Published October 2000
| Version v1
Book
Disruptive core relocation analysis of PHEBUS/FPT0 test with SAMPSON code
- 1. Nuclear System Analysis Technology R and D dept., Power and Industrial Systems R and D Center, Power Systems and Services Company, Toshiba Corporation, Kawasaki, Kanagawa (Japan)
- 2. Combustion and Heat Transfer Lab., Takasago R and D Center, Mitsubishi Heavy Industries, Ltd., Takasago, Hyogo (Japan)
- 3. Advanced Simulation Systems Dept., Nuclear Power Engineering Corporation, Tokyo (Japan)
Description
SAMPSON is an integration of twelve analysis modules under the final development phase (phase-2) and will be capable of simulating hypothesized severe accidents in a nuclear power plant. One of these modules, the Molten Core Relocation Analysis (MCRA) module, simulates the relocation behavior of a molten core during a severe accident. MCRA models severe accident phenomena by using mechanistic formulations for multi-phase, multi-component, and multi-velocity field. As one of the verification studies of SAMPSON in Phase-1, the in-core phenomena of PHEBUS/FPT0 was analyzed with three modules, MCRA, fuel rod heat up analysis (FRHA) module, and the analysis control module (ACM) of SAMPSON. (author)
Additional details
Publishing Information
- Publisher
- Atomic Energy Society of Japan
- Imprint Place
- Tokyo (Japan)
- Imprint Title
- Proceedings of NTHAS2: Second Japan-Korea symposium on nuclear thermal hydraulics and safety
- Imprint Pagination
- 800 p.
- Journal Page Range
- p. 567-574
Conference
- Title
- 2. Japan-Korea symposium on nuclear thermal hydraulics and safety
- Acronym
- NTHAS2
- Dates
- 15-18 Oct 2000
- Place
- Fukuoka (Japan)
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 32005505
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- COMPARATIVE EVALUATIONS; COMPUTERIZED SIMULATION; MELTDOWN; PHEBUS REACTOR; PWR TYPE REACTORS; REACTOR CORE DISRUPTION; REACTOR SAFETY; REACTOR SAFETY EXPERIMENTS; S CODES
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; ENRICHED URANIUM REACTORS; EVALUATION; POOL TYPE REACTORS; POWER REACTORS; REACTOR ACCIDENTS; REACTORS; RESEARCH AND TEST REACTORS; RESEARCH REACTORS; SAFETY; SIMULATION; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 10 refs., 11 figs., 1 tab.