All issues

The paper describes an analytical study of hydrodynamic processes taking place in the course of coolant flow through a fuel assembly of the core of a fast neutron sodium-cooled reactor. Within the framework of the study, a procedure and an analytical model were developed based on program complex FlowVision of computational fluid dynamics, which, using proved simplifications, permits to obtain a coefficient of rod lifting margin of a fuel assembly and to study hydrodynamic characteristics of processes taking place in the course of simulation of different initial events influencing motion of a reactor core fuel assembly.

For analytical justification a fuel assembly model was developed, which is equivalent by hydraulic resistance values and permits not to simulate explicitly a complicated full-scale fuel assembly design, thus, decreasing a number of computational cells in the model and, as a result, reducing computational and time resources.

Hydraulic parameters of the equivalent fuel assembly model in program complex FlowVision were analyzed in two stages. At the first stage, to determine the minimum rod lifting margin coefficient of a fuel assembly, steady-state analyses were performed, where various flowrate values were assigned at the model inlet and forces acting upon the assembly were analyzed. A series of dynamic mode analyses was performed at the second stage. Jump-like pressure increase being the initial event which could occur hypothetically in the fast neutron sodium cooled reactor plant was assigned in these modes. Hydrodynamic parameters and forces acting upon the fuel assembly were determined.

The results of the first stage of the analytical study proved the minimum coefficient of rod lifting margin of a fuel assembly of the fast neutron reactor justified in reactor plant design documentation. As a result of the second stage of the study, conclusions were made on impossibility for the fuel assembly to move at the initial event associated with jump-like pressure increase in the reactor pressure chamber.

The development of a high-temperature gas-cooled reactor (HTGR) constituting a part of nuclear power-and-process station and intended for large-scale hydrogen production is now in progress in the Russian Federation. One of the key objectives in development of the high-temperature gas-cooled reactor is the computational justification of the accepted design.

The article gives the procedure for the computational analysis of thermal and physical characteristics of the high-temperature gas-cooled reactor. The procedure is based on the use of the state-of-the-art codes for personal computer (PC).

The objective of thermal and physical analysis of the reactor as a whole and of the core in particular was achieved in three stages. The idea of the first stage is to justify the neutron physical characteristics of the block-type core during burn-up with the use of the MCU-HTR code based on the Monte Carlo method. The second and the third stages are intended to study the coolant flow and the temperature condition of the reactor and the core in 3D with the required degree of detailing using the FlowVision and the ANSYS codes.

For the purpose of carrying out the analytical studies the computational models of the reactor flow path and the fuel assembly column were developed.

As per the results of the computational modeling the design of the support columns and the neutron physical characteristics of the fuel assembly were optimized. This results in the reduction of the total hydraulic resistance of the reactor and decrease of the maximum temperature of the fuel elements.

The dependency of the maximum fuel temperature on the value of the power peaking factors determined by the arrangement of the absorber rods and of the compacts of burnable absorber in the fuel assembly is demonstrated.