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Title: Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical results were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.
Authors:
 [1] ;  [1] ;  [2] ;  [3] ;  [4]
  1. City College of New York, NY (United States)
  2. Montana State Univ., Bozeman, MT (United States)
  3. City College of New York, NY (United States); CUNY Energy Inst., New York, NY (United States)
  4. Idaho National Lab. (INL), Idaho Falls, ID (United States)
Publication Date:
Report Number(s):
INL/JOU-16-37821
Journal ID: ISSN 0029-5450
Grant/Contract Number:
AC07-05ID14517
Type:
Accepted Manuscript
Journal Name:
Nuclear Technology
Additional Journal Information:
Journal Volume: 196; Journal Issue: 3; Journal ID: ISSN 0029-5450
Publisher:
American Nuclear Society (ANS)
Research Org:
Idaho National Lab. (INL), Idaho Falls, ID (United States)
Sponsoring Org:
USDOE Office of Nuclear Energy (NE)
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; convection heat transfer; VHTR; high temperature gas reactor; DTHT; flow laminarization
OSTI Identifier:
1357608

Valentin, Francisco I., Artoun, Narbeh, Anderson, Ryan, Kawaji, Masahiro, and McEligot, Donald M.. Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results. United States: N. p., Web. doi:10.13182/NT16-46.
Valentin, Francisco I., Artoun, Narbeh, Anderson, Ryan, Kawaji, Masahiro, & McEligot, Donald M.. Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results. United States. doi:10.13182/NT16-46.
Valentin, Francisco I., Artoun, Narbeh, Anderson, Ryan, Kawaji, Masahiro, and McEligot, Donald M.. 2016. "Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results". United States. doi:10.13182/NT16-46. https://www.osti.gov/servlets/purl/1357608.
@article{osti_1357608,
title = {Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results},
author = {Valentin, Francisco I. and Artoun, Narbeh and Anderson, Ryan and Kawaji, Masahiro and McEligot, Donald M.},
abstractNote = {Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical results were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.},
doi = {10.13182/NT16-46},
journal = {Nuclear Technology},
number = 3,
volume = 196,
place = {United States},
year = {2016},
month = {12}
}