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首页> 外文期刊>Nuclear Technology >ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Units 2 and 3: German Contribution to the OECD/NEA BSAF Project, Phase 1
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ATHLET-CD/COCOSYS Analyses of Severe Accidents in Fukushima Daiichi Units 2 and 3: German Contribution to the OECD/NEA BSAF Project, Phase 1

机译：ATHLET-CD / COCOSYS分析福岛第一核电站2号和3号机组中的严重事故，德国对OECD / NEA BSAF项目第一阶段的贡献

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On behalf of the German Federal Ministry of Economics and Technology, Gesellschaft fur Anlagen- und Reaktorsicherheit (GRS) participated in the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) project titled Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF). Analysis of the severe accidents (SAs) that happened in the Fukushima Daiichi nuclear power plant (NPP) requires well-qualified methods and codes, e.g., ATHLET-CD and COCOSYS developed and applied at GRS. Coupled ATHLET-CD/ COCOSYS analyses for the SA progression during the first days for the similar Units 2 and 3 of Fukushima Daiichi have been provided as the German contribution to the OECD/NEA BSAF project, phase 1. ATHLET-CD is a detailed SA code based on the thermal-hydraulic code ATHLET of GRS to simulate the processes in the reactor circuit before and during core degradation. COCOSYS is focused on the simulation of design basis and SA progression in the containment and the surrounding buildings of the NPP. The focus is on selected results of the SA analyses in the boiling water reactors at the Fukushima Daiichi site especially with regard to the conditions in the torus-shaped wetwell (WW) of the primary containment and specific modeling needs. The GRS results obtained in this OECD/NEA BSAF project, phase 1, are encouraging in terms of capturing essential SA signatures like reactor and containment pressure, reactor water level, and WW temperature history for the first days of the accident in the analyzed Units 2 and 3. A detailed plant model was built up especially with a detailed torus nodalization allowing modeling of relevant phenomena like thermal stratification in the torus water pool and consideration of plant-specific details with regard to local water/steam injections into the torus water pool through safety systems and valves. As a result, the calculated accident progression of the best-estimate analyses for both units follows the accident time line quite closely. This is a prerequisite for reasonable core degradation calculations, as the time window available for the onset of core degradation between known points in time when safety injection stops and mobile pump injection into the reactor starts is small. The analyses are useful to identify areas that require further attention, to define information needs to be gained from the decommissioning, and to define further research needs with regard to experiments and code improvement.
机译：Gesellschaft fur Anlagen- und Reaktorsicherheit（GRS）代表德国联邦经济技术部参加了经济合作与发展组织/核能机构（OECD / NEA）题为“事故基准研究”的项目。福岛第一核电站（BSAF）。对福岛第一核电站（NPP）发生的严重事故（SA）进行分析需要合格的方法和规范，例如在GRS上开发和应用的ATHLET-CD和COCOSYS。作为德国对OECD / NEA BSAF项目第1阶段的贡献，已提供了福岛第一核电站第2和第3单元相类似的ATHLET-CD / COCOSYS对前几天SA进程的分析，这是德国对OECD / NEA BSAF项目第一阶段的贡献。该代码基于GRS的热工液压代码ATHLET来模拟堆芯退化之前和过程中电抗器回路中的过程。 COCOSYS专注于模拟基础的设计以及核电厂安全壳和周围建筑物的安全评估进展。重点是在福岛第一核电站的沸水反应堆中进行的SA分析的选定结果，特别是关于主要安全壳的圆环形湿井（WW）中的条件和特定的建模需求。 OECD / NEA BSAF该项目第一阶段获得的GRS结果令人鼓舞，因为在分析的单元2中捕获了事故发生第一天的基本SA特征，例如反应堆和安全壳压力，反应堆水位和WW温度历史记录。 3.建立了详细的工厂模型，特别是通过详细的环形节点化，可以对相关现象进行建模，例如环形水池中的热分层，以及考虑通过局部水/蒸汽注入环形水池的工厂特定细节。安全系统和阀门。结果，两个单位的最佳估计分析得出的事故进展非常接近事故时间线。这是进行合理堆芯退化计算的先决条件，因为安全注入停止和移动泵注入反应堆开始的已知时间点之间可用于发生堆芯退化的时间窗口很小。这些分析对于确定需要进一步关注的领域，定义退役中需要获得的信息以及确定有关实验和代码改进的进一步研究需求很有用。

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