Uncertainty Quantification of the RELAP5 Interfacial Friction Model in the Rod Bundle Geometry

Author:

Kinoshita Ikuo1,Torige Toshihide1,Yamada Minoru2

Affiliation:

1. Institute of Nuclear Safety System, Inc. (INSS), 64 Sata, Mihama-cho, Mikata-gun, Fukui 919-1205, Japan e-mail:

2. MHI Nuclear Engineering Co., Ltd. (MNEC), 3-3-1 Minatomirai, Nishi-ku, Yokohama, Kanagawa 220-8401, Japan e-mail:

Abstract

Interfacial friction in the core affects the two-phase mixture level and the distribution of the dispersed gas phase during a small-break loss-of-coolant accident (LOCA). The RELAP5/MOD3.2 code uses the drift flux model to describe the interfacial friction force in vertical dispersed flow, and the Chexal–Lellouche drift flux correlation is used for the rod bundle geometry. In the present study, the RELAP5 model uncertainty was quantified for the bubbly–slug interfacial friction model in the rod bundle geometry by conducting numerical analyses of void profile tests in the Thermal Hydraulic Test Facility (THTF) of the Oak Ridge National Laboratory (ORNL). The model uncertainty parameter was defined as a multiplier for the interfacial friction coefficient. Numerical analyses were performed by adjusting the multiplier so that the predicted void fractions agreed with the measured test data. The resultant distribution of the multipliers represented the interfacial friction model uncertainty.

Publisher

ASME International

Subject

Nuclear Energy and Engineering,Radiation

Reference13 articles.

1. Intentional Depressurization of Steam Generator Secondary Side during a PWR Small-Break Loss-of-Coolant Accident;J. Nucl. Sci. Technol.,1995

2. Core Liquid Level Responses Due to Secondary-Side Depressurization during PWR Small Break LOCAs;J. Nucl. Sci. Technol.,1998

3. Secondary-Side Depressurization during PWR Cold-Leg Small Break LOCAs Based on ROSA-V/LSTF Experiments and Analyses;J. Nucl. Sci. Technol.,1998

4. Effects of Secondary Depressurization on Core Cooling in PWR Vessel Bottom Small Break LOCA Experiments With HPI Failure and Gas Inflow;J. Nucl. Sci. Technol.,2006

5. Quantifying Reactor Safety Margins: Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large-Break Loss-of-Coolant Accident,1989

Cited by 1 articles. 订阅此论文施引文献 订阅此论文施引文献,注册后可以免费订阅5篇论文的施引文献,订阅后可以查看论文全部施引文献

同舟云学术

1.学者识别学者识别

2.学术分析学术分析

3.人才评估人才评估

"同舟云学术"是以全球学者为主线,采集、加工和组织学术论文而形成的新型学术文献查询和分析系统,可以对全球学者进行文献检索和人才价值评估。用户可以通过关注某些学科领域的顶尖人物而持续追踪该领域的学科进展和研究前沿。经过近期的数据扩容,当前同舟云学术共收录了国内外主流学术期刊6万余种,收集的期刊论文及会议论文总量共计约1.5亿篇,并以每天添加12000余篇中外论文的速度递增。我们也可以为用户提供个性化、定制化的学者数据。欢迎来电咨询!咨询电话:010-8811{复制后删除}0370

www.globalauthorid.com

TOP

Copyright © 2019-2024 北京同舟云网络信息技术有限公司
京公网安备11010802033243号  京ICP备18003416号-3