Assessment of RELAP5/MOD3.2 Using LOFT Large Break LOCA Test, LP–02–6 (NUREG/IA-0139)
On this page:
Download complete document
This page includes links to files in non-HTML format. See Plugins, Viewers, and Other Tools for more information.
Date Published: August 1998
T. S. Choi, J. H. Lee, B. S. Park, C. S. Cho, J.Y. Park/KNFC
Y. S. Bang, S. W Seul, H. J. Kim/KINS
Korea Nuclear Fuel Company
Yusong Gu, Daejeon City
Korea Institute of Nuclear Safety
Advanced Reactor Dept.
P.O. Box 114
Prepared as part of:
The Agreement on Research Participation and Technical Exchange under the
International Thermal-Hydraulic Code Assessment and Maintenance Program (CAMP)
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
The LOFT experiment LP-02-6 was simulated using the RELAP5/MOD3.2 code to assess its capability to predict the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was modeled with two core channels and split downcomer for a base calculation. The results of the base calculation show that the code can not predict the early bottom-up quenching which is a distinguished phenomenon of the experiment LP-02-6, mainly due to the deficiency of break flow model.
The discharge coefficient sensitivity study was performed to show that the calculated subcooled break flow which might significantly affect the early bottom-up quenching is dependent on the coefficient. More detailed modeling of the cross flow in the split downcomer was performed, but, resulted in little improvement on the predictability of bottom-up quenching. Additional calculation using the RELAP5/MOD3.1 instead of RELAP5/MOD3.2 showed that there is no large difference between the versions in the simulation of LBLOCA.