Japanese Sodium-cooled Fast Reactor Research Articles

R&D in Support of ASTRID and JSFR: Cross-Analysis and Identification of Possible Areas of Cooperation

The Commissariat à L’Énergie Atomique et aux Énergies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fast reactors (SFRs) within two decades. The common final goal of their respective programs is SFR commercialization. The target of a commercial SFR for CEA and JAEA is basically consistent with Generation IV goals. Based on their industrial backgrounds and feedback from past and existing reactor experiences, CEA and JAEA have selected pool and loop configurations for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) and the Japanese Sodium-Cooled Fast Reactor (JSFR), respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR, respectively). The analysis results show that both concepts are technologically feasible and meet design goals. From the viewpoint of collaboration, the analysis identifies a wide range of collaborative areas in several fields: design principles (e.g., design target and design standard); development of components and systems; development of component parts, devices, or subsystems; design methods; simulation tools; etc.

Advanced Reactor Concepts and Fuel Cycle Technologies

Nuclear power is an abundant source with zero CO 2 output, and expanded fission power is an attractive option especially in growing countries. Various reactor concepts are being studied to comply with requirements of future nuclear fuel cycle such as safety, sustainability, cost-effectiveness, and proliferation risk reduction. These generation IV reactor concepts include both thermal and fast reactors and are being developed in conjunction with advanced reprocessing technology. In addition, beyond-generation IV type reactors are being developed, which mitigate the concern on the nuclear waste issue by adopting very high-burnup fuel and long-life core. Consideringmovement towards future nuclear cycle, we have collected papers which cover the topical areas of new activities in advanced reactors and fuel cycle technologies and published these in this special issue on advanced reactor concepts and fuel cycle technologies. This special issue contains 10 research articles, including seven papers on advanced reactor concepts. The advanced reactor concept covers sodium-cooled fast reactor (SFR), gascooled fast reactor (GFR), supercritical water reactor (SCWR), and pressurized water reactor (PWR).The research topics include an advanced Korean SFR system design and analysis, improvement of Japanese SFR core performance by moderators in the blanket region, use of moderating materials to compensate the drawback ofminor actinide containing transmutation fuel in European fast reactors, a multiphysics modeling technique for in situ breeding and burning reactors, an ultralong fuel cycle compact GFR system design and analysis, a simulation strategy for the evaluation of neutronic properties of a Canadian SCWR fuel channel, and a thorium-plutonium fuel cycle for operating cycle extension of a PWR. The special issue also includes research results on the modeling forced-flow chemical vapor infiltration for fabrication of SiC-SiC composite materials, development of inactive demonstration facility for a proliferation-resistant pyroprocessing technology, and the economic analysis of a closed fuel cycle.

Study on Detailed Calculation and Experiment Methods of Neutronics, Fuel Materials, and Thermal Hydraulics for a Commercial Type Japanese Sodium-Cooled Fast Reactor

This paper discusses the objectives and results of a multiyear R&D project to improve the modeling accuracy for the detailed calculation of the Japanese Sodium-cooled Fast Reactor (JSFR), although the preliminary design of JSFR is prepared using conventional methods. For detailed design calculations, new methods are required because the JSFR has special features, which cannot be accurately modeled with existing codes. An example is the presence of an inner duct in the fuel assemblies. Therefore, we have developed new calculational and experimental methods in three areas: (1) for neutronics, we discuss the development of methods and codes to model advanced FBR fuel subassemblies, (2) for fuel materials, modeling and measurement of the thermal conductivity of annular fuel is discussed, and (3) for thermal hydraulics, we describe advances in modeling and calculational models for the intermediate heat exchanger and the calculational treatment of thermal stratification in the hot plenum of an FBR under low flow conditions. The new methods are discussed and the verification tests are described. In the validation test, measured data from the prototype FBR Monju is partly used.

Gas entrainment allowance level at free surface and gas dynamic behavior of sodium-cooled fast reactor

In a sodium-cooled fast reactor (SFR), inert gases exist in the primary coolant system either in a state of dissolved gas or free gas bubbles. The sources of the gas bubbles are entrainment and dissolution of the reactor cover gas (argon) at the vessel free surface and emission of the helium gas that is produced as a result of disintegration of B 4C control rod material. The gas in the primary system may cause disturbance in reactivity, nucleation site for boiling, etc. Therefore, it is a key issue from the design and safety viewpoint and the allowance level is necessary regarding the gas entrainment at the free surface and the gas bubble concentration in the primary system. In the present study, a gas entrainment allowance level at the free surface is discussed and rationalized for the Japanese SFR (JSFR) design. The influence of the gas entrainment is evaluated using the void fraction at the core inlet. Design criteria for the acceptable level of the gas entrainment and gas concentration are proposed in consideration of the background level of gasses in the coolant. For the purpose, a plant dynamics code VIBUL has been developed to apply to the JSFR design to evaluate the concentration distribution of the dissolved gas and the free gas bubble in the JSFR system. Using the plant dynamics code for the bubble behavior, the background level of the free gas (void fraction at the core inlet) has been obtained. Assuming that the total void fraction should be kept below 105% of the background level, a preliminary design allowance level of gas entrainment is proposed as the map in terms of the entrainment rate and the entrained bubble radius. Furthermore, the possibility of bubble removal and design requirement of the device is investigated to satisfy the allowance level. It is noted that the background level is already very low in comparison with the induced void reactivity by the void passing the reactor core.