柱形气冷微堆整体热水力装置设计与预试分析

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Zheng Huang , Miaoxin Jiao
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引用次数: 0

摘要

为了支持新型气冷微堆(GMR)的研发,设计了一个完整的热水力试验设施(名为INTALI),以探索GMR特有的关键热水力现象,从而为验证热水力和事故瞬态分析的计算机代码提供必要的数据。本文介绍了INTALI设备的初步设计、实验方法和测试前分析。INTALI设施由一个在原型温度和压力下运行的主回路和一个包含按比例缩小的模拟反应堆和一个被动堆芯冷却系统(PCCS)的测试部分组成。针对GMR的正常运行工况和强制冷却剂压力损失(PLOFC)事故工况,分别进行了稳态和瞬态试验。实验主要研究:(1)反应器与PCCS之间的耦合,特别是在PLOFC期间;(2)PCCS的运行特性和能量分布;(3)PCCS内部自然循环引起的重力方向的潜在热分层。利用COMSOL Multiphysics软件进行CFD模拟,对实验进行了预试分析。利用预测的反应器和PCCS的温度场和速度场的三维分布来确定仪器方案。模拟结果表明,RPV壁面没有明显的垂直热梯度。除了对流外,RPV向PCCS保温层的辐射传热在散热中也起着重要的作用。在PLOFC瞬态过程中,RPV的温度对PCCS的散热能力有显著影响。一旦获得实验数据,所开发的CFD模型也可以用于测试后的量化和验证。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
Design and pre-test analyses of an integral thermal–hydraulic facility for a prismatic gas-cooled micro reactor
To support the R&D of the novel gas-cooled micro reactor (GMR), an integral thermal–hydraulic test facility (named INTALI) is designed to explore the key thermal–hydraulic phenomena specific to the GMR, thereby providing necessary data to validate computer codes for thermal–hydraulic and accident transient analysis. This paper presents a preliminary design of the INTALI facility, experimental methodology, and pre-test analyses. The INTALI facility consists of a primary loop operating at the prototypical temperature and pressure and a test section containing a scaled-down simulated reactor and a passive core cooling system (PCCS). Steady-state and transient tests will be carried out, which correspond to the normal operation and the pressurized loss of forced coolant (PLOFC) accident condition of the GMR, respectively. The experiment is mainly to investigate: (i) the coupling between the reactor and the PCCS, especially during the PLOFC, (ii) the operational characteristics of the PCCS and the energy distribution, and (iii) potential thermal stratification in the gravitational direction caused by the natural circulation in the PCCS. The pre-test analyses of the experiment were performed by CFD simulations using the COMSOL Multiphysics software. The predicted 3D distributions of the temperature and velocity fields for both the reactor and the PCCS are used to determine the instrumentation scheme. The simulation results show that no significant vertical thermal gradient is observed on the RPV wall. The radiative heat transfer from the RPV to the PCCS insulation layer plays an important role in heat removal in addition to convection. The heat removal capability of the PCCS is significantly influenced by the RPV’s temperature during the PLOFC transient. The developed CFD model is also ready for post-test quantification and validation once the experimental data is available.
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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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