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Title: Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

Abstract

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well asmore » earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.« less

Authors:
 [1];  [2];  [3];  [1];  [1];  [3];  [3];  [3]
  1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
  2. Univ. of Tennessee, Knoxville, TN (United States)
  3. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Publication Date:
Research Org.:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1369369
Report Number(s):
INL/EXT-16-40050
TRN: US1703380
DOE Contract Number:
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; NUCLEAR POWER PLANTS; PROBABILISTIC ESTIMATION; FRACTURE MECHANICS; PRESSURE VESSELS; WATER MODERATED REACTORS; WATER COOLED REACTORS; Probabilistic Fracture Mechanics; Reactor Pressure Vessel Embrittlement

Citation Formats

Spencer, Benjamin, Backman, Marie, Williams, Paul, Hoffman, William, Alfonsi, Andrea, Dickson, Terry, Bass, B. Richard, and Klasky, Hilda. Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws. United States: N. p., 2016. Web. doi:10.2172/1369369.
Spencer, Benjamin, Backman, Marie, Williams, Paul, Hoffman, William, Alfonsi, Andrea, Dickson, Terry, Bass, B. Richard, & Klasky, Hilda. Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws. United States. doi:10.2172/1369369.
Spencer, Benjamin, Backman, Marie, Williams, Paul, Hoffman, William, Alfonsi, Andrea, Dickson, Terry, Bass, B. Richard, and Klasky, Hilda. 2016. "Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws". United States. doi:10.2172/1369369. https://www.osti.gov/servlets/purl/1369369.
@article{osti_1369369,
title = {Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws},
author = {Spencer, Benjamin and Backman, Marie and Williams, Paul and Hoffman, William and Alfonsi, Andrea and Dickson, Terry and Bass, B. Richard and Klasky, Hilda},
abstractNote = {This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.},
doi = {10.2172/1369369},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • Significant attention has been focused on the issue of reactor vessel pressurized thermal shock (PTS) for many years. Pressurized thermal shock transient events are characterized by a rapid cooldown at potentially high pressure levels that could lead to a reactor vessel integrity concern for some pressurized water reactors. As a result of regulatory and industry efforts in the early 1980`s, a probabilistic risk assessment methodology has been established to address this concern. Probabilistic fracture mechanics analyses are performed as part of this methodology to determine conditional probability of significant flaw extension for given pressurized thermal shock events. While recent industrymore » efforts are underway to benchmark probabilistic fracture mechanics computer codes that are currently used by the nuclear industry, Part I of this report describes the comparison of two independent computer codes used at the time of the development of the original U.S. Nuclear Regulatory Commission (NRC) pressurized thermal shock rule. The work that was originally performed in 1982 and 1983 to compare the U.S. NRC - VISA and Westinghouse (W) - PFM computer codes has been documented and is provided in Part I of this report. Part II of this report describes the results of more recent industry efforts to benchmark PFM computer codes used by the nuclear industry. This study was conducted as part of the USNRC-EPRI Coordinated Research Program for reviewing the technical basis for pressurized thermal shock (PTS) analyses of the reactor pressure vessel. The work focused on the probabilistic fracture mechanics (PFM) analysis codes and methods used to perform the PTS calculations. An in-depth review of the methodologies was performed to verify the accuracy and adequacy of the various different codes. The review was structured around a series of benchmark sample problems to provide a specific context for discussion and examination of the fracture mechanics methodology.« less
  • The OCA-P code is a probabilistic fracture-mechanics code that was prepared specifically for evaluating the integrity of pressurized-water reactor vessels when subjected to overcooling-accident loading conditions. The code has two-dimensional- and some three-dimensional-flaw capability; it is based on linear-elastic fracture mechanics; and it can treat cladding as a discrete region. Both deterministic and probabilistic analyses can be performed. For the former analysis, it is possible to conduct a search for critical values of the fluence and the nil-ductility reference temperature corresponding to incipient initiation of the initial flaw. The probabilistic portion of OCA-P is based on Monte Carlo techniques, andmore » simulated parameters include fluence, flaw depth, fracture toughness, nil-ductility reference temperature, and concentrations of copper, nickel, and phosphorous. Plotting capabilities include the construction of critical-crack-depth diagrams (deterministic analysis) and various histograms (probabilistic analysis).« less
  • The current regulations, as set forth by the United States Nuclear Regulatory Commission (NRC), to insure that lightwater nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to planned reactor 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. The technical basis for these regulations contains many aspects that are now broadly recognized by the technical community as being unnecessarily conservative and some plants are finding it increasingly difficult to comply with the current regulations. Consequently, a goalmore » of current NRC research is to derive a technical basis for a risk-informed revision to the current requirements that reduces the conservatism and also is consistent with the methods previously used to develop a risk-informed revision to the regulations for accidental transients such as pressurized thermal shock (PTS). Previous publications have been successful in illustrating potential methods to provide a risk-informed relaxation to the current regulations for normal transients. Thus far, probabilistic fracture mechanics (PFM) analyses have been performed at 60 effective full power years (EFPY) for one of the reactors evaluated as part of the PTS re-evaluation project. In these previous analyses/publications, consistent with the assumptions utilized for this particular reactor in the PTS reevaluation, all flaws for this reactor were postulated to be embedded. The objective of this paper is to review the analysis results and conclusions from previous publications on this subject and to attempt to modify/generalize these conclusions to include RPVs postulated to contain only inner-surface breaking flaws or a combination of embedded flaws and inner-surface breaking flaws.« less
  • The Nuclear Regulatory Commission (NRC) requested Oak Ridge National Laboratory (ORNL) to perform a pressurized-thermal-shock (PTS) probabilistic fracture mechanics (PFM) sensitivity analysis for the Yankee Rowe reactor pressure vessel, for the fluences corresponding to the end of operating cycle 22, using a specific small-break-loss- of-coolant transient as the loading condition. Regions of the vessel with distinguishing features were to be treated individually -- upper axial weld, lower axial weld, circumferential weld, upper plate spot welds, upper plate regions between the spot welds, lower plate spot welds, and the lower plate regions between the spot welds. The fracture analysis methods usedmore » in the analysis of through-clad surface flaws were those contained in the established OCA-P computer code, which was developed during the Integrated Pressurized Thermal Shock (IPTS) Program. The NRC request specified that the OCA-P code be enhanced for this study to also calculate the conditional probabilities of failure for subclad flaws and embedded flaws. The results of this sensitivity analysis provide the NRC with (1) data that could be used to assess the relative influence of a number of key input parameters in the Yankee Rowe PTS analysis and (2) data that can be used for readily determining the probability of vessel failure once a more accurate indication of vessel embrittlement becomes available. This report is designated as HSST report No. 117.« less