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Proceedings Papers

Effect of Contact Stress on Fretting Corrosion Behavior of Alloy 690 Tube Exposed to Simulated Secondary Water

Wenhua Zhuang, Zhongli Han, Xianglong Guo, Lefu Zhang

Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A054, August 8–12, 2022

Paper No: ICONE29-92395

DOI: https://doi.org/10.1115/ICONE29-92395

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Topics: Alloys, Corrosion, Stress, Water, Wear

Proceedings Papers

Analysis Method to Predict Grid-to-Rod Fretting Wear of Fuel Rod in Pressurized Water Reactors

Libing Zhu, Shufeng Zhao, Jingya Sun, Shiyin Xu, Fujun Gan, Sirui Li

Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A074, August 8–12, 2022

Paper No: ICONE29-93076

DOI: https://doi.org/10.1115/ICONE29-93076

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Topics: Fuel rods, Pressurized water reactors, Wear

Proceedings Papers

The Wear and General Corrosion Properties of Low-Temperature Plasma Nitrided 17-4PH Stainless Steel With a Work-Hardened Sublayer

Zhongli Han, Ping Lai, Wenhua Zhuang, Xianglong Guo, Lefu Zhang

Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A026, August 8–12, 2022

Paper No: ICONE29-90835

DOI: https://doi.org/10.1115/ICONE29-90835

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Topics: Corrosion, Low temperature, Nitriding, Plasmas (Ionized gases), Stainless steel, Wear

Proceedings Papers

Study of Pre-Oxidization Law and Fretting Wear Resistance of CZ2 Alloy Cladding

Guocheng Sun, Shi Lin, Xu Wang, Liutao Chen

Proc. ASME. ICONE29, Volume 2: Nuclear Fuel and Material, Reactor Physics and Transport Theory, and Fuel Cycle Technology, V002T02A088, August 8–12, 2022

Paper No: ICONE29-93804

DOI: https://doi.org/10.1115/ICONE29-93804

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Topics: Alloys, Cladding systems (Building), Wear, Wear resistance

Proceedings Papers

Study on the Application of Electromagnetic Stepping CRDM at Tilted Environment

Zhaohui Lu, Zewen Li, Jianming Zhou, Qingsong Liu

Proc. ASME. ICONE29, Volume 10: Advanced Methods of Manufacturing for Nuclear Reactors and Components, V010T10A005, August 8–12, 2022

Paper No: ICONE29-90150

DOI: https://doi.org/10.1115/ICONE29-90150

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Topics: Design, Engineering prototypes, Failure, Kinematics, Polishing equipment, Pressurized water reactors, Stress, Structural design, Surface finishing, Wear

Proceedings Papers

Research on Fretting Wear of Grid and Cladding Tube Under Simulated Primary Circuit Environment of Pressurized Water Reactor

Tingting Liu, Kun Li, Wenhui Ma, Yuxiang Zhang, Minli Chen, Xiaohang Wu

Proc. ASME. ICONE29, Volume 7A: Thermal-Hydraulics and Safety Analysis, V07AT07A044, August 8–12, 2022

Paper No: ICONE29-91561

DOI: https://doi.org/10.1115/ICONE29-91561

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Topics: Circuits, Cladding systems (Building), Pressurized water reactors, Wear

Proceedings Papers

Flow Induced Vibration and Fretting Wear Characteristics of Fuel Rods

Zhipeng Feng, Huanhuan Qi, Xuan Huang, Guo Chen, Shuai Liu, Yixiong Zhang

Proc. ASME. ICONE28, Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, V002T05A003, August 4–6, 2021

Paper No: ICONE28-62135

DOI: https://doi.org/10.1115/ICONE28-62135

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Topics: Flow-induced vibrations, Fuel rods, Wear, Fuels, Manufacturing, Design, Fluid density, Pressurized water reactors, Evaluation methods, Excitation

Proceedings Papers

Numerical and Experimental Investigation on Core Assembly Thermal-Gradient-Induced Deformation of Sodium-Cooled Fast Reactor

Zehua Ma, Yingwei Wu, G. H. Su, Wenxi Tian, Suizheng Qiu

Proc. ASME. ICONE26, Volume 9: Student Paper Competition, V009T16A063, July 22–26, 2018

Paper No: ICONE26-81891

DOI: https://doi.org/10.1115/ICONE26-81891

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Topics: Deformation, Manufacturing, Sodium fast reactors, Temperature gradient, Computation, Displacement, Finite element model, Friction, Reliability, Rods

