraulic conditions and freezing
induced by glaciation effects however may locally increase the release of cement
leachates in closure components in the upper part of the rock.
The long-term stability of montmorillonite under repository conditions is difficult to
assess quantitatively because of lack of theoretical and experimental knowledge.
Nonetheless, kinetic constraints and especially observations from natural systems
indicate long-term stability of montmorillonite at low temperatures over a large range of
geochemical conditions.
Chemical erosion of buffer and backfill
Chemical erosion of buffer and backfill in some deposition holes and deposition tunnels
due to the potential occurrence of dilute groundwater cannot currently be ruled out for
short times during an ice retreat and melting period. With the reference assumptions on
151
groundwater flow (a selected realisation of the DFN flow model) and evolving
groundwater composition, one canister position is calculated to undergo buffer erosion
during the first glacial cycle to an extent that advective conditions arise. This calculation
should be seen as illustrative, being based on only a single realisation of the DFN
groundwater flow model. An analysis of statistical distributions of flow-related
parameter values between canister positions shows that measures such as the mean and
90th percentile vary little between DFN realisations. However, the number of canister
positions experiencing advective conditions is determined by the tails of these
distributions, and is therefore subject to more uncertainty. Taking a more cautious view
on this and other uncertainties, buffer erosion might result in advective conditions in a
few canister positions.
The modelling is subject to a number of uncertain including:

whether sufficiently dilute conditions are attained and for how long;

the groundwater flow distribution;

the extent of chemical erosion;

threshold values for buffer and backfill loss before advective conditions are attained;

the implications of application of RSC criteria.
The consequences, if any for canister corrosion, are considered in Formulation of
Radionuclide Release Scenarios.
Evolution of the closure components
After 10,000 years, the backfill in the central tunnels will be completely saturated and
keep its safety functions regardless of the consequences of climate evolution. For the
upper parts of the closure, the following is concluded.

It cannot be excluded that the backfill in parts of the access tunnel will lose its clay
components due to chemical erosion over the long time scale considered. However,
this is not judged to jeopardise the overall safety functions of the closure backfill in
particular and of the closure components, as a whole.

Degradation of closure plugs is uncertain, but the swelling clays used in the lower
parts of the tunnels and shafts will ensure sufficient isolation capacity of the sealing
structures.

Freezing/thawing of the closure components would not impair their performance
relative to the closure performance targets. The access tunnel and shafts between
depths of 200 and 300 m backfilled with an in situ compacted swelling clayaggregate mixture may, in the far future, be subject to freeze/thaw cycles, but this
will have no major implications for the performance of the material. The material
filling the access tunnel and shafts above 200 m depth is in situ compacted crushed
rock. If frost heave develops, it will be of minor consequence, if the crushed rock
material is selected appropriately. Glacial erosion may have an effect on materials in
the upper part of the disposal facility, but the erosion rate, even during glacial cycles
is so slow (average of 8 mm in 100 years; Okko 1964) that it would take several
millions of years to erode the upper plugs and the material underneath.
152
Rock mechanics and mechanical impacts on the canister
Large earthquakes may occur in connection with ice-sheet retreat. There is the
possibility that shear displacements exceeding 5 cm could cause canister failure in the
context of such earthquakes. Although, by locating the deposition holes away from large
deformation zones and avoiding large fracture intersections in deposition holes, it is
possible to decrease the risk of canister failure due to an earthquake, it is estimated that
few tens of canisters could be in positions such that they could potentially fail in such an
event. On the other hand, the average annual probability of an earthquake large enough
potentially to lead to canister failure is estimated to be low, in the order of 10-7, given
that there are around five fault zones within and around the area of the repository that
could host such an earthquake. Thus, during the first glacial cycle, the probability of
occurrence of such an earthquake is low and it is very likely that the canisters will
remain intact, i.e. meet all its performance targets, for all conceivable mechanical loads.
For the next four glacial cycles, the average annual probability of an earthquake large
enough potentially to lead to canister failure is of a similar order of magnitude as during
the first glacial cycle, even if an earthquake were to occur during the first cycle. This is
because, with a low tectonic strain rate, it takes a long time to compensate for the loss of
stored strain energy after the stress relaxation due to a single earthquake. Furthermore,
as fracture propagation is limited, even if another earthquake were to occur, the same
sub-set of fractures is likely to be reactivated. This, together with the low annual
probability of large earthquakes, keeps the annual probability of canister failure low.
However, over a one million year time frame, the possibility of some canister failures
by this mechanism cannot be excluded.
In case of a small rock movement, the toughness, plasticity, creep and relaxation
properties of copper ensure that the overpack (now in contact with the insert) does not
break in spite of the shearing force causing deformation of the canister.
Canister corrosion
If the buffer is intact, a sulphide concentration of more than 500−700 mg/L is necessary
to completely corrode the copper shell thickness of 49 mm (or 35 mm considering the
minimum required copper thickness) in one million years for the most unfavourably
located deposition hole. This concentration is more than two orders of magnitude higher
than the pessimistic bounding estimate for sulphide concentration in groundwater
(3 mg/L).
Figure 6-9 shows the distribution of corrosion depth over a one million year time period
in all deposition holes satisfying the inflow RSC criteria, and also with no application of
RSC, for three cases. The first two consider diffusion of sulphide directly across the
buffer from a fracture intersecting the deposition hole:
a) a groundwater sulphide concentration of 3 mg/L and the distribution of flow from
groundwater flow modelling assuming no hydraulically significant damaged zone
around the deposition hole. In this case, most canisters experience very limited
corrosion (less than 0.1 mm of corroded copper) over one million years even if the
RSC criteria are not applied.
153
(a)
(b)
(c )
Figure 6-9. Number of canister positions as a function of corrosion depth over 1 Ma,
with and without the RSC applied, in the case of direct diffusion of sulphide across the
buffer from a fracture intersecting the deposition hole with (a) no hydraulically
significant rock damage around the deposition hole (b) increased flow around the
deposition hole due to rock damage, and in the case of (c) downward diffusion of
sulphide from the deposition tunnel. In this case also the limiting corrosion depth for
high groundwater flow in the deposition tunnel is given. For all cases the assumed
groundwater sulphide concentration is 3 mg/L.
154
b) the same groundwater sulphide concentration but an increased flow around the
deposition hole due to rock damage. In this case, corrosion depths are somewhat
higher (but still very small, in the order of 0.2 mm); the application of RSC has a
limited effect in this case.
c) the third case assumes a downward diffusion of sulphide from the deposition tunnel,
again with a 3 mg/L sulphide concentration. In this case, the distribution of
corrosion depths is markedly higher, but still not sufficient to lead to canister
failures.
The results show that the overall corrosion depth will not exceed a few tenths of a
millimetre even over one million years. Thus, if the buffer performs as designed, no
canister failures are expected even with high sulphide concentrations. Furthermore, even
if the buffer is affected by chemical erosion, few if any canister failures due to corrosion
are expected during the first glacial cycle, as long as conditions otherwise correspond to
the expected evolution (i.e. performance targets and target properties are met). The
calculated rate of corrosion and the calculated number of canister failures in these
circumstances depends on the assumptions made about groundwater flow and
composition, corrosion area, fracture apertures, the rate of buffer erosion and the
possibility of locally thinner parts of the copper overpack. Cautiously assuming a
sulphide concentration of 3 mg/L in the groundwater, but with realistic assumptions
concerning these other factors, chemical erosion of the buffer and subsequent corrosion
by sulphide is calculated to lead to no canister failures within the first glacial cycle, and
4−5 failures in the million year time frame. Based on more cautious assumptions,
around 3 canister failures are calculated to occur within the first glacial cycle, and a few
tens of failures in the million year time frame.
No canister corrosion failures are expected during the first glacial cycle, even if the
buffer is chemically eroded, as long as the conditions correspond otherwise to the
expected evolution (i.e. performance targets and target properties are met). With
pessimistic assumptions concerning the intact wall thickness (35 mm) of the canister,
the corrosion area, fracture aperture, high flows and duration of dilute conditions,
chemical erosion of the buffer in a small number of boreholes and subsequent corrosion
by sulphide might lead to a few canister failures in the time frame of one million years
(assuming sulphide concentration of 3 mg/L in groundwater). The number of canister
failures depends on the assumptions about groundwater flows, fracture apertures and
buffer erosion modelling.
Thus, most of the canisters will remain intact for the time period up to one million years
and provide complete containment of the spent nuclear fuel. As discussed above, for
most of the deposition holes the performance targets for the buffer, and also the target
properties of rock, are upheld. Incidental deviations for some deposition holes may
result in dilute conditions causing chemical erosion of the buffer, allowing advective
conditions that may lead to a few canister failures by corrosion on a timescale of several
hundred thousand years. Copper corrosion by water in oxygen-free conditions is still
under investigation to understand some of the results published in the literature
(Feature, Events and Processes, 4.2.5).
155
Subcriticality
In the long term, after canister failure, the probability of criticality is also very low
because there are no credible mechanisms that could cause the redistribution of the
fissile material into a critical configuration either inside or outside the canister.
Although long-term criticality is not considered in the formulation of release scenarios,
this issue is still under investigation.
6.3.2
Fulfilment of performance targets and target properties
As discussed above, for the expected evolution of conditions during the next glacial
cycle, i.e. more than 100,000 years into the future, the properties of the EBS and host
rock will conform to the performance targets and target properties with some incidental
deviations over most of the repository and most canister positions with a high degree of
confidence, and most of the canisters will remain intact. Even in the one million year
time frame, the components mostly still meet their performance targets. The number of
deposition holes with incidental deviations with respect to groundwater flow target
properties, salinity and sulphide levels, and potential for shear displacements, are
essentially the same as after the first glacial cycle.
All relevant FEPs and FEP interactions have been evaluated and considered in reaching
this conclusion. The key features and processes include:

changes in groundwater flows due to changes in sea level, temperature and
precipitation, permafrost and ice sheets;

long-term infiltration of meteoric water and inflows of glacial waters;

freezing and thawing of closure components;

chemical erosion of buffer and backfill;

stress redistribution and earthquakes related to ice-sheet retreat;

long-term generalised corrosion of the copper overpack, due to sulphide.
Groundwater flow modelling provides a quantitative evaluation of coupled thermal,
hydraulic and chemical processes, including the impacts of heterogeneity of host rock
properties. There are some possibilities for incidental deviations from the target
properties related to the flow conditions, especially during the high-flow conditions
related to ice-sheet retreat; although none of these are certain and these deviations can
only affect a few canister positions.
Dilute conditions may be experienced in some deposition holes during ice-sheet retreat,
but the number of such positions depends on the duration of the melt water intrusion
and on the modelling assumptions on the diffusional mass exchange between the
fracture water and the rock matrix, dilute conditions are possible and more widespread,
if matrix diffusion is either not accounted for or is assumed to be limited only to a short
distance from the flowing fractures.
In spite of its geological stability, a large earthquake could potentially occur near the
site, especially in connection with ice-sheet retreat. The number of deposition holes that
could experience shear displacements on intersecting fractures large enough to cause
156
canister failure in such an event is kept low by locating the deposition holes away from
the large deformation zones and avoiding large fracture intersections in the deposition
holes. Nevertheless, and although the average annual probability of an earthquake large
enough potentially to lead to canister failure is estimated to be low, in the order of 10-7,
the possibility of a limited number of canister failures by this mechanism cannot be
excluded over a one million year time frame.
As long as transport of corrodants through the buffer remains diffusion dominated, no
canister failures by corrosion are expected within 1 million years, even in the least
favourable canister locations and even with pessimistic assumptions regarding
uncertainties in groundwater flow and diffusion of sulphide from the backfill. However,
chemical erosion of buffer due to the occurrence of dilute groundwater cannot be
neglected. The calculated rate of corrosion and the calculated number of canister
failures in the event that erosion leads to advective conditions in the buffer depends on
the assumptions made about groundwater flow and composition, corrosion area, fracture
apertures, the rate of buffer erosion and the possibility of locally thinner parts of the
copper overpack. Cautiously assuming a sulphide concentration of 3 mg/L in the
groundwater, but with realistic assumptions concerning these other factors, chemical
erosion of the buffer and subsequent corrosion by sulphide is calculated to lead to no
canister failures within the first glacial cycle, and 4−5 failures in the million year time
frame.
6.4
Summary statement of performance and uncertainties
The previous sections have summarised how the properties of the EBS and host rock are
expected to conform to the performance requirements over more than 100,000 years.
Thus no radionuclide releases are expected over more than 100,000 years for the
expected evolution of conditions, primarily because, protected by the buffer, the copper
canisters are designed and expected to withstand all likely events and conditions over
this period.
A full exploration of uncertainties and variability in initial conditions of the EBS and
host rock, and site evolution, has identified some events and conditions that could lead
to incidental deviations from performance targets and target properties (Table 6-1). The
importance of these has been assessed using different release scenarios.
157
Table 6-1. Summary of incidental deviations from performance targets and target
properties that may occur and are relevant in each time window.
Deviations
Up to
closure
of the
disposal
facility
Up to
10,000
years
During
repeated
glacial
cycles
Possibility of an initial penetrating defect in one or a few canisters.



Higher flow rate or lower transport resistance than the target
ranges for a few deposition holes.



Groundwater composition outside the target ranges for a short
time during operation and soon after closure for a few deposition
holes.


–
Low density areas in the backfill where sulphate reduction to
sulphide cannot be ruled out.
–


Erosion of buffer in some deposition holes due to long-term
infiltration of meteoric water or dilute glacial meltwater.
–
–

Canister failure by corrosion due to unfavourable groundwater
conditions and buffer erosion.
–
–

Canister failure due to shear displacements in fractures during
ice-sheet retreat
–
–

158
159
7
FORMULATION OF RADIONUCLIDE RELEASE SCENARIOS AND
CALCULATION CASES
A scenario represents one time history of conditions (hereafter called a line of
evolution) or more than one time history (lines of evolution). The performance of the
repository system and its components has been analysed taking into account the
expected lines of evolution and the uncertainties involved (Chapter 6). Although in
Performance Assessment it is shown that no radionuclide releases are expected during
the first 10,000 years after emplacement and not even within 100,000 years,
uncertainties in the initial state of the barriers, groundwater evolution and the
occurrence of unlikely events have been examined by scoping calculations in
Performance Assessment. If the scoping calculations show that the line(s) of evolution
may lead to the failure of one or more barriers (i.e. the failure of one or more safety
functions) such that radionuclides could be released, then a set of calculation cases
(reference case, sensitivity cases, “what if” cases) is defined within the scenarios to
analyse the impact of the potential radionuclide releases. This is documented in
Formulation of Radionuclide Release Scenarios and summarised in this chapter. The
analysis of these scenarios and calculation cases is documented in Assessment of
Radionuclide Release Scenarios for the Repository System and Biosphere Assessment,
and summarised in Chapter 8 of this report.
7.1
Lines of evolution framing the scenarios and scenario formulation
As described in Section 2.3.6, scenarios for radionuclide release calculations are based
on defining lines of evolution for the repository system and surface environment, and
consistent with the definition of scenario types – Base scenario, Variant scenarios and
Disturbance scenarios – given in Guide YVL D.5.
For the TURVA-2012 safety case, lines of evolution of the disposal system are defined
following the timeline given by the climatic evolution (see Section 4.4) selected based
on the recommendations by Pimenoff et al. (2011). The climatic evolution defines the
time windows in which climate-driven processes may operate. Processes internal to the
disposal system (whether or not driven by external events and associated changes in
external conditions) are also taken into account.
The lines of evolution that comprise the expected climatic evolution and frame the
disposal system evolution are as follows.
1. The climatic line of evolution takes into account a temperate period (i.e. boreal
climate) similar to the current one, which is assumed to last from today until about
50,000 years after present according to Table 3 in Pimenoff et al. (2011).
Thereafter, effects of present-era human effects are assumed to have subsided and a
return to glacial-interglacial cycling is expected; this is represented by assuming a
repetition of the Weichselian glacial-interglacial cycle (which lasted about 120,000
years). This is a stylised representation, since in the last million years none of the
glacial cycles has been a repetition of another, but it is a reasonable and evidencebased choice. There could be shorter or longer permafrost periods and less or more
extensive ice cover. Very pessimistic climate conditions would be required,
however, for permafrost to reach repository depth (Hartikainen 2013). There are
160
uncertainties in the timing and duration of periods of permafrost, and therefore only
specific time windows of permafrost occurrence have been selected for detailed
studies. The most reliably characterised ice-sheet retreat period at the end of the
Weichselian was also selected for detailed study.
2. The evolution of the geosphere22 and surface environment includes the
hydrogeological evolution (groundwater and surface water; coupled to thermal and
mechanical evolution in the context of groundwater evolution); the evolution of
surface water gives the boundary conditions for the hydrogeological evolution of
groundwater (see Löfman & Karvonen 2012 and Sections 5.1, 6.1 and 7.1 in
Performance Assessment). The hydrogeochemical evolution, i.e. variations in
salinity and groundwater composition due to gradual dilution with meteoric water
and possible intrusion of dilute glacial meltwaters, is coupled to the hydrogeological
evolution. The rock mechanical evolution is coupled to the thermal evolution and
also to stress changes due to glacial loading. All these couplings are discussed in
Performance Assessment.
3. The evolution of the EBS (i.e. canister, buffer, backfill and closure components) is
coupled to thermal, climatic, hydrogeological, hydrogeochemical and mechanical
evolution for the same time windows. The evolution of the canister (with no initial
penetrating defects)23 is coupled to buffer and backfill evolution and consequently
to climatic and geosphere evolution. Again, these couplings are discussed in
Performance Assessment.
7.1.1
The link to scenario hierarchy
The base scenario, as defined in Section 2.3.6, includes the expected, or most likely,
lines of evolution of the repository system taking into account external conditions (i.e.
climate evolution), internal phenomena, and human actions. For the repository system
this includes the incidental deviation whereby one or a few canisters with an initial
penetrating defect are emplaced in the repository, giving rise to releases of
radionuclides. All other canisters are manufactured and emplaced according to design.
Surface environment scenarios are formulated by adopting a concept of credible lines of
evolution. This is based on the expected line of climatic evolution and physical changes
in the surface environment, e.g. sea-level change and development of natural
ecosystems, but employs a stylised representation of future human activities based on
present-day habits, e.g. regarding land use.
Evolution lines within variant scenarios consider the diminished performance of the
safety functions of the canister and/or the combined effects of the reduced performance
of more than one safety function of the other barriers, still within the broadly expected
conditions. Variant scenarios for the surface environment consider alternative credible
22
The evolution of the geosphere at the Olkiluoto site is discussed (see Sections 5.1, 6.1, and 7.1 in Performance Assessment ) for
the time window of the ongoing temperate period (boreal climate), for selected time windows of cold and dry climate
(permafrost), and for selected time windows of ice-sheet advance and retreat (especially ice-sheet retreat, which tends to be more
threatening to the isolation properties of the repository).
23
Note that the evolution of one canister or more with an initial undetected penetrating defect will also be coupled to the evolution
of the system as described above.
161
lines of evolution arising from reasonable variations of key FEPs, such as alternative
discharge locations, sea-level changes, land uses and human habits.
Lines of evolution including unlikely features, events and processes are considered
within disturbance scenarios and consider both unlikely natural phenomena (e.g. large
earthquakes, intrusion of dilute melt waters to repository depth to the extent that the
buffer becomes significantly eroded) and unlikely phenomena related to human actions
(e.g. inadvertent human intrusion and unlikey dietary profiles).
7.1.2
Process for identification of scenarios and cases
Repository system
Posiva’s approach to scenario formulation for the repository system follows a top-down
approach. This includes first identifying the safety functions, then considering the effect
of single FEPs or combination of FEPs on these functions, and also the effect of
uncertainties within the expected lines of evolution. The regulatory framework is taken
into account; it is prescriptive in terminology and definitions. Thus:

The safety functions for each of the repository system components are defined and a
range of values (performance targets and target properties) is given whenever
possible (see Section 2.2 and Design Basis).

FEPs that could adversely affect one or more safety functions at a given time or
place or under specific conditions within the repository are identified (i.e. FEPs that
are scenario drivers within the evolution of the repository system in time and space
(see Performance Assessment).

The effects of uncertainties in the expected evolution of the repository system are
taken into account (see Performance Assessment).

Lines of evolution that describe the evolution of the repository system and
ultimately lead to canister failure, form the basis for the definition of radionuclide
release scenarios. Each line of evolution is then classified using STUK’s scenario
terminology (Section 2.3.6).

For each of the scenarios, a set of calculation cases is defined to analyse the
potential radiological impact. The calculation cases take into account uncertainties
in the assumptions and data through variations in the models and parameter values.
The most relevant evolution-related FEPs that may affect the safety functions of the
repository system, and thus also affect migration-related FEPs, have been taken into
account in analysing and describing the expected evolution (see Performance
Assessment). Climate evolution (see Section 4.4) is the overarching FEP affecting the
whole disposal system.
In the expected evolution, the safety function of the canister(s) holds for hundreds of
thousands of years (see Performance Assessment) and no releases would occur within
the time window of the first several millennia. Nonetheless, according to regulations
(Government Decree 736/2008), a base scenario needs to be defined with a high
probability of causing radiation exposure, but of low consequences.
162
Therefore, the base scenario leading to radionuclide releases is defined as a scenario in
which one or a few canisters with an initial penetrating defect are emplaced in the
repository, while the safety functions of all the other canisters and the other repository
system components are maintained as described in Performance Assessment.
Variant scenarios are identified that consider plausible alternative assumptions for key
FEPs that could affect safety functions and the release of radionuclides. Disturbance
scenarios consider very unlikely events or processes that could lead to loss of safety
functions and release of radionuclides. Finally, complementary cases investigate key
uncertainties that help to give a better understanding of the modelled system or
subsystems. The complementary cases are fully documented in Formulation of
Radionuclide Release Scenarios and analysed in Assessment for Radionuclide Release
Scenarios for the Repository System and Biosphere Assessment.
The surface environment
Formulation of scenarios for the surface environment must be consistent with the
regulatory requirements, the methodology used in the formulation of scenarios for the
repository system, and the current radiation protection systems for humans and the
environment. Posiva’s approach for the surface environment is somewhat different from
that for the repository system, since the surface environment has no safety functions.
The process for identifying scenarios and cases is as follows.

Constraints on the scenarios arising from the regulatory framework are identified.

The most important FEPs (denoted key scenario drivers) with respect to the
evolution of the surface environment, fate of radionuclides in the surface
environment and/or the radiation exposure of humans, plants and animals are
identified. This work also comprises identifying other FEPs that affect the key
scenario drivers, either in isolation or combined, and could induce changes in a
timeline of evolution.

One or several lines of evolution are defined that describe in a timeline the surface
environment evolution from which one or more scenarios are formulated. One
credible line of evolution is identified and used to formulate the Base Scenario for
the surface environment.

Variant scenarios are formulated, mainly by considering reasonable deviations from
the lines of evolution underpinning the Base Scenario. Variant scenarios can include
additional scenario drivers (FEPs) with a potentially significant effect on the fate of
radionuclides in the surface environment and/or the radiation exposure of humans,
plants and animals.

Disturbance scenarios are formulated, mainly by identify unlikely FEPs or mainly
by considering unlikely deviations from the lines of evolution underpinning the
Base Scenario. Disturbance scenarios can include additional scenario drivers (FEPs)
with a potentially significant effect on the fate of radionuclides in the surface
environment and/or the radiation exposure of humans, plants and animals.

