Environmentally assisted cracking of light-water reactor materials

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Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator ... continued below

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7 p.

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Chopra, O.K.; Chung, H.M.; Kassner, T.F. & Shack, W.J. February 1, 1996.

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Description

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used.

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7 p.

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INIS; OSTI as DE96008405

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  • 58. annual meeting of the American power conference, Chicago, IL (United States), 9-11 Apr 1996

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  • Other: DE96008405
  • Report No.: ANL/ET/CP--89183
  • Report No.: CONF-960426--2
  • Grant Number: W-31109-ENG-38
  • DOI: 10.2172/288440 | External Link
  • Office of Scientific & Technical Information Report Number: 211322
  • Archival Resource Key: ark:/67531/metadc666016

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  • February 1, 1996

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  • June 29, 2015, 9:42 p.m.

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  • April 7, 2016, 7:46 p.m.

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Chopra, O.K.; Chung, H.M.; Kassner, T.F. & Shack, W.J. Environmentally assisted cracking of light-water reactor materials, article, February 1, 1996; Illinois. (digital.library.unt.edu/ark:/67531/metadc666016/: accessed June 20, 2018), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.