Pre-test prediction of FIX II intermediate break experiment with THYDE-P1
- 1. Japan Atomic Energy Research Inst., Tokai, Ibaraki. Tokai Research Establishment
Description
This report contains the results from the pre-test prediction calculation of FIX II Intermediate Break Experiment with the THYDE-Pl computer code. The experiment simulated the conditions corresponding to an intermediate size break loss-of-coolant accident (LOCA) in one of the recirculation lines of a boiling water reactor (BWR) with external pumps. The test was the first LOCA experiment in the FIX II facility and was chosen as the OECD CSNI International Standard Problem No. 15. Although an extensive verification calculations have been performed with THYDE-Pl for pressurized water reactors (PWRs), the present calculation is the first application to BWR. The major purposes of the present calculation are to verify the applicability of THYDE-Pl to BWR and to indentify the model deficiencies for further code modification. After the experimental data being released, the predicted results with THYDE-Pl have been compared with the experimental data. The comparison shows that the present calculation has adequately predicted the important phenomena observed in the experiment. (author)
Availability note (English)
MF available from INIS under the Report Number.Files
17047526.pdf
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Additional details
Additional titles
- Subtitle (English)
- CSNI international standard problem no. 15
Publishing Information
- Imprint Pagination
- 84 p.
- Report number
- JAERI-M--85-115
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 17047526
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Descriptors DEI
- BWR TYPE REACTORS; FORECASTING; HEAT TRANSFER; HYDRODYNAMICS; LIQUID FLOW; LOSS OF COOLANT; PUMPS; REACTOR CORES; REACTOR SAFETY EXPERIMENTS; T CODES; TRANSIENTS; VALVES
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; CONTROL EQUIPMENT; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EQUIPMENT; FLOW REGULATORS; FLUID FLOW; FLUID MECHANICS; MECHANICS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS