Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors

Understanding the Limits of Lattice Orientation Data Analysis in Environmental Degradation Studies

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_140 ◽

2016 ◽

pp. 2321-2333

Author(s):

Fabio Di Gioacchino ◽

João Quinta da Fonseca ◽

David Wright ◽

Fabio Scenini

Keyword(s):

Data Analysis ◽

Environmental Degradation ◽

Lattice Orientation ◽

Orientation Data ◽

Degradation Studies

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Evaluation of the Oxygen Diffusion Coefficient in Nickel-Base Alloys

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_90 ◽

2011 ◽

pp. 1463-1475

Author(s):

Hyo On Nam ◽

Jae Young Yoon ◽

Ji Hyun Kim ◽

Il Soon Hwang ◽

Kyu Hwan Lee

Keyword(s):

Diffusion Coefficient ◽

Oxygen Diffusion ◽

Oxygen Diffusion Coefficient ◽

Nickel Base

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SCC Crack Growth Rate of Alloy 82 in PWR Primary Water Conditions — Effect of a Thermal Treatment

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_68 ◽

2011 ◽

pp. 1127-1139 ◽

Cited By ~ 4

Author(s):

C. Guerre ◽

C. Duhamel ◽

M. Sennour ◽

J. Crépin ◽

M. Le Calvar

Keyword(s):

Growth Rate ◽

Crack Growth Rate ◽

Thermal Treatment ◽

Crack Growth ◽

Water Conditions ◽

Primary Water

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On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes and on Laboratory Specimens

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_61 ◽

2011 ◽

pp. 1007-1021

Author(s):

L. Legras ◽

O. de Bouvier ◽

F. Delabrouille ◽

E. Fargeas ◽

S. Miloudi ◽

...

Keyword(s):

Alloy 600

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Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_6 ◽

2011 ◽

pp. 71-89 ◽

Cited By ~ 7

Author(s):

David R. Tice ◽

Stuart L. Medway ◽

Norman Platts ◽

John W. Stairmand

Keyword(s):

Crack Growth ◽

Water Environment ◽

Alloy 690 ◽

Primary Water ◽

Cold Worked

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Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_54 ◽

2011 ◽

pp. 883-905 ◽

Cited By ~ 3

Author(s):

T. Lucas ◽

R. G. Ballinger ◽

H. Hanninen ◽

T. Saukkonen

Keyword(s):

Fracture Toughness ◽

Stainless Steel ◽

Thermal Aging ◽

Material Properties ◽

Steel Weld ◽

Weld Metals ◽

Stainless Steel Weld

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Stress Corrosion Cracking Behavior Near the Fusion Boundary of Dissimilar Weld Joint with Alloy 182-A533B Low Alloy Steel

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_50 ◽

2011 ◽

pp. 791-804

Author(s):

Hiroshi Abe ◽

Makoto Ishizawa ◽

Yutaka Watanabe

Keyword(s):

Stress Corrosion ◽

Alloy Steel ◽

Stress Corrosion Cracking ◽

Weld Joint ◽

Corrosion Cracking ◽

Low Alloy Steel ◽

Fusion Boundary ◽

Cracking Behavior ◽

Dissimilar Weld

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Microstructure and SCC of Alloy X-750

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_46 ◽

2011 ◽

pp. 719-744 ◽

Cited By ~ 2

Author(s):

Peter L. Andresen ◽

Juan Flores-Preciado ◽

Martin M. Morra ◽

Robert Carter

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Hydride Behavior in Zircaloy-4 During Thermomechanical Cycling

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_43 ◽

2011 ◽

pp. 683-688

Author(s):

Kimberly Colas ◽

Arthur Motta ◽

Mark R. Daymond ◽

Jonathan Almer ◽

Zhonghou Cai

Keyword(s):

Thermomechanical Cycling

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Effect of DH Concentration on Crud Deposition on Heated Zircaloy-4 in Simulated PWR Primary Water

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors ◽

10.1007/978-3-319-48760-1_38 ◽

2011 ◽

pp. 633-648

Author(s):

Hirotaka Kawamura

Keyword(s):

Primary Water

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Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors
Latest Publications

TOTAL DOCUMENTS

H-INDEX

Published By Springer International Publishing

Understanding the Limits of Lattice Orientation Data Analysis in Environmental Degradation Studies

Evaluation of the Oxygen Diffusion Coefficient in Nickel-Base Alloys

SCC Crack Growth Rate of Alloy 82 in PWR Primary Water Conditions — Effect of a Thermal Treatment

On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes and on Laboratory Specimens

Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals

Stress Corrosion Cracking Behavior Near the Fusion Boundary of Dissimilar Weld Joint with Alloy 182-A533B Low Alloy Steel

Microstructure and SCC of Alloy X-750

Hydride Behavior in Zircaloy-4 During Thermomechanical Cycling

Effect of DH Concentration on Crud Deposition on Heated Zircaloy-4 in Simulated PWR Primary Water

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Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water ReactorsLatest Publications

TOTAL DOCUMENTS

H-INDEX

Published By Springer International Publishing

Understanding the Limits of Lattice Orientation Data Analysis in Environmental Degradation Studies

Evaluation of the Oxygen Diffusion Coefficient in Nickel-Base Alloys

SCC Crack Growth Rate of Alloy 82 in PWR Primary Water Conditions — Effect of a Thermal Treatment

On the Microstructure of Alloy 600 SCC Cracks Observed by TEM on PWR SG Pulled Tubes and on Laboratory Specimens

Crack growth testing on Cold Worked Alloy 690 in Primary Water Environment

Effect of Thermal Aging on SCC, Material Properties and Fracture Toughness of Stainless Steel Weld Metals

Stress Corrosion Cracking Behavior Near the Fusion Boundary of Dissimilar Weld Joint with Alloy 182-A533B Low Alloy Steel

Microstructure and SCC of Alloy X-750

Hydride Behavior in Zircaloy-4 During Thermomechanical Cycling

Effect of DH Concentration on Crud Deposition on Heated Zircaloy-4 in Simulated PWR Primary Water

Proceedings of the 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems — Water Reactors
Latest Publications