Nuclear Engineering and Design: an international journal devoted to the thermal, mechanical, materials, and structural aspects of nuclear fission energy Journal

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Simulation of a loss of flow transient of a small Lead-Cooled reactor using a CFD-Based model. 412:112462-112462. 2023
Analysis of natural convection heat transfer and solidification within a corium simulant. 409:112318-112318. 2023
Nuclear reactor Thermal-Hydraulics Special and prospective topical areas. 409:112354-112354. 2023
Flow regime transitions for Aiding flow around a heated cylinder: A numerical and experimental investigation. 405:112206-112206. 2023
From NURETH-2013 to NURETH-2019: Non-Critical summary and brief contribution to the history of nuclear reactor thermal hydraulics. 403:112153-112153. 2023
Mixed convection around two vertically aligned horizontal cylinders: A numerical, experimental, and modeling investigation on the effect of local conditions on heat transfer. 394:111823-111823. 2022
Numerical modelling of the effect of appendages on turbulent mixing in connected nuclear fuel subchannels. 395:111788-111788. 2022
Development of a CFD-based model to simulate loss of flow transients in a small lead-cooled reactor. 392:111773-111773. 2022
Neutronics analysis of a source-driven subcritical system using G4-STORK. 388:111617-111617. 2022
Fluid-to-Fluid modelling of CHF at High-Pressure subcooled water conditions. 386:111577-111577. 2022
Baffle jetting modelling. 379:111233-111233. 2021
OECD nuclear energy agency CODAP database project on passive component operating experience. An international collaboration in materials research. 380:111280-111280. 2021
An experimental and numerical investigation into scaling considerations for moderator circulation experiments. 369:110829-110829. 2020
Experimental measurement and Monte Carlo code simulation of the gamma heating at different irradiation sites in a nuclear research reactor. 364:110690-110690. 2020
Extension and preliminary validation of the Polaris lattice physics code for CANDU analysis. 361:110540-110540. 2020
Evaluating adaptive vertical seismic isolation for equipment in nuclear power plants. 358:110399-110399. 2020
An experimental investigation of subcooled choked flow in actual steam generator tube cracks. 354:110144-110144. 2019
The life and the contribution of B. R. Sehgal, G. Yadigaroglu and G. Hewitt: Remembrance statements. 354:110252-110252. 2019
Mechanistic modelling of station blackout accidents for a generic 900 MW CANDU plant using the modified RELAP/SCDAPSIM/MOD3.6 code. 335:71-93. 2018
3D-thermo-structural simulation of pressure tube–calandria tube behaviour under accident conditions in PHWR using ABAQUS. 328:188-196. 2018
CFD analysis and experimental validation of steady state mixed convection sodium flow. 326:333-343. 2018
Effect of initial blast response on RC beams failure modes. 320:437-451. 2017
RELAP5 simulation of CANDU Station Blackout accidents with/without water make-up to the steam generators. 318:35-53. 2017
The effects of transient conditions on the onset of intermittent dryout during blowdown. 317:118-132. 2017
Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface. 315:104-116. 2017
Experimental investigation of quench and re-wetting temperatures of hot horizontal tubes well above the limiting temperature for solid–liquid contact. 311:167-183. 2017
Measurement and analysis of the re-wetting front velocity during quench cooling of hot horizontal tubes. 311:184-198. 2017
Probabilistic finite element investigation of prestressing loss in nuclear containment wall segments. 311:50-59. 2017
Experiments on the basic behavior of supercritical CO 2 natural circulation. 300:376-383. 2016
Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor. 298:78-89. 2016
Laminar simulation of intersubchannel mixing in a triangular nuclear fuel bundle geometry. 295:305-316. 2015
A comprehensive flow-induced vibration model to predict crack growth and leakage potential in steam generator tubes. 292:17-31. 2015
Turbulent jet erosion of a stably stratified gas layer in a nuclear reactor test containment. 292:133-148. 2015
Overview of methods to increase dryout power in CANDU fuel bundles. 287:131-138. 2015
Material challenges for advanced reactors. 280:651-651. 2014
Power distribution control of CANDU reactors based on modal representation of reactor kinetics. 278:323-332. 2014
Mass transfer in back to back elbows arranged in an out of plane configuration under annular two phase flow conditions. 276:142-152. 2014
Experimental investigation of heat transfer from a 2×2 rod bundle to supercritical pressure water. 275:205-218. 2014
Mass transfer in back-to-back elbows arranged in an out of plane configuration. 270:209-216. 2014
