ITER
G 84 RI 3 01-07-13 R1.0
ITER
Generic Site Safety Report
VOLUME III
RADIOLOGICAL AND ENERGY SOURCE TERMS
GSSR
ITER
G 84 RI 3 01-07-13 R1.0
PREFACE
The Generic Site Safety Report (GSSR) is an integrated, plant-level safety assessment of the
ITER design and implementation of a generic safety approach utilising a generic site intended
to support siting by any Party. After siting the ITER safety design and implementation will
be finalised in accordance with host country regulations and practices. The GSSR provides
site-independent input for an environmental impact assessment and for safety
characterisation. The main purpose of GSSR is to provide technical information to potential
host countries, in combination with other ITER documentation, to assist in preparing
regulatory submissions. The GSSR is not intended as the regulatory submission, i.e. it is not
a Preliminary Safety Analysis Report (PSAR) and does not follow a template or style of a
typical PSAR of any potential host country. GSSR is not a stand-alone document and
references other documents, such as the Plant Design Specification (PDS), Design
Requirements and Guidelines (DRG), and Plant Description Document (PDD) in providing
general safety criteria, plant descriptions, operational scenarios, detailed design, etc.
GSSR
ITER
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Executive Summary
This volume presents the radioactive sources, chemically hazardous or reactive substances,
and stored energy sources to be used or handled in ITER. These inventories and energies
consist of:
•
•
•
Radioactive materials (tritium fuel, neutron activation products in structures and in-vessel
dust, neutron activated corrosion products in coolant water);
Conventional (non-nuclear) hazardous materials (beryllium, hydrogen gas, etc.); and
Stored energy sources (plasma, magnets, nuclear energy (decay heat), thermal energy of
coolants, and chemical energy).
Tritium will be supplied from off-site sources. The total site inventory is less than 3 kg. For
each nominal 400 second pulse (effective time duration 440 s) at 500 MW fusion power, 0.39
g of tritium is burned. However the throughput of tritium per pulse maybe as high as 130 g
(burn fraction ~ 0.3%).
Co-deposition will be the dominant process for tritium uptake by carbon. Thick carbon
deposits with high levels of deuterium have been observed on graphite tiles exposed to
deuterium plasma discharges in TFTR, DIII-D, JET, TEXTOR, and other tokamaks. The
tritium retention at the "DTE1" tritium campaign at JET was initially 40% during tritium
fuelling by gas puffing. This large rate can be partly understood by isotopic exchange with
deuterium saturated walls. The longer term retention during fuelling of only deuterium was
about 20%. However, compared to ITER, results from present-day machines (with several
second pulse length) may be affected dispropotionally by transient effects such as impurity
generation and transport during plasma start-up and shutdown. In divertor tokamaks, the
main chamber wall is an area of net erosion, while both erosion and deposition occur in the
divertor. Note that there is a further significant difference between today's carbon tokamaks
and ITER which will operate with a beryllium FW and hence have a smaller source for
carbon errosion.
Estimating the co-deposited tritium inventory requires knowing the particle fluxes and target
design information, the sputtering yield, co-deposition rates (for example, hydrogen/carbon
ratios), and the ability to perform transient modeling of the full sputtering and re-deposition
processes. The calculations yield a tritium co-deposition rate of 2-5 g-T/400 s pulse. The Be
co-deposition rate is estimated to be in a range of 0.1-0.4 g-T/400 s pulse. Co-deposition of
tritium with tungsten is not observed. The present analyses are at most reliable for indicating
trends, not firm quantitative predictions.
Without removal of this tritium, kilograms of tritium could be retained in the co-deposited
layer of ITER. Therefore, active and effective methods to remove the co-deposited layers are
needed in ITER. The only effective methods for removing tritium known so far involve (1)
oxidation of the co-deposited layers (e.g., thermo-oxidative erosion > 250˚C, or oxygen
plasma discharges) or (2) physical removal.
A simplified yet conservative approach to mobilization of in-vessel tritium has been
established. Full mobilization is conservatively assumed for temperatures above 200ºC and a
linear behavior of mobilization is assumed between 0 and 200ºC.
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All in-vessel components are cooled by water. Diffusion of implanted tritium into the
primary first wall, baffle, and divertor during operation and baking will result in tritium
contamination of the cooling water of those components. Permeation is dominated by the
phases of baking when the in-vessel components are at elevated temperatures (240ºC). The
total tritium permeation into cooling loops has been estimated as 0.7 g per FW cooling loop
over the life time of ITER.
Table III.E-1 gives an overview of the Assessment Values for the tritium inventory in ITER.
Note that lower project guidelines for in-vessel (450 g) and fuel cycle (450 g) tritium
inventories are set to account for uncertainties.
TABLE III.E-1
Tritium Inventories for ITER: Assessment Values
Type of inventory
In-vessel mobilizable in-vessel (in plasma facing
components, dust, co-deposited etc.)
Fuel cycle circulating inventory
[g-T]
1000 *
Total site inventory
< 3000
700
* Not counting tritium bred in beryllium: 125 g (immobile for T<600C)
Neutrons form the fusion reactions activate surrounding materials. The basic operational
scenario for ITER decay heat and waste assessment assumes an integrated average neutron
fluence of 0.5 MWa/m2. This is the maximum fluence expected in ITER whereas the best
estimate of the ITER fluence is only 0.3 MWa/m2. For the activation calculations for nonpermanent components like the plasma facing material tungsten a maximum (poloidal
machine average) fluence of 0.13 MWa/m2 was used. Note that this corresponds to about 16
000 nominal plasma pulses of 440 s length. The majority of activation products will be bound
in solid metal structures of the in-vessel components. Smaller inventories will be found in
structures outside the vacuum vessel or circulating as corrosion products in First Wall /
Shield, and Divertor coolant streams. The most relevant source term of activation products is
activated dust originating from plasma facing material. Of all plasma facing materials
enviaged for use in ITER is it tungsten which has by far the largest radiological hazard
potential. Table III.E-2 lists the activation of tungsten based on an ITER fluence of 0.13
MWa/m2, which corresponds to the maximum lifetime of the plasma facing materials on the
divertor. The same table also shows the activation of activated corrosion product deposits in
the FW cooling loop where water is in contact with steel.
The Assessment Value for tungsten dust is 350 kg. Note that a smaller project guideline of
100 kg is set to cover uncertainties. Rough estimates for dust production indicate that the
administrative guideline of 100 kg for tungsten dust may be reached in about 500 plasma
pulses assuming a disruption frequency of 10%. Thus the administrative guideline may be
reached before the assumed replacement of the divertor. Dust diagnostics and removal
methods (to be developed) are required.
The baseline dust size specification, which has been set to envelope the size of the dust
collected from operating tokamaks and plasma disruption simulators, is given as a log normal
distribution with a "mass median diameter" of 2 micro-m.
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Table III.E-2: Activity of tungsten after 0.13 MWa/m2 ITER fluence at shutdown and
activity of activated corrsion products (ACP) in the FW/shield cooling loop after 0.5
MWa/m2 ITER fluence. Neutron flux: 0.65 MW/m2 outboard, 0.41 MW/m2 inboard.
10 mmW surface inboard
(plasma surface layer: 25 micro-m)
isotope
half life
activity
[y]
[Bq/kg]
W 187
W 185
W 185m
W 181
Re188
Re186
Re188m
W 179
Ta182
W 179m
Ta186
Ta183
Ta184
Ta182m
Ta179
Re184
Ta180
Hf183
2.72E-03
2.06E-01
3.17E-06
3.31E-01
1.94E-03
1.03E-02
3.54E-05
7.13E-05
3.14E-01
1.22E-05
2.00E-05
1.39E-02
9.92E-04
3.04E-05
1.61E+00
1.04E-01
9.22E-04
1.22E-04
5.24E+14
3.71E+13
3.64E+13
1.43E+13
6.01E+12
2.20E+12
5.79E+11
2.56E+11
1.54E+11
1.02E+11
6.34E+10
6.18E+10
4.34E+10
2.88E+10
2.74E+10
1.99E+10
1.15E+10
9.64E+09
ACP deposits (steel)
isotope
half life
[y]
Fe-55
Mn-54
Mn-56
Co-58
Co-60
Cr-51
Ni-57
Co-57
2.73E+01
8.55E-01
2.94E-04
1.94E-01
5.27E+01
7.59E-02
4.11E-03
7.44E-01
Ion and cruds in
deposit
solution activity
activity
[Bq/kg-deposit] [Bq/kg-Ion&Crud]
2.07E+12
9.61E+11
9.86E+10
3.49E+11
1.35E+12
1.19E+13
1.06E+11
3.92E+11
1.41E+11
2.39E+11
1.14E+11
4.54E+08
4.52E+10
8.85E+10
2.64E+11
4.96E+11
An activation corrosion products (ACP) assessment was carried out for a First Wall/Shield
cooling loop. The suspended cruds and dissolved ions are 1.8 g and 10.9 g respectively. The
total ACP deposit mass is ≈ 1.4 kg. To account for uncertainties especially with the divertor
cooling loops with copper water contact, the amount of activation corrosion product deposits
is assumed to be 10 kg per cooling loop in the analysis of reference events and 60 g ions and
cruds. The hazard potential of activation corrosion products inside the vacuum vessel HTS is
assumed to be 1% of the hazard potential of activation corrosion products of the FW/shield
cooling loop because the average n-flux is reduced by several orders of magnitudes inside the
vacuum vessel. This assumption is to be confirmed. The mobilization of activated corrosion
products is assumed as 1.3% of the activated corrosion products in the water spilled in offnormal situations.
Table III.E-3 gives a summary of the energy sources and typical time scales for release of
these energy sources. This table also indicates the way the project intends to control these
energy sources and prevent them from becoming a driving force in accident scenarios; more
details are in other volumes.
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TABLE III.3.6-1
Energy Inventories and Concerns for Release
Energy
Sources
Fusion
power
Plasma
Amount of
Energy
500 MW times
10 s =
5 GJ
0.7 GJ
Time Scale
for Release
Concerns
~10 s
Overheating of plasma
facing components;
In-vessel water ingress
Disruptions;
Limited evaporation
and melting of plasma
facing material
seconds to
Arcs; Quenches;
minutes
Localized magnet
melting;
Potential mechanical
damage
hours to years Heating of in-vessel
components;
depending
on concern Factor in waste
packaging
seconds to
Overheating of
hours
plasma-facing
components
H2 production
seconds to
Overpressurization of
minutes
confinement barriers
<1 s
Magnetic
50 GJ
Decay heat
for ITER
130 GJ in the first
day
330 GJ in first week
Chemical
energy
Wall (Be-steam):
500 GJ if react
Thermal
energy of
coolant
Baking - ~700 GJ
Normal operation ~300 GJ
GSSR
Control
Possibilities
• Normal operation coolant
systems
• Active Fusion Power
Shutdown System
• Passive shutdown in case of
larger disturbances
• Plasma control
• Disruption mitigation by
impurity pellets
• Rapid quench detection and
discharge system
• Grounding scheme
• Passive decay heat removal
based on radiation heat
transfer and natural
circulation
• Limiting off-normal
temperature of plasma-facing
components
• Overpressure suppression
systems limit pressures in
confinement volumes
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RADIOLOGICAL AND ENERGY SOURCE TERMS
TABLE OF CONTENTS
III.1
INTRODUCTION .........................................................................................................................................1
III.2
RADIOLOGICAL AND HAZARDOUS MATERIALS..........................................................................1
III.2.1
TRITIUM SOURCE TERM ..........................................................................................................................2
III.2.1.1
Tritium characteristics ................................................................................................................2
III.2.1.2
Tritium flows and inventories......................................................................................................3
III.2.1.3
In-vessel tritium inventories and mobilization potential...........................................................3
III.2.1.4
Hot cell, waste treatment, and tritium recovery tritium inventory ..........................................10
III.2.1.5
Coolant tritium inventories .......................................................................................................10
III.2.1.6
Tritium plant tritium inventories and mobilization potential ..................................................11
III.2.1.7
HT to HTO conversion ..............................................................................................................12
III.2.1.8
Summary of HTO/HT inventories .............................................................................................12
III.2.2
ACTIVATION PRODUCT SOURCE TERM .................................................................................................13
III.2.2.1
Activation product inventory.....................................................................................................13
III.2.2.2
Activation product mobilization................................................................................................19
III.2.2.3
Tokamak dust inventory and mobilization potential ................................................................20
III.2.2.4
Plasma-vaporized mobilization potential .................................................................................23
III.2.2.5
Activated corrosion product inventory and mobilization potential.........................................23
III.2.2.6
Oxidation-driven mobilization potential...................................................................................27
III.2.2.7
Activated gases ..........................................................................................................................28
III.2.3
CHEMICAL ............................................................................................................................................28
III.3
STORED ENERGY SOURCES ................................................................................................................29
III.3.1
III.3.2
III.3.3
III.3.4
III.3.5
III.3.5.1
III.3.5.2
III.3.5.3
III.3.5.4
III.3.5.5
III.3.6
III.4
PLASMA ENERGY SOURCES ..................................................................................................................30
MAGNET ENERGY SOURCES .................................................................................................................30
NUCLEAR ENERGY SOURCES ................................................................................................................31
THERMAL ENERGY OF THE COOLANT ..................................................................................................35
CHEMICAL ENERGY SOURCES ..............................................................................................................35
Hydrogen....................................................................................................................................35
Beryllium reaction rates............................................................................................................36
Carbon chemical reaction rates................................................................................................38
Tungsten chemical reaction rates .............................................................................................38
Ozone formation ........................................................................................................................38
SUMMARY OF ENERGY SOURCES..........................................................................................................39
REFERENCES ............................................................................................................................................41
FIGURES.....................................................................................................................................................................47
APPENDIX 1: DOSE CALCULATIONS IN SUPPORT OF GSSR...................................................................59
GSSR
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III.1 INTRODUCTION
This volume presents the radioactive sources, chemically hazardous or reactive substances,
and stored energy sources to be used or handled in ITER. These inventories and energies
consist of:
• Radioactive materials (tritium fuel, neutron activation products in structures and
in-vessel dust, neutron activated corrosion products in coolant water);
•
Potentially conventional (non-nuclear) hazardous materials (beryllium, hydrogen
gas, etc.); and
•
Stored energy sources (plasma, magnets, nuclear energy (decay heat), thermal
energy of coolants, and chemical energy).
Since ITER is an experimental machine, its design and operation must be flexible to explore
various physics conditions and to test different in-vessel components. Therefore, some
uncertainties exist in the amount of radioactive and other hazardous materials to be used in
ITER. In some cases two different estimates are provided to account for these uncertainties
and ensure that release estimates of such materials are appropriately bounded. Baseline
values are conservative estimates. These values are used in GSSR analyses unless otherwise
indicated. Best-estimate values include more realistic or optimistic assumptions about
uncertain parameters. They represent more realistic estimates of inventories than the
conservative baseline values and therefore give an indication where baseline values may
move as research progresses.
III.2 RADIOLOGICAL AND HAZARDOUS MATERIALS
Radioactive materials include tritium and neutron activation products. The estimates of
activation products are at end of life condition, i.e. at maximum activation. The hazard builds
up with time as the neutron fluence increases. For all radiological and chemically hazardous
materials, it is important to differentiate among the following:
•
The inventory of radionuclides and other hazardous material in the facility;
•
The maximum fraction of the inventory that could be mobilized within the
facility;
•
The fraction of material mobilized that could be subsequently released from the
facility to the environment.
There are various ways to characterise the hazard posed by a radioactive isotope. These
include:
GSSR
•
The activity (Bq) or equivalently the mass (kg) of the isotope; and
•
The theoretical dose (mSv) if humans are exposed in some manner.
page III-1
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III.2.1 TRITIUM SOURCE TERM
Tritium (H-3) is inherent in fusion energy reactions involving deuterium-tritium or
deuterium-deuterium. It is also produced by neutronic transmutations in beryllium.
III.2.1.1
Tritium characteristics
Tritium has a half life of 12.3 years. It emits a β−particle with a peak energy of 18.6 keV and
an average energy of 5.7 keV. The specific activity of tritium is approximately 3.6x1014 Bq/g
(9,670Ci/g).
As an isotope of hydrogen, tritium is the most mobile of the significant radioactive sources
that will be present in ITER and requires special handling and confinement procedures to
prevent its escape. Effective procedures for tritium control have been developed in various
other nuclear programs, and have been applied in large fusion tokamak programs, such as the
Tokamak Fusion Test Reactor (TFTR) and the Joint European Torus (JET).
The biological hazard of tritium strongly depends upon its chemical form. Both the gaseous
elemental and the oxide form will be present in a fusion reactor. The oxide form is readily
assimilated and distributed throughout the human body water, while the elemental form is
not.
The gaseous elemental forms of tritium (HT, DT or T2) are relatively difficult to contain and
can permeate through most materials including metals at high temperatures. Tritium from the
plasma or produced in beryllium can enter the first wall, limiter, and divertor coolant streams.
There it is converted to the oxide form whose escape via coolant leaks must be controlled.
The oxide form (HTO, DTO, or T2O) or organic forms of tritium are approximately ten
thousand times more radiotoxic than the elemental form, per gram of tritium taken into the
body. However, when natural conversion of gaseous elemental form to oxide form in the
environment is considered, the ratio of ten thousand decreases to a ratio of ten to one
hundred, per gram of tritium released from the facility [Tae91], [Bar92], [Vel00].
The effective biological half-life of oxidized tritium in the human body is about ten days,
though some may become organically bound in bone, fat, and other tissues (causing about
10% of the dose from HTO uptake), in which case it remains much longer (about 30 days to
one year). The residence time of water-borne tritium can be reduced by increasing the
normal fluid throughput.
Values of the committed effective tritium dose per unit intake for ingestion and for inhalation
are extracted from the 'International Basic Safety Standarts for Protection against Ionizing
Radiation and for the Safety of Radiation Sources' in Table III.2.1-1 for occupational
exposure and for public exposure [IBS96]. Based on the 20-mSv/a per person ITER limit on
plant personnel dose, the Derived Air Concentration (DAC) of tritium in the oxide form
(HTO) in plant air is 3.1 x 105 Bq/m3 (8.4 x 10-6 Ci/m3).
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TABLE III.2.1-1
Committed Effective Dose per Unit Intake (Sv/Bq) Versus Age (years)
Members of the Public
Nuclide
Ingestion:
HTO
organic
bound
tritium
Inhalation:
HTO
III.2.1.2
Workers
Age <1
Age 1-2
Age 2-7
Age 7-12
Age 12-17
Age >17
-
6.4E-11
1.2E-10
4.8E-11
1.2E-10
3.1E-11
7.3E-11
2.3E-11
5.7E-11
1.8E-11
4.2E-11
1.8E-11
4.2E-11
1.8E-11
4.2E-11
6.4E-11
4.8E-11
3.1E-11
2.3E-11
1.8E-11
1.8E-11
1.8E-11
Tritium flows and inventories
Tritium will be supplied from off-site sources. The total site inventory is less than 3 kg
[PSR00]. For each nominal 440 second pulse at 500 MW fusion power, 0.39 g of tritium is
burned. However the throughput of tritium per pulse maybe as high as 130 g (burn fraction
~ 0.3%).
