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Proceedings Papers

Coupling of Neutron Transport and Probabilistic Fracture Mechanics Codes for Analysis of Embrittled Reactor Pressure Vessels

Benjamin W. Spencer, William M. Hoffman, Benjamin S. Collins, Shane C. Henderson

Proc. ASME. PVP2020, Volume 3: Design and Analysis, V003T03A057, August 3, 2020

Paper No: PVP2020-21680

DOI: https://doi.org/10.1115/PVP2020-21680

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Topics: Embrittlement, Fluence (Radiation measurement), Fracture mechanics, Neutrons, Reactor vessels

Proceedings Papers

Non-Destructive Evaluation of RPV Embrittlement by Means of the Thermoelectric Power Method

M. Niffenegger, D. F. Mora, H. Kottmann

Proc. ASME. PVP2020, Volume 7: Non-Destructive Examination, V007T07A002, August 3, 2020

Paper No: PVP2020-21446

DOI: https://doi.org/10.1115/PVP2020-21446

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Topics: Embrittlement, Nondestructive evaluation, Steel

Proceedings Papers

Regression Analysis of Transition Temperature Shift Database for the Core Region Beltline Welds of WWER-1000 RPVs

Maksym Zarazovskii, Volodymyr Revka, Liudmyla Chyrko

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A018, August 3, 2020

Paper No: PVP2020-21748

DOI: https://doi.org/10.1115/PVP2020-21748

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Topics: Brittleness, Databases, Embrittlement, Phase transition temperature, Radiation (Physics), Regression analysis, Surveillance, Temperature, Welded joints, Silicon

Proceedings Papers

A Code Case Concerning the Effect of Embrittlement on Index Temperature Metrics

Mark Kirk, Marjorie Erickson

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A056, August 3, 2020

Paper No: PVP2020-21185

DOI: https://doi.org/10.1115/PVP2020-21185

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Topics: Embrittlement, Temperature, Accounting, Fracture toughness, Nuclear power stations, Safety, Uncertainty

Proceedings Papers

Extension of PASCAL4 Code for Probabilistic Fracture Mechanics Analysis of Reactor Pressure Vessel in Boiling Water Reactor

Kai Lu, Jinya Katsuyama, Yinsheng Li

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A062, August 3, 2020

Paper No: PVP2020-21421

DOI: https://doi.org/10.1115/PVP2020-21421

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Topics: Boiling water reactors, Fracture mechanics, Reactor vessels, Transients (Dynamics), Fracture (Materials), Pressurized water reactors, Surface cracks, Computational methods, Embrittlement, Irradiation (Radiation exposure)

Proceedings Papers

Bayesian Uncertainty Evaluation of Charpy Ductile-to-Brittle Transition Temperature for Reactor Pressure Vessel Steels

Hisashi Takamizawa, Yutaka Nishiyama, Takashi Hirano

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A065, August 3, 2020

Paper No: PVP2020-21698

DOI: https://doi.org/10.1115/PVP2020-21698

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Topics: Brittleness, Embrittlement, Irradiation (Radiation exposure), Notch testing, Phase transition temperature, Reactor vessels, Statistics, Steel, Uncertainty

Proceedings Papers

Cracking Behavior of a Decommissioned Material in Light Water Reactor Environment

Y. Chen, B. Alexandreanu, A. S. Rao

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A071, August 3, 2020

Paper No: PVP2020-21141

DOI: https://doi.org/10.1115/PVP2020-21141

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Topics: Embrittlement, Fracture (Materials), Fracture (Process), Fracture toughness, Irradiation (Radiation exposure), Light water reactors, Stress corrosion cracking

Proceedings Papers

Fracture Toughness Tests on Western RPV Steels Using Small Scale Specimen Technique

Florian Obermeier, Eberhard Altstadt, Hieronymus Hein, Aniruddh Das

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A102, August 3, 2020

Paper No: PVP2020-21481

DOI: https://doi.org/10.1115/PVP2020-21481

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Topics: Embrittlement, Fracture mechanics, Fracture toughness, Irradiation (Radiation exposure), Reactor vessels, Steel, Surveillance, Displacement, Brittle fracture, Cracking (Materials)

Proceedings Papers

Effect of Hydrogen Content in Natural Gas Blend on the Mechanical Properties of a L485-MB Low-Alloyed Steel

Hamza Ez-Zaki, Frederic Christien, Cédric Bosch, Laurent Briottet, Maxime Bertin, Olivier Levasseur, Alain Leriverain