Numerical investigation of quasi-periodic flow and vortex structure in a twin rectangular subchannel geometry using detached eddy simulation. 270:1-20. 2014
Analysis of implicit and explicit lattice sensitivities using DRAGON. 265:1-12. 2013
Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle. 264:119-125. 2013
Issues and future direction of thermal-hydraulics research and development in nuclear power reactors. 264:3-23. 2013
To the special issue on the 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14). 264:1-2. 2013
Performance evaluation of a 3-D kinetic model for CANDU reactors in a closed-loop environment. 243:76-86. 2012
Investigation of wall mass transfer characteristics downstream of an orifice. 242:353-360. 2012
Single and multi degree of freedom analysis of steel beams under blast loading. 242:63-77. 2012
Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels. 241:4045-4054. 2011
The reactor physics characteristics of a transuranic mixed oxide fuel in a heavy water moderated reactor. 241:3768-3776. 2011
Scoping assessment of design characteristics for an enhanced calandria tube. 241:2688-2697. 2011
Preface. 241:1287-1287. 2011
The hydrodynamic effects of single-phase flow on flow accelerated corrosion in a 90-degree elbow. 240:1589-1598. 2010
Measurements of critical heat flux in CANDU 37-element bundle with a steep variation in radial power profile. 240:290-298. 2010
Simulation of flow pulsations in a twin rectangular sub-channel geometry using unsteady Reynolds Averaged Navier–Stokes modelling. 239:2964-2980. 2009
SCWR subchannel code ATHAS development and CANDU-SCWR analysis. 239:1979-1987. 2009
OPDE—The international pipe failure data exchange project. 238:2115-2123. 2008
Effect of near-fault earthquakes on North American nuclear design spectra. 236:1928-1936. 2006
An experimental and analytical study of the effect of axial power profile on CHF. 236:1384-1395. 2006
Hydraulic study of turbulent flow in MTR-type nuclear fuel assembly. 236:975-984. 2006
Prediction of the obstacle effect on film-boiling heat transfer. 235:687-700. 2005
Pressure drops for steam and water flow in heated tubes. 235:53-65. 2005
A look-up table for fully developed film-boiling heat transfer. 225:83-97. 2003
Investigation of large amplitude stratified waves in a CANDU-type 37 rod nuclear fuel channel by a real-time neutron radiography technique. 200:221-231. 2000
The 1995 look-up table for critical heat flux in tubes. 163:1-23. 1996
Measurement of interfacial area concentration in subcooled liquid-vapour flow. 152:243-255. 1994
Spatio-temporal complexities and chaos in a two-phase flashing flow. 149:53-66. 1994
A new stochastic analysis for multiple supported MDOF secondary systems: I. Dynamic interaction effects. 147:235-249. 1994
A new stochastic analysis for multiple supported MDOF secondary systems: II. Tuning and spatial coupling effects. 147:251-262. 1994
Rewetting of hot horizontal tubes. 138:375-388. 1992
The SLOWPOKE licensing model. 137:191-198. 1992
Extensions to the approximation functions for the fast calculation of saturated water properties. 136:381-388. 1992
A modified CCFS approach for the seismic analysis of multiply supported MDOF secondary systems. 133:183-197. 1992
STGEN computer program methodology and application to the steam line break tests in the RD-12 steam generator. 126:151-173. 1991
On the development of a model for predicting phase separation phenomena in dividing two-phase flow. 123:67-75. 1990
Computation of single- and two-phase heat transfer rates suitable for water-cooled tubes and subchannels. 114:61-77. 1989
Simple functions for the fast approximation of light water thermodynamic properties. 113:21-34. 1989
A model for predicting diabatic pressure drops in multi-element fuel channels. 110:299-312. 1989
Two-phase pressure drop through obstructions. 105:349-361. 1988
On the thermal analysis of pressure tube/calandria tube contact in CANDU reactors. 104:197-206. 1987
Statistical uncertainty in the radiation diagnosis of two-phase flows. 85:261-275. 1985
Effect of multi-component excitation on the sliding response of unanchored components in nuclear facilities 2022
CFD analysis and experimental validation of mixed convection sodium flow 2017
Heat transfer profiles of a vertical, bare, 7-element bundle cooled with supercritical Freon R-12 2013
A generalized failure map for fuel elements subject to a power pulse 2011
Optimization of rib-roughened annular gas-coolant channels 2010
The 2006 CHF look-up table 2007
Thermalhydraulics studies examining the feasibility for introducing slightly enriched uranium fuel into the Embalse CANDU reactor 2007
Seismic qualification of expansion anchors to Canadian nuclear standards 2004

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0915 Interdisciplinary Engineering (FoR)
Energy (Science Metrix)