The following sections give a brief discussion of each of the significant tritium inventories,
followed by a table summarizing all the inventories.
III.2.1.3
In-vessel tritium inventories and mobilization potential
The tritium in the plasma and the divertor gas target does not contribute significantly to the
inventory that could be mobilized in case of an accident since it is less than a gram. All this
tritium is mobilizable during in-vessel events.
Tritium in the plasma-facing components is the most important inventory and the one that has
the largest uncertainty. The three major mechanisms for tritium to be in (or on) plasmafacing components are:
•
For metals, the dominant mechanism is implantation. Metals have associated
diffusion of tritium into the bulk material and trapping at impurities and defects
including those produced by neutron irradiation;
•
For beryllium, an additional mechanism is tritium breeding by neutron
transmutations; and
•
For carbon, the mechanism is tritium co-depositing as hydrogen-carbon layers.
For beryllium a similar process is co-implantation where tritium is trapped by
Be/BeO layers.
Other possible mechanisms have been studied, but they are not significant. For example,
tritium retention in disruption-produced carbon dust for tokamaks (ITER in particular) should
not significantly increase the tritium inventory [Cau94]. Also, diffusive movement of tritium
into graphite is not significant compared to co-deposition. A monolayer of tritium on the
beryllium and carbon surface (~900 m2) results in a tritium inventory of 0.5 g.
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page III-3
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Implantation
Implantation is generally the dominant mechanism for tritium uptake in plasma-facing
materials. Figure III.2.1-1 illustrates these processes. Tritium particles from the plasma with
energies varying from a few eV to a few hundred eV are implanted and can build up high
concentrations in the near surface region of plasma-facing materials. Because of this large
concentration of mobile tritium near the inner (plasma) wall surface, a concentration gradient
is established, causing tritium to diffuse into the bulk and eventually to the outer wall surface
where it can enter the coolant. Temperature gradients resulting from heat flux, neutron
interactions with the material, and surface erosion can also influence tritium retention and
transport properties [Fed96]. Tritium in the bulk material can be retained in solid solution
(so-called "mobile") or in traps.
Experimental evidence [Che96,Wam92,Gus96b,And92] supports the understanding that for
Be, saturation effects limit the mobile atom concentration at the implantation surface and thus
inhibit diffusion of tritium into the bulk beryllium. The apparent cause is surface texturing
and the development of interconnected porosity by migration of bubbles formed in the
implantation layer.
To estimate inventories and permeation rates in plasma-facing structures requires knowledge
of the intensity and time history of impinging particle fluxes, the materials, dimensions,
temperature history and hydrogen transport properties (diffusivity, solubility, recombination
coefficient, trap density) in the materials. These are all required to perform transient
calculations of inventory buildup using a code such as TMAP5 (see Volume XI).
The uncertainties associated with the implantation processes and the flexibility to use
different in-vessel materials results in uncertainty in the in-vessel inventory. The following
factors need consideration:
•
Plasma physics - The primary inputs are implantation rates, particle energies and
erosion rates. The erosion rates are largely uncertain, particularly in the divertor
area;
•
Transport processes - The movement of tritium at the surface and in the bulk is
principally governed by surface recombination (including saturation effects where
applicable) and diffusion. However, trapping at damage sites, oxide layers,
surface texturing, grain boundaries as well as other properties can affect transport.
For example, there is about an order of magnitude uncertainty in the trapping
effects and how these evolve with neutron irradiation. Traps in the beryllium can
exist because of material defects from processing the material or because of
neutron irradiation damage. The trap density in un-irradiated beryllium is
expected to be of the order of 0.001 atom fraction (0.1%) with an apparent
trapping energy of 1.8 eV, but traps could increase in neutron irradiated material.
•
Materials properties - Surface and bulk properties of the materials as well as
substrate materials are used in the transport models.
The tritium inventory in the beryllium first-wall due to implantation, diffusion, trapping, and
breeding after 12000 pulses will be of the order of 20 g [Fed00]. Tritium retention in n-
GSSR
page III-4
ITER
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induced traps could increase this inventory somewhat (e.g., up to ~90 g for a trap density of
0.1 at. % and a trap energy of 1.4 eV). It is expected that a conservative estimate of tritium
implantation in Be is within the range of 20 to 100 g. A sensitivity study yielded tritium
inventories up to 250 g for a trap density of 1 at. %.
Breeding in plasma facing components
Breeding is the generation of tritium in plasma facing components by nuclear transmutation
reactions. This is relevant only for beryllium (and to a lesser extent boron), not for carbon
nor for tungsten. The relevant reactions in beryllium are as follows:
9Be
+ n -> 7Li +3H
9Be
+ n -> 6He + 4He
6He -> 6Li + β (T =0.8 s)
1/2
6Li + n -> 4He +3H
The first of these reactions has a threshold at about 12 MeV. The second has a threshold at
600 keV and a strong resonance at 3 MeV. Therefore, tritium breeding depends strongly on
the neutron energy spectrum. The first reaction is dominant for plasma-facing components.
The analysis involves activation calculations and determination if any significant fraction of
the bred tritium escapes during normal operation. The 115 g tritium bred in the beryllium
after irradiation to an average fluence of 0.5 MWa/m2 is retained at operational temperatures
for full-density beryllium provided swelling is not excessive.
Co-deposition on plasma facing components
Co-deposition will be the dominant process for tritium uptake by carbon if this is used in the
machine. Thick carbon deposits with high levels of deuterium have been observed on
graphite tiles exposed to deuterium plasma discharges in TFTR, DIII-D, JET, TEXTOR, and
other tokamaks. Figure III.2.1-2 illustrates the basic mechanism.
Estimating the co-deposited tritium inventory requires knowing the particle fluxes and target
design information, the sputtering yield, co-deposition rates (for example, hydrogen/carbon
ratios), and the ability to perform transient modeling of the full sputtering and re-deposition
processes. The H/C ratio depends on the deposition conditions and surface temperature. At
room temperature it is as high as 1.4 [Jac98]. With respect to temperature dependence a wide
scattering in hydrogen concentration exist for different deposition condition. Recent
measurements reported a D/C ratio of 0.2 at 700ºC [Bal99].
The following experimental findings have been made in today's tokamaks:
(a) The tritium retention at the "DTE1" tritium campaign at JET was initially 40% during
tritium fuelling by gas puffing. This large rate can be partly understood by isotopic
exchange with deuterium saturated walls. The longer term retention during fuelling of
only deuterium was about 20%. However, compared to ITER, results from present-day
machines (with several second pulse length) may be affected dispropotionally by transient
effects such as impurity generation and transport during plasma start-up and shutdown.
First-wall erosion/re-deposition can be markedly different during start up and shutdown
period (limiter configuration) than during the flat-top period, where an equilibrium
GSSR
page III-5
ITER
(b)
(c)
(d)
(e)
G 84 RI 3 01-07-13 R1.0
divertor plasma is established. For ITER, the 400 s pulse length makes the startup/shutdown effects less significant. Another significant difference between JET and
ITER is that the ITER FW wall will be covered by solid beryllium and thus there is no
source of carbon from the FW for ITER.
In divertor tokamaks, the main chamber wall is an area of net erosion, while both erosion
and deposition occur in the divertor. Note that here is a further significant difference
between today's carbon tokamaks and ITER which will operate with a beryllium FW and
hence have a smaller source for carbon co-deposition as today's tokamaks.
Intense co-deposition of carbon and deuterium is found in many tokamaks in regions
which are shaded from the plasma, and in gaps between tiles, but are near carbon surfaces
receiving a high ion flux. Since ions cannot reach these shaded surfaces, this carbon
deposition must be due to neutral carbon atoms or molecules/radicals resulting from
dissociation of hydrocarbons released from carbon-surfaces.
Chemical erosion processes are known to be particularly efficient at releasing carbon
from surfaces. In particular, hydrogenic species, as ions or charge exchange neutrals,
readily form bonds with C and may be released from surfaces as hydrocarbon radicals
(CH, CH3, C2 H, C2 H3, etc) or methane (CH4). Several of these radicals have large
sticking coefficients and re-deposit on nearby cool surfaces forming co-deposited,
amorphous hydrocarbon films. Others have rather small sticking coefficients and may
travel considerable distances before sticking. These radicals may be capable of a-C:H
film formation in areas of the vessel well hidden from the plasma such as at the rear of
plasma facing components, in ports and so on.
Chemical erosion is a highly temperature sensitive effect and relatively small changes in
temperature can give rise to order of magnitude changes in erosion rates. Erosion
increases with temperature, suggesting that operation with cool surfaces (consistent with
minimizing many physical erosion processes) is beneficial. However, a-C:H film
formation is strongly enhanced on cooler surfaces and a balance, minimizing film and
flake formation may be difficult to find.
Calculations of erosion/co-deposition were made for the ITER design with emphasis on the
quantification of erosion during the flat-top burn-phase. A cross-sectional view of ITER is
given in Fig. III.2.1-3, and the main PFC operation parameters used for the analyses are
summarised in table III.2.1-2. Beryllium is the material for the first-wall, whereas tungsten is
the plasma-facing material for the divertor except near the strike plates. There, carbon (CFC)
is chosen. For the calculations, a plasma solution provided by the code B2-EIRENE [Cos97]
was used for a nominal ITER case with no impurity seeding, a gas puffing rate of
110 Pa·m3/s, and a separatrix density upstream of 3·1019 m–3. For the first-wall and baffle
region, erosion occurs due to physical sputtering by charge-exchange (CX) deuterium-tritium
(DT) neutrals from plasma recycling and gas puffing and by DT and impurity (e.g., He, C)
ions which are accelerated in the potential sheath above the sputtering threshold. Fast neutrals
from core plasma recombination also contributes to the formation of the high-energy tail of
the neutral distribution. To evaluate the neutral fluxes, the EIRENE code was run stand-alone
with modifications ensuring accurate computation of the energy spectra of the neutrals
impinging upon the surface, and the plasma inside the separatrix was assumed to have the
specified ITER profiles of temperature and density to provide the correct high-energy tails of
CX neutrals. The CX fluxes and the shapes of the spectra vary greatly around a poloidal
cross-section (see Fig. III.2.1-4). This has a strong effect on the wall erosion by sputtering
and hydrogen isotope implantation into the first-wall.
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TABLE III.2.1-2
Operation parameters used for the design of the PFCs of ITER.
Material
Area (m2)
Normal operation
Peak surface heat flux (MW/m2)
Peak particle flux (1023/m2s)
Divertor Target
Baffle/Dome
CFC(†) and W
55 and 60 (liner)
W
50/30
First-Wall (Startup Limiter)
Be
680 (~10)
10
~10
<3
<0.1
0.5 (~8 for ~100s)
0.01 (<0.1)
† Near vertical target strike-points ; tungsten elsewhere.
The total charge exchange flux Γ = ∫ Γ( E )de and the mean energy, Emean = ∫ Γ( E ) Ede Γ are shown in
Fig. III.2.1-5. The largest fluxes and the lowest Emean occur just at the top of the chamber due
to the localised puffing and recycling of gas there, and above the divertor due to the strong
recycling there. The CX spectra were then integrated over the angle of incidence and the
energy with the sputtering yield to calculate the net erosion by neutral particles. Sputtering
erosion by ions was found using the fluxes at the grid edge, temperatures scaled to the wall
with 3 cm e-folding length.
Divertor erosion/co-deposition including re-deposition in the divertor region was calculated
more thoroughly using REDAP/WBC codes ([Bro83], [Bro90]). Chemical sputtering of
carbon by DT ions, atoms, and molecules, more important here than physical sputtering, is
calculated using the yields described in [Fed99]. To calculate tritium co-deposition it is
assumed that (i) all material not re-deposited at the wall goes to the divertor (since ionisation
of wall-sputtered material in the SOL and subsequent transport along the field lines is
expected), (ii) the re-deposited material grows locally (does not erode further), (iii) the DT
flux to the growing surface is large and energetic enough to reach saturation at the
temperature-dependent saturated value for C and Be-O.
Erosion of the first-wall during normal operation may contribute to co-deposition in the
divertor region. Fig. III.2.1-6 shows the erosion rate for neutrals and ions along the first-wall
for Be and W. The beryllium peak erosion rate of ~0.1 nm/s (~0.3 cm/burn-yr) is acceptable
for the low duty-factor operation of ITER. Tungsten erosion, shown here only for
comparison, is between one and two orders of magnitude lower.
The calculations yield a tritium co-deposition rate of 2-5 g-T/400 s pulse. The present
analyses are at most reliable for indicating trends, not firm quantitative predictions. Material
mixing effects resulting from simultaneous erosion/re-deposition of different materials add
further uncertainties.
The integrated FW beryllium erosion adds up to 16 g per 400 s plasma pulses. The resulting
Be co-deposition rate, assuming that half of this beryllium builds up in lower temperature
areas (≤200˚C at the bottom of the divertor target and private region) is estimated to be in a
range of 0.1-0.4 g-T/400 s pulse.
Without removal of this tritium, kilograms of tritium could be retained in the co-deposited
layer of ITER. Therefore, active and effective methods to remove the co-deposited layers are
needed in ITER even if only a small fraction of the surface would be covered by carbon.
Carbon in the divertor can form coatings on other surfaces. Note that also co-deposition in Be
alone would require regular removal of tritium from in-vessel components although much
less frequent. The only methods effective for removing tritium known so far involve (1)
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oxidation of the co-deposited layers (e.g., thermo-oxidative erosion > 250˚C, or oxygen
plasma discharges) or (2) physical removal [Fed00]. For carbon co-deposited films, oxidation
rates strongly depend on the microstructure of the layers. Mixing of materials shows that
higher temperatures might be required for erosion of the films and release of the retained
hydrogenic-species. Therefore, although 240˚C may remove soft films, due to the variability
of film properties a baking capability at temperatures significantly greater than 300˚C would
be required for more complete removal. However, even at this temperature the efficiency of
removing the co-deposited films remains questionable, in particular, with regard to the effects
of mixed-materials and residual layers that could not be eroded and would become thicker
and thicker over several baking cycles. Moreover, frequent use of oxygen baking raises issues
of collateral damage on other in-vessel components, as well as recovery time for normal
plasma operation.
Co-deposition of tritium with tungsten is not observed [Gus96a], [Dol95].
Integrated in-vessel inventory
Since most of the mechanisms of in-vessel tritium source term build-up carries significant
uncertainties, it is difficult to make an accurate prediction of the expected source term. In
fact studying the source term in ITER should be viewed as one of the important missions of
the project. To cope with the uncertainties and the fact that a large amount of tritium may
build up inside the vacuum vessel during ITER operation, the approach is to set a guideline
for the maximum mobilizable amount of tritium inside the VV, to monitor the tritium source
term build-up during ITER operation and to remove tritium once the guideline amount is
reached. Methods for tritium removal need further development before they can be
employed in ITER.
The ITER project has set a guideline of 450 g tritium for the maximum mobilizable amounts
of tritium inside the vacuum vessel. The cryogenic pumps have a maximum tritium inventory
of 120 g. Thus the maximum amount of tritium remaining inside the VV has to be limited to
less than 330 g by regular removal. Further design studies have to show their feasibility.
These guidelines were set to push the design into a direction of minimizing tritium
inventories. Should further studies show that the guidelines need to be increased to allow
practical operation of ITER this would not invalidate the ITER safety approach since wide
margins are built into the confinement of source terms such that releases are significantly
below environmental release limits (see Volume IV and VII).
The Assessment Value used in Volumes VII and VIII is 1 kg of in-vessel tritium.
Tritium bred inside the beryllium of the first wall (up to 115 g) is considered to be immobile
in accidents because of the limited temperatures in ITER accidents as discussed below. The
same is true for the majority of implanted tritium inside beryllium.
Tritium co-deposition remains the major tritium retention mechanism for ITER, even if the
use of carbon is limited to the divertor strike plates. Retention by other mechanisms such as
implantation and surface adsorption, which may be significant for small short-pulse
machines, is expected to rapidly reach saturation in ITER. The rate of tritium uptake in the
co-deposited layers is plotted in Figure III.2.1-7 [Fed00], for two indicative T co-deposition
rates of 1 and 10 g/ 400 s, and is compared to other tritium retention mechanisms in the
metal-clad PFCs. The guideline for 330 g in-vessel tritium (not counting tritium in the
cryopumps) would be reached in 30 to 300 pulses. With 3000 shots per year estimated for a
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typical operation scenario [DRG-1], 10 to 100 in-vessel tritium cleaning campaigns have to
be envisaged each year. Thus, in the ITER design, it is recognized that, as long as carbon is
maintained in the design, techniques to control and/or mitigate the accumulation of tritium
retained in the carbon-containing co-deposited films are needed. Otherwise the in-vessel
tritium inventory hold-up could rapidly reach and exceed precautionary operating guidelines
set by safety considerations and/or exhaust the limited supply of tritium available for
operation.
Mobilization of in-vessel inventory
Mobilization of implanted tritium occurs in two ways. One is diffusion, which is slow. The
other is a non-diffusive burst release e.g. from the cryogenic pumps in case of air or steam
ingress into the vacuum vessel.
The mobilization of the diffusive inventory depends on time, temperature and the nature of
traps that are present. The time constant for inventory release is determined by the effective
value of hydrogen diffusivity in the material and by its thickness. Traps can make the
effective diffusivity very small, resulting in a long time needed to release the inventory.
Higher temperatures increase the release rate exponentially. For the primary first wall,
implanted tritium never diffuses deep into the material at operating temperatures. During a
thermal excursion event, its mobility is increased and some of the implanted tritium and
deuterium diffuse into the bulk of the material and become trapped. At temperatures below
500°C for beryllium, trapped tritium is not mobilized in any significant amount in the time
scale of interest for the event [Bal91, And97a]. From the mobilization point of view, the
neutron bred tritium will behave as trapped tritium.
The potential for "burst" mobilization from beryllium depends on temperature and microstructure, which itself depends on temperature, time at temperature, neutron
damage/swelling, and porosity [And96a, And97a]. Normally, burst releases are not seen
below 600°C and occur following an incubation time of several hours [Bal91]. Such
conditions are not reached in ITER off-normal situations.
However at high temperature increase rates (20-100 K/s) the temperature of tritium burst
release from beryllium irradiated with high neutron fluences (E>0.5 MeV) up to 10 22 cm-2
decreases down to 450ºC [And98]. This may apply to small fraction of the FW (~1 m2 or
0.2%) in case a VDE coincides with an accident (see e.g. [DDD1.7]).
The mobilization of carbon co-deposited tritium is very slow in vacuum or pure nitrogen but
is much faster when oxygen (or air), hydrogen, or water are present, primarily due to
oxidation and isotope exchange [Cau90], [Dav98], [Haa98], [Wan97], [Rom01]. Typically,
detritiation of carbon co-deposited tritium occurs in two stages, an initial fast mobilization
(minutes) followed by slow (hours-days) removal. Tritium is mobilized in the oxide form
when exposed to air or steam. At room temperature co-deposited layers have been measured
as being relatively stable. In [Wan97] a mobilization rate of 8.5*10-4/day has been reported.
