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Proceedings Papers

Damage Evaluations for BWR Lower Head in Severe Accident Based on Multi-Physics Simulations

Jinya Katsuyama, Yoshihito Yamaguchi, Yoshiyuki Nemoto, Takuya Furuta, Yoshiyuki Kaji

Proc. ASME. PVP2022, Volume 4A: Materials and Fabrication, V04AT06A030, July 17–22, 2022

Paper No: PVP2022-84609

DOI: https://doi.org/10.1115/PVP2022-84609

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Topics: Accidents, Boiling water reactors, Creep, Damage, Physics, Radiation (Physics), Simulation, Thermal hydraulics

Proceedings Papers

Study on Ductile Crack Extension and Fracture Behavior in Plate Specimen With a Semi-Elliptical Surface Crack Using a BWR Reactor Pressure Vessel Material

Takahiro Hayashi, Takuya Ogawa, Shuichi Yoshida, Masao Itatani, Toshiyuki Saito

Proc. ASME. PVP2022, Volume 4B: Materials and Fabrication, V04BT06A020, July 17–22, 2022

Paper No: PVP2022-84606

DOI: https://doi.org/10.1115/PVP2022-84606

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Topics: Boiling water reactors, Ductile fracture, Fracture (Materials), Fracture (Process), Fracture mechanics, Metals, Reactor vessels, Stress corrosion cracking, Surface cracks

Proceedings Papers

Large-Scale Parametric Modeling of Spent Nuclear Fuel Dynamics in the 30 cm Package Drop Scenario

Kevin Kadooka, James Fitzpatrick, Nicholas Klymyshyn

Proc. ASME. PVP2022, Volume 5: Operations, Applications, and Components; Seismic Engineering; ASME Nondestructive Evaluation, Diagnosis and Prognosis (NDPD) Division, V005T07A030, July 17–22, 2022

Paper No: PVP2022-84747

DOI: https://doi.org/10.1115/PVP2022-84747

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Topics: Dynamics (Mechanics), Modeling, Spent nuclear fuels, Fuels, Manufacturing, Simulation, Stress, Boiling water reactors, Cladding systems (Building), Fuel rods

Proceedings Papers

Computational Fluid Dynamics Modeling of Dry Cask Simulator With Crosswind

Ben J. Jensen, Sarah R. Suffield, Megan E. Higley

Paper No: PVP2022-84090

DOI: https://doi.org/10.1115/PVP2022-84090

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Topics: Computational fluid dynamics, Modeling, Spent nuclear fuels, Stress corrosion cracking, Computer software, Sensitivity analysis, Spent fuel storage, Boiling water reactors, Containment vessels, Flow (Dynamics)

Proceedings Papers

An Estimate of Spent Nuclear Fuel Mechanical Loads in the General 30 cm Package Drop Scenario

Nicholas Klymyshyn, Kevin Kadooka, Casey Spitz, James Fitzpatrick

Paper No: PVP2022-84353

DOI: https://doi.org/10.1115/PVP2022-84353

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Topics: Spent nuclear fuels, Stress, Fuels, Manufacturing, Damage, Modeling, Boiling water reactors, Boundary element methods, Collaboration, Design

Proceedings Papers

Seismic Test Results of the Main Steam Isolation Valve for Japanese Boiling Water Reactor Nuclear Power Plants (Part 2)

Hideaki Itabashi, Yoshitaka Tsutsumi, Koji Nishino, Shin Kumagai, Ryo Kubota, Shota Tanemura

Paper No: PVP2022-84190

DOI: https://doi.org/10.1115/PVP2022-84190

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Topics: Boiling water reactors, Nuclear power stations, Steam, Valves

Proceedings Papers

Study on Fracture Behavior and Assessment for Dissimilar Metal Weld of Low Alloy Steel and Ni-Base Alloy Weld Using a BWR Reactor Pressure Vessel Material

Takahiro Hayashi, Takuya Ogawa, Shuichi Yoshida, Masao Itatani, Yasuhiro Hattori, Tetsushi Yamaoka, Shigeaki Tanaka, Toshiyuki Saito

Proc. ASME. PVP2021, Volume 4: Materials and Fabrication, V004T06A047, July 13–15, 2021

Paper No: PVP2021-61467

DOI: https://doi.org/10.1115/PVP2021-61467

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Topics: Alloy steel, Alloys, Boiling water reactors, Fracture (Materials), Fracture mechanics, Fracture toughness, Metals, Reactor vessels, Stress corrosion cracking

