Development of Radionuclides Source Term for Spent Fuel in Geological
Disposal. Major Outcomes of the European Projects « In Can Processes » and
« Spent Fuel Stability »
C. Poinssot, JM. Cavedon,
Commissariat à l’Energie Atomique, Nuclear Energy Division, CEA-SACLAY, France
M. Cowper*,
AEAT Technology, Harwell, United Kingdom
B. Grambow
Ecole des Mines de Nantes, SUBATECH laboratory, France
T. McMenamin,
European Commission, Brussels
Summary
Direct disposal of spent nuclear fuel is a reference option for many European countries and is
also studied as an alternative option by the countries which currently reprocess their spent
fuel. Therefore, having an European reference knowledge on the evolution of spent fuel in
geological disposal is a major issue for the credibility and acceptance of nuclear energy. In the
framework of the 5th FWP, the European Commission supports two major research projects
dealing with the evolution of spent nuclear fuel in direct geological disposal:
•
•
First project entitled “Rates and Mechanisms of Radioactive Release and Retention
inside a Waste Disposal Canister” (referenced FIKW-CT-2000-00019), further
referred to as ICP.
Second project entitled “Spent Fuel Stability under Repository Conditions”
(referenced FIKW-CT-2001-00192), further referred to as SFS.
Their aims are (i) to identify the processes controlling the spent fuel alteration in the near-field
environment of a geological disposal, (ii) to develop robust and shared model to predict the
RadioNuclides (RN) release as a function of time. These models are supposed to be some
reference input data for any further national geological disposal performance assessment
exercises.
The aims of this paper is to depict the current major outcomes of these two projects in terms
of major scientific results and spent fuel performances in geological disposal conditions.
1. Presentation of the two EU projects
1.1 The ICP project
Launched in 2000, the first project was entitled “Rates and Mechanisms of Radioactive Release and
Retention inside a Waste Disposal Canister”. It finished in August 2003. It was coordinated by
*
present address : Serco Assurance, Risley, United Kingdom.
AEA Technology (M. Cowper) and involved 8 partners with a total budget of ~1384 k€: AEA
technology (40.6%), SKB (6.3%), VMO Konsult (6.0%), Chalmers Technical University (13.4%),
Stockholm University (7.5%), Uppsala University (6.2%), POSIVA OY (1.0%), VTT Chemical
Technology (19.1%). The overall objective of the programme was to improve the scientific
understanding of the processes that control release of radioactive species from spent fuel inside a
disposal canister and the chemical changes in those species that might limit release of radioactivity
from the canister. The programme studied the possible hydrogen overpressure effect, the potential
retention capability of the corrosion products and the influence of the corrosion products on the
whole redox balance within the container. The experimental work was carried out used both
simulated materials (including α-doped UO2 with various doping levels simulating old fuels) and
spent fuel itself. An isotope dilution method was developed and used to measure very low
dissolution rates [1]. More fundamental approaches were used to understand the chemical reactivity
of Fe(II) towards U(VI) and Np(V) by experimental measurements including the development and
use of Resonant Inelastic soft X-ray Scattering (RIXS) spectroscopy [2]. These data were supported
by modelling using quantum mechanics calculations [3].
1.2 The SFS project
Launched in 2001, the second project is entitled “Spent Fuel Stability Under Repository
Conditions”. It is managed by CEA (JM.Cavedon & C.Poinssot) and involves 13 partners for a total
budget of ~2725 k€ : CEA (15%1), ANDRA (8%), SUBATECH-ARMINES (6%), SKB (3%),
ENRESA (5%), CIEMAT (6%), ENVIROS (3%), Universitat Polytècnica de Catalunya (1%),
NAGRA (4%), ITU (25%), FZK (15%), SCK-CEN (7%) and Studsvik (2%). Its aims is to get a
reliable and shared RN source term able including the instantaneous release of RN at the water
ingress (Instant Release Fraction, IRF) and the much slower release governed by the dissolution of
the fuel matrix [4]. Focus is put on (i) the potential increase of the IRF with time due to an intrinsic
evolution of radionuclide distributions in the spent fuel before water ingresses, and (ii) the
radiolytic dissolution process of the matrix for which no generally agreed alteration model is yet
available. Experiments on simulated samples as well as irradiated UOX and MOX fuels are carried
out in both saline and argillaceous conditions. Significant focus is put on hydrogen affecting the
radiolytic fuel dissolution. This project will end in November 2004 and therefore, only partial and
preliminary results will be given.
