Break Size Effect on the Transient Thermalhydraulic Behaviour During the Steam Generator Tube Rupture Accident
NURETH-14 - 2011 September 25-30


Presented at:
NURETH-14
2011 September 25-30
Location:
Toronto, Canada
Session Title:
B7-1 Steam Generators Thermalhydraulics

Authors:
Kyoung-Ho Kang (Korea Atomic Energy Research Institute)
Hyun-Sik Park (Korea Atomic Energy Research Institute)
Seok Cho (Korea Atomic Energy Research Institute)
Nam-Hyun Choi (Korea Atomic Energy Research Institute)
In-Cheol Chu (Korea Atomic Energy Research Institute)
Byong Jo Yun (Korea Atomic Energy Research Institute)
Kyung-Doo Kim (Korea Atomic Energy Research Institute)
Yeon-Sik Kim (Korea Atomic Energy Research Institute)
Won-Pil Baek (Korea Atomic Energy Research Institute)
Ki-Y   

Abstract

In order to simulate the SGTR accident of the APR1400, integral effect tests were performed by simulating a double-ended rupture of a single and five U-tubes. Following the reactor trip, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened to reduce the secondary system pressure. Break area affected the timings of the major events observed in the tests. Less heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow.

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