Proceedings Papers

Thermal Analysis Evaluations Using the Dry Cask Simulator

S. R. Suffield, D. J. Richmond, J. A. Fort

Proc. ASME. PVP2021, Volume 5: Operations, Applications, and Components; Seismic Engineering; Non-Destructive Examination, V005T07A028, July 13–15, 2021

Paper No: PVP2021-61846

DOI: https://doi.org/10.1115/PVP2021-61846

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Topics: Thermal analysis, Fuels, Manufacturing, Computational fluid dynamics, Storage, Cladding systems (Building), Geometry, Heat, Stress, Temperature

Proceedings Papers

Probabilistic Fracture Mechanics Evaluation of a BWR Feedwater Nozzle

Kevin K. L. Wong, Garivalde Dominguez, Do Jun Shim, Steven K. Richter

Proc. ASME. PVP2021, Volume 1: Codes and Standards, V001T01A031, July 13–15, 2021

Paper No: PVP2021-62933

DOI: https://doi.org/10.1115/PVP2021-62933

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Topics: ASME Standards, Boiling water reactors, Feedwater, Fracture mechanics, Inspection, Nozzles, Shells

Proceedings Papers

Technical Basis for Code Case N-894 Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping With Thermal Fatigue Cracking

Stephen Marlette, Steven L. McCraken, Christopher Lohse

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A097, August 3, 2020

Paper No: PVP2020-21544

DOI: https://doi.org/10.1115/PVP2020-21544

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Topics: Fatigue cracks, Maintenance, Pipes, Stainless steel, Overlays (Materials engineering), Fatigue, Boiling water reactors, Pressurized water reactors, Stress corrosion cracking, Design

Proceedings Papers

Thermal Analysis Evaluations Using the Dry Cask Simulator

S. R. Suffield, D. J. Richmond, J. A. Fort

Proc. ASME. PVP2020, Volume 8: Operations, Applications, and Components, V008T08A037, August 3, 2020

Paper No: PVP2020-21146

DOI: https://doi.org/10.1115/PVP2020-21146

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Topics: Thermal analysis, Fuels, Manufacturing, Computational fluid dynamics, Storage, Cladding systems (Building), Geometry, Heat, Stress, Temperature

Proceedings Papers

Seismic Test Results of the Main Steam Isolation Valve for Japanese Boiling Water Reactor Nuclear Power Plants

Hideaki Itabashi, Yoshitaka Tsutsumi, Koji Nishino, Shin Kumagai

Proc. ASME. PVP2020, Volume 9: Seismic Engineering, V009T09A003, August 3, 2020

Paper No: PVP2020-21362

DOI: https://doi.org/10.1115/PVP2020-21362

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Topics: Boiling water reactors, Nuclear power stations, Steam, Valves

Proceedings Papers

Elimination of BWR Mark I Program Primary Containment Drywell-to-Wetwell Differential Pressure

Raymond M. Pace, Jason Ritzel

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A045, August 3, 2020

Paper No: PVP2020-21001

DOI: https://doi.org/10.1115/PVP2020-21001

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Topics: Boiling water reactors, Containment, Pressure, Vents, Accidents, Nitrogen, Water, Control rooms, Filters, Pipes

Proceedings Papers

Example Application of Code Case N-894 Rules for Weld Overlay of Thermal Fatigue Cracking

Christopher Lohse, Richard Bax, Minji Fong, Charles Fourcade, Do Jun Shim

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A100, August 3, 2020

Paper No: PVP2020-21760

DOI: https://doi.org/10.1115/PVP2020-21760

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Topics: Fatigue cracks, Overlays (Materials engineering), Fatigue, Filler metals, Maintenance, Boiling water reactors, Compression, Nickel alloys, Pressurized water reactors, Stainless steel

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

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

Akihiro Mano, Jinya Katsuyama, Yinsheng Li

Proc. ASME. PVP2020, Volume 1: Codes and Standards, V001T01A063, August 3, 2020

Paper No: PVP2020-21427

DOI: https://doi.org/10.1115/PVP2020-21427

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Topics: Fracture mechanics, Stress corrosion cracking, Water, Stress, Boiling water reactors, Collaboration, Failure, Fatigue, Nuclear power, Nuclear power stations

Proceedings Papers

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

Matthew Walter, Minghao Qin, Daniel Sommerville

Proc. ASME. PVP2020, Volume 4: Fluid-Structure Interaction, V004T04A009, August 3, 2020