2. Assessment of the Instant Release Fraction
2.1. Model of RN release
Depending on the heterogeneous distribution of radionuclides in the structure of spent fuel, their
release to groundwater is classically described by the contribution of two terms [5,6]:
•
•
1
A fast release of radionuclides which are not contained in the fuel matrix and which
are released more or less instantaneously when the confinement is breached and water
ingress the canister. This contribution is referred to as the Instant Release Fraction
(IRF).
A progressive and relatively slow release of the radionuclides which are embedded
within the fuel matrix. This contribution is referred to as the Matrix Fraction.
Corresponds to the percentage of the total budget allocated to each partner
Performance assessment exercises demonstrate that in most of the scenarios, the long term
radiotoxic impact is dominated by the IRF although it is representing the smallest part of the total
inventory [5]. In most of the past studies, the IRF was determined from the experimental results
obtained on fresh irradiated fuels (i.e. < 15 y.) considered as representative for older fuels which
will be contacted by water in geological disposal [7]. Recent results showed that the radionuclide
distribution in the spent fuel may evolve before water ingress the canister and therefore the IRF may
be different than that from the post-irradiation one [8, 9, 10].
2.2 Evolution of spent fuel prior to the water access
The radionuclide distribution in the heterogeneous spent fuel structure does not represent an
equilibrium state after irradiation and furthermore there are potential driving forces which could
lead to a significant radionuclide redistribution with time [8]. Among others, the importance of the
alpha decay has been evaluated. It leads to the formation of large quantities of helium within the
fuel pellet, the fate and consequences of which are not completely understood, and to the
accumulation of irradiation damages which are not recovered due to the relatively low temperature
[8, 10]. From the results of the studies performed by CEA [10], ITU and CRIEPI [12], the most
crucial issue seems to be for the long term:
• The fate of helium within the fuel pellet. Experimental results show that helium is more
mobile than fission gases and could therefore more easily diffuse within the fuel pellet
towards the grain boundaries, in geological time [8, 13].
• The occurrence of alpha-self irradiation enhanced diffusion (α−SIED) which potentially
contribute significantly to the mobility of the RN within the fuel pellet [6, 14, 15].
Conservative estimates for diffusion coefficient value are currently assumed to be in the
order of ~ 10-25 m2.s-1 after irradiation and decrease with decreasing alpha activity with
time[14, 15].
• Finally, there is currently a controversial debate on how to demonstrate the long-term
integrity of grains boundaries. Indeed, it seems difficult to allocate a strong long-term mass
transfer resistance to the grains boundaries due to their relative mechanical weakness after
irradiation, due to the segregation of numerous micro- or nano-bubbles and solid inclusions
which can cover up to 50% of the grains boundaries surfaces [10, 16]. On the other hand the
presence of high-pressurized fission gas bubbles indicates that grain boundaries are not so
easily accessible at relatively short term for mass transfer processes.
In the framework of the SFS project, a model was developed to account for most of these potential
processes and to get conservative quantitative assessment of the IRF inventory for the long term.
2.3. Model of IRF evolution for UOX fuel
A. IRF inventory soon after irradiation: IRF(time=0)
An IRF model has been developed by CEA and NAGRA within the SFS project by assessing the
respective confinement properties of the various microstructures present within the fuel pellet. The
fuel plenum, gap zone, fracture surfaces and grains boundaries are classically considered as nonconfining and the radionuclides (RN) within them are considered as part of the IRF. The RN present
within the highly porous rim zone of high burnup UOX fuel (> 40 GWd/t), have been quantified but
their allocation either to the IRF (no confinement properties – instant release) or the matrix
(confinement properties – delayed release) is controversial and correspond to two options of the
model. An exhaustive synthesis of all the available results on the RN location by both post
irradiation examination (PIE) and leaching experiments has been carried out [17].
B. IRF inventory evolution with time: IRF(∀time)
In order to assess the evolution of the RN redistribution with time, a Booth’s type model has been
developed assuming a α−SIED diffusion process from the grains to the grains boundaries and a
faster RN diffusion within the grains boundaries, i.e. a zero concentration boundary for mass
transfer from the grains to the grain boundaries [6, 14]. The model used the best estimate diffusion
coefficients for α−SIED and considered the precise RN location after irradiation previously
described [17]. Results for 55 GWd.t-1 UOX fuel are presented in the following table for different
dates of water ingress ranging from 1000 to 100000 y.:
Table 1: Bounding estimates of the IRF at t = 0 y, 1,000 y, 10,000 y and 100,000 y of key RN for a PWR UO2 fuel with a
burnup of 55 GWd/tIHM considering gap, rim and fractures within the IRF [15].
