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. 10.49-10.55. 1992
- PRESSURE LOSSES THROUGH OBSTRUCTION IN BOTH HORIZONTAL AND VERTICAL BUBBLY FLOWS.. American Society of Mechanical Engineers Fluids Engineering Division Publication FED. 11-17. 1985
- EFFECT OF STREAMLINING ON BUNDLE PRESSURE DROP.. undefined. 405-420. 1984
- EFFECT OF STREAMLINING ON BUNDLE PRESSURE DROP.. undefined. 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.. undefined. 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.