In Japan, the simplified hybrid method proposed by Kennedy is referred to (e.g., in the Atomic Energy Society of Japan standard (draft) on the risk assessment method for nuclear fuel facilities) as one of the seismic probabilistic risk assessment methods for nuclear fuel facilities. Although this method enables us to easily evaluate a plant’s annual probability of failure, the evaluated probability may occasionally be overestimated or underestimated depending on the analysis case. Such a tendency is a major obstacle to judging the appropriateness of the evaluation results of the method. In this paper, the cause of overestimation and underestimation is analyzed, and ways to improve the simplified hybrid method are studied and proposed to enable evaluation results within an acceptable range, which is defined in this paper, to be obtained.
In this paper, several modified stainless steels (SSs) with different contents of Ta (0.13–0.61%) and C (0.010–0.046%) were produced to measure the electrochemical potentiokinetic reactivation (EPR) ratio and investigate their crevice corrosion resistance, stress corrosion cracking (SCC) susceptibility and crack growth rate (CGR) in a simulated boiling water reactor (BWR) environment. As a result of the EPR tests, we found that the Ta/C ratio must be ≥ 13 to suppress sensitization by stabilization heat treatment. If the Ta/C ratio is ≥ 19, sensitization can be suppressed without stabilization heat treatment. In the crevice corrosion test under γ-ray irradiation, the maximum corrosion depth in the Ta-modified SSs was smaller than that in type 316L SS. Ta-modified SSs had better crevice corrosion resistance than 316L SS. In the creviced bent beam test, there was no SCC in any of the Ta-modified SSs, whereas cracks were found in four of seven specimens of 316L SS. The CGR test was conducted using 0.5 T-compact tension specimens. Crack growth rates of the Ta-modified SSs were lower than that of 316L SS. The crevice corrosion resistance and SCC resistance were improved by Ta addition. We assumed that Ta addition can improve the repassivation response.
SiC, which is a promising accident-tolerant fuel cladding, is a non-oxide, and it is known that passive oxidation occurs, where by a protective oxide film of SiO2 is formed under atmospheric conditions above 900℃. The reaction occurring at this high temperature is important in assessing the soundness of SiC during a severe accident, but the understanding of it is still insufficient. In this study, to evaluate the high-temperature oxidation behavior when SiC cladding is exposed to the atmosphere (105 Pa) during an accident involving a light-water reactor, an oxidation test was performed for up to 100 h at 1100 to 1500℃. As a result, a SiO2 oxide film was formed on the surface of SiC, but the formation of bubbles originating from impurities and cracks due to a phase transformation was confirmed. In addition, it was observed, for the first time in this research, that a multilayered SiO2 oxide film was formed at 1500℃. Therefore, it was shown that the oxidation reaction of SiC does not stop depending on the surrounding conditions under high temperature and atmospheric conditions.
After direct discharges of highly contaminated water from Units 2 and 3 of the Fukushima Daiichi Nuclear Power Plant (1F) from April to May 2011, Kanda suggested that relatively small amounts of run-off of radionuclides from the 1F port into the Fukushima coastal region subsequently continued, on the basis of his estimation method. However, the estimation period was limited to up to September 2012, and there has been no report on the issue since that work. Therefore, this paper focuses on the discharge inventory from the 1F port up to June 2018. In the missing period, the Japanese government and Tokyo Electric Power Company Holdings have continued efforts to stop the discharge, and consequently, the radionuclide concentration in seawater inside the 1F port has gradually diminished. We show the monthly discharge inventory of 137Cs up to June 2018 by two methods, i.e., Kanda’s method partially improved by the authors and a more sophisticated method using Voronoi tessellation reflecting the increase in the number of monitoring points inside the 1F port. The results show that the former always yields overestimated results compared with the latter, but the ratio of the former to the latter is less than one order of magnitnde. Using these results, we evaluate the impact of the discharge inventory from the 1F port into the coastal area and the radiation dose upon fish digestion.
The development of fabrication and inspection technologies for an oxidation-resistant fuel element was conducted referring to previous research to improve the safety of high-temperature gas-cooled reactors (HTGRs) during a severe oxidation accident. Herein, simulated coated fuel particles (CFPs), alumina particles, were coated with a mixture of Si and C powders and a small amount of resin to form “over-coated” particles that were subsequently molded and hot-pressed to sinter the simulated oxidation-resistant fuel elements with a SiC/C mixed matrix. SiC was formed by reaction bonding. Simulated oxidation-resistant fuel elements containing a matrix with a Si/C mole ratio of 1.00 were fabricated. Elemental Si and C peaks were not detected in X-ray diffraction of the matrix. The failure fraction of CFPs in fuel elements is a very important HTGR fuel inspection subject; it is essential that CFPs are extracted from fuel elements without additional failure. Herein, a method of extracting CFPs was developed. The dissolution of SiC using KOH or by pressurized acidolysis should be applied to extract CFPs. However, the outer high-density pyrolytic carbon layer should remain in spite of its transformation to SiC during sintering by reaction bonding with Si in the mixed powder.