A set of biosphere calculation cases is defined to analyse the surface environment
scenarios. These cases take into account uncertainties in assumptions and models,
and the uncertainties and variability in parameter values applied in the models.
163
7.2
Repository system scenarios
7.2.1
Base scenario for the repository system
Derivation of the base scenario
Considering scenarios for radionuclide release, the canister (which provides the safety
function of prolonged containment of the spent fuel) is the primary barrier since
radionuclide releases may only occur if the canister has failed.
Possible canister failure modes are generalised and localised corrosion, rock shear, and
presence of an initial undetected defect (see Chapter 6). As discussed in detail in
Performance Assessment, canister failure through corrosion is only possible at very long
times in the future. When using pessimistic assumptions, a few canisters may fail within
one million years after chemical erosion of the buffer leads to advective conditions and
to enhanced corrosion. Canister failures through rock shear are only possible as a result
of a large earthquake following rock stress changes due to glacial loading and
unloading, and thus will only occur in the time frame of about one hundred thousand
years or more. Although the average annual probability of an earthquake large enough
potentially to lead to canister failure is estimated to be low, in the order of 10-7, the
possibility of a limited number if canister failures by this mechanism cannot be
excluded over a one million year time frame.
It is expected that most of the canisters can be technically manufactured as designed
without any initial penetrating defect, and any penetrating defect greater than about 0.5
mm diameter is likely to be detected by weld inspection and testing (Holmberg &
Kuusela 2011, see also Section 3.5). The base scenario postulates that one or a few
defective canisters are emplaced in the repository. The currently available data are
insufficient, even when expert judgement is used, to make a reasonable estimate of the
probability of emplacing a defective canister in the repository. However, with additional
data on the welding process and continued development of the NDT process, it seems
practicable in the future to show that the probability of more than one initially defective
canister in the repository is less than one per cent. Thus, for the base scenario one
canister with an undetected penetrating defect of 1 mm diameter is assumed in the
Reference Case and in other cases addressing this canister failure mode. This is
consistent with YVL Guide D.5, which states that the base scenario shall assume the
performance targets for each safety function, taking account of incidental deviations
from the target values.
Thus, in the base scenario Reference Case, the undetected penetrating defect in one
canister is the incidental deviation, which acts as the main driver, whereas the
performance targets of most of the canisters and of all the other repository components
are expected to hold as shown in the performance assessment. The likelihood and
consequences of more than one defective canister being emplaced are, however,
considered in the complementary analysis described in Section 8.7.1.
It is assumed that the deposition tunnels have been excavated, and deposition holes
selected, successfully applying Rock Suitability Classification criteria. This implies that
during the disposal system evolution: 1) there will be a very low probability of
earthquakes leading to rock shear displacements along existing fractures or
164
discontinuities that could damage the engineered barrier system in most of the selected
deposition holes; 2) the evolving hydrogeological, hydrogeochemical, and mechanical
conditions will not impair the safety functions of the buffer, backfill, closure or host
rock for the majority of the deposition holes. Since initial canister defects are
uncorrelated to where the canister will be deposited, this means that for the Reference
Case it can be assumed that the canister with initial defect will experience rock and
engineered barrier systems with intact safety functions. The assumptions regarding the
EBS and host rock are shown in Table 7-1.
Nevertheless, there is substantial uncertainty in the radiological outcome of the base
scenario arising from the location of the defective canister. To analyse the case, the
canister position that may lead to radionuclide releases to a discharge location or
locations in the surface environment within the first millennia have been selected from a
chosen DFN realisation. This is valid since the values that would be assigned to
cautiously selected flow-related parameters have been shown not to vary greatly
between DFN realisations. Radiological impacts of multiple canister failures might or
might not coincide spatially and temporally (be additive) depending on location and
timing of release to the surface environment.
Lines of evolution, key processes and uncertainties
The climate evolution that encompasses the repository system evolution is the one
presented in Chapter 4 in Formulation of Radionuclide Release Scenarios and
summarised in Section 4.4 of this Synthesis. The evolution shows alternate temperate,
permafrost and ice-sheet periods. For the first tens of thousands of years and prior to the
first permafrost period at about 50,000 years after present, the hydrogeologically
effective precipitation data derived for a climate evolution for a CO2 concentration of
400 ppm or 280 ppm are very similar. Precipitation data are used for surface hydrology
modelling that serves as a boundary condition for groundwater flow modelling (Löfman
& Karvonen 2012). After 50,000 years permafrost may develop, but it is cautiously
assumed to have no effects on the release rate of radionuclides or on in the release paths
that could retard the transport of radionuclides to the surface.
The repository system in the base scenario follows the expected evolution depicted in
Chapters 5 to 8 in Performance Assessment in which the majority of the canisters, and
the buffer, backfill, closure components and host rock, maintain their respective safety
functions for the whole assessment period.
It is assumed that the transport of radionuclides from the defective canister to the rock
takes place through a fracture around the deposition hole, and that there is a damaged
zone around the deposition hole. Given that the transport path from the defective
canister to the near field is pessimistically assumed to become established within the
first millennia, the groundwater type is selected according to current observations and
modelling results (Löfman & Karvonen 2012), and thus is assumed to be brackish. As
stated above, the position(s) of the defective canister in the repository have been
cautiously selected from a chosen DFN realisation that takes into account the whole
repository system (see Section 6.2 in Assessment of Radionuclide Release Scenarios for
the Repository System).
165
Engineered Barrier System (EBS)
Bedrock
Table 7-1. Assumptions for the base scenario for the repository system (note also the
enveloping climate evolution discussed in Section 4.4).
Rock mass
RSC criteria are applied successfully and target properties hold during the
evolution.
Groundwater
Limited advection or inflows to repository level. Favourable groundwater
chemistry for the EBS and target properties for the groundwater chemistry
hold during the evolution.
Closure
Closure backfill and seals, including borehole seals are designed and
emplaced according to requirements, and performance targets are fulfilled
during the evolution.
Deposition tunnel
backfill
Deposition tunnel backfill and plugs are designed and emplaced according
to requirements. The backfill performance targets are fulfilled during the
evolution.
Buffer
The buffer is designed and emplaced according to requirements. The
buffer performance targets are fulfilled during the evolution.
Canister
Canisters are manufactured and emplaced according to design. As an
incidental deviation it is assumed that one canister is present with an initial
undetected penetrating defect of 1.0 mm diameter, which size will not
change in time.
Spent nuclear fuel
and cladding
Very low dissolution rate; no requirements or safety function in itself.
Figure 7-1 illustrates the repository base scenario and the uncertainties related to the
engineered and natural barriers and their safety functions addressed in the scenario. The
uncertainties are related to: 1) location of the defective canister within the repository,
and the corresponding flow-related transport parameters and discharge locations to the
surface environment; 2) the time of establishment of the transport path ending in the
selection of more than one time; 3) the speciation of several radionuclides as anions or
cations in the near and far field (Wersin et al. 2013a and b, Hakanen et al. 2013) ending
in the selection of more than one parameter (i.e. sorption, diffusivity) value for those
radionuclides.
166
Figure 7-1. The repository base scenario and the uncertainties addressed related to the
repository barriers and their safety functions.
7.2.2
Variant scenarios for the repository system
The reduced performance of any single safety function(s) of any component other than
the canister does not immediately give rise to canister failure and thus to radionuclide
releases. However, the reduced performance of the buffer may subsequently affect
radionuclide release and transport. The combined effect of the reduced performance of
the canister and the buffer is assessed in two variant scenarios, where the loss of the
safety function of the canister (initial penetrating defect, or failure by corrosion) is
combined with the reduced performance of the buffer.
Variant scenario 1 (VS1): Enlarging defect and degradation of the
buffer
In this scenario, one canister is assumed to have an initial penetrating defect of 1 mm
diameter at the time of emplacement that, due to corrosion, will enlarge up to 10 mm on
a timescale of 25,000 years. The degraded performance of the buffer is assumed to have
arisen as a consequence of a process or combination of processes that are likely to occur
within the first tens of thousands of years after emplacement, such as piping erosion,
and/or montmorillonite transformation (e.g. due to heat-transfer-induced cementation,
iron-clay interaction, and interaction with high pH water), which may lead to a reduced
effective buffer thickness.
Montmorillonite transformation would occur preferentially at the interface between the
canister and buffer (if induced by heat transfer) or at the interface between buffer and
167
host rock. Piping and erosion would occur preferentially at the host rock/buffer
interface. In either case, the whole thickness of bentonite would not be affected.
It is assumed that the transport of radionuclides from the defective canister to the rock
takes place though a fracture around the deposition hole, and that there is a damaged
zone around the deposition hole, as in the base scenario.
The uncertainties considered in this scenario are related to: 1) the composition of
groundwater, also reflected in the selection of the sorption, diffusion and solubility
values for the geosphere (see VS1-BRACKISH and VS1-HIPH in Table 7-3); 2) the
composition of porewater in the bentonite, reflected in sorption, diffusion and solubility
values for radionuclides in the canister, buffer, and backfill (see VS1-HIPH_NF in
Table 7-3). The buffer thickness is varied relative to the base scenario, as its
performance is assumed to be degraded because of loss of buffer due to piping and
erosion or because of montmorillonite transformation (see process 5.2.6 in Features,
Events and Processes).
Variant scenario 2 (VS2): Corrosion failure following buffer erosion
The line of evolution leading to this scenario includes chemical erosion of the buffer
followed by enhanced corrosion of the copper canister. It is assumed that all canisters
are initially intact, and that there are no processes adversely affecting the safety
functions of the EBS and the geosphere until after the advance and retreat of ice sheets
has made possible the conditions for penetration of dilute groundwaters to repository
depth. During the retreat of ice sheets, the intrusion of dilute glacial meltwater of low
ionic strength may reach a few deposition holes for a short period leading to chemical
erosion of bentonite (see Sections 7.1 to 7.5 in Performance Assessment). After
sufficient buffer material has been eroded, advective conditions may be established
between the canister and the rock; this is assumed to occur once the buffer mass loss
exceeds 1200 kg for the cautious assumption of incomplete homogenisation of the
buffer or fracture clogging (see Section 7.5.5 in Performance Assessment). Advective
groundwater flow then carries sulphide to the affected canister(s) and, after corrosion
has progressed sufficiently, the loss of canister containment and transport resistance.
There is uncertainty in the flow conditions around the deposition holes, and hence in the
number, locations and timings of canister failures. Four calculations cases are defined,
each corresponding to one of the four canisters that are calculated to fail within this time
frame based on a single realisation of the DFN groundwater flow model and a specific
set of groundwater flow model assumptions, as described in Performance Assessment.
Overall radionuclide releases in this scenario are obtained by superimposing the results
of these cases.
7.2.3
Disturbance scenarios for the repository system
In formulating disturbance scenarios, two main unlikely events are taken into account:
one is the occurrence of an earthquake capable of originating a rock shear large enough
to breach the canister, for which the probability of occurrence within different time
windows has been documented in Section 7.2 of Performance Assessment; the other is
inadvertent human intrusion, which is treated as a surface environment scenario (see
Section 7.3.3). FEPs that are likely to occur, but only detrimentally affect safety
168
functions if their rates are outside the expected range of possibilities, are also taken into
account. Further, the unlikely event of an accelerated insert corrosion rate, which allows
the insert to corrode faster and the corrosion products of the insert to expand and
mechanically breach the copper overpack resulting in a total loss of transport resistance,
has been considered.
FEP number 8.2.3 “reactivation – displacement along existing fractures” is closely
related to an earthquake-promoted rock shear, as rock shear is most likely to occur on
existing fractures (see 8.2.3 in Features, Events and Processes). This FEP may be
combined with other FEPs in defining disturbance scenarios (see below). Two other
FEPs that are likely to occur, but only detrimentally affect safety functions if acting at a
rate outside the expected range of possibilities, are corrosion coupled to mechanical
deformation (see processes 4.2.5/4.2.6 and 4.3.2 in Features, Events and Processes).
Disturbance scenario: Rock shear (RS)
The line of evolution of the repository system selected for this scenario is that of
expected evolution up to either 1) about 40,000 years AP or 2) about 155,000 years AP.
At this time, it is assumed that an earthquake occurs that causes a rock shear
displacement sufficient to breach a canister, but still keeping most of the buffer in place.
The selection of times of 40,000 and 155,000 years is based on the annual probability
for canister failure due to rock shear, being largest between 10,000 and 50,000 years AP
and after 100,000 years AP. The time 40,000 years is selected arbitrarily for comparison
to the results of a latter time. The significance of the latter time arises from the retreat of
an ice sheet and re-establishment of temperate conditions (see Ch. 4 in this Synthesis
and Section 7.2.4 in Performance Assessment), i.e. it relates to post-glacial fault
reactivation. The major difference between assuming an earthquake at 40,000 or at
155,000 years is the radionuclide inventory of the canister at that time, as the annual
probability of occurrence of canister failure is the same in both cases (see Section 7.2 in
Performance Assessment). Other uncertainties to be accounted for are groundwater flow
and composition, which may be substantially changed in the case of an earthquake after
ice-sheet retreat.
Disturbance scenario: Rock shear and buffer erosion (RS-DIL)
The line of evolution of the repository system selected for this scenario is, in part,
identical to the Rock shear scenario, i.e. normal or expected evolution up to either 1)
about 40,000 AP or 2) about 155,000 year AP. At this time, it is assumed that an
earthquake occurs that causes a rock shear displacement sufficient to breach a canister.
In addition, however, a perturbation of the fracture network due to the rock shear is
assumed, leading to inflow of dilute, low ionic strength, water reaching the positions of
the breached canisters at the time of canister failure and also later in association with
ice-sheet retreat and hence resulting in chemical erosion of the buffer. The penetration
of dilute water due to the long persistence of temperate conditions with meteoric water
infiltration to repository depth at 40,000 year AP is likely, but not necessarily at any
particular canister position (see Section 7.1.3 in Performance Assessment), and
therefore also not necessarily at the canister positions affected by a rock shear
displacement. There is a cautious assumption of incomplete homogenisation of the
buffer and either to fracture clogging. Advective conditions are assumed to be
established in the buffer once the buffer mass loss exceeds 1200 kg (see above).
169
Disturbance scenario: Accelerated insert corrosion (AIC)
In this scenario, it is assumed that a canister with an initial penetrating defect (1 mm
diameter) has been emplaced in the repository, as in the base scenario. The base
scenario assumes a reasonable expected corrosion rate for the cast iron insert (0.1 to
1 μm/year; Pastina & Hellä 2010) and a reasonable behaviour for the evolution of the
corrosion products (mostly magnetite) with no consequences for the copper overpack. In
the AIC scenario, however, an accelerated insert corrosion rate (> 1 μm/year) is
assumed, allowing the insert to corrode faster and the corrosion products of the insert to
expand and mechanically breach the copper overpack resulting in a total loss of
transport resistance after 15,000 years.
There is uncertainty over the tightness of the cast iron insert. If the insert is water-tight
then radionuclide releases cannot occur until the insert is breached (at 15,000 years); if
the insert is not tight (leaky) then releases may begin at 1000 years as in the Reference
Case, with total loss of transport resistance after 15,000 years. The alternatives are
represented by two calculation cases.
The assumptions of this scenario are very pessimistic. First, the minimum thickness of
the insert outer wall varies between 30 and 50 mm and thus it would take a minimum of
30,000 to 50,000 years for the insert to corrode at the corrosion rate of 1 μm/year.
Moreover, there is no evidence that the formation of the insert corrosion products could
breach the copper overpack in repository conditions since magnetite is porous and
yielding oxide. A third factor is that the corrosion of the iron insert would be limited by
the low rate of diffusion of water vapour through the defect. For these reasons the
scenario is judged very unlikely, and classified as a disturbance scenario.
7.2.4
Radionuclide release scenarios and cases
The radionuclide release scenarios and cases taken forward to quantitative analysis are
summarised in Tables 7-2 to 7-4. These present a systematic investigation of the main
uncertainties identified within each scenario.
Table 7-2. Calculation cases for the radionuclide release base scenario for the
repository system.
Scenario
BASE SCENARIO:
Incidental deviation of
introducing one or a
few canisters with a
penetrating defect of 1
mm diameter
Calculation case
Short description
BS-RC
Reference case (RC) – one canister with an initial
penetrating defect of 1 mm diameter.
Cautious position selected from a DFN realisation taking
into account the whole repository.
BS-LOC1
As RC, except alternative position_1 – investigates the
uncertainties in the selection of flow-related parameters
(uncertainty in DFN realisation).
BS-LOC2
As RC, except alternative position_2 – investigates the
uncertainties in the selection of flow-related parameters
(uncertainty in DFN realisation).
BS-ANNFF
As RC, except Ag, Mo, Nb migrate as anions in the near
and far field (i.e. geosphere) – investigates uncertainty in
the speciation of those elements.
BS-TIME
As RC, except uncertainty in the time needed to establish
a transport path from the defective canister is taken into
account (1000 years in RC and 5000 in TIME)
170
Table 7-3. Calculation cases for the radionuclide release variant scenarios.
Scenario
VARIANT SCENARIO 1:
Initial defect gradually
enlarges due to
corrosion
VARIANT SCENARIO 2:
No initial penetrating
defects (thin copper
canister wall 35 mm):
Erosion of buffer and
subsequent corrosion of
four canisters
Calculation case
Short description
VS1-BRACKISH
Cautious position as in the RC – initial penetrating defect
enlarging; degraded buffer; speciation for brackish water.
VS1-HIPH
Cautious position as in the RC – initial penetrating defect
enlarging; degraded buffer; speciation for high pH water
in the near and far field.
VS1-HIPH_NF
Cautious position as in the RC – initial penetrating defect
enlarging; degraded buffer; speciation for high pH water
in the near field alone.
VS2-H1
VS2-H2
VS2-H3
VS2-H4
Canisters in four positions fail due to corrosion after
buffer is chemically eroded. The four canisters that are
calculated to fail within this time frame are based on a
single realisation of the DFN groundwater flow model and
a specific (reference) set of groundwater flow model
assumptions.
Table 7-4. Calculation cases for the radionuclide release disturbance scenarios
Scenario
Calculation case
Short description
AIC-LI
The insert of a defective canister with an initial defect
starts to corrode at 1000 years after emplacement –
releases from a leaky insert start also at 1000 years.
Transport resistance is suddenly lost at 15,000 years.
AIC-TI
The insert of a defective canister with an initial defect
starts to corrode at 1000 years after emplacement – no
releases from a tight insert. Transport resistance is
suddenly lost at 15,000 years.
RS1
Canister(s) fail as a consequence of rock-shear at
40,000 years after emplacement.
RS2
Canister(s) fail as a consequence of rock-shear at
155,000 years after emplacement.
RS1-DIL
Canister(s) fail as a consequence of rock-shear at
40,000 years. Buffer erosion follows the event whenever
low ionic strength water is available.
RS2-DIL
Canister(s) fail as a consequence of rock-shear at
155,000 years. Buffer erosion follows the event
whenever low ionic strength water is available.
AIC
Accelerated Insert
Corrosion
RS
Rock Shear
RS-DIL
Rock Shear followed of
buffer erosion
7.3
Surface environment scenarios
This section gives an overview of the main assumptions used in formulating the
radionuclide release scenarios for the surface environment; these are discussed in detail
in the Formulation of Radionuclide Release Scenarios.
7.3.1
Base scenario for the surface environment
The surface environment scenarios are formulated independently from the repository
system and are limited to the dose assessment time window, hence covering the first ten
millennia after disposal. The base scenario for the surface environment is formulated
bearing in mind that this time window is relatively short compared with the whole
171
assessment time frame. The base scenario and its main assumptions are briefly
summarised below.
Key statements in the regulations are that the environmental changes due to sea-level
changes relative to the land (i.e. allowing for land uplift) should be considered, and that
the climate type as well as the human habits can be assumed to remain unchanged
(Guide YVL D.5, 307). Thus it is appropriate to assume the current climate type in the
region of the Olkiluoto site in the scenario formulation. Furthermore, Posiva judges that
it is appropriate to assume present-day demographic data and human habits, such as the
number of inhabitants in the region and land use, in the scenario formulation.
The past development of the climate and surface environment conditions, focusing on
the time span since the last deglaciation until present and current conditions in the
region of the Olkiluoto site are presented in Biosphere Description, along with the
current activities and habits of people in the region of the Olkiluoto site. This forms the
knowledge basis, especially on land uses and exposure pathways to people.
Furthermore, the regulations state (Guide YVL D.5, 317) that the assessment of
radiation exposures of flora and fauna shall assume present kinds of living terrestrial
and aquatic populations in the disposal site environment. Posiva interprets this as
making it appropriate to identify a set of representative species based on present-day
conditions at the Olkiluoto site and in the region. This is discussed in detail in Dose
Assessment for Plants and Animals.
The main features, events and processes (FEPs) assumed to drive the scenario
formulation (the key scenario drivers) are associated with the evolution of the natural
environment, the climate and how humans behave, especially how the land is utilised.
The two key scenario drivers identified are sea-level change (local) and land use.
The key scenario driver sea-level change (local) is an aggregated FEP. The FEPs
identified to have the strongest influence on sea-level change at the Olkiluoto site are
climate evolution (Features, Events and Processes, Section 10.2.1) and land uplift and
depression (Features, Events and Processes, Section 10.2.4). The line of evolution
selected for the base scenario for the key scenario driver sea-level change (local)
assumes that the current climate prevails.
The key scenario driver land use is also an aggregated FEP. The FEPs regarding land
use at the Olkiluoto site identified to be most important in scenario formulation are crop
type, irrigation procedures and livestock (addressed in Features, Events and Processes,
Section 9.2.4), forest and peatland management (Features, Events and Processes,
Section 9.2.5), construction of a well (Features, Events and Processes, Section 9.2.30)
and demographics (Features, Events and Processes, Section 9.2.33). The line of
evolution selected for the base scenario for the scenario key driver land use is based on
present-day conditions and prevailing land use practices.
The FEPs mentioned above are those with identified alternative lines of evolution for
the key scenario drivers that are primarily addressed when formulating variant and
disturbance scenarios. More FEPs (see Table 4-1 in Section 4) are taken into account in
the detailed formulation of the scenarios, and subsequently when defining the
calculation cases to be used in analysing the scenarios.
172
7.3.2
Variant scenarios for the surface environment
Variant scenarios for the surface environment are based on alternative credible lines of
evolution arising from reasonable variations of the FEPs affecting the key scenario
drivers. Consideration has also been given to additional scenario drivers. The variant
scenarios that are analysed in terms of doses are listed in Table 7-5 and briefly
described below. It should be noted that more scenarios are formulated and analysed in
the TESM, but not propagated further in the biosphere assemment modelling (these are
documented in Biosphere Assessment).
VS(A) − Discharge locations to the surface environment
Scenario driver: Discharge locations
The discharge locations of radionuclide release from a canister with an initial
penetrating defect may be affected by the position of the defective canister in the
repository, which affects the groundwater flow path(s) by which contaminants will be
carried to the surface environment. The implication of the uncertainty in the location of
a canister and groundwater flow paths on the radiological impact is addressed in this
scenario.
VS(D) – Land use (well)
Scenario driver: Land use
This variant scenario addresses uncertainties in how humans use the land, focusing on
the construction of wells to extract drinking water. In the Base Scenario, the number of
wells used in the model domain is consistent with the present-day average well density
in southwestern Finland. The implication of the uncertainty in the number of wells on
the radiological impact is addressed in this scenario.
Table 7-5. Variant scenarios identified for the surface environment, limited to the ones
analysed in terms of doses, the driver that the scenarios address and the most important
FEPs which uncertainties affect the drivers.
Variant scenario
Scenario driver
FEPs
VS-A
Discharge locations to
the surface environment
Discharge locations
“Defective canister location in the
(a)
repository layout”
VS-D
Land use (well)
Land use
VS-E
Route of radionuclide
transport
Element migration and
accumulation
VS-F
Exposure characteristics
Human habits
Construction of a well (9.2.29), Well
(9.2.30)
Alternative radionuclide transport routes
in biosphere terrestrial and aquatic
compartments affect a number of
terrestrial and aquatic processes
Dietary profile (9.2.32)
VS-G
Combined scenario
Sea-level change (local)
Land use
(a)
Agriculture and aquaculture (9.2.4), Climate evolution (10.2.1), Land uplift
and depression (10.2.4)
See BS-LOC1 and BS-LOC2 in Assessment of Radionuclide Release Scenarios for the Repository
System and Table 7-10.
173
VS(E) – Alternative radionuclide transport routes in biosphere compartments
Scenario driver: Element migration and accumulation
This variant scenario addresses uncertainties in the assumptions underlying the
radionuclide transport in the surface environment. In the Base Scenario it is assumed
that the radionuclide releases from the geosphere enter the biosphere through a deep
overburden in a terrestrial and agricultural ecosystem or into deep sediment in aquatic
ecosystems. In this variant scenario alternative compartments receiving the initial
releases are assumed (e.g. the radionuclides are released direct to the rooting zone in
terrestrial and agricultural ecosystems).
VS(F) – Exposure characteristics
Scenario driver: Human habits
This variant scenario addresses uncertainties in the human diet. In the Base Scenario, it
is assumed that all (contaminated) edibles possibly produced from the different
ecosystems at the site are consumed by humans, and that humans have no preferences
regarding the mix of foods consumed. This scenario assumes that humans in future
generations have the same preferences (dietary profile) regarding food consumption as
the present-day Finnish population.
VS(G) – Combined scenario
Scenario drivers: Sea-level change (local), Land use
This variant scenario addresses uncertainties in climate evolution, land uplift and
depression, agriculture and aquaculture. For each scenario driver, and the FEPs
affecting it, assumptions are made to individually maximise the arable land area, while
remaining consistent with current scientific understanding and within the reasonably
expected range of possibilities.
7.3.3
Disturbance scenarios for the surface environment
In the disturbance scenarios for the surface environment, unlikely lines of evolution that
may have a significant effect on the fate of radionuclides in the surface environment
and/or the radiation exposure of humans, plants and animals are addressed. The
identified disturbance scenarios that are analysed in terms of doses are listed in Table 76 and briefly described below. It should be noted that more disturbance scenarios are
formulated and analysed in the TESM, but not propagated further in the biosphere
assemment modelling (these are documented in Biosphere Assessment).
DS(D) – Exposure characteristics
Scenario driver: Biotope occupancy
This disturbance scenario assesses the impact on the doses to plants and animals due to
uncertainties in the biotope occupancy. In the Base Scenario it is assumed that plants
and animals have specific occupancy preferences (e.g. that some aquatic speices are
found in freshwater but not brackish water), but that they may be found in any suitable
part of the contaminated area of the model domain. This scenario cautiously assumes
constant occupancy of plants and animals in the most constraining (in terms of dose
rate) biotope.
174
Table 7-6. Disturbance scenarios identified for the surface environment, the driver the
scenarios address and the most important FEPs which uncertainties affect the drivers.
Disturbance scenario
Scenario driver
FEPs
DS(D)
DS(F)
Biotope occupancy
Human actions
Habitats
Human actions
Land use
Constructing a well
DS(G)
Exposure characteristics
Inadvertent Human
intrusion
Deep wells
DS(F) – Inadvertent Human intrusion
Scenario driver: Human actions
Human intrusion has been considered an issue in post-closure safety of solid radioactive
waste disposal for many years (NEA 1989). It has been concluded that the possibility of
human intrusion should not be ignored, but it is necessary to recognise the illustrative
nature of any assessment. ICRP (2000, paragraph 62) recommends that “one or more
typical plausible stylised {human intrusion} scenarios” should be considered to
evaluate the resilience of a repository to possible human intrusion events. In TURVA2012, the reference approach for evaluation of human intruder doses developed wihin
the BIOPROTA forum has been adopted (Smith et al. 2012). This includes the
assessment of scenarios for people having direct contact with contaminated material
brought to the surface by drilling, and others who might be exposed due to
contaminated material being left at the drill site.
DS(G) – Deep wells
Scenario driver: Land use
Wells are constructed in the bedrock (drilling or driving or dug) in the overburden,
which may be used for extracting household water, watering animals and irrigation
purposes. Here, the unlikely event that a deep (> 300 m) well is drilled at, or in the
vicinity of, the site is addressed. It is assumed that the well intersects a waterconducting feature somewhere deep in the geosphere and the water drawn from the well
has passed through the repository.
7.3.4
Calculation cases
As stated above it is not feasible to identify the most likely lines of evolution for the
entire surface environment, thus it is not possible to rank the base and variant scenarios
according to their likelihood of occurrence. In order to facilitate a clear communication
of the safety assessment results, the biosphere Reference Case (BSA-RC) is the only
one used to interpret the Base Scenario for the surface environment, and the other
biosphere calculation cases are identified as sensitivity cases arising under variant
scenarios, and what-if cases under disturbance scenarios.
The analysis of the biosphere calculation cases is done step-wise in a modelling chain
taking into account the connection between each biosphere assessment sub-process (see
Figure 5-8). 21 cases are first defined and calculated within the terrain and ecosystem
development modelling, which result is a series of projections of the development of the
surface environment (see Terrain and Ecosystem Development Modelling). It must be
noted that from the 21 cases that have been calculated (see Biosphere Assessment), only
175
the cases presented in Table 7-7 have been propagated to sub-sequent models in the
biosphere assessment process (the reason for not propagating all 21 cases is that most
projections are considered to be similar to the Reference Case projections or bounded
by the Terr_MaxAgri calculation case). Biosphere calculation cases, including the rest
of the chain in the biosphere modelling process, are then identified for each scenario
(Tables 7-8 and 7-9). In each sub-process the FEPs to be adressed given by the scenario
description are identified and models and parameter values are selected accordingly.
This is done independently for the surface- and near-surface hydrological model
(SHYD), the landscape model (LSM), and the dose models used in the radiological
impact assessment (RIA). It is also checked that the settings for each sub-process model
are consistent with both each other and with the scenario to be analysed.
The source term or input for the analysis of the biosphere calculation cases is the result
of the repository calculation cases that give radionuclide releases within the dose
assessment time window (i.e. up to 10,000 years after disposal). Most of the repository
calculation cases propagated to biosphere assessment are analysed with the biosphere
Reference Case (BSA-RC) (see Table 7-10).
The outcome of the Reference Case in the TESM is utilised to construct two landscape
models: one model for geosphere releases from the discharge locations north of the
present Olkiluoto Island, hence suitable for analysing the geosphere releases from BSRC, and all other repository calculation cases that are based on the same canister
location, and one model for geosphere releases from the alternative discharge locations
south of the present Olkiluoto Island (the geosphere releases in the cases BS-LOC1 and
BS-LOC2). These two landscape models are denoted REF and SOUTH in Table 7-8.
The biosphere objects receiving the direct releases differ in the cases BS-LOC1 and BSLOC2, which have implications for the radionuclide transport modelling part of the
landscape modelling sub-process. This is reflected by two model variants for the model
SOUTH (denoted Southern_1 and Southern_2 in the RNT column in Table 7-8). A
Reference Case model is set-up in the surface and near-surface hydrology (SHYD),
denoted REF in Table 7-8. In addition to this, two SHYD cases are identified to analyse
the variant scenario VS(D) addressing undertaintiesd in the number of wells at the site
(denoted MORE_WELLS and NO_WELLS in Table 7-8).
Table 7-7. Calculation cases assessed in terrain and ecosystem modelling (TESM).
TESM case
FEPs
taken
account
into
Reference Case
All relevant FEPs are
taken in to account as
specified in TESM.
YES
Terr_maxAgri
Sea level, Land uplift,
River boundary, Aquatic
erosion and Land use
differ from the Reference
Case
YES
Propagated
to SHYD
Comments
Combines maximised terrestrial area
within
reasonably
expected
development
(faster
land
uplift,
enhanched sedimentation in lakes and
maximum extent of agricultural land.
Uses climate simulation A2 (see Ch.4
in Biosphere Assessment)
176
In the terrain and ecosystem development line that maximises the land area for
agricultural practices (Terr_maxAgri; Table 7-7), alternative sub-models and data are
used for land uplift, climate specifics, sedimentation in lakes and the extent of
agricultural land. Also the water body connections, the surface and near-surface
hydrology and the landscape model change accordingly compared to the base scenario.
The only models that are not affected by this alternative TESM projection are the dose
models in the RIA.
In the base scenario the geosphere releases are assumed enter the biosphere in the
deepest layer of the overburden. To account for uncertainties in the configuration of the
overburden and the release path within it that may affect radionuclide migration and
accumulation, one case, RNT1, has been defined for the RNT part assuming that the
geosphere releases enter the biosphere directly in the rooting zone for terrestrial and
agricultural ecosystems, and directly to the water column in aquatic ecosystems.
In the radiological impact assessment (RIA) the reference model is used for most of the
biosphere calculation cases. Uncertainties in the dietary habits for humans are taken into
account by implementing a model assuming that future generations have the same
preferences regarding consumption of various food groups as the present-day Finnish
population.
Calculation cases within the base scenario
The Reference Case (BSA-RC) is the only one within the base scenario. This, and the
repository calculation reference case (BS-RC) connect the entire disposal system in
what is deemed the most expected and credible line of evolution.
Calculation cases within variant scenarios
Within the variant scenario VS(A), two calculation cases are defined that account for
uncertainties in the discharge locations to the surface environment (see Tables 7-8 and
7-10). In both cases, VS(A)-SOUTH1 and VS(A)-SOUTH2 the reference terrain and
ecosystem development modelling results are used, but uses a different landscape
model compared to the Reference Case, since the contaminated areas are located south
of the present Olkiluoto Island. The assumptions and models the RIA are the same as in
the Reference Case.
Two cases have been defined for the variant scenario VS(D), VS(D)-WELL that
assumes there to be more wells than in the base scenario, and VS(D)-NO_WELL that
assumes there are no wells (see Table 7-8).
In variant scenario VS(E), an alternative radionuclide transport route in biosphere
compartments are considered in VS(E)-RNT1. In this case radionuclide releases from
the geosphere to the biosphere is assumed to be introduced directly to the rooting zone
in terrestrial and agricultural ecosystems and to the water column in aquatic ecosystems.
Within the variant scenario VS(F), two cases are defined that differ from the Reference
Case only in the dietary profile assumed for future human generations. VS(F)-FINDIET
assumes that the consumed various food groups are the same as consumed by the
present-day Finnish population (based on the The National FINDIET 2007 Survey,
177
Paturi et al. 2008), and VS(F)-VEG assumes a vegetarian diet. No doses to animals and
plants will be calculated in these cases as they would remain identical to the reference
case.
VS(G)-COMBI is the only case defined within the variant scenario VS(G), where the
uncertainties in climate evolution, land uplift and depression, agriculture and
aquaculture are combined. In this case the extension of the areas dedicated to agriculture
is maximised. The terrain and ecosystem modelling (TESM), the results used are those
of Terr_maxAgri and so are the results of SHYD and LSM (Table 7-8).
Table 7-8. Biosphere calculation cases under the base and variant scenarios.
Calculation case
Calculation case in the sub-process modelling
Comments
TESM
SHYD
LSM
RNT
RIA
REF
REF
REF
REF
REF
Reference Case for
the Base scenario
for the entire
disposal system.
VS(A)-SOUTH1
REF
REF
SOUTH
Southern_1
REF
Uncertainties in
discharge locations
to the biosphere
VS(A)-SOUTH2
REF
REF
SOUTH
Southern_2
REF
Uncertainties in
discharge locations
to the biosphere
VS(D)-WELL
REF
MORE_WELLS
REF
REF
REF
More wells than in
the Reference Case
VS(D)-NO_WELL
REF
NO_WELLS
REF
REF
REF
No wells. Irrigation
and drinking water
from surface waters.
VS(E)-RNT1
REF
REF
REF
REF1
REF
Geosphere releases
get directly to
alternative
compartments
compared with the
BSA-RC. See main
text.
VS(F)-FINDIET
REF
REF
REF
REF
FINDIET
Dietary profile based
on present-day
average
consumption
statistics
VS(F)-VEG
REF
REF
REF
REF
VEG
Dietary profile with
no meat and fish.
Terr_max
Agri
Terr_maxAgri
Terr_max
Agri
Terr_maxAgri
REF
Combination of
uncertainties in sea
level, climate and
assumes extensive
agriculture
BSA-RC
VS(G)-COMBI
178
Calculation cases within disturbance scenarios
The disturbance scenario DS(D) takes into account unlikely exposure characteristics for
plants and animals, which depend on their habitat. One calculation case DS(D)HABITAT has been defined that considers that the most exposed plants and animals
live in the most contaminated object. In this case the same reference TESM, SHYD, and
LSM as in the reference case are used (see Table 7-9a). In DS(D)-HABITAT the
radiological impact assessment for humans remains as in the reference case.
In the inadvertent human intrusion scenario DS(F) six calculations cases are considered.
In all cases it is assumed that borehole drilling (e.g. for geothermal energy purposes), is
conducted somewhere within the footprint of the repository and reach repository depth.
In two of the six cases it is assumed that drilling hits an intact canister (DS(F)CANISTER) containing all spent fuel radionuclide inventory at the time. In one case all
the drill crew is exposed (DS(F)-CANISTER-D) and in the other a geologist is exposed
when examining drill core(s). In the other cases it is assumed that contaminated buffer
and backfill material is drilled and brought up to the surface (see DS(F)-BUFFER and
DS(F)-BACKFILL in Table 7-9b).
Table 7-9a. Biosphere calculation case under disturbance scenarios that is analysed
following all biosphere assessment sub-processes.
Calculation
cases
DS(D)-HABITAT
Calculation case in the sub-process modelling
TESM
SHYD
LSM
RNT
RIA
REF
REF
REF
REF
Occupancy
constraints
Comments
Only dose for plants
and animals
Table 7-9b. Biosphere calculation cases for analysing inadvertent human intrusion.
Calculation cases
Comments
DS(F)-HI-CANISTER-D
Drilling through an intact spent fuel canister; Drill
crew exposed in the process
DS(F)-HI-CANISTER-G
Drilling through an intact spent fuel canister;
Geologist exposed when examining a drill core
DS(F)-HI-BUFFER-D
Drilling through contaminated buffer material; Drill
crew exposed in the process
DS(F)-HI-BUFFER-G
Drilling through contaminated buffer material;
Geologist exposed when examining a drill core
DS(F)-HI-BACKFILL-D
Drilling through contaminated backfill material; Drill
crew exposed in the process
DS(F)-HI-BACKFILL-G
Drilling through contaminated backfill material;
Geologist exposed when examining a drill core
179
In the deep well scenario DS(G) all repository repository calculation cases that give
releases within the dose assessment time frame window are analysed as separate
calculation cases. These are not further addressed in this report, the analysis of these
cases are presented in detail in Biosphere Assessment to which the interested reader is
encouraged to consult.
Link between repository and biosphere calculation cases
There are nine repository calculation cases that give releases within the dose assessment
time window of 10,000 years. These cases are propagated to the biosphere assessment
and analysed with biosphere calculation cases. These repository calculation cases and
the type of biosphere calculation cases that are used to analyse them are summarised in
Table 7-10. The guiding principle is that the Reference Case for the repository system
(BS-RC) is analysed with all identified biosphere calculation cases, and all other
repository calculation cases are analysed with the Reference Case for the surface
environment (BSA-RC). An exception is that some repository calculation cases lead to
geosphere release to the south of the present-day Olkiluoto Island, so that is more
suitable to use the characteristics of this area in the biosphere analysis.
Table 7-10. Repository calculation cases propagated to the biosphere assessment.
Biosphere calculation cases used in the analysis and the name of the resulting
calculation case that is analysed with the biosphere full modelling chain.
Repository
calculation
case
Description of Repository
Calculation Case
Biosphere Calculation Case/s
used in the analysis (see also
Table 7-8 )
Name of the
resulting
calculation case
combination
BS-RC
Canister location 381,
leading to releases to the
surface environment north of
the present-day Olkiluoto
Island
All identified biosphere cases for
the relevant discharge locations
and human intrusion scenario
cases
BSA-RC (see
Table 7-8)
BS-LOC1
Sensitivity case assuming
alternative canister location.
Canister location 2418,
leading to releases to the
surface environment south of
the present-day Olkiluoto
Island
VS(A)-SOUTH1
VS(A)-SOUTH1
(see Table 7-8)
BS-LOC2
Sensitivity case assuming
alternative canister location.
Canister location 3829,
leading to releases to the
surface environment south of
the present-day Olkiluoto
Island
VS(A)-SOUTH2
VS(A)-SOUTH2
(see Table 7-8)
BS-ANNFF
Sensitivity case assuming
alternative near-field and
geosphere speciation.
BSA-RC
BSA-ANNFF
BS-TIME
Sensitivity case assuming
delayed establishment of
transport path.
BSA-RC
BSA-TIME
VS1BRACKISH
Sensitivity case assuming
reduced buffer thickness.
BSA-RC
BSA-BRACKISH
180
VS1-HIPH
Sensitivity case assuming
reduced buffer thickness and
high-pH groundwater.
BSA-RC
BSA-HIPH
VS1-HIPH_NF
Sensitivity case assuming
reduced buffer thickness and
high-pH (near-field).
BSA-RC
BSA-HIPH_NF
AIC-LI
What-if case assuming a
high insert corrosion rate,
causing a sudden loss of
transport resistance of the
defect after 15,000 years.
Identical to BS-RC in the
dose assessment time
window.
BSA-RC
BSA-AIC-LI
7.4
Summary and discussion on comprehensiveness
The formulation of radionuclide release scenarios for the repository system is the link
between Performance Assessment and Analysis of Radionuclide Release Scenarios for
the Repository System. All these make use of the Features, Events and Processes (FEPs)
that potentially could affect the disposal system and that have been defined in Features,
Events and Processes.
The formulation of radionuclide release scenarios for the surface environment brings
together Biosphere Description and the surface environment FEPs and is the link to the
assessment of the surface environment scenarios. The assessment of the surface
environment scenarios is summarised in Biosphere Assessment and discussed in detail
in:

