scholarly journals Overview of Post-Closure Criticality Safety – RD&D Topics in Switzerland

2021 ◽  
Vol 1 ◽  
pp. 111-112
Author(s):  
Madalina Wittel ◽  
Susanne Pudollek
Keyword(s):  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Current Phase ◽  
General Context ◽  
Spent Fuel ◽  
Research Topics ◽  
Regulatory Requirement ◽  
Research Activities ◽  
National Cooperative ◽  
Criticality Safety

Abstract. The demonstration of post-closure criticality safety of spent nuclear fuel in a deep geological repository is a regulatory requirement in Switzerland and many other countries. One of the main challenges stems from the very long timescale (1 million years in Switzerland) that has to be considered. Nagra, the Swiss National Cooperative for the Disposal of Radioactive Waste, is presently elaborating the technical and scientific foundation of the criticality safety assessment in view of the upcoming general licence application for the Swiss Spent Fuel and HLW repository. In this context, Nagra supports and pursues a focussed RD&D programme in collaboration with several renowned research institutes. Nagra's safety concept relies on natural and technical barriers. For the initial thermal phase of the repository, a steel canister assures complete containment of the spent fuel. The canisters are foreseen to remain intact for approximately 10 000 years; however, the subcriticality of the system has to be ensured for a much longer period. In this context, an important part of the research activities pursued by Nagra address the nearfield evolution and the formulation of scenarios for the corresponding evolution of the canister and spent fuel system. The role that variations in the canister design and material composition have on the system's reactivity are also investigated. Other research topics focus on developing a reliable methodology for carrying out the criticality safety assessment. This symposium contribution gives an overview of the post-closure criticality RD&D activities pursued and envisioned by Nagra. The general context and Nagra's fundamental approach to elaborating the current phase of the criticality safety assessment are presented first. Following this, the current RD&D landscape and the most important technical considerations underpinning Nagra's technical basis for the post-closure criticality safety assessment in particular are discussed. Future planned research topics and points of interest are also presented as an outlook of this presentation.

scholarly journals Strengthening the SSTC NRS Scientific and Technical Potential through Participation in the IAEA Coordinated Research Projects

2021 ◽  
pp. 5-13
Author(s):  
Yu. Balashevska ◽  
D. Gumenyuk ◽  
Iu. Ovdiienko ◽  
O. Pecherytsia ◽  
I. Shevchenko ◽  
...  
Keyword(s):  
Best Practices ◽  
Safety Assessment ◽  
Radiation Safety ◽  
Research Projects ◽  
Spent Fuel ◽  
Research Activities ◽  
Technical Potential ◽  
Technical Issues ◽  
Fuel Characterization ◽  
Definition Of

The State Scientific and Technical Center for Nuclear and Radiation Safety (SSTC NRS), a Ukrainian enterprise with a 29-year experience in the area of scientific and technical support to the national nuclear regulator (SNRIU), has been actively involved in international research activities. Participation in the IAEA coordinated research activities is among the SSTC NRS priorities. In the period of 2018–2020, the IAEA accepted four SSTC NRS proposals for participation in respective Coordinated Research Projects (CRPs). These CRPs address scientific and technical issues in different areas such as: 1) performance of probabilistic safety assessment for multi-unit/multi-reactor sites; 2) use of dose projection tools to ensure preparedness and response to nuclear and radiological emergencies; 3) phenomena related to in-vessel melt retention; 4) spent fuel characterization. This article presents a brief overview of the abovementioned projects with definition of scientific contributions by the SSTC NRS (participation in benchmarks, development of methodological documents on implementing research stages and of IAEA technical documents (TECDOC) for demonstration of best practices and results of research carried out by international teams).

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.

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.

scholarly journals Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Materials ◽  
2019 ◽  
Vol 12 (3) ◽  
pp. 494
Author(s):  
Alexander Vasiliev ◽  
Jose Herrero ◽  
Marco Pecchia ◽  
Dimitri Rochman ◽  
Hakim Ferroukhi ◽  
...  
Keyword(s):  
Fuel Assembly ◽  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Operating Conditions ◽  
Design Parameters ◽  
Fuel Loading ◽  
Pressurized Water ◽  
Fuel Assemblies ◽  
National Cooperative ◽  
Criticality Safety

This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

Axial Burnup Profile Influence on Criticality Safety of WWER Spent Fuel

2006 ◽  
Author(s):  
Surik Bznuni ◽  
Armen Amirjanyan ◽  
Shahen Poghosyan
Keyword(s):  
Safety Assessment ◽  
Safety Analysis ◽  
Spent Fuel ◽  
End Effect ◽  
Oak Ridge ◽  
Fuel Assemblies ◽  
Spent Fuel Storage ◽  
Fuel Storage ◽  
Criticality Analysis ◽  
Criticality Safety

Criticality safety assessment for WWER-440 NUHOMS® cask with spent nuclear fuel from Armenian NPP has been performed. The cask was designed in a such way that the neutron multiplication factor keff must be below 0,95 for all operational modes and accident conditions. Usually for criticality analysis, fresh fuel approach with the highest enrichment is taken as conservative assumption as it was done for ANPP. Nuclear and Radiation Safety Centre of Armenian Nuclear Regulatory Authority (NRSC ANRA) in order to improve future fuel storage efficiency, initiated research with taking into account burn up credit in the criticality safety assessment. Axial burn up profile (end effect) has essential impact on criticality safety justification analysis. However this phenomenon wasn’t taken into account in the Safety Analysis Report of NUHOMS® spent fuel storage constructed on the site of ANPP. Although ANRA doesn’t yet accept burn up credit approach for ANPP spent fuel storage, assessment of impact of axial burn up profile on criticality of spent fuel assemblies has important value for future activities of ANRA. This paper presents results of criticality safety analysis of spent fuel assemblies with axial burn up profile. Horizontal burn up profile isn’t taken account since influence of the horizontal variation of the burn up is much less than the axial variation. The Actinides and Actinides + Fission Products approach are discussed. The calculations were carried out with STARBUCS module of SCALE 5.0 code package developed at Oak Ridge National laboratory. SCALE5.0 sequence CSAS26 (KENO-VI) was used for evaluation the keff for 3-D problems. Obtained results showed that criticality of ANPP spent fuel cask is very sensitive to the end effect. Using Burn up profiles of Control Assemblies in both approaches leads to much more increasing than in case of Working Assemblies. Usually increasing burn up leads to decreasing Δkeff, hence decreasing end effect. However for WWER-440 Control Assemblies that worked only within 6th (operative) group increasing burn up leads to increasing of the end effect.

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.

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.

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.

Sign in / Sign up

Export Citation Format

Share Document