Nuclide
14
C
Cl
79
Se
129
I
135
Cs
36
Bounding IRF values (%)
After
Container failure time
irradiation
1,000 y
10,000 y
100,000 y
10
13
14
16
11
14
15
17
11
14
15
17
11
14
15
17
11
14
15
17
These results clearly evidence a much higher reference inventory for IRF than usually used in
performance assessment exercise (~5%). This is mainly due to the assumed unstable grain
boundaries, which yields also to consider the whole rim as part of the IRF. In addition, the
contribution of the α-SIED to the IRF inventory is relatively low (3-6%) as could be expected due
to the low diffusion coefficient (≤ 10-25 m2.s-1).
3. Intrinsic mechanisms of matrix dissolution and related performances
Both ICP and SFS projects are dealing with the radiolytic dissolution of matrix fraction. The goals
are both to accurately describe the elementary processes occurring at the fuel / water interface and
also to propose a more general matrix alteration model able to relate the fuel dissolution rate to any
accessible parameter like the residual alpha activity.
3.1 Qualitative and quantitative modelling of the radiolytic at the fuel / water interface
One of the first objectives of the SFS project was to get consensus based on all the available results
on the way the dissolution of spent fuel proceeds [18]. The irradiation of groundwater by alpha
particles from the spent fuel leads to the dissociation of water molecules along the alpha track
producing radiolytic reaction products. Part of these products are radical or molecular oxidants
which can oxidise the fuel surface resulting in the formation of U(VI) as evidenced by XPS results.
These U(VI)-species are subsequently release into solution as a function of the water chemistry.
-1
-1
Mass loss rate ( µg.L .h )
UO 2 /water
He
1000
2+
20 MeV
11
-2
3.3x10 cm .s
on
off
-1
500
-1
<1 µg.L .h
-1
0
-5
0
5
10
15
20
25
Time of water renewal (h)
Figure 1: Elementary processes occurring in the fuel
radiolytic dissolution in reducing conditions [18]
Figure 1: . Uranium mass loss rate per hour in aerated
deionised water before, under and after alpha
irradiation. Water is renewed each hour [19].
In order to quantify the production and consumption of radiolysis oxidants, several experiments
were conducted: (i) a UO2/H2O interface has been studied in a representative geometry with a
5MeV alpha particle flux supplied by a cyclotron (CEA, SUBATECH), (ii) external irradiation by
gamma source (60Co) or alpha solution (238Pu) (INE). The aim was to confirm the role of radiolysis
on long term spent fuel corrosion, to assess the influence of other chemical species on the radiolytic
reactions and to update the available radiolytic kinetic model. Focus was put on brines (one of the
German reference condition) and carbonate-rich water (relevant for argillaceous media like the
argillites).
Fig.2 shows the uranium mass loss rate obtained in one hour on the simulated interface with and
without irradiation [19]. The dissolution is clearly enhanced by alpha irradiation at high dose rate.
The precipitation of a hydrated uranium peroxide, studtite, was observed. Finally, experiments
performed out of irradiation but with similar H2 O2 have lower dissolution rate demonstrating the
common role of H2O2 and radicals.
New experimental results obtained on irradiated brines allowed to update the Christensen’s model
[20] (kinetic model) which has been implemented in the CHEMSIMUL or MAKSIMA code. New
reactions have been proposed to account for some additional reactions with contaminants like Fe 2+
and Ni2+.
3.2. Effect of decreasing radiation field: experiments on doped samples
In addition to the previous experiments, several sets of experiments on alpha doped samples have
been undertaken within the framework of both SFS and ICP projects with the aim of exploring the
relationship between the alpha dose rate and the dissolution rate. Activity was equivalent to ages
ranging between 150 and 500 000 y. with 233U as dopant.
