Overview of the French research on the evolution of spent fuel rod after discharge from the reactor
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Overview of the French research on the evolution of spent fuel rod after discharge from the reactor C. Ferry1, C. Cappelaere1, C. Jegou2, J.P. Piron3, M. Firon1 and A. Ambard4 1 CEA-Saclay, Nuclear Energy Division, 91191 Gif-sur-Yvette cedex, France 2 CEA-Marcoule, Nuclear Energy Division, BP 17171, 30207 Bagnols sur Ceze, France 3 CEA-Cadarache, Nuclear Energy Division, 13108 Saint-Paul lez Durance cedex, France 4 EDF, R&D division, les Renardieres, F-77818 Moret-sur-Loing Cedex, France ABSTRACT Since 2006, French research on spent fuel has focused on the main issues related to transport and extended in-pool storage of spent fuel assembly. Studies on creep behaviour of irradiated cladding have resulted in a new creep model which is valid over a wide domain of temperature, internal pressure and time. Under nominal conditions, no evolution of the spent fuel rod is expected during in-pool storage. In case of defective fuel rods in the storage pool, the consequences of fuel alteration on the initial defect of the cladding depend on the matrix alteration rate and nature of the secondary phases formed. Considering the optional scenario of direct disposal, the long-term behaviour of the spent fuel is investigated focusing on helium consequences before water contact on the one hand and on the influence of repository conditions on matrix alteration on the other hand. The aim of the on-going studies is to improve the safety margins initially introduced in the radionuclide source term models. INTRODUCTION Since 1999, CEA has been carrying out the research on the spent fuel (SF) long-term evolution in the framework of PRECCI program (acronym in French of the “Research Program on the Spent Fuel long-term Evolution”), with the support of Electricite de France (EDF) and AREVA-NP for the studies on the irradiated cladding. In 2006, the French Act for the management of radioactive wastes led to redefine the operational objectives of the project. In the French nominal scenario, reprocessing of spent fuel will be continued. Dry storage is abandoned whereas in-pool storage (extended up to 70 years) is chosen by EDF as the reference scenario to manage the flux of spent fuel before reprocessing (Figure 1). Direct disposal is still studied as an option, in collaboration with the French national agency of radioactive waste management (ANDRA, Agence Nationale pour la gestion des Dechets Radioactifs). Within this context, PRECCI has to provide the elements of response to the following operational questions: Regarding the retrievability of SF assemblies after transport and interim storage: The cladding is the first barrier of confinement before radionuclide release. The main risk of rupture at high temperature, during transport, is due to cladding creep under internal pressure. What are the creep laws and rupture criteria for irradiated claddings? And after transport, what are the consequences of hydruration on mechanical properties at ambient temperature for the irradiated cladding and structural parts of SF assembly? Some issues concern
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