Based on these data, a simplified yet conservative approach to mobilization has been
established. Because most of the in-vessel inventory is co-deposited tritium, mobilization of
this inventory is treated as if it were all co-deposited tritium using a simple temperature
criterion. The results are shown in Table III.2.1-3 and Figure III.2.1-8. Full mobilization is
conservatively assumed for temperatures above 200ºC and a linear behavior of mobilization
is assumed between 0 and 200ºC. To capture the time dependence of tritium mobilization, in
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events where it is conservative to delay the mobilization (thus releasing the tritium after
activation of the vacuum vessel suppression system, see Volume VII and Volume VIII), half
of the tritium is mobilized immediately and the remaining 50% is spread over six hours.
Figure III.2.1-9 shows some experimental data on the kinetics of tritium mobilization in
atmospheric air and steam measured with samples collected in the T-10 tokamak [Rom01].
At 300ºC the mobilization is fast and most tritium is released with the first hour of air
exposure. At 200ºC significant mobilization occurs within the time frame of a few hours. In
both cases the GSSR assumptions on mobilization are a conservative representation of the
data.
Table III.2.1-3: Generic in-vessel tritium mobilisation fractions
Inventory type
Mobilisation Mechanism
Mobilisation fraction
VV Cryopump inventory
Evaporation
100% instantaneous
In-vessel tritium
Primarily co-deposited layer
T [ºC]/200 for T<200ºC (d)
- air ingress (a)
removal (b)
100% for T> 200ºC
In-vessel tritium
Primarily co-deposited layer
T [ºC]/200 for T<200ºC (d)
- water ingress (a)
removal (b)
100% for T>200ºC
Hourly tritium release from inOutgasing
0.016 g/hr (c)
vessel after initial VV cleanup
(a) References of experimental studies of mobilisation of tritium in co-deposited layers:
[Cau90], [Dav98], [Haa98], [Wan97], [Rom01].
(b) When it is conservative for the in-vessel co-deposited tritium mobilisation to be delayed, half is
mobilised immediately and half spread over 6 hours.
(c) see also Vol. IV
(d) this means that the mobilization fraction increases linearly from 0 to 1 between 0ºC and 200ºC, e.g.
mobilization at 0ºC is 0%, at 100ºC is 50% and at 200ºC is 100%.
III.2.1.4
Hot cell, waste treatment, and tritium recovery tritium inventory
Before removal from the vacuum vessel in-vessel components are cleaned as much as
possible in-situ. They are then removed, transported to the hot cells, where most of the
tritium is recovered, the divertor cassettes are refurbished, and they are eventually re-used.
This reduces waste production from the divertor and also reduces the total on-site tritium
inventory. The hot cell and waste treatment facility have a tritium inventory guideline of
200 g. The tritium recovery and waste storage facility have a tritium inventory guideline of
50 g.
III.2.1.5
Coolant tritium inventories
All in-vessel components are cooled by water. Diffusion of implanted tritium into the
primary first wall, baffle, and divertor during operation and baking will result in tritium
contamination of the cooling water of those components. The analysis steps are the
implantation inventory analysis, the permeation through the material (based on diffusivity),
and finally the tritium balance in the water.
Permeation in carbon-based materials such as carbon fiber composites (CFC) is strongly
inhibited by the low permeability of bulk graphite. Some permeation may result from
alternate pathways such as grain surface diffusion or flow in open porosity but the rate is not
significant.
Permeation is dominated by the phases of baking when the in-vessel components are at
elevated temperatures (240ºC). The total tritium permeation into cooling loops has been
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estimated as 0.7 g per FW cooling loop [Ise01] by the end of plasma operation with an
accumulated fluence of 0.5 MWa/m2 in 10 years. The baking time has been estimated as
three month per year. Figure III.2.1-10 shows the resulting build up of tritium inside the three
FW cooling loops which is driven by the baking periods only. Table III.2.1-4 lists the
resulting guidelines for the maximum tritium concentration in primary coolant water of ITER
cooling loops. Note that the tritium concentration in these cooling loops would build up
slowly and sufficient time for upgrading water detritiation systems is available in the
unforeseen event that some of these inventories should be exceeded in the actual operation of
ITER.
Table III.2.1-4: Tritium concentration in primary cooling water
Vacuum vessel cooling system
In-vessel components cooling system
III.2.1.6
< 0.0001 g T/m3 water (0.001 Ci/kg)
< 0.005 g T/m3 water (0.05 Ci/kg)
Tritium plant tritium inventories and mobilization potential
The guideline for the maximum tritium inventory in the tritium plant for initial operation is
450 g and consists of a number of individual subsystems, whose tritium inventory is
generally below 100 g. The inventory in some systems are shown in Table III.2.1-5. The fuel
cycle inventory is listed in such a way that it represents a typical snap shot during ITER
operation. The amount of inventory may vary from system to system but the overall
circulating inventory will remain the same. During maintenance or decommissioning periods
some of the in-vessel inventory of Table III.2.1-5 may be transferred to the hot cell and waste
treatment. Another significant transfer of tritium inventory will occur during removal of invessel co-deposited tritium. Here up to 330 g of in-vessel inventory may be transferred to the
tritium plant and long term storage.
Of the in-vessel tritium 5% is conservatively assumed as hydrocarbons which is in line with
the design specification for the fuel cycle that <2.5% of the mol-fraction of tritated molecules
in the plasma exhaust is assumed to be in the form of hydrocarbons [DRG-1]. A recent
detailed investigation of the exhaust of JET during the tritium operation is documented in
[Per99]. Based on the latter reference:
~ 0.25% mol-fraction CQx were found in the regular exhaust from JET.
~ 0.5% mol-fraction CQx were found from the divertor pump exhaust
~ 1.4% mol-fraction CQx were found during warm up of LN panels.
Thus 2.5% seems a conservative design specification which is also used for safety studies.
2.5% mol-fraction corresponds to 5% tritium in the form of CQx if the majority of CQx is in
the form of methane CQ4 as measured in JET.
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Table III.2.1-5: Fuel cycle tritium inventory
Type of inventory
Isotope Separation System
Circulating tritium one plasma shot (effective burn
time 450 seconds),
(maximum cryopump inventory 120 g-Ta)
Pellet fuelling
Gas fuelling
Other
Total of fuel cycle tritium inventory
a
b
[g-T]
220
130
45
10
45
450b
5% of the in-vessel tritium inventory should be assumed as hydrocarbons molecules C(DT)
Assessment Value for Vol. VII and VIII is 700 g tritium to account for uncertainties
There are provisions for secure long-term storage of up to several months supply (900 g-T).
This tritium will be stored in hydride beds in a secure vault. The tritium inventory of each
hydride bed is limited to 100 g. Note that the Assessment Value for the fuel cycle tritium
inventory is set to 700 g to account for uncertainties.
III.2.1.7
HT to HTO conversion
In JET [JET95] the following assumptions have been applied for conversion of HT to HTO:
although the initial conversion in the first 1000 s is 0.2 %, a value of 1% is used for the first
few minutes after mobilization. The conversion is assumed to increase to 10% during the
first 12 hours. However in ITER most of the tritium in the vessel is expected from codeposited carbon layers. Those have been observed to release tritium in the form of HTO
[Cau90]. Conversion of DT from cryopumps may be slow, however the presence of hot
surfaces and radiation fields will increase the rate. Presence of steam will allow for isotopic
exchange. Therefore 100% conversion of HT to HTO is assumed prior to release from the
VV.
In the fuel cycle most events involve release of DT. At a rate of 10-9/s [Hou95], conversion
is limited to 0.01% during the first day. Here the JET assumption seem plausible for ITER
too: 100% of DT is assumed to oxidize for puff releases where flammable limits are exceeded
to address the potential for co-incident fire. For slow releases where flammable limits are not
expected, 1% conversion is assumed.
In the hot cell most outgasing is in the form of HTO. Thus all tritium sources should be
assumed as HTO.
III.2.1.8
Summary of HTO/HT inventories
Table III.2.1-6 gives an overview of the project guidelines for tritium inventories for ITER.
The total site inventory will be limited to 3 kg.
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TABLE III.2.1-6
Project Guidelines for Tritium Inventories in ITER
Type of inventory
In-vessel
Mobilizable in-vessel (in plasma facing
components, dust, co-deposited etc.)
Cryopumps open to VV
Subtotal in-vessel
Fuel cycle
Isotope Separation System
Circulating tritium one plasma shot (effective burn
time 450 seconds)
Pellet fuelling
Gas fuelling
Other
Sum of circulating inventory fuel cycle
Subtotal of fuel cycle tritium inventory
[g-T]
330 *
a
120 a
450 (project administrative guideline)b
220
130
45
10
45
450 (project administrative guideline)c
450
Long term storage
Hot cell and waste treatment
Tritium recovery and waste storage
Subtotal, hot cell and waste treatment
2 * 450
200
50
250
* Not counting tritium bred in beryllium: 125 g (immobile for T<600C)
a
5% of the in-vessel tritium inventory should be assumed as carbonised molecules C(DT)
b
Assessment Value :1000 g in-vessel tritium to account for uncertainties
c
Assessment Value: 700 g circulating fuel cycle tritium inventory to account for
uncertainties
III.2.2 ACTIVATION PRODUCT SOURCE TERM
III.2.2.1
Activation product inventory
Neutrons arise from the deuterium-tritium reaction: D + T ---> He + n. Because neutrons
have no charge, they leave the plasma and activate surrounding materials. The first step in
assessing activation hazards is calculating the activation inventory, as follows: neutron
source from plasma (neutron flux); time (neutron fluence, operating scenario, that is, the rate
that the fluence is accumulated and the time for decay between pulses); geometrical model of
materials being irradiated; neutron transport; and activation reactions (tracking all the nuclear
reactions, accumulation of isotopes, depletion of isotopes).
For ITER safety analysis the basic 1D activation calculations are reported in [Cep00] and
were checked by the Nuclear Analysis Group of the ITER Joint Central Team. For most
purposes discussed in GSSR 1-D activation calculations are sufficient to yield conservative
results. Exceptions are dose maps outside the cryostat for which more involved 3-D
neutronics transport calculations are needed. The specifications for the safety-relevant basic
1D activation calculations are listed in [Bar00]. Table III.2.2-1 shows the basic operational
scenario SA1 for ITER accident analysis and waste assessment with an integrated average
fluence of 0.5 MWa/m2. This corresponds to the maximum fluence expected in ITER
whereas the best estimate of the ITER fluence is only 0.3 MWa/m2. Note that this maximum
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fluence is also accumulated with 10 years whereas it is expected that the best estimate fluence
of 0.3 MWa/m2 is only reached within 20 years. Therefor it is judged that the SA1
operational scenario is a conservative envelope for the operation of ITER. For the activation
calculations for non-permanent components like the plasma facing material tungsten a
maximum (poloidal machine average) fluence of 0.13 MWa/m2 was used. Note that this
corresponds to about 16 000 nominal plasma pulses of 440 s length. For dose-rate
calculations which are used to estimate occupational doses a more realistic scenario called MDRG1 is used (for details see GSSR, Vol. VI.3.1).
Table III.2.2-2 lists the radial build of the 1D model for ITER (cylindrical midplane). The
following steps were followed to obtain activation results:
Sn transport calculations
• The neutron and gamma flux distribution have been obtained by means of the BonamiNitawl-XSDNRPM Sn coupled n-γ one dimensional discrete ordinates transport calculation
sequence from SCALE-4.4 computer code system [NUR98]. The Vitamin-ENEA-J (175n42γ groups) transport library, based on FENDL/E-2 data library is used for transport
calculation.
• The nuclear heating related to the different materials for all the radial zones have been
obtained from the transport calculation sequence using Kerma factors. They have been
obtained by processing nuclear basic data from the European Fusion File EFF-2.4 with
modules (e.g. HEATR and GAMINR, applying the energy balance method) of the NJOY
94.105, via GROUPR module, into the standard 175n-42γ VITAMIN-J energy structure.
Activation calculations
• The EASY-99 (FISPACT) activation package [EAS99] has been used to obtain the
activation characteristics of all the materials/zones.
The majority of activation products will be bound in solid metal structures of the in-vessel
components. Smaller inventories will be found in structures outside the vacuum vessel or
circulating as corrosion products in First Wall / Shield, and Divertor coolant streams. Some
products could also be generated in air inside the tokamak building by neutrons streaming
through penetrations.
Isotopes contributing significantly to total activity, contact dose, decay heat, or waste
clearance level (at least 1%) are listed in Table III.2.2-3 for the W divertor dome (the
tungsten region with highest activation), Cu FW, and SS316 FW. The errors are based on the
uncertainties in the nuclear libraries input data for activation calculation. The total
uncertainty for the activation of tungsten, copper and stainless steel is in the range of 3-9%.
The activation of water produces a few key isotopes. Tritium arises from (n,γ) capture of
deuterium normally in water as well as by any lithium addition in the water. C-14 arises
from oxygen isotopes in water via the reactions 17O(n,α)14C and 18O(n,n'α)14C. Two nitrogen
isotopes are important for the short term activation (life time ~ seconds). N-16 arises also
from oxygen isotopes via the reaction 16 O(n,p)16 N. N-17 is produced by the reaction
17
O(n,p)17N. Water activation is summarized in Table III.2.2-3a. The calculation assumed
stangant water for the activation calculation. The total activation in the water was divided by
the volume of the cooling loop to obtain the specific activity.
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The various activation products will produce an intense radiation field inside the cryostat and
vacuum vessel resulting in the requirement for remote maintenance for systems and
components inside these structures. The inventory of activation products in the ITER
structures will increase with time and with accumulation of fluence. In all safety assessments
end of lifetime conditions are conservatively assumed.
The activated structural materials will require carefully planned management activities
following completion of the planned experimental program; see Volume V. However, the 1D
assessment is not adequate to predict the activation of materials located behind the magnets,
because neutron streaming through gaps have to be considered in this case.
Table III.2.2-1: SA1 operation scenario
average n-flux FW
0.57 MW/m2
total operation time ITER minus last
campaign
10 a
fraction of fluence first 5 years
0.2
fraction of fluence second 5 years
0.8
fluence first 10 years
0.486 MWa/m2
last experimental campaign
maximum duty cycle
full operation 24 hours per day at maximum
duty cycle: 1 hour on, 2.33 hours off
fluence last campaign
total fluence ITER
GSSR
0.3
1 month
0.014 MWa/m2
0.5 MWa/m2
page III-15
ITER
G 84 RI 3 01-07-13 R1.0
Table III.2.2-2. Radial build of ITER - 1D Model
Zone
#
1
2
Zone
Identifier
VCTR
ITPL
3
CS
4
-
OTPL
5
6
TFCCRI
INSO
7
TFCW(5)
8
9
10
11
12
TFCW(4)
TFCW(3)
TFCW(2)
TFCW(1)
INSI
13
TFCCFI
15
THMLSI
16
17
18
19
VVRWI
VVSHLD(1)
VVSHLD(2)
VVFWI
20
21
22
23
24
25
26
27
28
29
30
31
32
33
BLBKI
BLKTI(15)
BLKTI(14)
BLKTI(13)
BLKTI(12)
BLKTI(11)
BLKTI(10)
BLKTI(9)
BLKTI(8)
BLKTI(7)
BLKTI(6)
BLKTI(5)
BLKTI(4)
BLKTI(3)
34
35
36
BLKTI(2)
BLKTI(1)
FWCUI
37
FWBEI
GSSR
Zone Name
Center Void
Inner Tie-Plates
Gap between Tie-plate and
CS
Central Solenoid
Gap between Tie-plate and
CS
Outer Tie-Plates
Gap between Tie-plate and
TF
TF Coil Case Rear Wall (In)
Insulator (Filler+Ground
Insulat.)
TFCW Inboard (Zone 5)
TFCW Inboard (Zone 4)
TFCW Inboard (Zone 3)
TFCW Inboard (Zone 2)
TFCW Inboard (Zone 1)
Insulator (Filler+Ground
Ins.)
TF Coil Case Front Wall
(In)
Gap
Thermal Shield
VV Rear Wall (Inboard)
VV Shield i.b.(Zone 1)
VV Shield i.b.(Zone 2)
VV front wall i.b.
gap
back of blanket i.b.
Blanket i.b. (Zone 15)
Blanket i.b. (Zone 14)
Blanket i.b. (Zone 13)
Blanket i.b. (Zone 12)
Blanket i.b. (Zone 11)
Blanket i.b. (Zone 10)
Blanket i.b. (Zone 9)
Blanket i.b. (Zone 8)
Blanket i.b. (Zone 7)
Blanket i.b. (Zone 6)
Blanket i.b. (Zone 5)
Blanket i.b. (Zone 4)
Blanket i.b. (Zone 3)
Gap (real gap: <0.5 mm; 9
mm simulate toroidal voids)
Blanket i.b. (Zone 2)
Blanket i.b. (Zone 1)
FW Cu heat sink i.b.
FW Be i.b.
plasma
Zone Materials
void
316LN (73%); Void (27%)
Void
Rin
(mm)
0
1197
1307
∆R
(mm)
1197
110
10
Incoloy (61%): Epoxy+glass (10%):
Cable (29%)
Void
1317
782
2099
10
316LN (73%): Void (27%)
void
2109
2169
60
30
316LN
Glass+Epoxy
2199
2424
225
12
316LN (41%): Conductor (44%);
Glass+Epoxy (15%)
Same as above
Same as above
Same as above
Same as above
Glass+Epoxy
2436
112
2548
2658
2768
2878
2988
110
110
110
110
12
316LN
3000
75
316LN (5%):304L (48%);Ag(0.018%)
3075
3180
105
55
3235
3295
3404
3513
3573
3598
3648
3682
3706
3726
3750
3778
3802
3832
3856
3884
3908
3927
3951
3971
60
109
109
60
25
50
34
24
20
24
28
24
30
24
28
24
19
24
20
9
3980
4006
4017
26
11
22
4039
4049
10
4372
316L(N)-IG
30467(60%), H2O(40%)
30467(60%), H2O(40%)
316L(N)-IG
316L(N)-IG(70%), void(30%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(62%), H2O(38%)
316L(N)-IG
316L(N)-IG(62%), H2O(38%)
316L(N)-IG
316L(N)-IG(65%), H2O(35%)
316L(N)-IG
DSCu(84%), 316L(N)-IG(1%),
H2O(12%), void (3%)
Be
page III-16
ITER
G 84 RI 3 01-07-13 R1.0
38
39
FWBEO
FWCUO
FW Be o.b.
FW Cu heat sink o.b.
