Coolability of collapsed core debris bed
Creators
- 1. Osaka Univ., Suita (Japan). Faculty of Engineering
Description
For gaining basic information on the post accident heat removal, measurement of dryout heat flux was made, with water as coolant, in a 50 mm i.d. pyrex glass cylinder vessel. The fuel debris bed subjected to decay heat was simulated by steel balls or pellets which were induction-heated with a power supply of 20 kHz and 30 kW. Emphases were placed on the effects of system pressure, particle size, size mixing and coolant flow. Mixtures of cylindrical steel pellets and ceramic balls were also used to simulate TMI-2 type degraded core. The experiment covered the ranges over the system pressure of 0.02 - 0.5 MPa, the steel ball diameter of 0.3 - 4.0 mm and the pellets of 5.5 mm dia. x 9 mm, 8 mm dia. x 12 mm and 11 mm dia. x 11 mm with about 2 mm Al2O3 balls. The experimental results, as a whole, agreed fairly well with the prediction based on Lipinski's O-D model in the dependence of dryout heat flux both on pressure and on particle size. In detail, however, the dryout heat flux deviates toward a lower value at a higher pressure while to a higher value for a small size particle bed. Comparison of the results between the free and fixed beds suggests that the deviation to the higher side will be attributed to the channeling and/or levitation. For mixture, a number-weighted average diameter is available as the equivalent hydraulic diameter, which means that smaller particles of large number govern the dryout. The dryout heat flux increases with coolant mass flux, approaching to the complete vaporization of coolant, and greatly enhanced by a small amount of flow or circulation head. (author)
Additional details
Publishing Information
- Publisher
- Atomic Energy Soc. of Japan.
- Imprint Place
- Tokyo (Japan)
- Imprint Title
- Proceedings of second international topical meeting on nuclear power plant thermal hydraulics and operations
- Imprint Pagination
- 1191 p.
- Journal Page Range
- p. 6/34-6/40.
Conference
- Title
- 2. international topical meeting on nuclear power plant thermal hydraulics and operations.
- Dates
- 15-17 Apr 1986.
- Place
- Tokyo (Japan).
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 18061819
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- AFTER-HEAT REMOVAL; COOLANTS; COOLING; DRYOUT; FLOW RATE; HEAT FLUX; MELTDOWN; PARTICLE SIZE; PRESSURE DEPENDENCE; REACTOR COOLING SYSTEMS; REACTOR CORES; SIMULATION; THREE MILE ISLAND-2 REACTOR
- Descriptors DEC
- ACCIDENTS; COOLING SYSTEMS; ENRICHED URANIUM REACTORS; POWER REACTORS; PWR TYPE REACTORS; REACTOR ACCIDENTS; REACTOR COMPONENTS; REACTORS; SIZE; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS