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Proceedings Papers

Research on Characteristics of Fluid Exciting Force of Reactor Internals

Zhipeng Feng, Liwen Deng, Xuan Huang, Pingchuan Shen, Shuai Liu, Fengchun Cai

Proc. ASME. PVP2021, Volume 3: Fluid Structure Interaction; High Pressure Technology, V003T04A005, July 13–15, 2021

Paper No: PVP2021-62128

DOI: https://doi.org/10.1115/PVP2021-62128

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Topics: Flow-induced vibrations, Fluids, Reactor vessels, Pressure, Computational fluid dynamics, Design, Excitation, Flow (Dynamics), Large eddy simulation, Nuclear reactors

Proceedings Papers

The Creep Crack Growth Behavior of GH3535 Superalloy at 650°C

Fan Guangcheng, Wang Wanxia, Tan Jianping, Shi Kexian, Wang Songlin, Lu Yanling

Proc. ASME. PVP2021, Volume 4: Materials and Fabrication, V004T06A011, July 13–15, 2021

Paper No: PVP2021-61787

DOI: https://doi.org/10.1115/PVP2021-61787

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Topics: Creep, Fracture (Materials), Superalloys, Alloys, Stress, China, Grain boundaries, High temperature, Mechanical properties, Molten salt reactors

Proceedings Papers

A Discussion of Strength Reduction Factor Development for Thermal Aging Effect on Nuclear Structural Alloys

Weiju Ren

Proc. ASME. PVP2021, Volume 4: Materials and Fabrication, V004T06A046, July 13–15, 2021

Paper No: PVP2021-61055

DOI: https://doi.org/10.1115/PVP2021-61055

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Topics: Alloys, Tensile strength, Design, Temperature, ASME Boiler and Pressure Vessel Code, Nuclear reactors, Safety, Service life (Equipment), Structural elements (Construction)

Proceedings Papers

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation - ATLAS+ Project: Status of the Activities of the WP3 on Modelling in 2021

Arnaud Blouin, Stéphane Marie, Tomas Nicak, Antti Timperi, Peter Gill

Proc. ASME. PVP2021, Volume 4: Materials and Fabrication, V004T06A029, July 13–15, 2021

Paper No: PVP2021-62010

DOI: https://doi.org/10.1115/PVP2021-62010

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Topics: Coolants, Fracture (Materials), Fracture toughness, Leak-before-break, Modeling, Nuclear reactors, Pipes, Pressure, Residual stresses

Proceedings Papers

An Update of the Assessment Methodology for Civil Ageing Management for LTO/CSO Based on International Standards and Engineering Judgement

F. H. E. de Haan - de Wilde, M. J. Janssen

Proc. ASME. PVP2021, Volume 5: Operations, Applications, and Components; Seismic Engineering; Non-Destructive Examination, V005T07A008, July 13–15, 2021

Paper No: PVP2021-61499

DOI: https://doi.org/10.1115/PVP2021-61499

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Topics: Amperes, Concretes, Containment, Engineering standards, Finite element analysis, Maintenance, Modeling, Nuclear power stations, Nuclear reactors, Performance

Proceedings Papers

Simplified Criteria With Reduced Testing Effort for Selecting Clad Materials for High Temperature Reactor Structural Components

B. Barua, R. I. Jetter, M. C. Messner, T.-L. Sham

Proc. ASME. PVP2021, Volume 1: Codes and Standards, V001T01A037, July 13–15, 2021

Paper No: PVP2021-61858

DOI: https://doi.org/10.1115/PVP2021-61858

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Topics: High temperature, Structural elements (Construction), Testing, Creep, Design, Temperature, Alloys, Corrosion, Corrosion resistant materials, Design methodology

Proceedings Papers

Prediction of J-Integrals at Defects in W-9CR Steel Sandwich-Type Cooling Pipes

Tristan Calvet, Yiqiang Wang, Minh-Son Pham, Catrin M. Davies

Proc. ASME. PVP2021, Volume 2: Computer Technology and Bolted Joints; Design and Analysis, V002T03A064, July 13–15, 2021

