Within the GIF sodium-cooled fast reactor (SFR) system arrangement, the SFR Safety and Operation (SO) project addresses the areas of safety technology and reactor operation technology developments. The aim of the SO project includes (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and codes” is devoted to the development of tools for the evaluation of safety, WP-SO-2 “Experimental Programs and Operational Experiences” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600, EBR-II and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as active and passive safety systems and other specific design features.This paper reports the following recent activities within the SO project.- Heat Transfer Analysis of CEFR Damaged Spent Fuel Assemblies in Closed Space Based on Fluid-Solid Coupling Analysis- PHENIX dissymmetric test description for system and coupled integral system simulations- Risk assessment methodology of decay heat removal function against external hazards- Evaluation method for Sodium-concrete reaction- Safety Analysis of CRW- Development of the COREMELT3D Code for Analyses of transients and Accidents in the SFR- Sodium Fire Validation in MELCOR- Analysis of EBR-II Tests BOP-301 and BOP-302R- SAS4A metal fuel model extension for simulation of molten fuel behavior over longer time periods- Safety analysis on ATWS events for PGSFR using MARS-LMR code- Thermal-Hydraulics in the Damaged Subassemblies Under the Operation of Decay Heat Removal System- Validation of computer codes for heat transfer performance on a finned-tube sodium-to-air heat exchanger- Influence of the non-isolated modes’ earthquake response on floor response spectra- Simplified PSA Level 1

Activities of the Gif Safety and Operation Project of Sodium-Cooled Fast Reactor Systems

Paper Proceedings: https://asmedigitalcollection.asme.org/ICONE/ICONE28/volume/85246

2021 28th International Conference on Nuclear Engineering (ICONE28)

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