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Proceedings Papers

Prediction of Pressure Propagation in Piping Systems After ISLOCA

Takeshi Yamada, Kohei Hisamochi, Taichi Takii, Kenichi Yasuda

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A047, July 7–11, 2014

Paper No: ICONE22-31086

DOI: https://doi.org/10.1115/ICONE22-31086

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Topics: Piping systems, Pressure, Pipelines, Transients (Dynamics), Valves, Accidents, Bifurcation, Boiling water reactors, Coolants, Failure

Proceedings Papers

Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor

Yanhua Zheng, Fubing Chen, Lei Shi

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A002, July 7–11, 2014

Paper No: ICONE22-30050

DOI: https://doi.org/10.1115/ICONE22-30050

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Topics: Very high temperature reactors, Cooling, Flow (Dynamics), Fuels, Temperature, Graphite, Safety, Accidents, Design, Helium

Proceedings Papers

Development of an Evaluation Methodology for the Fuel-Relocation Into the Coolant Plenum in the Core Disruptive Accident of Sodium-Cooled Fast Reactors

Kenji Kamiyama, Yoshiharu Tobita, Tohru Suzuki, Ken-ichi Matsuba

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A011, July 7–11, 2014

Paper No: ICONE22-30308

DOI: https://doi.org/10.1115/ICONE22-30308

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Topics: Accidents, Coolants, Fuels, Sodium fast reactors, Model validation, Water, Alloys, Fluid dynamics, Heat, Melting

Proceedings Papers

End-to-End Technology of Modeling a Melt-Structure Interaction During IVMR in VVER With HEFEST-URAN Toolkit

A. S. Filippov, N. I. Drobyshevsky, D. D. Kamenskay, A. E. Kisselev, E. V. Moiseenko

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A020, July 7–11, 2014

Paper No: ICONE22-30546

DOI: https://doi.org/10.1115/ICONE22-30546

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Topics: Modeling, Finite element methods, Finite element model, Stress, Vessels, Accidents, Creep, Erosion, Errors, Numerical analysis

Proceedings Papers

Spent Fuel Pool Under Severe Accident Conditions

Bernd Jaeckel, Jonathan Birchley, Leticia Fernandez-Moguel

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A029, July 7–11, 2014

Paper No: ICONE22-30729

DOI: https://doi.org/10.1115/ICONE22-30729

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Topics: Accidents, Spent nuclear fuels, Flow (Dynamics), Fuels, Coolants, Heat, Performance, Spent fuel storage, Boiling water reactors, Buoyancy

Proceedings Papers

Implications of Applying Semi-Scale Pump Test Data to Two-Phase RCP Performance Modeling for PWR LOCA Analyses

Mathew C. Jacob, Michael T. Coon, John A. Blaisdell, Ruben J. Espinosa

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A038, July 7–11, 2014

Paper No: ICONE22-30847

DOI: https://doi.org/10.1115/ICONE22-30847

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Topics: Modeling, Pressurized water reactors, Pumps, Emergency core cooling systems, Accidents, Coolants, Electricity (Physics), Engineering simulation, Engines, Motors

Proceedings Papers

MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis

Christian Royère, Michel Gonnet, Brahim Manchid

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A048, July 7–11, 2014

Paper No: ICONE22-31101

DOI: https://doi.org/10.1115/ICONE22-31101

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Topics: Safety, Transients (Dynamics), Accidents, Shapes, Steam, Boundary-value problems, Cross-flow, Design, Failure, Feedback

Proceedings Papers

Simulation of Molten Corium Concrete Interaction With the MOCO Code

Bo Lin, Sui-zheng Qiu, Guang-hui Su, Wen-xi Tian, Ya-pei Zhang

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A012, July 7–11, 2014

Paper No: ICONE22-30345

DOI: https://doi.org/10.1115/ICONE22-30345

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Topics: Concretes, Simulation, Water, Accidents, Cavities, Chemistry, Computers, Containment, Cooling, Heat transfer

Proceedings Papers

The Influence of Two-Phase Flow on Pore Clogging by Fine Particle Settlement During Ex-Vessel Debris Bed Formation in Severe Accident

