Light-water reactors (LWRs) are equipped with an emergency core cooling system (ECCS) that is designed to maintain the coolable geometry of the reactor core and finally minimize the release of radioactive fission products to the public and environment even in a loss-of-coolant accident (LOCA). Acceptance criteria for the ECCS of LWRs were determined to evaluate the safety function and performance in the design and to ensure a sufficient safety margin in the results of the evaluation. The latest revision of the criteria was made in 1981 in Japan, referring to the additional knowledge obtained after the previous revision. Fuel burnup has been extended by changing cladding materials, fuel design, etc., since the latest revision. Correspondingly, knowledge has been accumulated through studies on high-burnup fuel behavior under LOCA conditions to confirm the safety during the LOCA. This paper is a summary of the investigation and remaining issues on the applicability of the current Japanese ECCS acceptance criteria to high-burnup fuel, considering the history and basis of the current acceptance criteria. Results of the investigation conducted up to now reveal that the influence of burnup extension is small in terms of the cladding behavior of high-temperature oxidation and the fracture limit in quenching during the LOCA condition, and the current criteria are applicable even in the case of high-burnup fuel.
Various types of flame retardant cable are used in nuclear power plants. The cable insulation and sheath materials are generally categorized into two types: thermosetting (TS) and thermoplastic (TP) materials. Domestic cable fire damage criteria are set for each material type on the basis of the US guidelines, although they do not clearly define which type a TS/TP composite cable falls into. In this study, cable burning tests were conducted for TP and TS/TP composite cables used in domestic nuclear power plants to evaluate the validity of applying the US cable damage criteria to domestic cables. For varions cable damage scenarios, the tests measured the temperature/time dependence of insulation resistance between conductors and between cable and trays. The results of these tests justified the applicability of the US damage criteria. In addition, it was confirmed that the relationship between cable temperature and insulation resistance can be expressed by an Arrhenius plot. These findings are expected to be utilized for the Safety Improvement Evaluation periodically conducted by nuclear facility operators.
The mass discrimination effect in the isotope analyses of barium isotopes with natural abundance, from Ba-130 to Ba-138, was investigated by triple-quadrupole inductively coupled plasma-mass spectrometry (ICP-QQQ). The mass bias coefficients for Ba isotopes, denoted by ε(Ba), were determined from the slope of a linear relationship between the atomic mass differences and the ratios of ion count rates for a given isotope pair experimentally determined and calculated from natural abundances. The effect of the addition of a collision-reaction cell (CRC) gas such as helium, hydrogen, nitrous oxide, or carbon dioxide on ε(Ba) was examined. Large ε(Ba) values were observed in the case of He- or H2-CRC gas, and the values were from ＋0.8 to ＋1.7％ per atomic mass unit. On the other hand, ε(Ba) observed with CRC gasses containing N2O or CO2 was relatively small and below ＋0.3％ per atomic mass unit. In addition, the dependence of the energy discrimination potential (ED) applied between the CRC and the second quadrupole mass separator of ICP-QQQ on ε(Ba) was investigated. Finally, the analytical mass bias coefficient of the radioactive cesium nuclides 134Cs, 135Cs, and 137Cs, ε(Cs), was discussed for ε(Ba) in the same mass range as that of Ba isotopes.
The forward weighted consistent adjoint driven importance sampling (FW-CADIS) method has been proposed as a method for obtaining variance reduction parameters to estimate flux or dose rate distribution over a wide area, or responses at multiple localized detectors for a particle transport calculation based on the stochastic Monte Carlo method with a reasonable and high accuracy. The method has been applied to estimating the dose equivalent rate around the Japanese exclusive ship, “Seiei Maru”, which transports low-level radioactive waste. The particle transport calculation was performed using a mesh tally on the entire surface of the hatch cover above low-level radioactive waste packages stacked in the cargo hold and point detector tallies at each measurement point. The statistical error is spatially uniform for the mesh tally and is reduced only around the vicinity of each point detector tally. It is suggested that the estimated dose equivalent rate obtained by the calculation is equivalent to the measured result, and it is shown that this method is effective for the radiation safety evaluation of low-level radioactive waste transport ships.
The sodium–water reaction caused by failure of the steam generator tube of sodium-cooled fast reactors causes the wastage phenomenon, which is erosive and corrosive. Self-wastage takes place in the early stage of the sodium–water reaction event when a very small amount of water/steam penetrates a microcrack. When self-wastage proceeds to the inside wall of the tube, the failed area and water leakage rate will increase, whereby the area affected by the sodium–water reaction will be likely to extend. Thus, it is very important to clarify the self-wastage behavior for a locally affected region and detect water leakage in actual nuclear power plants. In this study, the authors performed self-wastage experiments under a high sodium temperature condition to evaluate the effects of the wastage form/geometry using two types of initial defect, i.e., the microfine pinhole and fatigue crack, and of the water leakage rate on the self-wastage rate. Taking into consideration the influence of crack type, we confirmed that the self-wastage rate did not strongly depend on the initial defect geometry. As a mechanism of the self-plug phenomenon, it is speculated that sodium oxide blocks and inhibits the progress of self-wastage. The dependence of the self-wastage rate on the initial sodium temperature was clearly observed, and a new self-wastage correlation was derived considering the initial sodium temperature.