Evaluation of Containment Failure Probability by Ex-Vessel Steam Explosion in Japanese LWR Plants

Abstract

The containment failure probability due to ex-vessel steam explosions was evaluated for Japanese BWR and PWR model plants. A stratified Monte Carlo technique (Latin Hypercube Sampling (LHS)) was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The evaluation was made for three scenarios: a steam explosion in the pedestal area or in the suppression pool of a BWR model plant with a Mark-II containment, and in the reactor cavity of a PWR model plant. The scenario connecting the generation of steam explosion loads and the containment failure was assumed to be displacement of the reactor vessel and pipings, and failure at the penetration in the containment boundary. We evaluated the conditional containment failure probability (CCFP) based on the preconditions of failure of molten core retention within the reactor vessel, relocation of the core melt into the water pool without significant interference, and a strong triggering at the time of maximum premixed mass. The obtained mean and median values of the CCPF were 6:4Â10 À2 (mean) and 3:9Â10 À2 (median) for the BWR suppression pool case, 2:2Â10 À3 (mean) and 2:8Â10 À10 (median) for the BWR pedestal case, and 6:8Â10 À2 (mean) and 1:4Â10 À2 (median) for the PWR cavity case. The evaluation of CCFPs on the basis of core damage needs consideration of probabilities for the above-mentioned preconditions. Thus, the CCFPs per core damage should be lower than the values given above. The specific values of the probability were most dependent on the assumed range of melt flow rate and fragility curve that involved conservatism and uncertainty due to simplified scenarios and limited information.

Review of thermal-hydraulic researches in severe accidents in light water reactors

“…Experimental and analytical researches are carried out using small-and large-scale experiments [11][12][13][14][15][16][17][18][19][20][21][22][23].…”

“…The containment failure probability due to exvessel steam explosions was evaluated for Japanese BWR and PWR model plants [21]. A stratified Monte Carlo technique was applied for the evaluation of steam explosion loads using steam explosion simulation code JASMINE.…”

“…The MELCOR code is composed of a control program and many modules, or packages, which model the major systems of a reactor plant and thermal-hydraulic response of the primary reactor coolant system, the reactor cavity, the containment, and the confinement buildings under offnormal or accidental conditions. This code is used for the development of systematic models for aerosol agglomeration and spray removal under SA conditions [46], a real-time prediction technique for accident progression in a PWR containment during a SA when water on the containment floor flows into the reactor cavity (wet cavity) [47], evaluation of containment failure probability by ex-vessel steam explosion in Japanese LWR [21], and evaluation of experiments on the release of gaseous iodine from g -irradiated aqueous CsI solution [48].…”

Simulation of Ex-Vessel Steam Explosion

“…T o b e a b l e t o p e r f o r m a s e r i e s o f s i m u l a tions of different ex-vessel steam explosion scenarios, the reactor cavity was modelled in a simplified 2D geometry, as is common practise (Meignen et al, 2003;Kawabata, 2004;Esmaili and Khatib-Rahbar, 2005;Moriyama et al, 2006;OECD/NEA, 2007). The 2D geometry has to be appropriately defined to assure that the 2D simulation results reflect qualitatively and quantitatively as closely as possible the conditions in a real 3D reactor cavity.…”