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

Takamatsu, Kuniyoshi  

The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.