TAILIEUCHUNG - THERMAL-HYDRAULIC IN NUCLEAR REACTOR

I. Introduction An important aspect of nuclear reactor core analysis involves the determination of the optimal coolant flow distribution and pressure drop across the reactor core. On the one hand, higher coolant flow rates will lead to better heat transfer coefficients and higher Critical Heat Flux (CHF) limits. On the other hand, higher flows rates will also in large pressure drops across the reactor core, hence larger required pumping powers and larger dynamic loads on the core components. Thus, the role of the hydrodynamic and thermal-hydraulic analysis is to find proper operating conditions that assure both safe and economical operation of. | THERMAL-HYDRAULIC IN NUCLEAR REACTOR GS. Trần Đại Phúc THERMAL-HYDRAULIC IN NUCLEAR REACTOR Summary Introduction Energy from fission Fission yield Decay heat Spatial distribution of heat sources Coolant flow & heat transfer in fuel rod assembly Enthalpy distribution in heated channel Temperature distribution in channel in single phase Heat conduction in fuel assembly Axial temperature distribution in fuel rod Void fraction in fuel rod channel Heat transfer to coolant THERMAL-HYDRAULIC IN NUCLEAR REACTOR I. Introduction An important aspect of nuclear reactor core analysis involves the determination of the optimal coolant flow distribution and pressure drop across the reactor core. On the one hand, higher coolant flow rates will lead to better heat transfer coefficients and higher Critical Heat Flux (CHF) limits. On the other hand, higher flows rates will also in large pressure drops across the reactor core, hence larger required pumping powers and larger dynamic loads on the core .

• Năng lượng
• thermal hydraulic
• nuclear reactor
• energy from fission
• fission yield
• decay heat
• coolant flow
• heat transfer
• Nuclear engineering and technology
• Sensitivity analysis of thermal hydraulic parameters
• The corrosion intensity in nuclear power plant steam generators
• Nuclear power plant steam generators
• The failure of steam generators
• The heat exchangers
• Multivariate analysis of critical parameters influencing the reliability
• Thermal hydraulic passive safety system
• Critical parameters influencing the reliability
• Two multivariate analysis methods
• The identified key CPs
• Water cooled steam generator
• Nuclear energy
• Gyeongju bentonite
• Thermal conductivity
• Hydraulic conductivity
• Temperature effect
• Thermal hydraulic analysis of core flow bypass
• Typical research reactor
• Core flow bypass
• Safety margins during normal operation
• Thermal hydraulic study of air cooled passive decay heat removal system
• Air cooled passive decay heat removal system
• APRþ under extended station blackout
• Extended Station Blackout
• Thermal hydraulic RELAP5 3D code
• Geometrical and material data
• Angra 2 final safety analysis report
• RELAP5 3D model
• Angra 2 PWR
• Thermal hydraulic analysis
• Fuel thermal conductivity
• Er doped assembly
• Erbia fully poisoned PWR
• Erbia credit super high burnup fuel concept
• WWR K reactor
• VVR KN fuel lead test
• Centre for Nuclear Energy Science and Technology
• Lead test fuel assemblies
• Thermal hydraulic cod
• Safety analysis methodology for aged candu nuclear reactors
• Safety analysis methodology
• Slow Loss of Regulation
• Heat Transport System
• The thermal hydraulic analysis
• Statistical method
• Interfacial shear stress
• Small Break LOCA
• Best Estimate Plus Uncertainty
• Nuclear Power Plant
• Thermal hydraulic behavior
• VVER core reflooding
• Fuel assembly mockup
• Two types of flooding
• VVER reactor core flooding
• Thermal hydraulic processes
• Test Facility ROSA 2/LSTF
• Calculation code CATHARE
• 2 V2 5 OECD/NEA ROSA 2
• Intermediate Break Loss Of Coolant Accident
• Physical models using in RELAP5 code
• FEBA experimental data
• Thermal hydraulic safety analysis
• Capor flow rate
• Emergency core cooling system
• Super critical light water reactor
• Linear stability
• Nonlinear stability
• Subcritical hopf
• Supercritical hopf
• Generalized hopf
• Nuclear engineering and design
• TRACE analyses
• Atmospheric pressures
• Thermal–hydraulic codes
• Temperature conditions
• Super Critical Water cooled Reactor
• Several thermal–hydraulic codes
• System code
• Safety analysis
• Real time interactive transient simulator
• Dalat nuclear research reactor
• Human machine interface
• The safety analysis report
• Best estimate thermal hydraulic