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Title: Grizzly Staus Report

Abstract

This report summarizes work during FY 2014 to develop capabilities to predict embrittlement of reactor pressure vessel steel, and to assess the response of embrittled reactor pressure vessels to postulated accident conditions. This work has been conducted a three length scales. At the engineering scale, 3D fracture mechanics capabilities have been developed to calculate stress intensities and fracture toughnesses, to perform a deterministic assessment of whether a crack would propagate at the location of an existing flaw. This capability has been demonstrated on several types of flaws in a generic reactor pressure vessel model. Models have been developed at the scale of fracture specimens to develop a capability to determine how irradiation affects the fracture toughness of material. Verification work has been performed on a previously-developed model to determine the sensitivity of the model to specimen geometry and size effects. The effects of irradiation on the parameters of this model has been investigated. At lower length scales, work has continued in an ongoing to understand how irradiation and thermal aging affect the microstructure and mechanical properties of reactor pressure vessel steel. Previously-developed atomistic kinetic monte carlo models have been further developed and benchmarked against experimental data. Initial work has beenmore » performed to develop models of nucleation in a phase field model. Additional modeling work has also been performed to improve the fundamental understanding of the formation mechanisms and stability of matrix defects caused.« less

Authors:
 [1];  [1];  [1];  [1];  [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:
1169239
Report Number(s):
INL/EXT-14-33251
M3LW-14IN0704013
DOE Contract Number:  
AC07-05ID14517
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 36 MATERIALS SCIENCE; Embrittlement; Grizzly; Reactor Pressure

Citation Formats

Spencer, Benjamin, Zhang, Yongfeng, Chakraborty, Pritam, Backman, Marie, Hoffman, William, Schwen, Daniel, Biner, S. Bulent, and Bai, Xianming. Grizzly Staus Report. United States: N. p., 2014. Web. doi:10.2172/1169239.
Spencer, Benjamin, Zhang, Yongfeng, Chakraborty, Pritam, Backman, Marie, Hoffman, William, Schwen, Daniel, Biner, S. Bulent, & Bai, Xianming. Grizzly Staus Report. United States. https://doi.org/10.2172/1169239
Spencer, Benjamin, Zhang, Yongfeng, Chakraborty, Pritam, Backman, Marie, Hoffman, William, Schwen, Daniel, Biner, S. Bulent, and Bai, Xianming. 2014. "Grizzly Staus Report". United States. https://doi.org/10.2172/1169239. https://www.osti.gov/servlets/purl/1169239.
@article{osti_1169239,
title = {Grizzly Staus Report},
author = {Spencer, Benjamin and Zhang, Yongfeng and Chakraborty, Pritam and Backman, Marie and Hoffman, William and Schwen, Daniel and Biner, S. Bulent and Bai, Xianming},
abstractNote = {This report summarizes work during FY 2014 to develop capabilities to predict embrittlement of reactor pressure vessel steel, and to assess the response of embrittled reactor pressure vessels to postulated accident conditions. This work has been conducted a three length scales. At the engineering scale, 3D fracture mechanics capabilities have been developed to calculate stress intensities and fracture toughnesses, to perform a deterministic assessment of whether a crack would propagate at the location of an existing flaw. This capability has been demonstrated on several types of flaws in a generic reactor pressure vessel model. Models have been developed at the scale of fracture specimens to develop a capability to determine how irradiation affects the fracture toughness of material. Verification work has been performed on a previously-developed model to determine the sensitivity of the model to specimen geometry and size effects. The effects of irradiation on the parameters of this model has been investigated. At lower length scales, work has continued in an ongoing to understand how irradiation and thermal aging affect the microstructure and mechanical properties of reactor pressure vessel steel. Previously-developed atomistic kinetic monte carlo models have been further developed and benchmarked against experimental data. Initial work has been performed to develop models of nucleation in a phase field model. Additional modeling work has also been performed to improve the fundamental understanding of the formation mechanisms and stability of matrix defects caused.},
doi = {10.2172/1169239},
url = {https://www.osti.gov/biblio/1169239}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}