It is pointed out that a vapor film on a premixed high-temperature droplet surface is needed to be collapsed to trigger vapor explosion. Thus, it is important to clarify the micromechanism of vapor film collapse behavior for the occurrence of vapor explosion. In a previous study, it is suggested experimentally that vapor film collapse behavior is dominated by phase change phenomena rather than by the surrounding fluid motion. In the present study, vapor film collapse behavior is investigated to clarify the dominant factor of vapor film collapse behavior with lattice gas automata of three-dimensional immiscible lattice gas model (3-D ILG model). First, in order to represent the boiling and phase change phenomena, the thermal model of a heat wall model and a phase change model is newly constructed. Next, the numerical simulation of vapor film collapse behavior is performed with and without the phase change effect. As a result, the computational result with the phase change effect is observed to be almost same as the experimental result. It can be considered that vapor film collapse behavior is dominated by phase change phenomena.
One of the key technology requirements to achieve the non-nuclear-grade hydrogen plants coupled to the HTTR is the development of control methods mitigating the effect of IS process-abnormal thermal load transients on the reactor operation. This study focused on developing the control methods for the thermal mitigation system in the HTTR-IS system. The control methods are developed to stabilize the steam generator helium gas outlet temperature increase against the high-temperature helium gas inflow. The key parameters and equipment are determined and the control operations are simulated. The simulation results show that the steam generator water temperature can be effectively decreased by interfacing the actuation of the air cooler inlet shut-off valve with the actuation of the diverter valve. Furthermore, the air cooler outlet shut-off valve is activated as the pressure difference approaches zero between the steam generator bottom and the down comer from the air cooler to the steam generator. It is demonstrated that the control methods developed enable the continuation of the nuclear reactor normal operation under IS process abnormal conditions and without operator intervention.
The reactor water was fed to the purge water of the mechanical seal on the original design of the primary loop recirculation pump. Because the mechanical seal had a short life due to the cruds in the reactor water, the clean purge water was adopted instead of the reactor water. After this modification, the shallow cracks were found on the surface of the pump shaft and casing cover due to the temperature fluctuation between the cold purge water and the hot pump discharge water. The fundamental mechanism and countermeasure were investigated by scale test, mock-up test and so on. The flow barrier with a heater was contrived through these tests. It has been introduced gradually in operating and constructing PLR pumps after its completion in 1995. The PLR pumps are overhauled around every 10 years in Japan. The first overhaul of the PLR pumps showed no cracks around the pump shaft and casing cover after 10 years' operation. This paper presents both its development process and inspection results.
The social research group of the 21st century COE program “Innovative Nuclear Energy Systems for Sustainable Development of the World” has studied under the theme coevolution of nuclear technology and society. As part of this study, this group conducted a questionnaire survey of 2,500 adults (collection rate of 22.0%; 551 replies) who live in the Tokyo metropolitan area. The purpose of this survey asking opinion about the relationship between attitude toward nuclear technology utilization and social awareness is to determine their request, exception and concern about nuclear technology utilization. The survey reveals that the differences of attitudes towards nuclear technology utilization can be explained in terms of differences of general views on the society, such as the directionality of social progress. Thus, it is necessary to argue with citizens about the strategy on nuclear technology utilization from the viewpoint of the directionality of the future society. The social decision-making process on nuclear technology utilization has to be renovated through dialogue among citizens as the partner taking on the achievement and contribution toward the directionality of the future society.
