Boiling water reactor building environments after containment failure

In this paper, the results of calculations performed for the LaSalle Unit II Level 3 probabilistic risk analysis (PRA) under the Phenomenology and Risk Uncertainty Evaluation Program are discussed. For accident sequences at LaSalle in which containment heat removal has failed but core cooling is working, the primary containment would pressurize until it is either vented at 5.17 bar (60 psig) or fails at ∼ 14.5 bar (195 psig). In either case, severe steam environments would be created in the reactor building, posing a threat to the various safety and nonsafety systems used for core cooling. To evaluate the survivability of the equipment, a detailed reactor building model was constructed for use in the MELCOR code, and calculations were performed for various containment venting and failure scenarios. The results of these calculations were used to quantify the probability of equipment failure, using the results of the NUREG-1150 expert review group that evaluated the probability of equipment survival for various harsh environments