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活动简介

In October 2016, the 11th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operations and Safety (NUTHOS-11) will take place in Gyeongju, Republic of Korea. The NUTHOS-11 is sponsored by the Korean Nuclear Society (KNS) and co-sponsored by the American Nuclear Society (ANS) in cooperation with other international nuclear societies including the Chinese Nuclear Society, the Canadian Nuclear Society, and others. We are well aware that the accident of the Fukushima Daiichi nuclear plant has been the focus of attention for the last several years. We learned invaluable lessons that the consequences and impacts of such severe nuclear accidents have trans-boundary features and also that the radiological damages can be handed from one generation to future generations. We must enhance nuclear safety based on the lessons learned from the accident without compromising the economy of clean and reliable nuclear power. Even after the Fukushima accident, nuclear energy remains critically important to cope with climate change. We need to redouble research and development efforts to achieve safety, efficiency, and economy of nuclear plant operation.

NUTHOS keeps you abreast of the most updated information in the advancement of science and technology in nuclear thermal hydraulics, operations and safety. NUTHOS aims at addressing key outstanding issues of future technical needs and direction of research to meet challenges in these areas.

The first NUTHOS was held in Taipei in 1984 followed by Tokyo (1986), Seoul (1988), Taipei (1994), Beijing (1997), Nara (2004), Seoul (2008), Shanghai (2010), Kaohsiung (2012) and Okinawa (2014). The success of the previous conferences clearly indicates the high interest of the international nuclear community in the NUTHOS and has led to the establishment of NUTHOS as one of the major international nuclear conferences. We believe that your researches and sharing of ideas in the discussions would contribute to the global nuclear safety.

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征稿范围

Fundamental Thermal-Hydraulics

  • Single- and Two-Phase Flow Fundamentals

  • Sub-channal Flow Dynamics and Analysis

  • CHF and post-CHF Heat Transfer

  • Boiling and Condensation Heat Transfer

  • Nano-Fluid Thermal-Hydaulics

 

Computational Thermal-Hydraulics

  • Advances in Numerical Methods and Modeling

  • Code Development and V&V

  • CFD Application to Nuclear Reactor System Design and Safety Analysis

  • Multi-Scale and Multi-Physics Calculations

  • Accuracy and Uncertainty Analysis including CFD Codes

 

Experimental Thermal-Hydraulics

  • Experiments for Code Verification and Validation

  • Containment Tests and Analysis

  • Steam Generator Thermal-Hydraulic Experiment and Analysis

  • Passive System Performance Test and Analysis

 

Multi-disciplinary Thermal-Hydraulics

  • Thermal Hydraulic Loads and Flow-Induced Vibration

  • Thermal Hydraulics Related to Nuclear Fuel Safety

  • Thermal Hydraulics with Materials and Water Chemistry

  • Multi-Physics Experiments and Analysis

 

Severe Accidents

  • Severe Accident Analysis and Accident Management

  • Degraded Core Thermal Hydraulics

  • Fuel-Coolant Interaction and Steam Explosion

  • Hydrogen and Fission Product Behavior

  • Advanced Design Features for Severe Accident Mitigation

 

Plant Operation and Maintenance

  • Plant Transients and Accidents Analysis and Testing

  • Plant Licensing Renewal and Life Extension

  • Plant Reliability Improvement and Safety Culture

  • Risk-Informed and Performance-Based Regulation

  • Application of Big Data for Plant Diagnosis

 

Plant Digonostics and Monitoring

  • Steam Generator Operation and Maintenance

  • Plant Simulators, Analyzers, Operator Training

  • PRA Applications to Design, Operation and Maintenance

  • Waste Management and Spent Fuel Pool Thermal Hydraulics

  • Environmental Nuclear Safety

 

Advances in Measurements and Instrumentations

  • Advanced Flow Visualization Techniques

  • Advanced Radiation Measurement Techniques

  • Advanced Measurement Techniques for Non-Aquous and Extreme Environment

  • Application of Innovative Measurement Technology

 

Thermal-Hydrulics and Safety of Advaced Reactors

  • SFR Thermal Hydraulics, Design and Safety Analysis

  • Gas Cooled Reactor Thermal Hydraulics, Design and Safety Analysis

  • MSR Thermal Hydraulics, Design and Safety Analysis

  • Supercritical Reactor Thermal Hydraulics

  • Research Reactor Operation and Thermal Hydraulics

  • SCO2-Based Technology

 

Special Session

  • Containment Thermal-Hydraulics

  • Hydrogen Management

  • SMR

  • Safety Analysis of Design Extension Conditions

  • Coolability of Damaged Fuels

  • Scaling Issues

  • CANDU Thermal-hydraulics and Safety 10.8  Filtered Containment Venting System (FCVS)

  • OECD/NEA International Programs

  • To be added

 

Invited Panel Sessions
 

Tutorials

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重要日期

  • 会议日期

    10月09日

    2016

    至

    10月13日

    2016

  • 10月13日 2016

    注册截止日期

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American Nuclear Society (ANS)
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American Nuclear Society (ANS)
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