Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power

Abstract

In a safety demonstration test involving the loss of both reactor-reactivity control and core cooling, the High-Temperature engineering Test Reactor (HTTR) demonstrated spontaneous stabilization of the reactor power. A test where all three gas circulators were tripped at 30% reactor power (9MW) without a reactor scram, called a loss-of-forced-cooling (LOFC) test, was analyzed, and the analytical results from a reactor kinetics code was reported. After that, the Japan Atomic Energy Agency (JAEA) acquired a large supercomputer and the STAR-CCM+ software developed by CD-adapco, which can be used to create a three-dimensional model of the HTTR to analyze steady-state and transient conditions. In this paper, a model of the three-dimensional thermal-hydraulics inside the reactor pressure vessel (RPV) during the LOFC test at 30% reactor power (9MW) is described and detailed analyses are performed using STAR-CCM+. With this tool, the effects of natural and forced convection on thermal-phenomena in the HTTR can be understood quantitatively. The finding is as follows: the downstream of forced convection caused by the helium purification system (HPS) pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection due to the HPS has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. And the effect can be neglected because the difference of the maximum velocity of the helium between the cases where the HPS is active and inactive, is very little over the course of LOFC test. The proposed three-dimensional analytical model will be useful for three-dimensional accident analyses.