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Journal Articles

Investigation on Applicability of Subchannel Analysis Code ASFRE to Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure of Sodium Cooled Fast Reactor

Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. July 2023, 9(3): 031401.
Paper No: NERS-21-1158
https://doi.org/10.1115/1.4056463
Published Online: February 8, 2023
View Article titled, Investigation on Applicability of Subchannel Analysis Code ASFRE to Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure of Sodium Cooled Fast Reactor
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Topics: Ducts, Flow (Dynamics), Fuels, Manufacturing, Sodium fast reactors, Temperature, Temperature distribution, Thermal hydraulics, Wire, Fuel rods
Journal Articles

Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Kazuo Yoshimura, Norihiro Doda, Kennichi Igawa, Masaaki Tanaka, Hidemasa Yamano
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. April 2023, 9(2): 021601.
Paper No: NERS-21-1160
https://doi.org/10.1115/1.4055328
Published Online: December 14, 2022
View Article titled, Validation of Feedback Reactivity Evaluation Models for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors
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Topics: Feedback, Sodium fast reactors, Temperature, Fuels, Sensitivity analysis, Density, Heat, Heat sinks, Dynamics (Mechanics), Coolants
Proceedings Papers

Application of First-Order Method to Estimate Structural Integrity in a Probabilistic Form of Component Subjected to Thermal Transient for Optimization of Design Parameter

Satoshi Okajima, Takero Mori, Norihiro Kikuchi, Masaaki Tanaka, Masashi Miyazaki
Proc. ASME. ICONE29, Volume 4: SMRs, Advanced Reactors, and Fusion, V004T04A001, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-88933
https://doi.org/10.1115/ICONE29-88933
View Paper titled, Application of First-Order Method to Estimate Structural Integrity in a Probabilistic Form of Component Subjected to Thermal Transient for Optimization of Design Parameter
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Topics: Design, Failure, Optimization, Probability, Transients (Dynamics)
Proceedings Papers

Study on Decay Heat Removal by Dipped-Direct Heat Exchanger in Sodium-Cooled Fast Reactor

Toshiki Ezure, Erina Hamase, Masaaki Tanaka, Antoine Gerschenfeld, Yannick Gorsse, Simon Li
Proc. ASME. ICONE29, Volume 7B: Thermal-Hydraulics and Safety Analysis, V07BT07A034, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-93119
https://doi.org/10.1115/ICONE29-93119
View Paper titled, Study on Decay Heat Removal by Dipped-Direct Heat Exchanger in Sodium-Cooled Fast Reactor
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Topics: Heat, Heat exchangers, Sodium fast reactors
Proceedings Papers

Development of Evaluation Method of Gas Entrainment on the Free Surface in the Reactor Vessel in Pool-Type Sodium-Cooled Fast Reactors -Gas Entrainment Judgment Based on Three-Dimensional Evaluation of Vortex Center Line and Distribution of Pressure Decrease

Kentarou Matsushita, Toshiki Ezure, Yasutomo Imai, Tatsuya Fujisaki, Masaaki Tanaka
Proc. ASME. ICONE29, Volume 8: Computational Fluid Dynamics (CFD) and Applications, V008T08A007, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-90288
https://doi.org/10.1115/ICONE29-90288
View Paper titled, Development of Evaluation Method of Gas Entrainment on the Free Surface in the Reactor Vessel in Pool-Type Sodium-Cooled Fast Reactors -Gas Entrainment Judgment Based on Three-Dimensional Evaluation of Vortex Center Line and Distribution of Pressure Decrease
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Topics: Evaluation methods, Pressure, Reactor vessels, Sodium fast reactors, Vortices
Proceedings Papers

Application of 1D-CFD Coupling Method to Unprotected Loss of Heat Sink Event in EBR-II Focusing on Thermal Stratification in Cold Pool

