Makoto Sobajima

Examination of the destructive forces in the chernobyl accident based on NSRR experiments

Abstract The possible causes of the destruction of the Chernobyl reactor core were examined by making use of the Nuclear Safety Research Reactor (NSRR) experimental results concerning the destructive forces generated by a fuel failure. A complementary experiment with Chernobyl reactor conditions was performed in order to observe the fuel failure behavior and the resultant vessel pressure rise, etc. Also, generation of hydrogen from the fuel rod cladding and the consequent system pressure rise were estimated based on the experiments. These examinations led to the conclusion that the most probable cause of the core pressure tube rupture in the accident was a static pressure rise due to rapid energy release from fragmented fuel. Other phenomena such as the hydrogen generation and molten fuel contact to the tube wall might have contributed to the tube rupture. The water hammer force is also estimated to have been large enough to break tubes even using conservative assumptions.

An Analysis of Transients in Experiments on Loss-of-Coolant Accidents

It was found that the results of the RELAP-3 code, which is one of the typical analytical codes for analysis of the loss-of-coolant accident (LOCA) of light water reactors, do not agree well with t...

Experimental study of effects of upward steam flow rate on quench propagation by falling water film.

Emergency core cooling (ECC) mater is carried up to the upper plenum and falls down again into the core during the reflood phase in PWR-LOCA. Therefore the quench front also propagates downward from the top of the core. The effect of upward steam flow rate on the top-down quench propagation was experimentally investigated. It was found that top-down quench velocity was delayed by upward steam flow. This effect is more significant when rod surface temperature is low and the falling water flow rate is small. The effect of the flow rate and the rod temperature on the quench velocity was correlated based on the experimental results under the conditions of atmospheric pressure, saturation temperature for water and steam, rod surface temperature of 350–600°C, down-ward water velocity of 0.01–0.1 m/s and upward steam velocity of 0–20 m/s.

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.

Carryover characteristic during reflood process in large scale separate effects tests

Abstract A study on carryover characteristic in the core and upper plenum during the reflood phase in PWR-LOCA was performed with the use of data from the slab core test facility (SCTF) having eight rod bundles scale. Void fraction distribution in the core was strongly related to quench propagation in rod bundles. A correlation for mass effluent rate out of core was derived from the void fraction distribution characteristic. The correlation was found to be widely applicable. On the other hand, the capture of entrained liquid in the upper plenum by structures and water pool is below 30% of the entrainment mass flow rate during most of the reflood phase and increase when the steam velocity decreases. Since entrainment rate into hot leg increases with increase of liquid flow out of the core, the reflood velocity should tend to be suppressed with time because of stronger steam binding effect.

Experimental study of two-dimensional thermal-hydraulic behavior in core during reflood phase of PWR LOCA.

Two-dimensional effects on the core cooling behavior during the reflood phase of a PWR-LOCA were experimentally studied by performing four tests with various radial core power profiles under the same total power and initial core stored energy conditions using the Slab Core Test Facility (SCTF). The heat transfer was enhanced and the cladding temperature was reduced for the higher and average power bundles in the steep radial power profile test especially at the upper elevation. The effect of radial power profile on the cladding temperature was quantitatively evaluated. For all tests with different radial power profiles, the collapsed water level in the upper plenum became higher in the hot leg side and the quench in the upper half of the core was delayed in the bundles corresponding to the outer bundles of a PWR core. The delay of the quench is considered to be caused by a flow stagnation trend in those bundles because the pressure in the outer bundles became higher than the pressure in the inner bundles ...

Coolability study on two-bundle scale flow blockage in the reflood process

Abstract A two-bundle scale channel flow blockage in the eight-bundle slab core test facility (SCTF) was examined for core cooling capability study in the reflood phase of a PWR-LOCA. The coplanar blockage with 60% blockage fraction resulted in promotion of quench for the elevation immediately above the blockage in relatively high reflood velocity tests. Conversely, this blockage led to delay of quench in low reflood velocity tests. Peak clad temperature, however, was not affected much by the existence of the blockage. These results are examined in comparison with the results of similar small scale test facilities. This examination revealed that the promotion of quench above the blockage was confined to a shorter length but the quench delay time was slightly longer for a large partial blockage than for a small blockage.

Spray Cooling Test: A Separate Effect Test with ROSA-III, a BWR-LOCA Simulation Facility

A steady separate effects test on BWR spray cooling was performed at relatively high system pressures using the ROSA-III test vessel. These tests were conducted in order to promote a better understanding of the thermal-hydraulic phenomena in LOCA experiments and to obtain information necessary for improvement of analytical codes. The fraction of entrainment or overflow for various spray conditions was obtained and the data of CCFL at the upper tie-plate were compared with correlations. It was shown that the occurrence of CCFL significantly diminished core cooling effects and that rod quench by fall back water was quite irregular and unstable. Reflood core cooling was also studied.

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.