Ku, J. H., Hong, H., Kim, J. S., & Cho, H. K. (2025). Wall heat partitioning model with bubble tracking method for nucleate boiling considering conjugate heat transfer coupled with OpenFOAM. International Communications in Heat and Mass Transfer, 160, 108364.
Lee, C. W., Ko, H., Hong, H., Yoo, J. S., Ku, J. H., Kim, G. W., Park, G. C. & Cho, H. K. (2024). Experiments of critical heat flux for helical finned rod under rolling conditions using R-134a. International Communications in Heat and Mass Transfer, 159, 107979.
Bicer, E., Hong, S. J., & Cho, H. K. (2024). A novel correlation for bubble size variation in the swarm region under pool scrubbing conditions. Progress in Nuclear Energy, 176, 105366.
Yoo, J. S., Lee, C. W., Hong, H., Ko, H., Ku, J. H., Bae, G. W., Kim, K., Partk, G. C. & Cho, H. K. (2024). Experimental study on flow boiling CHF of a helical finned rod under heaving conditions using simulant fluid R134a. International Journal of Heat and Mass Transfer, 232, 125927.
Bicer, E., Hong, S. J., & Cho, H. K. (2024). Investigation of Equivalent Spherical Bubble Diameter at High Inlet Velocity Pool Scrubbing Conditions. Nuclear Engineering and Technology.
Ko, Y., Seo, H., Cho, H. K., & Yi, K. (2024). Effect of inclination on SB LOCA transient of a floating nuclear reactor in MARS-KS analysis using moving reactor model. Nuclear Engineering and Design, 426, 113407.
Lee, J., Park, Y., Jeon, S. S., Park, J. Y., & Cho, H. K. (2024). Assessment of MARS-KS prediction capability for natural circulation flow in passive heat removal system. Nuclear Engineering and Technology.
Kim, S. Y., & Cho, H. K. (2024). Turbulence characteristics of corner flow and modification of elliptic relaxation function in Phit–fk–ε model. International Journal of Heat and Fluid Flow, 107, 109373.
Hong, H., Ku, J. H., Kim, J. S., & Cho, H. K. (2024). Boiling heat flux partitioning model with bubble tracking method considering bubble merger and stochastic characteristics. International Journal of Heat and Mass Transfer, 226, 125443.
Ko, Y., Seo, H., Cho, H. K., & Kim, I. (2024). SBLOCA analysis of a floating nuclear reactor under ocean conditions using MARS-KS moving reactor model. Progress in Nuclear Energy, 168, 105017.
Yoo, J. S., Lee, C. W., Hong, H., Ko, H., Ku, J. H., Kim, G. W., Park, G. C. & Cho, H. K. (2024). Experimental study on flow boiling CHF in annulus channel under heaving conditions using simulant fluid R134a targeting nuclear reactor applications. Applied Thermal Engineering, 236, 121906.
Yoo, J. S., Lee, C. W., Hong, H., Ko, H., Ku, J. H., Kim, G. W., Park, G. C. & Cho, H. K. (2024). Experimental Investigation of the Heaving Motion Effect on Critical Heat Flux Phenomena and Wall Temperature Fluctuation Using R134a for Floating Nuclear Applications. Nuclear Science and Engineering, 1-17.
Chae, M. S., Lee, J., Cho, H. K., Kelm, S., & Paladino, D. (2024). Analyses of Jet Buoyant Flow in a Multicompartment Containment Using an Open-Source Solver. Nuclear Technology, 1-16.
Kim, M. H., Cho, H. K., & Kim, B. J. (2024). Experimental and numerical investigation of flow dynamics in an upward bubbly flow in a tube undergoing oscillating rolling motion. Physics of Fluids, 36(1).
Lee, C. W., Yoo, J. S., Hong, H., Ko, H., Ku, J. H., Kim, G. W., Park, G. C. & Cho, H. K. (2023). Experimental investigation of helical fin and inclination effect on critical heat flux. International Journal of Heat and Mass Transfer, 215, 124482.
Lee, J., Jeon, S. S., Park, J. Y., & Cho, H. K. (2023). Effect evaluation on performance issues of passive safety system–Part Ⅱ. Passive emergency core cooling system. Nuclear Engineering and Design, 411, 112446.
