skip to main content
OSTI.GOV title logo U.S. Department of Energy
Office of Scientific and Technical Information

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
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. doi:10.2172/1169239.
Spencer, Benjamin, Zhang, Yongfeng, Chakraborty, Pritam, Backman, Marie, Hoffman, William, Schwen, Daniel, Biner, S. Bulent, and Bai, Xianming. Mon . "Grizzly Staus Report". United States. doi: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},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Sep 01 00:00:00 EDT 2014},
month = {Mon Sep 01 00:00:00 EDT 2014}
}

Technical Report:

Save / Share:
  • BPA proposes to clear unwanted vegetation from the rights of way and access roads for BPA’s Grizzly-Summerlake, Grizzly-Captain Jack Transmission Lines, beginning April 2002, and ending July 2002.
  • This notice announces BPA`s decision to construct, operate, and maintain the Grizzly Substation Fiber Optic Project (Project). This Project is part of a continuing effort by BPA to complete a regionwide upgrade of its existing telecommunications system. The US Forest Service and BPA jointly prepared the Grizzly Substation Fiber Optic Project Environmental Assessment (EA) (DOE/EA-1241) evaluating the potential environmental impacts of the Proposed Action, the Underground Installation Alternative, and the No Action Alternative. Based on the analysis in the EA, the US Forest Service and BPA have determined that the Proposed Action is not a major Federal action significantly affectingmore » the quality of the human environment, within the meaning of the National Environmental Policy Act (NEPA) of 1969. Therefore, the preparation of an Environmental Impact Statement (EIS) is not required and BPA is issuing this FONSI. The US Forest Service has separately issued a FONSI and Decision Notice authorizing BPA to construct, operate, and maintain the Project within the Crooked River National Grassland (Grassland).« less
  • Nuclear power currently provides a significant fraction of the United States’ non-carbon emitting power generation. In future years, nuclear power must continue to generate a significant portion of the nation’s electricity to meet the growing electricity demand, clean energy goals, and ensure energy independence. New reactors will be an essential part of the expansion of nuclear power. However, given limits on new builds imposed by economics and industrial capacity, the extended service of the existing fleet will also be required.
  • The Grizzly software application is being developed under the Light Water Reactor Sustainability (LWRS) program to address aging and material degradation issues that could potentially become an obstacle to life extension of nuclear power plants beyond 60 years of operation. Grizzly is based on INL’s MOOSE multiphysics simulation environment, and can simultaneously solve a variety of tightly coupled physics equations, and is thus a very powerful and flexible tool with a wide range of potential applications. Grizzly, the development of which was begun during fiscal year (FY) 2012, is intended to address degradation in a variety of critical structures. Themore » reactor pressure vessel (RPV) was chosen for an initial application of this software. Because it fulfills the critical roles of housing the reactor core and providing a barrier to the release of coolant, the RPV is clearly one of the most safety-critical components of a nuclear power plant. In addition, because of its cost, size and location in the plant, replacement of this component would be prohibitively expensive, so failure of the RPV to meet acceptance criteria would likely result in the shutting down of a nuclear power plant. The current practice used to perform engineering evaluations of the susceptibility of RPVs to fracture is to use the ASME Master Fracture Toughness Curve (ASME Code Case N-631 Section III). This is used in conjunction with empirically based models that describe the evolution of this curve due to embrittlement in terms of a transition temperature shift. These models are based on an extensive database of surveillance coupons that have been irradiated in operating nuclear power plants, but this data is limited to the lifetime of the current reactor fleet. This is an important limitation when considering life extension beyond 60 years. The currently available data cannot be extrapolated with confidence further out in time because there is a potential for additional damage mechanisms (i.e. late blooming phases) to become active later in life beyond the current operational experience. To develop a tool that can eventually serve a role in decision-making, it is clear that research and development must be perfomed at multiple scales. At the engineering scale, a multiphysics analysis code that can capture the thermomechanical response of the RPV under accident conditions, including detailed fracture mechanics evaluations of flaws with arbitrary geometry and orientation, is needed to assess whether the fracture toughness, as defined by the master curve, including the effects of embrittlement, is exceeded. At the atomistic scale, the fundamental mechanisms of degradation need to be understood, including the effects of that degradation on the relevant material properties. In addition, there is a need to better understand the mechanisms leading to the transition from ductile to brittle fracture through improved continuum mechanics modeling at the fracture coupon scale. Work is currently being conducted at all of these levels with the goal of creating a usable engineering tool informed by lower length-scale modeling. This report summarizes progress made in these efforts during FY 2013.« less
  • This report summarizes work conducted during fiscal year 2013 to work toward developing a full capability to evaluate fracture contour J-integrals to the Grizzly code. This is a progress report on ongoing work. During the next fiscal year, this capability will be completed, and Grizzly will be capable of evaluating these contour integrals for 3D geometry, including the effects of thermal stress and large deformation. A usable, limited capability has been developed, which is capable of evaluating these integrals on 2D geometry, without considering the effects of material nonlinearity, thermal stress or large deformation. This report presents an overview ofmore » the approach used, along with a demonstration of the current capability in Grizzly, including a comparison with an analytical solution.« less