Terrain and Ecosystems Development Modelling;

Surface and Near-Surface Hydrological Modelling;

Biosphere Radionuclide Transport and Dose Assessment; and

Dose Assessment for Plants and Animals.
7.4.1
Demonstrating that the set of scenarios is comprehensive
To claim that the set of formulated repository system scenarios is comprehensive it is
necessary to check if all the relevant FEPs have been taken into account. The scenarios
are formulated to ascertain the impact of uncertainties in the initial state and evolution
of the repository system, which have been highlighted in Performance Assessment. In
that report the most relevant evolution-related FEPs were taken into account, therefore it
remains to be checked what other FEPs have been included in the base, variant, and
disturbance scenarios.
Appendix 3 shows the repository system related FEPs with colour codes. Most of the
non-coloured FEPs have been taken into account in groundwater flow modelling (e.g.
degradation of auxiliary components, most backfill evolution-related processes, erosion
and sedimentation in fractures, etc.). Possible scenarios dealing with criticality will be
treated in a future stage after submitting the construction licence application to confirm
181
that, in the long term, the possibility of criticality within or outside waste packages is
indeed negligible. Freezing and thawing has been shown not to have detrimental effects
for the buffer, backfill, and auxiliary components.
In all the scenarios, the alteration and dissolution of the spent nuclear fuel matrix (FEP
3.2.8 in Features, Events and Processes), radioactive decay (FEP 3.2.1), and the release
of the labile fraction of the inventory (FEP 3.2.9) are taken into account. Heat
generation (FEP 3.2.2) and heat transfer (3.2.3) from the spent fuel are taken into
account in the design of the layout of the disposal facility as well as in the modelling of
groundwater flow (see FEPs in yellow in Appendix 3). Although precipitation and coprecipitation are conservatively not taken into account in radionuclide migration
calculations, they are nonetheless accounted for in calculating solubility limits. These
limits are applied in all the scenarios.
In the base scenario all migration FEPs are taken into account with the exception of
colloid transport (FEPs 4.3.6/5.3.6/6.3.6/8.3.6 in Features, Events and Processes report)
and advection in the canister, buffer and backfill (FEPs 4.3.5/5.3.5/6.3.5).
In variant scenario VS1, the partial degradation of the buffer is taken explicitly into
account due either to piping and erosion (FEP 5.2.3) or to montmorillonite
transformation (5.2.3). The influence of microbial activity (5.2.8/8.2.10) is taken into
account indirectly as a potential cause of higher than expected corrosion rates. All
migration FEPs are also taken into account, again with the exception of colloid transport
(see FEPs in yellow and green in Appendix 3).
In variant scenario VS2, the buffer is gradually degraded due to chemical erosion (FEP
5.2.4), and transported as colloids (5.3.6) because of adverse groundwater chemistry
(FEP 8.2.7). The canister is gradually corroded (FEPs 4.2.5, 4.2.6, 4.2.7) with or
without the aid of microbial activity (8.2.10) (see FEPs in yellow, green and blue in
Appendix 3).
The rock shear in the disturbance scenarios RS and RS-DIL take into account the
reactivation and displacement of fractures (FEP 8.2.3) as well as deformation of the
canister(s) (FEP 4.2.3) as it breaches (see also FEPs in orange and yellow). The canister
may fail easily as a consequence of rock shear if it has been affected by corrosion,
which is not explicitly taken into account, but the VS2 scenario encompasses this
possibility. In RS-DIL, the FEPs in blue and advection in the buffer (FEP in green) are
also taken into account.
The disturbance scenario AIC takes into account corrosion and deformation of the
canister. The accelerated insert corrosion could also be explained by microbial activity
(see also FEPs in yellow in Appendix 3).
The surface environment scenarios take into account the FEPs listed in Chapter 9 in
Features, Events and Processes to which the calculation cases in Sections 7.3.2 to 7.3.4
of this report refer.
Based on the explanations given above, it can be said that the repository system
scenarios are comprehensive, since all the FEPs influencing long-term safety have been
182
taken into account. To be confident that the most penalising conditions have been
assessed, the combinations of repository system scenarios and their implied FEPs are
considered in the next section.
The identified set of surface environment scenarios is not yet as thoroughly scrutinised
for comprehensiveness as the repository system scenarios are. This work is in progress
and will mature during the next iteration of the safety case. However, it is the view of
Posiva that the set of scenarios is sufficiently comprehensive to support an application
for construction.
7.4.2
Combinations of repository system scenarios
The radionuclide release scenarios discussed above have each been considered
individually. The scenarios are not, however, all mutually exclusive. For example, it is
not impossible that one of the earthquake/rock shear scenarios occurs when the
repository also contains one or more canisters with initial penetrating defects. If the rock
shear event affected the defective canisters, then the radionuclide release due to rock
shear would be reduced compared with a scenario of rock shear alone, since the IRF
radionuclides would already have been released. The likelihood of a rock shear event
affecting a defective canister is, however, much lower than that of it affecting an intact
canister, since there will be very few if any defective canisters present. Thus, it is
appropriate to treat them as independent, so that their impact in combination is the sum
the time-histories of releases from the two scenarios.
All possible binary combinations of the base, variant and disturbance scenarios have
been considered qualitatively (see Table 7-11). It is found, however, that many of the
combinations can be excluded from further analysis, for example, because the releases
due to one scenario are far larger than those due to another. Thus, only three
combinations are carried forward to analysis. This process is summarised and discussed
in detail in Assessment of Radionuclide Release Scenarios for the Repository System.
183
Table 7-11. Possible binary combinations of radionuclide release scenarios.
Combinations shown in red are excluded from further analysis for the reasons
summarised in the numbered notes, below. The results of the analysis of the
combinations shown are presented in Chapter 8.
Scenario
Base scenario (BS)
×
Enlarging defect/degradation
of buffer (VS1)
1
×
Corrosion failure following
buffer erosion (VS2)
Retained
for
analysis
3
×
Accelerated insert corrosion
rate (AIC)
1
3
Retained
for
analysis
×
Earthquake and rock shear
(RS)
Retained
for
analysis
2, 3
2
2
×
Rock shear followed by
buffer erosion (RS-DIL)
Retained
for
analysis
3
Retained
for
analysis
Retained
for
analysis
1
×
BS
VS1
VS2
AIC
RS
RS-DIL
Notes:
1: Excluded since the two scenarios mutually exclusive or inconsistent (e.g. relates to an uncertain
process that is assumed significant in one scenario but not in the other).
2: Excluded since combinations involving RS-DIL rather than RS are more penalising, and both include
the same rock shear canister failure mode.
3: Excluded since combinations involving AIC rather than VS1 are more penalising, and both include the
same key process of defect enlargement.
184
185
8
ASSESSMENT OF RADIONUCLIDE RELEASE SCENARIOS
This chapter summarises and presents overall conclusions from the assessment of the
radionuclide release scenarios and calculation cases as defined in Chapter 7. This
includes analyses of radionuclide release and transport in the repository system,
projections of the development of the surface environment, and analyses of potential
radiological impacts on humans, plants and animals. Base, variant and disturbance
scenarios are analysed and uncertainties within those scenarios are investigated by a
range of deterministic calculation cases as well as Monte Carlo simulations.
Probabilistic sensitivity analyses are carried out to assess sensitivities to parameter
values and to explore the consequences of alternative model assumptions. The
assessments of radionuclide release scenarios and cases are presented in full in
Assessment of Radionuclide Release Scenarios for the Repository System and Biosphere
Assessment.
8.1
Analysis of the Reference Case in the base scenario
The definition of the Reference Case for the base scenario for the repository system and
surface environment has been specified in the previous chapter. The radionuclide
release scenarios and cases taken forward to quantitative analysis are listed in Table 7-2.
The results of the analysis are presented below.
8.1.1
Results for the repository system
Figure 8-1 shows the calculated evolution of radionuclide release rates from the
repository near field to the geosphere in the Reference Case, summed over the three
release paths: from the buffer into geosphere fractures intercepting the deposition hole
(F-path); from the buffer to the EDZ of the deposition tunnel and thence into the
geosphere (DZ-path); from the buffer to the tunnel backfill and thence to the geosphere
(TDZ-path). The figure shows the evolution of total release rate, summed over all
calculated radionuclides, and the release rates of the five radionuclides that make the
largest contributions to the total: C-14, Cl-36, Ni-59, I-129 and Cs-135. The peak
release occurs at about 4500 years after closure and is dominated by C-14. At later
times, the release of C-14 decreases due to radioactive decay, so that beyond about
60,000 years the total release is dominated, first by Cl-36, and later by Cs-135 and
I-129. The peak release rate of Ni-59 is smallest of the five, and occurs at around
100,000 years.
186
Figure 8-1. Evolution of the total radionuclide release rate from the repository near
field to the geosphere in the reference case of the base scenario, summed over the F-,
DZ- and TDZ-paths, and the evolution of release rates of C-14, Cl-36, Ni-59, I-129 and
Cs-135, which are the radionuclides that make the largest contributions to the total.
Figure 8-2 shows the near-field release and geosphere release rates normalised with
respect to the radionuclide-specific constraints for releases to the environment defined
in STUK Guide YVL D.5. The figure indicates that during the dose assessment time
window (up to 10,000 years) the normalised activity release is almost four orders of
magnitude below the criterion of one defined in Para 313 of STUK’s Guide YVL D.5;
beyond a few tens of thousands of years the normalised activity release rate decreases to
between five and six orders of magnitude below one.
The limited role of the geosphere in attenuating the peak release rate is related to the
cautiously selected location of the deposition hole containing the defective canister. The
defective canister is, in reality, equally likely to be emplaced in any of the deposition
holes accepted for emplacement. As discussed in Section 8.1.1 of the Assessment of
Radionuclide Release Scenarios for the Repository System, a defective canister
emplaced in the majority of deposition holes would give rise to far lower peak release
rates, since geosphere transport would be sufficiently slow for substantial decay of the
C-14 to occur during transport.
187
Figure 8-2. Evolution of the near-field and geosphere release rates in the reference
case of the base scenario (BS-RC), with the release rates for each radionuclide
normalised with respect to the regulatory nuclide-specific constraints for radioactive
releases to the environment (see Section 2.3.7). For the period beyond the dose criteria
time window (taken to be 10,000 years in this and similar figures), the peak normalised
activity release rate, summed over all radionuclides, should be less than one in order to
satisfy the regulatory constraint on the overall release rate of activity from the
geosphere to the environment (Para 313 STUK Guide YVL D.5) and this is termed the
regulatory geo-bio flux constraint in this Figure and similar Figures in this chapter.
In summary, in the Reference Case, the normalised activity release rate is, at maximum,
about four orders of magnitude below one. The highest activity radionuclide release is
of C-14, which peaks at around 4500 years and then declines due to radioactive decay.
Beyond a few ten thousand years, longer-lived radionuclides − Cl-36, I-129 and Cs-135
− dominate radionuclide release. The dominant migration path is from the buffer
directly into fractures intercepting the deposition hole (the F-path); migration paths in
the EDZ of the deposition tunnel or in the tunnel backfill are less important. For the
chosen location of the defective canister, the geosphere has a limited role in attenuating
the peak release rate. It would have a greater role for the majority of deposition hole
locations, and the peak normalised activity release rate would be correspondingly
reduced.
Since releases occur within the 10,000-year dose assessment time window, the results of
the Reference Case are propagated to the biosphere assessment.
188
8.1.2
Results for the surface environment
This section summarises the results for the Reference Case for the surface environment.
Firstly, the discharge locations to the surface environment are addressed since this is a
key input from the geosphere modelling.
Discharge locations to the surface environment
As discussed above the Reference Case for the repository system analyses a single
defective canister in the repository, located in a relatively unfavourable deposition hole
(position 381 in Figure 8-3) in respect to the activity flux from the geosphere
(Assessment of Radionuclide Release Scenarios for the Repository System, Section 6.2).
Figure 8-3 shows the location of the canister in the repository and the corresponding
discharge locations to the surface environment via the F-, DZ- and TDZ-paths (see
Figure 5-6). Discharge takes place to the sea off the northern coast of the present island.
The discharge locations to the surface environment vary significantly between the F-,
DZ- and TDZ-paths and as a function of time. However, the discharge locations tend to
converge on a relatively limited area at later times (the green points in Figure 8-3).
These considerations have led to the decision to adopt the groundwater flow distribution
at 5000 AD as a basis for the selection of flow-related transport parameters for nearfield release and transport modelling and for geosphere transport modelling in the
Reference Case. Hence, the area covered by the three green points in Figure 8-3 is the
area where the radionuclides are introduced into the landscape model.
Figure 8-3. Discharge locations to the surface environment in the Reference Case via
the F-, DZ- and TDZ-paths, evaluated for groundwater flow conditions at 2000 AD,
3000 AD and 5000 AD. The present day outline of Olkiluoto Island and the layout of the
deposition tunnels considered in the safety case are shown in grey.
189
The key results in the biosphere modelling are the projections of the development of the
surface environment during the first 10,000 years and the potential radiological impacts
on humans, plants and animals living in that environment. The results from analysing
the Reference Case (BSA-RC) for the surface environment base scenario are presented
in Biosphere Assessment, Section 6.2, and summarised below.
Surface environment development
The projection of the development of the terrain and ecosystems in the surface
environment for the Reference Case is presented in detail in Terrain and Ecosystems
Development Modelling. Two illustrative examples are shown in Figures 8-4a and 8-4b.
Figure 8-4a. Surface environment projections for three time steps in the Reference
Case. The approximate location of central part the repository is indicated with a red
circle and the discharge locations with a green circle.
190
Figure 8-4b. Ecosystem projections in the Reference Case in 3520 and 12020 AD.
191
Doses to humans
The screening analysis performed on the geosphere releases in the repository case
BS-RC (Biosphere Assessment, Section 6.1) resulted in 6 of the 11 radionuclides with a
non-zero activity released into the surface environment being screened out from further
analysis with the landscape modelling and subsequent dose calculations. The five
radionuclides propagated all the way through the biosphere modelling chain in BSA-RC
are C-14, Cl-36, Mo-93, Ag-108m and I-129.
The annual doses to representative persons within the most exposed group (Emost_exp)
and among other exposed people (Eother) are presented in Figure 8-5a and 8-5b. The
dose maximum for Emost_exp is 2.0·10-7 mSv and occurs at about year 5000 and the
corresponding dose maximum for Eother is 1.3·10-9 mSv and occurs at about year 4000.
These results are about 6-7 orders of magnitude below the regulatory radiation dose
constraints. As seen in these figures, C-14 dominates the annual doses. This is a direct
effect of that C-14 dominates the geosphere releases in the repository calculation case
BS-RC during the dose assessment time window. Furthermore, the shapes of the dose
curves are more irregular compared with the shape of the release rate curves (see for
example the normalised geosphere release rates in Figure 8-6a). These extra structures
in the dose curves are mainly the effect of the dynamics in the landscape model and that
the dose is calculated by summing exposure pathway-specific contributions from
several biosphere objects. The dynamics in the development of biosphere objects,
especially changes in their geometries, have a strong influence on the resulting activity
concentrations in both environmental media in contaminated biosphere objects and in
the foodstuffs the objects produce. For example, when a lake develops into an
agricultural or terrestrial ecosystem it may lead to a steep increase in the activity
concentration in the shrinking water volume when it dries out.
Doses to plants and animals
The (typical) absorbed dose rate for plants and animals for the calculation case BSA-RC
for the most exposed organisms in freshwater, brackish, semi-aquatic and terrestrial
environments are presented in Figure 8-5c. The dose rate maximum over all organsims
is 2.6·10-7 mikroGy/h, which is observed for Pike in freshwater environment.
192
1,E-06
E_most_exp
BSA‐RC
C-14
I-129
Annual dose [mSv]
1,E-08
Cl-36
Ag-108m
Mo-93
1,E-10
1,E-12
1,E-14
1,E-16
2020
4020
6020
8020
10020
12020
Year
Figure 8-5a. The annual dose to a representative person within the most exposed group
(E_most_exp) and the contributions from each radionuclide for the calculation case
BSA-RC.
1,E-06
E_other
BSA‐RC
C-14
Annual dose [mSv]
1,E-08
I-129
Cl-36
1,E-10
Ag-108m
Mo-93
1,E-12
1,E-14
1,E-16
1,E-18
1,E-20
2020
4020
6020
8020
10020
12020
Year
Figure 8-5b. The annual dose to a representative person among other exposed people
(E_other) and the contributions from each radionuclide for the calculation case
BSA-RC.
193
1.E-06
Absorbed dose rate (microGy/h)
BSA-RC
Pike (freshwater)
Grey Seal (brackish water)
1.E-08
Beaver (semi-aquatic
environment)
Hazel grouse egg
(terrestrial environment)
1.E-10
1.E-12
1.E-14
2020
4020
6020
8020
10020
12020
Year
Figure 8-5c. Absorbed dose rate for plants and animals for the calculation case
BSA-RC for the most exposed organisms in freshwater, brackish, semi-aquatic and
terrestrial environments.
8.2
Analysis of other cases in the base scenario
In the base scenario, four sensitivity cases are defined to take into account: 1) the
uncertainty in the location of the defective canister in the repository system and the
corresponding uncertainty of the discharge location in the surface environment; 2) the
uncertainty in the speciation of silver, molybdenum and niobium radioisotopes; 3) the
uncertainty in the time for the establishment of a transport path between the canister
interior and exterior. See cases BS-LOC1 and BS-LOC2, BS-ANNFF, and BS-TIME in
Table 7-2.
8.2.1
Alternative canister positions BS-LOC1 and BS-LOC2
A canister with an initial penetrating defect is equally likely to be located in any of the
repository deposition holes that are accepted for disposal. The deposition holes vary
widely in their near-field flows, in the transport resistances of the potential radionuclide
migration paths through the host rock and in the discharge locations to the surface
environment. A systematic approach has been used to select the Reference Case
position for a defective canister and also alternative positions that, like the Reference
Case postion, are cautiously chosen with respect to the peak radionuclide release rates to
the surface environment, but discharge to different locations in the surface environment
(Assessment of Radionuclide Release Scenarios for the Repository System). Repository
system model assumptions and parameter values for calculating radionuclide release
rates from the alternative canister positions are identical to those of the Reference Case,
except for flow-related transport parameters that are specific to the deposition hole in
which the defective canister is assumed to be located. The deposition holes chosen as
alternatives to the Reference Case location of the defective canister are positions 2418
(BS-LOC1) and 3829 (BS-LOC2). The results for the geosphere releases are presented
in Figure 8-6a. The results from biosphere assessment for these cases are presented in
Section 8.4, since alternative canister positions lead to alternative discharge locations in
194
the surface environment, which are covered by the surface environment variant scenario
VS(A).
8.2.2
Alternative speciation BS-ANNFF / BSA-ANNFF
In the Reference Case, the only radionuclides considered to migrate in anionic form are
I-129, Cl-36 and Se-79. These anions migrate without retardation by sorption through
accessible pore space in the buffer and backfill, and this pore space is reduced due to
anion exclusion. Repository calculation case BS-ANNFF illustrates the impact of
uncertainties in the speciation of silver, molybdenum and niobium that could be present
in anionic form and thus also migrate in that form.
The assumption of migration in anionic form generally increases the maximum release
rates of the radioisotopes of these elements because they are treated as non-sorbing,
allowing faster migration and thus less time for radioactive decay. This leads to
significant increases in releases of Nb-94, Nb-93m and Mo-93, but the releases of these
radionuclides are still minor compared with the releases of C-14, Cl-36, I-129 and
Cs-135, so that the total radionuclide release rates are almost unchanged.
When the geosphere release rate is normalised with respect to the nuclide-specific
constraints for radioactive releases, the geosphere release rates in case BS-ANNFF are
indistinguishable from those of the Reference Case (Figure 8-6a).
The annual doses to representative persons within the most exposed group (Emost_exp)
and among other exposed people (Eother) for biosphere calculation case BSA-ANNFF
are presented in Figure 8-6b. The dose maximum for Emost_exp is 6.0·10-7 mSv and the
dose maximum for Eother is 2.2·10-7 mSv. It should be noted that the screening analysis
of the geosphere releases of BS-ANNFF results in the screening in of Nb-94, in addition
to the same five radionuclides as in the Reference Case.
The (typical) absorbed dose rate maxima for plants and animals for the calculation case
BSA-ANNFF is 1.0·10-5 mikroGy/h. This is observed for Reindeer Lichen in terrestrial
environments.
8.2.3
Delayed establishment of the transport path BS-TIME / BSA-TIME
In the Reference Case, it is assumed to take 1000 years for water to penetrate the
canister insert and fuel cladding and to contact the fuel and structural materials, and for
a transport pathway to be established between the canister interior and exterior. This is a
cautious assumption. In practice, the slow initial water ingress rate, the decrease of
ingress rate over time (e.g. due the build-up of internal pressure due to hydrogen gas
formation), and the barriers provided by the cast iron insert and the fuel cladding, are
expected to cause a longer delay before establishment of a transport path than assumed
in the Reference Case.
In the repository calculation case BS-TIME, a 5000-year delay is assumed before a
transport path is established between the canister interior and exterior, compared with
1000 years in the Reference Case. The 5000-year delay is arbitrarily chosen from within
the range of uncertainty for the purpose of propagating this case to the biosphere
195
assessment and analysing the radiological impact at that time. All other model
assumptions and parameter values are identical to those of the Reference Case.
The delay in establishing the transport path pushes back the time of first release,
previously at about 4500 years after disposal, to about 10,000 years after disposal.
Consequently, there is a decrease in peak normalised release rate, by about a factor of 2,
due to decay of C-14. Beyond about 20,000 years, the peak normalised geosphere
release rate in the case BS-TIME is indistinguishable from that in the Reference Case
(Figure 8-6a).
The annual doses to representative persons within the most exposed group (Emost_exp)
and among other exposed people (Eother) for biosphere calculation case BSA-TIME are
presented in Figure 8-6b. The dose maximum for Emost_exp is 1.2·10-7 mSv and the dose
maximum for Eother is 1.1·10-10 mSv. The screening analysis of the geosphere releases of
these cases results in only four radionuclides being screened in: C-14, Cl-36, Mo-93 and
I-129.
The (typical) absorbed dose rate maxima for plants and animals for the calculation case
BSA-TIME is 2.0·10-7 mikroGy/h. This is observed for Pike in freshwater
environments.
Figure 8-6a. Evolution of the geosphere release rates in the Reference Case (BS-RC)
and in all the base scenario sensitivity cases, with the release rates for each
radionuclide normalised with respect to the regulatory nuclide-specific constraints for
radioactive releases to the environment.
196
1.E-04
Emost_exp
BSA-RC
Annual dose [mSv]
BSA-TIME
BSA-ANNFF
1.E-06
1.E-08
1.E-10
1.E-12
2020
4020
6020
8020
10020
12020
Year
1.E-04
Eother
BSA-RC
Annual dose [mSv]
BSA-TIME
BSA-ANNFF
1.E-06
1.E-08
1.E-10
1.E-12
2020
4020
6020
8020
10020
12020
Year
Figure 8-6b. The annual doses to representative persons within the most exposed group
(Emost_exp) (top) and among other exposed people (Eother) (below) for the biosphere
Reference Case (BSA-RC), BSA-ANNFF and BSA-TIME (i.e. the three repository
calculation cases in the Base scenario that have the same discharge locations).
8.3
Analysis of the variant scenarios in the repository system
Two variant scenarios are identified that are considered plausible: an enlarging defect
and degradation of the buffer (VS1) and canister failure by corrosion following buffer
erosion (VS2). Sensitivity cases within VS1 investigate alternative assumptions for
porewater chemistry in the near field and geosphere. Sensitivity cases within VS2
investigate alternative locations for the failed canister.
8.3.1
Cases in Variant Scenario 1 (VS1)
Three calculation cases are analysed within this scenario, VS1-BRACKISH, VS1HIPH-NF and VS1-HIPH, which consider the influence of groundwater composition
(brackish or highly alkaline) on radionuclide releases.
197
Figure 8-7 shows the evolution of near-field release rates via the F-path, the DZ-path
and the TDZ-path in the Reference Case and in case VS1-BRACKISH.
The influence of groundwater composition (brackish or highly alkaline) on radionuclide
releases is shown to be relatively minor (see Figure 8-8a), and this can be explained in
terms of the differences in retention parameters. The maximum normalised release rate
to the surface environment occurs shortly after the dose criteria time window, and is
about an order of magnitude higher than in the Reference Case, but still almost three
orders of magnitude below the regulatory geo-bio flux constraint (Figure 8-8a). The
difference compared with the Reference Case is accounted for mainly by the assumption
in VS1 of an enlarging defect, rather than by the perturbation to the buffer, as shown in
a complementary analysis reported in Section 12.1.2 of Assessment of Radionuclide
Release Scenarios for the Repository System.
Figure 8-7. Evolution of radionuclide release rates from the near field via the F-, DZand TDZ-paths in the Reference Case and in case VS1-BRACKISH.
198
Figure 8-8a. Evolution of the geosphere release rates for the VS1 cases and for the
Reference Case (BS-RC), with the release rates for each radionuclide normalised with
respect to the regulatory nuclide-specific constraints for radioactive releases to the
environment.
The annual doses to representative persons within the most exposed group (Emost_exp)
and among other exposed people (Eother) for biosphere calculation cases
BSA-BRACKISH, BSA-HIPH, BSA-HIPH-NF are presented in Figure 8-8b. The dose
maximum for Emost_exp is 7.2·10-6 mSv and the dose maximum for Eother is 3.9·10-6 mSv
for the case BSA-BRACKISH. The dose maximum for Emost_exp is 1.9·10-6 mSv and the
dose maximum for Eother is 1.7·10-7 mSv for the case BSA-HIPH. The dose maximum
for Emost_exp is 1.5·10-6 mSv and the dose maximum for Eother is 1.5·10-9 mSv for the case
BSA-HIPH-NF.
The (typical) absorbed dose rate maxima for plants and animals for the calculation case
BSA-BRACKISH is 1.2·10-4 mikroGy/h. This is observed for Reindeer Lichen in
terrestrial environments. The (typical) absorbed dose rate maxima for plants and
animals for the calculation case BSA-HIPH is 2.2·10-5 mikroGy/h. This is observed for
Reindeer Lichen in terrestrial environments. The (typical) absorbed dose rate maxima
for plants and animals for the calculation case BSA-HIPH-NF is 2.6·10-6 mikroGy/h.
This is observed for Pike in freshwater environments.
199
1.E-04
Emost_exp
BSA-RC
Annual dose [mSv]
BSA-BRACKISH
BSA-HIPH_NF
1.E-06
BSA-HIPH
1.E-08
1.E-10
1.E-12
2020
4020
6020
8020
10020
12020
Year
1.E-04
Eother
BSA-RC
Annual dose [mSv]
BSA-BRACKISH
BSA-HIPH_NF
1.E-06
BSA-HIPH
1.E-08
1.E-10
1.E-12
2020
4020
6020
8020
Year
10020
12020
Figure 8-8b. The annual doses to representative persons within the most exposed group
(Emost_exp) (top) and among other exposed people (Eother) (below) for the biosphere
Reference Case (BSA-RC), BSA-BRACKISH, BSA-HIPH and BSA-HIPH-NF.
8.3.2
Cases in Variant Scenario 2 (VS2)
In Variant Scenario 2 (VS2) there are no initial penetrating defects in the canisters.
Rather, it is assumed that chemical erosion of the buffer takes place due to low ionic
strength water penetrating to repository depth, most likely in association with ice-sheet
retreat during periods of increased flow rates. Significant buffer erosion is considered
unlikely, but cannot currently be excluded in at least some of the deposition holes. Four
calculation cases are analysed within this scenario, VS2-H1, VS2-H2, VS2-H3 and
VS2-H4, which each consider the failure of a single canister at a given time and location
in the repository.
Failure within the assessment time frame occurs in those deposition holes that have
relatively high near field flows. The four calculation cases analysed for the VS2
scenario correspond to the four canisters that are calculated to fail within this time frame
based on a single realisation of the DFN groundwater flow model and a specific
200
(reference) set of groundwater flow model assumptions, as described in Performance
Assessment. The number and timing of canister failures are regarded as a reasonable
illustration, but are not to be interpreted as a precise prediction. Indeed, Appendix 2 of
Formulation of Radionuclide Release Scenarios presents arguments that suggest that
penetration of low ionic strength water to repository depth will probably not occur, and
so there will be no canister failures by corrosion following chemical erosion of the
buffer. Currently, however, a few canister failures in this scenario cannot be ruled out.
High near-field flows are correlated with low transport resistances in the geosphere
facture network. Thus, the deposition holes where canister failures might occur will, on
average, be less favourable than most with respect to geosphere transport. However, in
some instances, transport paths that leave a deposition hole via a fracture may
subsequently re-enter the engineered barrier system and there be subject to much slower
flows. This is, by chance, the case for calculation cases VS2-H2, -H3 and -H4 (although
not for VS2-H1). As a result, there is no radionuclide release from the geosphere in
these calculation cases within the one million year assessment time frame.
Figure 8-9 shows the normalised geosphere release rates in case VS2-H1. The main
radionuclide contributing to the normalised release is I-129, with smaller contributions
from Cl-36 and Se-79, i.e. all long-lived non-sorbing radionuclides. The modelled
geosphere release rates show periodic maxima, due to relatively rapid flushing of these
non-sorbing radionuclides from the geosphere during periods of high flow (as an ice
sheet retreats over the site). The maximum normalised release rate is more than three
orders of magnitude below the criterion of one.
This analysis considers only one transport path through the geosphere, in one realisation
of the geosphere fracture network, and higher peak normalised releases could
potentially arise if canister failure occurred in other deposition holes with more rapid
transport paths through the geosphere. Transport path uncertainties, as well as the
potential effects of irreversible uptake of radionuclides on bentonite colloids and of the
potential release of intrinsic colloids in this scenario are considered in complementary
analyses reported in Section 12.2 of Assessment of Radionuclide Release Scenarios for
the Repository System.
Overall, it is concluded that the low peak normalised release rates calculated for a single
failed canister in scenario VS2 indicate that the few canister failures that could
potentially arise in the more likely lines of evolution (or even the few tens of failures
that are calculated to occur in the Performance Assessment if highly pessimistic
assumptions are adopted) could easily be tolerated without exceeding the normalised
radionuclide release constraint of one.
201
.
Figure 8-9. Evolution of geosphere release rates for VS2-H1, with the release rates for
each radionuclide normalised with respect to the regulatory geo-bio flux constraints.
VS2-H2, -H3 and -H4 give no release to the surface environment within the one million
year assessment time frame. The figure shows calculation over four glacial cycles from
0.6 Ma to 1 Ma (with 3 glacial episodes in each cycle).
8.4
Analysis of the variant scenarios in the surface environment
The results for the analysis of all variant scenarios calculation cases in the surface
environment are presented in detail in Biosphere Assessment to which the interested
reader is encouraged to consult. Here the presentation is limited to the results from
analysing variant scenario VS(A), addressing uncertainties in the discharge locations to
the surface environment. The annual doses to representative persons within the most
exposed group (Emost_exp) and among other exposed people (Eother) for biosphere
calculation cases VS(A)-SOUTH1 and VS(A)-SOUTH2 are presented in Figure 8-10.
The dose maxima for Emost_exp are 6.2·10-5 mSv 1.7·10-4 mSv for the cases VS(A)SOUTH1 and VS(A)-SOUTH2, respectivelly. The dose maxima for Eother are
5.6·10-6 mSv 1.2·10-5 mSv for the cases VS(A)-SOUTH1 and VS(A)-SOUTH2,
respectivelly.
The (typical) absorbed dose rate maxima for plants and animals for the calculation case
VS(A)-SOUTH1 is 5.7·10-5 mikroGy/h. This is observed for Mallard in freshwater
environments. The (typical) absorbed dose rate maxima for plants and animals for the
calculation case VS(A)-SOUTH2 is 1.3·10-4 mikroGy/h. This is observed for Mallard
in freshwater environments.
202
1.E-03
BSA-RC
Annual dose [mSv]
VS(A)-SOUTH1
VS(A)-SOUTH2
1.E-05
1.E-07
1.E-09
1.E-11
2020
4020
6020
8020
10020
12020
Year
1.E-03
BSA-RC
Annual dose [mSv]
VS(A)-SOUTH1
VS(A)-SOUTH2
1.E-05
1.E-07
1.E-09
1.E-11
2020
4020
6020
8020
10020
12020
Year
Figure 8-10. The annual doses to representative persons within the most exposed group
(Emost_exp) (top) and among other exposed people (Eother) (below) for the biosphere
Reference Case (BSA-RC), VS(A)-SOUTH1 and VS(A)-SOUTH2.
8.5
Analysis of the disturbance scenarios in the repository system
Three disturbance scenarios are identified that are considered unlikely: accelerated
corrosion of the iron insert (AIC); rock shear leading to canister failure (RS); rock shear
leading to canister failure followed by buffer erosion (RS-DIL). A set of what-if cases
are considered within each of these scenarios assessing the condition of the insert (AIC
cases) and time of rock shear failure and buffer erosion.
8.5.1
Cases in the accelerated iron insert corrosion AIC scenario
The accelerated insert corrosion rate (AIC) scenario considers the possibility that an
initial penetrating defect in a canister becomes enlarged over time, due e.g. to faster
than expected corrosion of the insert, the corrosion products of which will occupy a
larger volume than the original metal. This possibility has also been considered in the
Monte Carlo simulations and in the VS1 variant scenario. In the VS1 scenario, the hole
203
grows gradually (which is more likely than rapid or instantaneous enlargement) between
1000 years and 25,000 years, and the peak release is about an order of magnitude higher
than in the Reference Case (Figure 8-8a). The AIC scenario is yet more pessimistic,
assuming a much higher than expected insert corrosion rate leading to a sudden
(instantaneous) loss of transport resistance of the canister after 15,000 years. The
analysis of this scenario focuses on the significance of whether or not a transport path
between the canister internal void space and the buffer exists prior to defect
enlargement. Two cases have been considered: case AIC-LI (leaky insert) and case
AIC-TI (tight insert).
In case AIC-LI, the defective canister evolves as in the Reference Case for the first
15,000 years; in case AIC-TI, there is no radionuclide transport path from the canister
internal void space to the buffer for the first 15,000 years. In both cases, near-field and
geosphere releases increase rapidly from around the time of assumed loss of transport
resistance (15,000 years). The releases peak somewhat higher in AIC-LI than in AICTI; this can be interpreted as due to the additional release, e.g. of C-14 from corrosion
of the other metal parts (fractional corrosion rate 1.0 × 10-3 per year) to the void space
in the canister interior in case AIC-LI before defect enlargement. The evolution of nearfield and geosphere releases after 15,000 years is virtually the same in AIC-LI and in
AIC-TI.
Figure 8-11 shows the overall geosphere release rates in the two AIC-cases and in the
Reference Case, with the release rates for each radionuclide normalised with respect to
the regulatory geo-bio flux constraints. In both the AIC-LI and AIC-TI cases, the
maximum normalised release rate is about three orders of magnitude higher than in the
Reference Case, but remains more than an order of magnitude below the regulatory geobio flux constraint.
204
Figure 8-11. Evolution of the geosphere release rates for the AIC-cases and for the
Reference Case (BS-RC), with the release rates for each radionuclide normalised with
respect to the regulatory geo-bio flux constraints.
8.5.2
Cases in the rock shear RS scenario
Olkiluoto is located in the Fennoscandian Shield away from active plate margins and is
currently seismically quiet. However, the possibility of large earthquakes especially at a
time of ice-sheet retreat cannot totally be excluded.
The rock shear (RS) scenario considers the possibility of canister failure due to
secondary shear movements on fractures intersecting the deposition holes in the event of
a large earthquake. Rock shear is assumed to adversely affect the flow-related transport
properties of the fracture intersecting the deposition hole where canister failure occurs.
The buffer, however, is assumed to continue to fulfil its safety functions. Two cases are
analysed: RS1 and RS2, in which rock shear and canister failure are assumed to occur at
40,000 and 155,000 years after disposal, respectively.
In both cases, upon canister failure, there is an almost immediate release of IRF
radionuclides, such as I-129 and Cl-36, and a more gradual release of e.g. Ni-59, which
is released congruently with the corrosion of zirconium alloy and other metal parts.
Later, the highest near-field and geosphere release rates are due to Ra-226. Figure 8-12
shows the single-canister normalised geosphere release rates in the RS1 and RS2 cases
and in the Reference Case. 1000-year centred moving averaging has been applied,
which is in accord with Finnish regulations (STUK Guide YVL D.5), and this reduces
the sharp pulses that occur at times of increased groundwater flow. Peak normalised
release rates, with and without averaging, are also shown. The highest normalised peak
release rates from the geosphere are in both cases more than two orders of magnitude
below one.
205
Figure 8-12. Evolution of the single-canister normalised geosphere release rates in the
RS1- and RS2-cases and in the Reference Case (BS-RC). 1000-year centred moving
averaging has been applied to the RS1 and RS2 curves, consistent with STUK Guide
YVL D.5.
According to Performance Assessment, up to some tens of canisters could potentially
fail in the event of a large earthquake. The overall average annual probability of such an
earthquake is around 10-7. Taking into account that the probability is not constant in
time, but is greatest following a period of ice-sheet retreat, and assuming that all
canisters that could potentially fail do in fact fail, the peak expectation value of the
normalised release rate in the RS scenario (i.e. the peak probability-weighted
normalised release rate) is at least around two orders of magnitude below the regulatory
guideline. This implies that more than one hundred canisters would have to fail
simultaneously before the regulatory geo-bio flux constraint is exceeded. This number
exceeds the few tens of canisters estimated to be in critical positions that are vulnerable
to failure in the event of a large earthquake (see section 6.3.1). Furthermore, the peak
expectation value of the normalised release rate (i.e. the peak probability-weighted
normalised release rate) taking into account the uncertain timing of the earthquake
leading to canister failure and assuming that all canisters that could potentially fail do in
fact fail, is also about two orders of magnitude or more below the regulatory guideline.
206
8.5.3
Case for the rock shear followed by buffer erosion in the RS-DIL
scenario
In the scenario of rock shear followed by buffer erosion (RS-DIL), the buffer is
assumed to undergo either immediate damage due to a rock shear event that causes
canister failure or longer-term erosion due to the penetration of low-ionic strength water
to repository depth. As the buffer erodes, radionuclides sorbed onto the surfaces of the
bentonite are released to the geosphere, either in solution or associated with bentonite
colloids. Once sufficient buffer erosion has taken place, advective conditions are
established between the canister interior and the geosphere, as in the scenario of
corrosion failure following buffer erosion (scenario VS2). RS-DIL is also taken to
encompass cases of rock shear where the buffer undergoes more minor perturbation due
to the rock shear event (deformation, local thinning).
The peak release rates for RS-DIL cases are about an order of magnitude higher than RS
cases, but the peak normalised release rate from a single failed canister is still more than
an order of magnitude below the regulatory geo-bio flux constraint (Figure 8-13).
Taking into account that the probability of rock shear events is not constant in time, and
assuming that all canisters that could potentially fail do in fact fail, the peak expectation
value of the normalised release rate in the RS-DIL scenario (i.e. the peak probabilityweighted normalised release rate) is around an order of magnitude below the regulatory
guideline.
The potential effects of irreversible uptake of radionuclides on bentonite colloids in this
scenario and of the potential release of intrinsic colloids in this and in the RS scenario
are considered in complementary analyses reported in Section 12.4 of Assessment of
Radionuclide Release Scenarios for the Repository System.
8.6
Analysis of the disturbance scenarios in the surface environment
The results for the analysis of all disturbance scenarios calculation cases in the surface
environment are presented in detail in Biosphere Assessment to which the interested
reader is encouraged to consult. Here only the results for the human intrusion scenario
are presented.
207
Figure 8-13. Evolution of the geosphere release rates in the cases RS1-DIL and
RS2-DIL and in the Reference Case (BS-RC), with the release rates for each
radionuclide normalised with respect to the regulatory nuclide-specific constraints for
radioactive releases to the environment. 1000-year centred moving averaging has been
applied to the RS1-DIL and RS2-DIL curves. The figure illustrates the selection of
normalised release rate FA, which is representative of the peak normalised release at
the time of canister failure, and FB, which is representative of later peaks at times of
ice-sheet retreat. For multiple canister failures, it also shows the probability-weighted
normalised release rate, taking into account the uncertain timing of the earthquake
leading to canister failure and assuming that all canisters that could potentially fail do
in fact fail.
208
8.6.1
Cases in the human intrusion scenario (DS(F)-HI)
Scenarios for inadvertent human intrusion caused by borehole drilling have been
formulated (see Section 7.3.4). Expectation values of (effective) doses to drilling
technicians and site geologists have been derived based on a stylised approach to the
dose calculations and estimation of indicative annual probabilities of the intrusion
occurring.
The peak expectation value of the dose in DS(F)-HI-CANISTER is around an order of
magnitude below the regulatory radiation dose constraint for the most exposed people
(Figure 8-14). The peak expectation value of the dose in DS(F)-HI-BUFFER and
DS(F)-HI-BACKFILL is several orders of magnitude below the regulatory radiation
dose constraint for the most exposed people; the detail results are found in Biosphere
Assessment.
Figure 8-14. Expectation value of the effective dose for the calculation cases
DS-HI-CANISTER-D (top) and DS-HI-CANISTER-G (below), including only the
radionuclides that at any time point exceeds a value of 10-6 mSv.
209
8.7
Complementary analyses
Uncertainties in the initial state of the repository system and in its evolution have been
taken into account in formulating the base, variant and disturbance scenarios.
Complementary analyses have been carried out to further investigate uncertainties
related to the model assumptions and parameter values. The complementary analyses
include deterministic analyses of complementary calculation cases, as well as scoping
calculations, Monte Carlo simulations and a probabilistic sensitivity analysis. The aim
of these analyses is to develop a better understanding of the modelled system or
subsystems. One key complementary analysis − the analysis of more than one defective
canister being present in the repository − is summarised below. After that, the results of
the Monte Carlo simulations and probabilistic sensitivity analysis for a single defective
canister are also summarised. These and other complementary analyses are described in
full in Assessment of Radionuclide Release Scenarios for the Repository System.
8.7.1
More than one defective canister in the repository
It is possible that there could be more than one canister with an initial penetrating defect
in the repository, and more than one of these canisters could be unfavourably located.
An illustrative probabilistic model for the reliability of the spent nuclear fuel final
disposal canister manufacture and testing has been developed by Holmberg & Kuusela
(2011), and this provides the basis for an illustrative probabilistic analysis of the
consequences of random emplacement of one or more defective canisters in the
repository.
To illustrate the consequences, the following quantities have been evaluated for three
example key radionuclides (I-129, Cl-36 and C-14):

the geo-bio release rate averaged over multiple realisations, where, in each
realisation, the number of defective canisters is sampled from the probability
distribution given in Holmberg & Kuusela (2011), and these defective canisters are
placed randomly in the repository;

