Toshio Fujishiro

Transient fuel behavior of preirradiated PWR fuels under reactivity initiated accident conditions

Abstract Since 1975, extensive studies on transient fuel behavior under reactivity initiated accident (RIA) conditions have been continued in the Nuclear Safety Research Reactor (NSRR) of Japan Atomic Energy Research Institute. A new experimental program with preirradiated LWR fuel rods as test samples has recently been started. In this program, transient behavior and failure initiation have been studied with 14 × 14 type PWR fuel rods preirradiated to a burnup of 20 to 42 MWd/kgU. The test fuel rods contained in a capsule filled with the coolant water were subjected to a pulse irradiation in the NSRR to simulate a prompt power surge in an RIA. The effects of preirradiation on the transient fission gas release, pellet-cladding mechanical interaction and fuel failure were clearly observed through the transient in-core measurements and postirradiation examination.

Dimensional Stability of Uranium Silicide Plate-Type Reactors at Transient Condition

This paper describes the result of transient experiments using low enriched uranium silicide plate-type fuel for research reactors. The pulse irradiation was carried out at Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Research Institute. The results obtained were: 1. At fuel plate temperature of below 400°C, a good dimensional stability of the tested fuel was kept. No fuel failure occurred. 2. At a plate temperature of about 540°C, a local crack was initiated on the AI-3%Mg alloy cladding. Once the cladding temperature exceeded the melting point of 640°C, the fuel plate was degraded much by increased bowing and cracking of the denuded fuel meat occurred after relocation of molten Al cladding. Despite of these degradation, neither fragmentation of the fuel plate nor mechanical energy generation occurred up to the cladding temperature of 971 °C. 3. At the temperatures of around 925°C, the reaction of silicide particles with molten Al in the matrix and that of cladding occurred, forming Al r...

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.

Generation of destructive forces during fuel/coolant interactions under severe reactivity initiated accident conditions

Abstract During a reactivity initiated accident (RIA) a large and prompt amount of energy is deposited within fuel rods. Consequent fuel melting and rod failure can lead to a fine dispersal of fuel melt in coolant, resulting in a violent thermal interaction between the fuel melt and coolant (i.e. fuel/coolant interaction — FCI). In the pre-mixing stage, the nature of the fuel release mode from failed rods is dominant factor influencing FCI energetics, where the internal rod pressure is a prime factor influencing melt jet characteristics. In this study, the generation of the destructive forces during FCIs in a severe RIA was demonstrated through in-pile experiments with LWR uranium dioxide fuels. The intensities of the destructive forces, i.e., the pressure pulse and the water hammer, were correlated with the initial internal pressure of the test fuel rods. Debris particle size distribution was also extensively examined. Rosin-Rammlers law was used to fit the distribution, and the correlation between the intensities of the destructive forces and the specific surface area of the debris is presented.

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.

Studies of the UO2-zircaloy chemical interaction and fuel rod relocation modes in a severe fuel damage accident

Abstract Experiments have been conducted in the Nuclear Safety Research Reactor (NSRR) at JAERI since 1975 in order to study fuel rod failure behavior under reactivity-initiated accident conditions. Recently the experiments have been focussed on fuel behavior under simulated severe fuel damage (SFD) accident conditions. UO2-Zircaloy reaction kinetics during very rapid transients at elevated temperatures was studied from a metallurgical point of view. Equilibrium was found to be established even in very rapid transients. The reaction rate equations developed in isothermal studies can be applied to interpret the experimental results. A fuel rod relocation criterion in connection with peak temperatures, environment conditions and initial fuel rod conditions was developed. According to the test results, fuel rod melt down due to liquefaction seems unlikely below the melting temperature of β-Zircaloy.

Hydrogen Generation during Cladding/Coolant Interactions under Reactivity Initiated Accident Conditions

Hydrogen generation in chemical reaction of Zircaloy cladding with coolant water under reactivity initiated accident conditions (RIAs) has been studied with the in-pile experiments in the Nuclear Safety Research Reactor (NSRR). PWR type segmented fuel rods were subjected to pulse irradiation to simulate the power excursion of an RIA. Transient measurements of the void fraction with a densimeter developed for the in-pile experiment detected prompt generation of hydrogen with an increase in cladding surface temperature. The hydrogen generation ceases in the initial 4 seconds of the power excursion even in the experiments resulting in the melting of the fuel cladding and severe damage. The total amount of hydrogen generated during the power burst was estimated by integration of the densimeter data and also by metallographic examinations of the oxidized cladding. The total amount of hydrogen produced increased with an elevation in the maximum temperature of the cladding surface, and the hydrogen generation wa...

Fuel behavior in simulated RIA under high pressure and temperature coolant condition

Abstract Fuel behavior in simulated reactivity initiated accident (RIA) was studied under high pressure and temperature coolant condition of LWR operating environment. Test fuel rods were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to simulate transient power generations at RIAs. From the results, it was clarified that the cladding collapsed under high external pressure but that the basic mechanisms and threshold enthalpies of incipient fuel failure were not different from those observed in the tests conducted under atmospheric pressure, room temperature and stagnant condition.

Effects gap heat transfer on LWR fuel behavior during an RIA transient: In-pile experimental results with helium and xenon fillet rods

Abstract Gap heat transfer characteristics and their effects on LWR fuel behavior during an RIA have been studied through the in-pile experiment with UO 2 pellet fuel rods. The report describes the experimental results obtained in the NSRR tests in which PWR type test fuel rods of helium and xenon filled as the gap gas have been irradiated in the pulse reactor, NSRR, to simulate the prompt heat up of RIAs. The relation between the cladding temperature history and the gap heat transfer conditions, and the effects of gap gas composition on fuel behavior and on the fuel failure threshold are discussed based on the in-pile experimental data.

Inpile Test Rigs for Ria Experiments in the NSRR

This report introduces the major inpile test rigs for the NSRR project conducted in Japan Atomic Energy Research Institute since 1975. The NSRR project is the inpile experimental program to understand the behavior of LWR fuels under reactivity initiated accident (RIA) conditions, LWR fuel rods are irradiated in the pulse reactor, NSRR, to realize a prompt power excursion in the RIA. Different types of inpile capsules and loops have been developed to accomodate the test conditions for the various parametric studies on the RIA fuel behavior. Major design features and test capabilities of these test rigs and the instrumentations for them are described.