1x10
-8
0.43
1x10
-9
0.04
1000
100
10
1
10
100
1000
10000
100000
1x10
1000000
Time, years
10
-10
0.00
233
U-UO 2(
233
U-10%)
238
U-UO 2(
233
U-10%)
238
U-UO 2
-11
U concentration (ppb)
U concentration (mol/Kg H 2O)
Dose rate (Gray/h)
10000
0.00
0
2000
4000
6000
8000
10000
Leaching time (days)
Figure 2 : Equivalent ages of the doped samples
prepared and studied within ICP and SFS project
(circle : SCK in SFS ; triangle : ITU in SFS ; diamond :
AEA in ICP ; square : experiments of Jegou et al. [21])
compared to the alpha decay of a 47 GWd.t -1 UOX fuel
(curve)
Figure 3 : U aqueous concentration in a bicarbonate
solution as a function of time for various doping levels :
1% 233U (~500 000y.), 10% 233U (3000y.) and no 233U.
Experiments were conducted in anoxic (Ar/CO2 atmosphere, Eh ~-50mV) or reducing (Ar/H2
atmosphere, Eh <-250mV, SCK•CEN and ITU) conditions in static and flow-through tests
(SCK•CEN). Most of the experiments are still ongoing. First results evidence a significant
influence of alpha doping level at least or the highest alpha activity although no definite conclusion
can be drawn (Fig.4). Several experiments seem to indicate that no obvious relationship can be seen
at low alpha activity, which could confirm the existence of a potential alpha activity threshold for
radiolytic dissolution below which radiolytic fuel dissolution ceases to be an effective dissolution
mechanism. In the absence of radiolysis, under reducing conditions, U(IV) solubility controlled fuel
matrix dissolution would dominate. Additional on-going experiments will allow a deeper
understanding of these processes.
3.3 Development of a Matrix Alteration Model (MAM) within SFS
One of the major objectives of the SFS project is to develop a generally agreed model able to
describe the long term fuel matrix alteration for future performance assessment exercises. The basis
of this model are: (i) modelling of the oxidants generation by a kinetic model of radiolysis [20]
considering that alpha radiation is dominating in the long term, (ii) oxidation of spent fuel surface
modelled by formal surface reactions, (iii) reduction of the aqueous oxidants (only molecular
species will be considered) and finally (iv) dissolution of spent fuel matrix and radionuclides release
according to the uranium speciation and groundwater composition [22]. G values and kinetic
constants for the radiolytic model will be those produced in SFS (INE).
The model is currently under development and should be achieved in mid-2004 for generic
calculations of spent fuel performance by the end of the SFS project.
In addition to this general approach, complementary investigations are conducted on the coupling
between electrochemical and geochemical models and on the comparison of radiolytic codes.
4. Coupling of the fuel dissolution with the near-field components
Near-field of a geological repository will be dominated by the large amount of metallic components
(canister or overpack materials), the corrosion of which will produce large amounts of hydrogen.
Hydrogen is also generated by radiolysis. The presence of hydrogen can significantly affect the
long-term alteration of spent fuel and have therefore to be accounted for. Experiments have been
conducted in both SFS and ICP project to assess these effects of hydrogen and of iron on spent fuel
stability.
4.1 Influence of hydrogen on fuel corrosion
Experiments on real spent fuel as well as doped pellets have been conducted in the SFS project in
presence of hydrogen to assess its potential inhibitor role (ITU, SKB, INE). Some of the results are
presented in the next figure:
1E-4
Cs
Static Test under
external H 2 Overpressure 3,2 bar
915 days
Conc. Mol/l
Sr
Reference test:
Pellet K4+Fe
2.7 bar H2
overpressure
1E-6
U
Tc
1E-8
Np
Pu
1E-10
Np (close to DL)
pH
4,9
1E-12
300
pH
6,5
pH
7,7
pH 6,3
pH 6,3
600
900
Am
pH 7,7
1200
Eu
pH 9,5
1500
1800
Time/days
Figure 4 : concentration of Cs, Sr, U, Tc, Np, Pu, Am and Eu as a
function of time in static leaching test of spent fuel in presence of
hydrogen overpressure [23]
Figure 5 : FIAP(Sr) per day as a function
of hydrogen final overpressure (bar) [23]
These results clearly show that in presence of relatively low concentrations of hydrogen (2.5
mmol.L-1 in the case of spent fuel and 0.5-13 mmol.L -1 for alpha doped UO2), the alteration of spent
fuel and of alpha doped UO2(s) (ITU-SKB) occurs with a very low rate as indicated by the very low
and practically constant (that is zero increase) concentrations of all redox sensitive elements in the
system. At the same time gas analysis of samples from the autoclaves indicates levels of oxygen
below detection limit although radiolytic oxygen would normally be detected in the absence of
hydrogen. This indicates that the radiolytic oxidants are consumed directly by hydrogen and can not
cause the oxidative dissolution of the UO 2(s) matrix of the spent fuel or of the doped oxide [24].
