Published: 01 January 1993
| ||Format||Pages||Price|| |
|PDF (244K)||12||$25||  ADD TO CART|
|Complete Source PDF (7.9M)||399||$103||  ADD TO CART|
Cite this document
The majority of pressurized water reactors (PWR) operating in the United States have a “design basis life” of 30 to 40 years. These design basis life estimations were not based on technical studies of material degradation in general but rather on fatigue usage factors for the most part. Recognizing this fact, the subject of operating an existing nuclear steam supply system (NSSS) for longer periods than originally intended has become an important issue worldwide. Radiation embrittlement of the reactor vessel beltline region (the area surrounding the height of the reactor core) is the main concern in extending the operation of the NSSS. Radiation damage, if any, to the reactor pressure vessel material is monitored by a material radiation surveillance program. If the data from the reactor vessel materials surveillance program indicate that the reactor pressure vessel will not meet the rules of the Code of Federal Register (CFR) and various regulatory guides (RG), there are a number of options a utility may take to ensure reactor pressure vessel design life attainment or extension. This paper describes the results from 12 surveillance capsule programs, which encompass four reactor vessel materials and five reactor vessel manufacturers.
reactor pressure vessels, irradiation, embrittlement, surveillance capsules, design basis, aging management, life attainment
Consulting engineer, Westinghouse Electric Corp., Pittsburgh, PA