17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics
Xi’an, Shaanxi, China, Sept. 3-8, 2017

Past NURETH Meeting

  • 2015 16th
  • Chicago, USA
  • 2013 15th
  • Pisa, Italy
  • 2011 14th
  • Toronto, CANADA
  • 2009 13th
  • Kanazawa, JAPAN
  • 2007 12th
  • Pittsburgh, USA
  • 2005 11th
  • Avignon, FRANCE
  • 2003 10th
  • Seoul, KOREA
  • 1999 9th
  • San Francisco, USA
  • 1997 8th
  • Kyoto, JAPAN
  • 1995 7th
  • Saratoga Springs, USA
  • 1993 6th
  • Grenoble, FRANCE
  • 1992 5th
  • Salt Lake City, USA
  • 1989 4th
  • Karlsruhe, Germany
  • 1985 3rd
  • Newport, USA
  • 1983 2nd
  • Santa Barbara, USA
  • 1980 1st
  • Saratoga Springs, USA

Click here to download the NURETH-17 Copy Right form.

Please fill out this form, rename this file with your paper ID, and send a scan copy to nureth17@xjtu.edu.cn by June 23, 2017.

 

  Important Date



The proceeding materials will be distributed in a flash drive. The limit for NURETH‐17 paper submissions is 14 pages and supplemented with a file size less than 10MB. 

 Selected papers will be published on the Special issues of NED, NT, etc.     

Please access our meeting system and submit your abstract to preferred track and session via: http://epsr.ans.org/meeting/?m=237.  
The Conference reserves the right to make final track/session adjustment.


The Abstract Template and Full Paper Template are avaible below:
    abstract_template.docx
    full_paper_template.docx

 

 


To download the General Call for Paper of NURETH-17 please clink the link:    Call for Papers

下载《NURETH-17论文发表保密审查意见表》:《NURETH-17论文发表保密审查意见表》



The Call for Papers for Special Topics and Mini-Symposium can be found below:

Special Topics


1. Heat Transfer and Other Potential Thermal-Hydraulic Issues Involving ATF
2. CFD Model and Benchmarking in Subchannel Systems
3. Rod Bundle CHF and Mixing Experiments
4. Thermal-Hydraulic Issues Related to Reactor Aging and Life Extension
5. Post Accident Rod Bundle Thermal- Hydraulic Behaviors
6. FHR & MSR
7. Containment Safety Experiments
8. Ocean Condition Thermal-Hydraulic
9. Measurement and Instrumentation for Thermal-Hydraulic Experiments
10. OECD R&D Status Report
11. CFD Modeling and Validation for Multi-phase Flows in Nuclear Reactor Systems
12. Space Reactor Thermal Hydraulics
13. Status of EASY Program 

14. Fukushima Related Topics
15. RELAP Codes: Past, Present, and future

 

  


Mini-Symposium


1.       SMR
2.       Fast Reactors
3.       SCWR
4.       V/V & UQ
 



Tracks and Sesssions

A.   TWO-PHASE FLOW AND HEAT TRANSFER FUNDAMENTALS

1.       Multifield Two-Phase Flow Modeling
2.       Computational Two-phase Flow
3.       Contact Angle and Wettability Phenomena
4.       Scaling Methods
5.       Two-phase Flow Experiments and Modeling
6.       Fluid Structure Interaction
7.       Supercritical Fluids Thermal Hydraulics
8.       Interfacial Area (data base, modeling, measurement techniques)
9.       Two-phase Flow Instrumentation Techniques
10.    Micro-scale Basic Phenomena, Fluid Flow and Heat Transfer
11.    Nano-Fluids Phenomena
 

B.   CODE DEVELOPMENTS AND APPLICATIONS

1.       Computational Fluid Dynamics and Verification/Validation/Applications (DNS, LES, RANS, etc.)
2.       Computational Multi-Fluid Dynamics and Validation/Verification/Applications
3.       Core Thermal-Hydraulics and Subchannel Analysis
4.       System Codes Development and Assessment
5.       Boron Dilution/Mixing
6.       Aerosol Transport, Deposition and Re- Entrainment
7.       Steam Generators Thermal-Hydraulics
8.       Containment Analysis
9.       Diffuse Interface methods and Interface Tracking Methods
10.     Uncertainties Analysis
11.     Experiments and Data Bases for Assessment and Verification

C.   SEVERE ACCIDENTS AND FIRES

1.       Molten Core Natural Convection and Physico-Chemical Phenomena, Modeling and Experiments
2.       Natural Convection and Mixing Phenomena, Modeling and Experiments
3.       In-Vessel Retention and Coolability
4.       Fuel Coolant Interaction, Modeling and Experiments
5.       Ex-Vessel Core Catchers and Ex-Vessel Cooling
6.       Fission Products Transport, Modeling and Experiments
7.       Direct Containment Heating by Dispersed Molten Fuel
8.       Debris Bed Cooling
9.       Combustion and Fires, Modeling and Experiments
 

D.   ADVANCED MODELLING AND COUPLING

1.       Fast Transient Modelling and Experiments
2.       Enhanced Near–Wall Flow and Heat Transfer Modeling
3.       Micro-, Meso- and Macro-Scale Coupling
4.       Neutronics/Thermal-Hydraulics Coupling
5.       Thermal-Hydraulic Coupling with Crud or Debris Modeling
6.       Thermal or Mechanical Interaction of Fluids and Structures
7.       Thermal-Hydraulic Dependent Corrosion, Erosion and Ablation
 

E.    OPERATION AND SAFETY OF EXISTING REACTORS

1.       Instabilities and Nonlinear Dynamics
2.       NPP Transients and Accidents Analysis
3.       Operating Water Reactor Thermal Hydraulics and Safety
4.       RBMK and VVER Safety Analysis

 

F.  EXPERIMENTAL THERMAL- HYDRAULICS

1.       Boiling Heat Transfer
2.       CHF and Post CHF Heat Transfer
3.       Condensation Heat Transfer with/without Non-Condensable Gases
4.       Flooding and CCFL
5.       Integral Testing
6.       Vibrations, Wear and Thermal Fatigue Phenomena
7.       Fast Reactor Experimental Thermal-Hydraulics
 

G.   ADVANCED REACTORS THERMAL- HYDRAULICS (GEN III+, -IV, INPRO and FUSION)

1.       Sodium Cooled Fast Reactors
2.       Small and Medium Reactors with/without On-Site Refueling
3.       Advanced PWRs, Advanced BWRs, Advanced CANDU Reactors
4.       Gas Cooled Fast Reactors and Very High Temperature Reactors
5.       Brayton Cycle Based Power Conversion Systems
6.       Accelerator Driven Reactors
7.       Lead and Lead-Bismuth Cooled Reactors
8.       Supercritical Water Reactors
9.       Molten Salt Reactors
10.     Fusion Reactors
 

H.   WASTE MANAGEMENT THERMAL- HYDRAULICS

1.       Long Term Storage Thermal Hydraulics
2.       Subsurface Repository Thermal Hydraulics
3.       Waste Management Processes Flow and Heat Transfer
4.       Deep Geological Repository Thermal- Hydraulics and Mass-Transfer
 

I.    THERMAL-HYDRAULICS OF NON ELECTRICITY GENERATING NUCLEAR SYSTEMS

1.       Nuclear Systems for Propulsion and Space Applications
2.       Research/Test/Demonstration Reactors
3.       District Heating
4.       Desalination
5.       Hydrogen Producing Nuclear Reactors
 


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