Hydrodynamics of the Coolant in the Outlet Section of a Fuel Cartridge with Heads of Different Designs of the Reactor Core RITM of a Low-Power Nuclear Plant

Abstract

The results of experimental studies and a comparative analysis of the coolant hydrodynamics in the outlet section of the fuel cartridge behind heads of different designs are presented. The considered fuel assemblies are designed for installation in the core of a RITM-type reactor of a small ground-based nuclear power plant. The aim of the work was to study the distribution of the axial velocity and flow rate of the coolant at the outlet of the fuel bundle, behind the heads of different designs, and in front of the coolant extraction pipe and in the holes of the upper base plate as well as to determine the areas of the fuel bundle from which the coolant flow is most likely to enter the sampling pipe and, accordingly, to the resistance thermometer installed in this pipe. The experiments were carried out on a research aerodynamic stand with an air working medium on a model of the outlet section of the fuel cartridge, which includes a fragment of the outlet part of the fuel bundle with spacer grids, dummies of two types of heads, an upper support plate, and a coolant extraction pipe. When studying the coolant flow rate in the outlet part of the fuel cartridge, the pneumometric method and the method of contrast impurity injection were used. The measurements were carried out over the entire cross section of the model. The hydrodynamic picture of the coolant flow is represented by cartograms of the distribution of axial velocity, coolant flow rate, and contrast impurities in the cross section of the model. The results of the research were used by specialists from the design and calculation departments of Afrikantov OKBM to justify engineering solutions in the design of new cores of RITM reactors. The results of the experiments were compiled into a database and used in the validation of the LOGOS CFD program developed by the employees of RFNC-VNIIEF and ITMP Moscow State University as analogues to foreign programs of the same class, which include ANSYS, Star CCM+, etc. Experimental data are also used to validate one-dimensional thermal-hydraulic codes used in Afrikantov OKBM when substantiating the thermal reliability of the cores of reactor installations; the thermal-hydraulic code KANAL also belongs to this class of programs.