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A085, August 3, 2020

Paper No: PVP2020-21228

DOI: https://doi.org/10.1115/PVP2020-21228

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Topics: Alloy steel, Hydrogen, Mechanical properties, Natural gas, Steel, Pressure, Fracture (Materials), Fracture (Process), Testing, Ductility

Proceedings Papers

Application of Probabilistic Fracture Mechanics Methodology for Japanese Reactor Pressure Vessels Using PASCAL4

Kai Lu, Jinya Katsuyama, Yinsheng Li, Shinobu Yoshimura

Proc. ASME. PVP2019, Volume 1: Codes and Standards, V001T01A057, July 14–19, 2019

Paper No: PVP2019-93935

DOI: https://doi.org/10.1115/PVP2019-93935

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Topics: Fracture mechanics, Reactor vessels, Failure, Neutrons, Embrittlement, Fluence (Radiation measurement), Irradiation (Radiation exposure), Light water reactors, Nondestructive evaluation, Nuclear power

Proceedings Papers

Recent Improvements in Toughness Prediction of Cast Duplex Stainless Steel Components

Patrick Le Delliou, Sébastien Saillet

Proc. ASME. PVP2019, Volume 1: Codes and Standards, V001T01A050, July 14–19, 2019

Paper No: PVP2019-93114

DOI: https://doi.org/10.1115/PVP2019-93114

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Topics: Fracture toughness, Stainless steel, Notch testing, Steel, Temperature, Casting, Cavities, Crack propagation, Embrittlement, Ferrites (Magnetic materials)

Proceedings Papers

RPV Irradiation Surveillance Programmes in China

Shuo Zhang, Haisheng Zhang, Kai Sun

Proc. ASME. PVP2019, Volume 6A: Materials and Fabrication, V06AT06A025, July 14–19, 2019

Paper No: PVP2019-93068

DOI: https://doi.org/10.1115/PVP2019-93068

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Topics: China, Irradiation (Radiation exposure), Surveillance, Pressurized water reactors, Safety, Nuclear power, Nuclear power stations, Construction, Design, Embrittlement

Proceedings Papers

Finite Element Verification of Engineering Equations for Prediction of Structural Strength of Annulus Spacers in CANDU Nuclear Reactors

Preeti Doddihal, Dennis Kawa, Douglas Scarth, Yu Chen

Proc. ASME. PVP2019, Volume 6A: Materials and Fabrication, V06AT06A013, July 14–19, 2019

Paper No: PVP2019-93671

DOI: https://doi.org/10.1115/PVP2019-93671

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Topics: Annulus, Finite element analysis, Nuclear reactors, Fracture (Materials), Fracture (Process), Strips, Fuels, Irradiation (Radiation exposure), Pressure, Stress

Proceedings Papers

Evaluation of Irradiation Embrittlement of the Chinese RPV Steels

Yupeng Cao, Yinbiao He, Binxi Wang, Yifeng Huang, Hui Li, Yan Yu, Hu Hui

Proc. ASME. PVP2019, Volume 6A: Materials and Fabrication, V06AT06A028, July 14–19, 2019

Paper No: PVP2019-93615

DOI: https://doi.org/10.1115/PVP2019-93615

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Topics: Embrittlement, Irradiation (Radiation exposure), Steel, Forging, China, Fluence (Radiation measurement), Fracture toughness, Mechanical properties, Notch testing, Nuclear power stations

Proceedings Papers

Susceptibility to Neutron Irradiation Embrittlement of Heat-Affected Zone of Reactor Pressure Vessel Steels

Hisashi Takamizawa, Jinya Katsuyama, Yoosung Ha, Tohru Tobita, Yutaka Nishiyama, Kunio Onizawa

Proc. ASME. PVP2019, Volume 6B: Materials and Fabrication, V06BT06A022, July 14–19, 2019

Paper No: PVP2019-94011

DOI: https://doi.org/10.1115/PVP2019-94011

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Topics: Embrittlement, Heat, Irradiation (Radiation exposure), Neutrons, Reactor vessels, Steel

Proceedings Papers

Consideration of Thermal Embrittlement in Alloy 316H for Advanced Non-Light Water Reactor Applications

Weiju Ren, Lianshan Lin

Proc. ASME. PVP2019, Volume 6B: Materials and Fabrication, V06BT06A016, July 14–19, 2019

Paper No: PVP2019-93431

DOI: https://doi.org/10.1115/PVP2019-93431

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Topics: Embrittlement, Indium alloys, Water, Alloys, Light water reactors, Temperature, Mechanical engineers, Mechanical properties, Safety

