1-12 of 12
Thermal Hydraulics and Computational Fluid Dynamics
Close
Save search
Follow your search
Access your saved searches in your account
Would you like to receive an alert when new items match your search?
Close Modal
Sort by
Proceedings Papers

The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor

Xinhe Qu, Xiaoyong Yang, Gang Zhao, Jie Wang
Proc. ASME. NUCLRF2018, ASME 2018 Nuclear Forum, V001T04A002, June 24–28, 2018
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2018-7384
https://doi.org/10.1115/NUCLRF2018-7384
View Paper titled, The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor
Open the PDF for The Features of Closed Brayton Cycle and Sub-Critical Combined Cycles Coupled With (Very) High Temperature Gas-Cooled Reactor in another window
Proceedings Papers

Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor

Grant L. Hawkes, Douglas S. Crawford, Gregory K. Housley
Proc. ASME. NUCLRF2018, ASME 2018 Nuclear Forum, V001T04A003, June 24–28, 2018
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2018-7600
https://doi.org/10.1115/NUCLRF2018-7600
View Paper titled, Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor
Open the PDF for Thermal Analysis Safety Margins Using ABAQUS for the MP-2 Experiment in the Advanced Test Reactor in another window
Topics: Safety, Thermal analysis, Fuels, Irradiation (Radiation exposure), Water, Aluminum, Finite element analysis, Plates (structures), Coolants, Finite element model
Proceedings Papers

Analysis of the Thermal-Hydraulic and Mechanical Performance of Intermediate Heat Exchangers Used in Systems With Very High Temperature Nuclear Reactors

Raciel de la Torre Valdés, Juan Luis Francois Lacouture
Proc. ASME. NUCLRF2018, ASME 2018 Nuclear Forum, V001T04A004, June 24–28, 2018
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2018-7620
https://doi.org/10.1115/NUCLRF2018-7620
View Paper titled, Analysis of the Thermal-Hydraulic and Mechanical Performance of Intermediate Heat Exchangers Used in Systems With Very High Temperature Nuclear Reactors
Open the PDF for Analysis of the Thermal-Hydraulic and Mechanical Performance of Intermediate Heat Exchangers Used in Systems With Very High Temperature Nuclear Reactors in another window
Topics: Equipment performance, Heat exchangers, High temperature, Nuclear reactors, Computational fluid dynamics, Design, Stress, Circuits, Compacting, Dimensions
Proceedings Papers

Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT

Xiaoyong Yang, Xiao Li, Jie Wang, Youjie Zhang
Proc. ASME. NUCLRF2018, ASME 2018 Nuclear Forum, V001T04A001, June 24–28, 2018
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2018-7381
https://doi.org/10.1115/NUCLRF2018-7381
View Paper titled, Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT
Open the PDF for Analysis of Loss-of-Flow Accidents in Pre-Cooler and Inter-Cooler of HTR-10GT in another window
Topics: Accidents, Flow (Dynamics), Helium, Temperature, Water, Water temperature, Heat exchangers, Safety, Very high temperature reactors, Brayton cycle
Proceedings Papers

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

Nolan A. Anderson, George L. Mesina
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A002, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3401
https://doi.org/10.1115/NUCLRF2017-3401
View Paper titled, Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables
Open the PDF for Improvement of the RELAP5-3D Model of Condensation in the Presence of Noncondensables in another window
Topics: Condensation, Vapors, Accidents, Density, Nuclear power stations, Pressure, Pressurized water reactors, Steam, Boilers, Coolants
Proceedings Papers

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

Yu Weng, Lang Liu, Yang Jiang, Hongfang Gu, Haijun Wang
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A003, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3452
https://doi.org/10.1115/NUCLRF2017-3452
View Paper titled, Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection
Open the PDF for Investigation of Transient Flow and Heat Transfer for Passive Nuclear Reactor Direct Safety Injection in another window
Topics: Heat transfer, Nuclear reactors, Safety, Unsteady flow, Flow (Dynamics), Reactor vessels, Transients (Dynamics), Accidents, Computational fluid dynamics, Nozzles
Proceedings Papers