Proceedings Papers

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

Lihua Wang, Qingxiang Yang, Ping Yang, Jiazheng Liu, Libing Zhu, Bo Yang, Hong Cao, Zhe Wang, Yang Ding, Jie Ding

Proc. ASME. ICONE25, Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application, V003T02A024, July 2–6, 2017

Paper No: ICONE25-66659

DOI: https://doi.org/10.1115/ICONE25-66659

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Topics: Fuels, Manufacturing, Design, Fuel rods, Nuclear power stations, Economics , Rods, Safety, Computer software, Coolants

Proceedings Papers

Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)

Huasong Cao

Proc. ASME. ICONE25, Volume 4: Nuclear Safety, Security, Non-Proliferation and Cyber Security; Risk Management, V004T06A014, July 2–6, 2017

Paper No: ICONE25-66637

DOI: https://doi.org/10.1115/ICONE25-66637

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Topics: Boilers, Pressurized water reactors, Simulation, Temperature, Coolants, Corrosion, Design, Flow (Dynamics), Pressure, Pumps

Proceedings Papers

Numerical Simulation of Flow Field Inside Reactor Upper Plenum for PWR With Code_Saturne

Jiawei Liu, Puzhen Gao, Tingting Xu, Jiesheng Min, Guofei Chen

Proc. ASME. ICONE25, Volume 6: Thermal-Hydraulics, V006T08A119, July 2–6, 2017

Paper No: ICONE25-68008

DOI: https://doi.org/10.1115/ICONE25-68008

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Topics: Computer simulation, Flow (Dynamics), Pressurized water reactors, Design, Computational fluid dynamics, Pressure, Wear, Computer software, Flow-induced vibrations, Manufacturing

Proceedings Papers

Testing the Impact on Emergency Diesel Performance of Water in Diesel Fuel

Robert Cryer, Mark O’Connell

Proc. ASME. ICONE24, Volume 1: Operations and Maintenance, Aging Management and Plant Upgrades; Nuclear Fuel, Fuel Cycle, Reactor Physics and Transport Theory; Plant Systems, Structures, Components and Materials; I&C, Digital Controls, and Influence of Human Factors, V001T01A004, June 26–30, 2016

Paper No: ICONE24-60214

DOI: https://doi.org/10.1115/ICONE24-60214

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Topics: Diesel, Emergencies, Testing, Water, Diesel engines, Engines, Computer software, Contamination, Nuclear power stations, Simulation

Proceedings Papers

HIPed Hard Facings for Nuclear Applications: Materials, Key Potential Defects and Mitigating Quality Control Measures

John Sulley, David Stewart

Paper No: ICONE24-61106

DOI: https://doi.org/10.1115/ICONE24-61106

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Topics: Foundry coatings, Quality control, Hardening, Wear, Erosion, Stress, Cobalt, Contamination, Corrosion resistance, Crack propagation

Proceedings Papers

Numerical Simulation of Fretting Wear of Heat-Transfer Tubes in Steam Generator Under Random Excitation Forces

Lichen Tang, Hao Qian, Yongcheng Xie, Yi Jing, Yongzhong Huo

Paper No: ICONE24-60630

DOI: https://doi.org/10.1115/ICONE24-60630

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Topics: Boilers, Computer simulation, Heat transfer, Random excitation, Wear, Displacement, Finite element methods, Finite element model, Fluids, Friction

Proceedings Papers

Improvement of FAC Maintenance Program Issued From BRT-CICERO™ via CFD Calculations

Elodie Gipon

Proc. ASME. ICONE24, Volume 4: Computational Fluid Dynamics (CFD) and Coupled Codes; Decontamination and Decommissioning, Radiation Protection, Shielding, and Waste Management; Workforce Development, Nuclear Education and Public Acceptance; Mitigation Strategies for Beyond Design Basis Events; Risk Management, V004T10A003, June 26–30, 2016

Paper No: ICONE24-60136

DOI: https://doi.org/10.1115/ICONE24-60136

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Topics: Computational fluid dynamics, Maintenance, Pipes, Computer software, Flow (Dynamics), Nuclear power stations, Algorithms, Corrosion, Feedback, Mass transfer

Proceedings Papers

Development of the Pump-Integrated Intermediate Heat Exchanger in Advanced Loop-Type Sodium-Cooled Fast Reactor for Demonstration

Katsunori Amano, Yasuhiro Enuma, Satoshi Futagami, Tomoyuki Inoue, Sota Watanabe

Proc. ASME. ICONE24, Volume 2: Smart Grids, Grid Stability, and Offsite and Emergency Power; Advanced and Next Generation Reactors, Fusion Technology; Safety, Security, and Cyber Security; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues, V002T06A004, June 26–30, 2016

