An improvement in RELAP5/MOD2 heat transfer calculation during reflood
- 1. Institute of Nuclear Energy Research, Lung-Tan (Taiwan)
Description
Analyses of two large-break loss-of-coolant-accident (LOCA) experiments, namely, LOFT L2-5 and Semiscale S-06-3, were performed with RELAP5/MOD2/cy 36.04. Excessive cooling, which occurred right before final quench, has been found in both calculations. The causes of the excessive cooling may be quite complex during large-break LOCA calculation. Among all the possible reasons, improper modeling of the heat transfer calculation in the reflood heat transfer package has been identified as a large contributor to the discrepancy between calculated and measured cladding temperatures. The extension of the correlations to the conditions far beyond the applicable ranges of the correlations and the usage of a partition factor in determining total heat transfer coefficient are the two aspects that will be improved in this study. A flow-regime-based heat transfer modeling has been incorporated into the RELAP5/MOD2 reflood heat transfer package. A brief description of the original model, the revised model, and the results from these two models are presented in this paper
Additional details
Publishing Information
- Journal Title
- Transactions of the American Nuclear Society
- Journal Volume
- 60
- Series
- Trans. Am. Nucl. Soc.
- Journal Page Range
- 762-764
- ISSN
- 0003-018X
- CODEN
- TANSA
Conference
- Title
- Winter meeting of the American Nuclear Society (ANS) and nuclear power and technology exhibit.
- Dates
- 26-30 Nov 1989.
- Place
- San Francisco, CA (USA).
INIS
- Country of Publication
- United States
- Country of Input or Organization
- United States
- INIS RN
- 22085718
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- CRITICAL HEAT FLUX; FLOW MODELS; HEAT TRANSFER; HYDRAULICS; LOFT REACTOR; LOSS OF COOLANT; MATHEMATICAL MODELS; PWR TYPE REACTORS; R CODES; REACTOR ACCIDENTS; REACTOR SAFETY; TEST FACILITIES
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; HEAT FLUX; POWER REACTORS; REACTORS; RESEARCH AND TEST REACTORS; SAFETY; TANK TYPE REACTORS; TEST REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Secondary number(s)
- CONF-891103--.