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Proceedings Papers

Experimental and Numerical Investigation on Debris Bed Quenching With Additional Injection of Non-Condensable Gas

Markus Petroff, Rudi Kulenovic, Jörg Starflinger

Proc. ASME. ICONE28, Volume 2: Nuclear Fuels, Research, and Fuel Cycle; Nuclear Codes and Standards; Thermal-Hydraulics, V002T07A031, August 4–6, 2021

Paper No: ICONE28-65512

DOI: https://doi.org/10.1115/ICONE28-65512

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Topics: Quenching (Metalworking), Particulate matter, Accidents, Cavities, Computer simulation, Concretes, Databases, Flow (Dynamics), Gases, Light water reactors

Proceedings Papers

Development and Assessment of Data-Driven Digital Twins in a Nearly Autonomous Management and Control System for Advanced Reactors

Linyu Lin, Pascal Rouxelin, Paridhi Athe, Nam Dinh, Jeffrey Lane

Proc. ASME. ICONE2020, Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering, V001T06A027, August 4–5, 2020

Paper No: ICONE2020-16813

DOI: https://doi.org/10.1115/ICONE2020-16813

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Topics: Control systems, Uncertainty, Databases, Flow (Dynamics), Modeling, Acceptance criteria, Accidents, Artificial neural networks, Design, Errors

Proceedings Papers

Inter-Unit Common Cause Failure Analysis Based on Data From Intra-Unit Cases

Eishiro Higo, Shota Soga, Hiromichi Miura

Proc. ASME. ICONE2020, Volume 2: Nuclear Policy; Nuclear Safety, Security, and Cyber Security; Operating Plant Experience; Probabilistic Risk Assessments; SMR and Advanced Reactors, V002T10A010, August 4–5, 2020

Paper No: ICONE2020-16819

DOI: https://doi.org/10.1115/ICONE2020-16819

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Topics: Failure, Failure analysis, Databases, Probability, Nuclear power stations, Performance, Probabilistic risk assessment

Proceedings Papers

Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident

Linyu Lin, Paridhi Athe, Pascal Rouxelin, Nam Dinh, Jeffrey Lane

Paper No: ICONE2020-16908

DOI: https://doi.org/10.1115/ICONE2020-16908

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Topics: Accidents, Control systems, Flow (Dynamics), Workflow, Databases, Safety, Breeder reactors, Decision making, Design, Engines

Proceedings Papers

Two-Phase Flow Regime Identification Using Fluctuating Force Signals Under Machine Learning Techniques

Yuta Saito, Shuhei Torisaki, Shuichiro Miwa

Proc. ASME. ICONE26, Volume 9: Student Paper Competition, V009T16A019, July 22–26, 2018

Paper No: ICONE26-81288

DOI: https://doi.org/10.1115/ICONE26-81288

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Topics: Machine learning, Signals, Two-phase flow, Flow (Dynamics), Artificial neural networks, Support vector machines, Algorithms, Databases, Design, Heat

Proceedings Papers

Validation of CTF Void Predictions Using the BFBT Database

Nathan W. Porter, Maria N. Avramova, Vincent A. Mousseau

Proc. ASME. ICONE26, Volume 9: Student Paper Competition, V009T16A059, July 22–26, 2018

Paper No: ICONE26-81869

DOI: https://doi.org/10.1115/ICONE26-81869

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Topics: Databases, Equilibrium (Physics), Geometry, Porosity, Boundary-value problems, Calibration, Design, Modeling, Oscillations, Steady state

Proceedings Papers

Modeling of Low and High Void Fraction Two-Phase Flow in a Vertical Pipe by Using the Two-Fluid Model

Shimo Yu, Xiao Yan, Junyi Zhang

Proc. ASME. ICONE26, Volume 8: Computational Fluid Dynamics (CFD); Nuclear Education and Public Acceptance, V008T09A019, July 22–26, 2018

Paper No: ICONE26-81583

DOI: https://doi.org/10.1115/ICONE26-81583

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Topics: Fluids, Modeling, Pipes, Porosity, Two-phase flow, Bubbles, Computational fluid dynamics, Databases, Flow (Dynamics), Heat transfer

Proceedings Papers

Design of a Closely-Spaced Rod Bundle for a Reference Direct Numerical Simulation

Afaque Shams, Tomasz Kwiatkowski

Proc. ASME. ICONE26, Volume 8: Computational Fluid Dynamics (CFD); Nuclear Education and Public Acceptance, V008T09A002, July 22–26, 2018

Paper No: ICONE26-81049

DOI: https://doi.org/10.1115/ICONE26-81049

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Topics: Computer simulation, Design, Fuel rods, Flow (Dynamics), Computational fluid dynamics, Reynolds-averaged Navier–Stokes equations, Heat transfer, Computation, Databases, Fluids