the probability that the maximum activity release rate from the geosphere as defined
above, exceeds the maximum activity release rate from the geosphere due to a single
defective canister in the Reference Case.
It is shown that the expectation value of the release rate from multiple defective
canisters randomly located in the repository is significantly less than the release rate in
the Reference Case, in which a single failed canister at a cautiously selected location is
postulated. This is because of the low probability of there being many defective
canisters and also because the majority of deposition holes have flow-related transport
properties that are significantly more favourable to limiting releases than the deposition
hole considered in the Reference Case. Furthermore, the probability that the release rate
maximum from multiple randomly-placed defective canisters exceeds the Reference
Case release rate maximum is low − estimated to be about 0.04 %. This provides
support for the Reference Case assumption of there being one defective canister in the
repository, unfavourably located.
210
8.7.2
Monte Carlo analyses and probabilistic sensitivity analysis
Monte Carlo analyses and a probabilistic sensitivity analysis have been performed to
assess the impact of uncertainty and variability in model parameters and to identify the
model parameters most responsible for the uncertainty in model outputs.
Case description and data for Monte Carlo simulations
Monte Carlo simulations with 10,000 realisations and probabilistic sensitivity analyses
(PSAs) have been carried out for two cases:
1. the “hole forever” case, where the initial penetrating defect in the canister overpack
remains unchanged over time. (The Reference Case can be viewed as one specific
realisation of the hole-forever case);
2. the “growing hole” case, where the defect becomes enlarged over time. This is
represented by assuming that the transport resistance provided by the defect is lost
entirely and instantaneously after a period that is varied between 5000 and 50,000
years after disposal.
The model outputs analysed are the activity release rates from the near field to the
geosphere and from the geosphere to the surface environment, normalised with respect
to the regulatory nuclide-specific constraints for radioactive releases to the environment
given in STUK Guide YVL D.5.
To perform the Monte Carlo simulations, probability density functions (PDFs) were
assigned to each of the model parameters judged to be affected by uncertainty, several
of which (solubilities, distribution coefficients, etc.) are element specific. A total of 160
parameter PDFs were defined in the hole forever case, plus an additional two
parameters (Time to loss of hole resistance and Length of canister failed) for the
growing hole case. The PDFs were chosen to provide a reasonable representation of the
full ranges of uncertainty and variability in the input data. The input data used and the
process followed to create the PDFs are presented in Cormenzana (2013b). The
parameters sampled are summarised in Table 8-1.
Consideration of parameter correlations in Monte Carlo simulations is potentially
important as their omission could, in principle, lead to extreme results if unreasonable
combinations of parameter values are sampled, which in turn could bias the results of
the PSA. This topic has been investigated in Cormenzana (2013b), where Monte Carlo
simulations are performed using both correlated and uncorrelated values for all the nearfield flows and geosphere flow-related parameters. The results of the Monte Carlo
simulations (expressed as cumulative distribution functions of the peak release rates
from the near field and to the biosphere of the main radionuclides and the whole
inventory) have, however, been found to be quite similar with and without correlations,
and only results without the inclusion of correlations are presented below.
211
Table 8-1. Summary of sampled parameters in the Monte Carlo simulations. Details in
Cormenzana (2013b).
Related to
Sampled parameters *
Canister
failure
Time to creation of transport pathway from canister interior to exterior. Diameter of defect
in copper shell.
Effective diffusion coefficient in the small hole.
Time to loss of hole resistance.
Length of canister lateral surface assumed to disappear when transport resistance is lost.
The spent
nuclear fuel
Instant release fractions (IRFs) of elements within each of the fuel matrix, zirconium alloy,
other metal parts and crud.
RN release rate from the fuel matrix, zirconium alloy and other metal parts.
Canister
interior
Volume of water in the canister interior.
Mass of buffer that enters into the canister interior.
Solubility of elements inside the canister.
Buffer
Porosity accessible to anions.
Effective diffusion coefficients.
Distribution coefficients.
Solubility correction factors.
Groundwaterbuffer
interface
Solubility correction factors (in damaged rock around the deposition hole)
Deposition
tunnel backfill
Porosity accessible for anions.
Effective diffusion coefficients.
Distribution coefficients.
Solubility correction factors.
Length of tunnel from the deposition hole to the exit fracture.
Near-field
flows
Equivalent water flows through the fracture that intersects: (a) the deposition hole, (b) the
deposition tunnel EDZ and (c) the deposition tunnel.
Water flux in the tunnel backfill
Geosphere
flow
parameters
Groundwater travel time, Transport resistance and Length for the F-path, for the DZ-path
and for the TDZ-path.
Peclet number in the 1-D geosphere paths.
Rock matrix
Accessible porosity, Effective diffusion coefficients and Maximum penetration depth in the
rock matrix.
Distribution coefficients.
*
Note that IRFs, solubilities, solubility correction factors and distribution coefficients (Kd) are each
element specific.
Results of the Monte Carlo simulations
Figure 8-15 shows the total normalised release rate to the surface environment for the
hole-forever and growing-hole cases, summed over the F-, DZ-, and TDZ-paths and
over all the calculated radionuclides. The results shown are the mean release, the 1st and
99th percentiles, the 5th and 95th percentiles, the median (50th percentile), and
maximum values as functions of time. The figure shows that, at any time, the
uncertainty in the value of the total normalised release rate is at least of four orders of
magnitude.
In the hole-forever case, the peak total normalised release rate to the surface
environment is more than two orders of magnitude below the regulatory geo-bio flux
212
constraint in all realisations. The peak normalised release rate of around 2 × 10-4 in the
Reference Case (also shown in Figure 8-15, upper graph) lies between the 95th and 99th
percentiles in the corresponding Monte Carlo simulation. This confirms that the
Reference Case uses a cautiously selected set of parameter values, giving a peak release
rate that is towards the upper end of the range calculated in the Monte Carlo simulation.
In the growing-hole case, the peak release rate also remains below the regulatory
constraint, although in some realisations the margin is less than an order of magnitude.
The model assumption of the growing-hole case that the transport resistance of the
defect is lost entirely and instantaneously at a given time is unrealistic and hypothetical.
Thus, the “growing-hole” case, as modelled here, can be considered as a bounding case
since the rate of enlargement of the defect over time, if it enlarges at all, is uncertain.
(The more realistic assumption of a gradually enlarging defect is considered in the
variant scenario VS1, see Section 8.3.1)
Figure 8-16 shows the mean normalised release rates to the surface environment in the
hole-forever and growing-hole cases, including the contributions of different
radionuclides. As in the Reference Case, C-14, Cl-36, I-129 and Cs-135 dominate the
release at different intervals within the assessment time window. In both the holeforever and the growing-hole cases, C-14 is the most important radionuclide
contributing to the peak release rate, followed by Cl-36 and I-129, with Cs-135 playing
a much smaller role. In the growing-hole case, the actinides and progeny are important
in the long term, while in the hole-forever case their contribution is negligible.
Overall, the Monte Carlo simulations show that radionuclides controlling the total
releases from the near field and to the biosphere are those with little or no sorption on
the buffer, backfill and rock matrix. Radionuclides with strong sorption on the buffer,
backfill and unaltered rock make practically no contribution to the peak total releases
from the near field and from the geosphere. These results are consistent with those for
the Reference Case.
213
1.E+00
Normalised release rate to the biosphere
1.E-01
Total
95 percentile
1.E-02
1.E-03
Regulatory geo-bio
flux constraint
th
Reference Case
Hole forever
1.E-04
th
1.E-05
1.E-06
99 percentile
Mean
Maximum
value
Median
th
50 percentile
1.E-07
1.E-08
th
5 percentile
1.E-09
1.E-10
1.E+01
st
1 percentile
1.E+02
1.E+03
1.E+04
1.E+05
Minimum
value
1.E+06
Time (years)
1.E+00
Normalised release rate to the biosphere
1.E-01
Total
th
95 percentile
1.E-02
1.E-03
Regulatory geo-bio
flux constraint
Growing hole
1.E-04
th
1.E-05
1.E-06
99 percentile
Maximum
value
Mean
1.E-07
Median
50th percentile
1.E-08
th
Minimum
value
5 percentile
1.E-09
1.E-10
1.E+01
st
1 percentile
1.E+02
1.E+03
1.E+04
1.E+05
1.E+06
Time (years)
Figure 8-15. Total normalised release rate to the surface environment in two Monte
Carlo cases with 10,000 realisations (hole-forever and growing-hole cases). The
Reference Case release rate is also shown in the upper figure.
214
Mean normalised release rate to the biosphere
1.E+00
Regulatory geo-bio
flux constraint
1.E-01
Hole forever
1.E-02
Rest of fission and
activation products
1.E-03
1.E-04
Actinides and
progeny
Total
I-129
1.E-05
1.E-06
Cl-36
1.E-07
C-14
Cs-135
1.E-08
1.E-09
1.E-10
1.E+01
1.E+02
1.E+03
1.E+04
Time (years)
1.E+05
1.E+06
Mean normalised release rate to the biosphere
1.E+00
Regulatory geo-bio
flux constraint
1.E-01
Growing hole
1.E-02
Rest of fission and
activation products
1.E-03
1.E-04
Total
Actinides and
progeny
I-129
1.E-05
C-14
1.E-06
1.E-07
Cl-36
Cs-135
1.E-08
1.E-09
1.E-10
1.E+01
1.E+02
1.E+03
1.E+04
Time (years)
1.E+05
1.E+06
Figure 8-16. Mean normalised release rates to the surface environment in Monte Carlo
cases with 10,000 realisations. Results for the most important radionuclides, the other
fission and activation products, the actinides and their progeny, and the total inventory.
215
Probabilistic sensitivity analysis (PSA)
The analysis of the PSA of the parameters used in the base scenario provides a rich
source of understanding of the sensitivity of model output to variation in input
parameter values, allowing the most important parameters and parameter combinations
to be determined. The most useful of the calculated sensitivity measures have proven to
be the rank correlation coefficients (RCCs), which identify the parameters whose
uncertainties have the greatest effect on the spread (uncertainty) in the model output.
These are identical to standardised rank regression coefficients (SRRCs) when all the
uncertain input parameters are independent (uncorrelated), a condition fulfilled for the
Monte Carlo simulations described here.
Of the graphical methods used, scatter plots and mean rank plots have proved
particularly informative. Realisations can be ranked according to the value taken by a
given model parameter. The mean rank of realisations in a subset giving the highest
10 % or lowest 10 % values of an output variable (e.g. the peak release rates of a given
radionuclide) indicates whether the parameter tends to take high or low values within
that subset. Figure 8-17 shows the example of Cs-135 peak release to the biosphere in
the hole-forever case as an output variable. The figure shows that the realisations with
the highest Cs-135 peak release rate to the biosphere are characterised, for example, by
low values of the distribution coefficient of Cs in the buffer and host rock (Kd(Cs)) and
high values of the defect (small hole) diameter and spent nuclear fuel alteration rate.
10000
9000
8000
Canister
failure
Hole forever
Waste and
canister
interior
Lowest 10%
Highest 10%
Kd(Cs)
WL/Q for F-path
Small hole
diameter
WL/Q for DZ-path
7000
Peclet
number
QF
Mean Rank
Fuel alteration rate
QDZ
6000
5000
4000
De in the
small hole
De(Cs) in backfill
WL/Q for TDZ-path
3000
2000
De
Kd(Cs) in buffer
Mass of buffer
in the cavity
Buffer
and
backfill
Near
field
flows
1000
Geosphere
flow
parameters
Unaltered
rock
0
Parameters
Figure 8-17. Mean ranks for all the parameters in the 10 % of realisations with the
highest/lowest Cs-135 peak release rate to the surface environment. Using the mean
ranks of the (more than one hundred) input parameters that are known to have no effect
on the output variable, it is possible to identify the range of values of the mean rank that
is not statistically significant: this is the grey band in the figure.
216
8.8
Combinations of repository radionuclide release scenarios
Possible combinations of radionuclide release scenarios are considered in Section 7.4,
and four scenario combinations are identified that require further analysis:
1. The base scenario (BS) in combination with corrosion failure following buffer
erosion (VS2) or in combination with the rock shear scenarios (RS and RS-DIL).
The analysis demonstrates that the increase in peak normalised release rate that occurs if
the canister affected by corrosion failure or failure due to rock shear is assumed to have
an initial penetrating defect is more than offset by the reduction that occurs when this
peak is multiplied by the very low probability of occurrence of these combined
scenarios. The combined scenarios are thus less penalising than the VS2, RS or RS-DIL
scenarios alone.
2. Accelerated insert corrosion rate (AIC) in combination with rock shear followed by
buffer erosion (RS-DIL).
The time of enlargement of the defect, if it occurs at all, is highly uncertain. Analysis
shows that to have any impact on the peak release due to the RS-DIL scenario, the
enlargement would need to occur shortly (i.e. within a few thousand years) before one
of the periods of increased water flow associated with ice-sheet retreat considered in the
RS-DIL scenario. The impact of this low probability occurrence would then be limited,
dominated by I-129 and Cl-36, since C-14 will have decayed to insignificance.
3. Corrosion failure following buffer erosion (VS2) in combination with rock shear
followed by buffer erosion (RS DIL).
The peak normalised releases in cases VS2-H1 and RS1-DIL both coincide with the
same period of increased groundwater flow associated with a particular ice-sheet retreat
period at around 600,000 years (see Figures 8-9 and 8-13). The peak releases can thus
simply be summed to give the peak release in the combined scenario. The combined
peak release is, however, dominated by the RS1-DIL component, the peak due to VS2H1 being around an order of magnitude smaller.
4. Accelerated insert corrosion rate (AIC) in combination with corrosion failure
following buffer erosion (VS2).
The peak normalised release in case AIC-LI is almost two orders of magnitude higher
than that of case VS2-H1. The AIC-LI release declines to insignificant levels by the
time that the canister fails in case VS2-H1. Thus, combining the two cases does not
affect the magnitudes of the peak releases from either.
8.9
Summary of safety assessment results and uncertainties
The scenarios analysed have considered a range of events and conditions under which
releases might occur. A range of calculation cases has been analysed for each scenario.
Case assumptions have been applied within each scenario that include taking
pessimistic views on the severity of initiating events, subsequent degradation of
engineered barriers and migration paths from defective or damaged canisters. Cautious
combinations of scenarios have also been analysed. Parameter uncertainties have been
217
investigated most thoroughly for the base scenario for radionuclide release using Monte
Carlo simulations and PSA to complement deterministic analyses. Model results are
found to be consistent with scientific understanding and indicate robust attenuation and
delay of radionuclide releases in the event of canister failures.
A set of biosphere calculation cases have been analysed for scenarios taking into
account uncertainties in the radionuclide releases discharge locations to the surface
environment, the development of the surface environment, the radionuclide transport in
the surface environment and in the dose calculations. The results are reported in
Biosphere Assessment. In this report only a sub-set of all results are presented.
8.9.1
Geosphere release rates
Figure 8-18 shows the peak normalised activity release rates from the geosphere to the
surface environment, and timing of the peak rates, for the base scenario, variant scenario
and disturbance scenario cases. All these results refer to failure of a single canister. For
cases RS1, RS2, RS1-DIL and RS2-DIL, 1000-year moving averaging has been applied
before calculation of the peak rates, consistent with STUK Guide YVL D.5.
Figure 8-18. Peak normalised geosphere release rates for all calculation cases within
the base, variant and disturbance scenarios, each assuming the failure of a single
canister. Colours are used to group cases by scenario. * indicates that 1000 year
averaging is applied, in these cases. The right hand subfigure shows ranges of values
for the peak probability-weighted normalised release rates in the RS and RS-DIL
scenarios. These ranges arise due to uncertainties in the numbers of canisters failing
due to rock shear, as well as in the timing of failure.
218
The lowest peak normalised release rates are for the Reference Case (BS-RC) and
sensitivity cases within the base scenario. In all cases, peak normalised release rates to
the surface environment are below the regulatory geo-bio flux constraint by around an
order of magnitude or more.
In any of these scenarios, more than one canister may fail. In the case of RS and RSDIL, quantitative estimates have been made of ranges of values for the peak probabilityweighted normalised release rates (see the right hand subfigure in Figure 8-18). These
ranges arise due to uncertainties in the numbers of canisters failing due to rock shear, as
well as the timing of failure. They are around an order of magnitude or more below the
regulatory geo-bio flux constraint. For the VS2 scenario, arguments are presented in
Formulation of Radionuclide Release Scenarios report that suggest that penetration of
low ionic strength water to repository depth will probably not occur, and so there will be
no canister failures by corrosion following chemical erosion of the buffer. Currently,
however, a few canister failures in this scenario cannot be ruled out. The results shown
in Figure 8-18 suggest that multiple failures could be tolerated without the regulatory
geo-bio flux constraint being exceeded. Other scenarios, including the base scenario,
postulate the existence of one or more canisters with initial penetrating defects. The
currently available data are insufficient, even when expert judgement is used, to make a
reasonable estimate of the probability of emplacing a defective canister in the
repository. However, with additional data on the welding process and continued
development of the NDT process, it seems practicable in the future to show that the
probability of more than one initially defective canister in the repository is less than one
per cent. Furthermore, even if there were more than one initially defective canister in
the repository, the likelihood that more than one of these would be placed in locations in
the repository as unfavourable as that assumed for the defective canister in the
Reference Case is very low. Most locations give peak release rates that are orders of
magnitude lower than those of the Reference Case.
Possible combinations of scenarios have also been considered. Many can be excluded
from detailed analysis on qualitative grounds. Where it is appropriate to sum the
releases of two different scenarios, the combined geo-bio flux still does not exceed the
regulatory geo-bio flux constraint.
Monte Carlo simulations, a probabilistic sensitivity analysis (PSA) and a number of
deterministic complementary analyses have been performed to obtain a better
understanding of the modelled system (see details in Assessment of Radionuclide
Release Scenarios for the Repository System). The importance of the properties of any
initial penetrating defect in the canister and its evolution over time has been highlighted
in these analyses.
Quality control and assurance measures have been adopted to ensure transparency and
traceability of the calculations performed and hence to promote confidence in the
analyses of the calculation cases. These are detailed in Assessment of Radionuclide
Release Scenarios for the Repository System.
219
8.9.2
Doses to humans, animals and plants
Figure 8-19 shows the annual dose maxima, and timing of the maxima, to a
representative person within the most exposed group and a representative person among
other people for the biosphere calculation cases presented in this report. The (typical)
absorbed dose rate maximum for plants and animals is 2.6·10-7 mikroGy/h in the
Reference Case (BSA-RC), observed for Pike in freshwater environment. The dose rate
maximum for plants and animals in all the calculation cases presented in this report is
1.3·10-4 mikroGy/h in, observed for Mallard in freshwater environments in the
calculation case VS(A)-SOUTH2.
Regulatory dose constraint
1.E-01
Annual dose [mSv]
1.E-02
1.E-03
VS(A)-SOUTH2
VS(A)-SOUTH1
1.E-04
BSA-BRACKISH
1.E-05
BSA-HIPH
1.E-06
BSA-HIPH-NF
BSA-RC
BSA-ANNFF
1.E-07
BSA-TIME
1.E-08
2020
4020
6020
8020
10020
12020
Year
Figure 8-19. The annual dose maxima to a representative person within the most
exposed group (Emost_exp) for the biosphere calculation cases presented in this report.
1.E-02
Regulatory dose constraint
Annual dose [mSv]
1.E-04
BSA-BRACKISH
VS(A)-SOUTH2
1.E-06
BSA-HIPH
BSA-ANNFF
1.E-08
BSA-HIPH-NF
BSA-RC
BSA-TIME
1.E-10
1.E-12
2020
VS(A)-SOUTH1
4020
6020
8020
10020
12020
Year
Figure 8-20. The annual dose maxima to a representative person among other people
(Eother) for the biosphere calculation cases presented in this report.
220
221
9
COMPLEMENTARY CONSIDERATIONS AND SUPPORTING
EVIDENCE
This chapter outlines complementary considerations that provide additional evidence for
the long-term safety of disposal according to the KBS-3 concept at the Olkiluoto site.
Complementary considerations and additional evidence related to the choice of the
geological disposal concept, the robustness of the KBS-3 method and the suitability of
the Olkiluoto site are summarised in Sections 9.1, 9.2 and 9.3, respectively. The
evidence is fully presented in Complementary Considerations. Results from evaluations
of a range of complementary indicators for the repository system are summarised in
Section 9.4. The evaluations are fully presented in Assessment of Radionuclide Release
Scenarios for the Repository System and Biosphere Assessment.
9.1
Choice of geological disposal
The choice of geological disposal as a concept for disposal of radioactive waste is
backed by technical experience and international consensus.
An appropriately chosen geological formation provides an environment that is stable
over many millions of years – geological timescales – and the nature of changes that can
occur is predictable from the geological sciences. A repository concept is developed that
is consistent with the chosen geological formation, taking advantage of the beningn or
beneficial qualities and designed to withstand expected and unlikely events and
processes that could affect the geological formation in the long term. The depth below
ground provides buffering of the repository system from processes occurring in the
surface environment and protection from unauthorised or inadvertent human actions.
9.2
Support for the robustness of the KBS-3 method
The KBS-3 method uses a few simple, common materials – copper and iron for the
canister, natural swelling clay for the buffer and backfill. This reduces the number of
materials whose properties need to be understood and the number of interactions
between the materials.
Demonstration of safety means that the long-term stability of the engineered barriers
must be robustly assured. This has been considered already when choosing materials
that are long lasting (natural occurrences/deposits of these materials that have persisted
over geological timescales) and with which there is already long experience of their use.
The very long timescales of interest – much longer than historical experience – means
that more than empirical experience is required and understanding of the processes
involved. For example, the understanding of copper corrosion is based on a combination
of experimental evidence and the study of natural analogues. Furthermore, the models
used for the interpretation of natural analogues and to make the safety assessment
calculations are based on the application of fundamental laws of nature, such as mass
and energy balances and the laws of thermodynamics, which provide a robust basis for
their use.
222
Our experience with natural analogues for both materials and processes provides a high
degree of confidence in our understanding of the disposal system and how it will evolve
over hundreds of thousands of years (see Chapter 8 in Complementary Considerations).
For each engineered barrier and key process, there is increasing analogue evidence to
support the models and parameters.
Copper is one of the few elements to occur in elemental form as a natural mineral –
native copper. Although there is no ‘perfect’ analogue, there is strong evidence from a
range of natural occurrences of native copper for very low corrosion rates of copper for
millions of years in groundwater and redox conditions similar to, or less favourable
than, those at Olkiluoto. Archaeological artefacts, while representing much more
variable and often more severe conditions, suggest low localised copper corrosion rates
that are not likely to be significant in determining canister longevity compared with
generalised corrosion.
Taken together, this supports the assertion that the copper canister can provide complete
containment for much longer than tens of thousands of years if the surrounding
environment maintains the favourable chemical conditions and protects the canister
from rock movements.
The bentonite buffer needs to maintain its low permeability and plasticity, and prevent
microbial activity (which could cause sulphate reduction). Studies of naturallyoccurring bentonite deposits show that mineral alteration processes that are detrimental
to the properties of low permeability and plasticity only occur significantly above about
150–200 °C even over geological timescales. Significant changes also depend on a
supply of potassium, which will be limited in the buffer (due to the favourable
groundwater composition and limited amount of potassium associated with foreign
materials to the system). Thus the bentonite buffer will remain stable during the
repository thermal period, in which a maximum buffer temperature of around
90−100 °C is foreseen.
Several analogue (and many experimental) studies have examined the chemical stability
of bentonite under various conditions, but interaction with cement leachate seems to be
the only potential detrimental consideration. The use of ordinary cement, either as a
component of concrete or cement-based grouts can be avoided by design, substituted by
other materials (silica sol grout, for example) or replaced (low pH concrete) so as to
ensure the minimum interaction.
There are several excellent illustrations of bentonite and other clays functioning as a
hydraulic barrier to preserve wood and human cadavers; these analogues also indicate
that microbial activity was significantly reduced.
One of the strengths of KBS-3 is its simplicity of components and materials. Foreign
materials – essentially anything that is not part of the EBS – introduce uncertainty as
they increase the number of chemical components and the number of possible
interactions between them. Thus, tracking (and minimising) all foreign materials offers
additional support to the robustness of the resulting disposal system.
223
9.3
Support for the suitability of geological disposal at the Olkiluoto
site
In Finland, there are limited choices of deep geological settings, leaving fractured
crystalline basement as the only realistic choice of repository host rock. Such rocks,
however, have also been considered as suitable for locating a deep geological repository
for long-lived radioactive waste in many other countries, including several countries in
which alternative rock formations (salt, basalt, tuff and a range of argillaceous
sediments) are available e.g. in Canada (see e.g. AECL 1994, p. 3-4 and Witherspoon &
Bodvarsson 2006), Switzerland (Nagra 1993) and the USA (Rechard et al. 2011).
The important safety functions provided by the host rock and geosphere at a site are:

physical isolation due to depth, which minimises the risks of future perturbations
(e.g. due to glaciation) and human intrusion;

mechanical protection of the engineered barrier system;

favourable geochemical conditions (key characteristics being redox, pH and salinity)
and sufficiently low groundwater flow rates; and

