SYMPOSIA PAPER Published: 01 January 1993

Utilization of Reactor Pressure Vessel Surveillance Data in Support of Aging Management


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.

Author Information

Mager, TR
Westinghouse Electric Corp., Pittsburgh, PA
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Developed by Committee: E10
Pages: 87–98
DOI: 10.1520/STP24766S
ISBN-EB: 978-0-8031-5228-1
ISBN-13: 978-0-8031-1478-4