Kazuo Yoshimura, Norihiro Doda, Tatsuya Fujisaki, Kennichi Igawa, Masaaki Tanaka, Hidemasa Yamano
Proc. ASME. ICONE29, Volume 8: Computational Fluid Dynamics (CFD) and Applications, V008T08A005, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-89259
https://doi.org/10.1115/ICONE29-89259
View Paper titled, Application of 1D-CFD Coupling Method to Unprotected Loss of Heat Sink Event in EBR-II Focusing on Thermal Stratification in Cold Pool
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Topics: Computational fluid dynamics, Heat sinks, Sodium fast reactors, Thermal stratification
Proceedings Papers

Quantitative Risk Assessment With CMMC Method on Abnormal Snowfall Incident for a Sodium-Cooled Fast Reactor

Risako Nakashima, Akari Koike, Takaaki Sakai, Norihiro Doda, Masaaki Tanaka
Proc. ASME. ICONE29, Volume 15: Student Paper Competition, V015T16A107, August 8–12, 2022
Publisher: American Society of Mechanical Engineers
Paper No: ICONE29-93039
https://doi.org/10.1115/ICONE29-93039
View Paper titled, Quantitative Risk Assessment With CMMC Method on Abnormal Snowfall Incident for a Sodium-Cooled Fast Reactor
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Topics: Risk assessment, Sodium fast reactors, Hazards, Meteorology, Temperature, Dynamics (Mechanics), Probability, Chain, Climate change, Coolants
Journal Articles

ARKADIA—For the Innovation of Advanced Nuclear Reactor Design

Hiroyuki Ohshima, Tai Asayama, Tomohiro Furukawa, Masaaki Tanaka, Akihiro Uchibori, Takashi Takata, Akiyuki Seki, Yasuhiro Enuma
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Design Innovation Paper
ASME J of Nuclear Rad Sci. April 2023, 9(2): 025001.
Paper No: NERS-21-1157
https://doi.org/10.1115/1.4054726
Published Online: June 15, 2022
View Article titled, ARKADIA—For the Innovation of Advanced Nuclear Reactor Design
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Topics: Design, Optimization, Safety, Thermal hydraulics, Nuclear design, Nuclear reactors, Computer simulation
Proceedings Papers

Validation of Evaluation Method of Feedback Reactivity for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors

Kazuo Yoshimura, Norihiro Doda, Masaaki Tanaka, Hidemasa Yamano, Kennichi Igawa
Proc. ASME. ICONE28, Volume 3: Computational Fluid Dynamics (CFD); Verification and Validation; Advanced Methods of Manufacturing (AMM) for Nuclear Reactors and Components; Decontamination, Decommissioning, and Radioactive Waste Management; Beyond Design Basis and Nuclear Safety; Risk Informed Management and Regulation, V003T09A001, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-62354
https://doi.org/10.1115/ICONE28-62354
View Paper titled, Validation of Evaluation Method of Feedback Reactivity for Plant Dynamics Analysis Code During Unprotected Loss of Heat Sink Event in Sodium-Cooled Fast Reactors
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Topics: Dynamics (Mechanics), Evaluation methods, Feedback, Heat sinks, Sodium fast reactors
Proceedings Papers

Investigation of Applicability of Subchannel Analysis Code ASFRE on Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure in Sodium Cooled Fast Reactor

Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka
Proc. ASME. ICONE28, Volume 3: Computational Fluid Dynamics (CFD); Verification and Validation; Advanced Methods of Manufacturing (AMM) for Nuclear Reactors and Components; Decontamination, Decommissioning, and Radioactive Waste Management; Beyond Design Basis and Nuclear Safety; Risk Informed Management and Regulation, V003T08A019, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-65662
https://doi.org/10.1115/ICONE28-65662
View Paper titled, Investigation of Applicability of Subchannel Analysis Code ASFRE on Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure in Sodium Cooled Fast Reactor
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Topics: Design, Ducts, Fuels, Manufacturing, Sodium fast reactors, Thermal hydraulics
Proceedings Papers

