Calculations of research reactor thermal hydraulics based on VVER-440 fuel assamblies

Having thoroughly analyzed the design features of VVER-type pressurized water reactors and VVR-type research reactors, the authors propose a design of a research reactor with low-enriched fuel based on deeply updated VVER-440 fuel assemblies. The research reactor is intended to solve a wide range of applied problems in nuclear physics, radiation chemistry, materials science, biology, and medicine. The calculated thermal hydraulics confirms the correctness of the fundamental approaches laid down in the reactor design. An equivalent reactor core model in the form of a thick-walled cylinder was considered, and the radial power density distribution was obtained. According to the heat power level, five groups of FAs were identified. For each group, the coolant mass flow rate was calculated, which ensures alignment with the outlet coolant temperature. The coolant flow regime was also estimated. It turned out that for the first row of FAs, the flow regime is in the transition region, while for the other rows the flow regime is laminar. A test by the Gr.Pr≥1.105 criterion showed its conformity (the calculated value was 1.96.106), indicating the transition to a viscous-gravitational regime. The FE surface overheating was calculated relative to the mixed coolant average temperature. The axial coolant flow temperature distribution is the same in all the FAs, the change in power is compensated by the corresponding change in the coolant flow. The maximum coolant overheating on the FE wall relative to the flow core is observed in the central FAs, reaching 31 °C, the boiling margin is about 15 °C. The estimates showed a significant dynamic pressure margin during natural thermal-convective circulation. By calculation, the values of the FE surface overheating during the reactor normal operation were obtained. An approximately 15-degree surface overheating margin relative to the saturation curve is shown, which guarantees the absence of cavitation wear of the FE claddings. In general, the performed calculations confirmed the correctness of the approaches laid down in the reactor design and made it possible to specify the core thermal hydraulics necessary for further developing the concept