We have worked on the development of a rapid detection method for radioactive cesium in the air and confirmed its basic performance by Monte Carlo simulations and experiments conducted in the “difficult-to-return zone” in Fukushima. In the stage of promoting its practical application, we considered that it is important to grasp the influence of the natural variation of radon and its daughter nuclides because the natural variation and influence are often discussed as problems in the field of environment radiation monitoring. Therefore, we carried out field verification tests to confirm the influence of the natural variation of radon in the difficult-to-return zone in Fukushima. As a result of continuous tests during three weeks in the zone, we confirmed that the variation of natural nuclides can be distinguished from the whole variation and that the variation of natural nuclides does not have an impact on the performance. Finally, we concluded that this method is a practical technology that is necessary in the difficult-to-return zone.
This study aims to improve the potential of an emergency response by analyzing the workload management during the accident at the Emergency Response Center (ERC) of TEPCO’s Fukushima Daiichi Nuclear Power Plant. Specifically, the research focused on the response of the ERC during the time between the discontinuation of Unit 3 core water injection and its recovery. It identified the different types of workload at the ERC had and how they had been managed based on the record of a TV conference. It also deduced the casual factors of the responses, supplementing the interview record of the director of ERC at the time by applying workload management analysis. On the basis of these findings, lessons to enhance the potential of the on-site emergency response have been obtained for ERC and outside organizations.
An accident of evaporation to dryness by boiling of high-level liquid waste (HLLW) is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, volatile radioactive materials, such as ruthenium (Ru), are released from the tanks with mixed vapor of water and nitric-acid into the atmosphere. In addition, nitrogen oxides are also released, formed by the thermal decomposition of metal nitrates of fission products (FP) in HLLW. It has been observed experimentally that nitrogen oxide strongly affects the transport behavior of Ru under the anticipated atmospheric conditions in cells and/or compartments of the facility building. Chemical reactions of nitrogen oxide with water and nitric acid are also recognized as complex phenomena that undergo simultaneously in the vapor and liquid phases. An analysis method has been developed by coupling two types of computer codes to simulate not only thermohydraulic behavior but also chemical reactions in the flow paths of carrier gases for quantitative estimation of the amount of Ru released to the environment. A simulation study has also been carried out with a typical facility building to demonstrate the feasibility of the developed simulation method.
MCNP whole-core reference solutions with pin-by-pin resolution were generated to numerically complement the operating and/or experimentally measured data of an Advanced BWR MOX core loaded with UO2 to 100% MOX fuels for the verification and validation (V&V) of BWR core design codes. To this end, (1) the degree of preciseness in geometrical and material modeling under the allowable storage limitations of MCNP was explored, (2) the correctness of predicted neutronics phenomena inside BWR cores by MCNP physics was investigated and the necessary enhancement was made for MCNP, (3) the smallness of uncertainties of results in MCNP due to its stochastic treatment and the nuclear data employed was ensured, (4) the representativeness of BWR core characteristics obtained from MCNP reference solutions was qualified. According to the results, the MCNP-generated reference solutions are applicable to validating the neutronics calculation performance of BWR core design codes as an alternative to measured data. In addition, extraction of the detailed neutronics quantities such as fuel-pin-wise flux enables us to verify the physics modules of the design codes such as the neutron flux solver. The MCNP whole-core reference solutions proved to be applicable in performing MCNP-based BWR design code V&V.