Kazuichiro Hashimoto

Experimental and Analytical Study on the Behavior of Cesium Iodide Aerosol/Vapor Deposition onto Inner Surface of Pipe Wall under Severe Accident Conditions

The WAVE experiments have been performed at JAERI to investigate the CsI deposition onto the inner surface of pipe wall under typical severe accident conditions. It was shown that relatively large amount of CsI was deposited at the upstream floor of the pipe and that larger amount of CsI was deposited on the ceiling than the floor at the downstream. Analyses of the experiments have also been conducted with the three-dimensional thermohydraulic code, SPRAC, and the radionuclide transport analysis code, ART. The experimental results were well reproduced with ART by using peripherally subdivided pipe cross section and associated representative thermohydraulic information from SPRAC prediction. It was clarified through the present experiment and analyses that major deposition mechanisms for the chemical form of CsI are thermophoresis and condensation. Accordingly, the coupling of the FP behavior and the detailed thermohydraulic analyses was found to be essential in order to accurately predict the CsI depositi...

Vapor Condensation and Thermophoretic Aerosol Deposition of Cesium Iodide in Horizontal Thermal Gradient Pipes

The aerosol deposition test series is being performed to investigate the deposition of FP vapor and aerosol onto the inner surface of reactor coolant piping during a severe accident of a light water reactor. Vapor and aerosol of CsI as an FP simulant was introduced into the horizontal test section pipes. No substantial decomposition of CsI was identified to occur both in the high temperature inert and superheated steam environments. The comparison between the results of the aerosol deposition test series and the thermo-fluiddynamic analysis with WINDFLOW implied that a profile of the CsI deposition due to vapor condensation and thermophoretic aerosol deposition was influenced by local thermo-fluiddynamic conditions. Deposition velocities were evaluated for CsI vapor and aerosol based on the deposition characteristics of CsI and the thermo-fluiddynamic analysis.

Experimental and analytical study on aerosol behavior in WIND project

The tests on fission product (FP) behavior in piping under severe accidents are being conducted in the wide range piping integrity demonstration (WIND) project at JAERI to investigate the piping integrity which may be threatened by decay heat from deposited FPs. In order to obtain the background information for future WIND experiment and to confirm analytical capabilities of the FP aerosol analysis codes, ART and VICTORIA, the FP behavior in safety relief valve (SRV) line of BWR during TQUX sequence was analyzed. The analyses showed that the mechanisms that control the FP deposition and transport agreed well between the two codes. However, the differences in models such as diffusiophoresis or turbulence, the treatment of chemical forms and aerosol mass distribution could affect the deposition in piping and, consequently, on the source terms. The WIND experimental analyses were also conducted with a three-dimensional fluiddynamic WINDFLOW, ART and an interface module to appropriately couple the fluiddynamics and FP behavior analyses. The analyses showed that the major deposition mechanism for cesium iodide (CsI) is thermophoresis which depends on the thermal gradient in gas. Accordingly, the coupling analyses were found to be essential to accurately predict the CsI deposition in piping, to which little attention has been paid in the previous studies.

Performance of Containment Sprays for Light Water Reactors and Evaluation of the Heat Transfer

Capacity curves, spray distributions, and droplet size distributions of light water reactor (pressurized water reactor (PWR) and boiling water reactor (BWR)) containment spray nozzles are obtained, and the heat removal effectiveness is evaluated by a computer program CONDENSE. It is revealed by the calculations that spray droplets from a PWR spray nozzle always attain the containment atmosphere temperature and spray droplets from a BWR spray nozzle attain the containment atmosphere temperature above approximately 70/degree/C. 16 refs.

Release of Radioactive Materials in Simulation Tests of a Postulated Solvent Fire in a Nuclear Fuel Reprocessing Plant

This paper reports on small- and large-scale fire tests performed to examine the adequacy of a safety evaluation method for a solvent fire in the extraction process of a nuclear fuel reprocessing plant. The test objectives were to obtain information on the confinement of radioactive materials during a 30% tri-n-butyl phosphate-n-dodecane fire while air ventilation is operating in the cell. The rates of release of cesium, strontium, cerium, ruthenium, and uranium from a burning solvent were determined. The quantities of species released were obtained from the solvent burning rate, smoke generation rate, partition coefficients of species between solvent and water, and coefficients of species entrainment to atmosphere in cell.

Iodine removal tests for BWR containment spray by large scale facility.

Iodine removal tests for a BWR containment spray were carried out with large-scale JAERI Model Containment Test Facility under LOCA simulated conditions. The tests consisted of two groups: “gas-phase based” tests mainly to obtain the initial iodine removal rate by the spray and “liquid-phase based” tests to obtain the iodine partition coefficient at equilibrium state. It was shown that the degree of iodine removal was largely influenced by pH-value of spray water. The results were discussed with calculated results by a code MIRA-PB using a dose reduction factor for the airborne iodine.

Fire Behavior and Filter Plugging During a Postulated Solvent Fire in the Extraction Process of a Nuclear Fuel Reprocessing Plant

A solvent fire in the extraction process of a fuel reprocessing plant is postulated. Because of the high concentration of fission products and large amount of nuclear fuel materials in the extraction process, it is necessary to demonstrate that these radioactive materials can be confined by the air ventilation system during a solvent fire. Large-scale tests are performed in a fire/filter facility to evaluate the effectiveness of a ventilation system including high-efficiency particulate air (HEPA) filters to confine radioactive materials. It is demonstrated that the integrity of the filters in the ventilation factor of HEPA filters for smoke particles, which might contain radioactive materials, is sufficiently high during a postulated solvent fire.

Summary of the Three Mile Island Unit 2 Analysis Exercise

AbstractAn overview is presented of the current activities within the international consortium participating in the Three Mile Island Unit 2 (TMI-2) Analysis Exercise, which is part of the Organization for Economic Cooperation and Development Nuclear Energy Agency/U.S. Department of Energy Joint Task Group program on TMI-2, formed to utilize the TMI-2 accident as a benchmark for severe accident computer codes. The participants have utilized various state-of-the-art severe core damage analysis computer codes to simulate the TMI-2 accident. The results of the analyses, although qualitatively similar, are quantitatively quite different. This indicates that continued development of these codes is desirable.

Effect of Microstructure on Failure Behavior of Light Water Reactor Coolant Piping under Severe Accident Conditions

In a severe accident of light water reactors, the reactor coolant system (RCS) piping might be subjected to thermal loads caused by the decay heat of the deposited fission products and the heat transfer from the hot gases, with an internal pressure in some accident sequences. Tests on the RCS piping failure were performed along with high temperature tensile and creep rupture tests including metallography to investigate the failure behavior. The prediction of the 0.2% proof stress by Arrhenius equation is in good agreement with the measured stress above 800°C for served RCS piping materials. The modified Nortons Law for the short term creep rupture model agrees with the experimental values between 800 and 1,150°C for type 316 stainless steel. The microstructural change was discussed with the effect of the very rapid formation and resolution of the precipitation on the strength at high temperature. The result of the piping failure tests which simulated the severe accident conditions, i.e., in short-term at...

Heat Removal Tests for Pressurized Water Reactor Containment Spray by Large-Scale Facility

Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m/sup 3/ volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm/sup 2/ X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition.