SR 97: Spent Fuel Alteration/Dissolution and the Influence of Near Field Hydrogen

2000 ◽  
Vol 663 ◽  
Author(s):  
Kastriot Spahiu ◽  
Patrik Sellin
Keyword(s):  
Uranium Dioxide ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Limiting Factor ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
S Matrix ◽  
The Matrix

ABSTRACTA discussion of the evaluation of the source term in the SR 97 safety assessment of a deep repository for spent nuclear fuel is presented. Since the majority of the radionuclides are embedded in the uranium dioxide fuel matrix, they will be released only after the alteration/dissolution of the matrix. Therefore a description of the process of alteration/dissolution of the spent fuel matrix is needed in a safety assessment.Under normal repository conditions, i.e. reducing environment and neutral to alkaline pH, uranium dioxide has a very low solubility in water. If solubility is assumed to be the limiting factor, the dissolution of the fuel matrix will proceed very slowly due to the low water exchange in the defective canister. On this basis, a solubility-limited model for the release of the radionuclides from the fuel may be formulated.The reducing conditions can be upset by the radioactivity of the spent fuel, which generates oxidizing products through water radiolysis. This causes the oxidative alteration/dissolution of the UO2(s) matrix. A model for fuel matrix conversion resulting from radiolytic oxidative dissolution is discussed, as well as parameter variations and the associated uncertainties.In a repository the spent fuel will come in contact with groundwater after the copper canister has breached. Large amounts of hydrogen are then produced through the anoxic corrosion of the cast iron insert. Recent data on spent fuel leaching in presence of repository relevant hydrogen pressures and the implications on the actual and future spent fuel dissolution modeling will also be discussed.

Mechanisms governing the release of radionuclides from spent nuclear fuel in geological repository: major outcomes of the European Project SFS

2006 ◽  
Vol 932 ◽  
Author(s):  
Christophe Poinssot ◽  
Cécile Ferry ◽  
Bernd Grambow ◽  
Manfred Kelm ◽  
Kastriot Spahiu ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Kinetic Scheme ◽  
Water Radiolysis ◽  
Large Set ◽  
Spent Fuel ◽  
Fuel Performance ◽  
Geological Repository ◽  
The Matrix ◽  
Release Model

ABSTRACTEuropean Commission supported a wide research project entitled “Spent Fuel Stability under repository conditions” (SFS) within the 5th FWP, the aim of which was to develop a common understanding of the radionuclides release from spent nuclear fuel in geological disposal and build a RN release model in order to assess the fuel performance. This project achieved by the end of 2004 focuses both on the Instant Release Fraction (IRF) model and the Matrix Alteration Model (MAM).A new IRF model was developed based on the anticipated performances of the various fuel microstructures (gap, rim, grains boundaries) and the potential diffusion of RN before the canister breaching. However, this model lets the choice to the end-user about the degree of conservativeness to consider.In addition, fuel alteration has been demonstrated to be linked to the production of radiolytic oxidants by water radiolysis at the fuel interface, the oxidation of the fuel interface by radiolytic oxidants and the subsequent release of uranium under the influence of aqueous ligands. A large set of experimental data was therefore acquired in order (i) to upgrade the current radiolytic kinetic scheme, (ii) to experimentally correlate the fuel alteration rate and the fuel specific alpha activity by performing experiments on alpha doped samples, (iii) to experimentally test the potential inhibitor effect of hydrogen on fuel dissolution. Based on these results, a new MAM was developed, which was also calibrated using the experiments on inactive UO2 samples. This model was finally applied to representative granitic, salt and clayey environment to predict spent fuel long-term fuel performance.

Modelling the Activation of H2 on Spent Fuel Surface and Inhibiting Effect of UO2 Dissolution

2013 ◽  
Vol 1518 ◽  
pp. 133-138 ◽  
Author(s):  
L. Duro ◽  
O. Riba ◽  
A. Martínez-Esparza ◽  
J. Bruno
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Fuel Pellet ◽  
Dissolution Behavior ◽  
Dissolution Mechanism ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Secondary Phases ◽  
The Matrix ◽  
Fuel Storage

