552 results on '"Tian, Wenxi"'
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2. Code development and preliminary validation for lead-cooled fast reactor thermal-hydraulic transient behavior
3. Development of a neutron space–time kinetics solver with improved quasi-static method based on OpenFOAM
4. Optimization analysis of nuclear thermal coupling for a small nuclear thermal propulsion (SNTP) reactor
5. Model development for oxidation and degradation behavior of accident tolerant Cr coating on Zr alloy cladding under high temperature steam atmosphere
6. Experimental investigation on heat transfer of nitrogen flowing in a circular tube
7. Simulation of threshold displacement energy in Fe-Cr-Al alloys using molecular dynamics
8. Extension of GeN-Foam to departure from nucleate boiling prediction and validation against the OECD/NRC PSBT benchmark
9. The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0
10. Uncertainty analysis of molten corium In-Vessel Retention under External Reactor Vessel Cooling
11. Coupled neutronics, thermal-hydraulics, and fuel performance analysis of dispersion plate-type fuel assembly in a cohesive way
12. Three-dimensional thermal-hydraulics/neutronics coupling analysis on the full-scale module of helium-cooled tritium-breeding blanket
13. Analysis of thermal hydraulic system code LOCUST V1.2 with ECC thermal mixing test facility
14. Structure optimization of helium-cooled blanket for fusion reactor based on three-dimensional full-scale thermal-hydraulic analysis
15. Investigation of plate fuel performance under reactivity initiated accidents with developed multi-dimensional coupled method
16. Numerical simulation of natural convection around the dome in the passive containment air-cooling system
17. Extension of GeN-Foam to modeling of boiling water and validation against the OECD/NRC PSBT benchmark
18. Thermal-hydraulic analysis of He–Xe gas mixture in 2×2 rod bundle wrapped with helical wires
19. Investigations on thermoelectric characteristic of heat pipe thermoelectric generator waste heat utilization device in nuclear power system
20. Study on iterative algorithm of full plate cross-flow and counter-flow printed circuit heat exchanger for fluoride-salt-cooled high-temperature advanced reactor
21. Analysis of containment pressure control strategy in HPR1000 NPP under severe accidents
22. Development of a hybrid model for fuel performance analysis of spherical fuel elements under normal conditions
23. CorTAF: A nuclear reactor core three-dimensional thermal-hydraulic characteristics analysis code based on OpenFOAM
24. Molecular dynamics study of the wettability effect on the evaporation of thin liquid sodium film
25. The evaporation of nanoscale sodium liquid film on the non-ideal nanostructure surface: A molecular dynamics study
26. Numerical simulation of corrosion phenomena in oxygen-controlled environment for a horizontal lead-bismuth reactor core
27. Melting and relocation behavior of metal and force balance analysis of metallic droplet
28. Experimental investigation of nano-particle deposited wick Structure's heat transfer characteristics
29. Multi-physics coupling analysis on neutronics, thermal hydraulic and mechanics characteristics of a nuclear thermal propulsion reactor
30. Water film covering characteristic on horizontal fuel rod under impinging cooling condition
31. SEINA: A two-dimensional steam explosion integrated analysis code
32. An improved MPS-DEM numerical model for fluid–solid coupling problem in nuclear reactor
33. Molecular dynamics study of liquid sodium film evaporation and condensation by Lennard-Jones potential
34. Thermal-hydraulic research on rod bundle in the LBE fast reactor with grid spacer
35. Preliminary development of a multi-physics coupled fuel performance code for annular fuel analysis under normal conditions
36. Study on the flow and heat transfer characteristics of liquid sodium in a hexagonal 7-rod bundle
37. Study on high-temperature hydrogen dissociation for nuclear thermal propulsion reactor
38. Parameter sensitivity study on startup characteristics of high temperature potassium heat pipe
39. Heat transfer analysis of heat pipe cooled device with thermoelectric generator for nuclear power application
40. Development of oxidation model for zirconium alloy cladding and application in the analysis of cladding behavior under loss of coolant accident
41. Study on startup characteristics of prototype once-through steam generator for China fast reactor
42. Assessment of the severe accident code MIDAC based on FROMA, QUENCH-06&16 experiments
43. CFD investigation of the cold wall effect on CHF in a 5 × 5 rod bundle for PWRs
44. Research on the nuclear fuel rods melting behaviors by alternative material experiments
45. CFD/RELAP5 coupling analysis of the ISP No. 43 boron dilution experiment
46. Numerical analysis of melt migration and solidification behavior in LBR severe accident with MPS method
47. Experimental study on the influence of heating surface inclination angle on heat transfer and CHF performance for pool boiling
48. Flow and heat transfer evaluation of lead-bismuth eutectic coolant for nuclear power application
49. Impact of axial power distribution on thermal-hydraulic characteristics for thermionic reactor
50. Parallelization and application of SACOS for whole core thermal-hydraulic analysis
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