selected scholarly activity
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academic article
- Thermal-hydraulics analysis of flow blockage events for fuel assembly in a sodium-cooled fast reactor. International Journal of Heat and Mass Transfer. 138:496-507. 2019
- A review on recent heat transfer studies to supercritical pressure water in channels. Applied Thermal Engineering. 142:573-596. 2018
- Numerical investigation of buoyancy effect on heat transfer to carbon dioxide flow in a tube at supercritical pressures. International Journal of Heat and Mass Transfer. 117:595-606. 2018
- Nonuniform heat transfer of supercritical water in a tight rod bundle – Assessment of correlations. Annals of Nuclear Energy. 110:570-583. 2017
- A predictive-corrective process for predicting forced convective heat transfer in heated tubes at supercritical pressures. International Journal of Heat and Mass Transfer. 110:374-382. 2017
- Improvement of buoyancy and acceleration parameters for forced and mixed convective heat transfer to supercritical fluids flowing in vertical tubes. International Journal of Heat and Mass Transfer. 106:1144-1156. 2017
- Heat transfer from a 2 × 2 wire-wrapped rod bundle to supercritical pressure water. International Journal of Heat and Mass Transfer. 97:486-501. 2016
- Experiments on the basic behavior of supercritical CO 2 natural circulation. Nuclear Engineering and Design. 300:376-383. 2016
- Assessment of Computational Tools in Support of Heat-Transfer Correlation Development for Fuel Assembly of Canadian Supercritical Water-Cooled Reactor. Journal of Nuclear Engineering and Radiation Science. 2. 2016
- Heat transfer effectiveness for cooling of Canadian SCWR fuel assembly under the LOCA/LOECC scenario. Annals of Nuclear Energy. 81:306-319. 2015
- Overview of methods to increase dryout power in CANDU fuel bundles. Nuclear Engineering and Design. 287:131-138. 2015
- Review of R&D for supercritical water cooled reactors. Progress in Nuclear Energy. 77:282-299. 2014
- Experimental investigation of heat transfer from a 2×2 rod bundle to supercritical pressure water. Nuclear Engineering and Design. 275:205-218. 2014
- Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle. Nuclear Engineering and Design. 264:119-125. 2013
- Issues and future direction of thermal-hydraulics research and development in nuclear power reactors. Nuclear Engineering and Design. 264:3-23. 2013
- Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels. Nuclear Engineering and Design. 241:4045-4054. 2011
- Coupled neutronics/thermal–hydraulics analysis of CANDU–SCWR fuel channel. Annals of Nuclear Energy. 37:58-65. 2010
- Subchannel analysis of CANDU-SCWR fuel. Progress in Nuclear Energy. 51:799-804. 2009
- SCWR subchannel code ATHAS development and CANDU-SCWR analysis. Nuclear Engineering and Design. 239:1979-1987. 2009
- PREDICTIONS OF CRITICAL HEAT FLUX USING THE ASSERT-PV SUBCHANNEL CODE FOR A CANFLEX VARIANT BUNDLE. Nuclear Engineering and Technology. 41:969-978. 2009
- Effect of CANDU Bundle-Geometry Variation on Dryout Power. Journal of Engineering for Gas Turbines and Power. 131. 2009
- A Study of the Mechanics of Porous Plga 85/15 Scaffold in Compression. Polymers and Polymer Composites Engineering plastics. 15:437-443. 2007
- An experimental and analytical study of the effect of axial power profile on CHF. Nuclear Engineering and Design. 236:1384-1395. 2006
- Lookup Tables for Predicting CHF and Film-Boiling Heat Transfer: Past, Present, and Future. Nuclear Technology. 152:87-104. 2005
- Prediction of the obstacle effect on film-boiling heat transfer. Nuclear Engineering and Design. 235:687-700. 2005
- Pressure drops for steam and water flow in heated tubes. Nuclear Engineering and Design. 235:53-65. 2005
- A look-up table for fully developed film-boiling heat transfer. Nuclear Engineering and Design. 225:83-97. 2003
- The 1995 look-up table for critical heat flux in tubes. Nuclear Engineering and Design. 163:1-23. 1996
- Crystallization of Bacillus subtilis Tryptophanyl-tRNA Synthetase. Journal of Molecular Biology. 230:1089-1090. 1993
- Computation of single- and two-phase heat transfer rates suitable for water-cooled tubes and subchannels. Nuclear Engineering and Design. 114:61-77. 1989
- A model for predicting diabatic pressure drops in multi-element fuel channels. Nuclear Engineering and Design. 110:299-312. 1989
- Two-phase pressure drop through obstructions. Nuclear Engineering and Design. 105:349-361. 1988
- Effect of flow-obstruction geometry on pressure drops in horizontal air-water flow. International Journal of Multiphase Flow. 9:73-85. 1983
- EVOLUTION OF THE CANADIAN SCWR FUEL-ASSEMBLY CONCEPT AND ASSESSMENT OF THE 64 ELEMENT ASSEMBLY FOR THERMALHYDRAULIC PERFORMANCE. CNL Nuclear Review. 1-18.
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conference paper
- NEW IAEA COORDINATED RESEARCH PROJECT ON THERMAL-HYDRAULICS OF SUPERCRITICAL WATER COOLED REACTORS. PROCEEDINGS OF THE 24TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 2016, VOL 2. 2016
- EXPERIMENTAL STUDY OF HEAT TRANSFER TO SUPERCRITICAL PRESSURE WATER FLOWING IN A 2x2 ROD BUNDLE. PROCEEDINGS OF THE 22ND INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2014, VOL 2A. 2014
- ASSESSMENT OF HEAT-TRANSFER CORRELATIONS AGAINST EXPERIMENTAL DATA OBTAINED WITH SUPERCRITICAL WATER IN VERTICAL ANNULI. PROCEEDINGS OF THE 20TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING AND THE ASME 2012 POWER CONFERENCE - 2012, VOL 3. 103-110. 2012
- THERMALHYDRAULICS AND SAFETY PROGRAMS IN SUPPORT OF THE CANDU SCWR DESIGN. PROCEEDINGS OF THE 18TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING 2010, VOL 6. 211-218. 2011
- COMPARISON AND IMPROVEMENTS OF CORRELATIONS FOR FILM BOILING IN TUBES. ICONE 17: PROCEEDINGS OF THE 17TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 3. 793-799. 2009
- ANALYTICAL AND EXPERIMENTAL PROGRAM OF SUPERCRITICAL HEAT TRANSFER RESEARCH AT THE UNIVERSITY OF OTTAWA. Nuclear Engineering and Technology. 107-116. 2008
- The 2006 CHF look-up table. Nuclear Engineering and Design. 1909-1922. 2007
- Thermalhydraulics studies examining the feasibility for introducing slightly enriched uranium fuel into the Embalse CANDU reactor. Nuclear Engineering and Design. 1628-1638. 2007