ABSTRACTThe dissolution of spent nuclear fuel is defined in two different time steps, i) the Instant Release Fraction (IRF) occurring shortly after water contacts the solid spent fuel and responsible of the fast release of those radionuclides that have been accumulated in the zones of the spent fuel pellet with low confinement, such as gap and grain boundaries and ii) the long term release of radionuclides confined in the spent fuel matrix, much slower and dependent on the conditions of the water that contacts the spent fuel.Several models have been developed to date to explain the dissolution behavior of spent nuclear fuel under disposal conditions. The Matrix Alteration Model (MAM) is one of the most evolved radiolytic models describing the dissolution mechanism in which an Alteration/Dissolution source term model is based on the oxidative dissolution of spent fuel. Under deep repository conditions and at the expected of water contacting time (after 1000 years of spent fuel storage), α radiation will be the main contributor to water radiolysis. In the current study, simulations evaluating the effect of surface area on the alteration/dissolution of spent fuel matrix are performed considering different particle sizes of spent fuel and simulations integrating the actinides dissolution have been performed considering the precipitation of secondary phases.

SIMFUEL and UO2 Solubility and Leaching Behavior Under Anoxic Conditions

1997 ◽  
Vol 506 ◽  
Author(s):  
J. Quiñones ◽  
J. Garcia-Serrano ◽  
J.A. Serrano ◽  
P. Díaz-Arocas ◽  
J.L.R. Almazan
Keyword(s):  
Steady State ◽  
Spent Nuclear Fuel ◽  
Solid Phase ◽  
Theoretical Calculations ◽  
Radioactive Decay ◽  
Experimental Tests ◽  
Spent Fuel ◽  
Complete Replacement ◽  
Reducing Conditions ◽  
The Matrix

ABSTRACTMost of performance assessment models for spent fuel repository safety consider radiolysis-self-oxidation to describe fuel matrix release. Nevertheless, due to radioactive decay, the matrix dissolution process under reducing conditions would be controlled by the solubility limit of the steady U solid phase. In this work, leaching behaviour under anoxic and reducing conditions of spent fuel unirradiated chemical analogues (natural U02 and SIMFUEL) in simulated groundwater is studied.The trial procedure was performed taking into account the possibility that the uranium oxide to be leachated had an initial outer layer with an oxidation state higher than the matrix. This oxidised layer would produce an overestimation on U concentration in solution for the solid studied. In order to avoid this effect, a complete replacement of the leaching solutions was carried out after several days of experimentation. After this initial experimental step, the steady state concentrations obtained in all tests were more than one order of magnitude lower than before. Uranium concentrations found in reducing and anoxic experiments for both U02 and SIMFUEL tests were very close. This fact is attributed to similarity in environmental conditions (pH, Eh, etc.). From that, it can be assured that steady state concentration obtained is independent of solid leached (UO2 or SIMFUEL). In order to assess which is the solid phase that could control the solubility of U, the experimental concentration obtained was compared with results from geochemical code EQ3/6. According with the theoretical calculations U4O9 would be the controlling pure phase formed in whole experimental tests described in this work. Comparisons with bibliography data from leaching experiments of spent nuclear fuel were made as well.

scholarly journals An Alternative Conceptual Model for the Spent Nuclear Fuel–Water Interaction in Deep Geologic Disposal Conditions

2021 ◽  
Vol 11 (18) ◽  
pp. 8566
Author(s):  
Barbara Pastina ◽  
Jay A. LaVerne
Keyword(s):  
Conceptual Model ◽  
Nuclear Fuel ◽  
Release Rate ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Water Radiolysis ◽  
Spent Fuel ◽  
Radiation Chemical ◽  
Radionuclide Release

For the long-term safety assessment of direct disposal of spent nuclear fuel in deep geologic repositories, knowledge on the radionuclide release rate from the UO2 matrix is essential. This work provides a conceptual model to explain the results of leaching experiments involving used nuclear fuel or simulant materials in confirmed reducing conditions. Key elements of this model are: direct effect of radiation from radiolytic species (including defects and excited states) in the solid and in the first water layers in contact with its surface; and excess H2 may be produced due to processes occurring at the surface of the spent fuel and in confined water volumes, which may also play a role in keeping the spent fuel surface in a reduced state. The implication is that the fractional radionuclide release rate used in most long-term safety assessments (10−7 year−1) is over estimated because it assumes that there is net UO2 oxidation caused by radiolysis, in contrast with the alternative conceptual model presented here. Furthermore, conventional water radiolysis models and radiation chemical yields published in the literature are not directly applicable to a heterogeneous system such as the spent fuel–water interface. Suggestions are provided for future work to develop more reliable models for the long-term safety assessment of spent nuclear fuel disposal.

