Kumiaki Moriya

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.

Application of the Resource-Renewable Boiling Water Reactor for TRU Management and Long-Term Energy Supply

The RBWR (resource-renewable boiling water reactor) is an innovative BWR that has a capability to breed and burn trans-uranium elements (TRUs) using a multi-recycling process. The RBWR can be used as a long-term energy supply, and it reduces the negative environmental impact that TRUs cause as they are otherwise long-lived radioactive wastes. Various design concepts of the RBWR core have been proposed. The RBWR-AC is a break-even reactor and the RBWR-TB and RBWR-TB2 are TRU burners. The RBWR-TB is designed to burn TRUs from the RBWR-TB itself and to burn almost all the TRUs by repeating their recycling. The RBWR-TB is assumed to be applied for a nuclear power phase-out scenario. The RBWR-TB2 is intended to burn TRUs from LWR spent fuels. The RBWR-TB2 is assumed to be applied for reducing the amount of TRUs to be managed in storage facilities. The RBWR cores achieve their TRU multi-recycling capability under the constraint that the void reactivity coefficient must be negative by introducing the parfait core concept. This chapter reviews details of the specific design and core characteristics of the RBWR.

Simulation tests of BWR non-loca transients using the tbl

Abstract In the last decade, a large number of experiments have been performed in order to understand the thermal-hydraulic response in a boiling water reactor (BWR) under postulated loss of coolant accident (LOCA) conditions. These experimental results showed that the core cooling effect under the LOCA conditions was significantly affected by three-dimensional and multi-bundle phenomena after emergency core cooling systems (ECCSs) started. Also, the peak cladding temperature (PCT) during the LOCA was kept below a specific value of the licensing acceptance criteria, 1473 K (1200°C). These key results of the experiments were incorporated into a computer code, SAFER, which was developed for the BWR LOCA/ECCS analyses under the cooperative studies of Hitachi Ltd, Toshiba Co., and General Electric Co. (GE). In a couple of years, the experimental study of multi-bundle phenomena was extended into the area of off-normal and non-LOCA transients. Thermal-hydraulic responses during boiling transition were studied using the TBL (Two Bundle Loop) test facility with two full-length bundles. The experimental results showed that interaction and feedback effects between the bundles were expected to be unaffected by core cooling during the typical off-normal and non-LOCA transients. Also, the SAFER showed good predictions for hydraulic responses in the bundles and temperature transients of the rod surfaces.