Comparison of the Concepts and Assumptions in Five Recent HLW/Spent Fuel Performance Assessments

1994 ◽  
Vol 353 ◽  
Author(s):  
F. B. Neall ◽  
P. Baertschi ◽  
I.G. McKinley ◽  
P. A. Smith ◽  
T. Sumerling ◽  
...  
Keyword(s):  
Safety Assessment ◽  
Performance Assessments ◽  
Spent Fuel ◽  
Environmental Radiation ◽  
Fuel Performance ◽  
Distant Future ◽  
Future Evolution ◽  
Radiological Hazards ◽  
Other Information ◽  
Integrated Performance

Integrated performance assessments of proposed repositories for HLW involve multidisciplinary studies which include identification of possible paths of future evolution of the repository system (scenario analysis) and quantification of the consequences of each scenario using deterministic or probabilistic modelling approaches. The results of such studies are commonly presented as estimates of consequence (e.g. dose to a representative individual) against time which extends into the distant future. Individual components of the model chains used can be tested to some extent, but how can the integrated assessment be evaluated to give additional confidence in the overall results? In order to address this issue, the Kristallin-l performance assessmentI has been compared with other HLW/spent fuel performance assessments. This technical comparison forms the basis of this paper which summarises the results from a more extensive comparison of a number of HLW/spent fuel disposal concepts, safety assessment methodologies and models2. It is part of a wider study which attempted to put the results of the Kristallin-I performance assessment into perspective by use of both this technical comparison with other performance assessments and by consideration of other information (e.g. natural analogue studies, environmental radiation data and risk assessments for radiological and non-radiological hazards) which allows the reasonableness of the results and significance of the predicted doses to be illustrated. Due to limitations of space, this aspect of the study has not been included here and the reader is referred to the “Kristallin-I: Results in Perspective” report2 for more details.

scholarly journals Is it Possible to Demonstrate Compliance with the Regulations for High-Level-Waste Repositories?

1992 ◽  
Vol 294 ◽  
Author(s):  
Felton W. Bingham
Keyword(s):  
Numerical Modeling ◽  
Data Collection ◽  
Performance Assessments ◽  
High Level Waste ◽  
Spent Fuel ◽  
Distant Future ◽  
Site Investigations ◽  
High Level ◽  
Waste Repositories ◽  
Shed Light

ABSTRACTThe regulations that currently govern repositories for spent fuel and high-level waste require demonstrations that are sometimes described as impossible to make. To make them will require an understanding of the current and the future phenomena at repository sites; it will also require credible estimates of the probabilities that the phenomena will occur in the distant future. Experts in many fields—earth sciences, statistics, numerical modeling, and the law—have questioned whether any amount of data collection can allow modelers to meet these requirements with enough confidence to satisfy the regulators. In recent years some performance assessments have begun to shed light on this question because they use results of actual site investigations. Although these studies do not settle the question definitively, a review of a recent totalsystem assessment suggests that compliance may be possible to demonstrate. The review also suggests, however, that the demonstration can be only at the “reasonable” levels of assurance mentioned, but not defined, in the regulations.

The SR 97 Safety Assessment of a KBS 3 Repository for Spent Nuclear Fuel – Overview, Review Comments and New Developments

2002 ◽  
Vol 713 ◽  
Author(s):  
Allan Hedin ◽  
Ulrik Kautsky ◽  
Lena Morén ◽  
Patrik Sellin ◽  
Jan-Olof Selroos
Keyword(s):  
Nuclear Fuel ◽  
Safety Assessment ◽  
Spent Nuclear Fuel ◽  
Bentonite Clay ◽  
Research Programme ◽  
High Strength ◽  
Functional Requirements ◽  
Future Evolution ◽  
Site Investigations ◽  
Potential Safety

