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1818-331X (Print), 2074-0565 (Online)
Thermal-hydraulic calculations of the WWR-SM research reactor
S. A. Baytelesov, F. R. Kungurov*, B. S. Yuldashev
Institute of Nuclear Physics, Academy of Sciences of the Republic of Uzbekistan, Tashkent, Uzbekistan
*Corresponding author. E-mail address: firstname.lastname@example.org
Abstract: The paper presents calculations of the thermal power distribution in the reactor core (RC) of the WWR-SM research nuclear reactor of the Institute of Nuclear Physics of the Academy of Sciences of the Republic of Uzbekistan, settlement Ulugbek, Tashkent, both for all fuel assemblies loaded into the core and for each fuel element of a separate fuel assembly. These calculations were carried out for RC configurations with a different number of fuel assemblies – 18, 20, and 24. The power distribution reactor core was performed using the IRT-2D code. A detailed simulation of the power distribution in the fuel element was performed using the MCNP4C code, while the fuel elements were modeled as square pipes with straight angles without rounding. The power distribution was calculated for each side of each fuel tube and divided into 15 axial nodes. The results of modeling of the thermal-hydraulic state were obtained using the PLTEMP code for various RC configurations. In the calculations, it was assumed, that the inlet water temperature is 45 and 48 °C for all RC configurations, the heat hydraulic parameters were taken from the calculation of the flow rate of the first circuit through the core 1250 m3/h. An analysis of the thermal power distribution of nuclear fuel in the reactor core of the WWR-SM research reactor showed that even with a conservative approach, permissible operating modes are not exceeded. During the operation of the three main circulation pumps, which provide the coolant flow through the core at the level of 1250 m3/h, the heat exchange crisis does not occur in the most energy-stressed fuel assemblies, namely, the temperature of the fuel rod clad and coolant remains below the permissible limits.
Keywords: reactor core, fuel assembly, fuel element, heat flux, thermal power.References:
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