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.

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.

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.

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.

scholarly journals The Three-­Dimensional Neutron-­Physical Model of Spent Nuclear Fuel Storage System

2021 ◽  
Vol 20 ◽  
pp. 51-59
Author(s):  
О. R. Trofymenko ◽  
◽  
І. M. Romanenko ◽  
М. І. Holiuk ◽  
C. V. Hrytsiuk ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Storage System ◽  
Three Dimensional ◽  
Monte Carlo Code ◽  
Storage Facility ◽  
Spent Fuel ◽  
Wide Range ◽  
Fuel Assemblies ◽  
Fuel Storage

The management of spent nuclear fuel is one of the most pressing problems of Ukraine’s nuclear energy. To solve this problem, as well as to increase Ukraine’s energy independence, the construction of a centralized spent nuclear fuel storage facility is being completed in the Chornobyl exclusion zone, where the spent fuel of Khmelnytsky, Rivne and South Ukrainian nuclear power plants will be stored for the next 100 years. The technology of centralized storage of spent nuclear fuel is based on the storage of fuel assemblies in ventilated HI-STORM concrete containers manufactured by Holtec International. Long-term operation of a spent nuclear fuel storage facility requires a clear understanding of all processes (thermohydraulic, neutron-physical, aging processes, etc.) occurring in HI-STORM containers. And this cannot be achieved without modeling these processes using modern specialized programs. Modeling of neutron and photon transfer makes it possible to analyze the level of protective properties of the container against radiation, optimize the loading of MPC assemblies of different manufacturers and different levels of combustion and evaluate biological protection against neutron and gamma radiation. In the future it will allow to estimate the change in the isotopic composition of the materials of the container, which will be used for the management of aging processes at the centralized storage of spent nuclear fuel. The article is devoted to the development of the three-dimensional model of the HI-STORM storage system. The model was developed using the modern Monte Carlo code Serpent. The presented model consists of models of 31 spent fuel assemblies 438MT manufactured by TVEL company, model MPC-31 and model HISTORM 190. The model allows to perform a wide range of scientific tasks required in the operation of centralized storage of spent nuclear fuel.

scholarly journals Prevention of Damage to Spent Nuclear Fuel during Handling Operations

2020 ◽  
pp. 62-71
Author(s):  
M. Sapon ◽  
O. Gorbachenko ◽  
S. Kondratyev ◽  
V. Krytskyy ◽  
V. Mayatsky ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Spent Nuclear Fuel ◽  
Spent Fuel ◽  
Long Term Storage ◽  
Fuel Storage ◽  
Storage Facilities

According to regulatory requirements, when carrying out handling operations with spent nuclear fuel (SNF), prevention of damage to the spent fuel assemblies (SFA) and especially fuel elements shall be ensured. For this purpose, it is necessary to exclude the risk of SFA falling, SFA uncontrolled displacements, prevent mechanical influences on SFA, at which their damage is possible. Special requirements for handling equipment (in particular, cranes) to exclude these dangerous events, the requirements for equipment strength, resistance to external impacts, reliability, equipment design solutions, manufacturing quality are analyzed in this work. The requirements of Ukrainian and U.S. regulatory documents also are considered. The implementation of these requirements is considered on the example of handling equipment, in particular, spent nuclear fuel storage facilities. This issue is important in view of creation of new SNF storage facilities in Ukraine. These facilities include the storage facility (SFSF) for SNF from water moderated power reactors (WWER): a Сentralized SFSF for storing SNF of Rivne, Khmelnitsky and South-Ukraine Nuclear Power Plants (СSFSF), and SFSF for SNF from high-power channel reactors (RBMK): a dry type SFSF at Chornobyl nuclear power plant (ISF-2). After commissioning of these storage facilities, all spent nuclear fuel from Ukrainian nuclear power plants will be placed for long-term “dry” storage. The safety of handling operations with SNF during its preparation for long-term storage is an important factor.

scholarly journals Nuclear and Radiation Safety of the Centralized Spent Fuel Storage Facility in Ukraine

2017 ◽  
pp. 3-10
Author(s):  
O. Hryhorash ◽  
O. Dybach ◽  
S. Kondratiev ◽  
O. Horbachenko ◽  
A. Panchenko ◽  
...  
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Radiation Safety ◽  
Storage Facility ◽  
Spent Fuel ◽  
Regulatory Documents ◽  
Thermal Regimes ◽  
Fuel Storage ◽  
Nuclear And Radiation Safety ◽  
Biological Protection

The paper presents the analysis of ensuring nuclear and radiation safety in the management of spent nuclear fuel at the Centralized SFSF and activities planned for Centralized SFSF lifecycle stages. There are results of comparing requirements of U.S. regulatory documents used by the HOLTEC Company to design Centralized SFSF equipment staff with relevant requirements of Ukrainian regulations, results based on analysis of the most important factors of Centralized SFSF safety (strength and reliability, nuclear safety, thermal regimes and biological protection) and verified expert calculations of the SSTC NRS. The paper includes issues to be considered in further implementation of Centralized SFSF project.

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.

A Comparative Evaluation of Licensing Requirements for Dry Storage of Spent Nuclear Fuel in the United States, Germany, Canada and the Russian Federation

2004 ◽  
Author(s):  
Edward Wonder ◽  
David S. Duncan ◽  
Eric A. Howden
Keyword(s):  
United States ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
The United States ◽  
Approval Process ◽  
Spent Fuel ◽  
Regulatory Requirements ◽  
Fuel Storage ◽  
The Russian Federation ◽  
The Republic

Technical activities to support licensing of dry spent nuclear fuel storage facilities are complex, with policy and regulatory requirements often being influenced by politics. Moreover, the process is often convoluted, with numerous and diverse stakeholders making the licensing activity a difficult exercise in consensus-reaching. The objective of this evaluation is to present alternatives to assist the Republic of Kazakhstan (RK) in developing a licensing approach for a planned Dry Spent Fuel Storage Facility. Because the RK lacks experience in licensing a facility of this type, there is considerable interest in knowing more about the approval process in other countries so that an effective, non-redundant method of licensing can be established. This evaluation is limited to a comparison of approaches from the United States, Germany, Russia, and Canada. For each country considered, the following areas were addressed: siting; fuel handling and cask loading; dry fuel storage; and transportation of spent fuel. The regulatory requirements for each phase of the process are presented, and a licensing approach that would best serve the RK is recommended.

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