Masayoshi Matsuura

Subchannel analysis to investigate the fuel assembly for the supercritical-water-cooled power reactor

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs. The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.

The Plant Feature and Performance of Double MS (Modular Simplified and Medium Small Reactor)

A new concept of a small and medium sized light water reactor, named the double MS: modular simplified and medium small reactor (DMS) was developed. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts, and effective use of proven technology. The decrease in the primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about 2 m compared with 3.7 m in the conventional boiling water reactor (BWR). The short active fuel length reduces the drop in core pressure and overcomes the natural circulation system. By using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to confirm transient performance, the DMS plant performance under transient conditions was evaluated using the TRACG code. TRACG code, which can treat multidimensional hydrodynamic calculations in a RPV, is well suited for evaluating the DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As a result, the maximum change in the minimum critical power ratio of the DMS was 0.14, almost the same as for the current advanced boiling water reactor (ABWRs). In order to improve safety efficiency developing an emergency core cooling system (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following: (1) eliminating highpressure injection systems, (2) adopting passive safety-related systems, and (3) optimizing distribution for the systems and power source for the ECCS. In this way, the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of the loss of coolant accident analysis, we confirmed that the core can be covered by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.

The Plant Feature and Performance of DMS (Double MS: Modular Simplified and Medium Small Reactor)

The Japan Atomic Power Company has initiative in developing the DMS concept as a 400MWe-class light water reactor. The main features of the DMS relative to overcoming the scale demerit are the miniaturization and simplification of systems and equipment, integrated modulation of construction, standardization of equipment layouts and effective use of proven technology. The decrease in primary containment vessel (PCV) height is achieved by reducing the active fuel length of the DMS core, which is about two meters compared with 3.7 meters in the conventional BWR. The short active fuel length reduces the drop in core pressure, and overcomes the natural circulation system. And by using the lower steam velocity in the upper plenum in the reactor pressure vessel (RPV), we can adopt a free surface separation (FSS) system. The FSS eliminates the need for a separator and thus helps minimize the RPV and PCV sizes. In order to improve safety efficiency, developing an Emergency Core Cooling System (ECCS) for the DMS was considered. The ECCS configuration in the DMS was examined to achieve core coverage and economic efficiency from the following. 1: Eliminating high-pressure injection systems. 2: Adopting passive safety-related systems. 3: Optimizing distribution for the systems and power source for the ECCS. In this way the configuration of the ECCS for the DMS was established, providing the same level of safety as the ABWR and the passive systems. Based on the results of Loss of Coolant Accident (LOCA) analysis, core cover can be achieved by this configuration. Therefore, the plant concept was found to offer both economic efficiency and safety.Copyright

Transient Performance of Medium Small LWR “DMS-400” Evaluated Using TRACG Code

A new concept of a small and medium sized light water reactor, named the DMS (double MS: modular simplified & medium small reactor) has been developed. The DMS features significantly simplified plant systems realized by adoption of a natural circulation system of coolant and a free surface separation (FSS) system that is based on the gravitational separation of steam and water. With these systems, reactor internal pumps and steam separators are not needed, reducing plant cost. In this study, the DMS plant performance under transient conditions has been evaluated using TRACG code. TRACG code, which can treat multi-dimensional hydrodynamic calculations in a reactor pressure vessel (RPV), is well suited for evaluating DMS reactor transient performance because it can evaluate the void fraction in the chimney and therefore evaluate the natural circulation flow. As critical transient cases, generator load rejection with total turbine bypass failure (LRNBP) and loss of feedwater heating (LFWH) were chosen to evaluate. LRNBP and LFWH are the most severely recognized events as a pressure increase event and a thermal power increase event, respectively. In case of LRNBP, heat flux increased to about 110% of rated power, and the natural circulation flow barely changed, resulting in a lower ΔMCPR than that of LFWH case. The reason that heat flux only increased to 110% was because the RPV of the DMS has a large steam region volume in the chimney region compared to the thermal power. As a result, the change in the void fraction with a pressure increase in the core was small. In case of LFWH, the maximum heat flux, calculated using the neutron flux, was 121% of rated power when a scram occurred, and ΔMCPR was 0.14, almost the same as for current ABWRs. Since the analysis conditions were set conservatively, these results show that the DMS performs as well for transient events as conventional BWRs.© 2008 ASME