Proceedings Papers

First Steps Towards the Development of a 3D Nuclear Fuel Behavior Solver With OpenFOAM

Alessandro Scolaro, Ivor Clifford, Carlo Fiorina, Andreas Pautz

Proc. ASME. ICONE26, Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, V003T02A051, July 22–26, 2018

Paper No: ICONE26-82381

DOI: https://doi.org/10.1115/ICONE26-82381

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Topics: Nuclear fuels, Temperature, Electrical conductance, Fuels, Materials properties, Cracking (Materials), Creep, Databases, Fracture (Process), Fuel rods

Proceedings Papers

First Attempt to Determine a Critical Heat Flux Correlation for Thermalhydraulic System Codes

Christophe Herer, Antoine Lejosne

Proc. ASME. ICONE26, Volume 6A: Thermal-Hydraulics and Safety Analyses, V06AT08A057, July 22–26, 2018

Paper No: ICONE26-81589

DOI: https://doi.org/10.1115/ICONE26-81589

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Topics: Critical heat flux, Boiling, Fluids, Temperature, Vapors, Nucleate boiling, Uncertainty, Accidents, Cooling, Databases

Proceedings Papers

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

Jun Huang, Junli Gou, Haifu Ma, Jie Fan, Jianqiang Shan

Proc. ASME. ICONE26, Volume 6B: Thermal-Hydraulics and Safety Analyses, V06BT08A014, July 22–26, 2018

Paper No: ICONE26-81945

DOI: https://doi.org/10.1115/ICONE26-81945

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Topics: Boilers, Heat transfer, Computers, Databases, Fluids, Heat exchangers, Nuclear power stations, Pressurized water reactors

Proceedings Papers

Two Stage Data Driven V&V for an Agile Thermohydraulic Analysis Method

Christopher Bennett, Scott Adams, Nick Alexander

Proc. ASME. ICONE26, Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation, V004T15A021, July 22–26, 2018

Paper No: ICONE26-82628

DOI: https://doi.org/10.1115/ICONE26-82628

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Topics: Architects, Boilers, Databases, Design, Dimensions, Engineering disciplines, Nuclear power stations, Nuclear reactors, Safety, Small modular reactors

Proceedings Papers

The International Experimental Thermal Hydraulic Systems Database (TIETHYS): A New NEA Validation Tool

Upendra Rohatgi, James Dyrda, Nicolas Soppera

Proc. ASME. ICONE26, Volume 4: Nuclear Safety, Security, and Cyber Security; Computer Code Verification and Validation, V004T15A022, July 22–26, 2018

Paper No: ICONE26-82631

DOI: https://doi.org/10.1115/ICONE26-82631

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Topics: Databases, Hydraulic drive systems, Nuclear reactors, Relational databases, Boiling water reactors, Computational fluid dynamics, Gas cooled reactors, High temperature, Instrumentation, Liquid metals

Proceedings Papers

Experimental and Numerical Analysis of Steam Generator Tube Rupture Event for MYRRHA Reactor in CIRCE Facility With SIMMER-IV Code

Alessio Pesetti, Mariano Tarantino, Nicola Forgione

Proc. ASME. ICONE26, Volume 5: Advanced Reactors and Fusion Technologies; Codes, Standards, Licensing, and Regulatory Issues, V005T05A028, July 22–26, 2018

Paper No: ICONE26-82503

DOI: https://doi.org/10.1115/ICONE26-82503

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Topics: Boilers, Numerical analysis, Rupture, Vessels, Pressure, Water, Coolants, Design, Flow (Dynamics), Liquid metals

Proceedings Papers

Real-Time Online Risk Monitoring and Management Method for Maintenance Optimization in Nuclear Power Plant

Anqi Xu, Zhijian Zhang, HuaZhi Zhang, Min Zhang, He Wang, Yingfei Ma, Sijuan Chen, Yan Wang, Gangyang Zheng

Proc. ASME. ICONE26, Volume 2: Plant Systems, Structures, Components, and Materials; Risk Assessments and Management, V002T14A021, July 22–26, 2018

Paper No: ICONE26-82472

DOI: https://doi.org/10.1115/ICONE26-82472

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Topics: Maintenance, Nuclear power stations, Optimization, Risk monitoring, Risk, Risk assessment, Errors, Modeling, Computational methods, Control systems

Proceedings Papers

Research Concept of Decommissioning Knowledge Management for the Fugen NPP

Yasuyoshi Taruta, Satoshi Yanagihara, Yukihiro Iguchi, Koichi Kitamura, Masashi Tezuka, Yuya Koda

Proc. ASME. ICONE26, Volume 7: Decontamination and Decommissioning, Radiation Protection, and Waste Management; Mitigation Strategies for Beyond Design Basis Events, V007T10A005, July 22–26, 2018