Paper No: ICONE24-60064

DOI: https://doi.org/10.1115/ICONE24-60064

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Topics: Heat exchangers, Pumps, Sodium fast reactors, Design, Safety, Coolants, Earthquakes, Leakage, Maintainability, Nuclear power

Proceedings Papers

Fretting Wear of the Zr Fuel Rod Cladding

Ladislav Pecinka, Jaroslav Svoboda, Vladimír Zeman

Proc. ASME. ICONE22, Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues, V001T03A027, July 7–11, 2014

Paper No: ICONE22-31103

DOI: https://doi.org/10.1115/ICONE22-31103

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Topics: Cladding systems (Building), Fuel rods, Wear, Zirconium, Fuels, Nuclear power stations, Pressurized water reactors, Damage, Failure, Flow (Dynamics)

Proceedings Papers

Application of Leakage-Flux Analysis for Three-Phase Induction Motors

Kohichi Nakayama, Yukio Watanabe, Kenji Osaki, Kotaro Hegi

Paper No: ICONE22-30593

DOI: https://doi.org/10.1115/ICONE22-30593

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Topics: Electromagnetic induction, Leakage, Motors, Bearings, Electric current, Vibration, Wear

Proceedings Papers

Development of Co-Reduced Valve Seat for Nuclear Power Plants

Toshiyuki Tazawa, Masahiro Saito, Nobuo Kojima

Proc. ASME. ICONE21, Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Radiation Protection and Nuclear Technology Applications, V001T02A045, July 29–August 2, 2013

Paper No: ICONE21-16714

DOI: https://doi.org/10.1115/ICONE21-16714

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Topics: Nuclear power stations, Valves, Alloys, Overlays (Materials engineering), Abrasion, Radiation (Physics), Boiling water reactors, Hardness (Materials), Leakage, Mechanical testing

Proceedings Papers

Fretting of Fuel Cladding Materials for Pb Cooled Fast Reactors: Long Term Prediction Using Fretting Maps

Mattia Del Giacco, Alfons Weisenburger, Georg Müller

Paper No: ICONE21-16905

DOI: https://doi.org/10.1115/ICONE21-16905

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Topics: Cladding systems (Building), Fast neutron reactors, Fuels, Steel, Corrosion, Wear, Lead alloys, Nuclear reactors, Stress, Temperature

NARROW

Effect of Contact Stress on Fretting Corrosion Behavior of Alloy 690 Tube Exposed to Simulated Secondary Water

Analysis Method to Predict Grid-to-Rod Fretting Wear of Fuel Rod in Pressurized Water Reactors

The Wear and General Corrosion Properties of Low-Temperature Plasma Nitrided 17-4PH Stainless Steel With a Work-Hardened Sublayer

Study of Pre-Oxidization Law and Fretting Wear Resistance of CZ2 Alloy Cladding

Study on the Application of Electromagnetic Stepping CRDM at Tilted Environment

Research on Fretting Wear of Grid and Cladding Tube Under Simulated Primary Circuit Environment of Pressurized Water Reactor

Flow Induced Vibration and Fretting Wear Characteristics of Fuel Rods

Numerical and Experimental Investigation on Core Assembly Thermal-Gradient-Induced Deformation of Sodium-Cooled Fast Reactor

Research and Engineering Application of Repaired Fuel Assembly Reusing Technique

Simulation of Small Steam Generator Tube Break (SSGTB) in a Small Pressurized Water Reactor (SPWR)

Numerical Simulation of Flow Field Inside Reactor Upper Plenum for PWR With Code_Saturne

Testing the Impact on Emergency Diesel Performance of Water in Diesel Fuel

HIPed Hard Facings for Nuclear Applications: Materials, Key Potential Defects and Mitigating Quality Control Measures

Numerical Simulation of Fretting Wear of Heat-Transfer Tubes in Steam Generator Under Random Excitation Forces

Improvement of FAC Maintenance Program Issued From BRT-CICERO™ via CFD Calculations

Development of the Pump-Integrated Intermediate Heat Exchanger in Advanced Loop-Type Sodium-Cooled Fast Reactor for Demonstration

Fretting Wear of the Zr Fuel Rod Cladding

Application of Leakage-Flux Analysis for Three-Phase Induction Motors

Development of Co-Reduced Valve Seat for Nuclear Power Plants

Fretting of Fuel Cladding Materials for Pb Cooled Fast Reactors: Long Term Prediction Using Fretting Maps

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