| Summary: | Advanced core materials and components for nuclear reactors of a new generation are tested in the BOR-60 reactor. A novel irradiation rig (IR) with a fuel heater is used for high-temperature tests of material samples. The irradiation rig features a number of advantages over ampoule-type irradiation rigs commonly used nowadays. Computational and experimental studies on an IR with a fuel heater have been conducted in the BOR-60 reactor core. The results of a dedicated methodical experiment have proved that it is possible to provide the required temperature conditions for irradiation of tested samples. MCU-RR, a precision code, was used for neutronic calculations, and thermohydraulic calculations were performed using the ANSYS CFX software system. A comparison of calculated temperature values against experimental data has shown a fit in the experimental error limits which confirms the applicability of the selected codes, models and procedures. Computational and experimental studies have also been conducted for the temperature distribution in the IR with a fuel heater following the withdrawal of the IR from the reactor and its placement in a dry cooling channel. The decay power in the IR fuel pins were calculated using the AFPA code and the temperature fields were calculated based on ANSYS CFX. It has been shown that the permissible temperature value on the fuel cladding is not exceeded in the IR withdrawn from the reactor following two-day cooling after the reactor shutdown.
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