Calculation of thermal dynamic characteristics of the research reactor on the basis of fuel assembles for the VVER-440 reactor
As a result of the analysis of the design features of water-cooled power reactors of the VVER type and research reactors of the VVER type, the design of a low-enriched fuel research reactor based on deeply modernized fuel assemblies of the VVER-440 reactor is proposed. The research reactor is designed to solve a wide range of applied problems in the field of nuclear physics, radiation chemistry, materials science, biology and medicine. The results of calculations of thermal dynamic characteristics are presented, confirming the correctness of the fundamental approaches incorporated in the reactor design.
An equivalent model of the reactor core in the form of a thick-walled cylinder is considered, and the distribution of heat generation density to a radius is obtained. Five groups of fuel assemblies have been identified in terms of heat output. The coolant mass flow rate for each of the groups was obtained, ensuring alignment with the outlet temperature of the coolant.
An evaluation was made of the flow regime of the coolant. It turned out that, for the first row of fuel assemblies, the flow mode lies in the transition region; for the others, the flow mode is laminar. The test by the Gr·Pr і 1Ч105 criterion showed its fulfillment (the calculated value is 1.96Ч106), which indicates a transition to a viscous-gravity regime. The calculations were also made of the overheating of the surface of the fuel rods relative to the average mixed temperature of the coolant. The coolant flow temperature-height distribution in all fuel assemblies is the same; the change in power is compensated by a corresponding change in the coolant flow rate. The maximum overheating of the coolant on the wall of the fuel elements relative to the flow core is observed for central fuel assemblies, reaching a value of 31°C, the margin to the boiling point is about 15°C.
The assessments showed a significant margin of the driving pressure during the natural thermoconvective circulation. By calculation, the values of the overheating of the surface of the fuel rods during reactor operation in normal mode were obtained. Approximately a 15-degree surface overheating margin relative to the saturation curve is shown, which guarantees the absence of cavitation wear of the fuel element claddings.
In general, the performed design justification confirmed the correctness of the approaches incorporated in the reactor design and made it possible to specify the thermal and hydraulic characteristics of the core necessary for the further development of the concept.
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