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.

The Assessment of the Long-Term Evolution of the Spent Nuclear Fuel Matrix by Kinetic, Thermodynamic and Spectroscopic Studies of Uranium Minerals.

1994 ◽  
Vol 353 ◽  
Author(s):  
Jordi Bruno ◽  
I. Casas ◽  
E Cera ◽  
R. C. Ewing ◽  
R. J. Finch ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Source Term ◽  
Spent Nuclear Fuel ◽  
Spectroscopic Studies ◽  
Spent Fuel ◽  
Term Evolution ◽  
Uranium Minerals ◽  
The Stability

AbstractThe long term behaviour of spent nuclear fuel is discussed in the light of recent thermodynamic and kinetic data on mineralogical analogues related to the key phases in the oxidative alteration of uraninite. The implications for the safety assessment of a repository of the established oxidative alteration sequence of the spent fuel matrix are illustrated with Pagoda calculations. The application to the kinetic and thermodynamic data to source term calculations indicates that the appearance and duration of the U(VI) oxyhydroxide transient is critical for the stability of the fuel matrix.

Landscape Modeling for Dose Calculations in the Safety Assessment of a Repository for Spent Nuclear Fuel

2007 ◽  
Author(s):  
Tobias Lindborg ◽  
Ulrik Kautsky ◽  
Lars Brydsten
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Landscape Model ◽  
Landscape Development ◽  
Surface Hydrology ◽  
Site Investigations ◽  
Time Period ◽  
The Individual

The Swedish Nuclear Fuel and Waste Management Co., (SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site (figure 1). Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the modeled time period, from sea to lake, and from lake to mire and finally, some objects are transformed into agricultural area due to favorable farming characteristics.

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.

Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague

2013 ◽  
Author(s):  
A. Meleshyn ◽  
U. Noseck
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Auxiliary Information ◽  
Opalinus Clay ◽  
High Level Waste ◽  
Stable Elements ◽  
High Level ◽  
Model Approach

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined — either by direct measurements or by involving established correlations — and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as 14C, 36Cl, 79Se, and 129I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries — Belgium, France, Germany, Japan, Netherlands, and Switzerland — an international effort can be recommended.

Review about the Effect of He on the Microstructure of Spent Nuclear Fuel in a Repository

MRS Advances ◽  
2016 ◽  
Vol 1 (62) ◽  
pp. 4147-4156 ◽  
Author(s):  
C. Ferry ◽  
J. Radwan ◽  
H. Palancher
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Critical Concentration ◽  
Fuel Pellet ◽  
Spent Fuel ◽  
Operational Model ◽  
Matrix Microstructure ◽  
Natural Analogues ◽  
Order Of Magnitude ◽  
Radionuclide Release

ABSTRACTHelium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

scholarly journals Approaches to Safety Justification for Loading of VSC-VVER Containers in ZNPP DSFSF

2019 ◽  
pp. 82-87
Author(s):  
Ya. Kostiushko ◽  
O. Dudka ◽  
Yu. Kovbasenko ◽  
A. Shepitchak
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Spent Fuel ◽  
Fuel Assemblies ◽  
Fuel Storage ◽  
The Government

The introduction of new fuel for nuclear power plants in Ukraine is related to obtaining a relevant license from the regulatory authority for nuclear and radiation safety of Ukraine. The same approach is used for spent nuclear fuel (SNF) management system. The dry spent fuel storage facility (DSFSF) is the first nuclear facility created for intermediate dry storage of SNF in Ukraine. According to the design based on dry ventilated container storage technology by Sierra Nuclear Corporation and Duke Engineering and Services, ventilated storage containers (VSC-VVER) filled with SNF of VVER-1000 are used, which are located on a special open concrete site. Containers VSC-VVER are modernized VSC-24 containers customized for hexagonal VVER-1000 spent fuel assemblies. The storage safety assessment methodology was created and improved directly during the licensing process. In addition, in accordance with the Energy Strategy of Ukraine up to 2035, one of the key task is the further diversification of nuclear fuel suppliers. Within the framework of the Executive Agreement between the Government of Ukraine and the U.S. Government, activities have been underway since 2000 on the introduction of Westinghouse fuel. The purpose of this project is to develop, supply and qualify alternative nuclear fuel compatible with fuel produced in Russia for Ukrainian NPPs. In addition, a supplementary approach to safety analysis report is being developed to justify feasibility of loading new fuel into the DSFSF containers. The stated results should demonstrate the fulfillment of design criteria under normal operating conditions, abnormal conditions and design-basis accidents of DSFSF components.  Thus, the paper highlights both the main problems of DSFSF licensing and obtaining permission for placing new fuel types in DSFSF.

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.

Long-term safety of the extended SFR - Methodology and conclusions from the SR-PSU project

MRS Advances ◽  
2016 ◽  
Vol 1 (61) ◽  
pp. 4075-4080
Author(s):  
Fredrik Vahlund
Keyword(s):  
Radioactive Waste ◽  
Nuclear Fuel ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Surface System ◽  
Radiological Safety ◽  
The Impact

ABSTRACTSince 1988 the Swedish Nuclear Fuel and Waste Management Co. operates a repository for low- and intermediate-level short-lived radioactive waste, SFR, in Forsmark, Sweden. Due to decommissioning of the nuclear power plants additional storage capacity is needed. In December 2014, an application to extend the repository was therefore submitted. One key component of this application was an assessment of post-closure safety of the extended SFR. For this safety assessment, a methodology based on that developed by SKB for the spent nuclear fuel repository was used and the impact of the degradation of repository components, the evolution of the surface system and changes of future climate on the radiological safety of the repository was assessed over a period of 100,000 years. The central conclusion of the SR-PSU safety assessment is that the extended SFR repository meets requirements on protection of human health and of the environment that have been established by the Swedish radiation safety authority for the final disposal of radioactive waste. Furthermore, the design of the repository was shown suitable for the waste selected and the applied methodology suitable for the safety assessment.

Sorption of Actinides in Well-Defined Oxidation States on Geologic Media

1981 ◽  
Vol 11 ◽  
Author(s):  
B. Allard ◽  
U. Olofsson ◽  
B. Torstenfelt ◽  
H. Kipatsi ◽  
K. Andersson
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Oxidation States ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Term Storage ◽  
Geologic Media

The long-lived actinides and their daughter products largely dominate the biological hazards from spent nuclear fuel already from some 300 years after the discharge from the reactor and onwards . Therefore it is essential to make reliable assessments of the geochemistry of these elements in any concept for long-term storage of spent fuel or reprocessing waste, etc.

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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