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Title: 3D J-Integral Capability in Grizzly

Abstract

This report summarizes work done to develop a capability to evaluate fracture contour J-Integrals in 3D in the Grizzly code. In the current fiscal year, a previously-developed 2D implementation of a J-Integral evaluation capability has been extended to work in 3D, and to include terms due both to mechanically-induced strains and due to gradients in thermal strains. This capability has been verified against a benchmark solution on a model of a curved crack front in 3D. The thermal term in this integral has been verified against a benchmark problem with a thermal gradient. These developments are part of a larger effort to develop Grizzly as a tool that can be used to predict the evolution of aging processes in nuclear power plant systems, structures, and components, and assess their capacity after being subjected to those aging processes. The capabilities described here have been developed to enable evaluations of Mode- stress intensity factors on axis-aligned flaws in reactor pressure vessels. These can be compared with the fracture toughness of the material to determine whether a pre-existing flaw would begin to propagate during a pos- tulated pressurized thermal shock accident. This report includes a demonstration calculation to show how Grizzly is usedmore » to perform a deterministic assessment of such a flaw propagation in a degraded reactor pressure vessel under pressurized thermal shock conditions. The stress intensity is calculated from J, and the toughness is computed using the fracture master curve and the degraded ductile to brittle transition temperature.« less

Authors:
 [1];  [1];  [1];  [1]
  1. Idaho National Lab. (INL), Idaho Falls, ID (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:
1169240
Report Number(s):
INL/EXT-14-33257
M3LW-14IN07040110
DOE Contract Number:  
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 97 MATHEMATICS AND COMPUTING; Fracture Mechanics; Grizzly; Reactor Pressure Vessel

Citation Formats

Spencer, Benjamin, Backman, Marie, Chakraborty, Pritam, and Hoffman, William. 3D J-Integral Capability in Grizzly. United States: N. p., 2014. Web. doi:10.2172/1169240.
Spencer, Benjamin, Backman, Marie, Chakraborty, Pritam, & Hoffman, William. 3D J-Integral Capability in Grizzly. United States. https://doi.org/10.2172/1169240
Spencer, Benjamin, Backman, Marie, Chakraborty, Pritam, and Hoffman, William. 2014. "3D J-Integral Capability in Grizzly". United States. https://doi.org/10.2172/1169240. https://www.osti.gov/servlets/purl/1169240.
@article{osti_1169240,
title = {3D J-Integral Capability in Grizzly},
author = {Spencer, Benjamin and Backman, Marie and Chakraborty, Pritam and Hoffman, William},
abstractNote = {This report summarizes work done to develop a capability to evaluate fracture contour J-Integrals in 3D in the Grizzly code. In the current fiscal year, a previously-developed 2D implementation of a J-Integral evaluation capability has been extended to work in 3D, and to include terms due both to mechanically-induced strains and due to gradients in thermal strains. This capability has been verified against a benchmark solution on a model of a curved crack front in 3D. The thermal term in this integral has been verified against a benchmark problem with a thermal gradient. These developments are part of a larger effort to develop Grizzly as a tool that can be used to predict the evolution of aging processes in nuclear power plant systems, structures, and components, and assess their capacity after being subjected to those aging processes. The capabilities described here have been developed to enable evaluations of Mode- stress intensity factors on axis-aligned flaws in reactor pressure vessels. These can be compared with the fracture toughness of the material to determine whether a pre-existing flaw would begin to propagate during a pos- tulated pressurized thermal shock accident. This report includes a demonstration calculation to show how Grizzly is used to perform a deterministic assessment of such a flaw propagation in a degraded reactor pressure vessel under pressurized thermal shock conditions. The stress intensity is calculated from J, and the toughness is computed using the fracture master curve and the degraded ductile to brittle transition temperature.},
doi = {10.2172/1169240},
url = {https://www.osti.gov/biblio/1169240}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}