Search Results | PVP | ASME Digital Collection

1-20 of 120

2 3 4 5

Sort by

Sort Order Select

Proceedings Papers

NuScale Pressure and Temperature Limits Methodology Using Finite Element Analysis

Proc. ASME. PVP2021, Volume 2: Computer Technology and Bolted Joints; Design and Analysis, V002T03A018, July 13–15, 2021

Paper No: PVP2021-61856

DOI: https://doi.org/10.1115/PVP2021-61856

View Paper

Topics: Finite element analysis, Fracture mechanics, Pressure, Temperature, Brittleness, Cylinders, Design, Ductility, Failure, Finite element model

Proceedings Papers

Effect of Several Design Parameters on the Sealing Performance of Reactor Pressure Vessel

Liangang Zheng, Xue Mi, Kaikai Shi, Xiaoming Bai, Hai Xie, Bin Zheng, Xiaohuan Yu

Proc. ASME. PVP2020, Volume 3: Design and Analysis, V003T03A015, August 3, 2020

Paper No: PVP2020-21208

DOI: https://doi.org/10.1115/PVP2020-21208

View Paper

Topics: Design, Reactor vessels, Sealing (Process), Flanges, Pressure vessels, Separation (Technology), Springs, Finite element analysis, Nuclear power stations, Nuclear reactor coolants

Proceedings Papers

Environmental Fatigue Screening for Subsequent License Renewal

Adam P. Walker, Gregory M. Imbrogno, Mark A. Gray, Charles A. Tomes

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A023, August 3, 2020

Paper No: PVP2020-21678

DOI: https://doi.org/10.1115/PVP2020-21678

View Paper

Topics: Fatigue, Licensing, Fatigue analysis, Cycles, Electric arc furnaces, Nuclear reactor coolants, Pressure, Transients (Dynamics), Water

Proceedings Papers

Study on the Influence of Crack Growth Behavior on Fatigue Life in Simulated Reactor Coolant Environment of Different Temperature

Tatsuru Misawa, Takanori Kitada, Takao Nakamura

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A040, August 3, 2020

Paper No: PVP2020-21172

DOI: https://doi.org/10.1115/PVP2020-21172

View Paper

Topics: Fatigue life, Fracture (Materials), Nuclear reactor coolants, Temperature, Pressurized water reactors, Crack propagation, Fatigue testing, High temperature, High pressure (Physics), Oxygen

Proceedings Papers

Acceptance Criteria for the Mechanical Integrity of Clad/Base Metal Interface for High Temperature Nuclear Reactor Cladded Components

B. Barua, M. C. Messner, A. Rovinelli, T.-L. Sham

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A071, August 3, 2020

Paper No: PVP2020-21494

DOI: https://doi.org/10.1115/PVP2020-21494

View Paper

Topics: Acceptance criteria, Base metals, High temperature, Nuclear reactors, Design, Shear stress, Temperature, Design methodology, Bonding, Corrosion resistance

Proceedings Papers

Design Methodologies for High Temperature Reactor Structural Components Cladded With Noncompliant Materials

B. Barua, V.-T. Phan, M. C. Messner, B. Jetter, T.-L. Sham, Y. Wang

Proc. ASME. PVP2019, Volume 1: Codes and Standards, V001T01A063, July 14–19, 2019

Paper No: PVP2019-93643

DOI: https://doi.org/10.1115/PVP2019-93643

View Paper

Topics: Design, High temperature, Structural elements (Construction), Corrosion resistant materials, Creep, Corrosion resistance, Deformation, Fatigue, Finite element analysis, Heat exchangers

Proceedings Papers

Vibration of a High Energy PWR Piping System due to Vibro-Acoustic Coupling: Evaluation and Mitigation

Gregory A. Meyer, Timothy K. Meneely, Jeremy R. Koether, Stephen D. Smith, David C. DiBasilio

Proc. ASME. PVP2019, Volume 4: Fluid-Structure Interaction, V004T04A021, July 14–19, 2019

Paper No: PVP2019-93594

DOI: https://doi.org/10.1115/PVP2019-93594

View Paper

Topics: Acoustics, Piping systems, Pressurized water reactors, Vibration, Pipes, Flow (Dynamics), Resonance, Boiling water reactors, Excitation, Junctions

Proceedings Papers

Demonstrate ASME Section XI Appendix G Margins for Pressurized Water Reactor Inlet and Outlet Nozzle Corner Regions

Anees Udyawar, J. Brian Hall, Justin Webb, Alexandria Carolan

Proc. ASME. PVP2018, Volume 1B: Codes and Standards, V01BT01A068, July 15–20, 2018

Paper No: PVP2018-84345

DOI: https://doi.org/10.1115/PVP2018-84345

View Paper

Topics: Corners (Structural elements), Nozzles, Pressurized water reactors, Temperature, Pressure, Finite element analysis, Fracture toughness, Stress, Reactor vessels, Ductility