ABSTRACTIn preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Nuclear Fuel and Waste Management Company, SKB has carried out the longterm safety assessment SR 97, requested by the Swedish Government. The repository is of the KBS-3 type, where the fuel is placed in isolating copper canisters with a high-strength cast iron insert. The canisters are surrounded by bentonite clay in individual deposition holes at a depth of 500 m in granitic bedrock. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock.The future evolution of the repository system is analysed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings, including climate, persist. The four other scenarios show the evolution if the repository contains a few initially defective canisters, in the event of climate change, in the event of earthquakes, and in the event of future inadvertent human intrusion.The principal conclusion of the assessment is that the prospects of building a safe deep repository for spent nuclear fuel in Swedish granitic bedrock are very good. The results of the assessment also serve as a basis for formulating requirements and preferences regarding the bedrock in site investigations, for designing a programme for site investigations, for formulating functional requirements on the repository's barriers, and for prioritisation of research.SR 97 has been reviewed both by an international group of OECD/NEA experts and by Swedish authorities. The NEA reviewers concluded that “SR 97 provides a sensible illustration of the potential safety of the KBS-3 concept”, and no issues were identified that need to be resolved prior to proceeding to the investigation of potential sites. The authorities' conclusions were in principal consistent with those of the NEA.Uncertainties and lack of knowledge in different areas identified in SR 97 have strongly influenced the contents and structure of SKBs most recent research programme, RD&DProgramme 2001.Since SR 97, the methodology for probabilistic consequence analyses have been further developed. Analytic approximations to the numerical transport models used in SR 97 have been developed. The new models have been used to extend the probabilistic calculations in SR 97.

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).

Uncertainty Quantification for the Development of Mechanistic Hydride Behavior Model for Spent Fuel Cladding Storage and Transportation

2020 ◽  
Author(s):  
Edward M. Duchnowski ◽  
Seokbin Seo ◽  
Nicholas R. Brown
Keyword(s):  
Temperature Profile ◽  
Total Concentration ◽  
Sensitivity Study ◽  
Diffusion Activation Energy ◽  
Local Concentration ◽  
Spent Fuel ◽  
Fuel Performance ◽  
Primary Coolant ◽  
Exponential Factor ◽  
The Impact

Abstract During operation of light water reactors hydrogen from the primary coolant is absorbed within the zirconium cladding and is able to migrate and redistribute within the cladding. The hydrogen in solid solution is able to precipitate, forming zirconium hydrides which results in a decrease of ductility of the cladding and ultimately an increase in likelihood of cladding failure, especially in conditions such as transport or storage of nuclear fuel rods. In collaboration with other universities, industries, and national laboratories the overarching goal of this project is to enhance the development for modeling hydride behavior to be implemented into the BISON fuel performance tool. The University of Tennessee-Knoxville (UTK) has been tasked with quantifying the uncertainty in the models developed within this project as well as quantifying the sensitivity to the most significant parameters of uncertainty. The BISON fuel performance code has been shown to overpredict the total concentration of hydrogen at cold regions of a temperature profile, thus a sensitivity study was performed to quantify the impact that key diffusion parameters have on the local concentration of hydrogen at the cold end of a 1-D model subjected to an asymmetric temperature profile. It is shown within this document that the diffusion activation energy and the pre-exponential factor, values within the diffusion coefficient of hydrogen, have a large impact on the local concentration of hydrogen and their importance increases for an increase in annealing time.

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 Safety assessment of spent fuel transportation in extreme environments

1982 ◽  
Vol 3 (1) ◽  
pp. 5-12 ◽  
Author(s):  
Robert P. Sandoval ◽  
John P. Weber ◽  
George J. Newton
Keyword(s):  
Safety Assessment ◽  
Extreme Environments ◽  
Spent Fuel

scholarly journals 1051 THE FUTURE INFORMATION SUPERHIGHWAY: HOW WIDE, HOW FAST, HOW ACCESSIBLE?

HortScience ◽  
1994 ◽  
Vol 29 (5) ◽  
pp. 579b-579
Author(s):  
Timothy J Ng
Keyword(s):  
High Speed ◽  
Information Services ◽  
Information Superhighway ◽  
Compressed Video ◽  
National Information Infrastructure ◽  
Future Evolution ◽  
Electronic Networks ◽  
Document Delivery ◽  
Other Information ◽  
Voice Data

The National Information Infrastructure (NII) initiative, more commonly known as the “information superhighway,” is intended to provide high-speed electronic access to a variety of voice, data, video, and other information services by the year 2000. Initial access to the superhighway will be prioritized for classrooms, libraries, hospitals and clinics. Eventually NII will provide for widespread and open availability of telecommunication and information services to everyone. Access to NII will be provided via electronic networks, telephone companies and cable operators. Many of the features proposed by NII are currently available in a developmental stage to horticulturists. The accessibility and future evolution of these services--which include those available via Internet (e.g., electronic messaging, data and information services, document delivery) and via telephone and cable companies (e.g. video/audio “on demand,” real-time polydirectional compressed video) --will be discussed.

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.

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