selected scholarly activity
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chapters
- Chapter 8 SuperCritical Water-cooled Reactors (SCWRs). 259-284. 2023
- SuperCritical Water-cooled Reactors (SCWRs). 259-284. 2023
- Super-Critical Water-Cooled Reactor (SCWR). 569-581. 2021
- Neutron Activation Analysis of Candidate Materials for High-Temperature Reactors. 793-800. 2017
- 8 Super-critical water-cooled reactors. 189-220. 2016
- Super-critical water-cooled reactors. 189-220. 2016
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conferences
- Assessment of Convective Heat Transfer Correlations Against an Expanded Database for Different Fluids at Supercritical Pressures. Journal of Nuclear Engineering and Radiation Science. 2018
- Summary on the Results of Two Computational Fluid Dynamic Benchmarks of Tube and Different Channel Geometries. Journal of Nuclear Engineering and Radiation Science. 2018
- Thermal-Hydraulics Program in Support of Canadian SCWR Concept Development. Journal of Nuclear Engineering and Radiation Science. 2018
- New IAEA Coordinated Research Project on Thermal-Hydraulics of Supercritical Water Cooled Reactors. Volume 2: Smart Grids, Grid Stability, and Offsite and Emergency Power; Advanced and Next Generation Reactors, Fusion Technology; Safety, Security, and Cyber Security; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues. 2016
- Effects of neutron activation on candidate fuel cladding for SCWR design. 36th Annual CNS Conference and 40th CNS-CNA Student Conference - Nuclear in the 21st Century: Global Directions and Canada's Role. 2016
- Gen IV SCWR cladding analysis project: Nickel content in SCWR cladding material. 2015 5th International Youth Conference on Energy (IYCE). 1-8. 2015
- Experimental Study of Heat Transfer to Supercritical Pressure Water Flowing in a 2×2 Rod Bundle. Volume 2A: Thermal Hydraulics. 2014
- Compilation of supercritical heat transfer data through the IAEA Coordinated Research Project. Canadian Nuclear Society - 33rd Annual Conference of the Canadian Nuclear Society and 36th CNS/CNA Student Conference 2012: Building on Our Past... Building for the Future. 1214-1225. 2012
- International contributions to IAEA-NEA heat transfer databases for supercritical fluids. International Congress on Advances in Nuclear Power Plants 2012, ICAPP 2012. 186-197. 2012
- Assessment of Heat-Transfer Correlations Against Experimental Data Obtained With Supercritical Water in Vertical Annuli. Volume 3: Thermal-Hydraulics; Turbines, Generators, and Auxiliaries. 103-110. 2012
- A Supercritical Water-Cooled Small Modular Reactor. ASME 2011 Small Modular Reactors Symposium. 243-250. 2011
- Research and development initiatives in support of the conceptual design for the CANDU supercritical water-cooled reactor. Canadian Nuclear Society - 31st Annual Conference of the Canadian Nuclear Society and 34th CNS/CNA Student Conference 2010. 1216-1229. 2010
- Advances in Pressure Tube Reactor Technology. 18th International Conference on Nuclear Engineering: Volume 6. 227-232. 2010
- Thermalhydraulics and Safety Programs in Support of the CANDU SCWR Design. 18th International Conference on Nuclear Engineering: Volume 6. 211-218. 2010
- Canada's NSERC/NRCan/AECL generation IV energy technologies program. Canadian Nuclear Society - 30th Annual Canadian Nuclear Society Conference and 33rd CNS/CNA Student Conference 2009. 1350-1362. 2009
- Optimisation of CANFLEX-SCWR bundle through subchannel analysis. Canadian Nuclear Society - 30th Annual Canadian Nuclear Society Conference and 33rd CNS/CNA Student Conference 2009. 2032-2043. 2009
- Thermalhydraulics projects in support of conceptual design and safety analyses of CANDU SCWR. Canadian Nuclear Society - 30th Annual Canadian Nuclear Society Conference and 33rd CNS/CNA Student Conference 2009. 1363-1371. 2009
- Comparison and Improvements of Correlations for Film Boiling in Tubes. Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance. 793-799. 2009
- Investigation of the combined effect of appendages and axial power profile on post-dryout heat transfer in tubes. Canadian Nuclear Society - 29th Annual Conference of the Canadian Nuclear Society and 32nd CNS/CNA Student Conference 2008. 1165-1175. 2008
- Subchannel analysis of candu-scwr fuel. International Conference on Advances in Nuclear Power Plants, ICAPP 2008. 1576-1584. 2008
- ANALYTICAL AND EXPERIMENTAL PROGRAM OF SUPERCRITICAL HEAT TRANSFER RESEARCH AT THE UNIVERSITY OF OTTAWA. Nuclear Engineering and Technology. 107-116. 2008
- Effect of CANDU Bundle-Geometry Variation on Dryout Power. Volume 3: Thermal Hydraulics; Instrumentation and Controls. 859-866. 2008
- Comparison of heat-flux and wall-temperature based correlations for predicting post-dryout surface temperature in tubes. Canadian Nuclear Society - 28th Annual Conference of the Canadian Nuclear Society and 31st CNS/CNA Student Conference 2007: "Embracing the Future: Canada's Nuclear Renewal and Growth". 1637-1648. 2007