Paper No: PVP2021-62878

DOI: https://doi.org/10.1115/PVP2021-62878

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Topics: Brazing, Cooling, Heat flux, Pipes, Steel, Thermal stresses, Fracture (Materials), Manufacturing, Stress, Tungsten

Proceedings Papers

Probabilistic Leak Before Break Using the R6 Methodology

Peter Gill

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A110, August 3, 2020

Paper No: PVP2020-21769

DOI: https://doi.org/10.1115/PVP2020-21769

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Topics: Leak-before-break, Stress, Computer software, Failure, Probability, Reliability, Fracture (Materials), Fracture mechanics, Inspection, Leakage

Proceedings Papers

An Assessment Methodology for Civil Ageing Management and Concrete for LTO/CSO Based on International Standards and Engineering Judgement

F. H. E. de Haan - de Wilde, C. G. M. de Bont

Proc. ASME. PVP2020, Volume 8: Operations, Applications, and Components, V008T08A013, August 3, 2020

Paper No: PVP2020-21838

DOI: https://doi.org/10.1115/PVP2020-21838

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Topics: Amperes, Concretes, Engineering standards, Containment, Pressurized water reactors, Steel, Finite element analysis, Maintenance, Modeling, Nuclear power stations

Proceedings Papers

High Flux Reactor Continued Safe Operation: Time Limited Ageing Analyses

L. Stefanini, F. H. E. de Haan - de Wilde

Proc. ASME. PVP2020, Volume 8: Operations, Applications, and Components, V008T08A010, August 3, 2020

Paper No: PVP2020-21032

DOI: https://doi.org/10.1115/PVP2020-21032

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Topics: Biomedicine, Equipment qualification, Fatigue, Irradiation (Radiation exposure), Isotopes, Leak-before-break, Manufacturing, Nuclear reactors, Reactor vessels

Proceedings Papers

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation: ATLAS+ Project — Status of the Activities of the WP3 on Modelling in 2020

Arnaud Blouin, Stéphane Marie, Tomas Nicak, Antti Timperi, Peter Gill

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A046, August 3, 2020

Paper No: PVP2020-21551

DOI: https://doi.org/10.1115/PVP2020-21551

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Topics: Coolants, Fracture (Materials), Fracture toughness, Leak-before-break, Modeling, Nuclear reactors, Pipes, Pressure, Residual stresses

Proceedings Papers

Ductile Tearing Prediction of Ferritic Pipe Material by GTN Model for ATLAS+ European Project

Kiminobu Hojo, Takatoshi Hirota, Naoki Ogawa, Satoshi Kumagai

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A047, August 3, 2020

Paper No: PVP2020-21645

DOI: https://doi.org/10.1115/PVP2020-21645

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Topics: Pipes, Fracture (Materials), Fracture (Process), Coolants, Ductile fracture, Fracture mechanics, Nuclear reactors, Pressure, Simulation, Stress

Proceedings Papers

Recent Argonne National Laboratory Estimated Thermal Expansion Coefficients for Reactor Pressure Boundary Base Metal, Similar and Dissimilar Metal Welds

Subhasish Mohanty, Joseph Listwan

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A024, August 3, 2020

Paper No: PVP2020-21828

DOI: https://doi.org/10.1115/PVP2020-21828

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Topics: Base metals, Metals, Pressure, Thermal expansion, Welded joints, Fillers (Materials), Nuclear reactors, Safety, Stress, Stress analysis (Engineering)

Proceedings Papers

Overview of UK Policy and Research Landscape Relevant to Deploying Advanced Nuclear Technologies in the UK

Nicholas Underwood, Paul Nevitt, Andrew Howarth, Nicholas Barron

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A043, August 3, 2020

Paper No: PVP2020-21790

DOI: https://doi.org/10.1115/PVP2020-21790

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Topics: Advanced materials, Boiling, Carbon, Climate change, Construction, Cooling, Cooling systems, Cutting, Cycles, Design

