Abstract
A pressure control system failure test series was conducted at the Rig of Safety Assessment (ROSA)-III test facility to evaluate the effect of the pressure control system on thermal-hydraulic phenomena during a small break loss-of-coolant accident (LOCA) of a boiling water reactor (BWR). The break was assumed at the recirculation pump suction line. The pressure control system had no effect for breaks greater than 5%. For breaks less than 5%, if the pressure control system was inactive, the core was uncovered temporarily because of the bubble collapse due to pressure rise by the closure of the main steam isolation valve (MSIV), and the fuel rod surface temperature rose high during this period. However, the peak cladding temperature (PCT), which occurred mainly during the later core uncovering by boil-off, was lower in a LOCA with pressure control system failure than in a corresponding LOCA with intact pressure control system. This is because the emergency core cooling systems (ECCSs) were actuated earlier in a LOCA with pressure control system failure due to lower system pressure. The PCT was well below the present safety criteria of 1,473K even if the pressure control system and high pressure core spray system (HPCS) (the severest single failure in ECCS) were assumed to be inactive.
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