Abstract A cross-comparison of four Generation IV reactor concepts and a PWR of 2nd generation is presented in this paper. 4 Gen IV reactor concepts are considered and are briefly presented in the first part of the paper: SFR of 1500 MWth, GFR of 2400 MWth, MSR of 3000 MWth and VHTR of 600 MWth. In order to perform this comparison, some simple common criteria related to accidental behavior of the reactors have been developed. The first kind of criteria aims at assessing the main physical thresholds to exceed in order to have a core degradation: phase changes of coolant and of core materials (including the effect of chemical reactions) for the various reactor concepts considered. The second set of criteria deals with kinetics aspects of the accident. On the basis of core power (after scram and without scram), on the coolant inventory and on the reactor capability to be passively cooled, the heating rate of the coolant and of the core materials are assessed thanks to simplified energy balances presented in the paper. As a result, for each reactor concept, the time to reach the physical thresholds defined above is evaluated. A third set of criteria deals with core features and aims at assessing the possible reactivity insertion that withstands each concept up to core melting (or boiling for the MSR) and the associated expected power peaks in case of coolant voiding/depressurization and in case of fissile material compaction. Finally, a last criterion set deals with the assessment of the possibility to jeopardize physical barriers confining fission products. These criteria deal with the risk of barrier loadings due to coolant and core material vaporization depending on the features of the coolant/fuel and on the operating point of each reactor concept. In the last part of the paper, a synthesis is made in order to underline the weak and strong points of each of the reactor concepts investigated in terms of severe accident prevention and mitigation.

中文翻译：

---

比较第四代核反应堆和 PWRS 事故行为的简化标准

摘要 本文介绍了四种第四代反应堆概念和第二代压水堆的交叉比较。本文第一部分考虑并简要介绍了 4 代第四代反应堆概念：1500 MWth 的 SFR、2400 MWth 的 GFR、3000 MWth 的 MSR 和 600 MWth 的 VHTR。为了进行这种比较，已经制定了一些与反应堆意外行为相关的简单通用标准。第一种标准旨在评估为使堆芯退化而超过的主要物理阈值：所考虑的各种反应堆概念的冷却剂和堆芯材料的相变（包括化学反应的影响）。第二组标准涉及事故的动力学方面。在核心功率的基础上（停堆后和不停堆），关于冷却剂存量和被动冷却的反应堆能力，由于文件中提出的简化的能量平衡，对冷却剂和堆芯材料的加热速率进行了评估。因此，对于每个反应堆概念，都会评估达到上述物理阈值的时间。第三组标准涉及堆芯特征，旨在评估可承受每个概念直至堆芯熔化（或 MSR 沸腾）的可能反应性插入以及在冷却剂排空/减压和裂变情况下相关的预期功率峰值材料压实。最后，最后一组标准涉及评估危害限制裂变产物的物理障碍的可能性。这些标准根据冷却剂/燃料的特性和每个反应堆概念的运行点处理由于冷却剂和堆芯材料汽化引起的屏障加载风险。在论文的最后一部分，进行了综合，以强调所研究的每个反应堆概念在严重事故预防和缓解方面的弱点和强项。