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Thermal-Hydraulic Analyses of Steam Generator Tube Rupture Accident for the Kori Nuclear Unit 1 Pressurized Thermal Shock Study
Volume 138 · Number 3 · June 2002 · Pages 273-283
Technical Paper · Thermal Hydraulics
Soon-Joon Hong, Jae-Hak Kim, Yong-Soo Kim, Goon-Cherl Park
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This paper discusses a thermal-hydraulic analysis methodology using RETRAN-3D and assembles system analyses for pressurized thermal shock resulting from a steam generator tube rupture accident in Kori Nuclear Unit 1. Through a systematic definition of sequences and thermal-hydraulic analyses using RETRAN-3D, the most important parameters on downcomer overcooling were identified. The break location that leads to the most significant overcooling was found to be the hot leg side in the loop that does not contain the charging flow inlet. The initial power level had a large effect on the downcomer overcooling. The closure failure of the pressurizer power operated relief valves and the termination failure of the safety injection were found to be the most significant operator actions. In contrast, auxiliary feedwater control failure had little effect on overcooling, and the steam dump valve closure failure merely resulted in a temperature rise in the latter half of the transient. Through these analyses, recommendations for sequence grouping and against downcomer overcooling are provided.
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