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Title: Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions

Abstract

Environmentally assisted cracking (EAC) is the most problematic form of localized corrosion in irradiated areas of nuclear power plants. EAC is any phenomenon where a combination of environment, material, and tensile stress result in cracking, including stress corrosion cracking (SCC), and hydrogen embrittlement. For this project an integrated EAC model based on first-principles electrochemistry and physical metallurgy was developed. The effect of neutron and gamma radiation dose is included explicitly so that irradiation-assisted SCC can be studied. The primary dependent variable in the model is the environmentally assisted crack growth rate as a function of chemical and mechanical parameters. The model integrates bulk and local chemistry models with the mechanical and chemical (alloy) behavior of the metal to allow the study of EAC in general. The model evaluates the water chemistry including radiolysis and hydrogen water chemistry and metal properties including radiation-induced segregation, radiation hardening, and the crack tip strain rate of a growing crack. The model inputs include thermal-hydraulic data such as flow rate, temperature, power level, dose, pressure, and initial water chemistry throughout the reactor, as well as dimension, stress/strain conditions, and initial sensitization data for reactor components.

Authors:
 [1]
  1. Massachusetts Inst. of Tech., Cambridge, MA (United States)
Publication Date:
OSTI Identifier:
678155
Report Number(s):
CONF-990605-
Journal ID: TANSAO; ISSN 0003-018X; TRN: 99:009140
Resource Type:
Journal Article
Journal Name:
Transactions of the American Nuclear Society
Additional Journal Information:
Journal Volume: 80; Conference: 1999 annual meeting of the American Nuclear Society (ANS), Boston, MA (United States), 6-10 Jun 1999; Other Information: PBD: 1999
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; 21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; IRON ALLOYS; NICKEL ALLOYS; CHROMIUM ALLOYS; HEAT RESISTING ALLOYS; CRACK PROPAGATION; CORROSION; REACTOR MATERIALS; HYDROGEN EMBRITTLEMENT; MATHEMATICAL MODELS; WATER CHEMISTRY; PHYSICAL RADIATION EFFECTS

Citation Formats

Pleune, T.T. Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions. United States: N. p., 1999. Web.
Pleune, T.T. Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions. United States.
Pleune, T.T. Wed . "Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions". United States.
@article{osti_678155,
title = {Integrated environmental degradation model for Fe-Ni-Cr alloys in irradiated aqueous solutions},
author = {Pleune, T.T.},
abstractNote = {Environmentally assisted cracking (EAC) is the most problematic form of localized corrosion in irradiated areas of nuclear power plants. EAC is any phenomenon where a combination of environment, material, and tensile stress result in cracking, including stress corrosion cracking (SCC), and hydrogen embrittlement. For this project an integrated EAC model based on first-principles electrochemistry and physical metallurgy was developed. The effect of neutron and gamma radiation dose is included explicitly so that irradiation-assisted SCC can be studied. The primary dependent variable in the model is the environmentally assisted crack growth rate as a function of chemical and mechanical parameters. The model integrates bulk and local chemistry models with the mechanical and chemical (alloy) behavior of the metal to allow the study of EAC in general. The model evaluates the water chemistry including radiolysis and hydrogen water chemistry and metal properties including radiation-induced segregation, radiation hardening, and the crack tip strain rate of a growing crack. The model inputs include thermal-hydraulic data such as flow rate, temperature, power level, dose, pressure, and initial water chemistry throughout the reactor, as well as dimension, stress/strain conditions, and initial sensitization data for reactor components.},
doi = {},
journal = {Transactions of the American Nuclear Society},
number = ,
volume = 80,
place = {United States},
year = {1999},
month = {9}
}