Although it is already known that very low levels of hydrogen are enough to cut the chains of
radiolytic oxidant production in presence of β/γ [25, 26], the mechanism has still to be explored by
which hydrogen counteracts the alpha radiolysis enhanced fuel dissolution.
Results from Chalmers in ICP (under 10bar hydrogen atmosphere) using spent fuel samples showed
decreasing [U] and increasing [Cs] in solution with time. These data show that hydrogen gas could
act as a reductant in these experiments. When Fe(II) was introduced into the experiments, the
aqueous U concentration fell by up to 3 orders of magnitude (to ~1×10-9 mol.L-1) with 99% of the U
reprecipitated on the surface of the spent fuel itself. 90Sr was continually released into solution
during the experiments showing that continued dissolution and subsequent reprecipitation of the
uranium was occurring.
4.2 Influence of actively-corroding iron
Within the ICP project, several set of experiments have been carried out under reducing condition
with spent fuel and alpha doped UO2 samples in presence of actively corroding iron, reproducing
part of the expected environment in the near-field. In order to measure very low dissolution rate, an
isotope dilution method was used by spiking the test solution with solutions with different 235U/238U
ratio [1]. Results from tests involving undoped uranium dioxide (0% 233U) and alpha-doped
uranium dioxide (to simulate ‘young’ spent fuel of ~2,500 (10% 233U)and 10,000 years (5% 233U)
alpha radiation dose) under both oxidizing (in air) and reducing conditions (produced by actively
corroding iron under nitrogen atmospheres) were compared with dissolution rates of spent fuel
(with the additional effects of beta and gamma radiation doses) under anaerobic conditions using a
hydrogen gas atmosphere. The results show that in the presence of iron, the dissolution rates are
very low (0.2 –0.3ppb uranium per day) and that current values used in safety assessments are
conservative values. In agreement with data from the SFS programme, the effect of alpha activity
using alpha-doped material under actively-reducing conditions was initially inconclusive, since
initial tests indicated higher dissolution rates at higher doping levels [1], but subsequent tests under
more controlled conditions showed no evidence of enhanced dissolution due to alpha radiolysis in
presence of actively corroding iron. This result is consistent with the breakdown of radiolysis due to
the presence of hydrogen produced in this experiment by iron corrosion.
4.3 Possible reduction of uranium and neptunium on corroded iron
Within ICP, experiments using an electrochemical cell were conducted to study the potential of
actively corroding iron to reduce oxidized aqueous U(VI) and Np(V) species to less soluble U(IV)
and Np(IV) in conditions representative of those expected to be found during the degradation of a
canister in a deep repository environment. In these experiments, uranium or neptunium was added
in a Allard modified groundwater to iron foils and the changes in the aqueous U or Np
concentrations were regularly measured. The results show that the solubility of uranium dioxide is
very low in the presence of iron (with uranium solution concentrations measured of less than
0.02ppb (<8.4×10-11 mol.L-1) [2]. This value is significantly lower than previously published data in
the scientific literature.
RIXS (resonant inelastic soft X-ray scattering) Spectroscopy was used to study and measure
quantitatively the reduction of aqueous U and Np species onto the corroding iron surface. Data
presented shows that U(VI) reduction can occur in solution rather than at the solid surface and that
carbonate complexes are reduced faster than hydroxyl complexes. These results were in agreement
with a new computer model that calculated from first principles the expected reaction path, and the
relative reaction rate for the reduction of U(VI) to U(IV) and Np(V) to Np(IV) in solutions with
various ligands [3]. These data show that even if more oxidised (and more soluble) U(VI) and
Np(V) species are released into solution from the waste form by radiolysis reactions, the reducing
conditions present due to the canister materials are sufficient to reduce them to less soluble products
significantly reducing the inventory of U and Np in the contaminated groundwaters which might
migrate through the host rock. This will reduce the expected radionuclide migration to the far field
and shows that the effect of canister materials must be included in performance assessment models
for high level waste disposal.