Proceedings Papers

Assessment of the Continued Need for Independent Requirements on Transition Temperature and Upper Shelf Charpy Impact Toughness Metrics in Regulations Pertaining to Nuclear Reactor Pressure Vessels

Mark Kirk, Masato Yamamoto, Marjorie Erickson

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A057, August 3, 2020

Paper No: PVP2020-21186

DOI: https://doi.org/10.1115/PVP2020-21186

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Topics: Fracture toughness, Nuclear reactors, Phase transition temperature, Pressure vessels, Regulations, Steel, Nuclear power stations, Pressure

Proceedings Papers

Development of Design Method for High Temperature Nuclear Reactor Cladded Components

B. Barua, M. C. Messner, R. I. Jetter, T.-L. Sham

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A067, August 3, 2020

Paper No: PVP2020-21469

DOI: https://doi.org/10.1115/PVP2020-21469

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Topics: Design methodology, High temperature, Nuclear reactors, Design, Coolants, Corrosion resistant materials, Corrosion, Creep, Failure, Finite element analysis

Proceedings Papers

Selection Criteria for Clad Materials to Use With a 316H Base Material for High Temperature Nuclear Reactor Cladded Components

B. Barua, M. C. Messner, R. I. Jetter, T.-L. Sham

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A070, August 3, 2020

Paper No: PVP2020-21493

DOI: https://doi.org/10.1115/PVP2020-21493

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Topics: High temperature, Nuclear reactors, Creep, Design, Alloys, Temperature, Cladding systems (Building), Design methodology, Elastic moduli, Failure

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

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

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

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

Alberto Sáez-Maderuelo, María Luisa Ruiz-Lorenzo, Francisco Javier Perosanz, Patricie Halodová, Jan Prochazka, Hygreeva Namburi

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A077, August 3, 2020

Paper No: PVP2020-21518

DOI: https://doi.org/10.1115/PVP2020-21518

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Topics: Alloys, Augers, Nickel, Nuclear reactors, Oxidation, Scanning electron microscopes, Water

2 3 4 5

NARROW

Research on Characteristics of Fluid Exciting Force of Reactor Internals

The Creep Crack Growth Behavior of GH3535 Superalloy at 650°C

A Discussion of Strength Reduction Factor Development for Thermal Aging Effect on Nuclear Structural Alloys

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation - ATLAS+ Project: Status of the Activities of the WP3 on Modelling in 2021

An Update of the Assessment Methodology for Civil Ageing Management for LTO/CSO Based on International Standards and Engineering Judgement

Simplified Criteria With Reduced Testing Effort for Selecting Clad Materials for High Temperature Reactor Structural Components

Prediction of J-Integrals at Defects in W-9CR Steel Sandwich-Type Cooling Pipes

Probabilistic Leak Before Break Using the R6 Methodology

An Assessment Methodology for Civil Ageing Management and Concrete for LTO/CSO Based on International Standards and Engineering Judgement

High Flux Reactor Continued Safe Operation: Time Limited Ageing Analyses

Advanced Structural Integrity Assessment Tools for Safe Long Term Operation: ATLAS+ Project — Status of the Activities of the WP3 on Modelling in 2020

Ductile Tearing Prediction of Ferritic Pipe Material by GTN Model for ATLAS+ European Project

Recent Argonne National Laboratory Estimated Thermal Expansion Coefficients for Reactor Pressure Boundary Base Metal, Similar and Dissimilar Metal Welds

Overview of UK Policy and Research Landscape Relevant to Deploying Advanced Nuclear Technologies in the UK

Assessment of the Continued Need for Independent Requirements on Transition Temperature and Upper Shelf Charpy Impact Toughness Metrics in Regulations Pertaining to Nuclear Reactor Pressure Vessels

Development of Design Method for High Temperature Nuclear Reactor Cladded Components

Selection Criteria for Clad Materials to Use With a 316H Base Material for High Temperature Nuclear Reactor Cladded Components

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

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

Study of the Influence of Microstructure and Intergranular Carbides on the Oxidation Behavior of a Nickel Base Alloy 690 TT in Supercritical Water Nuclear Reactor Conditions

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