Kim, G. W., Yoo, J. S., Lee, C. W., Hong, H., Park, G. C., & Cho, H. K. (2023). Critical heat flux correlations for tube and annulus geometries under inclination and rolling conditions. Applied Thermal Engineering, 225, 120131.
Lee, J., Jeon, S. S., Park, J. Y., & Cho, H. K. (2023). Effect evaluation on performance issues of passive safety system–Part Ⅰ. Passive heat removal system. Nuclear Engineering and Design, 403, 112160.
Jeong, M. J., Im, J., Lee, S., & Cho, H. K. (2023). Multiphysics analysis of heat pipe cooled microreactor core with adjusted heat sink temperature for thermal stress reduction using OpenFOAM coupled with neutronics and heat pipe code. Frontiers in Energy Research, 11, 1213000.
Im, J., Jeong, M. J., Choi, N., Kim, K. M., Cho, H. K., & Joo, H. G. (2023). Multiphysics Analysis System for Heat Pipe–Cooled Micro Reactors Employing PRAGMA-OpenFOAM-ANLHTP. Nuclear Science and Engineering, 197(8), 1743-1757.
Seo, H., Choi, M. H., Park, S. W., Kim, G. W., Cho, H. K., & Chung, B. D. (2022). Moving reactor model for the MULTID components of the system thermal-hydraulic analysis code MARS-KS. Nuclear Engineering and Technology, 54(11), 4373-4391.
Kim, G. W., Yoo, J. S., Lee, C. W., Hong, H., Park, G. C., & Cho, H. K. (2022). Critical heat flux characteristics of flow boiling on a heater rod under inclined and rolling conditions. International Journal of Heat and Mass Transfer, 189, 122670.
Kim, S. Y., & Cho, H. K. (2022). Turbulence model assessment and heat transfer phenomena inside a rectangular channel under forced and mixed convection. International Journal of Heat and Mass Transfer, 185, 122388.
Choi, C. J., & Cho, H. K. (2022). Numerical investigations of liquid film offtake by transverse gas flow in a downcomer annulus geometry. Frontiers in Energy Research, 10, 847458.
Kim, G. W., Park, G. C., & Cho, H. K. (2021). Scaling analysis for single-phase natural circulation under dynamic motion and its verification using MARS-KS code. Annals of Nuclear Energy, 159, 108308.
Kim, J. S., & Cho, H. K. (2021). Advanced boiling heat transfer model for a horizontal tube with numerical analysis of bubble behaviours. International Journal of Heat and Mass Transfer, 175, 121168.
Lee, C. W., Yoo, J. S., & Cho, H. K. (2021). Multi-scale simulation of wall film condensation in the presence of non-condensable gases using heat structure-coupled CFD and system analysis codes. Nuclear Engineering and Technology, 53(8), 2488-2498.
Kim, S. Y., Shin, D. H., Kim, C. S., Park, G. C., & Cho, H. K. (2021). Computational fluid dynamics analysis of buoyancy-aided turbulent mixed convection inside a heated vertical rectangular duct. Progress in Nuclear Energy, 137, 103766.
Kim, S. Y., Kim, C. S., & Cho, H. K. (2021). Local flow structure and turbulence quantities inside a heated rectangular riser in turbulent forced and mixed convection heat transfers. Experimental Thermal and Fluid Science, 122, 110297.
Choi, C. J., & Cho, H. K. (2020). Effect of asymmetric airflow on liquid film behavior and emergency core coolant bypass in the downcomer geometry of a nuclear reactor pressure vessel. International Communications in Heat and Mass Transfer, 117, 104765.
Yoon, H. Y., Park, I. K., Lee, J. R., Lee, S. J., Cho, Y. J., Do, S. J., Cho, H. K. & Jeong, J. J. (2020). A multiscale and multiphysics PWR safety analysis at a subchannel scale. Nuclear Science and Engineering, 194(8-9), 633-649.
Kim, J. S., Kim, Y. N., & Cho, H. K. (2020). Predicting the sliding bubble velocity on the lower part of a horizontal tube heater under natural convection based on force balance analysis. International Journal of Heat and Mass Transfer, 151, 119453.