Radionuclide Transport Calculations in the Safety Assessment SR 97

2000 ◽  
Vol 663 ◽  
Author(s):  
M. Lindgren ◽  
F. Lindström
Keyword(s):  
Input Data ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Near Field ◽  
Radionuclide Transport ◽  
Spent Fuel ◽  
Release Rates ◽  
Dose Limits ◽  
The Impact ◽  
Different Parts

ABSTRACTThis study treats radionuclide transport calculations for a canister defect scenario in the safety assessment SR 97, which concerns a deep repository for spent nuclear fuel of the KBS-3 type in Sweden. The aims of the calculations are to:Quantitatively describe the radionuclide transport.Show the impact of uncertainty in input data and show which parameters govern the calculated release rates.Compare three different real sites in Sweden (Aberg, Beberg and Ceberg) with each other and with dose limits given in Swedish regulations (none of the sites is considered in the on-going localization process). Only briefly described in this paper.Illustrate the impact of the different barriers in the system.Deterministic calculations illustrate the radionuclide transport for reasonable conditions. Uncertainty cases show the influence of the uncertainty for data related to different parts of the repository system by systematically giving them pessimistic values while all others are reasonable. Simplified probabilistic calculations have also been performed.The analysis shows that the most important parameters in the near field are the number of defective canisters and the instant release fraction. In the far field the most important uncertainties affecting release and retention are connected to permeability and connectivity of the fractures in the rock. The dose rate in the biosphere is essentially controlled by the possibilities of dilution.The calculated maximum doses for the hypothetical repositories are well below the dose limits, and hence they meet the acceptance criteria for a deep repository for spent fuel.

Near-Field Modelling in the Safety Assessment SR-Can

2008 ◽  
Vol 1107 ◽  
Author(s):  
C. Fredrik Vahlund
Keyword(s):  
Numerical Modelling ◽  
Input Data ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Fractured Rock ◽  
Near Field ◽  
Spent Fuel ◽  
Radionuclide Migration ◽  
Field Modelling ◽  
Swelling Clays

AbstractSpent nuclear fuel from the Swedish energy programme will be stored in an underground repository situated in saturated fractured rock at a depth of approximately 500 m. This paper describes numerical simulations of radionuclide migration in the near-field (consisting of a canister filled with spent fuel and an engineered system backfilled with swelling clays) for the recently completed safety assessment SR-Can [1] using a Matlab / Simulink code. Handling of input data for the models from the site descriptive programme from on-going investigations at two candidate sites and the numerical modelling concept are discussed.

Assessment of Redox Conditions in the Near Field of Nuclear Waste Repositories: Application to the Swiss high-level and intermediate level waste disposal concept

2003 ◽  
Vol 807 ◽  
Author(s):  
Paul Wersin ◽  
Lawrence H. Johnson ◽  
Bernhard Schwyn
Keyword(s):  
Solid Phase ◽  
Near Field ◽  
Intermediate Level ◽  
Redox Conditions ◽  
High Level Waste ◽  
Redox Potentials ◽  
Spent Fuel ◽  
Reducing Conditions ◽  
High Level ◽  
Waste Repositories

ABSTRACTRedox conditions were assessed for a spent fuel and high-level waste (SF/HLW) and an intermediate-level waste (ILW) repository. For both cases our analysis indicates permanently reducing conditions after a relatively short oxic period. The canister-bentonite near field in the HLW case displays a high redox buffering capacity because of expected high activity of dissolved and surface-bound Fe(II). This is contrary to the cementitious near field in the ILW case where concentrations of dissolved reduced species are low and redox reactions occur primarily via solid phase transformation processes.For the bentonite-canister near field, redox potentials of about -100 to -300 mV (SHE) are estimated, which is supported by recent kinetic data on U, Tc and Se interaction with reduced iron systems. For the cementitious near field, redox potentials of about -200 to -800 mV are estimated, which reflects the large uncertainties related to this alkaline environment.

scholarly journals An Automated SeaFAST ICP-DRC-MS Method for the Determination of 90Sr in Spent Nuclear Fuel Leachates

Molecules ◽  
2020 ◽  
Vol 25 (6) ◽  
pp. 1429 ◽  
Author(s):  
Víctor Vicente Vilas ◽  
Sylvain Millet ◽  
Miguel Sandow ◽  
Luis Iglesias Pérez ◽  
Daniel Serrano-Purroy ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Analytical Approach ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Chemical Purification ◽  
Radiochemical Method ◽  
Good Agreement

To reduce uncertainties in determining the source term and evolving condition of spent nuclear fuel is fundamental to the safety assessment. ß-emitting nuclides pose a challenging task for reliable, quantitative determination because both radiometric and mass spectrometric methodologies require prior chemical purification for the removal of interfering activity and isobars, respectively. A method for the determination of 90Sr at trace levels in nuclear spent fuel leachate samples without sophisticated and time-consuming procedures has been established. The analytical approach uses a commercially available automated pre-concentration device (SeaFAST) coupled to an ICP-DRC-MS. The method shows good performances with regard to reproducibility, precision, and LOD reducing the total time of analysis for each sample to 12.5 min. The comparison between the developed method and the classical radiochemical method shows a good agreement when taking into account the associated uncertainties.