40
41
BLKTO(1)
BLKTO(2)
42
43
44
45
46
47
48
49
50
51
52
53
54
55
BLKTO(3)
BLKTO(4)
BLKTO(5)
BLKTO(6)
BLKTO(7)
BLKTO(8)
BLKTO(9)
BLKTO(10)
BLKTO(11)
BLKTO(12)
BLKTO(13)
BLKTO(14)
BLKTO(15)
BLBKO
56
57
VVFWO
VVSHDO(1)
Blanket o.b. (Zone 1)
Blanket o.b. (Zone 2)
Gap (real gap: <0.5 mm; 9
mm simulate toroidal voids)
Blanket o.b. (Zone 3)
Blanket o.b. (Zone 4)
Blanket o.b. (Zone 5)
Blanket o.b. (Zone 6)
Blanket o.b. (Zone 7)
Blanket o.b. (Zone 8)
Blanket o.b. (Zone 9)
Blanket o.b. (Zone 10)
Blanket o.b. (Zone 11)
Blanket o.b. (Zone 12)
Blanket o.b. (Zone 13)
Blanket o.b. (Zone 14)
Blanket o.b. (Zone 15)
back of blanket o.b.
Gap
VV front wall o.b.
VV shield o.b. (Zone 1)
58
59
60
61
VVSHDO(2)
VVSHDO(3)
VVRWO
THMLSO
62
63
64
TFCCRO
INSO
TFCWO(1)
65
66
67
68
69
70
71
TFCWO(2)
TFCWO(3)
TFCWO(4)
TFCWO(5)
INSI
TFCCFO
THMLS2
72
CRYO
73
74
75
76
BIOSLD(1)
BIOSLD(2)
BIOSLD(3)
BIOSLD(4)
GSSR
VV shield o.b. (Zone 2)
VV shield o.b. (Zone 3)
VV Rear Wall (Outboard)
Thermal Shield
Gap
TF coil case rear wall (o.b.)
Insulator (Filler & Gr. Ins)
TFCW Outboard
TFCW Outboard
TFCW Outboard
TFCW Outboard
TFCW Outboard
Insulator (Filler+Gr. Ins.)
TF coil case front wall o.b.
Thermal Shield
Gap
Cryostat
Gap
Biological Shield (1)
Biological Shield (2)
Biological Shield (3)
Biological Shield (4)
Be
DSCu(84%), 316L(N)-IG(1%),
H2O(12%), void (3%)
316L(N)-IG
316L(N)-IG(65%), H2O(35%)
8421
8431
10
22
8453
8464
8490
11
26
9
316L(N)-IG
316L(N)-IG(62%), H2O(38%)
316L(N)-IG
316L(N)-IG(62%), H2O(38%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(75%), H2O(25%)
316L(N)-IG
316L(N)-IG(70%), void(30%)
8499
8519
8543
8562
8586
8614
8638
8668
8692
8720
8744
8764
8788
8822
8872
8897
8957
20
24
19
24
28
24
30
24
28
24
20
24
34
50
25
60
220
9177
9382
9587
9647
9702
10299
10419
10435
205
205
60
55
597
120
16
104
10539
10649
10759
10869
10979
10995
11115
11170
13497
13547
13667
13967
14267
14567
110
110
110
110
16
120
55
2327
50
120
300
300
300
300
316L(N)-IG
30467(42%); 316L(N)-IG(11%);
H2O(47%)
30467(42%); 430(24%), H2O(34%)
30467(42%); 430(24%), H2O(34%)
316L(N)-IG
316LN (5%):304L(48%); Ag(0.018%)
316LN
Glass & Epoxy
316LN (41%): Conductor (44%),
Glass+Epoxy (15%)
Same as above
Same as above
Same as above
Same as above
Glass+Epoxy
316LN
316LN (5%):304L(48%); Ag(0.018%)
304L
Concrete (97%): Steel Rebar (3%)
Concrete (97%): Steel Rebar (3%)
Concrete (97%): Steel Rebar (3%)
Concrete (97%): Steel Rebar (3%)
page III-17
ITER
G 84 RI 3 01-07-13 R1.0
Table III.2.2-3: Maximum isotopic concentrations for most relevant ITER materials
after 0.5 MWa/m2 pulsed average fluence at shutdown. Neutron flux: 0.65 MW/m2
outboard, 0.41 MW/m2 inboard. Selection criteria: half life>1 min, AND
contribution to activity, or contact dose, or decay heat, or clearance index at least
1%. The tungsten activity is based on an average machine fluence of 0.13 MWa/m2.
isotope
W 187
W 185
W 185m
W 181
Re188
Re186
Re188m
W 179
Ta182
W 179m
Ta186
Ta183
Ta184
Ta182m
Ta179
Re184
Ta180
Hf183
10 mmW surface inboard
(plasma surface layer: 25 micro-m)
half life
activity
[y]
[Bq/kg]
2.72E-03
5.24E+14
2.06E-01
3.71E+13
3.17E-06
3.64E+13
3.31E-01
1.43E+13
1.94E-03
6.01E+12
1.03E-02
2.20E+12
3.54E-05
5.79E+11
7.13E-05
2.56E+11
3.14E-01
1.54E+11
1.22E-05
1.02E+11
2.00E-05
6.34E+10
1.39E-02
6.18E+10
9.92E-04
4.34E+10
3.04E-05
2.88E+10
1.61E+00
2.74E+10
1.04E-01
1.99E+10
9.22E-04
1.15E+10
1.22E-04
9.64E+09
Total
isotope
Cu 64
Cu 62
Cu 66
Co 60m
Ni 65
Co 62
Co 60
Mg 27
Ni 63
Co 62m
Al 28
Na 24
Co 61
Zn 65
Zn 63
Sb122
As 76
Total
GSSR
FW heat sink outboard (Cu)
half life
activity
[y]
[Bq/kg]
1.45E-03
1.50E+14
1.85E-05
1.04E+14
9.70E-06
6.67E+13
1.99E-05
4.79E+12
2.87E-04
1.14E+12
2.85E-06
1.05E+12
5.27E+0
9.75E+11
1.80E-05
2.06E+11
9.9E+01
1.75E+11
2.64E-05
1.28E+11
4.26E-06
9.14E+10
1.71E-03
9.10E+10
1.88E-04
5.79E+10
6.69E-01
1.56E+10
7.30E-05
1.51E+10
7.38E-03
1.16E+10
3.00E-03
1.16E+10
N/A
2.93E+09
N/A
Error
[%]
5.72
5.01
7.13
30.37
29.83
30.47
22.09
5.00
23.91
30.29
27.03
23.02
196.20
14.07
24.21
27.17
18.01
3.4
6.21E+14
isotope
Mn 56
Fe 55
V 52
Cr 51
Co 58
Co 58m
Co 60m
Co 57
Mn 54
Al 28
Mo 99
Tc 99m
Mn 57
Mo101
Cu 64
Cr 55
Co 60
Tc101
V 53
Cu 62
Total
Error
[%]
9.47
25.35
48.90
23.04
31.73
28.52
34.73
28.60
28.90
36.65
42.27
20.59
29.66
34.73
28.55
29.56
32.86
54.03
8.6
FW heat sink outboard (SS316)
half life
activity
[y]
[Bq/kg]
2.94E-04
2.20E+13
2.73E+0
1.14E+13
7.12E-06
5.88E+12
7.59E-02
5.56E+12
1.94E-01
3.47E+12
1.02E-03
3.32E+12
1.99E-05
2.88E+12
7.44E-01
2.56E+12
8.55E-01
2.03E+12
4.26E-06
1.18E+12
7.52E-03
9.29E+11
6.86E-04
8.11E+11
3.06E-06
5.34E+11
2.78E-05
4.97E+11
1.45E-03
4.55E+11
6.73E-06
4.47E+11
5.27E+0
4.27E+11
2.70E-05
4.14E+11
3.08E-06
3.41E+11
1.85E-05
3.13E+11
N/A
5.34E+08
page III-18
Error
[%]
12.06
5.06
4.90
4.37
21.38
30.27
20.93
28.08
3.61
24.02
14.03
14.03
5.88
13.34
5.72
10.64
15.02
13.34
5.59
5.01
4.7
ITER
G 84 RI 3 01-07-13 R1.0
Isotope
Table III.2.2-3a: Summary of water activation
Inventory
FW/Blanket
Vacuum Vessel
-6
-9
8.4 x 10 GBq/cc
2.72 x 10 GBq/cc
3528 GBq total
1 GBq total
-6
-9
4.89 x 10 GBq/cc
8.98 x 10 GBq/cc
2054 GBq total
3 GBq total
-3
8.33 GBq/cc
3.58 x 10 GBq/cc
3.5 x 109 GBq total
3.5 x 109 GBq total
H-3
C-14
N-16
7.22 x 10-4 GBq/cc
N-17
2.82 x 10-7 GBq/cc
3 x 105 GBq total
FW/Blanket water hold-up = 420 m
VV water hold-up = 320 m3
III.2.2.2
90 GBq total
3
Activation product mobilization
Of the total inventories of activation products, the vast majority is tightly bound to the metal
structures so that they are not mobile, that is, not available for release in any credible accident
scenario. A small portion of the total activation product inventory is in a form that could be
mobilised in accident conditions. Five mobilisation pathways have been considered and will
be briefly discussed in the following sections. They are:
• Tokamak dust in the vacuum chamber and ducts resulting from plasma-facing
material evaporated previous disruptions and sputtered during normal operation;
•
Material mobilized during a disruption associated with the event sequence being
analyzed;
•
Activated corrosion products in the coolant loops;
•
Volatility or spallation of oxidation products that could form in case of accidents,
and
•
Activated gases such as divertor impurity gas, air, and cryogenic nitrogen.
The actual fraction of mobilizable activation products that becomes mobile depends on the
scenario and is described by the mobilization fraction. The mobilization source term refers to
the product of the inventory times the mobilization fraction.
Dust formation and plasma-vaporization are mechanical mechanisms; the elemental
composition of the mobilized material matches the original metal. Corrosion and oxidationdriven volatility are chemical mechanisms; the composition of mobilized material does not
typically match the original metal. Since the hazard varies among the different chemical
elements, the hazard from oxidation-driven volatility depends on the chemical mix of
volatilized material, which itself depends on the time and temperature of oxidation.
GSSR
page III-19
ITER
III.2.2.3
G 84 RI 3 01-07-13 R1.0
Tokamak dust inventory and mobilization potential
By definition, "dust" is loose material. Dust is defined as smaller than 100 µm in diameter.
Particles larger than 100 µm will not transport to the environment (see e.g. [Cam96]). This
radioactive dust could be mobilized either during maintenance inside the plasma chamber or
in accidents. It can arise from plasma vaporized material produced in a disruption that is part
of a sequence under investigation (see Section III.2.2.6), accumulated over time from
disruptions and sputtering (plasma erosion), or produced by mechanical operations such as
cutting during in-vessel maintenance. Another source of dust production is sputtering of
plasma facing material. Of current tokamaks, the highest reported dust inventory has been
observed during an earlier venting of TFTR, "several kilograms" [LaM93]. Lower amounts
(on the order of grams) have been reported in JET [Cha88], [Cha92] and JT-60 [Oda96].
The analysis steps for dust production from disruptions requires the disruption energy, the
time scale, target materials, impact of vapor shielding, and thermal calculation to determine
how much material is melted and vaporized. Only vaporized material is considered to have
the potential to produce dust.
Dust inventories
Inventory guidelines are set:
-
to limit the mobile activation product inventory inside the VV,
to ensure that chemical reactivity is adequately controlled, and
to avoid the hazard of dust explosions [Gae94],
The guidelines for dust amounts involve the mass (by chemical element), the size, and the
location of the dust. The size is important in determining transport during accidents and
moderately important for dust chemical reactivity. Dust is defined as smaller than 100 µm in
diameter. Particles larger than 100 µm will not transport to the environment. The location is
important for chemical reactivity because it determines the dust temperature; it is moderately
important for transport. Thus, one cannot simply specify the total mass of all dust.
The carbon and beryllium dust limits stem primarily from chemical reactivity concerns. The
tungsten dust limit is determined by both chemical reactivity and activation hazards.
Table III.2.2-4 shows the guidelines for dust inventories. Note that for the radiologically most
hazardous tungsten dust the Assessment Value is set to 350 kg to account for uncertainties.
Dust on the divertor surface is listed separately because the elevated temperature on the
divertor during off-normal situations lead to likely oxidation and subsequent hydrogen
production in case of steam ingress into the vacuum vessel. Complete reaction of the Be, C,
and W dust along the guidelines on hot divertor surfaces would produce 2.5 kg H 2, which is
below the deflagration limit for the ITER vacuum vessel (4 kg).
Table III.2.2-4: Guidelines for Dust Inventory (kg)*
Be
C
Location
Basis
Dust
Dust
plasma facing components of
6
6
H2 (2.5 kg)
the divertor
Total mobilisable dust within
Environmental release limits
100
200
1st confinement barrier
(<100µm)
*In addition 5 kg of dust (diameter 0.1 micro-m) is assumed to be produced by a disruption.
a
W
Dust
6
100a
Assessment Value: 350 kg tungsten dust to account for uncertainties
GSSR
page III-20
ITER
G 84 RI 3 01-07-13 R1.0
These limits need further review and R&D to demonstrate feasibility to monitor the dust and
remove it when necessary. Dust inside the divertor tile castellation gaps may be chemically
less active due to steam supply limitations. The dust in these gaps is also expected to be less
mobile.
Although there are significant uncertainties associated with dust production mechanisms and
rates and extrapolation from present machines to next-generation tokamaks, a very crude
(best) estimate of the dust production rates is given here (see also [DDD1.7]), resulting from
normal operation and off-normal transients, expected in ITER.
The main assumptions used for these estimates are as follows.
For physical and chemical sputtering during normal operation:
- use peak net-erosion rates calculated for the ITER ([Fed00a] i.e., ~0.6 nm/s for the Be first
wall, ~500 m2), and ~7 nm/s for the detached portion of the carbon target near the strike
point (~20 m2). Assume negligible sputtering for tungsten.
- assume that a fraction up to 30% (conservative) of the net-erosion rate ends-up in dust,
the rest (70%) being re-deposited. These films will build-up to a certain thickness and
then eventually flake generate large size debris, depending on several conditions.
For off-normal transients (i.e., disruptions, VDEs):
-
the thermal quench dissipates the 400 MJ thermal energy content of the plasma within
1 ms. Assume that in 70% of the cases the disruptions delivers all thermal energy to the
divertor region, i.e. ~20 MJ/m2 will strike the carbon-clad portion of the divertor (~20 m2
near strike-points) and as a result of secondary radiation ~3 MJ/m 2 reaches ~100 m2 of
nearby W-clad components (e.g., baffle and liner).
In the remaining 30% of the disruptions all thermal energy is assumed to impact the firstwall (assume 20 m2 ~20 MJ/m2).
VDEs load the FW with 20 MJ/m2 on an area of 20 m2 during 100 ms.
The following frequencies of off-normal events used are: 10% of pulses for disruptions
and 1% for VDEs.
- for carbon, erosion of 5 µm per disruption is calculated resulting from direct plasma
impact and including vapour shielding effects. For tungsten on the nearby components
(e.g., baffle and liner) a vaporised thickness of ~2 µm per disruption is calculated resulting
from secondary radiation and neglecting vapour shielding effects. For beryllium on the
FW a vaporised thickness of up to ~140 microns forms during a VDE.
- assume that a fraction of 50% of only the vaporised material generate dust, whereas the
rest is re-deposited. Melting of metal surfaces during disruptions will not generates dust
but rather films and droplets, which can be splashed away and end-up being firmly
attached to the surface, and then may flake.
The results of these estimates are shown in Fig. III.2.2-1. The administrative guideline of 100
kg for tungsten dust is reached in about 500 plasma pulses. Thus the administrative guideline
is reached before the assumed replacement of the divertor. Dust diagnostics and removal
methods (to be developed) are required.
GSSR
page III-21
ITER
G 84 RI 3 01-07-13 R1.0
Size of dust
The size of dust is important for both chemical reaction analysis (surface/volume) and aerosol
transport. The smaller the particles are, the higher is the hazard. The dust size specifications
for accident analyses are shown in Table III.2.2-5. The baseline dust size specification,
which has been set to envelope the size of the dust collected from operating tokamaks and
plasma disruption simulators, is given as a log normal distribution, consistent with tokamak
dust that has been characterised [Pet00]. Limited experiments with carbon indicate the
presence of very small (<0.01µm) particles [Rom01], however they account for a small
percentage of the total mass, and all dust on hot surfaces is assumed to react anyway. Note
that carbon dust is of little radiological concern. Figure III.2.2-2 shows the dust size
distribution and dust sizes collected from operating tokamaks and plasma guns. Dust
diagnostics and removal is addressed in [DDD1.7].
TABLE III.2.2-5
Dust Sizes and Surface Areas Used in Accident Analyses
Dust defined as particles ≤ 100 µm
Dust inventory (e.g., W)
Baseline
Log-normal distribution
CMD=0.5 µm
GSD=2.0
CRSF=2 (metals)
=20 (carbon)
0.1 µm
Plasma-vaporised mass during a disruption as part of
the event sequence being analysed
The log normal distribution is given by:
 (ln d p − ln CMD)2 
Fraction
1

=
exp −
2
∆ ln(d p )
2π ln GSD


2(ln GSD)


dp=particle diameter, CMD = Count Median Diameter; GSD = Geometric Standard
Deviation
Log-normal distribution with CMD=0.5 µm and GSD=2 (mass median diameter of
2.11 µm)
CRSF=chemical reactivity shape factor
CRSF=2 (metals), envelopes various simple forms and shapes
CRSF=20 (carbon), matching general data
Dust mobilization
Many values of the mobilization fraction of dust are found in the literature; few are directly
relevant to ITER conditions that include water flashing in the vacuum vessel while under
vacuum and air ingress to the vacuum vessel while under vacuum.
Mobilization during rapid fluid ingress into a vacuum has been measured in air-into-vacuum
tests at FZK with ~ 460 µm mean size particles. The mobilization fraction is a function of
the rate of pressure rise; see Figure III.2.2-3 [Özd94]. The FZK data cannot be used directly
for ITER because one must scale the results at test pressure rise rates to the pressure rise rates
possible in ITER. For simplification, we did not try to adjust FZK data for each transient.
Pessimistic geometries imply 100% mobilization. Complete mobilization has also been
measured in experiments carried out at JAERI [Kun98] with dust particles of 6 µm size. On
the other hand dust in cups have smaller mobilization fraction. Under pessimistic
assumptions only 10% of dust in cups was mobilized [Kun98].
GSSR
page III-22
ITER
G 84 RI 3 01-07-13 R1.0
Only dust with a mobilization potential will be included in the assessment of dust inventories
in ITER and 100% mobilization of dust is assumed for air or steam ingress events.
Handbook values are used for determining dust mobilization when steam or air mixes with
nitrogen/air in the vacuum vessel [Doe94]. Steam ingress into nitrogen is taken as "venting of
powders or confinement failure at pressures to 0.17 MPa-gauge or less." The handbook
values, given in terms of bounding, median, and low values are 0.5%, 0.2%, and 0.1%
mobilization [DOE94, p. 4-8, p. 4-73]. Air mixing with nitrogen is taken as "suspension of
bulk powder by debris impact and air turbulence from falling objects" with values of 1%,
~0.1%, and 0.01% [DOE94, p. 4-10, p. 4-85/7]. The "median" values are interpreted as
baseline.