an additional barrier to retard radionuclides, if they are released, due to sorption on
fracture minerals and diffusion into the rock matrix.
The Olkiluoto site is situated within the Fennoscandian Shield, away from active plate
margins. In general, the density and magnitude of earthquakes in Finland is very low;
earthquake magnitudes have never exceeded 5 (M=~5) since records began in the
1880s. Further, according to the data from historical earthquakes, the Olkiluoto area is
located within a zone of lower seismicity, between two seismically more active belts.
There have been only nine recorded earthquakes within 100 km, with the nearest event
(M=3.1) at 35−40 km from Olkiluoto in 1926.
An important consideration is to find a sufficient volume of rock, with generally low
and minor fracturing, to accommodate the spatial extent of the repository. Deposition
tunnels must be placed to avoid shear zones or heavily fractured zones, although the
access tunnel or shafts may cross such zones. At Olkiluoto, several options have been
considered and suitable volumes of rock have been defined such that the deposition
tunnels can be placed on a single level at a depth of 400−450 m.
The rock at Olkiluoto is geotechnically suitable for the construction of self-supporting
tunnels requiring only light rock support. Water inflows at depth are low and zones of
inflow can be treated by local grouting. Significant local experience exists from
construction of the ONKALO at Olkiluoto.
Evidence from drillholes and the ONKALO show that, in the natural situation,
groundwaters at repository depth are reducing, and also have otherwise favourable
hydrochemistry (low to medium levels of salinity, chloride and sum of cations). Such
conditions are indicative of long periods of negligible groundwater flows and suitable
for the longevity and continued functioning of the canister, buffer and backfill. At
repository depth, levels of sulphide, which is expected to be the main agent for canister
corrosion, and sulphate in the groundwaters that might be reduced to sulphide, are both
224
suitably low. There is no evidence that dilute meltwater has reached repository depth
during past glacial cycles.
The performance of both the EBS and the geosphere as retardation barriers for
radionuclides is improved considerably when conditions are chemically reducing. This
is because corrosion processes are generally slower under such conditions and many key
radionuclides are both less soluble and more highly sorbed.
Total salinity can play a significant role in defining the effectiveness of the barrier
system. Most information supporting the design and performance of the EBS and
radionuclide retention in the geosphere is defined for conditions of low to medium
salinity. However, the presence of old, very dense brines provide evidence of geological
stability.
The pH of groundwater is a less sensitive parameter and is unlikely to have much direct
impact on performance of the total system unless it lies at the high (>10) or very low
(<4) end of the range. It is therefore advantageous that the host rock has the capacity to
buffer any pH excursions caused by either perturbations to the rock itself or directly by
the repository.
The overall favourability of the Olkiluoto site has been discussed in detail in the Site
Description, Chapter 11; the stability of the host rock is also discussed Complementary
Considerations, Chapter 7. The characteristics of the site have been determined through
the extensive programme of site investigation, and the key features of the site are well
understood from site investigation and modelling. Although some uncertainties remain,
these are bounded and are allowed for in the safety case. It has been found that the rate
of groundwater flow at the planned repository depth is low and geochemical conditions
are favourable to the engineered barrier system with reducing conditions, low levels of
sulphide and moderate salinity of about 10−20 g/L. Sufficient volumes of rock will be
delineated suitable for the construction of repository panels and deposition tunnels.
Evidence that a site like Olkiluoto is appropriate to host a repository can be seen by the
presence of numerous ore bodies throughout the crystalline Fennoscandian Shield. Their
existence indicates the fundamental barrier properties of fractured crystalline rocks
where the geochemical environment of the host rock is appropriate for radionuclide
retention – as it most certainly is at Olkiluoto. Even at a disturbed site such as Palmottu
(see Chapter 7 in Complementary Considerations), penetration of oxidising glacial
meltwaters into the site was buffered by the host rock after only a hundred metres depth,
indicating the suitability of the crystalline host rock for a repository.
Like Palmottu, the Olkiluoto site also shows indications of the presence of glacial
meltwaters at depth, but not at repository depth. And also as in the case of Palmottu,
meltwaters appear to have mixed with deeper, more mineralised groundwaters or to
have been buffered by rock-water interactions so preventing dilute meltwaters reaching
the repository horizon.
Although data are still sparse (e.g. Pedersen 2008, Pedersen et al. 2010), microbial
populations are generally low at the Olkiluoto site, in line with values for most deep
225
crystalline groundwaters (e.g. Smith et al. 2001, Fukuda et al. 2010). Additional work in
the ONKALO facility will allow this feature to be evaluated more thoroughly in future.
9.4
Safety and complementary indicators
Posiva uses the term safety indicator for the quantities derived in safety assessment to
assess compliance with the regulatory radiation protection constraints. A range of
complementary indicators has been evaluated to highlight the performance of
components of the disposal system, and to provide an alternative line of argument for
safety.
The calculated evolution of the activities in disposal system compartments illustrates
where the majority of the activity resides at any given time (Figure 9-1). This
demonstrates that the majority of activity is contained within the fuel, zirconium alloy
and other metal parts at all calculated times. It illustrates the effectiveness of the waste
form and canister in providing long-term containment, even in the presence of a small
penetrating defect. Long periods of retention and the slowness of transport processes
mean that substantial radioactive decay takes place for the majority of radionuclides
before any eventual release to the surface environment can occur.
Figure 9-1. Evolution of the total activity, summed over all calculated radionuclides, in
the canister, buffer, backfill, geosphere and global biosphere in the Reference Case.
Activity in the ‘global biosphere’ is the time-integrated activity released from the
geosphere, taking into account radioactive decay and ingrowth.
226
Calculated activity concentrations in the buffer and backfill in the Reference Case have
been found to be similar to those that occur naturally in the host rock, although
calculated activity concentrations in the VS1 scenario, where enlargement of the defect
increases the release rate of radionuclides to the buffer and backfill, are up to around
two orders of magnitude higher. Activity concentrations in the buffer and backfill in the
Reference Case are also similar to typical values for coal and fly-ash, which are
examples of naturally-occurring radioactive material (NORM).
Activity fluxes from the near field and geosphere have been compared with the
naturally occurring activity flux due to dissolved radionuclides (most significantly,
Ra-226) in Olkiluoto groundwater24. Releases rates from the near field and geosphere in
the base scenario have been found to be more than an order of magnitude below the
lower bound of the range of uncertainty/variability of Ra-226 flux through the
repository area at repository depth. Release rates from the near field are shown in
Figure 9-2.
Ra-226 and its progeny have the highest dose conversion factor of all radionuclides
considered for the stylised deep wells (Biosphere Assessment, Section 6.4) in the
biosphere assessment. Thus, the fact that the activity flux from the repository, summed
over all radionuclides, is less than the natural Ra-226 flux indicates that the activity flux
from the repository is no more radiologically toxic than the natural flux, based on a
measure of toxicity (the dose conversion factor s for the hypothetical wells) that is
relevant to the site. For the repository variant scenarios, geosphere release rate maxima
have been found either to be lower than, or to lie within, the range of naturally occurring
Ra-226 fluxes. Only for the unlikely, disturbance scenarios are peak geosphere release
rates calculated that are above the naturally occurring Ra-226 flux range.
The C-14 released from the repository can be compared with the natural atmospheric
C-14 that is taken up by plants and animals in an area similar to the repository footprint.
The highest calculated releases of C-14 from the repository to the environment occur
when considering, as a complementary case, the possibility of gas-mediated release
from the repository. In the case of gas-mediated transport of C-14 from a defective
canister through the repository near field and geosphere to the surface environment, the
peak annual release rate, averaged over 1000 years, is around a factor of four below the
regulatory geo-bio flux constraint (Section 12.3.2 of Assessment of Radionuclide
Release Scenarios for the Repository System). To put this peak annual release rate in
perspective, it is almost an order of magnitude less than the annual uptake of natural
atmospheric C-14 by a forest ecosystem over the repository footprint, as demonstrated
in Section 13.4.2 of Assessment of Radionuclide Release Scenarios for the Repository
System.
24
Ra-226 concentrations are only based on four samples taken at Olkiluoto from different groundwater types/depths. These
concentrations are relatively high compared with other investigation sites in Finland (e.g. Romuvaara, Kivetty), where at the same
depth range Ra-226 concentration is no higher than 0.025 Bq/L. Nonetheless, the radium concentration is generally higher in the
more saline waters which are typical in coastal areas and in Ca-rich groundwaters such as Olkiluoto (Pitkänen et al. 2003, see also
e.g. Pöllänen 2003).
227
Figure 9-2. Near-field release rates for the various base scenario calculation cases,
compared with range of natural Ra-226 flux through the repository area at around
repository depth.
Finally, it has been found that the performance of each of the repository barriers can
often be characterised by two sets of element-specific parameters: transfer coefficients
and delay times. In the Reference Case, the lowest transfer coefficients (i.e. the highest
resistances to transfer) are from the canister to the buffer, indicating the importance of
the assumed size of defect (hole) in limiting releases of the key radionuclides. The
assumption of a 1 mm defect, as in the Reference Case and other cases is cautious, and
improvements in non-destructive testing methods are likely to reduce the maximum size
of defect that could possibly remain undetected.
228
229
10
COMPLIANCE WITH LEGAL REQUIREMENTS AND REGULATIONS
AND ASSOCIATED UNCERTAINTIES
10.1
Compliance with legal and regulatory requirements
The TURVA-2012 safety case demonstrates that Posiva’s repository design and
analyses of performance and safety are fully consistent with all the legal and regulatory
requirements related to long-term safety as set out in Government Decree 736/2008 and
STUK Guide YVL D.525.
A detailed trail showing that each of the legal and regulatory requirements is fulfilled is
contained within the body of the TURVA-2012 portfolio and summarised in
Appendix 2 of this report. Key features of the demonstration are summarised below.
The Posiva repository design is based on a robust system of multiple barriers. For the
expected evolution of conditions in the Olkiluoto bedrock and engineered barriers, the
copper canisters, in which the spent nuclear fuel is contained, are expected to contain all
radionuclides for over one million years. The location of the repository, at a depth of
about 400 to 450 m below ground, will provide isolation from the surface environment
and protection against inadvertent intrusion.
The mutually complementary barriers provide well-defined safety functions and the
barriers are arranged so that the detrimental impact of a deficiency in any individual
barrier on its safety functions will be compensated for by other safety functions.
Similarly, the system of complementary barriers and safety functions provides
robustness with respect to external events and processes, including geological and
climatic changes. The requirements for the reliable operation of each safety function are
expressed in terms of performance targets for the engineered barriers and target
properties for the host rock. These lead to design requirements for the engineered
barriers and definition of a Rock Suitability Classification system (RSC) by which the
local suitability of the rock for development of underground openings and deposition of
spent nuclear fuel can be assessed.
A comprehensive examination has been made of the features, events and processes that
could affect the evolution of the disposal system (repository system plus surface
environment), and the performance of individual barriers or the fulfilment of their safety
functions. Understanding of the changes due to construction and operation of the
repository, and understanding of the longer-term natural processes (mainly related to
climate changes) that will control the evolution of the natural setting of the repository,
leads to the definition of future lines of evolution of the repository and its setting.
The performance of the repository system has been systematically analysed in different
time windows. The analyses take account of the uncertainties in the initial state and
expected thermal, hydraulic, mechanical and chemical evolution of the repository
system, and uncertainties in the expected future lines of evolution, and also the
occurrence of unexpected or disruptive events. The analyses show that, under most
conditions and lines of evolution of the host rock and engineered barriers, all
25
As agreed with STUK, the licence application is based on draft 4 (the version from 17.3.2011 has been used).
230
performance requirements will be met. In this case the copper canisters will remain
intact and no releases of radionuclides will occur over at least one million years. Up to
50,000 years, release of radionculides can only occur if a canister with an initial
penetrating defect is emplaced. In the longer term, glacial episodes at the site may cause
hydrogeological and hydrochemical changes and seisimic disturbances leading to rock
shear, such that a few canister failures might occur in less favourable locations within
the repository.
Although releases of radionuclides to the environment are not expected, the safety
analyses focus on the cases in which releases of radionuclides could occur. It is shown
that even accounting for unlikely combinations of emplacement of a canister with an
initial penetrating defect in less favourable local rock conditions, peak normalised
radionuclide releases to the surface environment are orders of magnitude below the
radionuclide-specific regulatory constraints specified in the STUK Guide YVL D.5. In
the long term (approximately 100,000 years or more), calculated radionuclide release
rates remain below the regulatory constraint for the radioactive release to the
environment, even for pessimistic and unlikely combinations of damage to canisters by
rock shear events and erosion of buffer material due to dilute groundwater conditions.
The results of biosphere assessment show that the annual doses to representative
persons both within the most exposed group and among other people are orders of
magnitude below the regulatory constraints specified in the STUK Guide YVL D.5.
Absorbed dose rates for a range of plants and animals have been calculated and the
results show that exposures remain clearly below the levels which, on the basis of the
best available scientific knowledge, would cause decline in biodiversity or other
significant detriment to any living population.
Overall, it is concluded that the TURVA-2012 safety case demonstrates compliance
with the legal and regulatory requirements for the planned and designed disposal facility
for spent nuclear fuel at Olkiluoto. Some uncertainties still remain in the data and
models and some of these are unlikely to be eliminated. However, the analyses
performed have shown that the repository system is robust against these uncertainties,
and that the conclusions drawn about the compliance with the safety requirements hold
even when these uncertainties are taken into account.
10.2
The main research and development needs during the coming
years
The TURVA-2012 safety case assesses the performance and long-term safety of a
KBS-3 type spent nuclear fuel disposal facility at Olkiluoto. The safety case also
addresses the known uncertainties that may have an impact on the performance of the
facility. The TURVA-2012 safety case forms the basis for the construction licence
application, in which Posiva proposes that the construction of the repository can be
started. Some uncertainties still remain, but these do not affect the conclusions on longterm safety. Additional research and development will, however, help increase the
reliability of the safety case to be compiled for the operational licence application. The
focus of the research and development in the coming years are on the:
231

better understanding of the processes affecting canister corrosion and erosion of
buffer and backfill;

rock conditions in potential volumes of rock for the repository and the application of
RSC criteria for the selection of repository panels, tunnels and deposition holes;

demonstration of the implementation of the components of the repository system at
full scale according to the technical design and quality performance requirements.
Further investigations of the properties of the rock in the repository area will reduce the
probability of locating the canisters in unfavourable positions with respect to future
loads. The processes affecting the performance of the engineered barriers will continue
to be experimentally studied. Technical tests will be applied to demonstrate that the
repository can be implemented according to the assumptions made in the safety case.
232
233
11
STATEMENT OF CONFIDENCE
The proposed repository for the disposal of spent nuclear fuel is sited and designed to
provide long-term containment and isolation of the radionuclides in the fuel from the
surface environment.
A safety case has been developed to show, at a level of detail appropriate to the
repository construction licence application, that the safety concept can be implemented
through the KBS-3 method at the Olkiluoto site. The safety case is documented in a
report portfolio, the purpose of which is to show understanding of the disposal system,
including the initial state and future evolution, the main features, events and processes
that drive the evolution of the system, and the system’s performance, including the main
uncertainties.
It has been shown that the system will perform according to targets set for the EBS and
the host rock. Nevertheless, scenarios that could potentially lead to radionuclide release
have been identified and assessed. The radiological impact of these release scenarios
has been evaluated and shown to comply with the criteria given by the nuclear safety
authorities. In addition to compliance with regulatory criteria, factors contributing to
confidence in the long-term safety of the disposal system are considered such as quality
assurance aspects and compliance with international guidance on safety case
development. The knowledge and experience gathered thus far are sufficient to submit
the construction licence application for the disposal facility. Remaining uncertainties
will be addressed through further research and technological development (RTD)
activities to either resolve them through a modified design or gather further data to
better understand their effect on long-term safety. Based on the quantitative and
qualitative results summarised in the previous chapters, Posiva is confident that the
proposed disposal facility for spent nuclear fuel at Olkiluoto complies with the longterm safety criteria set by the authorities.
Posiva’s safety concept for the disposal of spent nuclear fuel is based on long-term
containment and isolation of the hazardous materials in the spent nuclear fuel. Longterm containment is achieved by packaging the spent fuel in canisters with long
expected lifetimes. Isolation is provided by depth of disposal and a multi-barrier system
consisting of both engineered and natural barriers.
The purpose of the safety case is to show, at a level of detail appropriate to the
repository construction licence application, that Posiva’s safety concept can be
implemented through the KBS-3 method, taking into account the characteristics of the
spent fuel, the site, and in the framework of the regulatory context in Finland.
Posiva’s methodology for the safety case (Chapter 2) consists of:

Description of the KBS-3 method and the Olkiluoto site

Documentation of the design basis

Assessment of performance and of long-term safety

Uncertainty management
234

Quality assurance.
This methodology is implemented through a portfolio of reports, the main results of
which are synthesised in the present report.
The implementation of the safety concept is described in Design Basis along with the
safety functions of the natural and engineered barriers. Performance targets (for
engineered barriers) and target properties (for the host rock) are also defined and
repository design requirements are set in a way such that the long-term safety
requirements will be met in the expected envelope of future conditions. The disposal
system is described in detail in Description of the Disposal system along with the initial
state of the system, i.e. the state at the time when direct control of the waste package
ceases.
Drivers for the evolution of the disposal system and related uncertainties are identified
in Features, Events and Processes. The models and data used in the performance
assessment and in the analysis of radionuclide release scenarios along with their quality
and associated uncertainties are compiled and described in Models and Data for the
Repository System and in Biosphere Data Basis. Identified uncertainties in models
(including numerical codes) and data that have an impact on long-term safety are
followed up in the RTD programme for 2013−2015.
Performance Assessment makes a comprehensive evaluation of the evolution of the
repository system in various time windows. In particular, it demonstrates that, under the
expected lines of evolution, the performance targets and target properties will be met in
each time window. Circumstances in which performance targets and target properties
might not be met at some locations and times (incidental deviations) are identified and
the potential for such deviations to lead to radionuclide releases is considered.
Uncertainties and deviations that could potentially lead to the release of radionuclides or
affect the radiological consequences of such releases are considered in Formulation of
Radionuclide Release Scenarios. Among the scenarios formulated, the “base scenario”
addresses the most likely lines of evolution. It includes both the possibility that no
canisters fail within the assessment time frame (Performance Assessment), and also the
incidental emplacement of one or a few canisters having an initial undetected
penetrating defect. In addition, several “variant” scenarios are formulated to deal with
uncertainties in the performance of the system, and “disturbance” scenarios are
postulated related to unlikely events that could lead to loss of one or more safety
functions.
The scenarios leading to radionuclide releases are analysed in Assessment of
Radionuclide Release Scenarios for the Repository System and their radiological
impacts in Biosphere Assessment. The analysis of radionuclide release scenarios for the
repository shows that all scenarios, including the most unfavourable ones, comply with
the regulatory criteria. In spite of the cautious assumptions used in formulating the
calculation cases and in the selection of the data, the margin of safety with respect to the
regulatory constraint on radionuclide release rate to the biosphere is several orders of
magnitude for the base and variant scenarios. Even for the “what if” cases in the
235
unlikely disturbance scenarios the margin of safety is still around one order of
magnitude.
The results of biosphere assessment show that the annual doses to humans comply with
the regulatory constraints and that radiation exposure of plants and animals remain
clearly below the levels which, on the basis of the best available scientific knowledge,
would cause decline in biodiversity or other significant detriment to any living
population.
Complementary Considerations provides additional arguments and evidence for the
long-term safety of the disposal system. This includes evidence for the stability of the
host rock conditions, the robustness of the KBS-3 method, the suitability of the
repository design and materials, and the limited rates of radionuclide migration in the
repository system. A range of safety and complementary indicators has been evaluated.
Radionuclide concentrations in the buffer and backfill have been shown to be
comparable to those in naturally occurring radioactive material, and radionuclide release
rates to be comparable to naturally occurring activity fluxes in groundwater at the site.
Finally, the present Synthesis report provides an overview of the main results of these
other reports, draws conclusions on the compliance of the disposal system and of the
safety case with the regulatory criteria and guidelines, identifies the remaining
uncertainties and thus provides the basis for the present statement of confidence. Based
on the quantitative and qualitative analyses presented in the safety case portfolio,
compliance of the repository performance with the regulatory requirements and
guidelines is shown.
Inevitably, uncertainties remain. The impact of these uncertainties has been assessed
through a number of scenarios and quantitative analyses that cover a broad range of
conditions and uncertainties. Identified remaining uncertainties are such that they do not
have an immediate impact on safety for the construction phase of the repository and
thus on the construction licence application. These uncertainties will be addressed
through further research and technological development (RTD) activities to either
resolve them through a modified design or gather further data to better understand their
long-term safety impact. So far, no uncertainties have been identified that cannot be
adequately resolved before the operational licence application. As yet, unidentified
issues cannot be excluded and their early detection is a key aim of a programme of
demonstration and pilot activities as well as a monitoring programme.
Quality control and assurance measures have been adopted to ensure transparency and
traceability of the calculations performed. All research, development and technical
design work at Posiva is subject to a certified ISO 9001 management system, which has
been augmented by a graded approach based on the safety significance of various
actions and processes. As regards the safety case activities, special emphasis is given to
the documentation of the process, critical assessment of the models and data used, and
the expert review process that each safety case report undergoes before publication.
Furthermore, an uncertainty management strategy has been implemented structured
around a stepwise approach to repository system development, construction and
operation as well as an iterative approach between long-term safety and RTD activities.
236
Based on the quantitative and qualitative results summarised in the previous chapters
and the strategy for management of remaining uncertainties, Posiva considers that the
TURVA-2012 safety case demonstrates that the proposed repository design provides a
safe solution for the disposal of spent nuclear fuel and that the performance and safety
assessments are fully consistent with all the legal and regulatory requirements related to
long-term safety as set out in Government Decree 736/2008 and in the guidance from
the regulatory − STUK. Moreover, Posiva considers that the level of confidence in the
demonstration of safety is appropriate and sufficient to submit the construction licence
application to the authorities. The assessment of long-term safety includes uncertainties,
but these do not affect the basic conclusions on the long-term safety of the repository.
237
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Formulation of Radionuclide Release Scenarios
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Terrain and Ecosystem Development Modelling
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TURVA-2012 Portfolio SUPPORTING reports
