Experiments and validation analyses of HTTR on loss of forced cooling under 30% reactor power

In a safety demonstration test involving the loss of both reactor reactivity control and core cooling, the high-temperature engineering test reactor (HTTR) demonstrates spontaneous stabilization of the reactor power. The test and analytical results of tripping one or two out of three gas circulators without reactor scram have already been reported. Moreover, the pre-analytical result of tripping all three gas circulators without reactor scram has been presented. On the other hand, the test and analytical results of tripping all three gas circulators without reactor scram are shown in this paper. About experiments, at an initial reactor power of 30% (9 MW), when all three gas circulators were tripped without reactor scram to reduce the coolant flow rate to zero, the fuel temperature did not show a large increase because the large heat capacity of the graphite core could absorb heat from the fuel in a short period. Moreover, the decay heat could be transferred through the graphite core and the reactor pressure vessel (RPV), emitted by thermal radiation from its outer surface and removed to the active vessel cooling system; therefore, the core at 9 MW was never exposed to the danger of a core melt, and the reactor power was stabilized spontaneously. About analyses, the reactivity performance is important for predicting the converging level of reactor power that affects the fuel temperature during a loss of forced cooling (LOFC) without reactor scram. With regard to thermal hydraulics, the performances of graphite heat conduction in the reactor core and thermal radiation from the RPV surface to the reactor cavity cooling system are crucial for predicting the temperature behavior of the fuel and RPV in the LOFC condition. It was confirmed that reactor kinetics coupled with heat transfer could be applied to reactor safety and accident analysis based on the comparison between the experiments and the analyses.