JET performed tests with slow, low-velocity air ingress: 5-h vent to air of 200 m3, would
scale to ~ 3 days for ITER. The results were 0.011% (graphite) and 0.0065% (average of
metals) [Cha88], [Cha92].
Table III.2.2-6 summarizes the dust mobilization fractions.
TABLE III.2.2-6
Dust Mobilisation Fractions
Initial Mobilisation
Mechanism
Resuspension:
Steam or air ingress into vacuum
Steam ingress into N2/air atmosphere
Maintenance activities in progress, air mixing with VV
atmosphere
Quiet undisturbed conditions
III.2.2.4
Mobilisation Fraction
100%
0.2%
0.1%
0.01% time-integrated or 0.004%/h
Plasma-vaporized mobilization potential
In very rapid disruptions, some plasma-facing material will melt and even vaporize. The
total amount that can be vaporized in a disruption depends on the plasma-facing material and
on the assumed thermal quench time. In vacuum, all of the vaporized material will
eventually condense on the walls or form dust particles.
This mobilization mechanism is differentiated from accumulated dust because it is not
subject to measurement (or control) prior to a particular pulse. Thus, allowance for it in the
dust limits must be made as implied in preceding Table III.2.2-4. The amount of material
melted is not expected to contribute to dust production as it will immediately re-freeze. The
material vaporized by a disruption that is part of an event sequence is considered totally
mobilized. Plasma-vaporized material is modeled as having very small particle sizes
(0.1 µm). Experiments with metals indicate sizes in the submicron to micron range [Sha97].
For ITER less than 5 kg of dust is produced in a single disruption (see [DDD1.7]).
III.2.2.5
Activated corrosion product inventory and mobilization potential
Background and analysis
Activated corrosion products (ACP) will be present in the various in-vessel and vacuum
vessel coolant loops as well as in any coolant loops related to test modules, auxiliary heating
GSSR
page III-23
ITER
G 84 RI 3 01-07-13 R1.0
or diagnostics equipment. These products impact occupational exposure, routine effluents to
the environment, and potential releases during accidents.
Calculations of activated corrosion product inventories by the PACTITER code (see Volume
XI, customized version of PACTOLE) require the development of a geometrical model of the
cooling system, specification of water chemistry, and material selection. Oxide porosity and
solubilities of relevant ions in water are calculated by a code routine.
Figure III.2.2-4 illustrates the terminology and corrosion processes used by PACTITER.
PACTITER differentiates among the following forms of activated corrosion products:
Coolant activity = ions in solution plus cruds in suspension,
Deposits = loose material on piping and component walls, and
Oxides = fixed, adherent oxide growing on corroding surfaces
The fixed oxides grow by simple corrosion of the base metal. Mechanical erosion of the
oxides is neglected; thus they do not directly contribute to coolant activity in PACTITER.
The coolant activity arises by dissolution of the base metal and transport through open
porosity in the oxide layer into the water, subject to the solubility of each element in water.
The open porosity is the only user-defined variable. The insoluble ions in water form
deposits on top of the oxides. The in-vessel oxides and deposits are more radioactive than the
ex-vessel ones. The ex-vessel oxide activity in the ITER PHTS temperature range is small
(< 0.1%) relative to the total ex-vessel wall activity.
PACTITER's treatment is quasi-steady state and like most corrosion product models for
water-cooled fission reactors is based on movement of metal species into the water
dominated by the solubility of relevant species in the water throughout the loop.
Figure III.2.2-5 shows the solubility of key elements as a function of temperature for the
ITER steel/water chemistry conditions.
PACTOLE has undergone validation under water cooled fission reactor conditions. The
"open porosity" parameter when properly adjusted, gives good agreement for fission reactors
(see also GSSR, Vol. XI). Data show that in power plant conditions, the outer oxide (Fe-Ni
rich) is adequately thick such that the activity in the water is controlled by the solubility (and
solubility change as the temperature varies around the loop) rather than diffusion.
There is no evidence of a qualitative change of mechanism at temperatures lower than power
plants (280-320°C). Data from French PWRs show that during shutdown as the temperature
decreases, the activity in the coolant increases as expected because solubility increases as the
temperature decreases [Ant94]. Verification of the code for the temperature regimes of
interest is expected from the CORELE experiment, intended to determine the stainless steel
release rate in the 100 – 150 °C temperature range.
There is less understanding on the Cu/water corrosion product behavior. Cu is in direct
contact with water only in the divertor & limiter and NBI PHTS. The basic Cu/water
behavior differs from steel/water. Because of the limited understanding in this area and in
order to increase the confidence of using Cu PACTITER estimates for ITER safety analyses,
an experimental campaign was carried out in 1996. The aim was to determine the valence
states of the copper (Cu0, Cu+ or Cu++) and to obtain solubility in steady state conditions
[You97].
GSSR
page III-24
ITER
G 84 RI 3 01-07-13 R1.0
An ACP assessment was carried out for a FW/shield cooling loop [DiP00] and the
divertor/limiter cooling loop [DiP01]. Table III.2.2-7 summarizes the activated corrosion
product radioactive inventory at shut down for one of the three FW/shield coolant loops
based on the SA1_acp activation scenario (320 days of full coolant flow) [DiP00]. The
hazard of the ACP is typically dominated by Mn-56 that arises from Mn and Fe in the base
alloy steel, by Co-57, Co-58, that arise from Ni in the base steel and by Co-60 that arises
from both Ni and Co in the base steel. The suspended cruds and dissolved ions are 1.8 g and
10.9 g respectively. The total ACP deposit mass is ≈ 1.4 kg. Table III.2.2-7a lists the
resulting specific activity of deposits and ions and cruds in solution.
To account for uncertainties especially with the divertor cooling loops with copper water
contact, the amount of activation corrosion product deposits is assumed to be 10 kg per
cooling loop in the analysis of reference events and 60 g ions and cruds. The hazard potential
of activation corrosion products inside the vacuum vessel HTS is assumed to be 1% of the
hazard potential of activation corrosion products of the FW/shield cooling loop because the
average n-flux is reduced by several orders of magnitudes inside the vacuum vessel. This
assumption needs confirmation.
TABLE III.2.2-7
Activated Corrosion Products radioactive inventory for a FW/shield loop
Parameter Units Fe-55
Mn-54
Mn-56
Co-58
Suspended GBq/t 1.73E-02 1.43E-03 5.20E-02 1.61E-03
cruds *
Soluble
GBq/t 7.96E-02 3.37E-02 1.15E+00 3.79E-02
ions *
GBq/t 9.70E-02 3.51E-02 1.20E+00 3.95E-02
Total
Coolant
Activity *
GBq 2.96E+03 1.41E+02 1.93E+03 1.51E+02
Total
Deposit
Activity**
* total coolant mass: 126 t (excluding pressurizer),
Mass of suspended cruds and ions: 12.7 g
** Mass of deposits: 1.4 kg
Co-60
Cr-51
Ni-57
Co-57
Total
9.82E-04 1.98E-06 3.97E-04 2.04E-03 7.58E-02
2.31E-02 4.38E-05 8.52E-03 4.80E-02 1.38E+00
2.41E-02 4.58E-05 8.91E-03 5.01E-02 1.46E+00
2.01E+02 1.63E+02 6.46E+01 3.78E+02 5.99E+03
TABLE III.2.2-7a
Specific activity of Activated Corrosion Products for a FW/shield loop
GSSR
isotope
half life
[y]
Fe-55
Mn-54
Mn-56
Co-58
Co-60
Cr-51
Ni-57
Co-57
2.73E+01
8.55E-01
2.94E-04
1.94E-01
5.27E+01
7.59E-02
4.11E-03
7.44E-01
deposit
activity
[Bq/kg-deposit]
2.07E+12
9.86E+10
1.35E+12
1.06E+11
1.41E+11
1.14E+11
4.52E+10
2.64E+11
Ion and cruds in solution
activity
[Bq/kg-Ion&Crud]
9.61E+11
3.49E+11
1.19E+13
3.92E+11
2.39E+11
4.54E+08
8.85E+10
4.96E+11
page III-25
ITER
G 84 RI 3 01-07-13 R1.0
TABLE III.2.2-8
Activated Corrosion Products radioactive inventory for the divertor loop
Fe-59
Parameter
Mn-54
Co-58
Co-57
Co-60
Cr-51
GBq/t 3.60E-06 5.18E-05 1.55E-04 8.48E-06 1.13E-03 1.28E-07
Activity in
suspension*
Activity in
GBq/t 5.41E-04 1.42E-02 4.23E-02 2.32E-03 3.09E-01 3.89E-05
solution*
Coolant
GBq 8.55E-02 2.23E+00 6.67E+00 3.66E-01 4.86E+01 6.12E-03
activity*
Total deposit
GBq 1.28E+00 4.51E+01 2.37E+02 1.33E+01 1.01E+03 2.36E+01
activity **
* total coolant mass: 157 t (excluding pressurizer)
Mass of suspended cruds and ions: 42 g
** Mass of deposits: 8.9 kg
Mn-56
Cu-64
Total
1.99E-03 8.54E-01 8.57E-01
5.42E-01 1.37E+00 2.28E+00
8.54E+01 3.48E+02 4.91E+02
1.22E+03 5.05E+04 5.30E+04
TABLE III.2.2-8a
Specific activity of Activated Corrosion Products for the divertor loop
isotope
half life
[y]
Fe-59
Mn-54
Co-58
Co-57
Co-60
Cr-51
Mn-56
Cu-64
1.23E-01
8.55E-01
1.94E-01
7.44E-01
5.27E+01
7.59E-02
2.94E-04
1.45E-03
deposit
activity
[Bq/kg-deposit]
1.44E+08
5.07E+09
2.66E+10
1.49E+09
1.13E+11
2.65E+09
1.37E+11
5.67E+12
Ion and cruds in solution
activity
[Bq/kg-Ion&Crud]
2.04E+09
5.33E+10
1.59E+11
8.70E+09
1.16E+12
1.46E+08
2.03E+12
8.31E+12
Activated corrosion product mobilization
In safety assessments, the particle size of corrosion products is used for aerosol transport
calculations. Table III.2.2-9 shows the values considered; although these are proving not to
be particularly important because ACP is only mobilized in very wet (saturated steam)
conditions and steam condensation rapidly erases memory of the initial ACP particle size.
TABLE III.2.2-9
Activated Corrosion Products Particle/Droplet Size and
Mobilisation Potential
Baseline
Particle/droplet size
2 µm
Mobilisation
1.3% of ACPs in
water spilled
The ACP "mobilisation fraction" is the product of three factors:
GSSR
•
Lift-off of corrosion product deposits into water;
•
Carry-over or entrainment of ions and particles into water droplets (fog);
page III-26
ITER
G 84 RI 3 01-07-13 R1.0
•
Fraction of water flashing into volume.
The "lift-off" fraction for the "deposits" is typically 100% but zero for the "oxides" [DiP00]
in severe depressurization (loss of coolant) events. For extremely slow, less violent
depressurizations (e.g., heat exchanger tube rupture), 20% is assumed. Data for mobilization
of suspended activity as liquid droplets or fog come from the Water Aerosol Leakage
Experiments (WALE) [Flu90a], [Flu90b]. These used CsNO3 as the tracer in water.
Neglecting samples with direct line of sight of the impinging water steam (which reached
2.6%), the highest data point was 1.3%. This is used as the baseline mobilization fraction for
ITER safety analysis. The lowest data point was 0.3%. One handbook bounding value is 7%
[DOE94] for "depressurization of superheated liquid, 50-100°C superheat. (The
corresponding "median" suggested value was 1.4%.) The highest mobilization was in a test
with 100 cm3 of 0.85 MPa (161°C) water producing a mobilization fraction of 6% (airborne
release fraction of 9% times respirable fraction of 69%) [DOE94]. The small size of the test
article makes extrapolation to ITER somewhat questionable. The maximum data point in the
WALE tests (1.3% [Flu90a], [Flu90b]) is taken as a conservative estimate (regardless of
superheat).
III.2.2.6
Oxidation-driven mobilization potential
Most of the activation products involved have higher vapour pressure and volatility as oxides
or hydroxides than as metals. Thus, oxidation of activated material can enhance the mobility
in two ways:
•
Through volatility of oxidation products that become released as a vapor and can
subsequently undergo nucleation (gas to particle conversion) and form aerosols;
and
•
Through spallation of oxidation products that become released in the form of
small flakes.
Oxidation-driven mobilization occurs only in accident scenarios in which air, water or steam
enters the plasma chamber and oxidizes the hot plasma facing components and other invessel surfaces.
In-vessel components include five materials. Beryllium and carbon are not relevant from an
oxidation-driven mobilization standpoint as the activation levels are so low. Their
mobilization quantities might still influence the aerosol behavior. There is no steel or copper
directly exposed to the plasma as beryllium, carbon, or tungsten cover such surfaces. The
total amount of bare steel and copper is minimal, and accident temperatures are low enough
that oxidation-driven mobilization of these materials is not a significant contributor to the
source term. Oxidation-driven mobilization of tungsten is considered because there is bare
tungsten in the Divertor.
Mass flux data and activation product inventory combine to produce the hazard as a function
of temperature per unit time. Figure III.2.2-6 shows the theoretical early dose per area and
time to the maximum exposed individual from W as a function of temperature, in a steam
environment, assuming no radioactivity confinement and ground level release[McC97b].
Assuming a temperature of 400°C for two days, and 200 m2 tungsten, the most conservative
GSSR
page III-27
ITER
G 84 RI 3 01-07-13 R1.0
dose scenario would result in a total dose due to oxidation-driven mobilization of tungsten,
without taking credit for confinement, of about 30 µSv [McC97b].
Besides volatility, oxides can spall from the surface and in this way contribute to the
mobilization source term. Spallation mainly occurs as a result of cool down and subsequent
thermal stress. The total mass of small (transportable) particles that could be produced by
spallation processes was estimated to be very low: between 0.1 g and 5 g [Hag94].
III.2.2.7
Activated gases
There are some gases in the ITER facility that will activate. By definition, these are mobile.
The following two subsections briefly describe the activation of air between the cryostat and
bioshield, impurity gases in the divertor.
Activated air from between the cryostat and bioshield
The space between the outer wall of the Cryostat and the Bioshield contains air that will
become activated. The rate of air activation is given in Table III.2.2-10. These gases are
exhausted; a significant inventory does not accumulate.
TABLE III.2.2-10
Air Activation Rate for the 1998 ITER design
Isotope
Half-life time
C-14
5600 a
N-13
10 min
N-16
7.4 seconds
Ar-37
35 days
Ar-39
260 a
Ar-41
1.8 h
Activation Rate
3
(Bq/cm -air per second irradiation)
1.87 x 10-9
1.53 x 10-4
3.18 x 10-2
3.03 x 10-9
1.04 x 10-11
5.44 x 10-5
Activated divertor exhaust gases
Noble gases may be injected into the divertor to improve line-radiation. These gases will
become activated by neutrons in the divertor region during a pulse. The gas will be pumped
together with the other exhaust gases by the normal vacuum pumps. The activation depends
on the residence time in the divertor, which is related to the divertor gas volume, density, and
injection rate.
III.2.3 CHEMICAL
Table III.2.3-1 lists the major chemicals at the ITER facility. There is extensive industrial
experience in safely handling each of these. There are few details yet available on the
inventory of each, except for beryllium (below), hydrogen (Section III.3.5.1), and ozone
(Section III.3.5.5).
GSSR
page III-28
ITER
G 84 RI 3 01-07-13 R1.0
Beryllium is a hazardous material for humans (for example, berylliosis in the respiratory
tract, a severe chemical pneumonitis which can be fatal), a possible carcinogen for humans,
and a known carcinogen for animals. Only from 1-5% of the human population is thought to
be sensitive to beryllium exposure [Gre90], [Lon94]. Acute health effects may occur due to
the inhalation of beryllium fume or dust and may cause a serious lung condition if the
beryllium concentration is above 100 µg/m3 [JET89]. A concentration of 10 mg/m3 is
considered an immediate danger to life and health [Lon94].
TABLE III.2.3-1
Chemicals Present at the ITER Facility
Chemical
Beryllium, carbon, iron, tungsten, titanium
Hydrogen
Ozone
Cryogenic helium and nitrogen
Freon used in refrigerant systems
SF6 used as insulating gas
Vacuum pump oil
Oil for electrical equipment
Lead shielding
Acid and alkaline cleaning and pickling solutions
Solvents like kerosene, acetone, and ethyl alcohol
Diborane
Demineralizing substances and spent resin from the
demineralizing beds in the CVCS
Carbon dioxide
Orthophosphates and calcium phosphate inhibitors
Commercial products to control bacterial, fungal, and
algae slimes in cooling tower water
Comment
Hazardous as small, respirable particles
Chemically inert but can cause effects similar to
thermal burns if exposed to body tissue
Poisons HT to HTO catalysis
If used in pumping systems it might become tritium
contaminated.
In the work shops
In special safety cans
Hazardous and flammable gas potentially used for wall
conditioning
For fire protection systems
To control corrosion in the cooling water systems
The current design requires approximately 13 tons of beryllium on plasma-facing
components as a 10-mm protective layer. Within the ITER facility, there may be respirable
dust of the bare metal, carbide, oxide, and possibly nitride. The ITER guideline for the
maximum amount of beryllium tokamak dust that will be allowed to accumulate inside the
plasma chamber has been established to be no more than 100 kg (see table 2.2-4).
III.3 STORED ENERGY SOURCES
ITER will contain a number of energy sources that could potentially cause release of
radioactivity or hazardous materials in an accident. The principal energy sources are:
GSSR
•
The plasma;
•
Energy stored in the magnetic fields;
page III-29
ITER
G 84 RI 3 01-07-13 R1.0
•
Decay heat from activation products;
•
Thermal energy of the water coolant used to cool the Primary First Wall, Baffle,
Limiter, and Divertor;
•
Potential for hydrogen, dust, and ozone explosions;
•
Chemical energy from beryllium reactions with air, water, or steam and resultant
hydrogen productions.
Energy sources such as fuel or gas tanks will be handled in the same manner as in
conventional plants. These sources will be located far away from the tokamak.
III.3.1 PLASMA ENERGY SOURCES
During normal operation the plasma delivers fusion power of 500 MW (nominal power). The
power level can potentially be increased by a factor of four in fusion power transients (see
Volume VII). The exact amount and time scales involved are dependent on the plasma
operational point and assumptions made on the β-limit.
The plasma current is nominally 15 MA with a maximum of 17 MA. When the plasma
disrupts it will deposit its thermal energy (thermal quench) and magnetic energy (current
quench) to the divertor and the other plasma-facing components. The machine is designed
for thousands of disruptions. The thermal plasma energy of 0.4 GJ will be released in a time
scale between 1 to 10 ms, and the magnetic energy of 0.3 GJ in 30 ms. There are disruptions
involving vertical motion of the plasma. A major issue is electromagnetically induced forces
from these "vertical displacement events" (VDEs). Another issue is the formation of a beam
of relativistic electrons ('runaway electrons') during the current quench. These electrons may
cause local damage on in-vessel components.
