A series of structural analyses of disposal containers for the direct disposal of spent fuel was carried out to provide preliminary estimates of the required thickness for adequate pressure resistance of the disposal container. Disposal containers were designed to contain either 2, 3 or 4 spent fuel assemblies in linear, triangular or square arrangements, respectively. The required pressure resistance thickness was evaluated using the separation distance of the housing space for each spent fuel assembly as a key model parameter to obtain the required thickness of the body as well as the lid of the disposal container. In this work, we also provide additional technical knowledge on the structural analysis of disposal containers, such as the validity of the setting of the stress evaluation line and the effect of the model length on the analysis. These can then be referred to and used again in the future as a basis for conducting similar evaluations under different conditions or proceeding with more detailed evaluations.
A deliberative poll (DP) is a new type of democratic process in which citizens selected at random from the voters are educated on and discuss given political issues, and then the results are reflected in policy. An example of this occured in 2012 when the Japanese government conducted a DP on nuclear power generation goals for 2030. As a result, support for abolishing nuclear power increased significantly. On the other hand, in 2017, the Korean government conducted a DP on the construction of the Shin-Gori nuclear reactors and on future nuclear policy. As a result, support for the continued construction of the reactors and phase out of nuclear power was strengthened. In Japan’s DP compared with that of Korea’s, Japan’s selected citizens were far less representative of the population, stakeholder’s involvement was nonexistent, and the government was much less committed.
Understanding the behavior of melted volatile fission products (FPs) on the fuel contributes to the precise assessment of the release behaviour during a severe nuclear accident. A previous study revealed that liquid CsI shows abnormally high wettability with measured contact angles of almost zero degrees against the polycrystalline UO2 solid surface. [K. Kurosaki et al., Sci. Rep. 7, Article number: 11449 (2017).]. In this study, we focus on the melting behavior of CsIO3 and revealed that liquid CsIO3 also shows high wettability on the polycrystalline UO2 solid surface. However, after melting, CsIO3 decomposed and only Cs reacted with the polycrystalline UO2 solid surface and I was only absorbed on the solid surface. When the CsI had melted on the polycrystalline UO2 solid surface, both Cs and I were able to penetrate inside the UO2 pellets. In short, when Cs and I exist as CsIO3, Cs and I will be separately released during severe accidents. These findings suggest that the release mechanisms of Cs and I could be strongly affected by the chemical species in the irradiated fuels.
The gamma camera is a useful measuring device for grasping the distribution of contamination, but estimating the radioactivity of contamination is difficult. If the total energy absorption peak count rate of γ-rays can be measured with a pinhole gamma camera, we should be able to estimate the radioactivity. We can calculate it analytically by using the distance between the source and measurement points, and the peak count rate of direct γ-rays. The estimation method of 137Cs radioactivity using a pinhole gamma camera was studied using Hitachi’s gamma camera (HDG-E1500), which can be used to measure the total energy absorption peak count rate of γ-rays.
Best-estimate evaluation with detailed models of complicated phenomena that occur during accidents has been introduced into the safety evaluation of nuclear power plants. A system analysis code, which has physical models for the realistic prediction of events during accidents, is necessary for the safety evaluation. The analysis code should also be validated for individual phenomena and their combined behaviors at the actual plant scale during accidents. In this study, the system analysis code “AMAGI”, which is applicable to the evaluation of events from anticipated operational occurrences to design extension conditions, has been developed from its basic design. The thermal hydraulic model, heat conduction model, control model, and thermal power model were implemented into AMAGI as primary analytical functions. By conducting analyses of experiments with AMAGI, its fundamental models were validated.
The Kyoto University Accelerator-based Neutron Source (KUANS) is a compact neutron source that is mainly used for spectrometer and detector development. In addition, it is also suited for experiments to study the neutronic design of moderators owing to the relatively low neutron generation yield by 9Be(p,n). We present a neutronic design of the neutron moderator on a reentrant-hole configuration for KUANS to enhance the neutron emission, and some experiments are conducted at KUANS for verification. A polyethylene moderator on a reentrant-hole configuration is designed by PHITS calculation and is introduced to KUANS to obtain intense oblong neutron beams. The intensity of the pulsed neutron beam is experimentally measured. The results reveal that the intensity becomes approximately 1.9 times stronger than that of the conventional rectangular design. In addition, the ratio of its intensity to the conventional intensity increases to approximately threefold as the neutron wavelength increases. It is interesting to note that the longer the neutron wavelength, the more efficiently they are extracted from the inside of the moderator owing to the existence of the reentrant-hole configuration.