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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. Thu . "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 = {Thu Sep 01 00:00:00 EDT 2016},
month = {Thu Sep 01 00:00:00 EDT 2016}
}

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