Published September 1990
| Version v1
Journal article
Subchannel analysis of a critical power test, using simulated BWR 8x8 fuel assembly
- 1. Nippon Atomic Industry Group Co. Ltd., Kawasaki, Kanagawa (Japan)
- 2. Toshiba Corp., Kawasaki, Kanagawa (Japan)
- 3. Toshiba Corp., Yokohama (Japan)
- 4. Nuclear Power Engineering Test Center, Tokyo (Japan)
- 5. Tokyo Inst. of Tech. (Japan)
- 6. Tokyo Univ. (Japan)
Description
Critical power predictions have been compared with the critical power test data obtained in simulated BWR 8x8 fuel rod assemblies. Two analytical methods for the critical power prediction in rod assemblies are used in the prediction, which are the subchannel analysis using the COBRA/BWR subchannel computer code with empirical critical heat flux (CHF) correlations and the liquid film dryout estimation using the CRIPP-3F 'multi-fluid' computer code. Improvements in both the analytical methods were made for spacer effect modeling, though they were specific for application to the current BWR rod assembly type. In general a reasonable agreement was obtained, though comparisons, between the prediction and the obtained test data. (orig.)
Additional details
Publishing Information
- Journal Title
- Nuclear Engineering and Design
- Journal Volume
- 122
- Journal Issue
- 1-3
- Series
- Nucl. Eng. Des.
- Journal Page Range
- 235-254
- ISSN
- 0029-5493
- CODEN
- NEDEA
Conference
- Title
- 3. international topical meeting on nuclear power plant (NPP) thermal hydraulics and operations.
- Dates
- 14-17 Nov 1988.
- Place
- Seoul (Republic of Korea).
INIS
- Country of Publication
- Netherlands
- Country of Input or Organization
- Netherlands
- INIS RN
- 22019514
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BWR TYPE REACTORS; C CODES; CONTINUITY EQUATIONS; CORRELATIONS; CRITICAL HEAT FLUX; CRITICAL PRESSURE; CROSS SECTIONS; DRYOUT; FLOW RATE; FOURIER HEAT EQUATION; FUEL ASSEMBLIES; MASS TRANSFER; POWER DISTRIBUTION; REACTOR CHANNELS; REACTOR CORES; TRANSITION BOILING; TWO-PHASE FLOW
- Descriptors DEC
- BOILING; COMPUTER CODES; DIFFERENTIAL EQUATIONS; DISTRIBUTION; ENRICHED URANIUM REACTORS; EQUATIONS; FLUID FLOW; HEAT FLUX; PARTIAL DIFFERENTIAL EQUATIONS; PHASE TRANSFORMATIONS; PHYSICAL PROPERTIES; POWER REACTORS; REACTOR COMPONENTS; REACTORS; SPATIAL DISTRIBUTION; THERMAL REACTORS; THERMODYNAMIC PROPERTIES; WATER COOLED REACTORS; WATER MODERATED REACTORS