Evaluation Of Risk Associated With Intergranular Stress Corrosion Cracking In Boiling Water Reactor Internals (NUREG/CR-6677) On this page:

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Publication Information

Manuscript Completed: June 2000
Date Published: July 2000

Prepared by:
A-G. Ware, D.K. Morton, M.E. Nitzel, S.A. Eide

Idaho National Engineering and Environmental Laboratory
Bechtel BWXT Idaho, LLC
Idaho Falls, ID 83415-3129

T.Y. Chang, NRC Project Manager

Prepared for:
Division of Engineering Technology
Office of Nuclear Regulatory Research
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001

NRC Job Code W6677

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Intergranular stress corrosion cracking (IGSCC) has been detected in a number of Boiling Water Reactor (BWR) vessel internals. The number of different types of components that have experienced cracking has increased over the years. Based on evaluations submitted by General Electric, BWR owners, and the BWRVIP group, the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission has determined that the short-term risk significance of cracking of BWR vessel internals is acceptable. The purpose of this report is to begin an evaluation of the long-term risk associated with IGSCC of BWR vessel internals. The differences in BWR types were studied, the history of IGSCC cracking in reactor internals was catalogued, and tables of accident scenarios involving cracked internals were developed. Screening methods were developed to reduce the number of accident scenarios that represent more significant risk to a more manageable number, and a final list of potential concerns was developed which ranked the scenarios as having a high, medium, or low impact on increasing the core damage frequency based on a qualitative risk assessment. In order to narrow the scope of the problem, the investigation was limited to a single BWR type (high-power BWR/4). Using several screening methods, the scenarios to investigate were reduced by about 2/3. To further concentrate the investigation and develop a methodology to be used, scenarios resulting from the failure of a single component (jet pump) were evaluated. The evaluation was then extended to the remaining reactor internals components. It was concluded that with the current BWRVIP inspection, monitoring, and repair proposals, there is expected to be no significant increase in CDF (< 5 x 10-6 events/yr) caused by failures of BWR internals. That is, IGSCC problems can be identified and evaluated or corrected, to preclude a significant increase in the CDF.

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