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[en] How Source Term Evaluation is addressed? • Understanding important challenges to the integrity of the fuel, reactor pressure vessel and the containment requires system analysis and models of plant response to various SA sequences; • A realistic modelling of severe accident source term requires modelling of a wide range of phenomena associated with core melt progression, containment performance and fission product release and transport; • The amount, composition, chemistry, and timing of fission product release through the RCS and into containment result in the in-containment source term; • Estimating the transport of source term requires understanding the phenomena which can challenge the integrity of the fuel cladding, the RPV and the containment under harsh conditions; • These factors affect on-site and off-site consequence analysis and possible protective actions which need to be planned for; • The source term may be also affected by the type of fuel, its amount and its burn-up; • Effectiveness of design feature (e.g., venting) and measures (e.g., filtration) can be determined with validated codes; • Many national and international programmes have been undertaken to address severe accidents and hence source term evaluation; their results have been shared through international networks (e.g., SARNET); • CSNI played a major role in promoting and organizing cooperative research projects, International Standard Problems (ISPs), Benchmarking activities and Workshops in the area of severe accidents since TMI-2 accident. • Source Term evaluation addressed within: – The OECD Joint Projects in Nuclear Safety; – The International Standard Problems; – The Benchmarking Activities; – The Workshops; – Reports, in particular State-of-the-Art Reports (SOARs).
[en] The OECD/NEA paved the way for the development and assessment of BEPU for about 40 years, through concrete tasks: International Standard Problems (ISPs), Benchmarking activities, Development of Validation Matrices, Joint Safety Research Projects, and Specialist meetings. Several NEA related Best-Estimate Plus Uncertainties (BEPU) programmes have been successfully completed: Uncertainty Methods Study (UMS), Best-Estimate Methods - Uncertainty and Sensitivity Evaluation (BEMUSE), Safety Margin Assessment and Application (SM2A), Uncertainty Analysis in Modeling (UAM) Benchmark. New Programmes are underway to address pending issues (e.g., input uncertainties, uncertainties in coupled codes). The present Workshop may highlight new issues to be addressed (e.g., uncertainty analysis for CFD codes). Document available in the slides-form only
[en] The purpose of the Technical Meeting is to provide a platform for detailed presentations and technical discussions on recent progress in R&D activities on in-vessel melt retention (IVMR) and ex-vessel corium cooling (EVCC) during severe accidents at water-cooled reactors (WCRs).
[en] Expected outputs: 1. Meeting webpage: • to introduce the meeting to Member States’ experts on the IAEA website; • TM presentations and major outcomes will be shortly uploaded after the meeting. 2. TM Report: • A draft report will be prepared by the Scientific Secretaries and is expected to be reviewed by all participants. 3. IAEA Technical Document (IAEA TECDOC) will: • summarize in detail the presentations and the discussions; • provide meeting conclusions and recommendations; • include all the presentations in an attached CD; • be distributed to all the participants for checking; • be subject to IAEA QA; • be distributed to all who are interested in Member States.
[en] The Nuclear Energy Agency (NEA)/Committee for the Safety of Nuclear Installations' (CSNI) Working Group on Risk Assessment (WGRISK) is tasked with supporting the improved use of Probabilistic Safety Assessment (PSA) in risk informed regulation and safety management through the analysis of results and the development of perspectives regarding potentially important risk contributors and associated risk reduction strategies. The task consists of the following major activities: Development, distribution, and completion of survey questionnaires; Analysis of survey questionnaire results at a task workshop; Preparation of the final task report. The main objectives of this task, as proposed by WGRISK and approved by CSNI, are the following: - Identification and characterization of the current uses of OECD data project products and data in support of PSA. In this context, the term 'products' refers to data analysis results, technical reports, and other project outputs. - Identification and characterization of technical and programmatic characteristics that either support or impede use of data project products in PSA. This includes an assessment of which PSA parameters could be potentially estimated from the various data project products and gaps between available product information and PSA data needs. - Identification of recommendations for enhancing the usefulness of data project products and the coordination between WGRISK and the data projects. This task report consists of the following sections: - Chapter 1 Provides a general overview of motivation and approach used for this task. - Chapter 2 Describes scope and objectives of the task. - Chapter 3 Provides an overview of the ICDE, FIRE, OPDE/CODAP, and COMPSIS data projects. For each project, the project objectives, project history, data collection methodology and quality assurance, project status, example PSA Applications, and information related to project participation is provided. - Chapter 4 Describes the methodology used for this task, including a more detailed description of the survey questionnaire and task group meeting. - Chapter 5 Summarizes the analysis of survey responses, including discussion of data challenges and best practices. Also included is a discussion of enhancing project participation, new data and analysis needs, data project success factors for PSA applications, and summary of key issues and potential resolutions. - Chapter 6 Provides a summary of key conclusions from the surveys and task group meeting. - Chapter 7 Summarizes key recommendations. - Appendices Several appendices are provided to provide more detailed information on the CSNI activity proposal sheet governing this task, copies of the surveys sent to WGRISK and Data Project representatives, summary results from each survey, the complete survey responses provided by each responding organization, and contact information for task participants. The following information is provided in the Appendices: - Appendix A: CSNI Activity Proposal Sheet WGRISK (2011)-1, 'Use of OECD Data Project Products in Probabilistic Safety Assessment (PSA)'; - Appendix B: Survey questionnaire for WGRISK members and observers; - Appendix C: Survey questionnaire for OECD joint data project representatives; - Appendix D: Summary of OECD joint data project publicly available information; - Appendix E: Summary of WGRISK member and observer responses; - Appendix F: Survey responses from OECD joint data project representatives; - Appendix G: Complete set of WGRISK member and observer survey questionnaire responses; - Appendix H: Contact information for task participants
