Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT. (December 2018)

• Record Type: Journal Article
• Authors: Guo, Yuchuan; Wang, Guanbo; Qian, Dazhi; Yu, Heng; Hu, Bo; Guo, Simao; Mi, Xiangmiao; Ma, Jimin
• Is Part Of: Annals of nuclear energy. Volume 122(2018)
• Page Start: 125
• Page End: 136
• Publication Date: 2018-12
• Subjects: Flow blockage -- RELAP5/MOD3.4 -- FLUENT -- DNB; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2018.08.031
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: Flow blockage in an assembly in JRR-3M reactor is simulated combining RELAP5/MOD3.4 and FLUENT codes. It is necessary for the analysis of the assembly blockage to consider the power distribution in space. Accident consequence depends heavily on the blockage ratio. Abstract: The main purpose of the paper is to execute the thermal-hydraulic analysis of flow blockage in a single fuel assembly in the JRR-3M research reactor (Upgraded Japan Research Reactor No. 3). Using the one-dimensional system analysis code RELAP5/MOD3.4, flow region in a standard fuel assembly (SFA) is lumped into a coolant channel. 5 similar coolant channels representing 5 independent assemblies are modeled to consider the interaction between blocked channel and adjacent channels. The rest of the core is lumped into 1 average channel and 1 bypass channel. The coolant system is also modeled in detail. Meanwhile, a three-dimensional model of heated part in the assembly is built using the computational fluid dynamics (CFD) code FLUENT. Results calculated by RELAP5/MOD3.4 are used as the boundary conditions of the 3D model. The user-defined function (UDF) is adopted to describe phase change and the power distribution in axial and radial direction in the assembly. Results indicate that it is necessary for assembly blockage to consider the influence of power distribution on accident consequence. When the blockage ratio is 64%, coolant in the hottest subchannel is still supercooled while coolant temperature at outlet is close to saturation temperature. It is the critical blockage ratio to ensure the reactor safety. When the blockage ratio is 70%, departure from nucleate boiling (DNB) will occur. According to the FLUENT code, the process of bubble generation and growth is discussed. It can be found that if the bubbles largely generate, there will be obvious impacts on heat transfer of fuel plate and coolant flow. Fuel plates may damage locally.