Thermo-mechanical behavior of an ablated reactor pressure vessel wall in a Nordic BWR under in-vessel core melt retention
Hongdi Wang, Walter Villanueva, Yangli Chen, Artem Kulachenko, Sevostian Bechta
Abstract
The reactor pressure vessel (RPV) of a nuclear reactor is one of the key safety barriers preventing radioactive environmental releases during a severe accident. One of the promising strategies of severe accident management (SAM) is to retain the molten core having continuous decay heat inside the RPV by natural water cooling of the external vessel surface. The feasibility of such a strategy relies on complex safety analyses including accurate prediction of vessel thermo-mechanical behavior which can be assessed by mechanical stresses and strains. In this paper, we present the stress–strain response of an ablated RPV of a Nordic boiling water reactor (BWR) to dynamic thermomechanical loads set by expanding volumetrically heated molten pool inside the RPV cooled by water at the external surface. MELCOR 2.2.9541 severe accident code is used to simulate the in-vessel behavior and provides the input conditions for dedicated structural analysis of the RPV using ANSYS® Mechanical APDL 19.2. A creep model of the SA533B1 vessel steel is validated against uniaxial creep tests carried out by INEL (Idaho National Engineering Laboratory) and creep tests performed at CEA (French Alternative Energies and Atomic Energy Commission) as part of the OLHF (OECD Lower Head Failure) Project. Two generic severe accident scenarios are considered: (i) Station Blackout (SBO) and (ii) Station Black-out and Loss-of-coolant Accident (SBO + LOCA). In both scenarios, we found that the RPV has maintained structural integrity considering two failure criteria: stress-based and strain-based.