Heavy-Section Steel Irradiation Program. Volume 2, No. 1: Semiannual progress report, October 1990--March 1991

Technical Report · Fri Jul 01 04:00:00 EDT 1994

DOI:https://doi.org/10.2172/10173358· OSTI ID:10173358

Corwin, W R ^[1]

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1. Oak Ridge National Lab., TN (United States)

Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the capabilities and limitations of the integrity inherent in the RPV. For this reason, the Heavy-Section Steel Irradiation (HSSI) Program has been established with its primary goal to provide a thorough, quantitative assessment of the effects of neutron irradiation on the material behavior, and in particular the fracture toughness properties, of typical pressure-vessel steels as they relate to light-water reactor pressure-vessel integrity. The HSSI Program is arranged into nine tasks: (1) program management, (2) K{sub ic} curve shift in high-copper welds, (3) K{sub ia} curve shift in high-copper welds, (4) irradiation effects on cladding, (5) K{sub ic} and K{sub ia} curve shifts in low upper-shelf (LUS) weld, (6) irradiation effects in a commercial LUS weld, (7) microstructural analysis of irradiation, (8) in-service aged material evaluations, and (9) correlation monitor materials. During this period, additional analyses on the effects of precleavage stable ductile tearing on the toughness of high-copper welds 72W and 73W demonstrated that the size effects observed in the transition region are not due to substantial differences in ductile tearing behavior. Possible modifications to irradiated duplex crack-arrest specimens were examined to increase the likelihood of their successful testing. Characterization of a second batch of 72W and 73W welds was begun and results of the Charpy V-notch testing is provided. A review of literature on the annealing response of reactor pressure vessel steels was initiated.

• Research Organization: Nuclear Regulatory Commission, Washington, DC (United States). Div. of Engineering; Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: Nuclear Regulatory Commission, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10173358
• Report Number(s): NUREG/CR--5591-Vol.2-No.1; ORNL/TM--11568-Vol.2-No.1; ON: TI94016626
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
210100
210200
22 GENERAL STUDIES OF NUCLEAR REACTORS
220900
36 MATERIALS SCIENCE
360106
COPPER ALLOYS
FRACTURE PROPERTIES
MECHANICAL TESTS
MIDLAND-1 REACTOR
MITIGATION
PHYSICAL RADIATION EFFECTS
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
BOILING WATER COOLED
POWER REACTORS
NONBREEDING
LIGHT-WATER MODERATED
NONBOILING WATER COOLED
PRESSURE VESSELS
PROGRESS REPORT
RADIATION EFFECTS
REACTOR ACCIDENTS
REACTOR MATERIALS
REACTOR SAFETY
RELIABILITY
STAINLESS STEELS
STEELS
WATER COOLED REACTORS
WELDED JOINTS