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Title: Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits. 2: Simulation of operating reactors

Abstract

The DAMAGE-PREDICTOR computer code, which has the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential (ECP), and the crack growth rate (CGR) of a reference crack in sensitized Type 304 stainless steel, is used to evaluate the responses of the Dresden-2 and Duane Arnold boiling water reactors (BWRs) to hydrogen water chemistry (HWC). The HWC simulations for these two BWRs are carried out for feedwater hydrogen concentrations (H{sub 2}, O{sub 2}, H{sub 2}O{sub 2}, etc.), ECP, and CGR are predicted for various components in the heat transport circuits (HTCs) of the two reactors. It is found that while 1.3 ppm of feedwater hydrogen is needed to protect part of the lower downcomer, the recirculation system, and the lower plenum in Dresden-2 from intergranular stress corrosion cracking, only 0.3 ppm is needed to achieve the same goal in Duane Arnold. However, it is also found that the ECP in many regions (core channel, core bypass, upper plenum, downcomer, etc.) in the HTCs cannot be lowered to below the critical corrosion potential of {minus}0.23 V{sub SHE} for sensitized Type 304 stainless steels, even when [H{sub 2}]{sub FW} is as high as 2.0 ppm.

Authors:
;  [1]
  1. Pennsylvania State Univ., University Park, PA (United States). Center for Advanced Materials
Publication Date:
OSTI Identifier:
264347
Resource Type:
Journal Article
Journal Name:
Nuclear Science and Engineering
Additional Journal Information:
Journal Volume: 123; Journal Issue: 2; Other Information: PBD: Jun 1996
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 36 MATERIALS SCIENCE; STAINLESS STEEL-304; STRESS CORROSION; CRACK PROPAGATION; INTERGRANULAR CORROSION; DRESDEN-2 REACTOR; WATER CHEMISTRY; REACTOR COOLING SYSTEMS; DUANE ARNOLD-1 REACTOR; FEEDWATER; CHEMICAL COMPOSITION; D CODES; HYDROGEN; OXYGEN; HYDROGEN PEROXIDE

Citation Formats

Yeh, T K, and Macdonald, D D. Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits. 2: Simulation of operating reactors. United States: N. p., 1996. Web.
Yeh, T K, & Macdonald, D D. Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits. 2: Simulation of operating reactors. United States.
Yeh, T K, and Macdonald, D D. Sat . "Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits. 2: Simulation of operating reactors". United States.
@article{osti_264347,
title = {Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits. 2: Simulation of operating reactors},
author = {Yeh, T K and Macdonald, D D},
abstractNote = {The DAMAGE-PREDICTOR computer code, which has the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential (ECP), and the crack growth rate (CGR) of a reference crack in sensitized Type 304 stainless steel, is used to evaluate the responses of the Dresden-2 and Duane Arnold boiling water reactors (BWRs) to hydrogen water chemistry (HWC). The HWC simulations for these two BWRs are carried out for feedwater hydrogen concentrations (H{sub 2}, O{sub 2}, H{sub 2}O{sub 2}, etc.), ECP, and CGR are predicted for various components in the heat transport circuits (HTCs) of the two reactors. It is found that while 1.3 ppm of feedwater hydrogen is needed to protect part of the lower downcomer, the recirculation system, and the lower plenum in Dresden-2 from intergranular stress corrosion cracking, only 0.3 ppm is needed to achieve the same goal in Duane Arnold. However, it is also found that the ECP in many regions (core channel, core bypass, upper plenum, downcomer, etc.) in the HTCs cannot be lowered to below the critical corrosion potential of {minus}0.23 V{sub SHE} for sensitized Type 304 stainless steels, even when [H{sub 2}]{sub FW} is as high as 2.0 ppm.},
doi = {},
url = {https://www.osti.gov/biblio/264347}, journal = {Nuclear Science and Engineering},
number = 2,
volume = 123,
place = {United States},
year = {1996},
month = {6}
}