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Past and Future Research at IRSN on Corium Progression and Related Mitigation Strategies in a Severe Accident

机译：严重事故中IRSN的进展和相关缓解策略的过去和未来研究

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摘要

Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Surete Nucleaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core-concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors. This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions. In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis. This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management.
机译：在过去的三十年中，反应堆堆芯的降解以及容器内和容器外前皮质的行为一直是主要的研究课题，通过大型研究计划的协调或做出的贡献，放射性保护研究所和核仁保卫研究所（IRSN）对此做出了重要贡献严重事故（SA）ASTEC代码的开发和验证。近年来，研究工作之间的平衡趋向于分析利弊和评估缓解措施。描述了在法国进行的风险显着性分析（包括燃料-冷却剂相互作用（FCI），氢燃烧和熔融核芯-混凝土相互作用（MCCI）风险）的结果以及对皮质行为的研究。这些天的重点是（1）未来反应堆概念的容器内熔体保留（IVMR）策略，以及在现有反应堆中实施时必须确立此类策略的可靠性，以及（2）现有反应堆的内含钴冷却。本文总结了与理解和建模有关的主要成就和尚待解决的问题（1）退化核的回采，尽管通过研究计划获得了大量知识，但仍需要付出更多的努力来确定这种措施的效率以及与之相关的风险大大退化的核心； （2）反应堆压力容器（RPV）下端的皮质行为，尽管有经济合作与发展组织/核能机构（OECD / NEA）MASCA计划的成果，但仍需要做出努力来预测RPV产生的RPV热负荷有无IVMR措施（内部和/或外部冷却）时，皮质层的演变和RPV弹性； （3）尽管取得了OECD / NEA SERENA计划的结果，其FCI知识不足以自信地评估诱发的安全壳失效风险； （4）MCCI，有关通过顶部和/或底部水驱使安全壳中的钙冷却的知识不足以得出有关此类措施效率的结论。顶部洪水特别令人关注的是进水和珊瑚喷发过程。特别是对于顶部洪水，在反应堆条件下仍需量化进水和喷发过程的影响。为了支持这些活动，IRSN也正在进行大量努力，以不断改进和验证核材料特性数据库，这些数据库是进行皮质行为分析的关键工具。本文介绍了IRSN或国际组织在IRSN的协调或参与下正在进行的和将来的研究计划，以解决剩余的问题，并总结了SA代码建模，风险分析和SA管理方面的预期进展。

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  • 来源
    《Nuclear Technology》 |2016年第2期|161-174|共14页
  • 作者

    Didier Jacquemain; Didier Vola; Renaud Meignen; Jean-Michel Bonnet; Florian Fichot; Emmanuel Raimond; Marc Barrachin;

  • 作者单位

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

    Institut de Radioprotection et de Surete Nucleaire (IRSN), Nuclear Safety Division/Safety Research/Severe Accident Department, Saint Paul Lez Durance, 13115, France;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);美国《生物学医学文摘》(MEDLINE);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
  • 关键词

    In-vessel melt retention; fuel-coolant interaction; molten core-concrete interaction;

    机译：容器内熔体保留;燃料-冷却剂相互作用;熔融核-混凝土相互作用;

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