III.3.2 MAGNET ENERGY SOURCES
There are 18 toroidal field (TF) coils, one Central Solenoid (CS), and six additional Poloidal
Field (PF) coils. The CS is divided into 6 modules called CSU1,2,3 and CSL1,2,3 (U: upper,
L: lower). Table III.3.2-1 summarizes the magnet energy parameters for the nominal
operating conditions and the maximum induced current in a single coil in case of a short
[DDD1.1]. The total stored energy is 51 GJ distributed over 8700 tonnes. Other energy
sources related to the superconducting magnet systems are the cryogenic coolant that, when
released, could evaporate and pressurize the cryostat.
All TF coils are wired in series. The CS, like all PF coils, is charged and discharged each
pulse. The Poloidal Field (PF) coils are powered by power supplies of several hundreds of
MW.
GSSR
page III-30
ITER
Initial
Energy/coil
Max Short
Energy
G 84 RI 3 01-07-13 R1.0
Table III.3.2-1
Energy Stored in a Coil with an External Short, GJ
TF CSU3 CSU2 CSU1 PF1 PF2 PF3 PF4
PF5
PF6
2.3
1.0
1.12
1.11
0.92
0.15
1.30
0.80
0.92
1.32
6
1.5
2.0
2.1
1.3
1.1
3.2
3.3
2.0
2.8
III.3.3 NUCLEAR ENERGY SOURCES
The decay heat is a function of the operating scenario, which has already been described in
Section III.2.2. A longer pulse operation, such as few thousand second plasma burn, is
possible in actual operation as long as the resultant global ITER decay heat stays below the
baseline ITER decay heat curve. The decay heat densities in the ITER FW and shielding
blanket as a function of time are listed in Table III.3.3-1 [Cep00]. This table also lists the
global decay heat and nuclear heating densities for ITER.
Because the decay heat curve is theoretically calculated, there are several potential sources of
errors. The errors are generally less than 10% for stainless steel and less than 10-15% for
copper decay heat predictions for the first 3 days after irradiation [Tay99, Mor01] Figure
III.3.3-1 shows the comparison of experimental data and calculated decay heat for one week
after shutdown.
Potential errors introduced by transport calculation modeling depend on modeling
sophistication and configuration control. Two approaches, a 1D analysis and 2D analysis,
were used to address this issue. The modeling errors potentially introduced by these two
independent methodologies indicate that the difference is less than ~ 10% [Saj97]
With all of these uncertainties taken together, the best estimate of the accuracy of the global
ITER decay heat calculation based on presently available data is less than ±20%.
All these statements on accuracy depend on the particular design data, including the radial
build data, material specification, operational scenario, and the level of detail in the
modeling. Accurate representation of the design in the first ~ 20 cm of material from the
plasma is critical for an accurate prediction of the decay heat. Also note that the current
decay heat values are for the SA1 operational scenario, which is still very conservative, by
assuming 100% availability with 30% duty cycle for the last month with 3600 second pulses.
A more realistic scenario will result in several tens of percent lower decay heat after one day
to several months as shown in Appendix-2 of [Saj97].
GSSR
page III-31
ITER
G 84 RI 3 01-07-13 R1.0
Table 3.3-1:
Decay heat densities [MW/m3] and integrated decay heat power [MW] and energy [MJ]
for the ITER FW and shielding blanket. Zones names correspond to table 2.2-2
Part (1/3)
material
fraction
BLBKI
BLKTI(15)
BLKTI(14)
BLKTI(13)
BLKTI(12)
BLKTI(11)
BLKTI(10)
BLKTI(9)
BLKTI(8)
BLKTI(7)
BLKTI(6)
BLKTI(5)
BLKTI(4)
BLKTI(3)
BLKTI(2)
BLKTI(1)
FWCUI
FWBEI
length of
zone
[mm]
50
34
24
20
24
28
24
30
24
28
24
19
24
20
26
11
22
10
0.7
1
0.75
1
0.75
1
0.75
1
0.75
1
0.62
1
0.62
1
0.65
1
0.9
1
Nuclear
Heating
[MW/m3]
6.70E-02
1.50E-01
1.70E-01
2.60E-01
2.80E-01
4.40E-01
5.00E-01
8.00E-01
9.40E-01
1.51
1.57
2.57
2.52
4.07
4.14
5.78
6.56
4.05
FWBEO
FWCUO
BLKTO(1)
BLKTO(2)
BLKTO(3)
BLKTO(4)
BLKTO(5)
BLKTO(6)
BLKTO(7)
BLKTO(8)
BLKTO(9)
BLKTO(10)
BLKTO(11)
BLKTO(12)
BLKTO(13)
BLKTO(14)
BLKTO(15)
BLBKO
10
22
11
26
20
24
19
24
28
24
30
24
28
24
20
24
34
50
1
0.9
1
0.65
1
0.62
1
0.62
1
0.75
1
0.75
1
0.75
1
0.75
1
0.7
5.38
8.10
7.65
5.64
5.55
3.48
3.53
2.17
2.08
1.30
1.11
6.90E-01
6.00E-01
3.80E-01
3.50E-01
2.30E-01
2.00E-01
9.00E-02
Name of
Zone
Sum [MW]
Energy [GJ]
GSSR
622
N/A
Time:1 s
MW/m3
5 min
MW/m3
30 min
MW/m3
60 min
MW/m3
3h
MW/m
3
7.27E-04
1.39E-03
2.72E-03
2.93E-03
4.18E-03
4.35E-03
6.75E-03
7.28E-03
1.19E-02
1.44E-02
2.65E-02
2.78E-02
4.19E-02
4.31E-02
6.37E-02
6.42E-02
3.86E-01
2.35E-02
6.82E-04
1.30E-03
2.58E-03
2.76E-03
3.94E-03
4.06E-03
6.31E-03
6.72E-03
1.10E-02
1.32E-02
2.46E-02
2.54E-02
3.84E-02
3.86E-02
5.70E-02
5.55E-02
2.67E-01
8.38E-05
5.84E-04
1.12E-03
2.24E-03
2.39E-03
3.42E-03
3.50E-03
5.46E-03
5.77E-03
9.47E-03
1.13E-02
2.12E-02
2.17E-02
3.30E-02
3.27E-02
4.82E-02
4.60E-02
8.91E-02
6.03E-05
5.09E-04
9.82E-04
1.96E-03
2.09E-03
2.99E-03
3.07E-03
4.78E-03
5.06E-03
8.30E-03
9.93E-03
1.86E-02
1.91E-02
2.90E-02
2.88E-02
4.24E-02
4.05E-02
6.31E-02
5.66E-05
3.19E-04
6.09E-04
1.20E-03
1.29E-03
1.84E-03
1.90E-03
2.95E-03
3.15E-03
5.15E-03
6.19E-03
1.15E-02
1.19E-02
1.79E-02
1.80E-02
2.65E-02
2.58E-02
5.37E-02
4.89E-05
3.12E-02
5.29E-01
8.61E-02
8.23E-02
5.63E-02
5.41E-02
3.64E-02
3.47E-02
1.91E-02
1.58E-02
9.76E-03
9.02E-03
5.83E-03
5.59E-03
3.92E-03
3.63E-03
1.84E-03
9.59E-04
1.07E-04
3.68E-01
7.27E-02
7.22E-02
4.96E-02
4.90E-02
3.29E-02
3.19E-02
1.74E-02
1.45E-02
8.97E-03
8.41E-03
5.43E-03
5.26E-03
3.69E-03
3.43E-03
1.73E-03
8.97E-04
7.38E-05
1.14E-01
5.94E-02
6.04E-02
4.16E-02
4.17E-02
2.80E-02
2.74E-02
1.49E-02
1.25E-02
7.68E-03
7.26E-03
4.68E-03
4.56E-03
3.19E-03
2.98E-03
1.49E-03
7.68E-04
6.89E-05
7.46E-02
5.25E-02
5.31E-02
3.67E-02
3.67E-02
2.46E-02
2.40E-02
1.31E-02
1.10E-02
6.73E-03
6.36E-03
4.09E-03
3.99E-03
2.79E-03
2.60E-03
1.30E-03
6.70E-04
5.92E-05
6.25E-02
3.37E-02
3.35E-02
2.32E-02
2.28E-02
1.54E-02
1.49E-02
8.17E-03
6.81E-03
4.21E-03
3.93E-03
2.54E-03
2.46E-03
1.72E-03
1.60E-03
8.09E-04
4.20E-04
1.12E+01 8.61E+00 4.96E+00 4.07E+00 2.75E+00
1.12E-02 2.98E+00 1.32E+01 2.13E+01 4.58E+01
page III-32
ITER
G 84 RI 3 01-07-13 R1.0
Table 3.3-1:
Decay heat densities [MW/m3] and integrated decay heat power [MW] and energy [MJ]
for the ITER FW and shielding blanket. Zones names correspond to table 2.2-2
Part (2/3)
5h
MW/m3
10 h
MW/m3
1d
MW/m3
3d
MW/m3
1w
MW/m3
1m
MW/m3
3m
MW/m3
2.10E-04
3.95E-04
7.66E-04
8.28E-04
1.18E-03
1.23E-03
1.90E-03
2.06E-03
3.32E-03
4.03E-03
7.34E-03
7.69E-03
1.15E-02
1.18E-02
1.73E-02
1.73E-02
4.83E-02
4.37E-05
9.56E-05
1.71E-04
3.10E-04
3.45E-04
4.81E-04
5.30E-04
7.94E-04
9.10E-04
1.42E-03
1.77E-03
3.03E-03
3.34E-03
4.86E-03
5.35E-03
7.74E-03
8.43E-03
3.71E-02
3.58E-05
5.34E-05
8.89E-05
1.45E-04
1.70E-04
2.30E-04
2.76E-04
3.96E-04
4.93E-04
7.30E-04
9.56E-04
1.48E-03
1.77E-03
2.47E-03
3.01E-03
4.28E-03
5.19E-03
1.85E-02
2.51E-05
4.56E-05
7.65E-05
1.26E-04
1.48E-04
2.00E-04
2.41E-04
3.48E-04
4.36E-04
6.47E-04
8.52E-04
1.32E-03
1.59E-03
2.21E-03
2.71E-03
3.86E-03
4.72E-03
3.53E-03
1.27E-05
3.98E-05
6.75E-05
1.12E-04
1.32E-04
1.80E-04
2.17E-04
3.14E-04
3.94E-04
5.89E-04
7.79E-04
1.21E-03
1.46E-03
2.04E-03
2.50E-03
3.57E-03
4.37E-03
2.34E-03
9.27E-06
3.33E-05
5.70E-05
9.56E-05
1.12E-04
1.53E-04
1.84E-04
2.67E-04
3.34E-04
5.00E-04
6.61E-04
1.03E-03
1.24E-03
1.74E-03
2.13E-03
3.04E-03
3.71E-03
2.30E-03
8.51E-06
2.71E-05
4.65E-05
7.83E-05
9.11E-05
1.24E-04
1.46E-04
2.12E-04
2.61E-04
3.92E-04
5.11E-04
8.07E-04
9.55E-04
1.34E-03
1.62E-03
2.31E-03
2.78E-03
2.25E-03
7.86E-06
1 year
MW/m
3
1.83E-05
3.18E-05
5.51E-05
6.23E-05
8.48E-05
9.52E-05
1.39E-04
1.61E-04
2.44E-04
3.05E-04
5.01E-04
5.61E-04
8.04E-04
8.99E-04
1.29E-03
1.42E-03
2.03E-03
6.66E-06
FWBEO
FWCUO
BLKTO(1)
BLKTO(2)
BLKTO(3)
BLKTO(4)
BLKTO(5)
BLKTO(6)
BLKTO(7)
BLKTO(8)
BLKTO(9)
BLKTO(10)
BLKTO(11)
BLKTO(12)
BLKTO(13)
BLKTO(14)
BLKTO(15)
BLBKO
5.27E-05
5.61E-02
2.29E-02
2.22E-02
1.53E-02
1.48E-02
1.01E-02
9.56E-03
5.34E-03
4.41E-03
2.75E-03
2.53E-03
1.65E-03
1.57E-03
1.11E-03
1.02E-03
5.24E-04
2.76E-04
4.30E-05
4.34E-02
1.16E-02
1.03E-02
7.18E-03
6.42E-03
4.47E-03
4.02E-03
2.40E-03
1.91E-03
1.23E-03
1.07E-03
7.17E-04
6.47E-04
4.65E-04
4.15E-04
2.28E-04
1.27E-04
3.03E-05
2.21E-02
7.42E-03
5.99E-03
4.21E-03
3.41E-03
2.46E-03
2.04E-03
1.33E-03
1.01E-03
6.81E-04
5.43E-04
3.78E-04
3.13E-04
2.32E-04
1.96E-04
1.20E-04
7.12E-05
1.60E-05
4.99E-03
6.79E-03
5.45E-03
3.82E-03
3.07E-03
2.22E-03
1.82E-03
1.19E-03
8.95E-04
6.04E-04
4.78E-04
3.32E-04
2.73E-04
2.02E-04
1.70E-04
1.03E-04
6.10E-05
1.22E-05
3.63E-03
6.33E-03
5.06E-03
3.54E-03
2.84E-03
2.04E-03
1.68E-03
1.09E-03
8.17E-04
5.48E-04
4.33E-04
2.98E-04
2.46E-04
1.80E-04
1.52E-04
9.13E-05
5.33E-05
1.12E-05
3.58E-03
5.38E-03
4.31E-03
3.01E-03
2.42E-03
1.74E-03
1.43E-03
9.23E-04
6.94E-04
4.64E-04
3.67E-04
2.53E-04
2.09E-04
1.53E-04
1.29E-04
7.71E-05
4.46E-05
1.03E-05
3.50E-03
3.98E-03
3.24E-03
2.26E-03
1.85E-03
1.32E-03
1.11E-03
7.09E-04
5.39E-04
3.61E-04
2.91E-04
2.00E-04
1.68E-04
1.24E-04
1.06E-04
6.25E-05
3.61E-05
8.71E-06
3.16E-03
1.93E-03
1.70E-03
1.20E-03
1.06E-03
7.48E-04
6.65E-04
4.10E-04
3.27E-04
2.17E-04
1.87E-04
1.28E-04
1.14E-04
8.35E-05
7.37E-05
4.24E-05
2.42E-05
Sum [MW]
Energy [GJ]
1.99E+00
6.29E+01
1.12E+00
9.09E+01
6.01E-01
1.34E+02
3.53E-01
2.17E+02
3.13E-01
3.32E+02
2.72E-01
9.13E+02
2.16E-01
2.18E+03
1.35E-01
6.34E+03
Name of
Zone
BLBKI
BLKTI(15)
BLKTI(14)
BLKTI(13)
BLKTI(12)
BLKTI(11)
BLKTI(10)
BLKTI(9)
BLKTI(8)
BLKTI(7)
BLKTI(6)
BLKTI(5)
BLKTI(4)
BLKTI(3)
BLKTI(2)
BLKTI(1)
FWCUI
FWBEI
GSSR
page III-33
ITER
G 84 RI 3 01-07-13 R1.0
Table 3.3-1:
Decay heat densities [MW/m3] and integrated decay heat power [MW] and energy [MJ]
for the ITER FW and shielding blanket. Zones names correspond to table 2.2-2
Part (3/3)
3a
MW/m3
10 a
MW/m3
30 a
MW/m3
100 a
MW/m3
1000 a
MW/m3
10,000 a
MW/m3
100,000 a 1e6 a
MW/m3 MW/m3
1.29E-05
2.25E-05
3.93E-05
4.38E-05
5.96E-05
6.52E-05
9.50E-05
1.06E-04
1.62E-04
1.96E-04
3.29E-04
3.54E-04
5.10E-04
5.38E-04
7.62E-04
7.70E-04
1.56E-03
5.43E-06
5.02E-06
8.74E-06
1.54E-05
1.70E-05
2.32E-05
2.51E-05
3.66E-05
4.05E-05
6.15E-05
7.38E-05
1.25E-04
1.32E-04
1.91E-04
1.96E-04
2.76E-04
2.68E-04
6.23E-04
3.14E-06
3.65E-07
6.40E-07
1.13E-06
1.25E-06
1.71E-06
1.84E-06
2.70E-06
2.97E-06
4.53E-06
5.42E-06
9.24E-06
9.73E-06
1.41E-05
1.44E-05
2.03E-05
1.95E-05
4.70E-05
8.44E-07
2.86E-09
7.46E-09
1.95E-08
1.89E-08
2.97E-08
2.60E-08
4.60E-08
4.15E-08
7.94E-08
8.80E-08
2.01E-07
1.81E-07
3.02E-07
2.47E-07
3.86E-07
2.72E-07
1.42E-06
1.59E-08
2.59E-10
4.33E-10
7.47E-10
8.28E-10
1.13E-09
1.25E-09
1.83E-09
2.06E-09
3.14E-09
3.84E-09
6.54E-09
7.03E-09
1.02E-08
1.09E-08
1.57E-08
1.63E-08
2.66E-09
2.59E-10
1.40E-10
2.34E-10
4.05E-10
4.45E-10
6.09E-10
6.56E-10
9.62E-10
1.06E-09
1.61E-09
1.92E-09
3.32E-09
3.46E-09
5.05E-09
5.08E-09
7.20E-09
6.82E-09
6.72E-11
1.84E-10
2.87E-11
5.10E-11
9.60E-11
1.02E-10
1.44E-10
1.48E-10
2.26E-10
2.36E-10
3.79E-10
4.41E-10
8.15E-10
8.11E-10
1.22E-09
1.16E-09
1.67E-09
1.45E-09
6.13E-11
1.43E-10
8.74E-13
1.30E-12
1.87E-12
2.21E-12
2.80E-12
3.37E-12
4.49E-12
5.48E-12
7.41E-12
9.23E-12
1.27E-11
1.52E-11
1.98E-11
2.39E-11
3.18E-11
3.73E-11
2.81E-11
9.64E-11
FWBEO
FWCUO
BLKTO(1)
BLKTO(2)
BLKTO(3)
BLKTO(4)
BLKTO(5)
BLKTO(6)
BLKTO(7)
BLKTO(8)
BLKTO(9)
BLKTO(10)
BLKTO(11)
BLKTO(12)
BLKTO(13)
BLKTO(14)
BLKTO(15)
BLBKO
7.15E-06
2.43E-03
9.75E-04
9.47E-04
6.79E-04
6.46E-04
4.55E-04
4.26E-04
2.57E-04
2.13E-04
1.41E-04
1.26E-04
8.69E-05
7.94E-05
5.83E-05
5.23E-05
2.97E-05
1.69E-05
4.26E-06
9.69E-04
3.26E-04
3.33E-04
2.42E-04
2.37E-04
1.67E-04
1.60E-04
9.58E-05
8.05E-05
5.34E-05
4.84E-05
3.33E-05
3.08E-05
2.26E-05
2.04E-05
1.16E-05
6.57E-06
1.20E-06
7.31E-05
2.36E-05
2.44E-05
1.77E-05
1.75E-05
1.23E-05
1.18E-05
7.03E-06
5.92E-06
3.91E-06
3.57E-06
2.44E-06
2.27E-06
1.66E-06
1.50E-06
8.45E-07
4.78E-07
2.28E-08
2.12E-06
3.11E-07
4.42E-07
2.91E-07
3.63E-07
2.22E-07
2.52E-07
1.12E-07
1.02E-07
5.39E-08
6.03E-08
3.42E-08
3.93E-08
2.50E-08
2.57E-08
9.83E-09
3.76E-09
2.86E-10
3.98E-09
2.14E-08
2.00E-08
1.41E-08
1.32E-08
9.19E-09
8.53E-09
5.08E-09
4.17E-09
2.75E-09
2.44E-09
1.67E-09
1.52E-09
1.11E-09
9.95E-10
5.75E-10
3.41E-10
2.01E-10
1.00E-10
8.21E-09
8.62E-09
6.23E-09
6.25E-09
4.37E-09
4.24E-09
2.49E-09
2.11E-09
1.39E-09
1.28E-09
8.72E-10
8.11E-10
5.92E-10
5.38E-10
3.09E-10
1.83E-10
1.55E-10
9.12E-11
1.69E-09
1.94E-09
1.38E-09
1.49E-09
1.01E-09
1.03E-09
5.65E-10
4.92E-10
3.10E-10
2.98E-10
1.96E-10
1.92E-10
1.36E-10
1.27E-10
6.72E-11
3.75E-11
1.04E-10
4.05E-11
4.72E-11
4.02E-11
3.05E-11
2.54E-11
1.97E-11
1.66E-11
1.21E-11
9.81E-12
7.27E-12
5.98E-12
4.50E-12
3.74E-12
2.94E-12
2.50E-12
1.72E-12
1.14E-12
Sum [MW]
Energy [GJ]
8.69E-02
3.27E-02
2.43E-03
5.20E-05
1.24E-06
5.60E-07
1.29E-07
3.82E-09
1.33E+04
2.65E+04
3.76E+04
4.03E+04
4.11E+04
4.13E+04
4.23E+04
4.42E+04
Name of
Zone
BLBKI
BLKTI(15)
BLKTI(14)
BLKTI(13)
BLKTI(12)
BLKTI(11)
BLKTI(10)
BLKTI(9)
BLKTI(8)
BLKTI(7)
BLKTI(6)
BLKTI(5)
BLKTI(4)
BLKTI(3)
BLKTI(2)
BLKTI(1)
FWCUI
FWBEI
GSSR
page III-34
ITER
G 84 RI 3 01-07-13 R1.0
III.3.4 THERMAL ENERGY OF THE COOLANT
The thermal energy of the coolant can be released if the water is spilled or leaked. Therefore
all volumes surrounding the primary HTS either self-contain that pressure, vent it to other
volumes, or send the fluid to a pressure suppression system. Table III.3.4-1 summarises the
inventory and temperature of water coolant. Some minor evolution of these parameters is
expected in the design process. To limit the pressurization of the vacuum vessel below 200
kPa, a water pool type suppression system is added to the ITER design. Its size of 650 m3 of
water at 30ºC is determined by a postulated break during baking.
