In this study, we aim to approximately evaluate the effect of nonlinearity of reinforced concrete structures through seismic response analysis using the equivalent linear analysis method. A simulation analysis was performed for the ultimate response test of the shear wall of the reactor building used in an international competition by OECD/NEA in 1996. The equivalent stiffness and damping of the shear wall were obtained from the trilinear skeleton curves proposed by the Japan Electric Association and the hysteresis curves proposed by Cheng et al. The dominant frequency, maximum acceleration response, maximum displacement response, inertia force-displacement relationship, and acceleration response spectra of the top slab could be simulated well up to a shear strain of approximately γ＝2.0×10^－3. The equivalent linear analysis used herein underestimates the maximum displacement response at the time of ultimate fracture of approximately γ＝4.0×10^－3. Moreover, the maximum shear strain of the shear wall could not capture the locally occurring shear strain compared with that of the nonlinear analysis. Therefore, when employing this method to evaluate the maximum shear strain and test results, including those during the sudden increase in displacement immediately before the fracture, sufficient attention must be paid to its applicability.

In this study, we perform model analyses assuming the Japanese power supply portfolio in 2050 to evaluate the economic efficiency of nuclear power generation under mass introduction of variable renewable energy (VRE) ― such as solar PV and wind ― and of hydrogen power generation in 2050. As a result, this study shows that even if the unit cost of VRE falls significantly by 2050, not only existing nuclear power plants, but also new construction, will have economic efficiency. Its benefit would become much larger when 100% carbon-free generation is mandated, but in that case, the role of nuclear energy as the base load power generation would be changed. On the other hand, in the case where hydrogen power generation will be deployed, the nuclear energy would be smaller than those in other 100% carbon-free cases, but its base load operation would be maintained.

Burnup characteristics of a fast-spectrum chloride molten salt reactor (Cl-MSR) are studied to utilize the trans-uranium element (TRU) and depleted uranium effectively. The fuel salt used is a mixed salt of NaCl, CaCl₂, UCl₃ and TRUCl₃. This fuel salt is expected to have a solidification temperature of about 530℃ and a heavy metal solubility of about 30 mol％. When the uranium inventory is 66.0 tons and the reprocessing rate is 18.3 liter/day, the fuel conversion ratio of Cl-MSR becomes about 1.0 after 16 years of reactor operation supplying both ²³⁸U and TRU. After that, Cl-MSR can continue to operate on ²³⁸U supply only and can confine TRU in its primary loop for a long period of time without increasing or decreasing its amount. By using TRU and depleted uranium currently existing in Japan, 14 Cl-MSRs can continue to generate 14 GW_e electricity for more than 1,000 years. These burnup characteristics of Cl-MSR are analyzed by SRAC2006 with SRACLIB-JDL40.

We estimate the monthly discharge inventory of tritium from the port of Fukushima Daiichi Nuclear Power Plant (1F) from Jun. 2013 to Mar. 2020 using the Voronoi tessellation scheme, following the tritium monitoring inside the port that started in Jun. 2013. As for the missing period from the initial month, Apr. 2011 to May 2013, we calculate the tritium discharge by utilizing the ratio of tritium concentration to ¹³⁷Cs concentration in stagnant contaminant water during the initial direct run-off period to Jun. 2011 and the discharge inventory correlation between tritium and ¹³⁷Cs for the next-unknown continuous-discharge period up to May 2013. From all the estimated results over 9 years, we found that the monthly discharge inventory sharply dropped immediately after closing the seaside impermeable wall in Oct. 2015 and subsequently coincided well with the sum of those of drainage and subdrain etc. By comparing the estimated results with those in the normal operation period before the accident, we point out that the discharge inventory from the 1F port after the accident is not very large. Even the estimation for the year 2011 is found to be comparable to the maximum of operating pressurized water reactors releasing relatively large inventories in the number of digits. In the national level, the total domestic release inventory in Japan significantly decreased after the accident owing to the operational shutdown of most plants. Furthermore, 1F and even the total Japanese discharge inventory are found to be minor compared with those of nuclear reprocessing plants and heavy-water reactors on a worldwide level. From the above, we suggest that various scenarios can be openly discussed regarding the management of tritium stored inside 1F with the help of the present estimated data and its comparison with the past discharge inventory.

The air-fed suit is a type of personal protective equipment that provides purified air through a hose and protects a worker from radiation contamination and internal exposure. In the Nuclear Fuel Cycle Engineering Laboratories of the Japan Atomic Energy Agency, the suit is used to protect a worker from radiation hazards during the size reduction and dismantlement of radioactive gloveboxes and equipment, which are contaminated particularly with plutonium. Although the suit has been widely adopted in similar activities, there still exist potential risks due to the limiting features of the suit itself and its supplemental system. In fact, we faced several unexpected events regarding such restricted aspects during dismantling activities. To address these failure potentials, we have implemented various countermeasures and improvements to enhance worker safety. Relevant examples include the exchange of the hose to one that is free from flattening while bending, the deployment of an air compressor to feed chilled air, and the installation of a monitoring system of the supplied air. We describe the disadvantages of the air-fed suit system and positive feedback regarding our implementations.

The author independently conducted a detailed analysis of plant conditions, based on a new meltdown process, during the 14 hours after all power loss at the Fukushima Daiichi Nuclear Power Plant Unit 1. After 3:30 PM on March 11, 2011, all power loss caused reactor core cooling function loss and the vaporization of water in the reactor vessel (RPV), and the RPV was filled with high-temperature hydrogen gas generated by decay heat and Zr-H₂O reactions. The reactor core melted and the peripheral stainless-steel structures and nuclear vessel walls were heated by the radiation energy of decay heat. Then, the shrouds and other components near the reactor core started melting and the wall temperature of the RPV was raised. Along with the wall temperature rising to nearly 650℃, the aluminum insulation near the wall started melting (aluminum melting point, about 650℃). The heat from the collapsed insulation caused the superheated state of the containment pressure and temperature of 0.84 MPa and 400℃, respectively, around 3 AM from the saturated state of 0.6 MPa around 1 AM on March 12. After that, the containment vessel was depressurized gradually and kept under stable cooling 14 hours after all power loss, for the time from 4 to 6 AM. By the first water injection to the reactor core at 4 AM, radioactivity release increased slightly and the containment pressure was stable. However, the situation changed significantly after continuous water injection to the reactor core started at around 6 AM. The water injection into the high-temperature reactor core generated film boiling (poor heat transfer) and good reactor core cooling was no longer available; thus, increased heating by the Zr-H₂O reactions started. The results of the author’s analysis showed good consistency with the measured reactor pressure, containment pressure, containment temperature and radioactivity near the main gate. Although decay heat cannot be decreased intentionally, the occurrence of the Zr-H₂O reactions can be inhibited. It is proposed that the best way to mitigate the effects of a meltdown is to stop water injection to the core after detecting the initiation of film boiling or inferring that the fuel rods are not covered with water.