Results of safety analysis concerning level gauge of pressurizer and related measures for Ohi Nuclear Power Station

The results of safety analysis concerning the level gauge of a pressurizer and the related measures for the Ohi Nuclear Power Station, which were reported by the Ministry of International Trade and Industry on May 1, 1979, and then investigated and presented by the Nuclear Reactor Safety Special Investigation Committee, are regarded as appropriate by the Atomic Energy Safety Commission (AESC) on May 19, 1979. This analysis and investigation were conducted in relation to the accident in the Three Mile Island (TMI) No. 2 nuclear power plant, occurred on March 28, 1979. In this investigation the influences of the problem concerning the level gauge for a pressurizer of the Ohi Nuclear Power Station on the function and the performance of emergency core cooling system (ECCS) for pressurized water reactors were analyzed, simulating the initial phenomena of the accident in TMI. As for the adequacy of the prior conditions of safety analysis and the analytical codes, the following three points were investigated; 1) the adequacy of accident conditions which were set as the subjects of safety analysis, 2) the adequacy of the assumption supposing that the auxiliary feed water pump is started by manual operation fifteen minutes after accident occurrence, and 3) the adequacy of the analytical Codes MARVEL and SATAN-6. The analytical results showed that the relief valve on a pressurizer does not operate at the similar accident to TMI in the Ohi Nuclear Power Station, because the difference of design exists between both plants, and also even if the relief valve on a pressurizer operates and does not close, the core is cooled sufficiently in the Ohi Nuclear Power Station. (Nakai, Y.)