Generation IV concepts: Japan

Abstract Japan's effort for developing Generation IV reactors is focused on sodium-cooled fast reactors. With the targets for a high level of safety and economy, a plant design concept has been developed and is called Japan sodium-cooled fast reactor (JSFR). It is a loop type and is characterized, in terms of safety, by a self-actuated (passive) reactor shut-down system, a re-criticality free core design, and a natural-circulation decay heat removal system. It is also characterized, from the viewpoint of economy, as a two-loop heat transport system, integrated intermediate heat exchanger/pump component, and others. Ten key innovative technologies were identified and research and development were conducted to confirm the feasibilities of these technologies. After the Fukushima-Daiichi accident, the design has been updated for pursuing further safety enhancement reflecting the lessons learned from the accident.

[1]  K Kurisaka.  Probabilistic Safety Assessment of Japanese Sodium-cooled Fast Reactor in Conceptual Design Stage , 2006 .

[2]  Mamoru Konomura,et al.  An Innovative Concept of the Sodium-Cooled Reactor to Pursue High Economic Competitiveness , 2002 .

[3]  M. Inoue,et al.  Progress in the R and D Project on Oxide Dispersion Strengthened and Precipitation Hardened Ferritic Steels for Sodium Cooled Fast Breeder Reactor Fuels , 2007 .

[4]  Ikken Sato,et al.  Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments , 2004 .

[5]  Shoji Kotake,et al.  Development of Passive Shutdown System for SFR , 2010 .

[6]  Hideaki Monji,et al.  Proposal of Design Criteria for Gas Entrainment from Vortex Dimples Based on a Computational Fluid Dynamics Method , 2008 .

[7]  Tai Asayama,et al.  Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (2) Development of the Material Strength Standard of 316FR Stainless Steel , 2013 .

[8]  Ryodai Nakai,et al.  Safety Strategy of JSFR Eliminating Severe Recriticality Events and Establishing In-Vessel Retention in the Core Disruptive Accident , 2011 .

[9]  M. Inoue,et al.  Effects of neutron irradiation on tensile properties of oxide dispersion strengthened (ODS) steel claddings , 2011 .

[10]  U. S. Doe.  A Technology Roadmap for Generation IV Nuclear Energy Systems , 2002 .

[11]  Yasushi Okano,et al.  Numerical Methodology to Evaluate Fast Reactor Sodium Combustion , 2001 .

[12]  Shoji Kotake,et al.  The result of a wall failure in-pile experiment under the EAGLE project , 2007 .

[13]  Tai Asayama,et al.  Development of 2012 Edition of JSME Code for Design and Construction of Fast Reactors: (1) Overview , 2013 .

[14]  Kenichi Kurisaka,et al.  Technological Feasibility of Two-Loop Cooling System in JSFR , 2010 .

[15]  Osamu Watanabe,et al.  ICONE23-1577 WATER EXPERIMENTS ON THERMAL STRIPING IN REACTOR VESSEL OF JAPAN SODIUM-COOLED FAST REACTOR : COUNTERMEASURES FOR SIGNIFICANT TEMPERATURE FLUCTUATION GENERATION , 2015 .

[16]  S. Okamura,et al.  Seismic Isolation Design for JSFR , 2011 .

[17]  Takafumi Aoyama,et al.  Distinguished Achievements of a Quarter-Century Operation and a Promising Project Named MK-III in JOYO , 2005 .

[18]  Takumi Kawamura,et al.  A volume-conservative PLIC algorithm on three-dimensional fully unstructured meshes , 2013 .

[20]  Yuzuru Eguchi,et al.  Development of an evaluation methodology for the natural circulation decay heat removal system in a sodium cooled fast reactor , 2015 .

[21]  Shoji Kotake,et al.  Evaluation of JSFR Key Technologies , 2012 .

[22]  H. Kamide,et al.  Influence of elbow curvature on flow structure at elbow outlet under high Reynolds number condition , 2011 .

[23]  Masaaki Tanaka,et al.  Numerical Investigation on Large Scale Eddy Structure in Unsteady Pipe Elbow Flow at High Reynolds Number Conditions with Large Eddy Simulation Approach , 2012 .

[24]  Shoji Kotake,et al.  Demonstration of Control Rod Holding Stability of the Self Actuated Shutdown System in Joyo for Enhancement of Fast Reactor Inherent Safety , 2007 .

[25]  Hideyuki Morita,et al.  Development of the Integrated IHX/Pump Component 1/4-Scale Vibration Testing , 2006 .

[26]  T. Inoue,et al.  In-pile creep rupture properties of ODS ferritic steel claddings , 2009 .

[27]  Yoshitaka Fukano,et al.  Three-Pin Cluster CABRI Tests Simulating the Unprotected Loss-of-Flow Accident in Sodium-Cooled Fast Reactors , 2011 .

[28]  Werner Maschek,et al.  The Development of SIMMER-III, An Advanced Computer Program for LMFR Safety Analysis, and Its Application to Sodium Experiments , 2006 .

[29]  H. Kamide,et al.  Transient Behavior of Gas Entrainment Caused by Surface Vortex , 2008 .

[30]  H. Kamide,et al.  Experimental Study on Gas Entrainment at Free Surface in Reactor Vessel of a Compact Sodium-Cooled Fast Reactor , 2008 .

[31]  Hiroyuki Ohshima,et al.  Validation of a Computational Simulation Method for Evaluating Thermal Stratification in the Reactor Vessel Upper Plenum of Fast Reactors , 2011 .

[32]  Hideki Kamide,et al.  ICONE23-1666 JSFR DESIGN PROGRESS RELATED TO DEVELOPMENT OF SAFETY DESIGN CRITERIA FOR GENERATION IV SODIUM-COOLED FAST REACTORS : (1) OVERVIEW , 2015 .

[33]  Mamoru Konomura,et al.  Progress on the Plant Design Concept of Sodium-Cooled Fast Reactor , 2007 .

[34]  H. Ohshima,et al.  Improvement of Gas Entrainment Prediction Method—Introduction of Surface Tension Effect— , 2010 .

[35]  Hiroyuki Ohshima,et al.  Development of V2UP (V&V plus uncertainty quantification and prediction) procedure for high cycle thermal fatigue in fast reactor—Framework for V&V and numerical prediction , 2016 .

[36]  Masaaki Tanaka,et al.  Numerical simulation of thermal striping phenomena in a T-junction piping system for fundamental validation and uncertainty quantification by GCI estimation , 2015 .

[37]  Shoji Kotake,et al.  Safety Design Requirements for Safety Systems and Components of JSFR , 2011 .