DOE`s annealing prototype demonstration projects
One of the challenges U.S. utilities face in addressing technical issues associated with the aging of nuclear power plants is the long-term effect of plant operation on reactor pressure vessels (RPVs). As a nuclear plant operates, its RPV is exposed to neutrons. For certain plants, this neutron exposure can cause embrittlement of some of the RPV welds which can shorten the useful life of the RPV. This RPV embrittlement issue has the potential to affect the continued operation of a number of operating U.S. pressurized water reactor (PWR) plants. However, RPV material properties affected by long-term irradiation are recoverable through a thermal annealing treatment of the RPV. Although a dozen Russian-designed RPVs and several U.S. military vessels have been successfully annealed, U.S. utilities have stated that a successful annealing demonstration of a U.S. RPV is a prerequisite for annealing a licensed U.S. nuclear power plant. In May 1995, the Department of Energy`s Sandia National Laboratories awarded two cost-shared contracts to evaluate the feasibility of annealing U.S. licensed plants by conducting an anneal of an installed RPV using two different heating technologies. The contracts were awarded to the American Society of Mechanical Engineers (ASME) Center for Research and Technology Development (CRTD) and MPR Associates (MPR). The ASME team completed its annealing prototype demonstration in July 1996, using an indirect gas furnace at the uncompleted Public Service of Indiana`s Marble Hill nuclear power plant. The MPR team`s annealing prototype demonstration was scheduled to be completed in early 1997, using a direct heat electrical furnace at the uncompleted Consumers Power Company`s nuclear power plant at Midland, Michigan. This paper describes the Department`s annealing prototype demonstration goals and objectives; the tasks, deliverables, and results to date for each annealing prototype demonstration; and the remaining annealing technology challenges.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Brookhaven National Lab. (BNL), Upton, NY (United States)
- OSTI ID:
- 467940
- Report Number(s):
- NUREG/CP-0157-VOL.2; CONF-9610202-Vol.2; ON: TI97004274; TRN: 97:008406
- Resource Relation:
- Conference: 24. water reactor safety information meeting, Bethesda, MD (United States), 21-23 Oct 1996; Other Information: PBD: Feb 1997; Related Information: Is Part Of Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity; Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)]; PB: 443 p.
- Country of Publication:
- United States
- Language:
- English
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