14.41. When present nuclear power plants were put into service, a 40- year operating lifetime was projected, essentially on an arbitrary basis. However, a maximum of 40 years was established by Congress in 1954 as the duration of an operating license. With no new plants being built and some plants approaching 30 years of service, operating license renewal for an additional 20 years has been recommended by the U. S. Department of Energy [5]. To allow ample time for regulatory decisions, NRC has de­veloped rules for considering license renewal. Essentially, the principle applies that the present adequate level of safety must be maintained during any extended operating period [6].

14.42. Extended life studies have concluded that it is feasible to maintain safety levels, regardless of age, through repair, refurbishment, and re­placement. Degradation in component performance can be detected by extensive routine surveillance and testing. Various aging mechanisms were identified which fall into categories of fatigue, erosion, corrosion, service wear, radiation embrittlement, thermal embrittlement, chemical effects, material creep, and environmental degradation [7]. Among these, PWR pressure vessel radiation embrittlement is probably the most important and will be discussed further in the next section. Corrosion problems have required the replacement of some PWR steam generators. Although such replacement is expensive, with about a 6-month outage required, it is a proven procedure.

14.43. In 1985, the NRC added rule 10 CFR 50.61 entitled “Fracture Toughness Requirements for Protection Against Thermal Shock Events.” The rule establishes pressure vessel screening criteria and calculation meth­ods for a reference temperature for nil-ductility transition (§7.16). Of con­cern is the possibility of brittle fracture during a pressurized thermal shock transient as a result of the sudden injection of cold water during operation of the Emergency Core Cooling Systems. For plates and axial welds, the maximum NDT temperature is 132°C (270°F), while 149°C (300°F) is the limit for circumferential welds at the core beltline. Licensees are required to submit projected values up the time of license expiration. Should the criteria be exceeded, neutron flux reduction programs are required.

14.44. The pressure vessel wall fast-neutron flux exposure can be re­duced by lower power reactor operation, but a preferred approach is to use special reload core designs to provide very low neutron leakage. A number of design options are available but generally feature loading the most depleted fuel or special blanket assemblies around the periphery of the core (§10.26).

14.45. In-place thermal annealing of a reactor vessel to remove some or all of the effects of neutron embrittlement is a “last ditch” option being investigated [8]. A technique that has been used successfully in Russia on defueled PWRs involves the use of special heaters in the region of critical vessel welds to raise the metal temperature to 454°C for about 6 days. Wet annealing is another possibility but would lead to lesser benefits. In this technique, the vessel and primary cooling circuit is held for about 6 days at a temperature of 340°C using main reactor cooling pump friction heat. Regulatory considerations associated with annealing methods would re­quire resolution.