Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits - I: the damage-predictor algorithm

Tsung Kuang Yeh, Digby D. Macdonald, Arthur Thompson Motta

Research output: Contribution to journalArticle

41 Citations (Scopus)

Abstract

A computer code with the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential, and the kinetics of growth of a reference crack in sensitized Type 304 stainless steel is developed for the heat transport circuits of boiling water reactors (BWRs). The primary objective of this code, DAMAGE-PREDICTOR, is to theoretically evaluate the effectiveness of hydrogen water chemistry (HWC) in the BWRs as a function of feedwater hydrogen concentration and reactor power level. The power level determines various important thermal-hydraulic parameters and the neutron and gamma energy deposition rate in the core and near-core regions. These input parameters are estimated using well-established algorithms, and the simulations are carried out for full-power conditions for two reactors that differ markedly in their responses to HWC. The DAMAGE-PREDICTOR code is found to successfully account for plant data from both reactors using a single set of model parameter values.

Original languageEnglish (US)
Pages (from-to)468-482
Number of pages15
JournalNuclear Science and Engineering
Volume121
Issue number3
DOIs
StatePublished - Jan 1 1995

Fingerprint

Electrochemical corrosion
Boiling water reactors
Crack propagation
Hydrogen
Networks (circuits)
Water
Radiolysis
Deposition rates
Neutrons
Stainless steel
Hydraulics
Cracks
Kinetics
Hot Temperature

All Science Journal Classification (ASJC) codes

  • Nuclear Energy and Engineering

Cite this

@article{737c7fc0a2be461ca96107ba70d4aedc,
title = "Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits - I: the damage-predictor algorithm",
abstract = "A computer code with the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential, and the kinetics of growth of a reference crack in sensitized Type 304 stainless steel is developed for the heat transport circuits of boiling water reactors (BWRs). The primary objective of this code, DAMAGE-PREDICTOR, is to theoretically evaluate the effectiveness of hydrogen water chemistry (HWC) in the BWRs as a function of feedwater hydrogen concentration and reactor power level. The power level determines various important thermal-hydraulic parameters and the neutron and gamma energy deposition rate in the core and near-core regions. These input parameters are estimated using well-established algorithms, and the simulations are carried out for full-power conditions for two reactors that differ markedly in their responses to HWC. The DAMAGE-PREDICTOR code is found to successfully account for plant data from both reactors using a single set of model parameter values.",
author = "Yeh, {Tsung Kuang} and Macdonald, {Digby D.} and Motta, {Arthur Thompson}",
year = "1995",
month = "1",
day = "1",
doi = "10.13182/NSE95-A24148",
language = "English (US)",
volume = "121",
pages = "468--482",
journal = "Nuclear Science and Engineering",
issn = "0029-5639",
publisher = "American Nuclear Society",
number = "3",

}

TY - JOUR

T1 - Modeling water chemistry, electrochemical corrosion potential, and crack growth rate in the boiling water reactor heat transport circuits - I

T2 - the damage-predictor algorithm

AU - Yeh, Tsung Kuang

AU - Macdonald, Digby D.

AU - Motta, Arthur Thompson

PY - 1995/1/1

Y1 - 1995/1/1

N2 - A computer code with the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential, and the kinetics of growth of a reference crack in sensitized Type 304 stainless steel is developed for the heat transport circuits of boiling water reactors (BWRs). The primary objective of this code, DAMAGE-PREDICTOR, is to theoretically evaluate the effectiveness of hydrogen water chemistry (HWC) in the BWRs as a function of feedwater hydrogen concentration and reactor power level. The power level determines various important thermal-hydraulic parameters and the neutron and gamma energy deposition rate in the core and near-core regions. These input parameters are estimated using well-established algorithms, and the simulations are carried out for full-power conditions for two reactors that differ markedly in their responses to HWC. The DAMAGE-PREDICTOR code is found to successfully account for plant data from both reactors using a single set of model parameter values.

AB - A computer code with the capability of simultaneously estimating the concentrations of radiolysis species, the electrochemical corrosion potential, and the kinetics of growth of a reference crack in sensitized Type 304 stainless steel is developed for the heat transport circuits of boiling water reactors (BWRs). The primary objective of this code, DAMAGE-PREDICTOR, is to theoretically evaluate the effectiveness of hydrogen water chemistry (HWC) in the BWRs as a function of feedwater hydrogen concentration and reactor power level. The power level determines various important thermal-hydraulic parameters and the neutron and gamma energy deposition rate in the core and near-core regions. These input parameters are estimated using well-established algorithms, and the simulations are carried out for full-power conditions for two reactors that differ markedly in their responses to HWC. The DAMAGE-PREDICTOR code is found to successfully account for plant data from both reactors using a single set of model parameter values.

UR - http://www.scopus.com/inward/record.url?scp=0029406351&partnerID=8YFLogxK

UR - http://www.scopus.com/inward/citedby.url?scp=0029406351&partnerID=8YFLogxK

U2 - 10.13182/NSE95-A24148

DO - 10.13182/NSE95-A24148

M3 - Article

AN - SCOPUS:0029406351

VL - 121

SP - 468

EP - 482

JO - Nuclear Science and Engineering

JF - Nuclear Science and Engineering

SN - 0029-5639

IS - 3

ER -