REACTOR COOLANT SYSTEM BRANCH LINE PIPE BREAK ANALYSIS FOR SHIN-KORI UNITS 3&4
Shin-Kori Units 3&4 (SKN 3&4) are the first APR1400 (Advanced Power Reactor 1400 MWe) nuclear power plants being under construction in Korea. The capacity and size of main components in the APR1400 nuclear steam supply system have been increased from the OPR1000 (Optimized Power Reactor 1000 MWe) which is the reference plant in operation. Especially, the increased pipe size of the main steam and the economizer feedwater lines cause more severe excitations such as the nozzle thrust, jet impingement, and nozzle reactions when those lines are postulated to break at terminal ends or intermediate locations. In this paper, the major mechanical effects on the Reactor Coolant System (RCS) due to the branch line pipe breaks for SKN 3&4 are investigated; the support loads of the reactor vessel, steam generators, reactor coolant pumps are evaluated. In this investigation, the RCS is modeled using 3-dimensional lumped mass and beam elements, and the non-linear time-history structural analyses are performed. The analysis results show that the RCS dynamic responses are governed by the main steam line breaks and feedwater economizer line breaks.
Young Jin Byun Young Sil Su Kwang Won Lee Jong Tae Seo
NSSS Engineering & Development Division Korea Power Engineering Company, Inc.
国际会议
18th International Conference on Nuclear Engineering(第18届国际核能工程大会 ICONE 18)
西安
英文
1-7
2010-05-17（万方平台首次上网日期，不代表论文的发表时间）