- Predictions of critical heat flux using the ASSERT subchannel code for a CANFLEX variant bundle. Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12. 2007
- Tube-CHF database compilation and assessment. Proceedings - 12th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-12. 2007
- 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
- Modelling of pressure distributions over fuel-strings for fuel irradiation experiments in the NRU reactor. 27th Annual Conference of the Canadian Nuclear Society and 30th Canadian Nuclear Society - Nuclear Energy A World of Service to Humanity. 2006
- CRITICAL HEAT FLUX IN AXIALLY NON-UNIFORM-HEATED CHANNELS. Boiling. 15. 2006
- Design and qualification of the bruce CANFLEX® low void reactivity fuel bundle - An overview. Annual Canadian Nuclear Society Conference. 381-392. 2004
- International collaboration to study the feasibility of implementing the use of slightly enriched uranium fuel in the embalse CANDU reactor. Abstracts of the Pacific Basin Nuclear Conference. 120. 2004
- Thermalhydraulic studies in support of qualification of low-void reactivity fuel. Annual Canadian Nuclear Society Conference. 407-416. 2004
- Separate Effects on Film-Boiling Heat Transfer. Proceeding of International Heat Transfer Conference 12. 6. 2002
- A GENERALIZED PREDICTION METHOD FOR CRITICAL HEAT FLUX IN CANDU FUEL-BUNDLE STRINGS. Proceeding of International Heat Transfer Conference 11. 15-20. 1998
- Low-void reactivity CANDU fuel bundle. Annual International Conference - Canadian Nuclear Association. 1992
- PRESSURE LOSSES THROUGH OBSTRUCTION IN BOTH HORIZONTAL AND VERTICAL BUBBLY FLOWS.. American Society of Mechanical Engineers, Fluids Engineering Division (Publication) FEDSM. 11-17. 1985
- EFFECT OF STREAMLINING ON BUNDLE PRESSURE DROP.. 405-420. 1984
- EFFECT OF STREAMLINING ON BUNDLE PRESSURE DROP.. 67-69. 1983
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journal articles
- Fluid-to-Fluid modelling of CHF at High-Pressure subcooled water conditions. Nuclear Engineering and Design. 386:111577-111577. 2022
- A phenomenological CHF model for mixing-vane spacers in a subchannel of a rod bundle. Annals of Nuclear Energy. 142:107445-107445. 2020
- Special Section: Selected Papers From the Ninth International Symposium on Supercritical Water-Cooled Reactors. Journal of Nuclear Engineering and Radiation Science. 6. 2020
- A mechanistic bubble crowding model for predicting critical heat flux in subchannels of a bundle. Annals of Nuclear Energy. 137:107085-107085. 2020
- Assessment of a Theoretical Model for Predicting Forced Convective Critical Heat Flux in Rod Bundles. Frontiers in Energy Research. 7. 2019
- 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
- Professor Pavel Leonidovich Kirillov on His 90th Birthday. Journal of Nuclear Engineering and Radiation Science. 4. 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
- Heat transfer of supercritical carbon dioxide flowing in a rectangular circulation loop. Applied Thermal Engineering. 98:39-48. 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. 2:011006. 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
- ICONE19-44057 A Study of Non-Dimensional Parameters for the Oscillatory Instability Boundary In Supercritical Flow. The Proceedings of the International Conference on Nuclear Engineering (ICONE). 2011.19:_ICONE1944-_ICONE1944. 2011
- Measurements of critical heat flux in CANDU 37-element bundle with a steep variation in radial power profile. Nuclear Engineering and Design. 240:290-298. 2010
- 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. 131:022906. 2009
- A Study of the Mechanics of Porous Plga 85/15 Scaffold in Compression. Polymers and Polymer Composites. 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
- Comparison of CHF measurements in horizontal and vertical tubes cooled with R-134a. International Journal of Heat and Mass Transfer. 45:4435-4450. 2002
- 1995 Look-up Table for Calculating Critical Heat Flux in Tubes. Thermal Engineering (English translation of Teploenergetika). 44:823-840. 1997
- 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 on two-phase pressure drops in both horizontal and vertical annular flows. 1985
- Pressure losses through obstruction in both horizontal and vertical bubbly flows.. IN: FUNDAMENTAL ASPECTS OF GAS-LIQUID FLOW, PRESENTED AT ASME WINTER ANNUAL MEETING, (MIAMI BEACH, U.S.A.: NOV. 17-22, 1985),. 29 ):11-17. 1985
- Effect of flow-obstruction geometry on pressure drops in horizontal air-water flow. International Journal of Multiphase Flow. 9:73-85. 1983
- AN ASSESSMENT OF ROUND TUBE CORRELATIONS FOR CONVECTIVE HEAT TRANSFER AT SUPERCRITICAL PRESSURE. CNL NUCLEAR REVIEW. 1-17.
- EVOLUTION OF THE CANADIAN SCWR FUEL-ASSEMBLY CONCEPT AND ASSESSMENT OF THE 64 ELEMENT ASSEMBLY FOR THERMALHYDRAULIC PERFORMANCE. CNL NUCLEAR REVIEW. 1-18.
- GENERAL ASSESSMENT OF CONVECTION HEAT TRANSFER CORRELATIONS FOR MULTIPLE GEOMETRIES AND FLUIDS AT SUPERCRITICAL PRESSURE. CNL NUCLEAR REVIEW. 1-20.