Backfill Production Line report
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tunnel backfill and plug. Eurajoki, Finland: Posiva Oy. POSIVA 2012-18. ISBN 978951-652-199-5.
Biosphere Description
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Canister Production Line report
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Eurajoki, Finland: Posiva Oy. POSIVA 2012-16. ISBN 978-951-652-197-1.
Closure Production Line report
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Finland: Posiva Oy. POSIVA 2012-19. ISBN 978-951-652-200-8.
Site Description
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239
Underground Openings Production Line report
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the underground openings. Eurajoki, Finland: Posiva Oy. POSIVA 2012-22. ISBN 978951-652-203-9.
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www.nrc.gov/about-nrc/regulatory/research/radiological-toolbox.html
250
251
APPENDIX 1: GOVERNMENT DECREE (736/2008)
Government Decree (736/2008)
on the safety of disposal of nuclear waste
Issued in Helsinki 27 November 2008
—————
According to the Government decision made on the submission by the Ministry of Employment and the
Economy, the following provisions are issued under Section 7 q of the Nuclear Energy Act (990/1987),
issued on 11 December 1987, in the form laid down in Act 342/2008:
Chapter 1
Scope of application and definitions
Section 1
Scope of application
This Decree shall apply to the disposal of spent
nuclear fuel and other nuclear waste, originating
in a nuclear facility, into a facility to be
constructed in bedrock.
The decree shall also apply to radioactive waste
as referred to in section 10 of the Radiation Act
(592/1991), if such waste is sited in a disposal
facility for nuclear waste, as referred to in
subsection 1.
Provisions on the handling and storage of spent
nuclear fuel and other nuclear waste in a nuclear
facility attached to a nuclear power plant are laid
down in the Government Decree on the Safety of
Nuclear Power Plants (733/2008).
Section 2
Definitions
For the purposes of this Decree:
1) nuclear waste facility shall refer to a nuclear
facility used for the encapsulation of spent nuclear
fuel or conditioning of other nuclear waste for
disposal, and to a disposal facility for spent
nuclear fuel or other nuclear waste;
2) disposal facility shall refer to an entirety
comprising the rooms for disposal of the waste
packages (emplacement rooms) and the adjoining
underground and above-ground auxiliary
facilities.
3) disposal site shall refer to the location of the
disposal facility and, after disposal has been
completed, the area entered in the real estate
register in accordance with Section 85 of the
Nuclear Energy Decree (161/1988), and the
underlying ground and bedrock.
4) short-lived waste shall refer to nuclear waste, the
activity concentration of which after 500 years is
below the level of 100 megabecquerels (MBq) per
kilogram in each disposed waste package, and
below an average value of 10 MBq per kilogram of
waste in one emplacement room;
5) long-lived waste shall refer to nuclear waste,
the activity concentration of which after 500 years
is above the level of 100 megabecquerels (MBq)
per kilogram in a disposed waste package, or
above an average value of 10 MBq per kilogram
of waste in one emplacement room;
6) annual dose shall refer to the sum of the
effective dose arising from external radiation
within the period of one year, and of the
committed effective dose from the intake of
radioactive substances within the same period of
time;
7) long-term safety shall refer to the safety of
disposal after the operational period of the
disposal facility, taking account of radiological
impacts on man and the environment;
8) safety case shall refer to documentation for
demonstrating compliance with the long-term
safety requirements;
9) safety functions shall refer to factors preventing
and limiting the releases and migration of
disposed radioactive materials;
10) barrier shall refer to an engineered or natural
structure or material used for achieving safety
functions;
11) assumed operational occurrence shall refer to
such incident influencing the safety of a nuclear
waste facility that can be expected to occur at
least once during any period of a hundred
operating years;
12) postulated accident shall refer to such
incident influencing the safety of a nuclear waste
facility that can be assumed to occur more rarely
than once during any period of a hundred
operating years; postulated accidents are grouped
further into two classes on the basis of their
frequency:
a) class 1 postulated accidents, which can be
assumed to occur at least once during any period
of a thousand operating years;
b) class 2 postulated accidents, which can be
assumed to occur less frequently than once during
any period of a thousand operating years;
252
13) expected evolution scenario shall refer to such
change affecting the performance of barriers that
has a high probability of causing radiation
exposure during the assessment period and which
can be caused by interactions occurring in the
disposal facility, by geological or climatic
phenomena or by human action; and
14) unlikely events impairing long-term safety
shall refer to such potential events significantly
affecting the performance of barriers that have a
low probability of causing radiation exposure
during the assessment period and which can be
caused by geological phenomena or by human
action.
Chapter 2
Radiation safety
Section 3
Operation of nuclear waste facility
A nuclear waste facility and its operation shall be
designed so that:
1) the radiation exposure of workers at the facility
is limited by all practicable means and so that the
maximum values laid down in the Radiation
Decree (1512/1991) are not exceeded;
2) as a consequence of undisturbed operation of
the facility, releases of radioactive materials into
the environment remain insignificantly low;
3) as a consequence of assumed operational
occurrences, the annual dose to the most exposed
people other than workers of the facility remains
below the value of 0.1 millisievert (mSv); and
4) as consequence of a postulated accident, the
annual dose to the most exposed people other than
workers of the facility remains below:
a) the value of 1 mSv when a Class 1 postulated
accident occurs;
b) the value of 5 mSv when a Class 2 postulated
accident occurs.
When applying this section, radiation doses
arising from natural radioactive materials in the
host rock of or released from groundwater bodies
into the underground rooms of the disposal
facility, shall not be taken into account.
Section 4
Long-term radiation impacts of disposal
Disposal of nuclear waste shall be planned so that
radiation impacts arising as a consequence of
expected evolution scenarios will not exceed the
constraints given in subsections 2 and 3.
In any assessment period, during which the
radiation exposure of humans can be assessed
with sufficient reliability, and which shall extend
at a minimum over several millennia:
1) the annual dose to the most exposed people
shall remain below the value of 0.1 mSv; and
2) the average annual doses to other people shall
remain insignificantly low.
During assessment periods after the period
referred to above in subsection 2, average
quantities of radioactive materials over long time
periods, released into the living environment from
the disposed nuclear waste, shall remain below
the maximum values specified separately for each
radionuclide by the Radiation and Nuclear Safety
Authority (STUK). These constraints shall be
specified so that:
1) at a maximum, radiation impacts caused by
disposal can be equivalent to those caused by
natural radioactive materials in earth’s crust; and
2) on a large scale, the radiation impacts remain
insignificantly low.
Section 5
Consideration of unlikely events
The significance of unlikely events impairing
long-term safety shall be assessed by evaluating
the reality, probability and possible consequences
of each event. Whenever possible, the
acceptability of the expectancies of radiation
impacts caused by such events shall be evaluated
in relation to the annual dose and release rate
constraints of radioactive materials, as referred to
in section 4.
Chapter 3
Design requirements for a nuclear waste
facility
Section 6
Handling of spent nuclear fuel and other nuclear
waste
Spent nuclear fuel and other nuclear waste shall
be conditioned and packed in accordance with
disposal specifications. Waste packages shall be
classified on the basis of their characteristics.
Constraints and other quality specifications shall
be defined for each class, necessary in terms of
the operational safety of the nuclear waste facility
and the long-term safety of disposal, and which
the waste packages are required to meet.
253
The nuclear waste facility shall employ effective
radiation protection arrangements in order to limit
the occupational radiation exposure and radiation
impacts caused in the environment of the facility.
In waste handling, releases of radioactive
materials inside the facility and into the
environment shall be prevented and limited as
necessary with containment, recovery and
filtering systems. Sufficient radiation protection
shall be ensured in handling of spent nuclear fuel
or other highly irradiating nuclear waste by using
remote handling and radiation shielding.
In handling of spent nuclear fuel, any damage to
the fuel and occurrence of a self-sustaining chain
reaction of fissions shall be prevented, and
sufficient cooling of the fuel shall be ensured, to a
high degree of certainty.
Section 7
Safety classification
The systems, structures and components of a
nuclear waste facility shall be classified on the
basis of their significance in terms of the
operational safety of the facility, or the long-term
safety of disposal. The required quality level of
each classified object, and the inspections and
testing necessary for verifying the quality, shall
be adequate as regards the significance of the
object in terms of safety.
Section 8
Prevention of operational occurrences and
accidents
In order to prevent operational occurrences and
accidents, the design, construction and operation
of a nuclear waste facility shall employ proven or
otherwise carefully examined high quality
technology. A nuclear waste facility shall
encompass systems that facilitate quick and
reliable detection of an operational occurrence or
accident and prevent the aggravation of any event.
Effective technical and administrative measures
shall be provided for the mitigation of the
consequences of potential accidents.
The functions at a nuclear waste facility, the
failure of which could result in a significant
release of radioactive materials or radiation
exposure of personnel at the facility, shall be
ensured. Ensuring the functions important to
safety shall primarily be based on inherent safety
features, alongside systems and components that
do not require external power supply or which, as
a consequence of a loss of power supply, will
settle into a state deemed preferable from the
safety point of view.
The design of a nuclear waste facility shall take
account of any impacts caused by potential
natural phenomena and other events external to
the facility. As external events, even unlawful
activities aiming at damaging the facility shall be
taken into account.
In a nuclear waste facility, the placement and
protection of systems alongside operative
methods shall ensure that fire, explosions or other
events inside the facility do not pose a threat to
safety.
Section 9
Disposal operations
The transfer of waste packages into the
emplacement rooms shall be carried out so that
the possibility of accidents remain low and the
packages cannot be damaged in any way that
would affect long-term safety.
The disposal package containing spent nuclear
fuel shall be designed so that no self-sustaining
chain reaction of fissions can occur, even in the
disposal conditions.
The emplacement activities shall be separated
from the excavation and construction work of the
disposal facility in such a manner as to ensure that
excavation and construction work cannot have
any harmful impact on the operational safety of
the facility or the long-term safety of disposed
waste.
The long-term performance of barriers shall be
confirmed by establishing an investigation and
monitoring programme, to be implemented during
the operational period of the final disposal
facility. A record shall be maintained of disposed
waste, including waste package specific data on
the waste type, radioactive materials, location
within the waste emplacement room, and other
necessary data. The Radiation and Nuclear Safety
Authority (STUK) shall arrange the permanent
recording of information concerning the disposal
facility and disposed waste.
An adequate protection zone shall be reserved
around the disposal facility as a provision for the
prohibitions on measures referred to in paragraph
6, section 63(1) of the Nuclear Energy Act.
Chapter 4
Long-term safety of disposal
Section 10
General requirements concerning disposal
254
Disposal shall be implemented in stages, with
particular attention paid to aspects affecting longterm safety. The planning of the construction,
operation and closure of a disposal facility shall
take account of reduction of the activity of
nuclear waste through interim storage, the
utilisation of high-quality technology and
scientific data and the need to ensure long-term
safety via investigations and monitoring.
However, the implementation of the various
stages of disposal shall not be unnecessarily
postponed.
Section 11
Multibarrier principle
The long-term safety of disposal shall be based on
safety functions achieved through mutually
complementary barriers so that a deficiency of an
individual safety function or a predictable
geological change will not jeopardise the
longterm safety.
Safety functions shall effectively prevent releases
of disposed radioactive materials into the bedrock
for a certain period, the length of which depends
on the duration of the radioactivity in waste. For
short-lived waste, this period shall be at least
several hundreds of years, and for long-lived
waste, at least several thousands of years.
Section 12
Disposal site
The geological characteristics of the disposal site
shall, as a whole, be favourable to the isolation of
the radioactive substances from the environment.
Any area with a feature that is substantially
adverse to long-term safety shall not be selected
as the disposal site.
The planned final disposal site shall contain
sufficiently large, intact rock volumes that
facilitate the construction of the waste
emplacement rooms. For the purposes of disposal
facility design and acquiring data required for
safety assessments, the geological characteristics
of the host rock at the site shall be characterized
through investigations at the intended disposal
depth, in addition to surface based investigations.
The layout, excavation, construction and closure
of underground facilities shall be implemented so
that the characteristics of the host rock deemed
important in terms of long-term safety are
retained, as far as possible.
The depth of the waste emplacement rooms shall
be selected appropriately as regards the waste
type and local geological conditions. The goal
related to disposal depth shall be that any impacts
on the long-term safety of above-ground events,
activities and environmental changes will remain
minor and that intrusion into the waste
emplacement rooms will be difficult.
Chapter 5
Demonstration of compliance with safety
requirements
Section 13
Operational safety of nuclear waste facility
Compliance with safety requirements concerning
the operation of a nuclear waste facility shall be
proven in connection with commissioning as far
as possible. Insofar as this is not possible,
operational safety shall be demonstrated through
experimental or computational methods, or via a
combination thereof. Computational methods
shall be selected so that the actual risk or harm
remains below the results of calculations, with a
high degree of certainty. Computational methods
shall be reliable and validated for dealing with the
events under analysis. The selection of
operational occurrences and accidents to be
analysed shall take account of their estimated
probabilities.
Section 14
Long-term safety
Compliance with the requirements concerning
long-term radiation safety, and the suitability of
the disposal method and disposal site, shall be
proven through a safety case that must analyse
both expected evolution scenarios and unlikely
events impairing long-term safety. The safety case
comprises a numerical analysis based on
experimental studies and complementary
considerations insofar as quantitative analyses are
not feasible or involve considerable uncertainties.
Compliance with the radiation exposure
constraints for the most exposed people, as
referred to in section 4 above, shall be proven by
considering a community that derives nutrition
from the immediate surroundings of the disposal
site and is most exposed to radiation. In addition
to impacts on people, possible impacts on flora
and fauna shall be analysed.
Section 15
Reliability of the safety case
The input data and models utilised in the safety
case shall be based on high-quality research data
255
and expert judgement. Data and models shall be
validated as far as possible, and correspond to the
conditions likely to prevail at the disposal site
during the assessment period.
The basis for selecting the computational methods
used shall be that the actual radiation exposure
and quantities of radioactive materials released
remain below the results of safety analyses, with a
high degree of certainty. The uncertainties
involved in the safety analysis, and their
significance, shall be separately assessed.
Section 16
Presentation of, and updates to, the safety case
The safety case shall be presented in connection
with the construction licence application and the
operating licence application of the nuclear waste
facility. The safety case shall be updated at 15
year intervals unless otherwise provided in the
licence conditions. Furthermore, the safety case
shall be updated prior to the permanent closure of
the facility.
Chapter 6
Construction and operation of the nuclear
waste facility
Section 17
Construction and commissioning
The holder of a construction licence for the
nuclear waste facility shall ensure that the facility
will be constructed in compliance with the
approved plans and procedures. Moreover, the
licensee shall ensure that the plant supplier and
subcontractors producing services and products
important in terms of safety act in an appropriate
manner.
In connection with the commissioning of a
nuclear waste facility, the licensee shall ensure
that the systems, structures and components and
the facility as a whole operate in the planned
manner. The licensee shall also ensure that an
expedient organisation is in place for the future
operation of the facility, alongside a sufficient
number of qualified personnel and instructions
suitable for the purpose.
Section 18
Operation
The operation of a nuclear waste facility shall be
based on written instructions that correspond to
the current structure and state of the facility.
Instructions shall be made available for the
identification and control of operational
occurrences and accidents. Significant events
influencing safety shall be documented so as to
facilitate their later analysis.
The Technical Specifications of a nuclear waste
facility shall include the technical and
administrative requirements for ensuring the
operation of the facility in compliance with design
bases. The licensee shall operate the facility in
compliance with these requirements and
restrictions, and supervise compliance and report
any deviations from them.
The nuclear waste facility shall have a condition
monitoring and maintenance programme for
ensuring the integrity and reliable operation of
systems, structures and components. Written
orders and appended instructions shall be issued
for the service and repair of components.
Compliance with requirements concerning the
operational radiation safety of the nuclear waste
facility shall be ensured through continuous or
periodic measurements inside the facility, in
possible significant release routes and in the
environs of the facility.
Chapter 7
Organisation and personnel
Section 19
Safety culture
When designing, constructing, operating and
decommissioning or closing a nuclear waste
facility, a good safety culture shall be maintained.
In its decisions and operations, the management
of the organisation concerned shall demonstrate
its commitment to procedures and solutions
promoting safety. Personnel shall be motivated to
perform responsible work and an open working
atmosphere shall be promoted in the working
community, in order to encourage the
identification, reporting and elimination of factors
endangering safety. Personnel shall be given the
opportunity to contribute to the continuous safety
enhancement.
Section 20
Safety and quality management
Organisations participating in the design,
construction, operation and decommissioning or
closure of a nuclear waste facility shall employ a
management system for ensuring the management
of safety and quality. The objective of the
256
management system is to ensure that safety is
prioritised without exception, and that quality
management requirements are commensurate with
the significance to safety of the activity. This
management system shall be systematically
assessed and further developed.
Safety and quality management shall cover all
activities influencing the safety of the nuclear
waste facility. For each activity, requirements
significant in safety terms shall be identified, and
planned measures described in order to ensure
compliance with requirements. The processes and
procedures shall be systematic and based on
instructions.
Systematic procedures shall be in place for
identifying and correcting deviations significant
in safety terms.
The licensee shall commit and oblige its
employees and suppliers, subcontractors and other
partners contributing to safety relevant activities
to engage in systematic safety and quality
management.
development and maintenance of the professional
skills of the persons working in these positions,
and adequate command of the knowledge
required for the positions shall be verified.
Chapter 8
Miscellaneous provisions
Section 22
Disposal in the ground
If nuclear waste, as referred to in the Nuclear
Energy Act, will be disposed of in a facility
constructed in the ground, said disposal shall be
planned and implemented in compliance with the
requirements laid down in sections 3—9 and 13—
21 herein. Only very low-level waste, the average
activity concentration of which does not exceed
the value of 100 kBq per kilogram, and the total
activity of which does not exceed the limits laid
down in section 6(1) of the Nuclear Energy
Decree, can be placed in a facility constructed in
the ground.
Section 21
Lines of management, responsibilities and
expertise
Section 23
Entry into force
The lines of management in the organisation of a
nuclear waste facility, alongside the positions and
related responsibilities of employees, shall be
defined and documented. The organisation shall
have access to the professional expertise and
technical knowledge required for the safe
operation of the nuclear waste facility and longterm safety of nuclear waste disposal.
Duties significant to safety shall be designated.
Training programmes shall be prepared for the
This Decree enters into force on 1 December
2008.
This Decree repeals the Decision of the Council
of State on the general regulations for the safety
of a disposal facility for reactor waste (398/1991),
issued on 14 February 1991, and the Government
Decision on the safety of disposal of spent nuclear
fuel (478/1999), issued on 25 March 1999.
Measures required for the enforcement of the
Decree can be undertaken prior to the entry into
force of the Decree.
Issued in Helsinki, 27 November 2008
Mauri Pekkarinen, Minister of Economic Affairs
Pasi Mustonen, Senior Adviser
The results of the safety analysis show
that, for the expected evolution scenarios,
the annual doses to humans during the
next 10,000 years remain below the
radiation dose constraints given in the
Government Decree. The dose
assessment (for humans) is presented in
the:
 Biosphere Radionuclide Transport
and Dose Assessment
 Biosphere Assessment
3
RADIATION PROTECTION
3.2 Long-term safety
Radiation dose constraints
306. Disposal of nuclear waste shall be planned so that as a
consequence of expected evolution scenarios
1) the annual dose to the most exposed people shall remain
below the value of 0.1 mSv
2) the average annual doses to other people shall remain
insignificantly low.
These constraints are applicable in an assessment period, during
which the radiation exposure of humans can be assessed with
sufficient reliability, and which shall extend at a minimum over
several millennia (GD 736/2008).
Chapter 2: Radiation safety
Section 4 – Long-term radiation impacts of
disposal
In any assessment period, during which the
radiation exposure of humans can be assessed with
sufficient reliability, and which shall extend at a
minimum over several millennia:
1) the annual dose to the most exposed people
shall remain below the value of 0.1 mSv; and
2) the average annual doses to other people shall
remain insignificantly low.
Climatic conditions are expected to remain
similar to present-day conditions within the
next 10,000 years. The effects of isostatic
uplift − a relative fall in sea level and hence
transition from coastal towards terrestrial
and freshwater ecosystems − have been
taken into account in dose assessments.
Human habits, nutritional needs and
metabolism have been assumed to remain
as at present, i.e. unchanged, during the
dose criteria time window. See:
 Terrain and Ecosystems
Development Modelling
 Biosphere Assessment.
The calculation of transport path(s), exit
location(s) and releases to the surface
environment is presented in Assessment of
Radionuclide Release Scenarios for the
307. In applying the dose constraints, such environmental changes
need to be considered that arise from changes in ground level in
relation to sea. The climate type as well as the human habits,
nutritional needs and metabolism can be assumed to remain
unchanged.
308. In applying the constraints, the exposure shall be assumed to
arise from radioactive substances released from the repository,
transported to near-surface groundwater bodies and further to
above-ground watercourses. At least the following potential
and summarised in the
 Synthesis.
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Legal requirements set out in the Government
Decree (736/2008)
APPENDIX 2: AUDIT OF LEGAL AND REGULATORY REQUIREMENTS RELATED TO THE LONG-TERM SAFETY CASE
257
During assessment periods after the period referred
to above … average quantities of radioactive
materials over long time periods, released into the
living environment from the disposed nuclear waste,
shall remain below the maximum values specified
separately for each radionuclide by the Radiation
and Nuclear Safety Authority (STUK). These
constraints shall be specified so that:
1) at a maximum, radiation impacts caused by
disposal can be equivalent to those caused by
natural radioactive materials in earth’s crust;
and
2) on a large scale, the radiation impacts remain
insignificantly low.
Legal requirements set out in the Government
Decree (736/2008)
Activity releases to the environment (“geo-
The doses to larger groups of people are
based on the dose to each exposed
individual and the number of exposed
individuals. The doses to larger groups of
people meet the requirement in YVL D.5,
paragraph 310, see the Biosphere
Assessment.
310. In addition, the average annual doses to such larger groups of
people shall be addressed, who live at a regional lake or a coastal
site and are exposed to the radioactive substances transported into
these watercourses. The acceptability of these doses depends on
the number of exposed people; however, these doses shall not be
more than one hundredth – one tenth of the constraint for the most
exposed individuals given above.
312. The nuclide specific constraints for the radioactive releases to
the environment (average release of radioactive substances per
This is taken into account in the
assessment of radiological impacts, see
Biosphere Assessment.
309. The dose constraint for the most exposed individuals, 0.1 mSv
per year, stands for an average dose e.g. in a self-sustaining family
or small village community living in the environs of the disposal site,
where the highest radiation exposure arises via the various
pathways. In the living environment of this community, i.a. a small
lake and shallow water well is assumed to exist.
In Assessment of Radionuclide Release
Scenarios for the Repository System, the
releases arising from the expected
evolution scenarios over long time periods
have been shown to remain below the
constraints specified in YVL D.5, paragraph
312. See also responses to 312 and 313
below.
Repository System. The exposure
pathways defined in paragraph 308 of YVL
D.5 are considered in Biosphere
Assessment.
exposure pathways shall be considered:

