Lifetime estimation of a BWR core shroud in terms of IGSCC

Balachov, Iouri; Fekete, Balazs; Macdonald, Digby D.; Spencer, Benjamin

doi:10.1016/j.nucengdes.2020.110831

Lifetime estimation of a BWR core shroud in terms of IGSCC

Journal Article · Thu Sep 10 00:00:00 EDT 2020 · Nuclear Engineering and Design

DOI:https://doi.org/10.1016/j.nucengdes.2020.110831· OSTI ID:1668311

Balachov, Iouri ^[1]; Fekete, Balazs ^[2]; Macdonald, Digby D. ^[2]; ^[3]

4D Power, LLC, Menlo Park, CA (United States)
Univ. of California, Berkeley, CA (United States)
Idaho National Lab. (INL), Idaho Falls, ID (United States)

The continued operation of aging Boiling Water Reactors (BWRs) worldwide requires gradually increasing cost of inspection, maintenance, and repair. Intergranular Stress Corrosion Cracking (IGSCC) in sensitized austenitic stainless steel piping first became a major issue for BWRs in the 1980s, resulting in recognition of the susceptibility of reactor internals to IGSCC. Shroud cracking identified in 1993–1994 confirmed that IGSCC of internals is a significant issue for BWRs. IGSCC is a time-dependent, material degradation process, which is caused and accelerated by the presence of residual stresses, material sensitization, irradiation, cold work, elevated temperature, and corrosive environments. This paper emphasizes the importance of accounting for corrosive environments, or more exactly, electrochemical phenomena in modelling IGSCC and predicting the service life of BWR in-vessel components, and stresses the necessity of performing such modeling not just for a single state point under full power conditions, but for the whole operating history of the reactor, including startups and shutdowns. Overall, this paper demonstrates that ignoring electrochemical considerations may result in underestimating component lifetime and lead to unnecessary expenses for inspection and repair.

View Accepted Manuscript (DOE)

Research Organization:: Idaho National Laboratory (INL), Idaho Falls, ID (United States)

Sponsoring Organization:: USDOE; USDOE Office of Nuclear Energy (NE)

Grant/Contract Number:: AC07-05ID14517; NE0008541

OSTI ID:: 1668311

Alternate ID(s):: OSTI ID: 1659342

Report Number(s):: INL/JOU--19-52431-Rev000

Journal Information:: Nuclear Engineering and Design, Journal Name: Nuclear Engineering and Design Journal Issue: n/a Vol. 368; ISSN 0029-5493

Publisher:: ElsevierCopyright Statement

Country of Publication:: United States

Language:: English

References (14)

