Masayoshi Shiba

ROSA-III Base Test Series for a Large Break Loss-of-Coolant Accident in a Boiling Water Reactor

The rig of safety assessment (ROSA)-III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) tests. It is confirmed from the experimental results obtained so far that the ROSA-III test facility can simulate major aspects of a BWR LOCA, such as boiling transition by lowering of the mixture level in the core, rewetting by the lower plenum flashing, and final quenching by the ECCS. The overall agreement between the calculated results by the RELAP5/ MOD0 code and the experimental results is good; however, the calculated lower plenum flashing rewetted the whole core and the calculated cladding temperature considerably underpredicts the measured value at the upper part of the core.

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Abstract Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.

ROSA-III Double-Ended Break Test Series for a Loss-of-Coolant Accident in a Boiling Water Reactor

The Rig of Safety Assessment (ROSA) III facility is a volumetrically scaled (1/424) boiling water reactor (BWR) system with an electrically heated core designed for integral loss-of-coolant accident (LOCA) and emergency-core-cooling-system (ECCS) tests. Experimental results obtained so far confirm that the severest single failure assumption in ECCS is the high-pressure core spray system failure even in a large-break LOCA in a BWR. The measured peak cladding temperature was well below the present safety criterion of 1473 K, even with the single failure assumption in ECCS, and the effectiveness of ECCS for core cooling during a double-ended-break LOCA has been confirmed. The overall agreement between the results calculated by the RELAP4/MOD6/U4/J3 computer code and the experimental results is good. The similarity between the ROSA-III test and a BWR LOCA has been confirmed through the comparison of calculated results for the ROSA-III facility and a BWR system.

LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates (∼ 30 kg/m2·s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.

ROSA-IV/LSTF Program at JAERI for PWR Small-Break LOCA and Operational Transient Experiments

The Japan Atomic Energy Research Institute (JAERI) has initiated the Rig of Safety Assessment, Number 4 (ROSA-IV), program to study small-break loss-of-coolant accidents and operational transients. This paper describes the design features of LSTF and the results of the pre-analysis of LSTF tests.

Verification study on alternative ECCS concepts for a PWR

Abstract An experimental study for alternative ECCSs for a PWR was performed with the ROSA-II facility. It was found through the tests that the combined injection of hot water into the upper plenum and cold water into the lower plenum accompanied by a low pressure coolant injection system in the hot legs is quite effective for core cooling through the whole period of a LOCA in the case of a cold leg break. The test results were compared with analytical results of the RELAP4J code. The code is found capable of estimating discharge flow behavior fairly well and can predict the overall fluid behavior in the tested method of the improved ECCSs. However, the calculated core heat transfer disagrees with the test data when the counter-cuurent flow of the two phases on the core is dominant.