Solution of Governing Equations for Six-Field System Code

Glenn A. Roth, George L. Mesina, Fatih Aydogan
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A004, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3515
https://doi.org/10.1115/NUCLRF2017-3515
View Paper titled, Solution of Governing Equations for Six-Field System Code
Open the PDF for Solution of Governing Equations for Six-Field System Code in another window
Topics: Accidents, Algebra, Boiling, Bubbles, Computer software, Computers, Condensation, Coolants, Cooling, Design
Proceedings Papers

Flow Testing and Analysis of the FSP-1 Experiment

Grant L. Hawkes, Warren F. Jones, Wade Marcum, Aaron Weiss, Trevor Howard
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A005, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3639
https://doi.org/10.1115/NUCLRF2017-3639
View Paper titled, Flow Testing and Analysis of the FSP-1 Experiment
Open the PDF for Flow Testing and Analysis of the FSP-1 Experiment in another window
Topics: Flow (Dynamics), Testing, Fuels, Plates (structures), Uranium, Alloys, Aluminum, Molybdenum, Orifices, Pressure
Proceedings Papers

Experimental Study on Heat Transfer to Supercritical CO 2 Flowing in Vertical Upward Tube at Medium Mass Flux

Qian Zhang, Huixiong Li, Xiangfei Kong, Jun Zhang, Xianliang Lei, Weiqiang Zhang
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A006, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3664
https://doi.org/10.1115/NUCLRF2017-3664
View Paper titled, Experimental Study on Heat Transfer to Supercritical CO 2 Flowing in Vertical Upward Tube at Medium Mass Flux
Open the PDF for Experimental Study on Heat Transfer to Supercritical CO 2 Flowing in Vertical Upward Tube at Medium Mass Flux in another window
Topics: Heat transfer, Supercritical carbon dioxide, Heat flux, Pressure, Buoyancy, Carbon dioxide, Wall temperature, Heat transfer coefficients, Density, Fluids
Proceedings Papers

A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP

John P. McCloskey, Richard J. Smith
Proc. ASME. NUCLRF2017, ASME 2017 Nuclear Forum, V009T03A001, June 26–30, 2017
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2017-3266
https://doi.org/10.1115/NUCLRF2017-3266
View Paper titled, A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP
Open the PDF for A Statistical Method for Benchmarking Nuclear Plant Models, Using ACAP in another window
Topics: Computers, Coolants, Design, Errors, Fluids, Nuclear power stations, Nuclear reactors, Steady state, Transients (Dynamics), Uncertainty
Proceedings Papers

The Effects of Subcooling on Quenching of a Vertical Brass Cylinder With Heating Power

Yuan-Hong Ho, Ming-Xi Ho, Chin Pan
Proc. ASME. NUCLRF2015, ASME 2015 Nuclear Forum, V001T04A001, June 28–July 2, 2015
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2015-49307
https://doi.org/10.1115/NUCLRF2015-49307
View Paper titled, The Effects of Subcooling on Quenching of a Vertical Brass Cylinder With Heating Power
Open the PDF for The Effects of Subcooling on Quenching of a Vertical Brass Cylinder With Heating Power in another window
Topics: Brass (Metal), Cylinders, Heating, Quenching (Metalworking), Subcooling, Film boiling, Coolants, Heat, Nuclear power stations, Water
Proceedings Papers

Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor

Grant L. Hawkes, James W. Sterbentz, Binh T. Pham
Proc. ASME. NUCLRF2015, ASME 2015 Nuclear Forum, V001T04A002, June 28–July 2, 2015
Publisher: American Society of Mechanical Engineers
Paper No: NUCLRF2015-49698
https://doi.org/10.1115/NUCLRF2015-49698
View Paper titled, Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor
Open the PDF for Sensitivity Evaluation of the Daily Thermal Predictions of the AGR-2 Experiment in the Advanced Test Reactor in another window
Topics: Finite element analysis, Fluence (Radiation measurement), Fuels, Graphite, Heat, Neutrons, Particulate matter, Stainless steel, Temperature, Thermal conductivity