ROLE OF GRAIN BOUNDARY CARBIDES IN CRACKING BEHAVIOR OF Ni BASE ALLOYS
Tài liệu tham khảo
IAEA Technical Document No. NP-T-3.13, Stress Corrosion Cracking in Light Water Reactors: Good Practices and Lessons Learned, p. 1, Vienna, Austria IAEA (2011).
2003, 1007832
NUREG 1823, U.S. Plant Experience With Alloy 600 Cracking and Boric Acid Corrosion of Light-Water Reactor Pressure Vessel Materials, U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research Washington, DC 20555-0001, (2005)
1994
S.S. Hwang, et. al., Failure Analysis on Steam Generator Drain Nozzle in Yonggwang Unit 4-Final Report, KAERI/CR-329/2009, (2009).
Bruemmer, 1988, Corrosion, 44, 782, 10.5006/1.3584948
Tsai, 1993, Corrosion science, 34, 741, 10.1016/0010-938X(93)90097-Z
Tsai, 1994, Corrosion, 50, 98, 10.5006/1.3293507
Tsai, 1996, Corrosion Science, 38, 33, 10.1016/0010-938X(96)00097-2
ASTM G38, ‘Standard Practice for Making and Using C-ring Stress-Corrosion Test Specimens', (1984).
E647-00, Standard Test Method for Measurement of Fatigue Crack Growth Rates, ASTM International, PO Box C700, West Conshohocken, PA 19428, (2000)
Hicks, 1982, A Comparison of Theoretical and Experimental Methods of Calibrating the Electrical Potential Drop Technique for Crack Length Determination, Int. J. of Fracture, 20, 91, 10.1007/BF01141259
Kawamura, 1999, Corrosion Engineering, 48, 99
H.Coriou, et. al. ‘Historical Review of the Principal Research Concerning the Phenomena of Cracking of Nickel Base Austenitic Alloys.’ Proceedings of Conference on Fundamental Aspects of Stress Corrosion Cracking and Hydrogen Embrittlement of Iron Base Alloys. Unieux-Firminy France, June 12-16, 1973, NACE-5.
1983
1988
Garriga Majo, D., et al. Prediction of the In-service behavior of alloy 600 tubes used in steam generators of pressurized water reactors, Colloque international, Fontevraud II, SFEN, Sept. 10-14, (1990).
Sarver, 1989, Corrosion, 44, 288, 10.5006/1.3583939
1994, 7
Hwang, 1999, of Nuclear Materials, 275, 28, 10.1016/S0022-3115(99)00111-7
Hwang, 2008, Corrosion Science and Technology, 7, 189
Z.Szkalska-Smialowska, Factors influencing IGSCC of alloy 600 in primary and secondary waters of PWR steam generators, 4th Symposium on environmental degradation of materials in nuclear power systems-Water Reactors, p.6-1Jekyll Island, Georgia, August 6-10, 1989, NACE, Houston, TX, (1990).
Rebak, 1996, The mechanism of stress corrosion cracking of alloy 600 in high temperature water, Corrosion Science, 38, 971, 10.1016/0010-938X(96)00183-7