Contingency Options for the Drying, Conditioning and Packaging of Magnox Spent Fuel in the UK

2009 ◽  
Author(s):  
Jenny Morris ◽  
Stephen Wickham ◽  
Phil Richardson ◽  
Colin Rhodes ◽  
Mike Newland
Keyword(s):  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Ambient Temperatures ◽  
Long Term Storage ◽  
Reducing Conditions ◽  
The Us ◽  
Potential Formation ◽  
The Uk ◽  
Interim Storage ◽  
Metallic Uranium

The UK Nuclear Decommissioning Authority (NDA) is responsible for safe and secure management of spent nuclear fuel. Magnox spent fuel is held at some Magnox reactor sites and at Sellafield where it is reprocessed using a number of facilities. It is intended that all Magnox fuel will be reprocessed, as described in the published Magnox Operating Plan (MOP) [1]. In the event, however, that a failure occurs within the reprocessing plant, the NDA has initiated a programme of activities to explore alternative contingency options for the management of wetted Magnox spent fuel. Magnox fuel comprises metallic uranium bar clad in a magnesium alloy, both of which corrode if exposed to oxygen or water. Consequently, contingency options are required to consider how best to manage the issues associated with the reactivity of the metals. Questions of whether Magnox spent fuel needs to be dried, how it might be conditioned, how it might be packaged, and held in temporary storage until a disposal facility becomes available, all require attention. A review of potential contingency options for Magnox fuel was conducted by Galson Sciences Ltd, UKAEA and the NDA. During storage in the presence of water, the corrosion of Magnox fuel produces hydrogen (H2) gas, which requires careful management. When uranium reacts with hydrogen in a reducing environment, the formation of uranium hydride (UH3) may occur, which under some circumstances can be pyrophoric, and might create hazards which may affect subsequent retrieval and/or repackaging (e.g. for disposal). Other factors that may affect the choice of a viable contingency option include criticality safety, environmental impacts, security and Safeguards and economic considerations. At post-irradiation examination (PIE) facilities in the UK, Magnox spent fuel is dried as a result of storage in air at ambient temperatures. Early French UNGG (Uranium Naturel Graphite Gaz) fuel was retrieved from pond storage at Cadarache, dried using a hot gas drying technique, oxidised and packaged in sealed canisters and placed in interim storage at the CASCAD (CASemate CADarache) facility. In the US, spent fuels including the Zircaloy clad Hanford N-Reactor fuels were cold vacuum dried and Idaho legacy aluminium clad metallic uranium fuels were hot vacuum dried; the dried fuel was then packaged in sealed and vented canisters (at Hanford and Idaho, respectively) for interim storage. With regard to conditioning and packaging, several different approaches have been reviewed, including encapsulation in cementitious grout or polymer, high-temperature vitrification or ceramicisation, and solution in acid or alkali solution followed by cementation or vitrification (without reprocessing). All of these approaches require further research in order to be evaluated and developed further for application to formerly wetted Magnox fuel. A variety of containers have been developed for the transport, storage and/or disposal of spent fuel in radioactive waste management programmes worldwide. Wetted Magnox spent fuel could be packaged in a container, with reservations about the potential formation of UH3 in a sealed environment where reducing conditions may develop. The applicability of different combinations of drying, conditioning and packaging techniques to the preparation of Magnox spent fuel for long-term storage and eventual disposal are discussed.

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

2008 ◽  
Vol 1104 ◽  
Author(s):  
Claude Degueldre ◽  
Wolfgang Wiesenack
Keyword(s):  
Near Field ◽  
Fission Products ◽  
Minor Actinides ◽  
Spent Fuel ◽  
Geological Disposal ◽  
Inert Matrix Fuel ◽  
Inert Matrix ◽  
Fuel Pellets ◽  
Reducing Conditions ◽  
Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

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