A coupled environment model for stress corrosion cracking in sensitized type 304 stainless steel in LWR environments MacDonald, Digby D.; Urquidi-MacDonald, Mirna Corrosion Science, Vol. 32, Issue 1 https://doi.org/10.1016/0010-938X(91)90063-U	journal	January 1991
Prediction of crack growth rate in Type 304 stainless steel using artificial neural networks and the coupled environment fracture model Shi, Jiangbo; Wang, Jihui; Macdonald, Digby D. Corrosion Science, Vol. 89 https://doi.org/10.1016/j.corsci.2014.08.011	journal	December 2014
Modular system for probabilistic fracture mechanics analysis of embrittled reactor pressure vessels in the Grizzly code Spencer, B. W.; Hoffman, W. M.; Backman, M. A. Nuclear Engineering and Design, Vol. 341 https://doi.org/10.1016/j.nucengdes.2018.10.015	journal	January 2019
An advanced coupled environment fracture model for hydrogen-induced cracking in alloy 600 in PWR primary heat transport environment Fekete, Balazs; Ai, Jiahe; Yang, Jie Theoretical and Applied Fracture Mechanics, Vol. 95 https://doi.org/10.1016/j.tafmec.2018.02.013	journal	June 2018
Remaining life prediction of the core shroud due to stress corrosion cracking failure in BWRs using numerical simulations Li, Yongkui; Lu, Shanping; Li, Dianzhong Journal of Nuclear Science and Technology, Vol. 52, Issue 1 https://doi.org/10.1080/00223131.2014.930703	journal	June 2014
Effect of Loading Rate on Environmentally Controlled Cracking of Sensitized 304 Stainless Steel in High Purity Water Ford, F. P.; Silverman, M. CORROSION, Vol. 36, Issue 11 https://doi.org/10.5006/0010-9312-36.11.597	journal	November 1980
Viability of Hydrogen Water Chemistry for Protecting In-Vessel Components of Boiling Water Reactors Macdonald, D. D. CORROSION, Vol. 48, Issue 3 https://doi.org/10.5006/1.3315925	journal	March 1992
Effect of Water Impurities on Stress Corrosion Cracking in a Boiling Water Reactor Ljungberg, Lars G.; Cubicciotti, Daniel CORROSION, Vol. 41, Issue 5 https://doi.org/10.5006/1.3582006	journal	May 1985
Stress Corrosion Cracking of Sensitized AISI 304 Stainless Steel in Oxygenated High Temperature Chloride Solutions Containing Cupric (Cu ²⁺ ) and Lead (Pb ²⁺ ) Ions Macdonald, D. D.; Eghan, A. K.; Szklarska-Smialowska, Z. CORROSION, Vol. 41, Issue 8 https://doi.org/10.5006/1.3583830	journal	August 1985
Effects of Impurities on the IGSCC of Stainless Steel in High-Temperature Water Ljungberg, L. G.; Cubicciotti, D.; Trolle, M. CORROSION, Vol. 44, Issue 2 https://doi.org/10.5006/1.3583909	journal	February 1988
Apparatus for Controlled Hydrodynamic Electrochemical and Corrosion Studies in High-Temperature Aqueous Systems Macdonald, D. D.; Mankowski, J.; Karaminezhaad-Ranjbar, M. CORROSION, Vol. 44, Issue 3 https://doi.org/10.5006/1.3583923	journal	March 1988
Thin-Layer Mixed-Potential Model for the Corrosion of High-Level Nuclear Waste Canisters Macdonald, D. D.; Urquidi-Macdonald, M. CORROSION, Vol. 46, Issue 5 https://doi.org/10.5006/1.3585122	journal	May 1990
Effects of Impurities and Supporting Electrolytes On SCC of 304 Stainless Steel in High Temperature Aqueous Environments Andresen, Peter L.; Indig, Maurice E. CORROSION, Vol. 38, Issue 10 https://doi.org/10.5006/1.3593856	journal	October 1982
Effect of Sulfuric Acid, Oxygen, and Hydrogen in High Temperature Water on Stress Corrosion Cracking of Sensitized AISI 304 Stainless Steel Ruther, W. E.; Soppet, W. K.; Ayrault, G. CORROSION, Vol. 40, Issue 10 https://doi.org/10.5006/1.3593887	journal	October 1984

Similar Records

Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

Technical Report · Thu Feb 29 23:00:00 EST 1996 · OSTI ID:236061

Use of an IGSCC damage index for cost benefit evaluation of BWR pipe cracking remedies

Conference · Fri Jun 01 00:00:00 EDT 1984 · Am. Soc. Mech. Eng., (Pap.); (United States) · OSTI ID:6280559

An integrated approach for the evaluation of BWR vessel internal corrosion

Conference · Thu Dec 31 23:00:00 EST 1992 · Transactions of the American Nuclear Society; (United States) · OSTI ID:6960055

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
42 ENGINEERING
BWR core shroud
Boiling Water Reactor
Core shroud
Deterministic model
IGSCC
Stress corrosion cracking
Water chemistry
fracture

Lifetime estimation of a BWR core shroud in terms of IGSCC

Citation Formats

References (14)

Similar Records

Related Subjects