TABLE III.3.4-1
Thermal-Hydraulic Parameter of ITER Water Coolant Loops
Tokamak Water Cooling System
Number
of Loops
Volume
Divertor/Limiter
1
(m3/loop)
170
First Wall/Blanket
3
Vacuum Vessel
2
Pressure
(MPa)
Temperature (°C)
Operation
Bakeout
4.3-5.7
100-150
240±10
140
3.0-5.7
100-150
240±10
160
1.5-5.2
100-110
200±10
III.3.5 CHEMICAL ENERGY SOURCES
The highest stored energy source in ITER is chemical, including the following sources:
•
Hydrogen inventories in various systems connected to the tritium plant;
•
Potential reactions of Be, C, or W with air or steam inside the vacuum vessel at
elevated temperature, releasing energy. (The C-steam reaction is endothermic.)
The steam reactions will also form hydrogen with potential for delayed chemical
energy release if H2-air mixtures form co-incident with an ignition source. Dust
explosions are also conceivable [Gae94] and dust inventory limits are set to avoid
this hazard; and
•
Formation of ozone in liquid/frozen air or nitrogen inside the Cryostat with
potential for delayed chemical energy release if the molar concentration reaches
values above four percent. Therefore no liquid nitrogen is used inside the cryostat.
The following subsections describe the nature of these chemical energy sources, the stored
energy, and (where relevant) the rate of energy release.
III.3.5.1
Hydrogen
There are potential hydrogen sources from chemical reactions involving in-vessel walls and
dust with steam; these are discussed in more detail in the following subsections. The invessel inventories and any hydrogen produced from in-vessel chemical reactions should be
within the PSR limit of 4 kg-H2 after converting from deuterium and tritium to protium on a
mole basis [PSR00]. The 4 kg-H2 in-vessel limit was set to limit the maximum pressure from
hydrogen-air deflagrations inside the vacuum vessel below 200 kPa [Ise00] postulating local
explosive mixtures. Detonations loads from postulated clouds of hydrogen and air will be
GSSR
page III-35
ITER
G 84 RI 3 01-07-13 R1.0
larger but here the relevant parameter is the time integrated momentum transfer to the
structures. The maximum time integrated momentum load per area was determined as 3200
Pa·s and the impact time is about 2 ms ([Ise00], see also Volume VIII). An additional source
of hydrogen can come from interaction between lithium, lithium-lead and Be-pebbles used in
the Test Blanket Modules (TBMs). Here the safety approach is to limit the inventories of
lithium and lithium-lead such that the maximum hydrogen production from each of these
sources stays below 2.5 kg (see also [PSR00]).
Table III.3.5-1 lists the hydrogen inventories in the ITER fuel cycle for a typical plasma
operation. No specific ignition sources have been identified so far. However, whenever an
explosive mixture is reached in any transient analysis the presence of an ignition source is
postulated.
TABLE III.3.5-1
Hydrogen Inventories (P, D, and T) during plasma operation
System
Isotope Separation System (ISS)
Storage and Delivery System (SDS) (gas)
Storage and Delivery System (SDS) (hydride)
Tokamak Exhaust Processing (TEP)
Water detritiation
Others: Gas supply lines, mechanical pumps and lines
Subtotal, tritium plant
Protium, Deuterium, and Tritium
Inventory
mole
520
650
320
60
60
<20
1630
Long Term Storage (LTS) (hydride)
200
Neutral Beam Injectors (all 3 and diagnostic NBI)
Co-deposited tritium and deuterium
Torus pumping
Fuelling
Subtotal, in-vessel
TOTAL
370
150
40
10
570
2390
III.3.5.2
Beryllium reaction rates
Be-water reactions are highly exothermic:
Be + H2O ==> BeO + H2 - 370 kJ/mole of Be reacted with water
This is 41 GJ/ton; times 13 tons of first wall beryllium (Section III.2.3) giving a stored
energy of 500 GJ. Be-steam reactions and the subsequent H2 production is also a major
concern for ITER safety at the rate of 0.22 kg-H2/kg-Be reacted.
The exothermic character of the Be-steam reactions leads to the possibility of self-sustained
Be-steam reactions. The potential to degrade to a self-sustained reaction in case of accidental
conditions and how this is prevented in the ITER design are assessed under the accident
analysis in Volume VII and VIII
GSSR
page III-36
ITER
G 84 RI 3 01-07-13 R1.0
The magnitude of the beryllium hazard depends on the reaction rate and the geometry. The
geometry determines heat sinks and heat transfer; ITER has significant beneficial heat sinks
that are not present in reaction rate experiments so that one has to be careful in interpreting
reaction rate experiments directly. Instead, the rate must be extracted and then used in
geometry-specific models. The reaction rate is a strong function of the temperature and
pressure as well as the physical condition of the beryllium. The Be-steam reaction rate is a
strong function of porosity.
Tests have been done with fully dense Be [And96c], 88% porous Be [Smo92], and irradiated
Be [And97a]. Figure VIII.3.5-1 summarizes the data and shows the interpretation for ITER.
The 88% porous Be was the most reactive form of Be, about 1-3 orders of magnitude more
reactive than the fully dense Be over the test temperature range (500-700°C). The main
motivation for the porous Be tests was to investigate porous Be for use as a multiplier inside
breeding blankets. Porous Be provides a lower thermal conductivity which is desirable for
the blanket beryllium to increase the operating temperature to enhance tritium release. The
lower thermal conductivity beryllium had interconnected porosity, explaining the large
differences in reaction rates for the porous and dense samples. The reactivity of 94% dense
plasma-sprayed Be was between the fully dense and porous, about a factor of 100 higher than
dense samples at lower test temperatures (400-600°C) and similar to dense samples at higher
test temperatures (900°C and above) [And97b].
Irradiation, gas production (via hydrogen and helium production), and then swelling is
another potential mechanism to increase porosity. Tests with irradiated beryllium show no
enhanced reaction versus unirradiated samples at and below 600°C [And97a]. The fluence
received by the test samples was estimated to be larger than for ITER conditions and the He
production was comparable. Enhanced reaction was noted for tests at 700°C. In these
experiments, the tritium and helium release behavior indicated that the specimen swelled and
developed a surface-connected porosity network prior to steam exposure. In this test
geometry, the exothermic heat of reaction increased the sample temperature to 1000°C or
above, further accelerating the reaction [And97a]. Similar behavior has been noted by 3H
mobilization experiments performed on irradiated Be samples [Bal91].
Below a specific temperature, depending on the form of Be (dense, plasma-sprayed,
irradiated, etc.), the reaction rate of Be in steam was observed to be parabolic as a function of
time [And96c], [And97a], [And97b]. The reaction rates are linearized over the test time for
use in safety assessments. This is generally conservative because the test times are usually
shorter than the accident time.
The baseline Be-steam reaction rate data are described in detail in [McC97a] and is shown in
Figure III.3.5-1. The first wall Be-steam reaction rates are expected to be close to those of the
dense Be samples as long as temperatures stay below 500°C. Above 600°C, we have adopted
a baseline reaction rate as the geometrical mean of the fully dense and 88% dense Be to
account for possible porosity and irradiation effects until enough irradiated data are available
for a new correlation. A linear interpolation is used between 500°C and 600°C. A safety
factor of 2 is used for Volume VII safety analyses providing reasonable margin against
unexpected new reaction rate data. For Be dust-steam reactions, a correlation based on a
curve fit over data at and below 600°C is used [McC97a]. The pressure dependence is nearly
linear (~p0.9) in the pressure range between zero and 200 kPa based on a literature survey
[Boi94a], [Boi94b].
GSSR
page III-37
ITER
G 84 RI 3 01-07-13 R1.0
Solid beryllium weakly interacts with air at 800°C; with significant weight gain at 900°C
[Dav96]. At 1000°C, beryllium precipitated from the vapor phase (simulating plasma
sprayed beryllium) has an reaction with air, abruptly rising because of the enhanced porosity
[Dav96]. Beryllium powder, tested to simulate beryllium dust, is more reactive. The percent
of 14-31 µm beryllium power reacting over several hours was 1-2% (500°C), 9-86% (800°C),
and 94-98% (1000°C). At 800°C, 9-13% reacted in the first half hour, reaching 23-30% in
the following 3-4 hours. 5 hours into the experiment, a rise in the reaction rate was observed,
and the measured fraction of powder reacted was 75-86% [Dav96]. Temperatures in air
ingress events are well below these values, thus little oxidation is expected.
Figure III.3.5-2 shows the nominal hydrogen production rate for solid beryllium, carbon, and
tungsten as a function of temperature.
III.3.5.3
Carbon chemical reaction rates
Carbon-steam reactions are endothermic and therefore not prone to self-sustained reactions.
However, they do pose potential for producing H2 at the rate of 0.17 kg-H2/kg-C reacted.
Figure III.3.5-2 shows the H2 production rate. The baseline C-steam reaction rate data are
described in more detail elsewhere [Smo90].
An air ingress might lead to potential self-sustained carbon-air reactions (in short-term, non
decay heat driven sequences) but several studies have shown that for the temperature ranges
expected for ITER such a scenario is not credible [Jah90].
III.3.5.4
Tungsten chemical reaction rates
W-steam reactions pose some potential for producing H2. Figure III.3.5-2 shows the H2
production rate, at 0.033 kg-H2/kg-W. The baseline W-steam reaction rate data are described
in more detail elsewhere [Smo96b]. Typically, production of H2 from W-steam reactions is
less important than Be-steam or C-steam because the hydrogen production per kilogram
reacted is 5-7 times lower and the buildup of tungsten dust is limited by confinement
concerns.
W-air reactions are not a significant energy source as the reaction and reaction rates are
modest [Smo96a].
III.3.5.5
Ozone formation
A potential chemical and energy hazard in cryogenic systems is the formation of ozone in the
presence of radiation fields [Bre89], [Boi94]. Ozone is formed from oxygen by electric
discharge, radiation, or chemical reaction. Thus, ozone can be formed in solid air frozen on
the cold surfaces resulting from previous air leaks into the Cryostat. At low temperatures
(< 90 K), the ozone yield is directly proportional to the irradiation time and dose rate. The
G(O3) value (the number of molecules produced for 100 eV of energy absorbed) is about
12.5, for example, in liquid nitrogen with oxygen impurities.
Ozone has a very high oxidation potential and is unstable. It decomposes by the reaction 2
O3 -> 3 O2, which is exothermic (144 kJ/mole of ozone) [Boi94]. Experience has shown that
at low temperatures ozone is prone to detonation. The ozone detonation velocity is 6000 m/s,
and the stored energy is 2.94 kJ/g-ozone. The explosion limit of ozone depends on
concentration, temperature, pressure and the presence of contaminants or catalysts. The
higher the ozone concentration, the lower the energy that is needed to initiate detonation.
GSSR
page III-38
ITER
G 84 RI 3 01-07-13 R1.0
Gram quantities of ozone could be an explosion hazard if these quantities concentrate and
detonate. When ozone is in a solid phase it can not migrate and the risk of high concentration
and explosion is small. In case of melting of the frozen air formed on the cold magnet
surfaces at some conditions, like current quench, a liquid air with ozone could concentrate in
the bottom of the Cryostat. In liquid air, a hazardous situation may develop when the ozone
concentration reaches about 4 mol% [Boi94].
The key uncertainties concerning ozone hazards in ITER are:
• The number of ozone molecules produced in frozen air in case of absorption of
definite amount of energy;
•
The pressure and temperature conditions in the Cryostat;
•
The circumstances by which an ozone enriched phase may appear.
The main source of ozone formation could develop due to an air leak into the Cryostat
resulting in the formation of solid air on the magnets. A potentially hazardous inventory of
ozone is avoided via monitoring accumulation of frozen air in the cryostat [Top95].
The operating limit for leakage of air into the cryostat is ~0.1 Pa-m3/s. About 50 g of ozone
could be produced in ~ 40 kg of frozen air during one year of operation at 10% plasma
operation time and 40% magnet surfaces facing to the Vacuum Vessel [Top95]. This is a
stored energy of 0.15 MJ. Leaks larger than 0.1 Pa-m3/s (or 1.2 mg/s) would be detected and
repaired. To set this into perspective it should be noted that the cryostat design leak is only
1.2*10-7 g/s. The minimum detectable air leak into the cryostat has been estimated as
6*10–6 /s. With such a leak rate the ozone production in one year of typical ITER operation
(3000 plasma shots per year) would amount to 50 mg.
III.3.6 SUMMARY OF ENERGY SOURCES
The amount of energy, time scales of concern for its release, and the potential consequences
are shown in Table III.3.6-1. This table also indicates the way the project intends to control
these energy sources and prevent them from becoming a driving force in accident scenarios;
more details are in other volumes.
GSSR
page III-39
ITER
G 84 RI 3 01-07-13 R1.0
TABLE III.3.6-1
Energy Inventories and Concerns for Release
Energy
Sources
Fusion
power
Plasma
Amount of
Energy
500 MW times
10 s =
5 GJ
Time Scale
for Release
Concerns
~10 s
Overheating of plasma
facing components;
In-vessel water ingress
0.7 GJ
<1 s
Magnetic
50 GJ
Decay heat
for ITER
FW and
Shield
Blanket
130 GJ in the first
day
330 GJ in first week
Chemical
energy
Wall (Be-steam):
500 GJ if react
seconds to
hours
Thermal
energy of
coolant
Baking - ~700 GJ
Normal operation ~300 GJ
seconds to
minutes
GSSR
seconds to
minutes
min. to years
depending
on concern
Disruptions;
Limited evaporation
and melting of plasma
facing material
Arcs; Quenches;
Localized magnet
melting;
Potential mechanical
damage
Heating of in-vessel
components;
Factor in waste
packaging
Overheating of
plasma-facing
components
H2 production
Overpressurization of
confinement barriers
Control
Possibilities
• Normal operation coolant
systems
• Active Fusion Power
Shutdown System
• Passive shutdown in case of
larger disturbances
• Plasma control
• Disruption mitigation by
impurity pellets
• Rapid quench detection and
discharge system
• Grounding scheme
• Minimize decay heat
• Several normal operation
coolant systems
• Passive decay heat removal
based on radiation heat
transfer and natural
circulation
• Limiting off-normal
temperature of plasma-facing
components
• Overpressure suppression
systems limit pressures in
confinement volumes
page III-40
ITER
G 84 RI 3 01-07-13 R1.0
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GSSR
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GSSR
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L.N. Topilski, 'Ozone formation at air leak into the cryostat', S 81 RI 2 95-03-27
F1, March 21, 1995.
[Vel00]
M. Velarde and J.M. Perlado, 'Disometric effects in the diffusion of the tritium in
reactors of nuclear fusion', presented at 14th TOFE, Park City, Utah, USA,
October 2000.
[Wam92] W.R. Wampler, 'Trapping of Deuterium in Beryllium', Journal of Nuclear
Materials, Vol. 196-197, pp. 981-985, 1992.
[Wan97] W. Wang, W. Jacob, J. Roth, 'Oxidation and hydrogen isotope exchange in
amorphous, deuterated carbon films', J. Nuc.Mat 245 (1997) 66-71.