Paper No: ICONE26-81228

DOI: https://doi.org/10.1115/ICONE26-81228

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Topics: Knowledge management, Nuclear decommissioning, Nuclear power stations, Databases, Accidents, Nuclear power, Construction, Contamination, Decontamination, Design

Proceedings Papers

ALFRED Steam Generator Assessment: Design and Pre-Test Analysis of HERO Experiment

Pierdomenico Lorusso, Alessio Pesetti, Mariano Tarantino

Proc. ASME. ICONE26, Volume 6B: Thermal-Hydraulics and Safety Analyses, V06BT08A006, July 22–26, 2018

Paper No: ICONE26-81824

DOI: https://doi.org/10.1115/ICONE26-81824

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Topics: Boilers, Design, Water, Engineering simulation, High pressure (Physics), Safety, Separation (Technology), Simulation, Circuits, Coolants

Proceedings Papers

Experimental Activities on Thermal Hydraulics of Heavy Liquid Metal Flow in Typical Elements of Nuclear Power Stations

O. Kashinsky, A. Kurdyumov, P. Lobanov, A. Svetonosov, M. Vorobyev, N. Pribaturin

Proc. ASME. ICONE26, Volume 6B: Thermal-Hydraulics and Safety Analyses, V06BT08A055, July 22–26, 2018

Paper No: ICONE26-82407

DOI: https://doi.org/10.1115/ICONE26-82407

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Topics: Flow (Dynamics), Liquid metals, Nuclear power stations, Thermal hydraulics, Coolants, Temperature, Junctions, Rods, Manufacturing, Computational fluid dynamics

Proceedings Papers

Modeling and Simulating Supercritical CO2 Brayton Cycle in SMR Using Modelica

Yinglin Yang, Qixun Guo, Jianshu Lin, Yaoli Zhang, Kai Ye, BoShen Bian, Zhuocheng Li

Proc. ASME. ICONE25, Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application, V003T13A009, July 2–6, 2017

Paper No: ICONE25-66649

DOI: https://doi.org/10.1115/ICONE25-66649

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Topics: Brayton cycle, Modeling, Supercritical carbon dioxide, Carbon dioxide, Fluids, Pressure, Simulation, Temperature, Compression, Computer simulation

Proceedings Papers

Application of GIS in the General Layout of a Nuclear Power Plant

Peng Fu, Ming Wang

Proc. ASME. ICONE25, Volume 3: Nuclear Fuel and Material, Reactor Physics and Transport Theory; Innovative Nuclear Power Plant Design and New Technology Application, V003T13A012, July 2–6, 2017

Paper No: ICONE25-67164

DOI: https://doi.org/10.1115/ICONE25-67164

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Topics: Nuclear power stations, Design, Databases, Nuclear construction, Nuclear design, Optimization, Safety, Three-dimensional modeling

2 3 4 5

NARROW

Experimental and Numerical Investigation on Debris Bed Quenching With Additional Injection of Non-Condensable Gas

Development and Assessment of Data-Driven Digital Twins in a Nearly Autonomous Management and Control System for Advanced Reactors

Inter-Unit Common Cause Failure Analysis Based on Data From Intra-Unit Cases

Development and Assessment of a Nearly Autonomous Management and Control System During a Single Loss of Flow Accident

Two-Phase Flow Regime Identification Using Fluctuating Force Signals Under Machine Learning Techniques

Validation of CTF Void Predictions Using the BFBT Database

Modeling of Low and High Void Fraction Two-Phase Flow in a Vertical Pipe by Using the Two-Fluid Model

Design of a Closely-Spaced Rod Bundle for a Reference Direct Numerical Simulation

First Steps Towards the Development of a 3D Nuclear Fuel Behavior Solver With OpenFOAM

First Attempt to Determine a Critical Heat Flux Correlation for Thermalhydraulic System Codes

Development of a Thermal-Hydraulic Analysis Code for Helical Coiled Once Through Steam Generator

Two Stage Data Driven V&V for an Agile Thermohydraulic Analysis Method

The International Experimental Thermal Hydraulic Systems Database (TIETHYS): A New NEA Validation Tool

Experimental and Numerical Analysis of Steam Generator Tube Rupture Event for MYRRHA Reactor in CIRCE Facility With SIMMER-IV Code

Real-Time Online Risk Monitoring and Management Method for Maintenance Optimization in Nuclear Power Plant

Research Concept of Decommissioning Knowledge Management for the Fugen NPP

ALFRED Steam Generator Assessment: Design and Pre-Test Analysis of HERO Experiment

Experimental Activities on Thermal Hydraulics of Heavy Liquid Metal Flow in Typical Elements of Nuclear Power Stations

Modeling and Simulating Supercritical CO2 Brayton Cycle in SMR Using Modelica

Application of GIS in the General Layout of a Nuclear Power Plant

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