TRACE code core reflood thermal-hydraulics phenomena benchmarking against the NRC–PSU Rod Bundle Heat Transfer (RBHT) test facility

IF 1.9 3区 工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY
Grant R. Garrett , Douglas J. Miller , Turki Almudhhi , Fan-Bill Cheung , Brian R. Lowery , Molly K. Hanson , Stephen M. Bajorek , Kirk Tien , Chris L. Hoxie
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引用次数: 0

Abstract

This paper evaluates the performance of the U.S. Nuclear Regulatory Commission’s (NRC’s) thermal hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) against experimental reflood data from the NRC/Pennsylvania State University (NRC/PSU) Rod Bundle Heat Transfer (RBHT) test facility, as an integral step in verification of code accuracy. This paper is an extension of the NURETH-20 conference paper by the first author (Garrett et al., 2023) that has been recommended for consideration and submission to Nuclear Engineering and Design. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), was conducted with data collected in the NRC/PSU RBHT test facility, located at the Pennsylvania State University. A series of 16 benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. These 16 tests were segmented into 11 open tests, followed by five blind tests. This paper covers the five blind tests as the 11 open tests were covered by Garrett et al. at the NURETH-19 conference (Garrett et al., 2021).

For TRACE code benchmarking, a numerical model with the same dimensions as the RBHT facility was used. The initial and boundary conditions for this model were taken from experimental measurements. Many of the test conditions were chosen to examine sensitivities to important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. The wide range of test conditions served to test the code and provide insight to its strengths and potential areas of improvement. These novel experiments were vital in this effort.

Simulations were made for five reflood tests and comparisons between predicted and measured results were made for the transient cladding temperatures, vapor temperature, bundle liquid mass fraction, carryover fraction, and steam exhaust fraction. The comparison presented in this paper has provided useful insight into code improvements. Studies to more accurately model reflood phenomena are currently underway as a result of the work presented in this paper.

根据 NRC-PSU 棒束传热 (RBHT) 试验设施对 TRACE 代码堆芯回流热液压现象进行基准测试
本文评估了美国核管理委员会 (NRC) 的 TRAC/RELAP 高级计算引擎 (TRACE) 热液压代码与 NRC/宾夕法尼亚州立大学 (NRC/PSU) 棒束传热 (RBHT) 试验设施的实验回流数据的对比性能,作为验证代码准确性的一个不可或缺的步骤。本文是第一作者 NURETH-20 会议论文(Garrett 等人,2023 年)的延伸,该论文已被推荐给《核工程与设计》考虑和提交。由核能机构(NEA)事故管理与分析工作组(WGAMA)赞助的再充水热工水力学国际研究,利用位于宾夕法尼亚州立大学的 NRC/PSU RBHT 试验设施收集的数据进行。共进行了 16 次基准测试,测试条件涵盖了精心选择的振荡、可变阶跃和恒定速率再注水速度范围。这些独特的条件有助于验证代码和改进模型。这 16 次测试分为 11 次公开测试和 5 次盲测。本文介绍的是五次盲测,因为 Garrett 等人在 NURETH-19 会议上已经介绍了 11 次公开测试(Garrett 等人,2021 年)。为了进行 TRACE 代码基准测试,使用了与 RBHT 设施尺寸相同的数值模型。该模型的初始条件和边界条件取自实验测量结果。选择许多测试条件是为了检查重要参数的敏感性,如回流液体过冷度、回流速率和系统压力。广泛的测试条件有助于测试代码,并深入了解其优势和潜在的改进领域。对五次再充水试验进行了模拟,并对瞬态包层温度、蒸汽温度、管束液体质量分数、携带分数和蒸汽排出分数的预测结果和测量结果进行了比较。本文中的比较为改进代码提供了有益的启示。根据本文介绍的工作成果,目前正在研究如何更准确地模拟回流现象。
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
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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