Cold leg Small Break LOCA experiments at HPITF
- 1. China Inst. of Atomic Energy, Beijing (China)
Description
Three cold-leg small-break loss-of-coolant accident (SBLOCA) experiments, NSB-4, NSB-6 and NSB-7, were carried out at the High Pressure Integral Test Facility (HPITF) to clarify the thermal-hydraulic phenomena of primary system during SBLOCA and to investigate the break area effects on system thermal-hydraulic responses. In these three experiments, the simulated break which correspondingly provided a break area of 5%, 2% or 1% of the 1/1000 scaled PWR cold-leg cross-sectional area respectively was oriented at the bottom of the horizontal cold leg. The break flow rate was affected by the break area significantly, and the event chronology and accidental severity of the three tests was considerably different; the heated-rod surface temperature excursion was the most severely for the NSB-7 experiment. The three experimental results were used to assess the RELAP5/MOD2 code in micro-computer version. The major phenomena that occurred during the experiments were calculated by RELAP5/MOD2 code although there were some differences between the data and predictions
Additional details
Publishing Information
- Imprint Title
- Proceedings of fifth international topical meeting on nuclear thermal hydraulics, operations and safety
- Imprint Pagination
- 1493 p.
- Journal Page Range
- p. GG2.1-GG2.6
Conference
- Title
- 5. international topical meeting on nuclear thermal hydraulics, operations and safety
- Dates
- 14-18 Apr 1997
- Place
- Beijing (China)
INIS
- Country of Publication
- China
- Country of Input or Organization
- China
- INIS RN
- 29048955
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BENCH-SCALE EXPERIMENTS; COMPARATIVE EVALUATIONS; FLOW RATE; HEAT TRANSFER; HYDRAULICS; LOSS OF COOLANT; NUCLEAR POWER PLANTS; PRESSURE MEASUREMENT; PRIMARY COOLANT CIRCUITS; PWR TYPE REACTORS; R CODES; SAFETY ANALYSIS; TEMPERATURE MEASUREMENT; TEST FACILITIES; TRANSIENTS
- Descriptors DEC
- ACCIDENTS; COMPUTER CODES; COOLING SYSTEMS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; EVALUATION; NUCLEAR FACILITIES; POWER PLANTS; POWER REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTOR COOLING SYSTEMS; REACTORS; THERMAL POWER PLANTS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS