Rough Estimates of Severe Accident Containment Loads Accompanying Vessel Breach in Boiling Water Reactors

Several probabilistic risk assessments (PRAs) have identified containment loads accompanying reactor vessel failure as a major contributor to the probability of early containment failure during severe accidents. Two significant contributors to these loads are phenomena referred to as “steam spike” and “direct containment heating.” To date, direct application of experimental and analytical studies of these phenomena to boiling water reactors (BWRs) are constrained by two limitations: (a) they are based on applications of large, complex containment response analysis computer codes, for which values of many major input parameters are highly uncertain, or (b) they only address pressurized water reactor containment designs. Relatively simple, parametric models are developed which allow a PRA analyst to evaluate the range of conditions under which steam spike or direct containment heating may be important contributors to containment loads for postulated severe accidents in BWRs. The models have been applied to a representative BWR/4 Mark I containment design to illustrate calculated results.