Safety margin predictions of reactor pressure vessel integrity assessment procedures for pressurized thermal shocks

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2025-09-04 DOI:10.1016/j.nucengdes.2025.114435

Klaus Heckmann, Jens Arndt, Jürgen Sievers, Ulrike Läuferts, Uwe Jendrich, Sara Beck, Frank Michel

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Abstract

The integrity assessment of a reactor pressure vessel in a loss of coolant scenario with pressurized thermal shock (PTS) loading of the vessel wall is a crucial part of the safety review for the long-term operation, since the embrittlement of the vessel steel due to neutron irradiation continues during the ongoing operation. The procedures for the safety demonstration in different countries select different options for this assessment, namely concerning crack postulates, weld residual stresses, fracture toughness model, and warm pre-stress model. In order to understand the impact of these differences, the deterministic procedures of eight European countries are applied to two different example cases, which are based on the Konvoi and the VVER-440 plant type but are partly fictitious and not reflecting any particular unit. The safety margin in terms of ductile–brittle transition temperature is computed for each assessment, allowing to draw conclusions regarding the impact of different options within an assessment. In addition, it is possible to identify general trends of the procedures concerning their safety margin prediction.

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压力热冲击下反应堆压力容器完整性评估程序的安全裕度预测

由于中子辐照导致的容器钢脆性在持续运行过程中持续存在，因此，在冷却剂损失情况下，对容器壁进行加压热冲击（PTS）负荷的反应堆压力容器的完整性评估是长期运行安全审查的关键部分。不同国家的安全论证程序选择了不同的评估方案，即裂纹假设、焊缝残余应力、断裂韧性模型和热预应力模型。为了了解这些差异的影响，将八个欧洲国家的确定性程序应用于两个不同的案例，这两个案例基于Konvoi和VVER-440电站类型，但部分是虚构的，不反映任何特定的单元。根据每次评估计算韧性-脆性转变温度的安全裕度，从而得出关于评估中不同选项影响的结论。此外，还可以确定有关其安全裕度预测的程序的一般趋势。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Nuclear Engineering and Design 工程技术-核科学技术

CiteScore

3.40

自引率

11.80%

发文量

377

审稿时长

5 months

期刊介绍： Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.