Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor Metadata

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Title

  • Main Title Revisiting the Integrated Pressurized Thermal Shock Studies of an Aging Pressurized Water Reactor

Creator

  • Author: Bryson, J. W.
    Creator Type: Personal
  • Author: Dickson, T. L.
    Creator Type: Personal
  • Author: Malik, S. N. M.
    Creator Type: Personal
  • Author: Simonen, F. A.
    Creator Type: Personal

Contributor

  • Sponsor: United States. Department of Energy. Office of Energy Research.
    Contributor Type: Organization
    Contributor Info: USDOE Office of Energy Research (ER)
  • Sponsor: USNRC Office of Nuclear Regulatory Research
    Contributor Type: Organization

Publisher

  • Name: Oak Ridge National Laboratory
    Place of Publication: Oak Ridge, Tennessee

Date

  • Creation: 1999-08-01

Language

  • English

Description

  • Content Description: The Integrated Pressurized Thermal Shock (IPTS) studies were a series of studies performed in the early-mid 1980s as part of an NRC-organized comprehensive research project to confirm the technical bases for the pressurized thermal shock (PTS) rule, and to aid in the development of guidance for licensee plant-specific analyses. The research project consisted of PTS pilot analyses for three PWRs: Oconee Unit 1, designed by Babcock and Wilcox; Calvert Cliffs Unit 1, designed by Combustion Engineering; and H.B. Robinson Unit 2, designed by Westinghouse. The primary objectives of the IPTS studies were (1) to provide for each of the three plants an estimate of the probability of a crack propagating through the wall of a reactor pressure vessel (RPV) due to PTS; (2) to determine the dominant overcooling sequences, plant features, and operator actions and the uncertainty in the plant risk due to PTS; and (3) to evaluate the effectiveness of potential corrective actions. The NRC is currently evaluating the possibility of revising current PTS regulatory guidance. Technical bases must be developed to support any revisions. In the years since the results of IPTS studies were published, the fracture mechanics model, the embrittlement database, embrittlement correlation, inputs for flaw distributions, and the probabilistic fracture mechanics (PFM) computer code have been refined. An ongoing effort is underway to determine the impact of these fracture-technology refinements on the conditional probabilities of vessel failure calculated in the IPTS Studies. This paper discusses the results of these analyses performed for one of these plants.
  • Physical Description: 11 Pages

Subject

  • Keyword: Calvert Cliffs-1 Reactor
  • STI Subject Categories: 21 Nuclear Power Reactors And Associated Plants
  • Keyword: Robinson-2 Reactor
  • Keyword: Thermal Shock
  • Keyword: Oconee-1 Reactor

Source

  • Conference: ASME Pressure Vessels and Piping Conference, Boston, MA, August 1-5, 1999

Collection

  • Name: Office of Scientific & Technical Information Technical Reports
    Code: OSTI

Institution

  • Name: UNT Libraries Government Documents Department
    Code: UNTGD

Resource Type

  • Article

Format

  • Text

Identifier

  • Other: DE00006153
  • Report No.: ORNL/CP-102439
  • Grant Number: AC05-96OR22464
  • Office of Scientific & Technical Information Report Number: 6153
  • Archival Resource Key: ark:/67531/metadc692874
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