Abstract In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a Reactor Vault Cooling System (RVCS) is prepared to maintain the structural integrity of the reactor vessel and the concrete during normal operation as well as to mitigate accidents during severe accident conditions. In this research a series of analysis was described during the RVCS design refinement process. The one-dimensional design code analysis was used to examine various design parameters that are able to enhance RVCS cooling capability without deteriorating RVCS performance in severe accident condition. As a results RVCS gap distance was decreased by 10 mm. Multi-dimensional thermal–hydraulic CFD analysis were utilized to analyze the real geometry effects on flow field. Based on the analysis, RVCS design refinement has been done within an acceptable range. In order to evaluate the RVCS design refinement to design requirement under more realistic condition, a Head Access Area (HAA)-RVCS integrated reactor system model was built and an integrated simulation was performed at normal operating conditions. It was found that proposed RVCS fulfills the design requirements of normal operation condition.

中文翻译：

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PGSFR反应堆拱顶冷却系统的设计与评价

摘要 在原型第四代钠冷快堆（PGSFR）中，准备了反应堆拱顶冷却系统（RVCS），以在正常运行期间保持反应堆容器和混凝土的结构完整性，并在严重事故中减轻事故。使适应。在本研究中，描述了 RVCS 设计改进过程中的一系列分析。一维设计规范分析用于检查各种设计参数，这些参数能够在严重事故工况下提高 RVCS 冷却能力而不恶化 RVCS 性能。结果 RVCS 间隙距离减少了 10 毫米。多维热工水力 CFD 分析被用来分析实际几何对流场的影响。根据分析，RVCS 设计改进已在可接受的范围内完成。为了在更现实的条件下评估 RVCS 设计改进是否符合设计要求，建立了水头接入区 (HAA)-RVCS 集成反应堆系统模型，并在正常运行条件下进行了集成模拟。结果表明，提出的 RVCS 满足正常运行条件的设计要求。