use of contaminated water as household water, as irrigation
water and for animal watering

use of contaminated natural or agricultural products originating
from terrestrial or aquatic environments.
Constraints for releases of radioactive substances
311. Disposal of nuclear waste shall be planned so that, as a
consequence of expected evolution scenarios, the average
quantities of radioactive substances over long time periods,
released into the environment from disposed waste, shall remain
below the constraints specified separately for each nuclide by the
Radiation and Nuclear Safety Authority. These constraints shall be
set so that:
1) at a maximum, the radiation impacts arising from disposal can
be equivalent to those arising from natural radioactive
substances in earth’s crust
2) on a large scale, the radiation impacts remain insignificantly
low (GD 736/2008).
These constraints are applied to limiting the radiation exposures
arising beyond the assessment period referred to in paragraph 306.
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
258
Section 5 – Consideration of unlikely events
The significance of unlikely events impairing longterm safety shall be assessed by evaluating the
reality, probability and possible consequences of
each event. Whenever possible, the acceptability of
the expectancies of radiation impacts caused by
such events shall be evaluated in relation to the
annual dose and release rate constraints of
radioactive materials, as referred to in section 4.
Legal requirements set out in the Government
Decree (736/2008)
The unlikely events that have been
considered include:
315. Unlikely events induced by natural phenomenon to be
considered shall include at least rock movements jeopardizing the
The release rates arising from the
expected evolution scenarios over long
time periods have been calculated. For
some of the considered cases, the activity
releases have been averaged over 1000
years.
The sum of the ratios between the nuclide
specific activity releases and the respective
constraints is less than one.
See:
 Assessment of Radionuclide Release
Scenarios for the Repository System.
313. These constraints apply to activity releases which arise from
expected evolution scenarios and which may enter the environment
earliest after several thousands of years. These activity releases
can be averaged over 1000 years at the most. The sum of the ratios
between the nuclide specific activity releases and the respective
constraints shall be less than one.
Unlikely events have been considered as
part of disturbance scenarios discussed in:
 Formulation of Radionuclide Release
Scenarios
 Assessment of Radionuclide Release
Scenarios for the Repository System
 Biosphere Assessment
See also responses to 315 and 316 below.
bio fluxes”) have been calculated in:
 Assessment of Radionuclide Release
Scenarios for the Repository System
The releases remain below the nuclide
specific release constraints (see also 311
and 313).
annum) … are as follows:

0.03 GBq/a for the long-lived, alpha emitting radium, thorium,
protactinium, plutonium, americium and curium isotopes

0,1 GBq/a for the nuclides Se-79, Nb-94, I-129 and Np-237

0,3 GBq/a for the nuclides C-14, Cl-36 and Cs-135 and for the
long-lived uranium isotopes

1 GBq/a for the nuclide Sn-126

3 GBq/a for the nuclide Tc-99

10 GBq/a for the nuclide Zr-93

30 GBq/a for the nuclide Ni-59

100 GBq/a for the nuclide Pd-107.
Unlikely events
314. The importance of unlikely events impairing long-term safety
shall be assessed by considering the reality, likelihood and possible
consequences of each event. Whenever practicable, the radiation
impacts caused by such events shall be assessed quantitatively
(GD 736/2008).
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
259
Legal requirements set out in the Government
Decree (736/2008)
316. The importance to safety of such incidental events shall be
assessed and whenever practicable, the resulting annual radiation
dose or activity release shall be calculated and multiplied by its
estimated probability of occurrence. The obtained expectation value
shall be below the radiation dose constraint given in paragraph 306
or the activity release constraint given in paragraph 312. The
probability of such radiation exposure which might imply
deterministic radiation impacts (at least a dose of 0.5 Sv), shall be
extremely low.
integrity of disposal canisters. Unlikely events caused by human
actions to be considered shall include at least boring a mediumdeep water well at the site and a core drilling or boring hitting a
disposal canister. In such cases it is assumed that the existence of
the disposed waste is not known and that the event cannot take
place earliest 200 years after the closure of the disposal facility.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Rock movements (the RS and RSDIL scenarios).
Boring of a medium-deep water well
(the well scenarios in Biosphere
Assessment).
Core drilling hitting a canister (HICANISTER scenario, i.e. a human
intrusion scenario).
The unlikely events have been considered
as part of disturbance scenarios and the
resulting radiation dose or activity release
calculated for the selected scenarios.
Probability-weighted normalised release
rates have been applied to some of the
cases. The release rates are, even in the
most pessimistic case, around an order of
magnitude below the constraints given in
YVL D.5, paragraph 312.
The lines of evolution in the biosphere that
are considered unlikely, including the
human intrusion (HI) scenarios, and the
resulting doses have been assessed and
the expectation values of the calculated
doses fulfil the requirements in YVL D.5
paragraph 306.
For reference see:
 Formulation of Radionuclide Release
Scenarios
 Assessment of Radionuclide Release
Scenarios for the Repository System
See:
 Formulation of Radionuclide Release
Scenarios
 Assessment of Radionuclide Release
Scenarios for the Repository System
 Biosphere Assessment.



Summary position and location of
supporting evidence
260
Section 9 – Disposal operations
… The disposal package containing spent nuclear
fuel shall be designed so that no self-sustaining
chain reaction of fissions can occur, even in the
disposal conditions. …
Chapter 3: Design requirements for a nuclear
waste facility
Section 6 – Handling of spent nuclear fuel and
other nuclear waste
Spent nuclear fuel and other nuclear waste shall be
conditioned and packed in accordance with
disposal specifications. Waste packages shall be
classified on the basis of their characteristics.
Constraints and other quality specifications shall be
defined for each class, necessary in terms of the
operational safety of the nuclear waste facility and
the long-term safety of disposal, and which the
waste packages are required to meet. …
Legal requirements set out in the Government
Decree (736/2008)
Protection of other living species
317. Disposal shall not affect detrimentally to species of fauna and
flora. This shall be demonstrated by assessing the typical radiation
exposures of terrestrial and aquatic populations in the disposal site
environment, assuming the present kind of living populations. The
assessed exposures shall remain clearly below the levels which, on
the basis of the best available scientific knowledge, would cause
decline in biodiversity or other significant detriment to any living
population.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Biosphere Assessment.
The possibility of spent nuclear fuel
criticality in the canister is prevented by the
design of the canister and by the loading
strategy of the different assemblies in each
canister. OL1−2 and LO1−2 canisters will
remain subcritical even in the case of zero
burn-up (fresh fuel) and if the canister is
filled with water. Demonstrating the sub-
Waste package classification included in
Posiva´s Construction License Application
in Appendix 8
The exposures remain clearly below the
levels which, on the basis of the best
available scientific knowledge, would cause
decline in biodiversity or other significant
detriment to any living population.
Radiation exposures of flora and fauna are
assessed in:
 Dose Assessment for Plants and
Animals
 Biosphere Assessment

Summary position and location of
supporting evidence
261
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
4
PLANNING OF THE DISPOSAL METHOD
4.1 Stepwise implementation
401. Disposal shall be implemented in stages, with particular
attention paid to aspects affecting long-term safety. The planning of
the construction, operation and closure of a disposal facility shall
take account of reduction of the activity of nuclear waste through
interim storage, the utilization of high-quality technology and
scientific knowledge and the need to ensure long-term safety via
investigations and monitoring. However, the implementation of the
various stages of disposal must not be unnecessarily postponed
(GD 736/2008).
Legal requirements set out in the Government
Decree (736/2008)
Chapter 4: Long-term safety of disposal
Section 10 – General requirements concerning
disposal
Disposal shall be implemented in stages, with
particular attention paid to aspects affecting longterm safety. The planning of the construction,
operation and closure of a disposal facility shall
take account of reduction of the activity of nuclear
waste through interim storage, the utilisation of
high-quality technology and scientific data and the
need to ensure long-term safety via investigations
and monitoring. However, the implementation of the
various stages of disposal shall not be
unnecessarily postponed.
The aim of long-term safety has guided the
development of the disposal system and its
implementation.
The construction, operation and closure of
the disposal facility will be implemented in
stages. Prior to disposal, the spent fuel is
planned to be stored for at least tens of
years.
Current technology and scientific
knowledge has been used for the planning
of the disposal activities, and the
development in these fields will be followed
and taken into account in the
implementation stages. Investigations of
the site and the engineered barrier system
as well as monitoring of the site have been
carried out and will be continued focusing
on the requirements and needs of the
implementation stage.
The schedule outlined in the early 1980s
has been followed. Disposal is planned to
commence around 2020.
See:

Facility design (Saanio et al. 2013)

Site Description
criticality of larger OL3-type fuel elements
for spent nuclear fuel disposal requires the
use of burn-up credit, which is not yet an
internationally established practice in
criticality analysis. However,
complementary work will be done in the
future. See Canister Production Line report
and Description of the Disposal System,
Section 6.3.6.
Summary position and location of
supporting evidence
262
Legal requirements set out in the Government
Decree (736/2008)
402. Implementation of disposal of nuclear waste includes the
following phases:

selection of the disposal concept

selection and characterization of the disposal site, which may
include construction of an underground research facility at the
site

design of the disposal facility with related research and
development work

construction of the disposal facility

waste emplacement activities and other operations of the
disposal facility

backfilling and closure of emplacement rooms and other
underground openings

post-closure institutional control measures, if required.
These phases may be partly parallel.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
YJH-2012 (Posiva 2012a)
Performance Assessment
Posiva’s reference design in the
construction licence application is based on
vertical emplacement of the canisters
(KBS-3V). Currently, an alternative
horizontal emplacement design (KBS-3H)
is being jointly developed by the Swedish
Nuclear Fuel and Waste Management
Company (SKB) and Posiva. The final
decision has not yet been taken between
3V and 3H designs. See Design Basis and
YJH-2012 (Posiva 2012a).
The Olkiluoto site has been selected as a
site for the underground disposal facility
(DiP 2001). Site investigations have been
carried out for over 25 years and are still
ongoing focussing on the repository depth.
The ONKALO underground research
facility has been constructed and is in
operation. See Site Description.
For the construction licence application, the
design, production and description of the
initial state for the disposal system
components are presented in Production
Line reports. Plans for the research and
development work during the coming years
until 2018 are presented in YJH-2012
(Posiva 2012a).
The plan for the implementation, operation
and closure of the disposal facility is
presented by Saanio et al. (2013) and the
activities related to repository design and
implementation including demonstrations
during the coming years until 2018 is
presented in YJH-2012.


Summary position and location of
supporting evidence
263
Section 11 – Multibarrier principle
The long-term safety of disposal shall be based on
safety functions achieved through mutually
complementary barriers so that a deficiency of an
individual safety function or a predictable geological
change will not jeopardise the long-term safety.
Safety functions shall effectively prevent releases of
disposed radioactive materials into the bedrock for
a certain period, the length of which depends on the
duration of the radioactivity in waste. For short-lived
waste, this period shall be at least several hundreds
of years, and for long-lived waste, at least several
thousands of years.
Legal requirements set out in the Government
Decree (736/2008)
As in the case of the preceding phases, the
coming phases of the disposal programme
have been scheduled by allowing several
iterations between long-term safety
assessments (in the safety case) and
design (Design Basis and YJH-2012
(Posiva 2012a)).
Safety functions for the EBS and the host
rock are presented in Section 5.1.2 of the
Design Basis. Safety functions have been
assigned so that the deficiency in an
individual safety function or a predictable
geological change does not jeopardise the
long-term safety as shown in Performance
Assessment, Formulation of Radionuclide
Release Scenarios and Assessment of
Radionuclide Release Scenarios for the
Repository System.
The engineered barriers and their safety
functions are presented in Section 5.1.2 of
Design Basis. The waste matrix is not
considered as an engineered barrier in the
Posiva safety concept, although it is
acknowledged that it has properties that
promote long-term safety (see Design
Basis). Rather, safety functions are
assigned to the defined engineered
barriers such that safety is assured for the
expected fuel types from Olkiluoto and
Loviisa.
4.2 Barriers and safety functions
404. The long-term safety of disposal shall be based on safety
functions achieved through mutually complementary barriers so that
a deficiency of an individual safety function or a predictable
geological change will not jeopardise the long-term safety (GD
736/2008).
405. Engineered barriers and their safety functions may consist of

waste matrix, in which radioactive substances are incorporated

hermetic, corrosion resistant and mechanically strong
container, in which the waste is enclosed

chemical environment around waste packages, which limits the
dissolution and migration of radioactive substances

material around waste canisters (the buffer), which provides
containment and yields minor rock movements

other containment structures in the emplacement rooms

backfilling materials and sealing structures, which limit
transport of radioactive substances through excavated
openings.
Construction of the disposal facility, waste
emplacement and closure of completed
sections of the disposal facility will be done
partly in parallel.
Summary position and location of
supporting evidence
403. The various phases of disposal shall be scheduled and
implemented giving priority to safety. Preparedness for moving to
the next phase shall be assessed as a whole taking account of the
suitability of the disposal concept and site, technical feasibility and in
particular the outcome of and confidence in the long-term safety
assessments.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
264
Legal requirements set out in the Government
Decree (736/2008)
Summary position and location of
supporting evidence
The repository host rock functions as a
natural barrier and its safety functions are
presented in Section 5.1.2 of Design Basis.
Performance requirements, i.e.
performance targets for the engineered
barriers and target properties for the host
rock have been formulated within Posiva’s
requirement management system VAHA.
Details on the reasoning and rationale,
including the design basis scenarios (the
time dependent conditions and loads,
taken into account in the definition of the
performance requirements) are described
in Design Basis.
The performance targets for the EBS of the
spent nuclear fuel disposal facility are
defined with the aim that the safety
functions are maintained and the EBS will
retain its functionality for hundreds of
thousands of years (Design Basis). For the
expected lines of evolution, the EBS
prevents the releases of radioactive
substances for at least 10,000 years. See
Performance Assessment.
In accordance with the repository concept,
the safety functions for the barriers are
designed so that they are not sensitive to
changes in bedrock and the characteristics
of the canisters or the disposal
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
406. Natural barriers and their safety functions may consist of

stable and intact rock with low groundwater flow rate around
disposal canisters

rock around waste emplacement rooms where low
groundwater flow, reducing and also otherwise favourable
groundwater chemistry and retardation of dissolved substances
in rock limit the mobility of radionuclides

protection provided by the host rock against natural
phenomena and human actions.
407. Targets based on high quality scientific knowledge and expert
judgement shall be specified for the performance of each safety
function. In doing so, the potential changes and events affecting the
disposal conditions during each assessment period shall be taken
into account. In an assessment period extending up to several
thousands of years, one can assume that the bedrock of the site
remains in its current state, taking however account of the changes
due to predictable processes, such as land uplift and those due to
excavations and disposed waste.
408. Performance targets for the safety functions of engineered
barriers shall be specified taking account of the activity level of
waste and the half-lives of dominating radionuclides. The safety
approach for disposal of spent fuel shall be that the safety functions
provided by the engineered barriers will limit effectively the release
of radioactive substances into bedrock for at least 10 000 years. ...
409. The design of the safety functions shall aim to provide a
disposal concept that is not sensitive to changes in the bedrock.
Another design objective shall be that the characteristics of waste
packages or the disposal environment will not evolve with time in a
way that may affect adversely the safety functions.
265
The overall suitability of the Olkiluoto site
and the properties of the host rock suitable
as a natural barrier are discussed in Site
Description, Complementary
Considerations and RSC-2012 (McEwen et
al. 2013). Although a few factors have
been identified that constrain the repository
layout, no factors indicating unsuitability of
the site have been found. The suitability of
the Olkiluoto site was discussed in the
safety assessment submitted in support of
the Decision-in-Principle of 2001.
The favourable properties of the host rock
with respect to EBS performance are
4.3 Disposal site and facility
410. The bedrock of the disposal site shall be such that it
adequately acts as a natural barrier, as specified in paragraph 406.
Factors indicating unsuitability of a disposal site may include at least