ARKADIA – for the Innovation of Advanced Nuclear Reactor Design

Hiroyuki Ohshima, Tai Asayama, Tomohiro Furukawa, Masaaki Tanaka, Takashi Takata, Yasuhiro Enuma
Proc. ASME. ICONE28, Volume 1: Operating Plant Challenges, Successes, and Lessons Learned; Nuclear Plant Engineering; Advanced Reactors and Fusion; Small Modular and Micro-Reactors Technologies and Applications, V001T03A007, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-64525
https://doi.org/10.1115/ICONE28-64525
View Paper titled, ARKADIA – for the Innovation of Advanced Nuclear Reactor Design
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Topics: Artificial intelligence, Computer simulation, Innovation, Nuclear design, Nuclear reactors, Optimization, Design, Safety, Carbon, Coolants
Proceedings Papers

Core Thermal-Hydraulic Analysis During Dipped-Type Direct Heat Exchanger Operation in Natural Circulation Conditions

Erina Hamase, Norihiro Doda, Ayako Ono, Masaaki Tanaka, Yasuhiro Miyake, Yasutomo Imai
Proc. ASME. ICONE28, Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, V002T07A005, August 4–6, 2021
Publisher: American Society of Mechanical Engineers
Paper No: ICONE28-63380
https://doi.org/10.1115/ICONE28-63380
View Paper titled, Core Thermal-Hydraulic Analysis During Dipped-Type Direct Heat Exchanger Operation in Natural Circulation Conditions
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Topics: Computational fluid dynamics, Flow (Dynamics), Heat exchangers, Sodium fast reactors, Fuels, Reactor vessels, Sodium, Coolants, Cooling, Cooling systems
Proceedings Papers

Analysis of Gas Entrainment Phenomenon From Free Liquid Surface for a Sodium-Cooled Fast Reactor Design: Validation of Velocity Profile and Strouhal Number in a Flow Field

Mao Uchida, Moe Hirakawa, Aaru Sano, Keisuke Inoue, Takaaki Sakai, Toshiki Ezure, Masaaki Tanaka
Proc. ASME. ICONE2020, Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors, V002T11A009, August 4–5, 2020
Publisher: American Society of Mechanical Engineers
Paper No: ICONE2020-16418
https://doi.org/10.1115/ICONE2020-16418
View Paper titled, Analysis of Gas Entrainment Phenomenon From Free Liquid Surface for a Sodium-Cooled Fast Reactor Design: Validation of Velocity Profile and Strouhal Number in a Flow Field
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Topics: Design, Flow (Dynamics), Sodium fast reactors, Vortices, Computational fluid dynamics, Reactor vessels, Water, Coolants, Evaluation methods, Particulate matter
Journal Articles

Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor

Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka, Hiroyuki Ohshima
Journal: Journal of Nuclear Engineering and Radiation Science
Publisher: ASME
Article Type: Research-Article
ASME J of Nuclear Rad Sci. April 2019, 5(2): 021001.
Paper No: NERS-17-1280
https://doi.org/10.1115/1.4042191
Published Online: March 15, 2019
View Article titled, Subchannel Analysis of Thermal-Hydraulics in a Fuel Assembly With Inner Duct Structure of a Sodium-Cooled Fast Reactor
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Topics: Ducts, Flow (Dynamics), Fuels, Manufacturing, Sodium, Temperature distribution, Temperature, Sodium fast reactors, Wire, Thermal hydraulics
Proceedings Papers

Study on Gas Entrainment From Unstable Drifting Vortexes on Liquid Surface

Moe Hirakawa, Yuichiro Kikuchi, Takaaki Sakai, Masaaki Tanaka, Hiroyuki Ohshima
Proc. ASME. ICONE26, Volume 9: Student Paper Competition, V009T16A091, July 22–26, 2018
Publisher: American Society of Mechanical Engineers
Paper No: ICONE26-82477
https://doi.org/10.1115/ICONE26-82477
View Paper titled, Study on Gas Entrainment From Unstable Drifting Vortexes on Liquid Surface
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Topics: Vortices, Flow (Dynamics), Computational fluid dynamics, Water, Reactor vessels, Sodium fast reactors, Coolants, Design, Design methodology, Evaluation methods
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
Publisher: American Society of Mechanical Engineers
Paper No: PVP2017-65601
https://doi.org/10.1115/PVP2017-65601
View Paper titled, Numerical Analysis of Flow-Induced Vibration of Large Diameter Pipe With Short Elbow
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Topics: Flow-induced vibrations, Numerical analysis, Pipes, Pressure, Design, Stress, Computer simulation, Flow (Dynamics), Construction, Cooling systems
Proceedings Papers

