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Proceedings Papers

A Weakly Supervised Time Series Analysis Framework for Anomaly Detection in Nuclear Power Plants

Feiyan Dong, Shi Chen, Kazuyuki Demachi, Masanori Yoshikawa, Akiyuki Seki, Shigeru Takaya

Proc. ASME. ICONE29, Volume 15: Student Paper Competition, V015T16A050, August 8–12, 2022

Paper No: ICONE29-91609

DOI: https://doi.org/10.1115/ICONE29-91609

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Topics: Condition monitoring, Nuclear power stations, Time series

Proceedings Papers

Proposal of Detailed Procedures of Determining Rational In-Service Inspection Requirements Based on System Based Code Concept

Hiroki Yada, Shigeru Takaya, Kyoichi Morohoshi, Shinobu Yokoi

Proc. ASME. ICONE29, Volume 1: Nuclear Plant Operation and Maintenance, Engineering and Modification, Operation Life Extension (OLE), and Life Cycle, V001T01A013, August 8–12, 2022

Paper No: ICONE29-90542

DOI: https://doi.org/10.1115/ICONE29-90542

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Topics: In-service inspection, Inspection, Sodium fast reactors

Journal Articles

Development of In-Service Inspection Rules for Liquid-Metal Cooled Reactors Using the System Based Code Concept

Shigeru Takaya, Tai Asayama, Hiroki Yada, A. Thomas Roberts, Frank J. Schaaf, Jr.

J. Pressure Vessel Technol. April 2020, 142(2): 021204.

Paper No: PVT-19-1071

DOI: https://doi.org/10.1115/1.4044344

Published Online: November 27, 2019

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Topics: In-service inspection, Liquid metals, Sodium, Pipes, Heat, Breeder reactors, Fracture (Materials), Welded joints, Reliability, Engineering prototypes

Proceedings Papers

Evaluation Method of Response Reliability During an Accident and its Applicability to Fast Reactor Plants

Masaaki Suzuki, Kazuyuki Demachi, Shigeru Takaya, Yoshitaka Chikazawa

Proc. ASME. ICONE26, Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation, V004T06A046, July 22–26, 2018

Paper No: ICONE26-82567

DOI: https://doi.org/10.1115/ICONE26-82567

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Topics: Accidents, Evaluation methods, Fast neutron reactors, Reliability, Safety, Nuclear power stations

Proceedings Papers

Proposal of a Simple Evaluation Method for Sloshing Impact Pressure on Flat Roofs

Shigeru Takaya, Tatsuya Fujisaki

Proc. ASME. ICONE26, Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management, V002T03A031, July 22–26, 2018

Paper No: ICONE26-82562

DOI: https://doi.org/10.1115/ICONE26-82562

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Topics: Evaluation methods, Pressure, Roofs, Sloshing, Computational fluid dynamics, Damage, Earthquakes, Fast neutron reactors, Reactor vessels, Slabs

Proceedings Papers

Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow

Shigeru Takaya, Tatsuya Fujisaki, Masaaki Tanaka

Proc. ASME. PVP2017, Volume 4: Fluid-Structure Interaction, V004T04A006, July 16–20, 2017

Paper No: PVP2017-65601

DOI: https://doi.org/10.1115/PVP2017-65601

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Topics: Flow-induced vibrations, Numerical analysis, Pipes, Pressure, Design, Stress, Computer simulation, Flow (Dynamics), Construction, Cooling systems

Proceedings Papers

Numerical Analysis of Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Advanced Loop-Type Sodium Cooled Fast Reactor

Shigeru Takaya, Tatsuya Fujisaki, Masaaki Tanaka

Proc. ASME. PVP2016, Volume 3: Design and Analysis, V003T03A003, July 17–21, 2016

Paper No: PVP2016-63845

DOI: https://doi.org/10.1115/PVP2016-63845

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Topics: Numerical analysis, Piping systems, Sodium fast reactors, Flow (Dynamics), Pipes, Design, Simulation results, Swirling flow, Computer simulation, Construction

Proceedings Papers

Implementation of Reliability Evaluation Into JSME Fast Reactor Codes: 1 — Current Status and Path Forward

Tai Asayama, Shigeru Takaya

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, V002T08A007, June 26–30, 2016