5. Conclusion
These two complementary projects allow to significantly increase the understanding and confidence
in the spent fuel performance in geological repository. Concerning the Instant Release Fraction, the
SFS project developed an overall model to describe the evolution as a function of time of the RN
location within the spent fuel pellet considering both parameter and model uncertainty. Considering
the performance allocated to each structural part of the rod, IRF inventories can hence be predicted,
allowing for different choices in some of the fundamental assumptions. Most conservative
approaches lead to much higher inventory (~15% for UOX fuel) than classically assumed (~4% for
UOX fuel). This work enlightened the potentially very strong impact of the rim region and of grain
boundaries long term instability. This topic will clearly have to be studied in order to potentially
reduce current conservatism.
Concerning the matrix performance, the significance of the radiolytic dissolution has been
experimentally demonstrated by several sets of experiments. A kinetic model of water radiolysis at
the fuel / water interface has been developed and completed in particular for brines and carbonaterich solutions. Although no definite results is available about the dose effect, first experimental
results seem to indicate that in the absence of hydrogen and actively corroding iron the fuel
alteration rate may be related to the alpha dose rate except at the lowest dose for which our data are
still inconclusive. Minor effects of alpha radiolysis at low dose rates may be related to the existence
of a dose threshold below which radiolytic dissolution is compensated by solubility-controlled
alteration. Modelling of the radiolytic fuel dissolution is on going and will focus on the two first
steps which are considered as the limiting ones: the production of radiolytic oxidants by water
radiolysis and fuel surface oxidation.
Finally, a strong focus was put on the possible interaction of some of the near field components on
the fuel alteration rate. In particular, several experiments were conducted to demonstrate the
inhibitor role of dissolved hydrogen on the radiolytic dissolution, probably due to the consumptions
of the radiolytic oxidants. Very low alteration rate (almost non measurable alteration rate) has been
observed on several long term experiments both on spent fuel and alpha doped UO2. However, this
process is not currently completely understood and will have to be more deeply studied in the nearfuture to increase the robustness of its influence. Results obtained within the ICP project also
evidence that metallic corrosion products can play a significant role in particular in the global redox
balance in the near-field, leading to actinides reduction and therefore limiting their mobility. More
work is necessary to quantify and model these effects in order to include them in repository
performance assessment.
6. Acknowledgements
This work was supported by the European Commission through the projects In Can Process
(FIKW-CT-2000-00019) and Spent Fuel Stability under repository conditions (FIKW-CT-2001000192).
References
[1]
Ollila K, Albinsson Y, Oversby V and Cowper M (2003), Dissolution rates of unirradiated
UO2, UO2 doped with 233U and spent fuel under normal atmospheric conditions and under
reducing conditions using an isotope dilution method, SKB TR-03-13.
[2]
Butorin, S Ollila K, Albinsson Y, Nordgren J and Werme L (2003), Reduction of uranyl and
neptunyl carbonate complexes with chemical-electrochemical methods and RIXS
spectroscopy, SKB TR-03-15.
[3]
Wahlgren, U. (2003), Investigating the thermodynamics of the reduction of U(VI) to U(V) by
Fe(II) using ab initio methods, SKB TR-03-14.
[4]
Poinssot C., Toulhoat P., Grambow B. (2002), Overview of the “Spent Fuel Stability under
Repository Conditions” European Project, Spent Fuel Workshop, Avignon, Sept.2002.
[5]
Johnson L., Shoesmith, D.W. (1988), radioactive waste forms for the future, Eds. Lutze W. &
Ewing R., Elsevier Science Publishers B.V., 11.
[6]
Poinssot C., Lovera P., Ferry C., Gras JM. (2002), Consequences of the anticipated long term
evolution of spent nuclear fuel for the assessment of the release rate of radionuclides, Mat.
Res. Soc. Symp. Proc Vol 757, Scientific basis for nuclear Waste Management XXVI, 35-42.
[7]
Grambow B., Loida A., Martinez-Esparza A., Diaz-Arocas P., Quinones J., de Pablo J. Casas
I., Paul J.L., Le Lous K., Marx G., Glatz J.P., Lemmens K., Ollilla K., Christensen H. (2000),
Source term for performance of assessment of spent nuclear fuel as a waste form, Final
Report, EUR 19140 EN, EC Luxembourg.
[8]
Piron, J.-P., et al. (2001), “Spent nuclear fuel evolution in a closed system”, ICEM’01, Proc.
of the 8th International Conference on Radioactive Waste Management and Environmental
Remediation (TABOAS, A., VANBRABRANT, R., BENDA,G., Eds), Bruges, Belgium,
Sept.30-Oct. 4, 2001.
