Radioactive materials were released to the general environment due to the accident at the Fukushima Daiichi nuclear power station. The released radioactive materials fell and contaminated the land mainly in the Tohoku and Kanto areas of Japan. We surveyed the air dose rate in relation to the pave condition of the land, and investigated the contamination level in some nonpaved areas at the center of Fukushima City, Koriyama City, and Nasushiobara City. From the survey results, the dose rates of the nonpaved areas were found to be higher than those of the paved areas, and the dose rates of the paved areas depend on the paving materials of the area. The contamination level of the nonpaved area in Nasushiobara City was below the regulation level of specific activities in a radiation-controlled area in Japan. However, the contamination levels in the nonpaved areas in Fukushima City and Koriyama City were above the regulation level.
There has been some concern in reviewing the effectiveness of making decisions on the implementation of protective measures in emergency exposure situations. After the Fukushima Daiichi nuclear power plant accident, temporal changes in the concentration of iodine 131 in tap water were studied using published data from several authorities in Fukushima, Ibaraki, and Tokyo. Averted doses to infants (1-year-old children) due to the intake of iodine 131 through tap water restrictions were also evaluated. Consequently, it was found that the apparent half-life of iodine 131 in tap water was 2.8 days. The averted equivalent doses to the thyroids of 1-year-old children were found to have a maximum value of 8.3 mSv in a local area of Fukushima. Hence, the tap water restrictions implemented by the authorities were considered to be effective in the early phase of the emergency exposure situation.
In Fukushima and neighboring prefectures, the distributions of dose rate and γ-ray count rate of radionuclides from the Fukushima Daiichi Nuclear Power Station were measured on expressways on March 15, 16, 17, and April 8, 2011, using an NaI(Tl) detector and a LaBr3 detector. A radioactive plume that contained 133Xe, 132Te, 132I, 131I, 134Cs, and 136Cs was observed at Koriyama on the afternoon of March 15. The plume arrived in the Nakadori region of Fukushima prefecture, which is surrounded by two mountain ranges, and most of the radioactivity there was deposited by rainfall. Although the distributions of 132Te, 132I, 134Cs, 136Cs, and 137Cs were similar, the distribution of 131I was different from the others. The effective nuclides for the dose rate measurement were 132Te and 132I on March 15-17 and 134Cs and 137Cs on April 8. The initial distribution profile of the dose rate on March 15-17 was retained on April 8 because the deposited radioactive material was not moved from the initial location and there was no additional effective deposition of radioactivity.
The time variations in the dose rate and γ spectrum of radio nuclides originating from the accident of Fukushima Daiichi Nuclear Power Station were measured at Tsukuba City, Ibaraki, during the period from 15th March to 9th April 2011. The radiation dose peaked three times during the period from 15th to 16th March (1.27 μSv/h at maximum). The contribution of Xe-133 to the dose rate was observed from the γ spectrum obtained from the 5 h measurement during the peaks on 15th and 16th March, indicating that radioactive plume passed through Tsukuba City at that time. After the peaks, a dose rate increase with rainfall was observed on 21th March, dominating the integral dose rate measured at Tsukuba City. The dose after the rainfall comes from I-131, Cs-134, and Cs-137 that can be observed as peaks in the spectra.
In-cell solvent fire has been treated globally as a typical accident of a reprocessing plant. In order for the fire to break out, leakage of a significant amount of solvent, a temperature rise above its flash point, and the presence of ignition source have to occur simultaneously. In this paper, such an accident was analyzed from a probabilistic viewpoint. The occurrence frequency was shown to be 10−7/y because a frequency or probability of each event is so small. Nevertheless, we assumed the fire occurring for 1 h. One hour should be enough for the assessment of the consequence because an operator can implement measures to extinguish the fire by inducing oxygen deficiency in the cell, such as closing a fire damper or stopping the exhausting fan. Since a certain amount of soot produced by fire does not deteriorate the function of a HEPA filter, the effective dose for the general public is less than 0.1 mSv. The results will be useful for judging the safety importance level of the safety measures against fire.
In nuclear engineering, fluid flows involving free surface were well studied, e.g., pipe thinning produced by liquid droplet impingement, and steam explosion triggered by molten metal immersed in water. The moving particle simulation (MPS) method was often used in past studies. In this method, the Poisson equation was solved to obtain the pressure field. Solving the Poisson equation becomes a dominant process. Reducing the time of calculation of the Poisson equation is important for using the CFD in the engineering fields. Thereat, we propose a new MPS method by which the pressure field is calculated explicitly. We call it the explicit (E-)MPS method. The E-MPS method was applied to a static water and dam break problem to show its adequacy. Besides, we compare the calculation time between the E-MPS method and traditional one.
At the Ningyo-Toge Environmental Engineering Center of JAEA, the uranium enrichment plant has entered the decommissioning stage now. The gas centrifuge in this plant is contaminated by intermediate uranium fluoride. For the reasonable decommissioning of this plant, it has been planned to carry out the IF7 treatment of the gas centrifuge for system decontamination before dismantling. Although a few studies about IF7 treatment have been carried out in the small scale, the details about the decontamination performance in terms of level and time, for example, have not been clarified. Therefore, in order to clarify the decontamination performance, this study was carried out for full-scale examination using the actual gas centrifuge cascade. This paper shows the investigation result of reasonable IF7 treatment condition and the full-scale examination result under this IF7 treatment condition. As a result, it could be confirmed that 99% of intermediate uranium fluoride in the cascade was removed after 60 days. Furthermore, in this study, we carried out the examination of IF7 gas production needed in the IF7 treatment examination, too. This paper shows the investigation result of reasonable IF7 production condition and the full-scale examination result. As a result, it could be proved that 3,355 kg-IF7 gas was produced.
In uranium enrichment plant decommissioning, it is important to treat and dispose the gas centrifuge reasonably. At the Ningyo-Toge Environmental Engineering Center of JAEA, it has been planned to carry out IF7 treatment for the gas centrifuge in the uranium enrichment plant for decontamination before dismantling. In order to carry out this IF7 treatment, a nondestructive measurement technology to estimate the uranium weight in the interior of the gas centrifuge was needed. However, it was difficult to estimate exactly uranium weight using former methods. Therefore, in this study, we investigated a new nondestructive measurement. As a result, we succeeded in constructing theoretically the methodology using γ-ray intensity measured outside the gas centrifuge and MCNP code. Furthermore, the UF6 and IF5 mixture was recovered by IF7 treatment. A technology to separate and refine IF5 from this mixture was needed in order to use as a material of IF7 gas. Therefore, in this study, we carried out IF5 separation and refinement examination. This paper shows the investigation result of reasonable separation condition and the full-scale examination result. As a result, it could be proved that 86% of IF5 was separated and refined from UF6 and IF5 mixture.
Analytical methods have been developed for the simple and rapid determination of radioactive nuclides, which are selected as important nuclides for the safety assessment of the disposal of wastes generated from research facilities. We advanced the development of a high-efficiency nondestructive measurement technique for γ-ray-emitting nuclides, simple and rapid methods for the pretreatment of hard-to-dissolve samples and subsequent radiochemical separation, and rapid determination methods for long-lived nuclides. In order to establish a system to analyze the important nuclides in various kinds of sample, actual radioactive wastes such as concentrated liquid waste, activated concrete, and metal pipes were analyzed by the present method. The results showed that the present method was well suited for a rapid and simple determination of low-level radioactive wastes generated from research facilities.