[en] This revision of the Knowledge Base for Emergency Core Cooling System Recirculation Reliability (NEA/CSNI/R (95)11) describes the current status (late 2012) of the knowledge base on emergency core cooling system (ECCS) and containment spray system (CSS) suction strainer performance and long-term cooling in operating power reactors. New reactors, such as the AP1000, EPR and APR1400 that are under construction in some Organization for Economic Co-operation and Development (OECD) member countries, are not addressed in detail in this revision. The containment sump (also known as the emergency or recirculation sump in pressurized water reactors (PWRs) and pressurized heavy water reactors (PHWRs) or the suppression pools or wet wells in boiling water reactors (BWRs)) and associated ECCS strainers are parts of the ECCS in both reactor types. All nuclear power plants (NPPs) are required to have an ECCS that is capable of mitigating a design basis accident (DBA). The containment sump collects reactor coolant, ECCS injection water, and containment spray solutions, if applicable, after a loss-of-coolant accident (LOCA). The sump serves as the water source to support long-term recirculation for residual heat removal, emergency core cooling, and containment atmosphere clean-up. This water source, the related pump suction inlets, and the piping between the source and inlets are important safety-related components. In addition, if fibrous material is deposited at the fuel element spacers, core cooling can be endangered. The performance of ECCS/CSS strainers was recognized many years ago as an important regulatory and safety issue. One of the primary concerns is the potential for debris generated by a jet of high-pressure coolant during a LOCA to clog the strainer and obstruct core cooling. The issue was considered resolved for all reactor types in the mid-1990s and the OECD/NEA/CSNI published report NEA/CSNI/R(95)11 in 1996 to document the state of knowledge of ECCS performance at that time. Subsequent to the publication of NEA/CSNI/R(95)11, a number of new issues (e.g., chemical effects, downstream effects and long-term effects) have been identified that have reopened the topic of strainer performance. This revised knowledge-base document has been developed to update the knowledge base by incorporating the considerable quantity of research completed, and the lessons learned, since 1996. It was recognized from the beginning that differences in the issue status and the methods (regulatory aspects, resolution of issues and research and development actions) used to address the strainer clogging remained a challenge, and the NEA Sump Clogging Task Team chose to focus on generic issues. The present report includes not only an update of the previous information, but also two new topics on chemical effects and downstream effects. In addition, while NEA/CSNI/R(95)11 focused on BWRs, the present update includes a significant amount of new information related to PWRs, leading in particular to a very much expanded Appendix on 'Experimental Investigations and Test Facilities'. This document was prepared by the NEA Sump Clogging Task Team
[en] This report presents the results of a joint task of the Working Groups on Risk Assessment (WGRISK) and on Human and Organisational Factors (WGHOF) of the OECD/NEA CSNI, to identify desirable attributes of Human Reliability Assessment (HRA) methods, and to evaluate a range of HRA methods used in OECD member countries against those attributes. The purpose of this project is to provide information that will support regulators and operators of nuclear facilities when making judgements about the appropriateness of HRA methods for conducting assessments in support of Probabilistic Safety Assessments (PSA). The task was performed by an international team of Human Factors, HRA and PSA experts from a broad range of OECD member countries. As in other reviews of HRA methods, the study did not set out to recommend or promote the use of any particular HRA method. Rather the study aims to identify the strengths and limitations of commonly used and developing methods to aid those responsible for production of HRAs in selecting appropriate tools for specific HRA applications. The study also aims to assist regulators when making judgements on the appropriateness of the application of an HRA technique within nuclear-related probabilistic safety assessments. The report is aimed at practitioners in the field of human reliability assessment, human factors, and risk assessment more generally
[en] The Committee on the Safety of Nuclear Installations (CSNI) formed the CCVM (Containment Code Validation Matrix) task group in 2002. The objective of this group was to define a basic set of available experiments for code validation, covering the range of containment (ex-vessel) phenomena expected in the course of light and heavy water reactor design basis accidents and beyond design basis accidents/severe accidents. It was to consider phenomena relevant to pressurised heavy water reactor (PHWR), pressurised water reactor (PWR) and boiling water reactor (BWR) designs of Western origin as well as of Eastern European VVER types. This work would complement the two existing CSNI validation matrices for thermal hydraulic code validation (NEA/CSNI/R(1993)14) and In-vessel core degradation (NEA/CSNI/R(2001)21). The report initially provides a brief overview of the main features of a PWR, BWR, CANDU and VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher). It then provides a general overview of the accident progression for light water and heavy water reactors. The main focus is to capture most of the phenomena and safety systems employed in these reactor types and to highlight the differences. This CCVM contains a description of 127 phenomena, broken down into 6 categories: - Containment Thermal-hydraulics Phenomena; - Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena; - Aerosol and Fission Product Behaviour Phenomena; - Iodine Chemistry Phenomena; - Core Melt Distribution and Behaviour in Containment Phenomena; - Systems Phenomena. A synopsis is provided for each phenomenon, including a description, references for further information, significance for DBA and SA/BDBA and a list of experiments that may be used for code validation. The report identified 213 experiments, broken down into the same six categories (as done for the phenomena). An experiment synopsis is provided for each test. Along with a test description and references, the synopsis also identifies the availability of the report and data, phenomena covered by the test, type of test (separate effect, combined effect or integral test), covers DBA and/or SA/BDBA conditions, range of key experimental parameters and past code validation/ benchmarks. This CCVM has identified experiments for 93% of the phenomena requiring validation. However, if only experiments suitable for CFD validation are considered, then only about half of the phenomena are covered by this CCVM. It is recommended that this work be reviewed in 5 years time to include new experiments and to attempt to close the identified experiment gaps (phenomena lacking suitable experiments for validation). (authors)