The presented paper discuses a study of core degradation behavior where a Medium Break Loss of Coolant Accident (MB LOCA) with a150 mm internal diameter along with station blackout (SBO) was the initiating event. One of the goals of this work was the examination of core degradation at different primary side break locations.

Two simulations of MB-LOCA scenarios were performed. The first analysis was performed with the break located in the hot leg and the other with the break in the cold leg. The purpose of the selected study was to investigate plant response with breaks of similar sizes located in different positions in the reactor.

This study was focused on in-vessel phase phenomena, such as: core uncovery, core heat up, the beginning of hydrogen generation based on the steam zirconium reaction and the oxidation of other core materials, core degradation, fuel cladding failure, melting of core materials with the formation of a molten pool in the reactor core, relocation of core materials to the bottom of the reactor vessel, and the formation of a molten pool containing corium in the lower head with further vessel failure.

The investigation demonstrates that the scenario with a break in the hot leg leads to faster HAs depletion and faster core degradation during first part of the accident with delays in the accidents progression during the later stages of the accident resulting in a later failure of the reactor vessel head due to better core cooling when the results are compared to the evaluation of the accident with a cold leg break.

The investigation of the severe accident sequence induced by a MB-LOCA was performed using the computer code ASTEC (Accident Source Term Evaluation Code) v2.1.1.6. The referenced nuclear power plant considered for this analysis is a water-water energetic (WWER)-1000 reactor.

The results of the paper could be used for improving severe accident management guidlines (SAMG) as well as for level 2 probabilistic safety analysis (PSA).

提出的论文讨论了核心退化行为的研究，其中，内径为150 mm的冷却剂中断裂损失事故（MB LOCA）以及停电事故（SBO）是引发事件。这项工作的目标之一是检查在不同的一次侧断位置处的堆芯退化。

进行了MB-LOCA方案的两个模拟。第一次分析是在热腿处有断裂，而另一次是在冷腿处有断裂。所选研究的目的是研究位于反应堆不同位置的类似大小断裂的植物反应。

这项研究的重点是船内相现象，例如：堆芯发现，堆芯加热，基于蒸汽锆反应和其他堆芯材料的氧化而开始产生氢，堆芯退化，燃料包层失效，堆芯熔化在反应堆堆芯中形成熔池，将堆芯材料转移到反应堆容器底部以及在下部封盖中形成含钴的熔池导致容器进一步损坏。

调查表明，热段断裂的情况导致事故第一部分的HA损耗更快，堆芯退化更快，事故后期阶段的事故进展延迟，导致反应堆容器后来发生故障。将结果与比较冷腿断裂事故的评估结果相比较时，由于更好的堆芯冷却而使头部变形。

使用计算机代码ASTEC（事故源术语评估代码）v2.1.1.6对MB-LOCA引发的严重事故序列进行了调查。考虑进行此分析的参考核电站是水-水高能（WWER）-1000反应堆。

本文的结果可用于改进严重事故管理指南（SAMG）以及用于2级概率安全分析（PSA）。