The Japan Atomic Energy Agency has been developing "Advanced Reactor Knowledge-and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA)" to offer the best solutions for challenges in the design and operation of nuclear plants. A part of ARKADIA for design study, which included design optimization of components, is named as ARKADIA-Design. In the development of ARKADIA-Design, we have been developing a process to automatically optimize design parameters of structural components subjected to various kinds of loads, including thermal transients. In this paper, we propose a simplified procedure to estimate the failure probability of components subjected to thermal transients for design optimization. An objective function of this optimization is defined on the basis of failure probability of the components, because failure probability can be commonly used as an indicator of component integrity for various mechanisms, and it helps future introductions of a risk-informed performance-based approach to component design. To enable the necessary number of estimations for design optimization with practical calculation time, we aimed to reduce the number of analyses required for one estimation. For this purpose, we adopted the first-order second-moment (FOSM) method as the estimation method for failure probability in the process of optimization. An orthogonal table in the experiment design method is utilized to define the conditions of the analyses for evaluation of the mean and variance of thermal transient stress, which are used as inputs in the FOSM method. The superposition of ramp responses is also utilized to evaluate the time history of thermal transient stress instead of finite element analysis. The proposed procedure was applied in a demonstration study to optimize the thickness of a cylindrical vessel subjected to thermal transients derived from shutdown. We confirmed that the procedure can evaluate the failure probability depending on the cylinder thickness with practical calculation time.

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Since the accident of the Fukushima Dai-ichi Nuclear Power Station, analysis evaluations for stress tests have been conducted on the prototype fast breed reactor MONJU. In these evaluations, it has been necessary to evaluate the plant characteristics including the Ex-Vessel fuel Storage Tank (EVST) under the severe accident conditions such as station blackout. The EVST is a coaxial cylindrical tank to store spent fuels (SFs) stored in fuel transfer pots (FTPs) until the radioactivity enough decays. It is able to store 252 fuel transfer pots in the rotating rack and cool them by internal natural circulation of sodium coolant under both the severe accident and normal operation. The heat produced by SFs removed by the helical cooling coils installed in the annular space. Evaluations of natural circulation in the EVST have been performed with a one-dimensional flow-network code. However, it would be difficult to predict its behavior exactly, because it would include multidimensional flow such as local natural convection. Then, in order to clarify the natural circulation behavior and multidimensional effects and evaluate appropriateness of this flow network model, we have performed a thermal-hydraulic analysis using a three-dimensional model which has high resolution meshes and the almost same geometry as the actual equipment. This model makes it possible to take into account the following multidimensional phenomena, the heat distribution of the FTPs, the mixing in plenums, the bypass flow through the flow holes and the other geometry effects. In this study, we have used a commercial computational thermal-hydraulics code, “FrontFlow/red”. As a result of steady analyses, we have confirmed the following: The coolant temperature in the plenums is almost uniform and its difference is in a few degrees. The influence of the flow holes is also limited because its flow rate is relatively low to main flow rate. On the other hand, pressure loss at supporting plates of the rotating rack, which are main causes of flow resistance in the EVST, is larger than the case without multidimensional effects because of the natural convection concentrated in the high temperature region near heated FTPs. The result leads to our presumption that the flow network model of the EVST is almost appropriate. It should be noted that flow resistance coefficient of the supporting plates or the heat transfer center of the cooling coils should be set to conservative for the safety analysis on the EVST.

This paper reports a part of realistic safety margin analysis of Japanese prototype fast breeder reactor MONJU, based on actually measured plant characteristics during a commissioning operation under up to 40% electric power generation mode. In order to evaluate realistic integrity margins against design-phase evaluations and against the limit values, the thermal stresses of critical component parts after a typical thermal transient event, i.e., a manual reactor trip, were derived through serially combined thermal transient and structural analyses. The results of thermal transient and structure analysis confirmed that the evaluation values of the structural stresses at the reactor vessel outlet nozzle, at the upper tube sheet of the intermediate heat exchanger, and at the evaporator feed water inlet tube sheet are well within the design evaluation values.

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