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.

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.

Integration of the H2 Inhibition Effect of UO2 Matrix Dissolution Into Radiolytic Models

2009 ◽  
Author(s):  
Lara Duro ◽  
Abel Tamayo ◽  
Jordi Bruno ◽  
Aurora Marti´nez-Esparza
Keyword(s):  
Spent Nuclear Fuel ◽  
Solid Phase ◽  
Dissolution Mechanism ◽  
Spent Fuel ◽  
Inhibition Effect ◽  
Matrix Surface ◽  
High Let Radiation ◽  
The Matrix ◽  
High Let ◽  
Source Term Model

Different models describing the dissolution mechanism of spent nuclear fuel under repository conditions have been developed in the last years. One of the most evolved ones is the Matrix Alteration Model (MAM), which is an Alteration/Dissolution source term model based on the oxidative dissolution of spent fuel. Oxidant and reducing species can be naturally or radiolytically-generated. The experimentally-observed inhibition of matrix dissolution by H2, was integrated into MAM by considering that H2 is able to consume the oxidant species responsible for UO2 oxidation, e.g. H2O2. As a consequence, MAM predicts lower H2O2 concentrations for systems containing larger amounts of dissolved H2. Radiolysis experiments carried out by Pastina and coworkers have shown that under specific conditions such as high linear energy transfer (LET) radiation and absence of solid phase, dissolved H2 has a negligible effect on the H2O2 concentration, thus suggesting that the H2 inhibition effect catalyzed by the matrix surface has not been properly implemented in MAM. Modelling exercises performed in this work confirm such point and reveal the necessity of considering H2-activation when modelling this kind of systems. In addition, it has been demonstrated that, for high LET radiation, a clear dependence exists between the extent of the H2 activation and the integral LET of the radiation. The integration of the function describing such dependence allows improving the implementation of the H2 inhibition effect in MAM.

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.

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.

Influence of the Evolution of the Surface Area Value on the Spent Nuclear Fuel Dissolution Rate for Performance Assessment Studies

2008 ◽  
Vol 1124 ◽  
Author(s):  
Javier Quiñones ◽  
Eduardo Iglesias ◽  
Nieves Rodriguez ◽  
Juan Manuel Nieto
Keyword(s):  
Surface Area ◽  
Spent Nuclear Fuel ◽  
Current Knowledge ◽  
Surface Oxidation ◽  
Initial Surface ◽  
Spent Fuel ◽  
Initial Oxidation ◽  
Evaluation Time ◽  
The Matrix ◽  
Initial Power

AbstractThis paper focuses on how to extrapolate current knowledge of spent fuel matrix alteration processes from laboratory to repository conditions, i.e., the influence of changes in both the initial surface oxidation level and the evolution of the specific surface area during the alteration process. Therefore, a spent fuel matrix alteration model allowing the alteration rate evolution to be predicted as a function of both the host rock considered and evaluation time scale of interest is described. At present, the model assumes that alteration of the spent fuel will start when the groundwater reaches the solid surface and that only the radiolytic species of the groundwater (oxidants generated by a-radiation of spent fuel) will produce the surface oxidation process and subsequent matrix dissolution; O2, H2O2 and OH are the species that react with UO2(s) for oxidation of the pellet surface. The dissolution process of the surface sites that are oxidized is modelled in two steps: first, a surface co-ordination of the oxidized layer with aqueous ligands and, second, detachment (dissolution) of the product species. Taking this mechanism into account, the model gives the evolution of the spent fuel matrix alteration rate over periods as long as 1,000,000 years. In this work is focussed on input the experimental results obtained of UO2 surface area behaviour (presented in previous MRS conference), on the MAM model. The matrix alteration rate results obtained, with MAM model, for repository granitic environment will be presented and compared to those performed for SFS project. Furthermore, a sensitivity analysis study has been performed on the influence of the following variables: Influence of the initial power size distribution and the initial oxidation state

Leaching of Spent Fuel under Anaerobic and Reducing Conditions

2002 ◽  
Vol 757 ◽  
Author(s):  
Yngve Albinsson ◽  
Arvid Ödegaard-Jensen ◽  
Virginia M. Oversby ◽  
Lars O. Werme
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Bentonite Clay ◽  
Fluid Phase ◽  
Inert Atmosphere ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Reducing Conditions ◽  
Reaction Vessels

ABSTRACTSweden plans to dispose of spent nuclear fuel in a deep geologic repository in granitic rock. The disposal conditions allow water to contact the canisters by diffusion through the surrounding bentonite clay layer. Corrosion of the canister iron insert will consume oxygen and provide actively reducing conditions in the fluid phase. Experiments with spent fuel have been done to determine the dissolution behavior of the fuel matrix and associated fission products and actinides under conditions ranging from inert atmosphere to reducing conditions in solutions. Data for U, Pu, Np, Cs, Sr, Tc, Mo, and Ru have been obtained for dissolution in a dilute NaHCO3 groundwater for 3 conditions: Ar atmosphere, H2 atmosphere, and H2 atmosphere with Fe(II) in solution. Solution concentrations forU, Pu, and Mo are all significantly lower for the conditions that include Fe(II) ions in the solutions together with H2 atmosphere, while concentrations of the other elements seem to be unaffected by the change of atmospheres or presence of Fe(II). Most of the material that initially dissolved from the fuel has reprecipitated back onto the fuel surface. Very little material was recovered from rinsing and acid stripping of the reaction vessels.