The study of a pyrochemical process by using alkalis molybdate melt has been carried out as a candidate reprocessing process for spent oxide fuels. In the previous studies, the UO2 pellet was dissolved into molten Na2MoO4-MoO3 mixtures and the dissolved uranium species in the melts were recovered by electrolysis. However, in a MOX dissolution test, we observed that iron species were dissolved into the melts from the stainless steel cladding tubes. The immersed corrosion tests of SUS316 tubes were thus carried out in molten Na2MoO4-MoO3 mixtures with MoO3 mole fractions ranging from 0.02 to 0.50 at 1023 K. Results showed that the concentration of dissolved iron and nickel species in the melt increased with an increase in MoO3 mole fraction, whereas the concentrations of dissolved chromium species remained low, independent of the MoO3 content. The cross-sectional morphology of the SUS316 specimen immersed for 2 h in the molten 0.5Na2MoO4-0.5MoO3 mixture at 1023 K clearly showed a layer of 22-29 nm thickness with a high content of chromium formed at the surface of the SUS316 specimen. It was thus found that the corrosion of SUS316 specimens in molten Na2MoO4-MoO3 mixtures at 1023 K proceeds with the iron and nickel preferably dissolved into the melt and with chromium-rich corrosion films formed at the surface.
The high-burnup BWR 9×9 lead use fuel assemblies (LUAs), which are designed for a maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiation in a commercial BWR to confirm the reliability of the current safety evaluation methodology and to accumulate data for judging the adequacy of its application to the future higher burnup fuel. The irradiation performance of 9×9 LUAs for two different designs, types A and B, is generally on the extended trend of 8×8 fuel, but some new findings in terms of fuel performance have been addressed after 5 cycle irradiations. Accelerated corrosion of cladding for the corner rods in Type-B fuel assemblies and spacers in both types is observed after 5 cycle irradiations. The increasing trend of high hydrogen concentration seems to be an issue, which should be paid much attention with respect to fuel integrity during high-burnup irradiation. The large difference in fission gas release rate between two types of fuel is confirmed after 3 and 5 cycle irradiations, and the release rate of Al-Si-O doped pellets is particularly higher than the others.
To date, human error has been the major cause of adverse events in nuclear power plants (NPPs), and thus, to decrease adverse events, it is crucial to prevent human errors. This study focused on worker errors, which are the most frequent human errors in NPPs, and a new method proposed for classifying measures into seven categories for error prevention. By using this new measure classification for the vertical axis of the matrix table and one of the classifications of human errors, such as background factors, error stages, and individual factors for the horizontal axis, three measure tables for preventing worker errors were created. The measures for preventing worker errors adopted by Japanese NPPs and general industry were arranged in the matrix tables. The tendencies of measures adopted by Japanese NPPs were then analyzed based on the measure tables, and several improvements to measures for decreasing worker errors were identified. The objective of this study was to decrease worker errors in NPPs by using these measure tables and by improving the measures based on the tendency analysis.
Failures on demand of a reactor core isolation cooling (RCIC) system in BWRs are the most frequent events of limiting conditions for operation during 1982-2006 in Japan, according to data gathered in Nuclear Information Archives (NUCIA). In this work, probabilities of failures of the RCIC system are analyzed by using the hierarchical Bayes method. The failures on demand of the RCIC system are classified into two groups; one is related to the demand at a periodical inspection test, which is performed almost every 13 months at the end of the periodical inspection of the nuclear power plant, and the other is related to the monthly surveillance test during plant operation. The hierarchical Bayes analysis shows the characteristics of probabilities of failures of each Japanese plant and also that probabilities of failures at the periodical inspection test are quite different from those at the surveillance test, comparing Japanese nuclear power plants with American ones. This paper provides a new approach to analyzing sparse failure data taken from nuclear power plants in Japan.
The melting temperatures of uranium plutonium mixed oxide fuel for fast reactor were investigated as functions of Pu content, Am content, and oxygen-to-metal (O/M) ratio using the thermal arrest technique. Rhenium inner was used for the measurement to prevent the reaction between the sample and capsule materials. The solidus temperatures decreased with increasing Pu and Am contents and increased with decreasing O/M ratio. It is considered that the maximum temperature in the U-Pu-O system varies in the hypostoichiometric composition region. The melting temperatures were evaluated by the ideal solid solution model derived to calculate the solidus and liquidus temperatures in the UO2-PuO2-AmO2-PuO1.7 system. The derived model reproduced the experimental data within an accuracy of ±25 K.
