Analytical studies for hydrogen distribution & management for 700 MWe IPHWR

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2025-04-21 DOI:10.1016/j.nucengdes.2025.114053

Sanjeev Kumar Sharma , D.K. Chandraker , Manvendra Singh , Vibha Hari , Sameer Hajela

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Abstract

Hydrogen generation and local accumulation of hydrogen in the containment atmosphere during a postulated accident scenario involving multiple failures could pose a threat to the integrity of the containment as the hydrogen can form a flammable mixture with air in the containment. One such unlikely postulated accident sequence is Large Break Loss–Of-Coolant Accident (LBLOCA) along with failure of Emergency Core Cooling System (ECCS). Moderator cooling system works as the heat sink for such accident sequences and hydrogen generation will be limited. Further as a part of Defense-In-Depth (DID) approach, the containment response has been evaluated under more severe conditions where failure of moderator cooling system is also postulated. During such a severe accident large amounts of hydrogen are expected to get generated.

A system thermal hydraulic computer code, Post Accident Containment System Response (Singh et al., 2023, Sharma et al., 2024) has been developed for the containment response calculation during normal operation as well as accident conditions including, severe accident.. The calculations for hydrogen management by using a combination of Passive Catalytic Recombiner Devices (PCRD) and provision of Passive Opening or Forced Mixing are performed for 700 MWe Indian Pressurized Heavy Water Reactors (IPHWRs) by using in house computer code PACSR- SI2.0. This paper presents the studies performed to assess the hydrogen behaviour along with optimised mitigating features during severe accident conditions. From the analysis, it is found that, with optimized number of PCRDs; hydrogen can be effectively managed during severe accident conditions. Lumped parameter approach is considered to be a good technique for the finalisation of the optimised number of PCRD,which can be further verified using 3-D CFD approach.

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700mwe水冷堆氢气分配与管理分析研究

在涉及多次故障的假想事故情景中，氢气的产生和氢气在安全壳大气中的局部积聚可能对安全壳的完整性构成威胁，因为氢气可以与安全壳中的空气形成可燃混合物。其中一个不太可能发生的事故序列是伴随紧急堆芯冷却系统（ECCS）失效的冷却剂大断裂损失事故（LBLOCA）。慢化剂冷却系统作为这种事故序列的散热器，氢的产生将受到限制。此外，作为纵深防御（DID）方法的一部分，在假定慢化剂冷却系统失效的更严重条件下，对安全壳响应进行了评估。在如此严重的事故中，预计会产生大量的氢气。已经开发了一个系统热液压计算机代码，事故后遏制系统响应（Singh等人，2023年，Sharma等人，2024年），用于正常运行以及事故条件下的遏制响应计算，包括严重事故。采用PACSR- SI2.0对700mwe印度压重水堆（IPHWRs）进行了无源催化重组装置（PCRD）和无源开启或强制混合装置组合的氢气管理计算。本文介绍了在严重事故条件下进行的研究，以评估氢的行为以及优化的缓解特征。分析发现，随着pcrd数量的优化；氢气可以在严重事故条件下有效管理。集总参数法被认为是确定PCRD优化数量的一种很好的技术，可以通过三维CFD方法进一步验证。

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.