Paper No: PVP2020-21633

DOI: https://doi.org/10.1115/PVP2020-21633

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Topics: Boiling water reactors, Sound pressure

Proceedings Papers

SCC Growth Prediction for Surface Crack in Welded Joint Using Advanced FEA

Takuya Ogawa, Masao Itatani, Takahiro Hayashi, Toshiyuki Saito

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A080, August 3, 2020

Paper No: PVP2020-21670

DOI: https://doi.org/10.1115/PVP2020-21670

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Topics: Finite element analysis, Surface cracks, Welded joints, Fracture (Materials), Fracture (Process), Shapes, Fitness-for-service, Metals, Stainless steel, Base metals

Proceedings Papers

Investigation on the Effects of Carbon Macro-Segregation of Mechanical Properties Using a Large-Scale Forged Low Alloy Steel for a BWR Reactor Pressure Vessel

Takahiro Hayashi, Takuya Ogawa, Rie Sumiya, Tetsushi Yamaoka, Shigeaki Tanaka, Yasuhiro Hattori, Atsushi Mori

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A082, August 3, 2020

Paper No: PVP2020-21857

DOI: https://doi.org/10.1115/PVP2020-21857

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Topics: Alloy steel, Boiling water reactors, Carbon, Fracture toughness, Iron and steel making, Mechanical properties, Pressure vessels, Reactor vessels

Proceedings Papers

Applicability of Formula of the Irradiation Assisted Stress Corrosion Cracking Rates for Neutron-Irradiated Type 316L Stainless Steels to Various Types of Stainless Steels Under Boiling Water Reactor Condition

Masato Koshiishi, Hitoshi Seto, Shigeaki Tanaka, Ryoji Obata

Proc. ASME. PVP2020, Volume 6: Materials and Fabrication, V006T06A073, August 3, 2020

Paper No: PVP2020-21332

DOI: https://doi.org/10.1115/PVP2020-21332

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Topics: Boiling water reactors, Irradiation (Radiation exposure), Neutrons, Stainless steel, Stress corrosion cracking, Fracture (Materials), Grain boundaries, Mechanical properties

2 3 4 5

NARROW

Damage Evaluations for BWR Lower Head in Severe Accident Based on Multi-Physics Simulations

Study on Ductile Crack Extension and Fracture Behavior in Plate Specimen With a Semi-Elliptical Surface Crack Using a BWR Reactor Pressure Vessel Material

Large-Scale Parametric Modeling of Spent Nuclear Fuel Dynamics in the 30 cm Package Drop Scenario

Computational Fluid Dynamics Modeling of Dry Cask Simulator With Crosswind

An Estimate of Spent Nuclear Fuel Mechanical Loads in the General 30 cm Package Drop Scenario

Seismic Test Results of the Main Steam Isolation Valve for Japanese Boiling Water Reactor Nuclear Power Plants (Part 2)

Study on Fracture Behavior and Assessment for Dissimilar Metal Weld of Low Alloy Steel and Ni-Base Alloy Weld Using a BWR Reactor Pressure Vessel Material

Thermal Analysis Evaluations Using the Dry Cask Simulator

Probabilistic Fracture Mechanics Evaluation of a BWR Feedwater Nozzle

Technical Basis for Code Case N-894 Alternative Rules for Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping With Thermal Fatigue Cracking

Thermal Analysis Evaluations Using the Dry Cask Simulator

Seismic Test Results of the Main Steam Isolation Valve for Japanese Boiling Water Reactor Nuclear Power Plants

Elimination of BWR Mark I Program Primary Containment Drywell-to-Wetwell Differential Pressure

Example Application of Code Case N-894 Rules for Weld Overlay of Thermal Fatigue Cracking

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

Probabilistic Fracture Mechanics Benchmarking Study Involving the xLPR and PASCAL-SP Codes: Analysis by PASCAL-SP

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

SCC Growth Prediction for Surface Crack in Welded Joint Using Advanced FEA

Investigation on the Effects of Carbon Macro-Segregation of Mechanical Properties Using a Large-Scale Forged Low Alloy Steel for a BWR Reactor Pressure Vessel

Applicability of Formula of the Irradiation Assisted Stress Corrosion Cracking Rates for Neutron-Irradiated Type 316L Stainless Steels to Various Types of Stainless Steels Under Boiling Water Reactor Condition

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