Lee, J. H., Cho, H. K., & Park, G. C. (2019). Three-dimensional looped network analysis code including core thermal analysis model for prismatic very high temperature gas-cooled reactor. International Journal of Thermal Sciences, 143, 76-91.
Beom, H. K., Kim, G. W., Park, G. C., & Cho, H. K. (2019). Verification and improvement of dynamic motion model in MARS for marine reactor thermal-hydraulic analysis under ocean condition. Nuclear Engineering and Technology, 51(5), 1231-1240.
Shin, D. H., Kim, S. Y., Kim, C. S., Park, G. C., & Cho, H. K. (2019). Development of a heat transfer coefficient correlation for buoyancy-aided turbulent mixed convection of air inside a vertical channel. Applied Thermal Engineering, 159, 113884.
Choi, C. J., & Cho, H. K. (2019). Investigation on emergency core coolant bypass with local measurement of liquid film thickness using electrical conductance sensor fabricated on flexible printed circuit board. International Journal of Heat and Mass Transfer, 139, 130-143.
Kim, S. Y., Shin, D. H., Kim, C. S., Park, G. C., & Cho, H. K. (2019). Flow visualization experiment in a two-side wall heated rectangular duct for turbulence model assessment in natural convection heat transfer. Nuclear Engineering and Design, 341, 284-296.
Choe, H. S., Kim, G. W., Park, G. C., Cho, H. K., & Im, K. H. (2018). Application of the mesh adaptation technique to effective heat capacity method for melting simulation of the first wall in breeding blanket under high heat flux condition. Fusion Engineering and Design, 136, 891-896.
Yang, J. H., Euh, D. J., Cho, H. K., & Park, G. C. (2018). Development of wall and interfacial friction models for two-dimensional film flow with local measurement methods. Nuclear Engineering and Design, 336, 141-153.
Lee, K. H., Kim, S. Y., Cho, H. K., & Park, G. C. (2018). Jet impingement model for system analysis code to enhance mixing behavior prediction in downcomer during DVI line break accident. Nuclear Engineering and Design, 334, 121-137.
Lee, J. H., Cho, H. K., & Park, G. C. (2018). Application of three-dimensional looped network analysis method to the core of prismatic very high temperature gas-cooled reactor. Annals of Nuclear Energy, 117, 12-24.
Choi, C. J., Yang, J. H., Euh, D. J., Park, G. C., & Cho, H. K. (2018). Effect of wall friction model on predicting emergency core coolant behavior in upper downcomer with direct vessel safety injection using MARS-KS. Annals of Nuclear Energy, 116, 395-406.
Lee, J., Park, G. C., & Cho, H. K. (2018). Simulation of wall film condensation with non-condensable gases using wall function approach in component thermal hydraulic analysis code CUPID. Journal of Mechanical Science and Technology, 32, 1015-1023.
Yoon, S. J., Kim, S. B., Park, G. C., Yoon, H. Y., & Cho, H. K. (2018). Application of CUPID for subchannel-scale thermal–hydraulic analysis of pressurized water reactor core under single-phase conditions. Nuclear Engineering and Technology, 50(1), 54-67.
Shin, D. H., Kim, C. S., Park, G. C., & Cho, H. K. (2017). Experimental analysis on mixed convection in reactor cavity cooling system of HTGR for hydrogen production. International Journal of hydrogen energy, 42(34), 22046-22053.
Shin, D. H., Yoon, S. J., Cho, H. K., Park, G. C., & Kim, T. (2017). Development of effective thermal conductivity models for Reserve Shutdown Control fuel block of prismatic HTGR for hydrogen production. International Journal of Hydrogen Energy, 42(29), 18614-18625.
Kim, G. W., Cho, H. K., Park, G. C., & Im, K. (2017). Melting and evaporation analysis of the first wall in a water-cooled breeding blanket module under vertical displacement event by using the MARS code. Fusion Engineering and Design, 118, 52-63.
Yang, J. H., Choi, C. J., Cho, H. K., Euh, D. J., & Park, G. C. (2017). Assessment of wall friction model in multi-dimensional component of MARS with air–water cross flow experiment. Nuclear Engineering and Design, 312, 106-120.
