Effect of Local Strain Distribution of Cold-Rolled Alloy 690 on Primary Water Stress Corrosion Crack Growth Behavior

• Autorzy: Kim S.-W. , Ahn T.-Y. , Lim Y.-S. , Hwang S.-S.

Identyfikatory

DOI

10.1515/amm-2017-0185

• Języki publikacji: EN

Abstrakty

EN

This work aims to study the stress corrosion crack growth behavior of cold-rolled Alloy 690 in the primary water of a pressurized water reactor. Compared with Alloy 600, which shows typical intergranular cracking along high angle grain boundaries, the cold-rolled Alloy 690, with its heterogeneous microstructure, revealed an abnormal crack growth behavior in mixed mode, that is, in transgranular cracking near a banded region, and in intergranular cracking in a matrix region. From local strain distribution analysis based on local mis-orientation, measured along the crack path using the electron back scattered diffraction method, it was suggested that the abnormal behavior was attributable to a heterogeneity of local strain distribution. In the cold-rolled Alloy 690, the stress corrosion crack grew through a highly strained area formed by a prior cold-rolling process in a direction perpendicular to the maximum principal stress applied during a subsequent stress corrosion cracking test.

Słowa kluczowe

EN

alloy 690; crack growth; electron backscatter diffraction; local strain distributions; stress corrosion cracking

• 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: 2017
• Tom: Vol. 62, iss. 2B
• Strony: 1243--1248
• Opis fizyczny: Bibliogr. 13 poz., rys.

Twórcy

autor

Kim S.-W.

• Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, Daedeok-Daero 989-111, Yuseong-Gu, Daejeon 34057, Korea

autor

Ahn T.-Y.

• Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, Daedeok-Daero 989-111, Yuseong-Gu, Daejeon 34057, Korea

autor

Lim Y.-S.

• Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, Daedeok-Daero 989-111, Yuseong-Gu, Daejeon 34057, Korea

autor

Hwang S.-S.

• Nuclear Materials Safety Research Division, Korea Atomic Energy Research Institute, Daedeok-Daero 989-111, Yuseong-Gu, Daejeon 34057, Korea

Bibliografia

• [1] P.M. Scott, in: D. Feron, J.M. Olive (Ed.), Corrosion issues in light water reactors – stress corrosion cracking, New York, Woodhead Publishing Ltd. (2007).
• [2] P.L. Andresen, M.M. Morra, A. Ahluwalia, J. Wilson, Effect of deformation and orientation on SCC of alloy 690, in: Proc. 14th Int. Conf. on Environmental Degradation Materials Nuclear Power Systems – Water Reactors, Virginia, USA (2009).
• [3] S.M. Bruemmer, M.J. Olszta, M.B. Toloczko, L.E. Thomas, Corrosion 69, 953 (2013).
• [4] P.L. Andresen, Unusual cold work and strain rate effects on SCC, in: Proc. 14th Int. Conf. on Environmental Degradation Materials Nuclear Power Systems – Water Reactors, Virginia, USA (2009).
• [5] S.W. Kim, S.S. Hwang, J.M. Lee, Corrosion 71, 1071 (2015).
• [6] ASTM Standard E 647, Standard test method for measurement of fatigue crack growth rates, West Conshohocken, PA, ASTM International (2005).
• [7] ASTM Standard E 1823, Standard terminology relating to fatigue and fracture testing, West Conshohocken, PA, ASTM International (2009).
• [8] J. Hou, Q.J. Peng, Z.P. Lu, T. Shoji, J.Q. Wang, E.-H. Han, W. Ke, Corros. Sci. 53, 1137 (2011).
• [9] Q.J. Peng, J. Hou, T. Yonezawa, T. Shoji, Z.M. Zhang, F. Huang, E.H. Han, W. Ke, Corros. Sci. 57, 81 (2012).
• [10] S.I. Wright, M.M. Mowell, D.P. Field, Microsc. Microanal. 17, 316 (2011).
• [11] G.E. Dieter, McGraw-Hill, Mechanical Metallurgy, SI metric ed., UK, 1988.
• [12] N. Hansen, Metal. Trans. A 16A, 2167 (1985).
• [13] Y.S. Lim, S.W. Kim, S.S. Hwang, H.P. Kim, C.H. Jang, Corros. Sci. 108, 125 (2016).

Uwagi

EN

Opracowanie ze środków MNiSW w ramach umowy 812/P-DUN/2016 na działalność upowszechniającą naukę (zadania 2017)