Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation. 1. Introduction The loss of coolant in nuclear reactors during scenarios such as operation at full or low power, shutdown, and refueling can cause excessive heatup of the nuclear fuel; this is a beyond design basis accident where the core meltdown is the main focus. The slump of the molten core into the lower head of the Reactor Pressure Vessel (RPV) may produce the failure of it. The accident in Unit 2 of Three Mile Island (TMI-2) in 1979 was an accident with core melt but without failure of the RPV and with intact containment; therefore, the radiological consequences were negligible. The studies of severe accident are oriented into two branches: In-vessel and Ex-vessel where the failure of the RPV or ruptures in some pipes, such as the main steam lines, recirculation pipes, or feed water among others, is the interface parameter. The In-vessel phenomena include the thermal-hydraulics with natural circulation, reflooding, and refilling of the core at high temperature, oxidation of the cladding, hydrogen production, relocation of the molten core into the lower plenum of the RPV, liquefaction of control rods, recriticality, and failure of the reactor vessel. The Ex-vessel phenomena involve steam explosion, melt dispersion, fission products transport and deposition on structures and components, direct containment heating (DCH) among others [1];