Kim, Y. N., Kim, J. S., Park, G. C., & Cho, H. K. (2017). Measurement of sliding bubble behavior on a horizontal heated tube using a stereoscopic image processing technique. International Journal of Multiphase Flow, 94, 156-172.
Lee, K. B., Kim, J. R., Park, G. C., & Cho, H. K. (2016). Feasibility test of a liquid film thickness sensor on a flexible printed circuit board using a three-electrode conductance method. Sensors, 17(1), 42.
Jeon, S. S., Hong, S. J., Cho, H. K., & Park, G. C. (2016). Development of heat transfer model package for horizontal U-shaped heat exchanger submerged in pool of passive safety system. Nuclear Technology, 196(2), 303-318.
Lee, J. H., Cho, H. K., & Park, G. C. (2016). Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200. Nuclear Engineering and Design, 307, 106-118.
Kim, G. W., Lee, J. H., Cho, H. K., Park, G. C., & Im, K. (2016). Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor. Fusion Engineering and Design, 103, 98-109.
Jeon, S. S., Hong, S. J., Cho, H. K., & Park, G. C. (2015). Prediction of nucleate boiling heat transfer on horizontal U-shaped heat exchanger submerged in a pool of water using MARS code. Nuclear Engineering and Design, 295, 317-337.
Yang, J. H., Cho, H. K., Kim, S., Euh, D. J., & Park, G. C. (2015). Experimental study on two-dimensional film flow with local measurement methods. Nuclear Engineering and Design, 294, 137-151.
Lee, J. H., Park, I. W., Kim, G. W., Park, G. C., Cho, H. K., & Im, K. (2015). Thermal-hydraulic analysis of water cooled breeding blanket of K-DEMO using MARS-KS code. Fusion Engineering and Design, 98, 1741-1746.
Lee, J. H., Yoon, S. J., Cho, H. K., Jae, M., & Park, G. C. (2015). Experimental investigation and CFD analysis on cross flow in the core of PMR200. Annals of Nuclear Energy, 83, 422-435.
Suh, J. K., Kim, J. W., Kwon, S. G., Lee, J. Y., Cho, H. K., & Park, G. C. (2015). Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly. Nuclear Engineering and Design, 289, 49-59.
Lee, J., Park, G. C., & Cho, H. K. (2015). Improvement of CUPID code for simulating filmwise steam condensation in the presence of noncondensable gases. Nuclear Engineering and Technology, 47(5), 567-578.
Cho, H. K., Cho, Y. J., & Yoon, H. Y. (2014). Heat structure coupling of CUPID and MARS for the multi-scale simulation of the passive auxiliary feedwater system. Nuclear Engineering and Design, 273, 459-468.
Yoon, H. Y., Lee, J. R., Kim, H., Park, I. K., Song, C. H., Cho, H. K., & Jeong, J. J. (2014). Recent improvements in the CUPID code for a multi-dimensional two-phase flow analysis of nuclear reactor components. Nuclear Engineering and Technology, 46(5), 655-666.
Kim, H., Kim, S. H., Lee, S. J., Park, I. K., Yoon, H. Y., Cho, H. K., & Jeong, J. J. (2014). Development of CUPID-SG for the analysis of two-phase flows in PWR steam generators. Progress in Nuclear Energy, 77, 132-140.
Park, I. K., Yoon, H. Y., & Cho, H. K. (2013). Simulations of air–water flow and subcooled boiling flow using the CUPID code. Journal of Nuclear Science and Technology, 50(8), 813-827.
Yoon, H. Y., Jeong, J. J., Cho, H. K., Bang, Y. S., & Seul, K. W. (2013). A multi-scale analysis of the transient behavior of an advanced safety injection tank. Annals of Nuclear Energy, 62, 17-25.
Cho, H. K., Lee, S. J., Yoon, H. Y., Kang, K. H., & Jeong, J. J. (2013). Simulation of single-and two-phase natural circulation in the passive condensate cooling tank using the CUPID code. Journal of Nuclear Science and Technology, 50(7), 709-722.