Proceedings Papers

Cooperative Design and Optimization of Reactor Coolant System Piping Supports Under Static and Dynamical Load Conditions

Fu-Rui Xiong, Bin Lan

Proc. ASME. PVP2018, Volume 3B: Design and Analysis, V03BT03A040, July 15–20, 2018

Paper No: PVP2018-84026

DOI: https://doi.org/10.1115/PVP2018-84026

View Paper

Topics: Design, Nuclear reactor coolants, Optimization, Pipes, Stress, Algorithms, ASME Boiler and Pressure Vessel Code, Boilers, Flexible structures, Pareto optimization

Proceedings Papers

Study on the Influence of Strain Rate on Crack Initiation and Growth in Simulated Reactor Coolant Environment of Type 316 Stainless Steel

Takahisa Nose, Takao Nakamura, Takanori Kitada

Proc. ASME. PVP2018, Volume 1A: Codes and Standards, V01AT01A022, July 15–20, 2018

Paper No: PVP2018-84490

DOI: https://doi.org/10.1115/PVP2018-84490

View Paper

Topics: Fracture (Materials), Nuclear reactor coolants, Stainless steel, Crack propagation, Fatigue, Fatigue cracks, Fatigue damage, Fatigue life, Fatigue testing, Maintenance

Proceedings Papers

ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

Anees Udyawar, Charles Tomes, Alexandria Carolan, Steve Marlette, Thomas Meikle, Warren Bamford

Proc. ASME. PVP2018, Volume 1A: Codes and Standards, V01AT01A020, July 15–20, 2018

Paper No: PVP2018-84346

DOI: https://doi.org/10.1115/PVP2018-84346

View Paper

Topics: Licensing, Piping systems, Pressurized water reactors, Fatigue, Transients (Dynamics), Cycles, In-service inspection, Inspection, Construction, Design

Proceedings Papers

Simulation of Acoustic-Structure Interaction by Using Finite Element Analysis in Reactor Pressure Vessel of PWR

Koji Maeta, Keisuke Matsuyama, Hirokazu Sugiura, Shigeyuki Watanabe, Hideyuki Morita, Akihisa Iwasaki, Yoshito Nishikawa

Proc. ASME. PVP2018, Volume 4: Fluid-Structure Interaction, V004T04A007, July 15–20, 2018

Paper No: PVP2018-84496

DOI: https://doi.org/10.1115/PVP2018-84496

View Paper

Topics: Acoustics, Finite element analysis, Pressurized water reactors, Reactor vessels, Simulation, Sound pressure, Resonance ionization spectroscopy, Pressure, Computer software, Nuclear reactor coolants

Proceedings Papers

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)

Sebastian Lindqvist, Kim Wallin, Dominique Moinereau, Mike Smith, Stéphane Marie, Peter Dillström, Elisabeth Keim, Szabolcs Szavai

Proc. ASME. PVP2018, Volume 6A: Materials and Fabrication, V06AT06A073, July 15–20, 2018

Paper No: PVP2018-84554

DOI: https://doi.org/10.1115/PVP2018-84554

View Paper

Topics: Coolants, Fracture mechanics, Materials properties, Nuclear reactor coolants, Nuclear reactors, Piping systems, Pressure, Reactor vessels, Residual stresses, Safety

Proceedings Papers

Numerical Calculation of Fatigue Crack Growth for the Structures of Elbow and Tee Pipe Based on RCC-MR A16 Code

Peng Tang, Jiacheng Luo, Pengzhou Li, Lei Sun, Juan Luo

Proc. ASME. PVP2018, Volume 6B: Materials and Fabrication, V06BT06A056, July 15–20, 2018

Paper No: PVP2018-85061

DOI: https://doi.org/10.1115/PVP2018-85061

View Paper

Topics: Fatigue cracks, Pipes, Pipeline systems, Stress, Coolants, Crack propagation, Fatigue, Flow-induced vibrations, Fracture (Materials), High pressure (Physics)

Proceedings Papers

Numerical Study on the Creep Strain Characteristics for TMSR Reactor Coolant Piping Under Thermal Loading

Wei Gong, Xiaochun Zhang, Mingqiang Xie, Yuan Fu, Xiao Wang

Proc. ASME. PVP2018, Volume 6B: Materials and Fabrication, V06BT06A058, July 15–20, 2018

Paper No: PVP2018-84333

DOI: https://doi.org/10.1115/PVP2018-84333

View Paper

Topics: Creep, Nuclear reactor coolants, Pipes, Fatigue, Cycles, Damage, Stress, Computational methods, Design, Design methodology

Proceedings Papers

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

Robert A. Leishear

Proc. ASME. PVP2017, Volume 4: Fluid-Structure Interaction, V004T04A012, July 16–20, 2017

