Technical Program

NURETH-14 Program (pdf 2.8 MB)

The List of Tracks and Sessions planned for NURETH-14

 

A. TWO-PHASE FLOW AND HEAT TRANSFER FUNDAMENTALS
  • Boiling and Condensation Fundamentals
  • Multifield Two-Phase Flow Modeling
  • Contact Angle and Wettability Phenomena
  • Mini Symposium on Flow-induced vibration in nuclear components
  • Supercritical fluids thermal hydraulics
  • Interfacial Area Transport (data base, modeling, measurement techniques)
  • Micro and Nano-Scale Basic Phenomena, Fluid Flow and Heat Transfer
  • Mini symposium on Thermal-hydraulics of non-unity Prandtl number flows
  • Coherent Large Scale Structures in the Gaps of Rod-Bundles
B. CODE DEVELOPMENTS AND APPLICATIONS
  • Computational Fluid Dynamics and Verification/Validation/Applications (DNS, LES, RANS, etc.)
  • Computational Multi-Fluid Dynamics and Validation/Verification/Applications
  • Core Thermal-Hydraulics and Subchannel Analysis
  • PlantSystem Codes Development and Assessment
  • Boron Dilution/Mixing
  • Steam Generators Thermal-Hydraulics
  • Containment Analysis
  • Uncertainties Analysis
  • Experiments and Data Bases for Assessment and Verification of 3D Models
  • Mini-symposium on Pressure Surges in Nuclear Power Plants
  • Development, Assessment and Applications of TRACE
C. SEVERE ACCIDENTS AND FIRES
  • Molten Core Natural Convection and Physico-Chemical Phenomena, Modeling and Experiments
  • Natural Convection and Mixing phenomena, Modeling and Experiments
  • Fuel Coolant Interaction, Modeling and Experiments
  • Direct Containment Heating by Dispersed Molten Fuel
  • Debris Bed Cooling
  • Combustion and Fires, Modeling and Experiments
  • Advanced Design Features for Severe Accident Mitigation
D. ADVANCED CODE DEVELOPMENTS
  • Fast Transient Modelling and Experiments
  • Enhanced Near–Wall Flow and Heat Transfer Modeling
  • Fluid and Structures Mechanical Interactions
  • Multi-scale multi-physics couplings
  • CASL- Thermal-hydraulics Activities in the Consortium for Advanced Simulation of LWRs
E. OPERATION AND SAFETY OF EXISTING REACTORS
  • Plant life extension and power up-rating
  • Instabilities and Nonlinear Dynamics
  • NPP Transients and Accidents Analysis
  • Safety of Sodium cooled RBMK and VVER Reactors
  • Natural Circulation Phenomena and Passive Safety Systems
F. EXPERIMENTAL THERMAL-HYDRAULICS
  • Boiling and Condensation Heat Transfer
  • CHF and Post CHF Heat Transfer, Flooding and CCFL
  • Instrumentation Technique
  • Integral Testing
  • Flow Visualization
G. ADVANCED REACTORS THERMAL-HYDRAULICS (GEN III+, -IV, INPRO and FUSION)
  • Sodium Cooled Fast Reactors Design and Safety
  • Small and Medium Reactors with/without On-Site Refueling
  • Advanced PWRs, Advanced BWRs, Advanced CANDU Reactors
  • Gas Cooled Fast Reactors and Very High Temperature Reactors
  • Lead and Lead-Bismuth Cooled Reactors
  • Supercritical Water Reactors
H. WASTE MANAGEMENT THERMAL-HYDRAULICS
I. THERMAL-HYDRAULICS OF NON ELECTRICITY GENERATING NUCLEAR EQUIPMENT
O. Special Topics (Organized Sessions) include:
  • Session on Thermal Hydraulics and Structural Integrity in Connection to Aging and Life Extension;
  • Session on BEPU (Best Estimate code Plus Uncertainty) method, CSAU, Statistical Methods;
  • Session on the Study of Pressurized Thermal Shock;
  • OECD session on Thermal Hydraulic Benchmarks;
  • OECD/NEA session on Analysis and Management of Accidents (BEMUSE, CFD Studies);
  • IAEA session on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel;
  • Westinghouse session on Full Spectrum LOCA – A Realistic LOCA Evaluation Model Applicable to the Full Range of Break Sizes using WCOBRA/TRAC-TF2 code;
  • Development, Assessment and Applications of TRACE;
  • Radiological Hazard Related Thermal Hydraulics;
  • Mini-symposium on Flow Induced Vibration in Nuclear Components;
  • Mini-symposium on Thermal Hydraulics of Non-Unity Prandtl Number Flows;
  • Mini-symposium on Pressure Surges in Nuclear Power Plants;
  • Thermal Hydraulics Activities in the Consortium for Advanced Simulation of LWRs (CASL).

 

Panels

Scheduled Panel Sessions include:

  • Issues and Future Directions of Thermalhydraulics Research and Developments
    • Panel Organizers: Dr. Pradip Saha, General Electric Co.and Dr. Nusret Aksan, University of Pisa
      Panel Members: Jin Yan, Jens Andersen, Robert B Lowrie, Rich Martineau, and Douglas B. Kothe.
  • Lessons Learned from the Fukushima Accident
    • Panel Organizer: Professor Bill Cheung and Dr. Jovica Riznic Panel Members: Hisashi Ninokata, John Luxat, Michael Corradini, Randall Gauntt, Gerri Frappier, Raj Sehgal.
  • Global Cooperation in Nuclear Education and Research
    • Panel Organizer: Dr. Jong H. Kim, KAIST

 

Keynote Addresses

Scheduled Keynote Addresses include:

  • Thermal Hydraulics of Sodium-cooled Fast Reactors - Key Issues and Highlights: Professor Hisashi Ninokata
  • Experimental and theoretical tools to support the development of high-performance LWR fuel elements: Professor Micheal Prasser
  • Korean development of advanced thermal-hydraulic codes for water reactors and  HTGRs: SPACE and GAMMA: Professor Hee Cheon No
  • Some Challenges in Water-Cooled Reactor Thermal-Hydraulics: Professor Xu Cheng
  • Status and perspective of a multiscale approach to nuclear reactor thermal hydraulic simulation: Dr. Dominique Bestion
  • Development of interfacial area transport equation - modeling and experimental benchmark: Professor Mamoru Ishii
  • Supercritical Flow and Heat Transfer in Advanced Reactors: Professor Michael Corradini
  • Thermal-Hydraulics and Safety Concepts of Supercritical Water Cooled Reactors: Professor Thomas Schulenberg

 

For additional information about the NURETH-14 Technical Program, please contact the Technical Program Chair at:  jovica.riznic@cnsc-ccsn.gc.ca