Development of Corrosion Product Behaviour and Radiation Fields at the Sizewell B PWR from 1995 to 2008

摘要

Sizewell B is a standard Westinghouse 4-loop 3411 MWth PWR with low cobalt Alloy 690 tubed steam generators (SGs). A very limited amount of Stellite hard facing alloy was eliminated (fourteen CVCS flow control valves), but the Stellite in the reactor pressure vessel and reactor coolant pumps (RCPs) was not replaced. It was commissioned in 1994/1995 using a modified Hot Functional Test (HFT) passivation procedure with both boric acid and hydrogen present, which was aimed at growing an initial oxide film on the RCS surfaces under correct redox conditions [1]. Care was taken to prevent any valve lapping debris entering the primary circuit (RCS) during construction and all circuits were flushed extensively [2]. No SG channel heads were electropolished.