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首页> 外文会议>International meeting on severe accident assessment and management 2012: Lessons learned from fukushima dai-ichi >SIMULATION OF THE CORE-DEGRADATION PHASE OF THE FUKUSHIMA ACCIDENTS USING THE ASTEC CODE
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SIMULATION OF THE CORE-DEGRADATION PHASE OF THE FUKUSHIMA ACCIDENTS USING THE ASTEC CODE

机译：使用ASTEC代码模拟福岛事故的核心降解阶段

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摘要

IRSN, the French Institute for Nuclear Safety and Radioprotection, attempts to simulate the Fukushima accidents using the ASTEC code. The first analysis carried out concerned the reactor number 2 transient. Results were considered as satisfactory being quite consistent with measures reported by TEPCO and similar computations performed with MELCOR or MAAP. Knowledge gained from PWR practise and different lectures available in the open literature for BWR confirm the trends observed or provide some potential explanations to formulate alternative assumptions. Leakage model from the containment up to the refuelling bay through the head flange seal was very efficient to retrieve pressure evolution inside the dry well. Extension of the model to reactor number 3 gave also results quite coherent with what similar codes computed. However for both reactors some figures characteristic of the transient as hydrogen production, are liable to vary a lot if models for bottom and top nozzles are added which has not been done in reference computation due to initial lack of data. Uncertainties with simulation of accident on reactor number 1 are rather large due to scarcity of data. On the other hand as the measurement points were quasi absent for most of the first 24 hours there is no reference to compare to. Bottom head failure is predicted but due to the high number of penetrations the mechanical failure models developed for PWR may not be so relevant for BWR.
机译：法国核安全与辐射防护研究所IRSN试图使用ASTEC代码模拟福岛事故。进行的第一次分析涉及2号反应堆的瞬态。结果被认为是令人满意的，与TEPCO报告的测量结果非常一致，并且使用MELCOR或MAAP进行了类似的计算。从压水堆实践中获得的知识以及公开文献中有关压水堆的不同演讲，可确认观察到的趋势或提供一些潜在的解释以建立替代假设。从安全壳一直到顶部加油口密封之间的泄漏模型对于恢复干井内部的压力变化非常有效。将模型扩展到3号反应堆也得到了与相似代码计算结果相当一致的结果。但是，对于这两个反应器，如果添加底部和顶部喷嘴的模型，由于氢气的最初缺乏，因此在参考计算中未完成，因此这两个反应器的某些数字会随着氢气的产生而具有很大的变化。由于缺乏数据，模拟1号反应堆事故的不确定性相当大。另一方面，由于在前24小时的大部分时间内几乎没有测量点，因此无法进行比较。可以预测井底钻头的故障，但是由于穿透次数高，为PWR开发的机械故障模型可能与BWR不太相关。

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