Assessment of RELAP5/MOD3.1 Using LSTF Ten-Percent Main Steam-Line-Break Test Run SB-SL-01 (NUREG/IA-0148)

On this page:

• Publication Information
• Abstract

Download complete document

• NUREG/IA-0148 (PDF 3.40 MB)

Publication Information

Date Published: September 1998

Prepared by:
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 Yusong, Taejon Korea 305–600

Japan Atomic Energy Research Institute
Tokai-Mura, Naka-Gun
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

Availability Notice

Abstract

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.