Shoji Katanishi

Performance Test Results of Mock-up Test Facility of HTTR Hydrogen Production System

For the purpose to demonstrate effectiveness of high-temperature nuclear heat utilization, Japan Atomic Energy Research Institute has been developing a hydrogen production system and has planned to connect the hydrogen production system to High Temperature Engineering Test Reactor (HTTR). Prior to construction of a HTTR hydrogen production system, a mock-up test facility was constructed to investigate transient behavior of the hydrogen production system and to establish system controllability. The Mock-up test facility with a full-scale reaction tube is an approximately 1/30-scale model of the HTTR hydrogen production system and an electric heater is used as a heat source instead of a reactor. After its construction, a performance test of the test facility was carried out in the same pressure and temperature conditions as those of the HTTR hydrogen production system to investigate its performance such as hydrogen production ability, controllability and so on. It was confirmed that hydrogen was stably produced with a hot helium gas about 120m3/h, which satisfy the design value, and thermal disturbance of helium gas during the start-up could be mitigated within the design value by using a steam generator. The mock-up test of the HTTR hydrogen production system using this facility will continue until 2004.

High Power Transient Characteristics and Capability of NSRR

In order to study fuel behavior under abnormal transients and accidents, the control system of the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute (JAERI) was modified to achieve high power transients. With this new operational mode, called Shaped Pulse (SP), transients at the maximum power of 10 MW can be conducted for a few seconds. This new operational mode supplements the previous Natural Pulse (NP) operation at the maximum power of 23 GW for milliseconds. For high power transient operation, a simulator using a point kinetic model was developed, and characteristics of the NSRR in the new operational mode were examined through tests and calculations. With the new operational mode, new types of fuel irradiation tests simulating power oscillations of boiling water reactors (BWRs) can be conducted in the NSRR. Reactor characteristics and capability, such as control rod worth, feedback reactivity, and operational limits of the NSRR for SP operations are discussed.

Quenching degradation in-pile experiment on an oxidized fuel rod in the temperature range of 1000 to 1260°C

Abstract Under core uncovery accident conditions, the cladding tube of a fuel rod will be oxidized and embrittled. The fuel degradation conditions due to the thermal shock during delayed reflooding need to be studied. In the Nuclear Safety Research Reactor (NSRR) of the Japan Atomic Energy Research Institute, the sequences in a severe accident were simulated to investigate the in-core fuel degradation due to quenching. With these in-pile experiments, the oxidation behavior of the Zircaloy cladding tube was clarified at temperatures ranging 1000–1260°C, and it was shown that there was fuel degradation due to the thermal shock by the reflooding after the cladding was exposed to high-temperature steam for a relatively long time. Analysis of the test results was also performed using the SCDAP code to evaluate the applicability of this code to these particular tests and to obtain supporting data for the test results. Generally, the calculated results agreed well with the test results. However, at lower elevation of the fuel rod, the predicted cladding temperature and oxide layer thickness overestimated the test results due to the modeling of the cooling effect by steam flow.