proximity of exploitable natural resources

abnormally high rock stresses with regard to the strength of the
rock

predictable anomalously high seismic or tectonic activity

exceptionally adverse groundwater characteristics, such as
lack of reducing buffering capacity and high concentrations of
substances which might substantially impair the safety
functions.
411. The characteristics of the host rock shall be favourable
regarding the long-term performance of engineered barriers. Such
Section 12 – Disposal site
The geological characteristics of the disposal site
shall, as a whole, be favourable to the isolation of
the radioactive substances from the environment.
Any area with a feature that is substantially adverse
to long-term safety shall not be selected as the
disposal site.
The planned final disposal site shall contain
sufficiently large, intact rock volumes that facilitate
the construction of the waste emplacement rooms.
For the purposes of disposal facility design and
acquiring data required for safety assessments, the
geological characteristics of the host rock at the site
shall be characterized through investigations at the
intended disposal depth, in addition to surface
based investigations. The layout, excavation,
environment will not evolve with time in a
way that could affect adversely the safety
functions (Design Basis and Performance
Assessment).
According to the performance requirements
and the technical design requirements of
the various barriers, the barriers shall not
adversely affect each other and deposition
tunnels and deposition holes shall be
located (through application of rock
suitability classification (RSC) criteria) so
that the performance of the EBS is
ensured. The long-term safety aspects are
strongly emphasised in the material
selection, design and manufacturing of the
EBS components and in the underground
construction, e.g. through the selection of
well understood materials.
See Design Basis, Production Line reports
and RSC-2012 (McEwen et al. 2013).
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Legal requirements set out in the Government
Decree (736/2008)
266
Current and predictable future conditions at
the selected repository depth of 400 to 450
m are considered to be favourable for the
long-term safety of the repository, see
RSC-2012 (McEwen et al. 2013).
The classification of systems, structures
and components has been presented in the
Classification Document, which will be
412. The location of the repository shall be favourable with respect
to the groundwater flow regime at the disposal site. The disposal
depth shall be selected giving priority to long-term safety, taking into
account the geological structures of the bedrock as well as the
trends with depth in hydraulic conductivity, groundwater chemistry
and rock stress - strength ratio. The repository for spent fuel shall be
located at the depth of several hundreds of metres in order to
mitigate adequately the impacts from aboveground natural
phenomena, such as glaciation, and human actions. (The
repositories for other long-lived wastes and those for short-lived
wastes shall be located at the depth of some tens of metres as a
minimum.)
5
DESIGN OF THE DISPOSAL FACILITY AND PRACTICES
5.2 Design of structures, systems and practices
See Chapter 3 above
described by the target properties. The
future conditions including geological
changes related to climate changes have
been taken into account in defining the
target properties (Design Basis and RSC2012 (McEwen et al. 2013)). No major
climate changes are expected within at
least several thousands of years
(Formulation of Radionuclide Release
Scenarios). The fulfilment of the target
properties has been evaluated in
Performance Assessment. The suitability of
the Olkiluoto host rock and its compatibility
with the EBS is discussed in Site
Description, Complementary
Considerations report and RSC-2012
report.
The effects of large-scale climate changes,
principally sea-level fall, and periods of
permafrost and glaciation, have also been
taken into account (see Performance
Assessment).
conditions in the bedrock as are of importance to long-term safety
shall be stable or predictable up to at least several thousands of
years. The range of geological changes which occur thereafter,
particularly due to the large scale climate changes, shall be
estimable and be considered in specifying the performance targets
for the safety functions.
construction and closure of underground facilities
shall be implemented so that the characteristics of
the host rock deemed important in terms of longterm safety are retained, as far as possible.
The depth of the waste emplacement rooms shall
be selected appropriately as regards the waste type
and local geological conditions. The goal related to
disposal depth shall be that any impacts on the
long-term safety of above-ground events, activities
and environmental changes will remain minor and
that intrusion into the waste emplacement rooms
will be difficult.
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Legal requirements set out in the Government
Decree (736/2008)
267
Legal requirements set out in the Government
Decree (736/2008)
Summary position and location of
supporting evidence
delivered to STUK in conjunction with the
construction licence application. The
classification takes account of the
classification principles given in the
requirement in paragraph 507 of YVL D.5.
The long-term safety has been considered
in the safety classification of those systems
(e.g. the EBS components) that have longterm safety relevance. The classification is
presented in the Classification Document,
The rock suitability classification (RSC)
system describes the approach to
identifying suitable locations for the
disposal panels, the deposition tunnels and
the deposition holes and the
characterisation studies related to RSC
application are discussed in the RSC-2012
report and future plans in YJH-2012
(Posiva 2012a).
The monitoring programme that has been
applied before the ONKALO construction
and during the ONKALO construction is
described in the Monitoring programme
report 2003 (Posiva 2003) and the results
in annual reports for each discipline. The
monitoring programme for the coming
years is presented in the Posiva report
“Monitoring at Olkiluoto - a Programme for
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Classifications
507. Systems, structures and components of a disposal facility shall
be classified according to their functional and structural importance
to safety. The classification shall be based besides the operational
safety, also [on] the long-term safety of disposal. The safety class
shall be considered in setting requirements for the design,
fabrication, installation, testing and inspection of a system, structure
or component. Structures and components shall also be classified
on the basis of resistance to environmental conditions. …
509. Regarding the long-term safety of disposal, the classification
shall be based on structures and functions which have considerable
impact on the safety functions referred to in paragraphs 405 and
406 or which may have such adverse impacts on long-term safety
as referred to in paragraphs 512. Structures and functions of
importance are notably waste packages with surrounding buffer
materials and containment structures, and the disposition,
excavation and injection of the underground openings in the
disposal facility.
Construction, operation and closure of the disposal facility
510. During the construction and operation of the disposal facility,
an investigation, testing and monitoring program shall be executed
to ensure the suitability for disposal of the rock to be excavated, to
determine safety relevant characteristics of the host rock and to
ensure long-term performance of barriers. This program shall
include at least

characterization of the rock volumes intended to be excavated

monitoring of rock stresses, movements and deformations in
rock surrounding the waste emplacement rooms

hydrogeological monitoring of rock surrounding the waste
emplacement rooms

monitoring of groundwater chemistry at the disposal site

monitoring of the behaviour of engineered barriers.
268
Legal requirements set out in the Government
Decree (736/2008)
Bedrock structures to be avoided are being
classified based on the rock suitability
classification (RSC) system. The RSC
includes also criteria for the selection of
suitable host rock for the underground
disposal facility, most importantly for
deposition tunnels and deposition holes.
The repository layout will be updated
based on the results of detailed
investigations and rock suitability
classification carried out in stages. The
classification process and its coupling with
layout design are described in the RSC2012 report (McEwen et al. 2013).
The requirements related to construction
and closure of the disposal facility are
defined in Posiva´s requirements
management system and presented in
Design Basis and in the Production Lines
Reports.
Discussed in Posiva´s CLA documentation
as a part of the document STUK1
responding to the fulfillment of the YVL D.5
512. The construction, operation and closure of the waste
emplacement rooms and other underground openings shall aim at
maintaining the rock characteristics important to long-term safety.
For this purpose, particularly in case of the implementation of spent
fuel disposal,

such rock construction methods shall be used that limit the
excavation disturbances in rock around waste emplacement
rooms

reinforcement and injection of host rock shall be done so that
no significant amounts of substances detrimental to the
performance of barriers enter the waste emplacement rooms

introduction of organic and oxidising substances to the waste
emplacement rooms shall be minimised

waste emplacement rooms shall be backfilled and closed as
soon as expedient with regard to the disposal activities and
related monitoring activities.
513. The layout of the disposal facility shall be designed so that the
waste emplacement activities are appropriately separated from the
transfers of excavated rock, backfill materials and heavy machinery.
the Period Before Repository Operation”
(Posiva 2012-01).
Summary position and location of
supporting evidence
511. Such structures and other characteristics of rock surrounding
the waste emplacement rooms which may have importance
regarding groundwater flow, rock movements or other factors
affecting long-term safety, shall be defined and classified.
Modifications of the layout of the underground openings shall be
provided for in case that the quality of rock surrounding the
designed excavations proves to be significantly inferior to the design
basis.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
269
Compliance with the radiation exposure constraints
for the most exposed people, as referred to in
section 4 above, shall be proven by considering a
community that derives nutrition from the immediate
Chapter 5: Demonstration of compliance with
safety requirements
Section 14 – Long-term safety
Compliance with the requirements concerning longterm radiation safety, and the suitability of the
disposal method and disposal site, shall be proven
through a safety case that must analyse both
expected evolution scenarios and unlikely events
impairing long-term safety. The safety case
comprises a numerical analysis based on
experimental studies and complementary
considerations insofar as quantitative analyses are
not feasible or involve considerable uncertainties.
Legal requirements set out in the Government
Decree (736/2008)
Guide requirements.
According to the design basis, the
canisters will be stored, transferred and
emplaced in a way that the copper shell is
not damaged (Design Basis). The buffer
and the backfill − as well as the related
quality assurance measures − have been
designed so that their quality remains
unchanged during the installation process.
The long-term safety aspects form the
basis for the design of all EBS production
lines, including transfer and installation
(Canister, Buffer and Backfill Production
Line reports).
The TURVA-2012 Safety Case portfolio is
composed of reports that satisfy the
requirements listed in 704 as follows:

see Description of the Disposal
System, Site Description and Design
Basis

see Design Basis

see Performance Assessment and
Formulation of Radionuclide Release
Scenarios

see Design Basis, Models and Data
for the Repository System, Biosphere
Data Basis, Terrain and Ecosystems
Development Modelling, Surface and
Near-Surface Hydrological Modelling,
Biosphere Transport and Dose
Assessment and Dose Assessment for
Excavation induced rock collapses or displacements in openings
where waste canister emplacement is underway or completed, shall
be prevented by careful excavation, rock support, and by keeping
these openings at sufficient distance from the excavation activities.
514. Transfer into the disposal position of a spent fuel canister or
other waste package with long-term durability requirements,
alongside the installation of buffer and backfill materials, shall be
performed so that no damage compromising the performance of the
engineered barriers will occur.
7
DEMONSTRATION OF COMPLIANCE WITH SAFETY
REQUIREMENTS
7.1 Principles for safety demonstration
Long-term safety
704. Compliance with the long-term radiation protection
requirements as well as the suitability of the disposal method and
site shall be demonstrated by means of a safety case that shall
include at least

description of the disposal system and the definition of barriers
and safety functions

determination of performance goals for the safety functions

definition of the evolutions describing the potential future
behaviour of the disposal system (scenario analysis)

functional description of the disposal system by means of
conceptual and mathematical modelling and the determination
of the input data needed in these models

analysis of the quantities of radioactive substances that are
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
270
The disposal system is described in detail
in Description of the Disposal System. The
characteristics of the host rock are
described in greater detail in the Site
Description and the EBS and the
underground openings in a series of
Production Line reports.
The barriers and their safety functions
including performance targets for the EBS
A03. The safety case shall define the safety concept, barriers and
safety functions and specify their performance targets. In doing so,
Plants and Animals
see Assessment of Radionuclide
Release Scenarios for the Repository
System and Biosphere Assessment
see Performance Assessment,
Assessment of Radionuclide Release
Scenarios for the Repository System
and Biosphere Assessment
see Performance Assessment,
Assessment of Radionuclide Release
Scenarios for the Repository System
and Complementary Considerations
see Assessment of Radionuclide
Release Scenarios for the Repository
System, Biosphere Assessment and
Synthesis.
A02. The safety case shall include a description of the disposal
system: quantities of radioactive materials, waste packages, buffer
materials, backfill materials, containment and blockage structures,
excavations, geological, hydrogeological, hydrochemical, thermal
and rock mechanical characteristics of the host rock, and the natural
environment at the disposal site.




Appendix A of YVL D.5 has been used as a
basis for planning the Safety Case
portfolio, which is described in Synthesis
Section 1.4. The Safety Case portfolio
includes all the required contents defined in
Appendix A.
released from the waste, penetrate the barriers and enter the
biosphere, and analysis of the resulting radiation doses

whenever practicable, estimation of probabilities of activity
releases and radiation doses arising from unlikely events
impairing long-term safety

uncertainty and sensitivity analyses and complementary
considerations

comparison of the outcome of the analyses with safety
requirements.

(See also paragraphs 309 ‘most exposed individuals’ and 317
‘protection of other living species’ above.)
Appendix A includes detailed requirements for the content of the
safety case.
surroundings of the disposal site and is most
exposed to radiation. In addition to impacts on
people, possible impacts on flora and fauna shall be
analysed.
Summary position and location of
supporting evidence
APPENDIX A: SAFETY CASE
A01. Compliance with the requirements concerning long-term
radiation safety, and the suitability of the disposal method and
disposal site, shall be proven through a safety case that must
analyze both expected evolution scenarios and unlikely events
impairing long-term safety. The safety case comprises a numerical
analysis based on experimental studies and complementary
considerations insofar as quantitative analyses are not feasible or
involve considerable uncertainties (GD 736/2008).
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Legal requirements set out in the Government
Decree (736/2008)
271
Legal requirements set out in the Government
Decree (736/2008)
Summary position and location of
supporting evidence
and target properties for the host rock are
defined in Design Basis. Temporal and
stochastic variations have been taken into
account.
The scenarios have been formulated so
that they consider:

the most likely lines of evolution

various situations where one or
several safety functions have
significantly degraded

lines of evolution with an extremely
low probability but which cannot be
completely ruled out. See A05 below.
The scenarios have been formulated by
considering the features, events and
processes (FEPs) that may be of
importance to long-term safety and the
interactions between the FEPs. These are
discussed in Features, Events and
Processes. The potentially significant FEPs
have been taken into account in
Performance Assessment and Formulation
of Radionuclide Release Scenarios, and
the analysis of radionuclide release
scenarios is reported in Assessment of
Radionuclide Release Scenarios for the
Repository System and in Biosphere
Assessment.
The base scenario assumes the targets
defined for each safety function
(performance targets, target properties and
safety functions are fulfilled), taking
account of incidental deviations from the
target values. The base scenario considers
the possibility that there is an undetected
initial penetrating defect in one or a few
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
temporal and stochastic variations due to e.g. geological and
climatic processes shall be taken into account.
A04. The safety case shall include a scenario analysis which covers
both the expected evolutions and unlikely events impairing longterm safety. The scenarios shall be constructed so that they cover
the features, events and processes which may be of importance to
long-term safety and which may arise from

interactions within the disposal system, caused by radiological,
mechanical, thermal, hydrological, chemical biological or
radiation induced phenomena

external factors, such as climate changes, geological
processes or human actions.
A05. The base scenario shall assume the performance targets
defined for each safety function, taking account of incidental
deviations from the target values. The influence of declined
performance of a single safety function or, in case of coupling
between safety functions, the combined effect of declined
performance of more than one safety function, shall be analysed by
means of variant scenarios. Disturbance scenarios shall be
constructed for the analysis of unlikely events impairing long-term
272
Section 15 – Reliability of the safety case
The input data and models utilised in the safety
case shall be based on high-quality research data
and expert judgement. Data and models shall be
Legal requirements set out in the Government
Decree (736/2008)
Summary position and location of
supporting evidence
canisters. The variant scenarios consider
the influence of declined performance of a
single safety function and also the
combined effect of declined performance of
more than one safety function. The
disturbance scenarios consider unlikely
events impairing long-term safety (see
Formulation of Radionuclide Release
Scenarios).
The models used for analysing the
radionuclide release and transport and the
performance of the repository system are
described in Models and Data for the
Repository System report. The key models
used for the biosphere assessment are
described in the respective modelling
reports (Terrain and Ecosystems
Development Modelling, Surface and NearSurface Hydrological Modelling, Biosphere
Radionuclide Transport and Dose
Assessment, Dose Assessment for Plants
and Animals) and regarding the conceptual
models in Biosphere Description.
The conceptual models as well as
numerical models are described as well as
the processes considered in the models.
Assumptions and simplifications made for
the modelling are also described. Such
simplifications are made following the
principle that the performance of safety
functions is neither overestimated nor
overly underestimated.
The models and data used in TURVA2012, and specific actions undertaken to
promote confidence in these, are described
in Models and Data for the Repository
System and in Biosphere Data Basis and
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
safety.
A06. In order to analyse the release and transport of disposed
radioactive substances, conceptual models shall be drawn up to
describe the physical phenomena and processes controlling the
safety functions. Besides the modelling of release and transport
processes, models are needed to describe the circumstances
affecting the performance of safety functions. From the conceptual
models, the respective computational models are derived, normally
with simplifications. Simplification of the models and the
determination of the required input shall be based on the principle
that the performance of a safety function will not be overestimated
while neither overly underestimated.
A07. Modelling and determination of input data shall be based on
high-quality scientific knowledge and expert judgement obtained
through experimental studies, such as laboratory experiments, site
investigations and evidence from natural analogues. The models
and input data shall be consistent with the scenario, assessment
273
modelling reports (Terrain and Ecosystems
Development Modelling, Surface and NearSurface Hydrological Modelling, Biosphere
Radionuclide Transport and Dose
Assessment, Dose Assessment for Plants
and Animals). At a more general level,
evidence for long-term safety from natural
analogues has been used, the latter being
reported in Complementary Considerations
report along with observations at the site.
A stochastic modelling approach has been
applied whenever random variations are of
significance either in the geosphere or
surface environment (Performance
Assessment, and Assessment of
Radionuclide Release Scenarios for the
Repository System).
As noted under A06, the modelling carried
out for TURVA-2012 aims to neither
overestimate nor overly underestimate
safety functions and retention properties.
Uncertainties are taken into account in
defining the range of scenarios and
calculation cases in Formulation of
Radionuclide Release Scenarios, as well
as in other complementary analyses of
sensitivities and uncertainties described in
Assessment of Radionuclide Release
Scenarios for the Repository System and in
Biosphere Assessment.
An assessment of confidence is included in
Synthesis.
Complementary considerations, including
analyses by simplified methods,
comparisons with natural analogues and
observations of the geological history of the
period and disposal system. Whenever the input data used in
modelling involve random variations due to e.g. heterogeneity of
rock, stochastic models may be employed.
A08. Selection of the computational methods, performance targets
and input data shall be based on principle that the actual radiation
exposures or quantities of released radioactive substances shall
with high degree of certainty be lower than those obtained through
safety analyses. The uncertainties included in the safety analysis
shall be assessed by means of appropriate methods, e.g. by
sensitivity analyses or probabilistic methods. The safety case shall
include an assessment of the confidence level with regard to
compliance with the safety requirements and of uncertainties with
most contribution to the confidence level.
A09.The importance to safety of such scenarios that cannot
reasonably be assessed by means of quantitative safety analyses
shall be examined by means of complementary considerations.
They may include e.g. analyses by simplified methods, comparisons
validated as far as possible, and correspond to the
conditions likely to prevail at the disposal site during
the assessment period.
The basis for selecting the computational methods
used shall be that the actual radiation exposure and
quantities of radioactive materials released remain
below the results of safety analyses, with a high
degree of certainty. The uncertainties involved in
the safety analysis, and their significance, shall be
separately assessed.
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
Legal requirements set out in the Government
Decree (736/2008)
274
Chapter 7: Organisation and personnel
Section 19 – Safety culture
When designing, constructing, operating and
decommissioning or closing a nuclear waste facility,
a good safety culture shall be maintained. In its
decisions and operations, the management of the
organisation concerned shall demonstrate its
commitment to procedures and solutions promoting
safety. Personnel shall be motivated to perform
responsible work and an open working atmosphere
shall be promoted in the working community, in
order to encourage the identification, reporting and
elimination of factors endangering safety. Personnel
shall be given the opportunity to contribute to the
continuous safety enhancement.
Section 16 – Presentation of, and updates to,
the safety case
The safety case shall be presented in connection
with the construction licence application and the
operating licence application of the nuclear waste
facility. The safety case shall be updated at 15 year
intervals unless otherwise provided in the licence
conditions. Furthermore, the safety case shall be
updated prior to the permanent closure of the
facility.
Legal requirements set out in the Government
Decree (736/2008)
see A10 and A11 below?
Olkiluoto site are included in
Complementary Considerations report.
Complementary indicators are also
discussed in Assessment of Radionuclide
Release Scenarios for the Repository
System report and safety indicators in
Biosphere Assessment.
with natural analogues or observations of the geological history of
the disposal site. The significance of such considerations grows as
the assessment period increases, and safety evaluations extending
beyond [a] time horizon of one million years can mainly be based
on the complementary considerations. Complementary
considerations shall also be applied parallel to the actual safety
assessment in order to enhance the confidence in results of the
analysis or [a] certain part of it.
See also A11 and A12 below.
The current safety case portfolio has been
compiled to be included in the CLA.
Synthesis provides the summary of the
whole safety case.
Synthesis also includes an executive
summary.
Plans for further implementations of the
safety case and underlying investigations
have been included in the YJH-2012 report
(Posiva 2012a).
Organisations and personnel are presented
in CLA Appendix 10 and 15, where
organizations for operating period and
construction time are presented,
respectively. Also, CLA Appendix 7
describes organization in perspective as
part of safety culture oriented organization.
In Appendix 8, VNA 736/2008 17 and 19 §
demands for organizations are
compensated. In addition, Posiva will
during 2013 submit a separate report on
safety culture and management.
Summary position and location of
supporting evidence
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
275
Summary position and location of
supporting evidence
The reports within the safety case have
been written to be read as standalone
reports and to provide references to other
reports for further reading. Key models and
data are collected in individual reports
Models and Data for the Repository
System, Biosphere Data Basis and the
biosphere modelling reports (Terrain and
Ecosystems Development Modelling,
Surface and Near-Surface Hydrological
Modelling, Biosphere Radionuclide
Transport and Dose Assessment, Dose
Assessment for Plants and Animals).
The production process, organisation and
Quality Assurance of the Safety Case
process are described in Section 2.5 of the
Synthesis. External experts have been
used in the review process. The key data
have been evaluated using the Expert
Elicitation process.
Regulatory guidance set out in STUK Guide YVL D.5 (Draft 4,
17.3.2011)
A10. The safety case shall be documented carefully. In each part of
the safety case, the basic assumptions, used methods, obtained
results and coupling to wholeness case shall be evident (clarity) and
the justifications for the adopted assumptions, input data and
models shall be easily found in the documentation (traceability).
A11. The quality of the safety case shall be ensured through the
management system related to the design, construction and
operation of the disposal facility. The implementer of the project
shall establish an expedient organisation, adequate competence
and appropriate information management system. The various
stages of the preparation of the safety case shall be planned
systematically and the reliability of the results of important studies
and analyses shall be confirmed by independent experts or
analyses.
Legal requirements set out in the Government
Decree (736/2008)
Section 20 – Safety and quality management
Organisations participating in the design,
construction, operation and decommissioning or
closure of a nuclear waste facility shall employ a
management system for ensuring the management
of safety and quality. The objective of the
management system is to ensure that safety is
prioritised without exception, and that quality
management requirements are commensurate with
the significance to safety of the activity. This
management system shall be systematically
assessed and further developed.
Safety and quality management shall cover all
activities influencing the safety of the nuclear waste
facility. For each activity, requirements significant in
safety terms shall be identified, and planned
measures described in order to ensure compliance
with requirements. The processes and procedures
shall be systematic and based on instructions.
Systematic procedures shall be in place for
identifying and correcting deviations significant in
safety terms.
The licensee shall commit and oblige its employees
and suppliers, subcontractors and other partners
contributing to safety relevant activities to engage in
systematic safety and quality management.
276
277
APPENDIX 3: REPOSITORY SYSTEM COMPONENTS FEPS AND
SCENARIOS
Components
SPENT FUEL
System
Evolution
Radioactive
decay (and ingrowth)
Heat
generation
Heat transfer
Structural
alteration of
fuel pellets
Radiolysis of
residual water
(i.c.)
Radiolysis of
the canister
water
Corrosion of
cladding
tubes
Alteration &
dissolution of
the fuel matrix
Release of
the labile
fraction of the
inventory
Production of
He gas
Criticality
Migration
CANISTER
Radiation
attenuation
Heat transfer
Deformation
Thermal
expansion of
the canister
Corrosion of
the copper
overpack
Corrosion of
the cast iron
insert
Stress
corrosion
cracking
AUXILIARY
COMPONENTS
BACKFILL
Heat transfer
Water uptake
and swelling
Heat transfer
Water uptake
and swelling
Heat transfer
Stress
redistribution
Piping and
erosion
Piping and
erosion
Reactivationdisplacements
Chemical
erosion
Chemical
erosion
Spalling
Radiolysis of
porewater
Creep
Montmorillonite transformation
Alteration of
accessory
minerals
Montmorillonite transformation
Alteration of
accessory
minerals
Microbial
activity
Freezing and
thawing
Aqueous
solubility and
speciation
Precipitation
and coprecipitation
Sorption
Diffusion
Advection
Colloid
transport
Gas transport
Aqueous
solubility and
speciation
Aqueous
solubility and
speciation
Microbial
activity
Freezing and
thawing
Aqueous
solubility and
speciation
Precipitation
and coprecipitation
Sorption
Diffusion
Precipitation
and coprecipitation
Sorption
Diffusion
Advection
Colloid
transport
Gas transport
Precipitation
and coprecipitation
Sorption
Diffusion
Advection
Colloid
transport
Gas transport
GEOSPHERE
1
BUFFER
Chemical
degradation
Physical
degradation
Erosion and
sedimentation
in fractures
Rock-water
interaction
Microbial
activity
Freezing and
thawing
Transport
through
auxiliary
components
Aqueous
solubility and
speciation
Precipitation
and coprecipitation
Sorption
Diffusion
Advection
Colloid
transport
Gas transport
All these FEPs are taken into account Implicitly or explicitly in all the radionuclide release scenarios All these FEPS are taken into account in the scenario Variant 1. Also in AIC excepting piping and erosion and montmorillonite transformation This FEP is taken explicitly into account in the scenario Variant 2 along with all the FEPs in green with the exception of piping and erosion. It is also taken into account in RS‐DIL These FEPs are taken explicitly into account in all RS scenarios
See explanation in the main text in Chapter 8 1) Methane hydrate formation and salt exclusion are neither explicitly or implicitly dealt with in any scenario
because they are highly unlikely
LIST OF REPORTS
15.2.2013
POSIVA-REPORTS 2012
_______________________________________________________________________________________
POSIVA 2012-01
Monitoring at Olkiluoto – a Programme for the Period Before
Repository Operation
Posiva Oy
ISBN 978-951-652-182-7
POSIVA 2012-02
Microstructure, Porosity and Mineralogy Around Fractures in Olkiluoto
Bedrock
Jukka Kuva (ed.), Markko Myllys, Jussi Timonen,
University of Jyväskylä
Maarit Kelokaski, Marja Siitari-Kauppi, Jussi Ikonen,
University of Helsinki
Antero Lindberg, Geological Survey of Finland
Ismo Aaltonen, Posiva Oy
ISBN 978-951-652-183-4
POSIVA 2012-03 Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Design Basis 2012 ISBN 978-951-652-184-1
POSIVA 2012-04
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Performance Assessment 2012
ISBN 978-951-652-185-8
POSIVA 2012-05
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Description of the Disposal System 2012
ISBN 978-951-652-186-5
POSIVA 2012-06
Olkiluoto Biosphere Description 2012
ISBN 978-951-652-187-2
POSIVA 2012-07
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Features, Events and Processes 2012
ISBN 978-951-652-188-9
POSIVA 2012-08
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Formulation of Radionuclide Release Scenarios 2012
ISBN 978-951-652-189-6
POSIVA 2012-09
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Assessment of Radionuclide Release Scenarios for the Repository
System 2012
ISBN 978-951-652-190-2
POSIVA 2012-10
Safety case for the Spent Nuclear Fuel Disposal at Olkiluoto - Biosphere
Assessment BSA-2012
ISBN 978-951-652-191-9
POSIVA 2012-11
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Complementary Considerations 2012
Posiva Oy
ISBN 978-951-652-192-6
POSIVA 2012-12
Safety Case for the Disposal of Spent Nuclear Fuel at Olkiluoto Synthesis 2012
ISBN 978-951-652-193-3