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Risk-Based Fracture Evaluation of Reactor Vessels Subjected to Cool-Down Transients Associated With Shutdown: An Examination of the Effects of Different Modeling Approaches on Estimated Failure Probabilities

[+] Author Affiliations
T. L. Dickson

Oak Ridge National Laboratory, Oak Ridge, TN

M. T. EricksonKirk

U.S. Nuclear Regulatory Commission, Washington, D.C.

Paper No. PVP2006-ICPVT-11-93813, pp. 565-570; 6 pages
doi:10.1115/PVP2006-ICPVT-11-93813
From:
  • ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference
  • Volume 3: Design and Analysis
  • Vancouver, BC, Canada, July 23–27, 2006
  • Conference Sponsors: Pressure Vessels and Piping Division
  • ISBN: 0-7918-4754-3 | eISBN: 0-7918-3782-3
  • Copyright © 2006 by ASME

abstract

The current regulations, as set forth by the United States Nuclear Regulatory Commission (USNRC), to insure that light-water nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned startup (heat-up) and shutdown (cool-down) transients are specified in Appendix G to 10 CFR Part 50, which incorporates by reference Appendix G to Section XI of the ASME Code. In 1999, the USNRC initiated the interdisciplinary Pressurized Thermal Shock (PTS) Re-evaluation Project to determine if a technical basis could be established to support a relaxation in the current PTS regulations. The PTS re-evaluation project included the development and application of an updated risk-based computational methodology that incorporates several advancements applicable to modeling the physics of vessel fracture due to thermal hydraulic transients imposed on the RPV inner surface. The results of the PTS re-evaluation project demonstrated that there is a sound technical basis to support a relaxation of the current PTS regulations. The results of the PTS re-evaluation are currently under review by the USNRC. Based on the promising results of the PTS re-evaluation, the USNRC has recently applied the updated computational methodology to fracture evaluations of RPVs subjected to planned cool-down transients, associated with reactor shutdown, derived in accordance with ASME Section XI – Appendix G. The objective of these analyses is to determine if a sound technical basis can be established to provide a relaxation to the current regulations for the derivation of bounding cool-down transients as specified in Appendix G to Section XI of the ASME Code. This paper provides a brief overview of these analyses, results, and the implications of the results.

Copyright © 2006 by ASME

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