Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR plants
Recent data from the HFIR vessel surveillance program indicate a substantial radiation embrittlement rate effect at low irradiation temperatures (/approximately/120/degree/F) for A212-B, A350-LF3, A105-II, and corresponding welds. PWR vessel supports are fabricated of similar materials and are subjected to the same low temperatures and fast neutron fluxes (10/sup 8/ to 10/sup 9/ neutrons/cm/sup 2//center dot/s, E > 1.0 MeV) as those in the HFIR vessel. Thus, the embrittlement rate of these structures may be greater than previously anticipated. A study sponsored by the NRC is under way at ORNL to determine the impact of the rate effect on PWR vessel-support life expectancy. The scope includes the interpretation and application of the HFIR data, a survey of all light-water-reactor vessel support designs, and a structural and fracture-mechanics analysis of the supports for two specific PWR plants of particular interest with regard to a potential for support failure as a result of propagation of flaws. Calculations performed thus far indicate best-estimate critical flaw sizes, corresponding to 32 EFPY, of /approximately/0.2 in. for one plant and /approximately/0.4 in. for the other. These flaw sizes are small enough to be of concern. However, it appears that low-cycle fatigue is not a viable mechanism for creation of flaws of this size, and thus, presumably, such flaws would have to exist at the time of fabrication. 59 refs., 128 figs., 49 tabs.
- Research Organization:
- US Nuclear Regulatory Commission (NRC), Washington, DC (United States). Office of Nuclear Regulatory Research; Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 6371359
- Report Number(s):
- NUREG/CR-5320; ORNL/TM-10966; ON: TI89007009; TRN: 89-005899
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
36 MATERIALS SCIENCE
PRESSURE VESSELS
SUPPORTS
EMBRITTLEMENT
TROJAN REACTOR
TURKEY POINT-3 REACTOR
CRACKS
DATA ANALYSIS
DEFECTS
EXPERIMENTAL DATA
FRACTURES
HFIR REACTOR
MECHANICAL PROPERTIES
NEUTRON FLUX
RADIATION EFFECTS
RECOMMENDATIONS
STRESS ANALYSIS
TEMPERATURE EFFECTS
CONTAINERS
DATA
ENRICHED URANIUM REACTORS
FAILURES
INFORMATION
IRRADIATION REACTORS
ISOTOPE PRODUCTION REACTORS
MECHANICAL STRUCTURES
NUMERICAL DATA
POWER REACTORS
PWR TYPE REACTORS
RADIATION FLUX
REACTORS
RESEARCH AND TEST REACTORS
RESEARCH REACTORS
TANK TYPE REACTORS
TEST REACTORS
THERMAL REACTORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
210200* - Power Reactors
Nonbreeding
Light-Water Moderated
Nonboiling Water Cooled
360103 - Metals & Alloys- Mechanical Properties
360106 - Metals & Alloys- Radiation Effects