Paper No: PVP2017-66279

DOI: https://doi.org/10.1115/PVP2017-66279

View Paper

Topics: Explosions, Fire, Nuclear power stations, Water hammer, Hydrogen, Oxygen, Pipes, Compression, Accidents, Combustion

Proceedings Papers

On-Site Vibration Test for Measuring Damping Ratio of PWR Reactor Coolant Loop

Takami Ishiguro, Takeshi Numata, Nobuyoshi Goshima, Masatsugu Monde, Hiroyuki Fuyama

Proc. ASME. PVP2017, Volume 8: Seismic Engineering, V008T08A008, July 16–20, 2017

Paper No: PVP2017-65047

DOI: https://doi.org/10.1115/PVP2017-65047

View Paper

Topics: Damping, Nuclear reactor coolants, Pressurized water reactors, Vibration tests, Earthquakes, Energy dissipation, Seismic analysis, Actuators, Boilers, Construction

Proceedings Papers

WJP Applicability Study for PWR Components As a PWSCC Mitigation Technique

Ryuji Kimura, Noboru Saito, Hisamitsu Hato, Akihiro Kanno, Masami Ando, Hiroshi Muto, Fujio Yoshikubo, Morio Sato, Hiroyuki Kawamoto

Proc. ASME. PVP2017, Volume 1B: Codes and Standards, V01BT01A058, July 16–20, 2017

Paper No: PVP2017-66164

DOI: https://doi.org/10.1115/PVP2017-66164

View Paper

Topics: Pressurized water reactors, Stress, Nozzles, Water, Shot peening, Flow (Dynamics), Reactor vessels, Cavitation, Inspection, Bubbles

Proceedings Papers

Assessment of Powder Metallurgy-Hot Isostatic Pressed Nozzle-to-Safe End Transition Joints

Ben Sutton, David Gandy

Proc. ASME. PVP2017, Volume 6A: Materials and Fabrication, V06AT06A014, July 16–20, 2017

Paper No: PVP2017-65776

DOI: https://doi.org/10.1115/PVP2017-65776

View Paper

Topics: Metallurgy, Nozzles, Alloys, Manufacturing, Metals, Alloy steel, Heat treating (Metalworking), Welded joints, Corrosion, Diffusion (Physics)

Proceedings Papers

Analysis of Water Hammer due to Sudden Rupture of Reactor Coolant System

Iaroslav Dubyk, Igor Orynyak

Proc. ASME. PVP2016, Volume 4: Fluid-Structure Interaction, V004T04A021, July 17–21, 2016

Paper No: PVP2016-63589

DOI: https://doi.org/10.1115/PVP2016-63589

View Paper

Topics: Nuclear reactor coolants, Rupture, Water hammer, Stress, Accidents, Cooling, Design, Impulse (Physics), Shells, Structural stability

2 3 4 5

NARROW

NuScale Pressure and Temperature Limits Methodology Using Finite Element Analysis

Effect of Several Design Parameters on the Sealing Performance of Reactor Pressure Vessel

Environmental Fatigue Screening for Subsequent License Renewal

Study on the Influence of Crack Growth Behavior on Fatigue Life in Simulated Reactor Coolant Environment of Different Temperature

Acceptance Criteria for the Mechanical Integrity of Clad/Base Metal Interface for High Temperature Nuclear Reactor Cladded Components

Design Methodologies for High Temperature Reactor Structural Components Cladded With Noncompliant Materials

Vibration of a High Energy PWR Piping System due to Vibro-Acoustic Coupling: Evaluation and Mitigation

Demonstrate ASME Section XI Appendix G Margins for Pressurized Water Reactor Inlet and Outlet Nozzle Corner Regions

Cooperative Design and Optimization of Reactor Coolant System Piping Supports Under Static and Dynamical Load Conditions

Study on the Influence of Strain Rate on Crack Initiation and Growth in Simulated Reactor Coolant Environment of Type 316 Stainless Steel

ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

Simulation of Acoustic-Structure Interaction by Using Finite Element Analysis in Reactor Pressure Vessel of PWR

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)

Numerical Calculation of Fatigue Crack Growth for the Structures of Elbow and Tee Pipe Based on RCC-MR A16 Code

Numerical Study on the Creep Strain Characteristics for TMSR Reactor Coolant Piping Under Thermal Loading

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

On-Site Vibration Test for Measuring Damping Ratio of PWR Reactor Coolant Loop

WJP Applicability Study for PWR Components As a PWSCC Mitigation Technique

Assessment of Powder Metallurgy-Hot Isostatic Pressed Nozzle-to-Safe End Transition Joints

Analysis of Water Hammer due to Sudden Rupture of Reactor Coolant System

Journals

Conference Proceedings

eBooks

Resources

Opportunities

Journals

Conference Proceedings

eBooks

Resources

Opportunities

This Feature Is Available To Subscribers Only