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Title: Modeling the development of damage in BWR primary coolant circuits

Abstract

Hydrogen water chemistry (HWC) has been explored as a remedial measure for inhibiting intergranular stress corrosion cracking (IGSCC), and for recently for mitigating irradiation assisted stress corrosion cracking (IASCC) in boiling water reactors over the past ten years. However, it is not clear if HWC can successfully protect all of the structural components in BWR primary heat transport circuits (HTCS) from IGSCC and LASCC. The authors have explored this issue using DAMAGE-PREDICTOR, which is a computer code that is capable of estimating the concentrations of radiolysis species, the electrochemical corrosion potential (ECP), and the growth rate of a reference crack in sensitized Type 304 stainless steel. This code was developed specifically for modeling the HTCs of BWRs. The primary objective of this code is to theoretically evaluate the effectiveness of HWC in BWRs as a function of feedwater hydrogen concentration and reactor power level. HWC simulations have been carried out for full power conditions for two reactors that differ markedly in their responses to HWC. It is found that DAMAGE-PREDICTOR can successfully account for plant data from both reactors using a single set of model parameter values.

Authors:
;  [1]
  1. Pennsylvania State Univ., University Park, PA (United States). Center for Advanced Materials
Publication Date:
OSTI Identifier:
203762
Report Number(s):
CONF-950816-
ISBN 1-877914-95-9; TRN: 96:009728
Resource Type:
Book
Resource Relation:
Conference: 7. international symposium on environmental degradation of materials in nuclear power plants: water reactors, Breckenridge, CO (United States), 6-10 Aug 1995; Other Information: PBD: 1995; Related Information: Is Part Of Seventh international symposium on environmental degradation of materials in nuclear power systems -- Water reactors: Proceedings and symposium discussions. Volume 2; Airey, G.; Andresen, P.; Brown, J. [eds.] [and others]; PB: 620 p.
Country of Publication:
United States
Language:
English
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 36 MATERIALS SCIENCE; BWR TYPE REACTORS; PRIMARY COOLANT CIRCUITS; WATER CHEMISTRY; STAINLESS STEEL-304; PHYSICAL RADIATION EFFECTS; STRESS CORROSION; CRACK PROPAGATION; INTERGRANULAR CORROSION; MITIGATION; MATHEMATICAL MODELS; D CODES; THEORETICAL DATA; COMPUTERIZED SIMULATION

Citation Formats

Yeh, T K, and Macdonald, D D. Modeling the development of damage in BWR primary coolant circuits. United States: N. p., 1995. Web.
Yeh, T K, & Macdonald, D D. Modeling the development of damage in BWR primary coolant circuits. United States.
Yeh, T K, and Macdonald, D D. 1995. "Modeling the development of damage in BWR primary coolant circuits". United States.
@article{osti_203762,
title = {Modeling the development of damage in BWR primary coolant circuits},
author = {Yeh, T K and Macdonald, D D},
abstractNote = {Hydrogen water chemistry (HWC) has been explored as a remedial measure for inhibiting intergranular stress corrosion cracking (IGSCC), and for recently for mitigating irradiation assisted stress corrosion cracking (IASCC) in boiling water reactors over the past ten years. However, it is not clear if HWC can successfully protect all of the structural components in BWR primary heat transport circuits (HTCS) from IGSCC and LASCC. The authors have explored this issue using DAMAGE-PREDICTOR, which is a computer code that is capable of estimating the concentrations of radiolysis species, the electrochemical corrosion potential (ECP), and the growth rate of a reference crack in sensitized Type 304 stainless steel. This code was developed specifically for modeling the HTCs of BWRs. The primary objective of this code is to theoretically evaluate the effectiveness of HWC in BWRs as a function of feedwater hydrogen concentration and reactor power level. HWC simulations have been carried out for full power conditions for two reactors that differ markedly in their responses to HWC. It is found that DAMAGE-PREDICTOR can successfully account for plant data from both reactors using a single set of model parameter values.},
doi = {},
url = {https://www.osti.gov/biblio/203762}, journal = {},
number = ,
volume = ,
place = {United States},
year = {Sun Dec 31 00:00:00 EST 1995},
month = {Sun Dec 31 00:00:00 EST 1995}
}

Book:
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