Nuclear Data for Sustainable Nuclear Energy

Final report of a coordinated action on nuclear data for industrial development in Europe (CANDIDE). The successful development of advanced nuclear systems for sustainable energy production depends on high-level modelling capabilities for the reliable and cost-effective design and safety assessment of such systems, and for the interpretation of key benchmark experiments needed for performance and safety evaluations. High-quality nuclear data, in particular complete and accurate information about the nuclear reactions taking place in advanced reactors and the fuel cycle, are an essential component of such modelling capabilities. In the CANDIDE project, nuclear data needs for sustainable nuclear energy production and waste management have been analyzed and categorized, on the basis of preliminary design studies of innovative systems. Meeting those needs will require that the quality of nuclear data files be considerably improved. The CANDIDE project has produced a set of recommendations, or roadmap, for sustainable nuclear data development. In conclusion, a substantial long-term investment in an integrated European nuclear data development program is called for, complemented by some dedicated actions targeting specific issues.JRC.D.5-Neutron physic

Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF

The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental-scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facilities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall system that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Several parameters were analyzed, like the criticality behavior (namely the K-eff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed