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In this study, we introduced new guidelines for enhancing the extrapolation capability of NN. Furthermore, the following tests are carried out at higher initial power stage therefore, it is needed to enhance the capability of extrapolation of the network model. After the learning, the network can predict the changes of reactor power and reactivity in the following tests. The actual tests data, which were previously carried out in the HTTR, were used for leaming the model of the plant dynamics. The inputs of the network are the changes of the central control rods position and other significant core parameters. This paper describes pre-test analysis method using a neural network (NN) to predict the changes of reactor power and reactivity. In the results, calculated void fraction distribution in horizontal plane agreed quantitatively with the measured one obtained from the advanced neutron radiography technique including the bridge formation of the liquid at the position of adjacent fuel rods where an interval is the narrowest.Д¬ontrol-rod withdrawal tests simulating reactivity insertion are carried out in the HTTR to verify the inherent safety features of HTGRs.

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And the modified TPFIT code was applied to large-scale water-vapor two-phase flow in tight lattice rod bundle. The vectorization and parallelization of TPFIT code was conducted to fit the large-scale simulations.

The present paper describes a part of the development of the two-phase flow simulation code TPFIT. As a part of this technology development, we are developing an advanced interface tracking method to improve the conservation of volume of fluid. Thus, we start to develop a predictable technology for thermal-hydraulic performance of the RMWR core using an advanced numerical simulation technology.

In the core, there is no sufficient information about the effects of the gap spacing and grid spacer configuration on the water-vapor two-phase flow characteristics. The reduced-moderation water reactor (RMWR) core adopts a hexagonal tight-lattice arrangement with about 1mm gap between adjacent fuel rods. The SIRIUS-F experimental results demonstrated stability characteristics such as a stabilizing effect of the power, and reviled a sufficiently large stability margin even under hypothetical conditions of power enlargement. The decay ratios and the resonance frequencies are in good agreement with those from the design analysis code, ODYSY. By utilizing this method, one can estimate stability at any specific operating point online without assuming excess conservative conditions.Д¬hannel and regional stability experiments were conducted for a wide range of operating conditions including maximum power points along the minimum pump speed line and the natural circulation line of the ABWR.

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A real-time simulation is performed for the modal-point kinetics of reactor neutronics and fuel-rod conduction on the basis of a measured void fraction in a reactor core section of the facility.ДЄ noise analysis method was performed to calculate decay ratios from dominant poles of transfer function on the basis of the AR method by applying time series of a core inlet flow rate. The SIRIUS-F facility was designed and constructed for highly accurate simulation of channel, core-wide and regional instabilities of an ABWR.