An analysis of ROSA-IV/LSTF 10% main steam line break test run SB-SL-01 using RELAP5/MOD3

Book · Mon Jul 01 04:00:00 EDT 1996

OSTI ID:250969

Lee, H; Seong, H ^[1]; Park, G ^[2]; Kumamaru, Hiroshige; Kukita, Yutaka ^[3]

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1. Korea Power Engineering Co., Inc., Kyunggi-Do (Korea, Republic of). Nuclear Engineering Dept.
2. Seoul National Univ. (Korea, Republic of). Nuclear Engineering Dept.
3. Japan Atomic Energy Research Inst., Ibaraki (Japan). Dept. of Nuclear Safety Research

This paper presents RELAP5/MOD3 code calculations of a 10% main steam line break test, designated as RUN SB-SL-01, conducted using the ROSA-4 Large Scale Test Facility (LSTF). The RELAP5/MOD3 input deck of LSTF, which includes 189 volumes, 200 junctions, and 180 heat slabs, was modeled to obtain best-estimate predictions of several important features during the main steam line break accident in order to property evaluate the consequences of this accident. The main conclusions drawn were that the results of RELAP5/MOD3 code calculations were in reasonable agreements with test RUN SB-SL-01, especially for the trends of key parameters. Detailed investigations indicated minor discrepancies in RCS pressure during the period of time that voiding occurred in the upper head. This is possible due to emptying of the pressurizer and voiding in the upper head. Sensitivity studies were also performed for the break junction discharge coefficient and the separator drain line loss coefficient. These parameters had significant effects on the steam quality on the secondary side and on the break flow through the change of water inventory on the secondary side. This phase separation process was adequately predicted during all transients with break junction discharge coefficient of 0.85 and separator drain line loss coefficient of 10.

• OSTI ID: 250969
• Report Number(s): CONF-960306--; ISBN 0-7918-1226-X
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
22 GENERAL STUDIES OF NUCLEAR REACTORS
COMPARATIVE EVALUATIONS
COMPUTERIZED SIMULATION
EXPERIMENTAL DATA
PERFORMANCE
PWR TYPE REACTORS
REACTOR ACCIDENTS
REACTOR COOLING SYSTEMS
REACTOR SAFETY EXPERIMENTS
SAFETY ANALYSIS
SENSITIVITY ANALYSIS
TEST FACILITIES
THEORETICAL DATA