Volumes | International Conference on Nuclear Engineering | American Society of Mechanical Engineers Digital Collection

ASME 2017 Nuclear Forum collocated with the ASME 2017 Power Conference Joint With ICOPE-17, the ASME 2017 11th International Conference on Energy Sustainability, and the ASME 2017 15th International Conference on Fuel Cell Science, Engineering and Technology

June 26–30, 2017

Charlotte, North Carolina, USA

Conference Sponsors:

ISBN:

978-0-7918-4059-7

In This Volume

Codes, Standards, Licensing and Regulatory Compliance (1)
Structures, Components and Materials (3)
Thermal Hydraulics and Computational Fluid Dynamics (6)

ASME 2017 Nuclear Forum

Codes, Standards, Licensing and Regulatory Compliance

ASME PTC 46 Performance Tests in Digital Plant Context

Olivier Le Galudec

ICONE 2017; V009T01A001https://doi.org/10.1115/NUCLRF2017-3245

Abstract

View Articletitled, ASME PTC 46 Performance Tests in Digital Plant Context

PDF

Topics: Testing performance , Doors , Performance test codes , Power stations , Transparency

Structures, Components and Materials

Safety Enhancement Study of CAP1400 Spent Fuel Storage Racks

Xiaoming He, Ziqiang Zhu, Changlei Shao, Ran Huang

ICONE 2017; V009T02A001https://doi.org/10.1115/NUCLRF2017-3164

Abstract

View Articletitled, Safety Enhancement Study of CAP1400 Spent Fuel Storage Racks

PDF

Topics: Safety , Spent fuel storage , Cooling , Neutrons , Storage , China , Computational fluid dynamics , Fukushima nuclear disaster, Japan, 2011 , Nuclear power stations , Pressurized water reactors

Thermo-Mechanical Performance Assessment of Selected Plates From MP-1 High Power Experiments

Hakan Ozaltun, Barry H. Rabin

ICONE 2017; V009T02A002https://doi.org/10.1115/NUCLRF2017-3271

Abstract

View Articletitled, Thermo-Mechanical Performance Assessment of Selected Plates From MP-1 High Power Experiments

PDF

Investigation on Creep Mechanisms of Alloy 709

Abdullah S. Alomari, Nilesh Kumar, Korukonda L. Murty

ICONE 2017; V009T02A003https://doi.org/10.1115/NUCLRF2017-3649

Abstract

View Articletitled, Investigation on Creep Mechanisms of Alloy 709

PDF

Topics: Alloys , Creep , Stress , Temperature , Grain boundary diffusion , Dislocations , Generation IV reactors , Iron , Mechanical properties , Nuclear reactors

Thermal Hydraulics and Computational Fluid Dynamics

A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

John P. McCloskey, Richard J. Smith

ICONE 2017; V009T03A001https://doi.org/10.1115/NUCLRF2017-3266

Abstract

View Articletitled, A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

PDF

Topics: Computers , Coolants , Design , Errors , Fluids , Nuclear power stations , Nuclear reactors , Steady state , Transients (Dynamics) , Uncertainty

Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

Nolan A. Anderson, George L. Mesina

ICONE 2017; V009T03A002https://doi.org/10.1115/NUCLRF2017-3401

Abstract

View Articletitled, Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

PDF

Topics: Condensation , Vapors , Accidents , Density , Nuclear power stations , Pressure , Pressurized water reactors , Steam , Boilers , Coolants

Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection

Yu Weng, Lang Liu, Yang Jiang, Hongfang Gu, Haijun Wang

ICONE 2017; V009T03A003https://doi.org/10.1115/NUCLRF2017-3452

Abstract

View Articletitled, Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection

PDF

Topics: Heat transfer , Nuclear reactors , Safety , Unsteady flow , Flow (Dynamics) , Reactor vessels , Transients (Dynamics) , Accidents , Computational fluid dynamics , Nozzles

Solution of Governing Equations for Six-Field System Code

Glenn A. Roth, George L. Mesina, Fatih Aydogan

ICONE 2017; V009T03A004https://doi.org/10.1115/NUCLRF2017-3515

Abstract

View Articletitled, Solution of Governing Equations for Six-Field System Code

PDF

Topics: Accidents , Algebra , Boiling , Bubbles , Computer software , Computers , Condensation , Coolants , Cooling , Design

Flow Testing and Analysis of the FSP-1 Experiment

Grant L. Hawkes, Warren F. Jones, Wade Marcum, Aaron Weiss, Trevor Howard

ICONE 2017; V009T03A005https://doi.org/10.1115/NUCLRF2017-3639

Abstract

View Articletitled, Flow Testing and Analysis of the FSP-1 Experiment

PDF

Topics: Flow (Dynamics) , Testing , Fuels , Plates (structures) , Uranium , Alloys , Aluminum , Molybdenum , Orifices , Pressure

Experimental Study on Heat Transfer to Supercritical CO₂ Flowing in Vertical Upward Tube at Medium Mass Flux

Qian Zhang, Huixiong Li, Xiangfei Kong, Jun Zhang, Xianliang Lei, Weiqiang Zhang

ICONE 2017; V009T03A006https://doi.org/10.1115/NUCLRF2017-3664

Abstract

View Articletitled, Experimental Study on Heat Transfer to Supercritical CO<sub>2</sub> Flowing in Vertical Upward Tube at Medium Mass Flux

PDF

Topics: Heat transfer , Supercritical carbon dioxide , Heat flux , Pressure , Buoyancy , Carbon dioxide , Wall temperature , Heat transfer coefficients , Density , Fluids

All Years

Codes, Standards, Licensing and Regulatory Compliance

ASME PTC 46 Performance Tests in Digital Plant Context

Structures, Components and Materials

Safety Enhancement Study of CAP1400 Spent Fuel Storage Racks

Thermo-Mechanical Performance Assessment of Selected Plates From MP-1 High Power Experiments

Investigation on Creep Mechanisms of Alloy 709

Thermal Hydraulics and Computational Fluid Dynamics

A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables

Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection

Solution of Governing Equations for Six-Field System Code

Flow Testing and Analysis of the FSP-1 Experiment

Experimental Study on Heat Transfer to Supercritical CO₂ Flowing in Vertical Upward Tube at Medium Mass Flux

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