Nuclear power plants have used many contract workers. Their safety and health conditions are very important in Japan. Several amendments, which deregulate temporary personnel service and employment agency markets, have been done in recent years. The number of contract and temporary help agency workers have been rapidly increasing especially since the 1990s. As a result, ensuring the level of safety and health education and training of workers becomes a serious problem. This paper examines the possibility that the level of safety training of the contract workers is less than that of the direct-hire employees in nuclear power plants. We show that (1) the use of contract workers could be less efficient for ensuring the level of safety training, and (2) nuclear power plants still use contract workers in some situations in spite of the loss of efficiency. We also study legislations and past cases relating to nuclear power generation. We find that there are some structural problems that might make the contract workers less trained.
Tritium migration behavior in the next-generation nuclear plant (NGNP) employing a high-temperature electrolysis (HTE) process to produce hydrogen is estimated by numerical analysis. Estimated tritium concentrations in the hydrogen product and tertiary heat transport fluid in heat exchangers in the HTE process are higher than the limit in drinking water defined by the U.S. Environmental Protection Agency (EPA) and in the effluent at the boundary of an unrestricted area defined by the U.S. Nuclear Regulatory Commission (NRC), respectively. The effects of some countermeasures (i.e., reducing tritium release rate, increasing purification system capacity, removing tritium at high-temperature positions in the heat transport fluids, reducing the permeability of heat exchangers, and hydrogen or water injection in the heat transport fluids) to reduce tritium concentrations in the hydrogen product and tertiary heat transport fluid are proposed and evaluated. The alternative countermeasure proposed in this study to decrease the tritium permeation rate by water injection, which produces HTO from HT according to an isotope exchange reaction (HT+H2O=H2+HTO) in the heat transport fluids, is effective for decreasing the tritium concentrations.
Supercritical carbon dioxide (CO2) is being investigated as a material for a secondary cooling system of sodium (Na)-cooled fast reactor to avoid Na/water reaction. In this type of reactor, however, it is necessary to consider the consequences of Na/CO2 reaction, which might occur in the case of tube rupture in a heat exchanger between primary and secondary systems. Experiments were carried out with test equipment for the Na/CO2 reaction, which can handle 1-5 g order of Na and measure temperatures using thermocouples. The solid products of the Na/CO2 reaction sampled from the equipment were analyzed by X-ray diffraction (XRD) and chemical analysis. The parts of exhaust gases were analyzed by gas chromatography. From these experimental results, we proved that the reaction proceeded between liquid Na and CO2. The Na/CO2 reaction stopped only the pool surface reaction with a small quantity of aerosol emission when the initial temperature of Na was lower than 570°C. On the other hand, the reaction continuously proceeded with an orange-colored flame and aerosol release when the Na initial temperature was higher than 580°C, and the reaction products expanded to the margin of the Na pool tray.
A modular high-temperature gas-cooled reactor (HTGR) coupled with a closed cycle helium gas turbine is expected as one of the promising power generation plants due to inherent safety and economics. Effects of main design parameters, particularly gas pressure, on cycle thermal efficiency and power generation cost were examined for the block-type HTGR power plant. A previous design of 600 MWt reactor with inlet and outlet temperatures of 460 and 850°C and a gas pressure of 6 MPa was assumed as a reference design. A computer analysis program combining a core thermal design and a gas turbine cycle was developed. From the results of parametric analyses regarding both the pressure and the fuel channel diameter based on this program, the reactor pressure vessel (RPV) cooling flow ratio of 1% was clarified to be sufficient. With increasing pressure, pressure drop decreased, and consequently, a higher cycle thermal efficiency and a lower power generation cost were achieved. Namely, it was clarified that an optimal pressure exists at around 8 MPa, which is different from the existing generally employed pressure of 6 or 7 MPa. In addition, the effects of fuel cost escalation were examined.