Eunho Kim, Jin Ho Park, Moo Hwan Kim, Hyun Sun Park

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A003, July 7–11, 2014

Paper No: ICONE22-30126

DOI: https://doi.org/10.1115/ICONE22-30126

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Topics: Accidents, Particulate matter, Two-phase flow, Vessels, Cavities, Reactor vessels, Water, Accident management, Bubbles, Convection

Proceedings Papers

Source Term Estimation in Nuclear Accidents Using 4D Variational Data Assimilation for Heterologous Atmospheric Condition

Yun Liu, Sheng Fang, Hong Li, Jingyuan Qu, Dong Fang

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A021, July 7–11, 2014

Paper No: ICONE22-30551

DOI: https://doi.org/10.1115/ICONE22-30551

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Topics: Accidents, Radioisotopes, Stability, Wind velocity, Computer simulation, Emergency response, Modeling, Optimization, Radiation measurement, Robustness

Proceedings Papers

Research on the Local Neutronics Thermal-Hydraulic Coupled Response of a PWR During Hypothetical Blockage Scenarios

Xiaorong Li, Shinian Peng

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A039, July 7–11, 2014

Paper No: ICONE22-30911

DOI: https://doi.org/10.1115/ICONE22-30911

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Topics: Feedback, Flow (Dynamics), Fuels, Manufacturing, Pressurized water reactors, Thermal hydraulics, Transients (Dynamics)

Proceedings Papers

An Alternative Approach to Prediction of Critical Heat Flux: Projection Support Vector Regression

Botao Jiang, Fuyu Zhao

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A030, July 7–11, 2014

Paper No: ICONE22-30747

DOI: https://doi.org/10.1115/ICONE22-30747

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Topics: Critical heat flux, Support vector machines, Artificial neural networks, Boilers, Boiling, Cooling systems, Design, Fuels, Trains

Proceedings Papers

Dynamic Correcting Dispersion Parameters of Lagrangian Puff Model in Atmospheric Tracer Experiments

Yuanwei Ma, Dezhong Wang, Zhilong Ji, Nan Qian

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A013, July 7–11, 2014

Paper No: ICONE22-30347

DOI: https://doi.org/10.1115/ICONE22-30347

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Topics: Accidents, Computer simulation, Errors, Genetic algorithms

Proceedings Papers

On Features of Thermal Design of Passive Evaporation-and-Condensation Systems of Reactor Thermal Shielding

A. N. Gershuni, A. P. Nishchik, E. N. Pis'mennyi, V. G. Razumovskiy, I. L. Pioro

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A004, July 7–11, 2014

Paper No: ICONE22-30133

DOI: https://doi.org/10.1115/ICONE22-30133

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Topics: Condensation, Design, Evaporation, Heat shielding, Cavities, Heat, Safety, Containment, Emergencies, Nuclear engineering

Proceedings Papers

Development of Inherently Safe Technologies for Large Scale BWRs: (5) Operation Support System for Plant Accidents

Masaki Kanada, Ryota Kamoshida, Yoshihiko Ishii, Tadaaki Ishikawa, Setsuo Arita, Kenichi Katono

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A049, July 7–11, 2014

Paper No: ICONE22-31104

DOI: https://doi.org/10.1115/ICONE22-31104

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Topics: Accidents, Boiling water reactors, Sensors, Signals, Simulation, Errors, Natural disasters, Nuclear reactors, Plant design, Simulation results

Proceedings Papers

RELAP5 Code Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure: Investigation on Base Case Result Appropriate for the Best Estimate Plus Uncertainty (BEPU) Application

Ikuo Kinoshita, Toshihide Torige, Michio Murase, Yoshitaka Yoshida, Takeshi Takeda, Akira Satou, Hideo Nakamura

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A022, July 7–11, 2014

Paper No: ICONE22-30559

DOI: https://doi.org/10.1115/ICONE22-30559

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Topics: Accident management, Boilers, Uncertainty, Accidents, Condensation, Coolants, Flow (Dynamics), High pressure (Physics), Pressure, Simulation analysis

Proceedings Papers

Influence of the Temperature History on Secondary Hydriding and Mechanical Properties of Zircaloy-4 Claddings: An Analysis of the QUENCH-LOCA Bundle Tests