Development of Numerical Estimation Method Using Spatial Connection Methodology for Thermal Striping in Upper Plenum of Reactor Vessel of an Advanced Loop-Type Sodium-Cooled Fast Reactor in Japan

Masaaki Tanaka, Satoshi Murakami
Proc. ASME. ICONE25, Volume 8: Computational Fluid Dynamics (CFD) and Coupled Codes; Nuclear Education, Public Acceptance and Related Issues, V008T09A055, July 2–6, 2017
Publisher: American Society of Mechanical Engineers
Paper No: ICONE25-67876
https://doi.org/10.1115/ICONE25-67876
View Paper titled, Development of Numerical Estimation Method Using Spatial Connection Methodology for Thermal Striping in Upper Plenum of Reactor Vessel of an Advanced Loop-Type Sodium-Cooled Fast Reactor in Japan
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Topics: Reactor vessels, Sodium fast reactors, Flow (Dynamics), Computer simulation, Fuels, Sodium, Fluids, High temperature, Instrumentation, Low temperature
Proceedings Papers

Thermal-Hydraulic Analysis of Fuel Assembly With Inner Duct Structure of an Advanced Loop-Type Sodium-Cooled Fast Reactor Using ASFRE Code

Norihiro Kikuchi, Yasutomo Imai, Ryuji Yoshikawa, Norihiro Doda, Masaaki Tanaka, Hiroyuki Ohshima
Proc. ASME. ICONE25, Volume 8: Computational Fluid Dynamics (CFD) and Coupled Codes; Nuclear Education, Public Acceptance and Related Issues, V008T09A054, July 2–6, 2017
Publisher: American Society of Mechanical Engineers
Paper No: ICONE25-67870
https://doi.org/10.1115/ICONE25-67870
View Paper titled, Thermal-Hydraulic Analysis of Fuel Assembly With Inner Duct Structure of an Advanced Loop-Type Sodium-Cooled Fast Reactor Using ASFRE Code
Open the PDF for Thermal-Hydraulic Analysis of Fuel Assembly With Inner Duct Structure of an Advanced Loop-Type Sodium-Cooled Fast Reactor Using ASFRE Code in another window
Topics: Ducts, Fuels, Manufacturing, Sodium fast reactors, Pins (Engineering), Flow (Dynamics), Temperature distribution, Design, Thermal hydraulics, Turbulence
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
Publisher: American Society of Mechanical Engineers
Paper No: PVP2016-63845
https://doi.org/10.1115/PVP2016-63845
View Paper titled, Numerical Analysis of Unsteady Phenomena in Upper Plenum and Hot-Leg Piping System of Advanced Loop-Type Sodium Cooled Fast Reactor
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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

Effect of Deflected Inflow on Flows in a Strongly-Curved 90 Degree Elbow

Yukiharu Iwamoto, Ryo Kusuzaki, Motosuke Sogo, Kazunori Yasuda, Hidemasa Yamano, Masaaki Tanaka
Proc. ASME. AJKFluids2015, Volume 1A: Symposia, Part 2, V01AT25A008, July 26–31, 2015
Publisher: American Society of Mechanical Engineers
Paper No: AJKFluids2015-25596
https://doi.org/10.1115/AJKFluids2015-25596
View Paper titled, Effect of Deflected Inflow on Flows in a Strongly-Curved 90 Degree Elbow
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Topics: Flow (Dynamics), Inflow, Pressure, Flow visualization, Reynolds number, Collisions (Physics), High pressure (Physics), Pressure measurement, Spectral energy distribution