French regulatory decision process for severe accident management and orientations for future European research.

Jorge Oliveri, Bernard Chaumont, FISA-2001 EU Resarch in reactor safety, 12-14 novembre 2001, Luxembourg.

In 1992, the French safety authority asked EDF to present its orientations in the field of prevention and mitigation of severe accidents for operating plants, focusing first on hydrogen risks, high pressure core melt scenarios (prevention of early containment failure risk) and on the definition and qualification of the instrumentation needed to manage such situations. In addition, identification of potential scenarios leading to containment bypass was required. Since then, safety analysis performed has led, among others, to the recommendation of hydrogen risk mitigation features implementation in all French NPPs. Besides, feasibility studies of high pressure core melt prevention by an intentional and reliable Reactor Coolant System depressurisation were performed. This paper presents the status of decisions making process and open issues for which specific R&D programs have been defined. For hydrogen risk, experimental investigations are or have been performed in the TOSQAN and MISTRA facilities for hydrogen distribution, in the RUT facility for hydrogen combustion and in the KALI and H2PAR facilities for passive autocatalytic recombiners behavior. All the knowledge gained will be incorporated into models of the TONUS and ASTEC codes under development at irsn in order to study hydrogen behavior inside the containment, from distribution to different combustion modes. To investigate potential solution for corium cooling, EDF and irsn are involved in extended R&D programs about in-vessel retention or ex-vessel cooling. In such a frame, particular attention is paid to the risk of containment failure as a consequence of a steam explosion. As far as steam explosions are concerned, experimental activities are devoted to improve the knowledge of fine physics to improve modeling in the MC3D code for thermalhydraulics and in the CASTEM-PLEXUS code for mechanical consequences. A more global validation of these codes is performed by comparison to the FARO, KROTOS, ECO and BERDA programs results. Finally irsn has undertaken a new evaluation of S3 source term taking into account updated data on fission product behaviour (PHEBUS). Dedicated R&D programs have been defined on molten core concrete interaction and iodine behaviour, which are considered as the main uncertainties for source term evaluation. Concerning future reactors, a common German and French approach was defined in 1993. Safety Plant design requirements and the associated R&D program to study specific severe accidents mitigation features are not described in this presentation