[9]
Poinssot C., Toulhoat P., Gras J.M., Vitorge P. (2002), Long term evolution of spent nuclear
fuel in long term storage or geological disposal. New findings from the French PRECCI R&D
program and implications for the definition of the RN source term in geological repository,
Journal of Nuclear Science and Technology, supp.3, 473-476.
[10] C. Poinssot, C. Ferry, JM. Gras (2003), key scientific issues related to the evolution of spent
nuclear fuel in long term dry storage and geological disposal, conference MRS’03, Scientific
Basis for Nuclear Waste Management XXVII, Kalmar, June 03.
[11] Poinssot, C., Toulhoat P., Grouiller J.P., Pavageau J., Piron J.P., Pelletier M., Dehaudt P.,
Cappelaere C., Limon R., Desgranges L., Jegou C., Corbel C., Maillard S., Fauré M.H.,
Ciccariello J.C., Masson M. (2001), synthesis on the long term behaviour of the spent nuclear
fuel, CEA report CEA-R-5958, ISSN 0429-3460.
[12] Yamakawa H., Wataru M., Saegusa T. (1999), Further research in CRIEPI for the storage of
high burn up and MOX spent fuel, Journal of Nuclear Materials Management XXVII(4), 2024.
[13] Ronchi C., Hiernaut J.P. (2004), Helium diffusion in uranium and plutonium oxides, Journal
of ,Nuclear Materials (325), 1-12.
[14] Lovera P., Ferry C., Poinssot C., Johnson L.(2004), Synthesis report on the relevant diffusion
coefficients of fission products and helium in spent nuclear fuels, CEA report CEA-R-6039
ISSN 0429-3460.
[15] Ferry C., Lovera P., Poinssot C., Johnson L. (2003), quantitative assessment of instant release
fraction (IRF) or fission gases and volatile elements as a function of burnup and time under
geological disposal conditions, conference MRS’03, Scientific Basis for Nuclear Waste
Management XXVII, Kalmar, June 03, in press.
[16] Thomas L.E. (1991), Condensed phase xenon and krypton in UO2 spent fuel, Fund. Aspects
of Inert gases in Solids, Ed. Donnelly S.E. & Evans J.H., Plenum Press, NY.
[17] Johnson, L., Ferry C., Poinssot C. (2004) Estimates of the instant release fraction for UO2 and
MOX fuel at t=0, NAGRA technical report, in press.
[18] Martinez-Esparza A., (2002), proceedings of the Avila meeting on the spent fuel alteration
model, ENRESA technical report.
[19] Corbel, C. Sattonnay, G. Lucchini, J.-F. Ardois, C. Barthe, M.-F. Huet, F. Dehaudt, P. Hickel
B. and Jegou C. (2001), NIM B 179, 225-229.
[20] Christensen H., Sunder S., and Shoesmith D.W. (1994), Development of a kinetic model to
predict the rate of oxidation and dissolution of nuclear fuel (UO2) by the radiolysis of water,
AECL Canada report 1994 (AECL-11102).
[21] Jegou C., Broudic V., Poulesquen A., Bart J.M. (2003), Effects of a and ß? radiolysis of water
on alteration of the spent UO2 nuclear fuel matrix, conference MRS’03, Scientific Basis for
Nuclear Waste Management XXVII, Kalmar, June 03, Proc. In press.
[22] Martinez-Esparza A., Bruno J., Cera E., Merino J., Quinones J., de Pablo J., Gimenez J.,
Casas I., Jegou C., Poulesquen A. (2003), Conceptual model of the matrix alteration model,
Enviros technical report, in press.
[23] Poinssot C. (2003), Spent Fuel stability under repository conditions - SFS second year annual
report, CEA report, in press.
[24] Spahiu K., Devoy J., Cui D., Lundstrom M. (2003), The reduction of U(VI) by near-feidl
hydrogen in the presence of UO2(s), Migration conference 2003, Gyeongju, Proceedings in
press.
[25] Pastina B., Isabey J. and Hickel B., (1999) J. Nucl. Mat. 264, 309-318 (1999).
[26] Pastina B., LaVerne,. B. and J. A. (2001), Effect of molecular hydrogen on hydrogen peroxide
in water radiolysis, J. Phys. Chem. A 105, 9316-9322 (2001).