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0013011609
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Study on lead-induced stress corrosion cracking of steam generator tubing under AVT water chemistry conditions
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Amelia Island, Florida
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H. Takamatsu, T. Matsunaga, B. Miglin, J. Sarver, P. Sherburne, K. Aoki, T. Sakai, Study on lead-induced stress corrosion cracking of steam generator tubing under AVT water chemistry conditions, in: Proceedings of Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Amelia Island, Florida, 1997, p. 216
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Takamatsu, H.1
Matsunaga, T.2
Miglin, B.3
Sarver, J.4
Sherburne, P.5
Aoki, K.6
Sakai, T.7
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2
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18444418391
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Stress corrosion cracking aspects of nuclear steam generator tubing materials in the water containing lead at high temperature
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Amelia Island, Florida
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S.S. Hwang, K.M. Kim, U.C. Kim, Stress corrosion cracking aspects of nuclear steam generator tubing materials in the water containing lead at high temperature, in: Proceedings of Eighth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Amelia Island, Florida, 1997, p. 200
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Hwang, S.S.1
Kim, K.M.2
Kim, U.C.3
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3
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0033489132
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Lead-induced SCC propagation rates in Alloy 600
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Newport Beach, California
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M. Wright, M. Mirzai, Lead-induced SCC propagation rates in Alloy 600, in: Proceedings of Ninenth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Newport Beach, California, 1999, p. 657
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Wright, M.1
Mirzai, M.2
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4
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0033489404
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Lead assisted stress corrosion cracking of alloy 690
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Newport Beach, California
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M. Psaila-Dombrowski, F. Hua, P. Doherty, Lead assisted stress corrosion cracking of alloy 690, in: Proceedings of Ninth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Newport Beach, California, 1999, p.703
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Psaila-Dombrowski, M.1
Hua, F.2
Doherty, P.3
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5
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0041535940
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Effect of lead on the OD degradation of steam generator tubes
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Fontevraud, France
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A. Rocher, F. Cattant, D. Buisine, B. Prieux, M. Helie, Effect of lead on the OD degradation of steam generator tubes, in: Proceedings of International Symposium on Fontevraud III, Fontevraud, France, 1994, p. 538
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Rocher, A.1
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0027848377
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Lead assisted stress corrosion cracking of Alloy 600, 690 and 800
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San Diego, California
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M. Helie, Lead assisted stress corrosion cracking of Alloy 600, 690 and 800, in: Proceedings of Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, San Diego, California, 1993, p. 179
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(1993)
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Helie, M.1
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7
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0027885307
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Influence of lead contamination on the stress corrosion resistance of nickel alloys
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San Diego, California
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M.L. Castano-Marin, D. Gomez-Briceno, F. Hernandez-Arroyo, Influence of lead contamination on the stress corrosion resistance of nickel alloys, in: Proceedings of Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, San Diego, California, 1993, p. 189
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Castano-Marin, M.L.1
Gomez-Briceno, D.2
Hernandez-Arroyo, F.3
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8
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0013008676
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Some considerations about the possible mechanisms of lead assisted stress corrosion cracking of steam generator tubing
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Breckenridge, Colorado
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M. Helie, I. Lambert, G. Santarini, Some considerations about the possible mechanisms of lead assisted stress corrosion cracking of steam generator tubing, in: Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Breckenridge, Colorado, 1995, p. 247
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Helie, M.1
Lambert, I.2
Santarini, G.3
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9
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0006449744
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Comparative behaviour of Alloy 600, 690 and 800 in caustic environments
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Breckenridge, Colorado
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F. Vaillant, D. Buisine, B. Prieux, Comparative behaviour of Alloy 600, 690 and 800 in caustic environments, in: Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Breckenridge, Colorado, 1995, p. 219
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Vaillant, F.1
Buisine, D.2
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11
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0002028252
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Model boiler testing to evaluate inhibitors for caustic induced stress corrosion cracking of Alloy 600 tubes
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Breckenridge, Colorado
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J. Daret, J.P. Paine, M. Partridge, Model boiler testing to evaluate inhibitors for caustic induced stress corrosion cracking of Alloy 600 tubes, in: Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Breckenridge, Colorado, 1995, p. 177
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(1995)
Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors
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Daret, J.1
Paine, J.P.2
Partridge, M.3
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12
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0009056276
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Mechanism and effectiveness of inhibitors for SCC in a caustic environment
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Breckenridge, Colorado
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J. Lumsden, S. Jeanjaguet, J.P.N. Paine, A. Mcilree, Mechanism and effectiveness of inhibitors for SCC in a caustic environment, in: Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Breckenridge, Colorado, 1995, p. 317
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Lumsden, J.1
Jeanjaguet, S.2
Paine, J.P.N.3
McIlree, A.4
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13
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0008966537
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SCC of Alloy 600 in complex caustic environments
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Breckenridge, Colorado
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M. Miglin, J. Monter, C. Wade, M. Psaila-Dombrowski, A. Mcilree, SCC of Alloy 600 in complex caustic environments, in: Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, Breckenridge, Colorado, 1995, p. 277
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(1995)
Proceedings of Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors
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Miglin, M.1
Monter, J.2
Wade, C.3
Psaila-Dombrowski, M.4
McIlree, A.5
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14
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0027813674
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Slow strain rate testing to evaluate inhibitors for stress corrosion cracking of Alloy 600
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San Diego, California
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B. Miglin, J.P. Paine, Slow strain rate testing to evaluate inhibitors for stress corrosion cracking of Alloy 600, in: Proceedings of Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, San Diego, California, 1993, p. 303
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(1993)
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Miglin, B.1
Paine, J.P.2
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