Paper No: ICONE24-60936

DOI: https://doi.org/10.1115/ICONE24-60936

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Topics: Fast neutron reactors, Japan Society of Mechanical Engineers, Reliability, Design, In-service inspection, Buckling, Design methodology, Engineering standards, Inspection, Stress

Proceedings Papers

Implementation of Reliability Evaluation Into JSME Fast Reactor Codes: 2 — Development of a Code Case for Seismic Buckling Evaluation Based on LRFD

Shigeru Takaya, Naoto Sasaki, Tai Asayama, Yoshio Kamishima

Paper No: ICONE24-60964

DOI: https://doi.org/10.1115/ICONE24-60964

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Topics: Buckling, Fast neutron reactors, Japan Society of Mechanical Engineers, Reliability, Design, Stress, Probability, Vessels, Civil engineering, Earthquake risk

Proceedings Papers

Development of Integrated Numerical Analysis Model for Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Japan Sodium-Cooled Fast Reactor

Shigeru Takaya, Masaaki Tanaka, Tatsuya Fujisaki

Proc. ASME. PVP2015, Volume 3: Design and Analysis, V003T03A001, July 19–23, 2015

Paper No: PVP2015-45248

DOI: https://doi.org/10.1115/PVP2015-45248

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Topics: Numerical analysis, Piping systems, Sodium fast reactors, Flow (Dynamics), Stress, Vibration analysis, Computational fluid dynamics, Design, Engineering prototypes, Flow-induced vibrations

Journal Articles

Study on Minimum Wall Thickness Requirement for Seismic Buckling of Reactor Vessel Based on System Based Code Concept

Shigeru Takaya, Daigo Watanabe, Shinobu Yokoi, Yoshio Kamishima, Kenichi Kurisaka, Tai Asayama

J. Pressure Vessel Technol. October 2015, 137(5): 051802.

Paper No: PVT-14-1003

DOI: https://doi.org/10.1115/1.4029322

Published Online: October 1, 2015

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Topics: Buckling, Design, Earthquakes, Reliability, Stress, Wall thickness, Probability, Statistical distributions, Reactor vessels, Earthquake risk

Journal Articles

Evaluation of Fatigue Strength of Similar and Dissimilar Welded Joints of Modified 9Cr–1Mo Steel

Shigeru Takaya

J. Pressure Vessel Technol. February 2016, 138(1): 011402.

Paper No: PVT-15-1003

DOI: https://doi.org/10.1115/1.4030932

Published Online: August 26, 2015

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Topics: Fatigue life, Steel, Welded joints, Metals, Fatigue, Base metals

Journal Articles

Development of Creep–Fatigue Evaluation Method for 316FR Stainless Steel

Yuji Nagae, Shigeru Takaya, Tai Asayama

J. Pressure Vessel Technol. August 2015, 137(4): 041407.

Paper No: PVT-14-1096

DOI: https://doi.org/10.1115/1.4028497

Published Online: August 1, 2015

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Topics: Creep, Fatigue, Fatigue life, Stainless steel, Damage

Journal Articles

Application of the System Based Code Concept to the Determination of In-Service Inspection Requirements

Shigeru Takaya, Tai Asayama, Yoshio Kamishima, Hideo Machida, Daigo Watanabe, Satoru Nakai, Masaki Morishita

ASME J of Nuclear Rad Sci. January 2015, 1(1): 011004.

Paper No: NERS-14-1020

DOI: https://doi.org/10.1115/1.4026392

Published Online: February 9, 2015

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Topics: Creep, Design, Failure, Fatigue, Fracture (Materials), In-service inspection, Probability, Reliability, Safety, Stress

Proceedings Papers

Development of Structural Codes for JSFR Based on the System Based Code Concept

Tai Asayama, Takashi Wakai, Masanori Ando, Satoshi Okajima, Yuji Nagae, Shigeru Takaya, Takashi Onizawa, Kazuyuki Tsukimori, Masaki Morishita

Proc. ASME. PVP2014, Volume 1: Codes and Standards, V001T01A064, July 20–24, 2014

Paper No: PVP2014-28853

DOI: https://doi.org/10.1115/PVP2014-28853

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Topics: Construction, Design, Fast neutron reactors, Fitness-for-service, Industrial research, Japan Society of Mechanical Engineers, Leak-before-break, Reliability, Sodium fast reactors, Welding

