Stress Corrosion Cracking Properties of Steam Generator Tubing Alloys in Crevice Environment

• Autorzy: Shin Jung-Ho , Kim Dong-Jin

Identyfikatory

DOI

10.24425/amm.2019.127575

• Języki publikacji: EN

Abstrakty

EN

The safe and reliable operation of pressurized water reactors (PWRs) depends on the integrity of structural material. In particular, the failure of steam generator (SG) tubes on the secondary side is one of the major concerns of operating nuclear power plants. To establish remediation techniques and manage damage, it is necessary to articulate the mechanism through which various impurities affect the SG tubes. This research aims to understand the effect of impurities (e. g., S, Pb, and Cl) on the stress corrosion cracking of Alloy 600 and 690.

Słowa kluczowe

EN

stress corrosion; nickel base alloy; nuclear plants; secondary side; passive film

• Wydawca: Institute of Metallurgy and Materials Science of Polish Academy of Sciences ; Committee of Materials Engineering and Metallurgy of Polish Academy of Sciences
• Czasopismo: Archives of Metallurgy and Materials
• Rocznik: 2019
• Tom: Vol. 64, iss. 2
• Strony: 543--545
• Opis fizyczny: Bibliogr. 9 poz., rys., tab.

Twórcy

autor

Shin Jung-Ho

• Korea Atomic Energy Research Institute, Nuclear Materials Research Division, (989-111 Daedeok-Daero, Yuseong-Gu,Daejeon 34057, Republic of Korea)

autor

Kim Dong-Jin

• Korea Atomic Energy Research Institute, Nuclear Materials Research Division, (989-111 Daedeok-Daero, Yuseong-Gu,Daejeon 34057, Republic of Korea)

Bibliografia

• [1] Guideline for Tube Selection Removal and Examination, EPRI Draft Report (1997).
• [2] D. Gómez-Briceno, M. L. Castano, M. S. Carcía, Nucl. Eng. Des.165, 161-169 (1996).
• [3] R. S. Dutta, J. Nucl. Mater. 393, 343-349 (2009).
• [4] R. L. Tapping, Steam generator aging in CANDUs: 30 years of operation and R&D, presented at the in 5th CNS International Steam Generator Conference, Toronto, ON, Canada, (2006).
• [5] R. W. Staehle, J. A. Gorman, Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors: Part 1. Corrosion 59 (11), 931-994 (2003).
• [6] C. O. Ruud, D. J. Snoha, A. R. Mcllree, Experimental mechanics, p. 54. (1989)
• [7] Effect of Alumino Silicate on the Stress Corrosion Cracking of Alloy 600 and 690, EPRI, Palo Alto, Ca: 1002782, 2002.
• [8] M. G. Burke, R. E. Hermer, M. W. Phaneuf, Microstructural Analysis of Lead-Induced Transgranular SCC of Alloy 690 in PbO+ 10% NaOH Solution, Microscopy and Microanalysis 14, 2, 644-645 (2008).
• [9] L. Lindsay et al., Characterisation of stress corrosion cracking and internal oxidation of alloy 600 in high temperature hydrogenated steam, Proc. 16th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Houston, TX: NACE, 2013.

Uwagi

EN

1. This work was financially supported by the Korean Nuclear R&D Program organized by the National Research Foundation (NRF) in support of the Ministry of Science and ICT (2017M2A8A4015155), and by the R&D Program of Korea Atomic Energy Research Institute (KAERI).

PL

2. Opracowanie rekordu w ramach umowy 509/P-DUN/2018 ze środków MNiSW przeznaczonych na działalność upowszechniającą naukę (2019).