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.

Modeling Heat Transfer in Spent Fuel Transfer Cask Neutron Shields: A Challenging Problem in Natural Convection

2010 ◽  
Author(s):  
James A. Fort ◽  
Judith M. Cuta ◽  
Chris S. Bajwa ◽  
Emilio Baglietto
Keyword(s):  
Heat Transfer ◽  
Experimental Data ◽  
Steady State ◽  
Natural Convection ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Challenging Problem ◽  
Spent Fuel ◽  
Neutron Shield

In the United States, commercial spent nuclear fuel is typically moved from spent fuel pools to outdoor dry storage pads within a transfer cask system that provides radiation shielding to protect personnel and the surrounding environment. The transfer casks are cylindrical steel enclosures with integral gamma and neutron radiation shields. Since the transfer cask system must be passively cooled, decay heat removal from the spent nuclear fuel canister is limited by the rate of heat transfer through the cask components, and natural convection from the transfer cask surface. The primary mode of heat transfer within the transfer cask system is conduction, but some cask designs incorporate a liquid neutron shield tank surrounding the transfer cask structural shell. In these systems, accurate prediction of natural convection within the neutron shield tank is an important part of assessing the overall thermal performance of the transfer cask system. The large-scale geometry of the neutron shield tank, which is typically an annulus approximately 2 meters in diameter but only 10–15 cm in thickness, and the relatively small scale velocities (typically less than 5 cm/s) represent a wide range of spatial and temporal scales that contribute to making this a challenging problem for computational fluid dynamics (CFD) modeling. Relevant experimental data at these scales are not available in the literature, but some recent modeling studies offer insights into numerical issues and solutions. However, the geometries in these studies, and for the experimental data in the literature at smaller scales, all have large annular gaps that are not prototypic of the transfer cask neutron shield. This paper proposes that there may be reliable CFD approaches to the transfer cask problem, specifically coupled steady-state solvers or unsteady simulations; however, both of these solutions take significant computational effort. Segregated (uncoupled) steady state solvers that were tested did not accurately capture the flow field and heat transfer distribution in this application. Mesh resolution, turbulence modeling, and the tradeoff between steady state and transient solutions are addressed. Because of the critical nature of this application, the need for new experiments at representative scales is clearly demonstrated.

Modeling spent nuclear fuel alteration and radionuclide migration in disposal conditions

2006 ◽  
Vol 94 (9-11) ◽  
Author(s):  
Laurent de Windt ◽  
H. Schneider ◽  
C. Ferry ◽  
H. Catalette ◽  
V. Lagneau ◽  
...  
Keyword(s):  
Reactive Transport ◽  
Spent Nuclear Fuel ◽  
Transport Model ◽  
Near Field ◽  
Radioactive Decay ◽  
Steel Corrosion ◽  
Spent Fuel ◽  
Fuel Pellets ◽  
Physico Chemical ◽  
Integrated Performance

A physico-chemical model developed for spent fuel alteration was integrated in a global reactive transport model of a spent fuel disposal system, considering both decaying and stable isotopes, corroded steel canisters, bentonite backfills and a clayey host-rock. Fuel evolution took into account radiolytic-enhanced corrosion and long-term solubility-controlled dissolution as well as instantaneous release fractions. The calculations show that spent-fuel dissolution has no significant alteration effect on the near-field components except an oxidizing plume in the vicinity of the waste packages. The dissolved uranyl species, partly precipitate as schoepite on the fuel pellets, and partly diffuse in the near-field where magnetite and pyrite reduce U(VI) to yield uraninite precipitation. Under disposal conditions, preliminary calculations indicate that steel corrosion may generate sufficient dissolved hydrogen as to react with radiolytic oxidants and inhibit fuel dissolution. The formation of a protective schoepite layer could also reduce the alteration of fuel pellets. Radionuclides migration (Am, Cs, I) in the near-field is discussed in a second stage discriminating between sorption, precipitation and radioactive decay processes. The migration of Cs is translated in terms of cumulative activity profiles useful for integrated performance assessment.

Radionuclides Release Model for Performance Assessment Studies of Spent Nuclear Fuel in Geological Disposal

2004 ◽  
Vol 824 ◽  
Author(s):  
Christophe Poinssot ◽  
Patrick Lovera ◽  
Cécile Ferry
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Hydrogen Effect ◽  
Spent Fuel ◽  
Geological Disposal ◽  
The Matrix ◽  
Simple Kinetic Model ◽  
Scientific Results ◽  
Release Model

AbstractIn the framework of the research conducted on the long term evolution of spent nuclear fuel in geological disposal conditions, a source term model has been developed to evaluate the instantaneous release of RN (Instant Release Fraction IRF) and the delayed release of the RN which are embedded within the matrix. This model takes into account all the scientific results currently available in the literature except the hydrogen effect. IRF was assessed by considering the evolution with time of the RN inventories located within the fuel microstructure to which no confinement properties can be allocated on the long term (rim, gap, grain boundaries). It allows to propose some reference bounding values for the IRF as a function of time of canister breaching and burnup. The matrix radiolytic dissolution was modeled by a simple kinetic model neglecting the radiolytic species recombination and the influence of aqueous ligands and radiolytic oxidants were supposed to completely react with the fuel surface. Spent fuel performance was therefore demonstrated to deeply depend on the reactive surface area.

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