Evaluation of cooling characteristics for spent fuel pool accidents. Analysis of loss-of-cooling and loss-of-coolant accident in SFP with MAAP code
- 1. Central Research Inst. of Electric Power Industry, Nuclear Technology Research Lab., Komae, Tokyo (Japan)
Description
In the present study, the spent fuel pool (SFP) accident analyses using the integral severe accident code 'MAAP' were performed to predict the event sequence of both loss-of-cooling and loss-of-pool water accidents. The SFP model in MAAP5 code can calculate time to boiling, time to uncover and heatup of fuel assemblies, Zr-water and Zr-O2 oxidation. Moreover, a gas natural circulation can be calculated when the water level decreases below the bottom of the fuel rack or the fuel rack is melted away. The total decay heat power of fuel assemblies in the SFP and the leak rate of pool water are treated as variables in the present study. Under the severe thermal condition that the half of the SFP inventory is consisted of the fuel assemblies whose cooling time is only 7 days after discharge from the core, the pool water temperature reaches the saturation temperature at about 11 hours after the onset of the loss-of-cooling accident, and the pool level drops to the top of active fuels at about 89 hours. Then the peak cladding temperature reaches at 1200deg-C in 7.7 hours after the fuels are uncovered. It is found from the loss-of-pool water accident analyses that ensuring natural circulation of air in the spent fuel assemblies at early phase of the accident sequence is essential in preventing fuel damages. Even in case of the large leak rate or the instantaneous loss-of-pool water scenario, small decay heat can be cooled by natural circulation of air in the spent fuel assemblies. (author)
Additional details
Publishing Information
- Journal Title
- Denryoku Chuo Kenkyusho Hokoku
- Journal Issue
- no.L12007
- Journal Page Range
- p. 1-4, 1-31
- ISSN
- 1340-4652
INIS
- Country of Publication
- Japan
- Country of Input or Organization
- Japan
- INIS RN
- 48046254
- Subject category
- S42: ENGINEERING;
- Descriptors DEI
- AFTER-HEAT REMOVAL; COMPUTERIZED SIMULATION; FUEL ASSEMBLY DISMANTLING; FUEL STORAGE POOLS; FUKUSHIMA DAIICHI NUCLEAR POWER STATION; FUKUSHIMA-4 REACTOR; HYDROGEN; LOSS OF COOLANT; LOSS OF FLOW; MELTDOWN; NATURAL CONVECTION; NUCLEAR POWER PLANTS; POWER RANGE 01-10 MW; REACTOR CORE DISRUPTION; TEMPERATURE RANGE 1000-4000 K; ZIRCONIUM
- Descriptors DEC
- ACCIDENTS; BWR TYPE REACTORS; CONVECTION; ELEMENTS; ENERGY TRANSFER; ENRICHED URANIUM REACTORS; HEAT TRANSFER; MASS TRANSFER; MEGAWATT POWER RANGE; METALS; NONMETALS; NUCLEAR FACILITIES; POWER PLANTS; POWER RANGE; POWER REACTORS; REACTOR ACCIDENTS; REACTOR SITES; REACTORS; REMOVAL; SIMULATION; TEMPERATURE RANGE; THERMAL POWER PLANTS; THERMAL REACTORS; TRANSITION ELEMENTS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 5 refs., 52 figs., 4 tabs.; This record replaces 45008309