J. Stuckert, M. Große, C. Rössger, M. Steinbrück, M. Walter

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A031, July 7–11, 2014

Paper No: ICONE22-30792

DOI: https://doi.org/10.1115/ICONE22-30792

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Topics: Mechanical properties, Temperature, Hydrogen, Cladding systems (Building), Transients (Dynamics), Embrittlement, Accidents, Cooling, Design, Fluids

Proceedings Papers

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

Pavlin P. Groudev, Antoaneta E. Stefanova, Petya I. Vryashkova

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A040, July 7–11, 2014

Paper No: ICONE22-30941

DOI: https://doi.org/10.1115/ICONE22-30941

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Topics: Quenching (Metalworking), Fuels, Accidents, Reactor vessels, Accident management, Cladding systems (Building), Melting, Modeling, Porosity, Temperature

Proceedings Papers

Experience Gained From the Living PSA Project for Dukovany NPP

Stanislav Hust’ák

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A014, July 7–11, 2014

Paper No: ICONE22-30391

DOI: https://doi.org/10.1115/ICONE22-30391

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Topics: Nuclear power stations, Computer software, Design, Event history analysis, Model development, Probabilistic safety assessment, Risk assessment, Risk monitoring, Risk-based decision making, Safety

Proceedings Papers

To Analyze the Core Damage Scenario via the RISMC-Based Metatheory and Reliability Theory

Gangyang Zheng, Yu Gong, Zhijian Zhang, Zibin Liu

Proc. ASME. ICONE22, Volume 3: Next Generation Reactors and Advanced Reactors; Nuclear Safety and Security, V003T06A023, July 7–11, 2014

Paper No: ICONE22-30564

DOI: https://doi.org/10.1115/ICONE22-30564

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Topics: Damage, Reliability theory, Nuclear power stations, Safety, Reliability, Risk, Risk-based design

2 3 4 5

NARROW

Prediction of Pressure Propagation in Piping Systems After ISLOCA

Analysis of Bypass Effect on a 250 MW High Temperature Gas-Cooled Reactor

Development of an Evaluation Methodology for the Fuel-Relocation Into the Coolant Plenum in the Core Disruptive Accident of Sodium-Cooled Fast Reactors

End-to-End Technology of Modeling a Melt-Structure Interaction During IVMR in VVER With HEFEST-URAN Toolkit

Spent Fuel Pool Under Severe Accident Conditions

Implications of Applying Semi-Scale Pump Test Data to Two-Phase RCP Performance Modeling for PWR LOCA Analyses

MTC 3D: A Fully Coupled Methodology for MSLB Safety Analysis

Simulation of Molten Corium Concrete Interaction With the MOCO Code

The Influence of Two-Phase Flow on Pore Clogging by Fine Particle Settlement During Ex-Vessel Debris Bed Formation in Severe Accident

Source Term Estimation in Nuclear Accidents Using 4D Variational Data Assimilation for Heterologous Atmospheric Condition

Research on the Local Neutronics Thermal-Hydraulic Coupled Response of a PWR During Hypothetical Blockage Scenarios

An Alternative Approach to Prediction of Critical Heat Flux: Projection Support Vector Regression

Dynamic Correcting Dispersion Parameters of Lagrangian Puff Model in Atmospheric Tracer Experiments

On Features of Thermal Design of Passive Evaporation-and-Condensation Systems of Reactor Thermal Shielding

Development of Inherently Safe Technologies for Large Scale BWRs: (5) Operation Support System for Plant Accidents

RELAP5 Code Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure: Investigation on Base Case Result Appropriate for the Best Estimate Plus Uncertainty (BEPU) Application

Influence of the Temperature History on Secondary Hydriding and Mechanical Properties of Zircaloy-4 Claddings: An Analysis of the QUENCH-LOCA Bundle Tests

MELCOR Study of VVER-1000 Behavior in Case of Overheated Reactor Core Quenching

Experience Gained From the Living PSA Project for Dukovany NPP

To Analyze the Core Damage Scenario via the RISMC-Based Metatheory and Reliability Theory

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