342 results on '"Park, Goon-Cherl"'
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2. Experimental study on flow boiling CHF in annulus channel under heaving conditions using simulant fluid R134a targeting nuclear reactor applications
3. Experimental investigation of helical fin and inclination effect on critical heat flux
4. Critical heat flux correlations for tube and annulus geometries under inclination and rolling conditions
5. Corrigendum to “Experimental study on flow boiling CHF in annulus channel under heaving conditions using simulant fluid R134a targeting nuclear reactor applications” [Applied Thermal Engineering 236 (2024) 121906]
6. Critical heat flux characteristics of flow boiling on a heater rod under inclined and rolling conditions
7. Scaling analysis for single-phase natural circulation under dynamic motion and its verification using MARS-KS code
8. Computational fluid dynamics analysis of buoyancy-aided turbulent mixed convection inside a heated vertical rectangular duct
9. Three-dimensional looped network analysis code including core thermal analysis model for prismatic very high temperature gas-cooled reactor
10. Verification and improvement of dynamic motion model in MARS for marine reactor thermal-hydraulic analysis under ocean condition
11. Development of a heat transfer coefficient correlation for buoyancy-aided turbulent mixed convection of air inside a vertical channel
12. Flow visualization experiment in a two-side wall heated rectangular duct for turbulence model assessment in natural convection heat transfer
13. 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
14. Development of wall and interfacial friction models for two-dimensional film flow with local measurement methods
15. Validation of Wall Friction Model in Multidimensional Component of Mars with Two-Phase Flow Experiments Describing ECC Behavior in Downcomer
16. Jet impingement model for system analysis code to enhance mixing behavior prediction in downcomer during DVI line break accident
17. Application of three-dimensional looped network analysis method to the core of prismatic very high temperature gas-cooled reactor
18. Effect of wall friction model on predicting emergency core coolant behavior in upper downcomer with direct vessel safety injection using MARS-KS
19. Application of CUPID for subchannel-scale thermal–hydraulic analysis of pressurized water reactor core under single-phase conditions
20. Measurement of sliding bubble behavior on a horizontal heated tube using a stereoscopic image processing technique
21. Experimental analysis on mixed convection in reactor cavity cooling system of HTGR for hydrogen production
22. Development of effective thermal conductivity models for Reserve Shutdown Control fuel block of prismatic HTGR for hydrogen production
23. Melting and evaporation analysis of the first wall in a water-cooled breeding blanket module under vertical displacement event by using the MARS code
24. Assessment of wall friction model in multi-dimensional component of MARS with air–water cross flow experiment
25. Development of the loss coefficient correlation for cross flow between graphite fuel blocks in the core of prismatic very high temperature reactor-PMR200
26. Simulation of wall film condensation with non-condensable gases using wall function approach in component thermal hydraulic analysis code CUPID
27. Experimental Study on Flow Boiling Chf in Annulus Channel Under Heaving Conditions Using Simulant Fluid R134a Targeting Nuclear Reactor Applications
28. EXPERIMENTAL INVESTIGATIONS ON EVAPORATIVE HEAT TRANSFER OF WATER FILM IN A VERTICAL DUCT WITH COUNTER-CURRENTLY PASSED AIR CONDITION
29. A STUDY ON THE THERMAL-HYDRAULIC PHENOMENA OF DECAY HEAT REMOVAL SYSTEM USING THE CLOSED TWO-PHASE THERMOSIPHON LOOP
30. EXPERIMENTAL STUDY ON PHASE DISTRIBUTION IN INCLINED SUBCOOLED BOILING ANNULUS
31. Experimental Investigation of Helical Fin and Inclination Effect on Critical Heat Flux
32. Development of thermal-hydraulic analysis methodology for multiple modules of water-cooled breeder blanket in fusion DEMO reactor
33. Finite element simulation of creep crack growth using combined plastic-creep damage model
34. Prediction of nucleate boiling heat transfer on horizontal U-shaped heat exchanger submerged in a pool of water using MARS code
35. Experimental study on two-dimensional film flow with local measurement methods
36. Reduction of circulation power for helium-cooled fusion reactor blanket using additive CO2 gas
37. Experimental investigation and CFD analysis on cross flow in the core of PMR200
38. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly
39. Assessment of TAPINS and TASS/SMR codes and application to overpower transients in REX-10
40. Improvement of core effective thermal conductivity model of GAMMA+ code based on CFD analysis
41. Numerical and experimental investigation on labyrinth seal mechanism for bypass flow reduction in prismatic VHTR core
42. Assessment of horizontal in-tube condensation models using MARS code. Part II: Annular flow condensation
43. Experimental Assessment of GOTHIC Code for Local Hydrogen Combustion Analysis
44. Assessment of horizontal in-tube condensation models using MARS code. Part I: Stratified flow condensation
45. Characteristics of critical heat flux under rolling condition for flow boiling in vertical tube
46. Design and analysis of a high pressure and high temperature sulfuric acid experimental system
47. The effects of crossflow gap and axial bypass gap distribution on the flow characteristics in prismatic VHTR core
48. Development of a thermal–hydraulic system code, TAPINS, for 10 MW regional energy reactor
49. Condensation dynamics of submerged steam jet in subcooled water
50. Critical Heat Flux Correlations for Tube and Annulus Geometries Under Inclination and Rolling Conditions
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