Pre-Phase 1 Aging Assessment of the BWR Isolation Condenser System

Abstract

The isolation condenser system (ICS) is part of the emergency core cooling system in five U.S. boiling-water reactors. In the event that the reactor pressure vessel becomes isolated from the main condenser, the ICS removes decay heat from the reactor. The ICS is important to reactor safety because it is relied on to help mitigate core damage during a loss-of-coolant accident. In support of the U.S. Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, staff from the Pacific Northwest Laboratory researched the aging of the ICS by reviewing available industry databases. Each component of the ICS was evaluated to 1) identify applicable aging issues and also to 2) determine if the component had already been studied as a part of other NPAR assessments. The results of this preliminary study indicate that most of the critical ICS components have been previously evaluated by the NPAR program. The one ICS component that has not been specifically studied is the isolation condenser itself. There is little evidence in the databases to suggest that there have been problems with the isolation condenser. Only one plant, Millstone Unit 1, has ever had an isolation condenser tube failure problem recorded. This instance resulted from events that occurred early in the life of the plant. The problem was remedied through tube replacement. The isolation condenser and the pressurized-water reactor (PWR) steam generator were compared to illustrate that even though the isolation condenser is a heat exchanger, it is not subjected to the same service dynamics as the PWR steam generator. The isolation condenser operates for most of its service life in a relatively benign, static environment, resulting in a comparatively good service record. PNL staff recommend that the results of this research be used to continue studying the ICS to determine if the aging isolation condenser tubes are being adequately maintained. This new study should include an evaluation of the current inspection methods and a verification that they are effective in identifying tube aging degradation, such as intergranular stress corrosion cracking. Continued study may also provide beneficial input into the design of the Simplified Boiling Water Reactor.