SMART100设计中反应堆冷却剂系统流量完全损失的整体效应试验及其代码模拟

IF 2.1 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Jin-Hwa Yang, Byong-Guk Jeon, Hwang Bae, Hyun-Sik Park
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引用次数: 0

摘要

利用SMART-ITL的综合效应测试设备,成功地测试了反应堆冷却剂系统流量完全损失(CLOF)的情景。稳态条件较好地满足了试验要求中提出的初始试验条件;精确地模拟了其边界条件,并在SMART100设计中适当地再现了CLOF场景。CLOF测试进行了长达60000 s的时间,以了解SMART100设计的被动式余热排出系统(PRHRS)的长期行为。使用MARS-KS代码分析了SMART-ITL的测试结果,以评估其模拟SMART100设计的CLOF场景的能力。由于被动安全系统的运行时间不少于72小时,因此精确预测SMART100反应堆冷却剂系统(RCS)和被动安全系统的长期热水力行为对SMART100设计的安全性分析至关重要。使用SMART-ITL对CLOF测试中测量的热液数据进行了MARS-KS代码的适当模拟,表明其具有合理的模拟能力,可以模拟带有PRHRS的SMART100设计的CLOF场景。在长期模拟过程中,热液现象更多地取决于RCS和PRHRS回路中的流动阻力以及RCS、蒸汽发生器、PRHRS回路和应急冷却罐的热损失。认为适当考虑回路内的流动阻力和热结构的热损失是可以进行精确模拟的。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
An integral effect test and its code simulation on the complete loss of reactor coolant system flowrate for the SMART100 design
A complete loss of reactor coolant system flowrate (CLOF) scenario was successfully tested using the integral effect test facility of SMART-ITL. The steady-state conditions were well achieved to satisfy initial test conditions presented in the test requirement; its boundary conditions were accurately simulated, and the CLOF scenario was reproduced properly for the SMART100 design. The CLOF test was performed over a long period of 60,000 s to understand the long-term behavior of the passive residual heat removal system (PRHRS) of the SMART100 design. The test results from the SMART-ITL were analyzed using the MARS-KS code to assess its capability to simulate a CLOF scenario for the SMART100 design. As the passive safety systems will operate for no less than 72 h to fulfill their function, the precise prediction of thermal–hydraulic behaviors in reactor coolant system (RCS) and passive safety systems of the SMART100 in the long-term sense is very crucial for the safety analyses of the SMART100 design. The measured thermal–hydraulic data from the CLOF test using the SMART-ITL were properly simulated using the MARS-KS code, which showed its reasonable simulation capability for the CLOF scenario of the SMART100 design with PRHRS. The thermal–hydraulic phenomena depend more on flow resistance both in RCS and PRHRS loops and heat losses from the RCS, steam generator, PRHRS loop and emergency cooldown tank during the long-term simulation. It is considered that accurate simulation is possible with proper consideration on flow resistance in the loops and heat loss through the heat structure.
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来源期刊
Nuclear Engineering and Design
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.
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