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Volume 112

Number 3

Validation of the Assert Subchannel Code: Prediction of Critical Heat Flux in Standard and Nonstandard Candu Bundle Geometries

M. B. Carver, J. C. Kiteley, R. Q.-N. Zhou, S. V. Junop, D. S. Rowe

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 299-314

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35156

Prediction of Dryout Performance for Boiling Water Reactor Fuel Assemblies Based on Subchannel Analysis with the Rings Code

P. Knabe, F. Wehle

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 315-323

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35157

A Perspective on Large eddy Simulation of Problems in the Nuclear Industry

Yassin A. Hassan, John M. Pruitt, David A. Steininger

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 324-330

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35158

The THYC Three-Dimensional Thermal-Hydraulic Code for Rod Bundles: Recent Developments and Validation Tests

Sylvie Aubry, Christian Caremoli, Jean Olive, Paul Rascle

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 331-345

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35159

The Application of Computational Fluid Dynamics to Critical Heat Flux

James H. Stuhmiller, Paul J. Masiello, Govinda S. Srikantiah, Lance J. Agee

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 346-354

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35160

A Utility Perspective on Subchannel Analysis

Gregg B. Swindlehurst

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 355-358

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35161

Equilibrium Quality and Mass Flux Distributions in an Adiabatic Three-Subchannel Test Section

George Yadigaroglu, Athan Maganas

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 359-372

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35162

Reactor Subchannel Analysis—Electric Power Research Institute Perspective

G. Srikantiah

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 373-381

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35163

The Japanese Utilities’ Expectations for Subchannel Analysis

Akio Toba, Akira Omoto

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 382-387

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35164

Void Fraction Distribution in A Boiling Water Reactor Fuel Assembly and the Evaluation of Subchannel Analysis Codes

Akira Inoue, Masanobu Futakuchi, Makoto Yagi, Toru Mitsutake, Shin-Ichi Morooka

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 388-400

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35165

Subchannel Thermal-Hydraulic Analysis at AP600 Low-Flow Steam-Line-Break Conditions

T. Morita, C. A. Olson, Y. X. Sung, J. F. Connelley, Jr., E. H. Novendstern, S. Kapil, P. W. Rosenthal

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 401-411

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35166

Pressurized Water Reactor Fuel Assembly Subchannel Void Fraction Measurement

Yoshiei Akiyama, Keiichi Hori, Keiji Miyazaki, Kaichiro Mishima, Shigekazu Sugiyama

Nuclear Technology / Volume 112 / Number 3 / December 1995 / Pages 412-421

Technical Paper / Heat Transfer and Fluid Flow / dx.doi.org/10.13182/NT95-A35167

 

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