After the Fukushima daiichi nuclear power plant accident, various countermeasures were taken for Beyond Design Basis Events (BDBE) in the system safety field. These included portable devices, additional backup facilities and accident management. They are different from approaches for Design Basis Events (DBE). In the field of structural mechanics; however, efforts were focused on strengthening to prevent failures for both DBE and BDBE in the same way. This approach will lead to limitless requirements for strength and expensive plants.As a breakthrough approach in structural mechanics for BDBE, we propose failure mitigation methods through the application of passive safety structures, where preceding failures release loadings and mitigate subsequent failures. When preceding failure modes have small impacts on safety performance, such as small deformation and crack initiation, and subsequent ones are catastrophic modes such as collapse and break, the passive safety structure improves safety and resilience. This idea is the utilization of passive characteristics of structures without additional equipment and electric power, allowing for simple and reliable plants.To demonstrate this idea, passive safety structures were applied to next-generation fast reactors, subject to high temperature and low-pressure conditions. In the case of loss-of-heat-removal accidents, high temperature conditions accelerate the creep deformation of structures. When deformation redistributes loadings and reduces stresses at important positions such as coolant boundaries, progression to creep rupture of boundaries can be mitigated. When an excessive earthquake occurs, plastic deformation and buckling become dominant, due to low pressure and, therefore, a thin-wall structure. The above-mentioned failure modes reduce rigidity and natural frequency. When the natural frequency becomes lower than the input frequency, vibration energy is hardly transferred to structures and the subsequent failures of structures, such as collapse and break, are mitigated.
Understanding the dissolution behavior of the fuel debris is necessary for the safe decommissioning of Fukushima Daiichi Nuclear Power Plants. The dissolution behavior of FeUO4 compounds formed by a high-temperature reaction of UO2 with iron, a stainless-steel component of reactor structural materials, was investigated under atmospheric conditions. The compounds were prepared in an electric furnace using U3O8 and Fe3O4 as starting materials, and their solid states were analyzed using X-ray diffraction, scanning electron microscopy-energy-dispersive X-ray spectroscopy, and X-ray absorption fine-structure spectroscopy. The fission products were produced via thermal-neutron irradiation. The concentration of nuclides dissolved in water was examined by performing static leaching tests of FeUO4 compounds for up to three months. A redox reaction was proposed to occur between trivalent Fe and pentavalent U ions in the early stage of FeUO4 dissolution. It was thermody-namically deduced that the reduced divalent Fe ion was finally oxidized into a trivalent ion in the presence of dissolved oxygen, and iron hydroxide limited the solubility of Fe. Meanwhile, the concentration of hexavalent U (i.e., uranyl ion) was limited owing to the presence of secondary minerals such as metaschoepite and sodium uranate and subsequently decreased, possibly owing to sorption on Fe oxides, for example. The concentrations of multivalent ions of fission products, such as Ru and Ce, also decreased, likely for the reason above. By contrast, the concentration of soluble Cs ions did not decrease. The validity of this interpretation was supported by comparing the results with the dissolution behavior of a reference sample (Fe-free U3O8).
The Japan Atomic Energy Agency (JAEA) is developing an evaluation method for a two-phase flow in the reactor core using simulation codes based on the volume of fluid (VOF) method. However, it is impossible to simulate boiling on the heating surface in large-scale domains, such as fuel assemblies, using this simulation method because simulating boiling based on the VOF method requires fine meshes to resolve the initiation of boiling. Therefore, the JAEA started developing a simplified boiling model (SBM) for the two-phase flow simulation in fuel assemblies. In the SBM, the motion and growth equations of a bubble are solved to obtain their diameter and time length at the detachment, the size scale of which is within/around the calculation grid size of the numerical simulation. This information is given as the vapor injection flow rate for the boundary condition on the heating surface. JUPITER calculates the bubble behavior with a scale of more than several millimeters. Using the developed SBM, this study simulates convection boiling on a vertical heating surface. A comparison between the simulation and experimental results showed good reproducibility of the heat flux and velocity dependency on the passage period of the bubble.
Sample reactivity measurements of CaH2 samples as the integral experiments were conducted in the Kinki University Reactor (UTR-KINKI). The sample reactivity worth was determined from a difference between two excess reactivities of the respective reactors with and without the sample. The reactivity worth measurements were repeated for 21.20 g, 41.00 g, 72.78 g and 164.4 g of CaH2 sample. Furthermore, the sample reactivity worth was calculated using the continuous energy Monte Carlo codes MVP3.0 with the nuclear library JEFF-3.1, where each cross section of H and Ca constituting a molecule of CaH2 was taken into the thermal neutron scattering law. The ratios of calculated to experimental values (C/E) were 1.119±0.587, 0.806±0.305, 0.874±0.165 and 0.894±0.099 for 21.29 g, 41.00 g, 78.72 g and 164.4 g of the samples, respectively.
