Time histories of the I-131 concentration in air at monitoring posts in Fukushima prefecture in March 2011 were estimated using the pulse height distribution of a NaI(Tl) detector, which was opened to the public. Several corrections to the pulse height distribution were necessary owing to high count rates. The contribution to the count rates from I-131 accumulated around the monitoring post was estimated on the basis of the time history of the peak count rate by the method proposed by the authors. The concentrations of I-131 in air were converted from the peak count rates using the calculated response of the NaI(Tl) detector with egs5 for a model of a plume containing I-131 uniformly. The obtained time histories of the I-131 concentration in air at a fixed point in March 2011 were the first ones for Fukushima prefecture. The results at 3 monitoring posts, Naraha Town Shoukan, Hirono Town Futatunuma and Fukushima City Momijiyama, which can be analyzed during almost all of March, show that a plume including I-131 arrived after March 15. The results at other monitoring posts near Fukushima Daiichi Nuclear Power Station are used to characterize plume diffusion at the early period of the accident before March 15. The I-131 time-integrated concentrations in air at several monitoring posts were compared with those given in UNSCEAR 2013 ANNEX A, which were obtained using estimated time-dependent rates of release to the atmosphere. The agreement between the two results varies depending on the places compared, owing to the large uncertainties in the estimated release rate used in UNSCEAR. The results obtained in this study can be used to increase the accuracy of the time-dependent release rate estimation.
As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained.
Recently, digital instrumentation and control systems have been increasingly installed to the important safety features in nuclear power plants such as the reactor trip system and the actuation system of the engineered safety features. On the other hand, it has been found that it is difficult to model the digital equipment reliability in probabilistic risk assessment (PRA) by the conventional Fault Tree Analysis technique. The OECD/NEA CSNI Working Group of Risk Assessment (WGRisk) set up the task group DIGREL to discuss several issues including quantitative dynamic modeling. This paper shows that, taking account of the relationship among the reactor trip system failures, demand after the initiating event, detection of the reactor trip system fault by self-diagnostic or surveillance tests, repair of the failed reactor trip system components and plant shutdown operation by the plant operators as a stochastic process, the anticipated transient without scram (ATWS) event can be modeled by the event logic fault tree and Markov state-transition diagrams in the case of the 2-out-of-4 digital reactor trip system.
An accident of evaporation to dryness by boiling of high-level liquid waste (HLLW) is postulated to be one of the severe accidents that may occur as a result of the loss of cooling function at a fuel reprocessing plant. In this case, some amount of nonvolatile fission products (FPs) will be transferred in the form of mist to the vapor phase in the tank, and could be released to the environment. Therefore, the quantitative estimation of the transfer rate is one of the key issues in the assessment of the accident consequences. To resolve this issue, a mechanistic correlation of the entrainment rate with upward vapor flow has been developed based on the data obtained from the experiments using simulated and actual HLLW.
Boron carbide (B4C) used for BWR or EPR absorbers could cause phenomena that never occur in PWR with silver-indium-cadmium absorbers during a severe accident. B4C would undergo a eutectic interaction with stainless steel and enhance core melt relocation. Boron oxidation could increase H2 generation, and the change of liberated carbon to CH4 could enhance the generation of organic iodide (CH3I). HBO2 generated during B4C oxidation could be changed to cesium borate (CsBO2) by combining it with cesium. This may increase cesium deposition into the reactor coolant system. There could be differences in the configuration, surface area, and stainless-steel to B4C weight ratio between the B4C powder absorber and pellet absorber. The present task is to clarify the effect of these differences on melt progression, B4C oxidation, and the iodine or cesium source term. Advancement of this research field could contribute to further sophistication of prediction tools for melt progression and source terms of the Fukushima accident, and the treatment of organic iodide formation in safety evaluation.
In a volume reduction process for the decontamination of contaminated soil, the performance degradation of a filter press is expected owing to material deterioration under high-dose irradiation. Eleven-stock selection of candidate materials including polymers, fibers and rubbers for the filter press was conducted to achieve a high performance of volume reduction of contaminated soil and the following results were derived. Crude rubber and nylon were selected as prime candidates for packing, diaphragm and filter plate materials. Polyethylene was also selected as a prime candidate for the filter cloth material.
After the accident at TEPCO’s Fukushima Daiichi Nuclear Power Stations on March 11, 2011, severe accident countermeasures and regulations have been discussed in various organizations as well as the Secretariat of the Nuclear Regulation Authority (S/NRA) in Japan. For severe accident management, spray systems or alternative spray systems have become increasingly important for reducing radionuclide release from the containment. In the present study, a simplified model was developed for aerosol removal by the spray system, considering a nonsprayed region in the containment. The effect of the nonsprayed region was estimated by the simplified model with a single volume, although multivolumes were used in the past. The model was verified through comparison with the analytical solutions for typical containment spray conditions.