Proceedings Papers

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using a Newly Developed Statistical Thermal Aging Model

M. Uddin, C. Sallaberry, G. Wilkowski

Proc. ASME. PVP2019, Volume 1: Codes and Standards, V001T01A051, July 14–19, 2019

Paper No: PVP2019-93711

DOI: https://doi.org/10.1115/PVP2019-93711

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Topics: Stainless steel, Fracture toughness, Pipes, Stress, Pressurized water reactors, Casting, Ferrites (Magnetic materials), Databases, Dimensions, Embrittlement

Proceedings Papers

Verification of Probabilistic Fracture Mechanics Analysis Code Through Benchmark Analyses

Yinsheng Li, Shumpei Uno, Koichi Masaki, Jinya Katsuyama, Terry Dickson, Mark Kirk

Proc. ASME. PVP2018, Volume 1B: Codes and Standards, V01BT01A075, July 15–20, 2018

Paper No: PVP2018-84963

DOI: https://doi.org/10.1115/PVP2018-84963

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Topics: Fracture mechanics, Failure, Probability, Embrittlement, Irradiation (Radiation exposure), Neutrons, Nuclear power, Reactor vessels, Thermal shock

Proceedings Papers

Failure Analysis of Furnace Tube in the Radiation Section of Ethylene Cracking Furnace

Wen Liu, GuoDong Sun, Jian Xing, Zhenlong Hu, Luowei Cao, Fakun Zhuang

Proc. ASME. PVP2018, Volume 3B: Design and Analysis, V03BT03A009, July 15–20, 2018

Paper No: PVP2018-84037

DOI: https://doi.org/10.1115/PVP2018-84037

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Topics: Failure analysis, Fracture (Process), Furnaces, Radiation (Physics), Carburization, Pipes, Failure, Brittle fracture, Embrittlement, Fracture (Materials)

Proceedings Papers

Evaluation of Reactor Internals Integrity and Lifetime According to the NTD ASI

Milan Brumovsky

Proc. ASME. PVP2018, Volume 1A: Codes and Standards, V01AT01A056, July 15–20, 2018

Paper No: PVP2018-84140

DOI: https://doi.org/10.1115/PVP2018-84140

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Topics: Control systems, Cracking (Materials), Creep, Design, Dimensions, Embrittlement, Failure, Ferrites (Magnetic materials), Fracture (Process), Fracture mechanics

2 3 4 5

NARROW

Coupling of Neutron Transport and Probabilistic Fracture Mechanics Codes for Analysis of Embrittled Reactor Pressure Vessels

Non-Destructive Evaluation of RPV Embrittlement by Means of the Thermoelectric Power Method

Regression Analysis of Transition Temperature Shift Database for the Core Region Beltline Welds of WWER-1000 RPVs

A Code Case Concerning the Effect of Embrittlement on Index Temperature Metrics

Extension of PASCAL4 Code for Probabilistic Fracture Mechanics Analysis of Reactor Pressure Vessel in Boiling Water Reactor

Bayesian Uncertainty Evaluation of Charpy Ductile-to-Brittle Transition Temperature for Reactor Pressure Vessel Steels

Cracking Behavior of a Decommissioned Material in Light Water Reactor Environment

Fracture Toughness Tests on Western RPV Steels Using Small Scale Specimen Technique

Effect of Hydrogen Content in Natural Gas Blend on the Mechanical Properties of a L485-MB Low-Alloyed Steel

Application of Probabilistic Fracture Mechanics Methodology for Japanese Reactor Pressure Vessels Using PASCAL4

Recent Improvements in Toughness Prediction of Cast Duplex Stainless Steel Components

RPV Irradiation Surveillance Programmes in China

Finite Element Verification of Engineering Equations for Prediction of Structural Strength of Annulus Spacers in CANDU Nuclear Reactors

Evaluation of Irradiation Embrittlement of the Chinese RPV Steels

Susceptibility to Neutron Irradiation Embrittlement of Heat-Affected Zone of Reactor Pressure Vessel Steels

Consideration of Thermal Embrittlement in Alloy 316H for Advanced Non-Light Water Reactor Applications

Flaw Evaluation Procedure for Cast Austenitic Stainless Steel Materials Using a Newly Developed Statistical Thermal Aging Model

Verification of Probabilistic Fracture Mechanics Analysis Code Through Benchmark Analyses

Failure Analysis of Furnace Tube in the Radiation Section of Ethylene Cracking Furnace

Evaluation of Reactor Internals Integrity and Lifetime According to the NTD ASI

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