Assessment of RELAP5/MOD3.1 Using LSTF Ten-Percent Main Steam-Line-Break Test Run SB-SL-01 (NUREG/IA-0148)
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: September 1998
J. G. Oh, H. D. Lee, K. K. Jee, S. K. Kang/KOPEC
Y. S. Bang, K. W Seul/KINS
H. Kumamaru, Y. Anoda/JAERI
|Korea Power Engineering Company
150 Duckjin-Dong, Yusong-Ku
Taejon, Korea 305–353
|Korea Institute of Nuclear Safety
P.O. Box 114
Japan Atomic Energy Research Institute
Ibaraki-Ken 319–1195, Japan
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
Results produced by the RELAP5/MOD3.1 computer code were compared with the experimental data from JAERI's LSTF Test Run SB-SL-01 for a 10% main steam line break transient in a pressurized water reactor. The code simulation for the base case included a total of 189 fluid control volumes and 199 flow junctions to model the transient two-phase flow phenomena. Also, a total of 180 heat slabs were used to model the system heat transfer. The code predictions of the experimental results are generally satisfactory for the trends of key parameters.
Sensitivity studies performed for the break discharge coefficient, the separator drain line loss coefficient, and the number of steam generator nodes did not reveal any strong dependencies. Nevertheless, optimal values of these parameters that led to the lowest overall statistical error were obtained, and these values were subsequently used in the "Base Case" analysis.