In the framework of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors (SFRs) have been developed as part of the worldwide deployment of SFRs. Japan Atomic Energy Agency (JAEA) and the Mitsubishi FBR Systems, Inc. (MFBR) have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump-integrated Intermediate Heat Exchanger (IHX) is evaluated as a safety measure for the advanced loop-type SFR. Furthermore, maintainability and reparability of the safety measures are taken into account for the advanced loop-type SFR design study. The pump-integrated IHX has been modified to satisfy these requirements. This paper describes the modifications to withstand severe earthquake, primary coolant leak and sodium-water reaction. Also, this paper includes evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes for they have been adversely affected by the modifications.
- Nuclear Engineering Division
Development of the Pump-Integrated Intermediate Heat Exchanger in Advanced Loop-Type Sodium-Cooled Fast Reactor for Demonstration
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Amano, K, Enuma, Y, Futagami, S, Inoue, T, & Watanabe, S. "Development of the Pump-Integrated Intermediate Heat Exchanger in Advanced Loop-Type Sodium-Cooled Fast Reactor for Demonstration." Proceedings of the 2016 24th International Conference on Nuclear Engineering. Volume 2: Smart Grids, Grid Stability, and Offsite and Emergency Power; Advanced and Next Generation Reactors, Fusion Technology; Safety, Security, and Cyber Security; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues. Charlotte, North Carolina, USA. June 26–30, 2016. V002T06A004. ASME. https://doi.org/10.1115/ICONE24-60064
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