摘要：The TVS-X fuel rod model designed by NSC KIPT as an alternative fuel for subcritical assembly (SCA, KIPT, Kharkov) and research reactor (WWR-M, INR, Kiev) is described. The model is a program that allows calculating the tempera-ture distribution on the radius and height of the fuel element containing both uranium oxide pellets and dispersion fuel based on the UO2+Al composition with different contents of the fuel phase, as well as the different geometric characteristics of the fuel element and the values of the coolant parameters: the temperature at the entrance to the hydraulic channel and the coolant speed. Comparative calculations of tempera-ture distribution during operation are carried out. As a result, it has been shown that for conditions of operation in the SCA (linear power of fuel rod is 2.62 kW/m), the fuel center temperature reaches ~140 °C for UO2 and ~112 °C for the UO2+Al composition. For operating condi-tions in the WWR-M reactor (linear power of fuel rod is 12.1 kW/m), the fuel center temperature reaches ~626 °C for ceramic (UO2) and ~381 °C for metal-ceramic fuel (UO2+Al). The calculations show a significant effect of the type of fuel material (UO2 or UO2+Al dispersion composition) on the fuel center temperature, taking into account the operating conditions in the subcritical assembly and the WWR-M re-search reactor. The maximum temperature of the cladding for the WWR-M operating conditions was 86.5 °C, and the maximum temperature of the cladding for the SCA operating conditions is 27 °C, which does not exceed the boiling point (vaporization) under the nominal condi-tions of their operation. Cross-section area of fuel rods, heat transfer coefficient and temperature distribution of the coolant are calculated. The software module allowed to estimate the temperature distribution of fuel element with different types of nuclear fuel for the conditions of research nuclear assemblies.
其他摘要：Наведено опис моделі стрижневого твела паливної збірки ТВЗ-Х, спроектованої в НТК ЯПЦ ННЦ ХФТІ, як альтернативного палива для ядерної установки, заснованої на підкритичній збірці (ПКУ, ННЦ ХФТІ, м. Харків) і дослідницького реактора (ВВР-М, ІЯД, м Київ). М
关键词：fuel rods; model; thermal conductivity; uranium oxide; dispersion fuel; volume fraction of the fuel phase
