Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks

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This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak ... continued below

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Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott & Hsu, C. August 1, 1999.

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Description

This paper presents our assessment of field experience related to pressurized water reactor (PWR) primary system leaks in terms of their number of rates, how aging affects frequency of leak events, the safety significance of such leaks, industry efforts to reduce leaks, and effectiveness of current leak detection systems. We have reviewed the licensee event reports to identify the events that took place during 1985 to the third quarter of 1996, and reviewed related technical literature and visited PWR plants to analyze these events. Our assessment shows that USNRC licensees have taken effective actions to reduce the number of leak events. One main reason for this decreasing trend was the elimination or reportable leakages from valve stem packing after 1991. Our review of leak events related to vibratory fatigue reveals a statistically significant decreasing trend with age (years of operation), but not in calendar time. Our assessment of worldwide data on leakage caused by thermal fatigue cracking is that the fatigue of aging piping is a safety significant issue. Our review of leak events has identified several susceptible sites in piping having high safety significance; but the inspection of some of these sites is not required by the ASME Code. These sites may be included in the risk-informed inspection programs.

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  • 1999 ASME Pressure Vessels and Piping Conference,Boston, MA,08/01/1999,08/05/1999

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  • Report No.: INEEL/CON-99-00037
  • Grant Number: DE-AC07-99ID-13727
  • Office of Scientific & Technical Information Report Number: 911384
  • Archival Resource Key: ark:/67531/metadc881776

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Office of Scientific & Technical Information Technical Reports

Reports, articles and other documents harvested from the Office of Scientific and Technical Information.

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  • August 1, 1999

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  • Sept. 22, 2016, 2:13 a.m.

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  • Nov. 22, 2016, 8 p.m.

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Shah, Vikram Naginbhai; Ware, Arthur Gates; Atwood, Corwin Lee; Sattison, Martin Blaine; Hartley, Robert Scott & Hsu, C. Assessment of Field Experience Related to Pressurized Water Reactor Primary System Leaks, article, August 1, 1999; [Idaho Falls, Idaho]. (digital.library.unt.edu/ark:/67531/metadc881776/: accessed December 11, 2017), University of North Texas Libraries, Digital Library, digital.library.unt.edu; crediting UNT Libraries Government Documents Department.