[You97]
GSSR
D. You, S. Lefervre, P. Gerlinger, and A. Cheniere, 'Thermodynamik stability and
solubility of copper and copper oxides', CEA, France, LPCC-15662, RT-SCECF
420 (May 1997).
page III-46
ITER
G 84 RI 3 01-07-13 R1.0
Figures
Implantation
Trapping/ Detrapping
Permeation
Recombination
tritium atom
lattice atom
Diffusion
Traps or
lattice damage
Surface damage
and erosion
Plasma Implantation
Zone
Armor/Substrate
Coolant
FIGURE III.2.1-1
Implantation of Tritium into Plasma-Facing Material
Tritium
Carbon
Plasma
C erosion
T co-deposition
C redeposition
Original surface
Accumulated co-deposited layer
Eroded surface
FIGURE III.2.1-2
Mechanism of Tritium Co-Depositing with Carbon
GSSR
page III-47
ITER
G 84 RI 3 01-07-13 R1.0
10 20
FIGURE III.2.1-3
ITER-FEAT poloidal cross-section showing vacuum vessel (inner contour), divertor
vertical target (VT), divertor baffle (B), and divertor private region consisting of
dome (D) liner (L), and inner (IW) and outer (OW) first-wall. Start-up limiters (2
modules) are located at the equatorial level. SOL magnetic surfaces including
separatrix are shown (numbers are distance in metres along the poloidal coordinate).
a = 0.0m
b = 3.2m
c = 8.0m
10 17
d = 13.7m
e
e = 18.5m
b d
10 11
10 14
Γ(E)
c
a
100
10 1
10 2
Energy (eV)
10 3
FIGURE III.2.1-4
Energy spectra Γ (E) of the charge-exchange neutral flux at different poloidal
locations (see also Fig. III.2.1-3).
GSSR
page III-48
G 84 RI 3 01-07-13 R1.0
10 3
10 22
ITER
10 0
1018
10 19
Γ
10 1
Γcx(1/m2s)
10 20
10 2
Emean(eV)
10 21
E
0.
2.
4.
6.
8.
10.
12.
Distance (m)
14.
16.
18.
10 0
FIGURE III.2.1-5
Total CX flux Γ and Emean versus poloidal distance (figure III.2.1–3), ne= 3x1019 m-3.
Net erosion rate (nm/s)
10 -4
10 -2
Be
10 -6
W
ions
neutrals
total
0.
2.
4.
6.
8.
10.
12.
Distance (m)
14.
16.
18.
FIGURE III.2.1-6
Sputtered erosion rate (from DT neutrals and DT and impurity ions) versus poloidal
distance (Fig. III.2.1-3, divertor at 0.0 m and 18.0 m) for Be, W.
GSSR
page III-49
ITER
G 84 RI 3 01-07-13 R1.0
10000
1000
100
10
1
10
100
1000
10000
Number of Pulses (400 s)
FIGURE III.2.1-7
Comparison of the rate of tritium accumulation in the beryllium first-wall
including breeding (curve (a) without n-effect; curve (b) with 0.1% traps; curve
(c) with 1% traps) and T-co-deposition for two indicative co-deposition rates of 1
g/pulse (curve (d)) and 10 g/pulse (curve (e)). Each pulse is assumed to be 400 s.
GSSR
page III-50
ITER
G 84 RI 3 01-07-13 R1.0
1
air
0.8
H3 release
fraction
0.6
ITER-safety analysis
T-10 (32 hours)
TFTR (30 minutes)
TFTR-10 h / 2.1 kPa O2
JET 10h/2.1 kPa O2
D-III-D 10h/2.1 kPa O2
D-III-D 2h/21 kPa O2
AU (2 hours)
0.4
0.2
0
0
100
200
300
400
500
T [ºC]
1
0.8
H3 release
fraction
0.6
ITER-safety-analysis
T-10 (32hours)
TFTR 7 hours/2.1 kPa
0.4
0.2
0
0
100
200
T [ºC]
300
400
500
FIGURE III.2.1-8
Tritium mobilization data from codeposited layers formed in tokamaks and the
derived assumtion used in ITER safety analysis. Athmospheric air/steam
pressures were used during mobilization experiments unless stated otherwise.
The times given are indicative; most of the mobizisation occurs fast. The
following references were used: T-10 [Rom01], TFTR-30minutes [Cau90], TFTR
-10 hours/JET/D-III-D [Dav98], [Haa98], AU [Wan98]
GSSR
page III-51
ITER
G 84 RI 3 01-07-13 R1.0
100
[%]
Air
80
60
40
GSSR-assumption-1
GSSR-assumption-2
T=200ºC, T-10 Sample1
T=200ºC, T-10 Sample2
T=300ºC, T-10 Sample1
T=300ºC, T-10 Sample2
20
0
0
5
10
15
20
time [h]
100
[%]
Steam
80
60
GSSR-assumption-1
40
GSSR-assumption-2
T=200ºC, T-10 Sample 1
T=200ºC, T-10 Sample 2
20
T=300ºC, T-10 Sample 1
T=300ºC, T-10 Sample 2
0
0
5
10
15
20
time [h]
FIGURE III.2.1-9
Kinetics of tritium mobilization in atmospheric air and steam from co-deposited
samples collected in the T-10 tokamak at 200ºC and 300ºC. The two lines show the
ITER safety analysis assumptions. Whatever is the more conservative mobilization
scenario is chosen in the accident analysis.
GSSR
page III-52
ITER
G 84 RI 3 01-07-13 R1.0
2.5
[g-T]
2
1.5
1
0.5
0
0
2
4
6
8
10
time [a]
FIGURE III.2.1-10
Tritium permeation from the first wall to the coolant. The fluence of 0.5
MWa/m2 is distributed uniformly over 10 years. Different operational states are
lumped together in each year for this calculation. Significant permeation occurs
at baking only. For periods without baking the tritium concentration decreases
slightly due to radioactive decay.
GSSR
page III-53
ITER
G 84 RI 3 01-07-13 R1.0
Dust Production Rates
(conservative)
200
180
160
140
120
100
80
60
40
20
0
carbon
tungsten
beryllium
carbon with flakes
0
200
400
600
800
1000
No. of Pulses
FIGURE III.2.2-1
Estimated dust production rate as a function of number of cycles during operation.
(fraction of vaporized, eroded, or sputtered material: 0.3)
10
CMOD
DIIII-D
SIRENS Cu
SIRENS W
JET
1
SIRENS
316 SS
SIRENS Al
TFTR
ITER Safety Analysis Specification
0.1
FIGURE III.2.2-2
Comparison of Dust Size Distribution Data versus GSSR Analysis Specification
GSSR
page III-54
ITER
G 84 RI 3 01-07-13 R1.0
Percent mobilized
100%
80%
60%
40%
20%
0%
0
10
20
30
40
Air pressurization rate (kPa/s) in test apparatus
FIGURE III.2.2-3
Dust Mobilization Experimental Data
NEUTRON FLUX
Ions in solution
Fe
Suspended particulates
Fe3O4 or
complex ferrite
Mn
Co
Deposits ("internal", less
subject to mobilization) Boundary layer
Deposits ("external",
prone to mobilization)
thickness
e
Fixed oxide (not mobile)
Metal ions through
oxide porosity
Piping base metal
FIGURE III.2.2-4
Basic Corrosion Model in the PACTOLE Code
GSSR
page III-55
ITER
G 84 RI 3 01-07-13 R1.0
Elemental equilibrium solubilities
[kg/kg-water]
10
10
10
10
Fe; pH=7.1
Fe; pH=5.7
Ni; pH=7.1
Ni; pH=5.7
Co; pH=7.1
Co; pH=5.7
Cr; pH=7.1
Cr; pH=5.7
Mn; ph=7.1
-6
-8
-10
-12
0
50
100
150
200
250
Temperature [C]
300
350
Early dose (mSv), per hour at
temperature, per m2 exposed area
FIGURE III.2.2-5
Equilibrium Solubilities as Function of Temperature
(Bold = baseline water chemistry with pH300ºC =5.7 (pH20°C=7.0))
1E+00
Total
1E-01
W
1E-02
1E-03
Re
1E-04
Ta
1E-05
1E-06
1E-07
1E-08
1E-09
400
500
600
700
800
900
1000
Temperature (°C)
FIGURE III.2.2-6
Early Dose [mSv] from Tungsten, Conservative Weather, Ground-Level Release,
Steam, 0.3 MW-a/m2 Fluence, No Credit for any Radioactivity Confinement
GSSR
page III-56
ITER
G 84 RI 3 01-07-13 R1.0
a)
C-E/E
0
ACT-4/ FENDL/A-2
ANITA-4M/ FENDL/A-2
-0.05
DKR-PULSAR2.0/ FEND L/A-2
FISPACT/ FEND L/A-2
REAC-3/ FENDL/A -2
-0.1
0
1
2
3
4
time, days
5
6
7
b)
0.1.
C-E/E
.
ACT-4/ FENDL/A-2
ANITA/ FENDL/A-2
0
DKR-PULSAR2.0/ FENDL/A-2
FISPACT/ FENDL/A-2
REAC-3/ F ENDL/A-2
.
-0.1.
.
0
1
2
3
4
5
6
7
time, days
FIGURE III.3.3-1
Comparison of calculated and experimental decay heat in a) SS316 and b) copper for a
14 MeV n-flux.
GSSR
page III-57
ITER
G 84 RI 3 01-07-13 R1.0
1.E+02
1 0 0 0oC
800oC
700 o C
6 0 0oC
500oC
4 0 0oC
1.E+01
1.E+00
1.E-01
[l-STP/m2-s]
1.E-02
INEEL-correlation-dense
INEEL-correlation-porous
ITER-FW-correlation
ITER-FW-safety
INEEL-1992-dense-Be
INEEL-1996-dense-Be
1.E-03
1.E-04
1.E-05
INEEL-1996-irradiated-Be
INEEL-1992-88%porous-Be
TRW-dense-Be
1.E-06
1.E-07
6
8
10
12
14
16
Inverse temperature [1/K*10000]
Hydrogen production rate (kg/m 2 -s)
FIGURE III.3.5-1
Measured hydrogen production rates for steam-beryllium chemcial reactions and
interpretation for the ITER FW. Solid beryllium will be used for the FW plasma facing
material. Up to 500ºC irradiation effects have no influence on the reaction rates. For
larger temperature excursions, a geometrical mean between the solid beryllium reaction
rate data and 88% porous reactions rate data is used, because maximum swelling of
beryllium in ITER conditions is expected to be less than 6%, yielding an expected open
porosity of 94%. Such swelling only occurs at temperatures above 600ºC.
10-2
10-3
10-4
10-5
Be
C
W
10-6
10-7
10-8
10-9
10-10
10-11
400
600
800
1000
1200
1400
Temperature (°C)
FIGURE III.3.5-2
Hydrogen Production Rates for Beryllium (dense), Carbon, and Tungsten
GSSR
page III-58
ITER
G 84 RI 3 01-07-13 R1.0
Appendix 1: Dose calculations in support of GSSR
This appendix reports doses per unit release of tritium or activation products calculated for
ITER. Dose release calculations have been performed deterministically for three different
conditions:
-
ground level releases taking building wake effects into account,
release from the top of the building with high speed exhaust
from a 100 m height stack without building wake
The dimensions of the tokamak building are 51 m width, 80 m length, 56 m height. The
baseline exhaust point would be a vertical pipe or duct at the NE corner of the tokamak
building, extending a few meters (4 meters) above the height of the top of the building (which
is ~56 m). The diameter of this single point release vent duct is about 1.8 m, and the velocity
is about 20 m/s. The flow rate depends on the circumstances, ranging from 55,000 Nm3/hr to
as high as 200,000 Nm3/hr. The velocity would be maintained by use of a discharge blower.
The general framework of dose-release calculations follows [Pie96]. For off-normal releases
the following two dispersion parameter sets have been used (release duration >1 hour):
-
MOL [Bul72] (a conservative representation of EU HT dispersion parameters);
P-G (Pasquill-Gifford) (official US dispersion set used by NRC, EPA, see e.g., [Jow90]);
The following three types of off-normal doses have been calculated: early dose (7 days,
inhalation including skin absorption, groundshine, cloudshine), chronic dose without
ingestion (50 years exposure), chronic dose with ingestion.
For normal operational releases the European Cadarache site weather statistics was used.
Normal operation doses are chronic doses including all pathways.
The site boundary is assumed at 1 km.
The source terms considered are listed in Table III.A1-1.
Table III.A1-1: ITER source terms
Release condition
Normal operation
Normal operation
Normal operation
Normal operation
Off-normal
Off-normal
Off-normal
Off-normal
Acronym
SN1
SN1b
SN2
SN3
SO1
SO1b
SO2
SO3
Type
Tritium
Tritium
ACP
Tungsten dust
Tritium
Tritium
ACP
Tungsten dust
Quantity
1 g/year
1 g/year
1 g/year
1 g/year
1g
1g
1g
1g
Comment
HTO
HT
Specific activity Table III.E-2, 7th column
Specific activity, Table III.E-2, 3rd column
HTO
HT
Specific activity Table III.E-2, 6th column
Specific Activity Table III.E-2, 3rd column
The results are documented in more detail in [Ras01]. A summary of the most important
results is given below. In a first set of investigations, the relevant worst case weather
conditions for each of the combinations of release height and dispersion parameter set were
calculated.
GSSR
page III-59
ITER
G 84 RI 3 01-07-13 R1.0
The computer program UFOTRI [Ras90], [Ras93] has been used for the dose assessment of
tritium releases. Processes such as the conversion of tritium gas (HT) into tritiated water
(HTO) in soil, reemission after deposition and the conversion of HTO into organically bound
tritium (OBT) are considered. During the time period of the first few days, all relevant
transfer processes between the compartments of the biosphere (atmosphere, soil, plants,
animals) are described dynamically. A first order compartment model calculates the longer
term pathways of tritium in the foodchains.
Calculations for released activation products were performed with the version NL/95 of the
program system COSYMA [Cos91], including extended data sets for activation products
[Hey96]. It was assumed that the nuclides appear in aerosol form with a mean diameter of 1
micro-m AMAD, and the corresponding dry deposition velocity is set to 1 mm/s. The doses
by ingestion of contaminated foodstuffs are calculated assuming the local production and
consumption method; that means, all foodstuffs are consumed in the grid element where they
are harvested/produced. The foodchain information from the German model ECOSYS has
been used in the calculations [Hey96].
It should be noted that in case of the MOL dispersion parameter set, the reference height for
the wind speed was always set to 10 m. The wind speed in release height is calculated
according to a power law approach implemented in UFOTRI and COSYMA. In case of the
P-G dispersion parameter set, the wind speed indicated is the wind speed in the release
height. Only in case of the jet release, the no-rain case was applied with the reference height
of 10 m and the wind speed was adjusted to the release height. The precipitation intensity
was assumed to be 1 mm/h.
Table III.A1-2a-d summarizes the resulting dose per unit release factors. For off-normal
releases the most conservative results are listed. The atmospheric dispersion is often
described by the 'Chi/Q' value (if multiplied by a release rate, one gets the atmospheric
concentration). This value is not meaningfull for off-normal releases since the doses to the
public are typically dominated by rain (wash-out) and the subsequent wet deposition.
Therefore Chi/Q values are only given for normal operation releases for HT and HTO in
Table III.A1-3. The small differences espacially for ground level releases are due to the
larger deposition velocity (into the soil) for HTO.
GSSR
page III-60
ITER
G 84 RI 3 01-07-13 R1.0
Table III.A1-2a: Normal operation doses per unit release (1 g/year) at 1 km for a
generic site [micro-Sv]
Source Term
Tritium as HTO, 1 g/year
Tritium as HT, 1 g/year
Tungsten, AP, 1 g
ACP, 1 g
60 m exhaust
"jet release" 20 m/s
Ground level
100 m stack
no building wake
2.67
0.0952
0.0603
0.0383
26.6
1.66
0.759
0.468
3.01
0.107
0.0764
0.0499
Table III.A1-2b: Off-normal doses per unit release (1 g) at 1 km for a generic site,
Chronic dose without ingestion [micro-Sv]
Source Term
Tritium as HTO, 1 g
Tritium as HT, 1 g
Tungsten, AP, 1 g
ACP, 1 g
60 m exhaust
"jet release" 20 m/s
Ground level
100 m stack
no building wake
83.1
0.81
16.1
16.7
637
5.92
20.5
20
42.8
0.182
16.3
16.7
Table III.A1-2c: Off-normal doses per unit release (1 g) at 1 lm for a generic site,
Chronic dose with ingestion [micro-Sv]
Source Term
Tritium as HTO, 1 g
Tritium as HT, 1 g
Tungsten, AP, 1 g
ACP, 1 g
60 m exhaust
"jet release" 20 m/s
Ground level
100 m stack
no building wake
1890
38.2
61.5
28.6
2980
207
74.2
34
1840
15.5
61.8
28.6
Table III.A1-2d: Off-normal doses per unit release (1 g) at 1 km for a generic site,
Early dose [micro-Sv]
Source Term
Tritium as HTO, 1 g
Tritium as HT, 1 g
Tungsten, AP, 1 g
ACP, 1 g
60 m exhaust
"jet release" 20 m/s
Ground level
100 m stack
no building wake
83.1
0.8
13.2
0.142
636
5.8
17.1
0.491
42.8
0.176
13.5
0.145
Table III.A1-3: Chi/Q [s/m3] values for normal operation effluents at 1 km
Source Term
Tritium as HTO
Tritium as HT
GSSR
60 m exhaust
"jet release" 20 m/s
Ground level
100 m stack
no building wake
3.48E-07
3.49E-07
3.71E-06
5.71E-06
4.11E-07
4.14E-07
page III-61
ITER
G 84 RI 3 01-07-13 R1.0
References
[Bul72]
[Cos91]
[Hey96]
[Jow90]
[Pie96]
[Ras90]
[Ras93]
[Ras01]
GSSR
H. Bultnyck and L.M. Malert, 'Evaluation of Atmospheric Dilution Factors for
Effluents Diffused from an Elevated Continuous Point Source", Tellus XXIV,
pp 455-472 (1972).
COSYMA: A new program Package for Accident Consequence Assessment.
A Joint Report of KfK and NRPB, Commission of the European
Communities, Report EUR-13028 EN (1991).
S.M. Heywood, J. Brown, J.A. Jones, C. Fayers, J. Smith, G. Pröhl, M.
Bleher, P. Jacob, and H. Müller, 'Databases for activities in foodstuffs, for
external exposure from the ground and for dose per unit intake, for fusion
radionuclides for input to the COSYMA ACA system', National Radiological
Protection Board, Chilton UK, Report NRPB-M634, 1996.
H.-N. Jow et al., 'MELCOR Accident Consequence Code System (MACCS)',
NUREG/CR-4691, Sand86-1562, Vol.2, Sandia National Lab, 1990.
S. Piet, 'Input Parameters for Off-site Dose Calculations', ITER San Diego
JWS, S 81 MD 81 96-11-25, November 25, 1996.
W. Raskob, 'UFOTRI: Program for Assessing the Off-Site Consequences from
Accidental Tritium Releases', Report KfK-4605, Kernforschungszentrum
Karlsruhe, Germany, 1990.
W. Raskob, 'UFOTRI: Description of the New Version of the Tritium Model
UFOTRI 4.0, including user guide', Report KfK-5194,
Kernforschungszentrum Karlsruhe, Germany, 1993.
W. Raskob and I. Hasemann, 'Deterministic calculations for source terms from
ITER-FEAT', ITER Task G81TD05 (D452-EU), Subtask-4, Forschungszentrum Karlsruhe GmbH, Germany, May 2001.
page III-62