Proceedings Papers

Material Strength Evaluation for 60 Years Design in Japanese Sodium Fast Reactor

Yuji Nagae, Takashi Onizawa, Shigeru Takaya, Takuya Yamashita

Proc. ASME. PVP2014, Volume 1: Codes and Standards, V001T01A063, July 20–24, 2014

Paper No: PVP2014-28689

DOI: https://doi.org/10.1115/PVP2014-28689

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Topics: Design, Sodium fast reactors, Strength (Materials), Steel, Stress, Temperature, Boilers, Forgings (Products), Fracture (Materials), Fracture (Process)

Proceedings Papers

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (3) Guidelines on Structural Reliability Evaluation for FBR

Shigeru Takaya, Hideo Machida, Yoshio Kamishima

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, V001T07A005, July 7–11, 2014

Paper No: ICONE22-30570

DOI: https://doi.org/10.1115/ICONE22-30570

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Topics: Japan Society of Mechanical Engineers, Reliability, Failure, Breeder reactors, Crack propagation, Creep, Damage, Fatigue, Modeling, Probability

Proceedings Papers

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (4) Joint Efforts of JSME and ASME

Tai Asayama, Shigeru Takaya, Masaki Morishita, Frank Schaaf

Paper No: ICONE22-30572

DOI: https://doi.org/10.1115/ICONE22-30572

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Topics: Japan Society of Mechanical Engineers, ASME Boiler and Pressure Vessel Code, Flow (Dynamics), In-service inspection, Liquid metals, Mechanical engineers, Safety, Sodium fast reactors

Proceedings Papers

Evaluation of Fatigue Strength of Similar and Dissimilar Welded Joints of Modified 9Cr-1Mo Steel

Shigeru Takaya

Paper No: ICONE22-30022

DOI: https://doi.org/10.1115/ICONE22-30022

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Topics: Fatigue strength, Steel, Welded joints, Fatigue life, Base metals, Breeder reactors, Evaluation methods, Failure, Fatigue, Mechanical properties

Journal Articles

Development of Creep-Fatigue Evaluation Method for Modified 9Cr-1Mo Steel

Shigeru Takaya, Yuji Nagae, Tai Asayama

J. Pressure Vessel Technol. June 2014, 136(3): 031404.

Paper No: PVT-13-1199

DOI: https://doi.org/10.1115/1.4026497

Published Online: February 27, 2014

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Topics: Creep, Fatigue, Relaxation (Physics), Stress, Steel, Failure, Ductility, Evaluation methods

Journal citation

NARROW

A Weakly Supervised Time Series Analysis Framework for Anomaly Detection in Nuclear Power Plants

Proposal of Detailed Procedures of Determining Rational In-Service Inspection Requirements Based on System Based Code Concept

Development of In-Service Inspection Rules for Liquid-Metal Cooled Reactors Using the System Based Code Concept

Evaluation Method of Response Reliability During an Accident and its Applicability to Fast Reactor Plants

Proposal of a Simple Evaluation Method for Sloshing Impact Pressure on Flat Roofs

Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow

Numerical Analysis of Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Advanced Loop-Type Sodium Cooled Fast Reactor

Implementation of Reliability Evaluation Into JSME Fast Reactor Codes: 1 — Current Status and Path Forward

Implementation of Reliability Evaluation Into JSME Fast Reactor Codes: 2 — Development of a Code Case for Seismic Buckling Evaluation Based on LRFD

Development of Integrated Numerical Analysis Model for Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Japan Sodium-Cooled Fast Reactor

Study on Minimum Wall Thickness Requirement for Seismic Buckling of Reactor Vessel Based on System Based Code Concept

Evaluation of Fatigue Strength of Similar and Dissimilar Welded Joints of Modified 9Cr–1Mo Steel

Development of Creep–Fatigue Evaluation Method for 316FR Stainless Steel

Application of the System Based Code Concept to the Determination of In-Service Inspection Requirements

Development of Structural Codes for JSFR Based on the System Based Code Concept

Material Strength Evaluation for 60 Years Design in Japanese Sodium Fast Reactor

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (3) Guidelines on Structural Reliability Evaluation for FBR

Elaboration of the System Based Code Concept — Activities in JSME and ASME: (4) Joint Efforts of JSME and ASME

Evaluation of Fatigue Strength of Similar and Dissimilar Welded Joints of Modified 9Cr-1Mo Steel

Development of Creep-Fatigue Evaluation Method for Modified 9Cr-1Mo Steel

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