This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: ""Experimental Study of RCCS for the HTGR"". Contributions include va...
Graduation date: 2017The Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system ...
© 2020 American Nuclear Society. Investigating thermal stratification in the upper plenum of a sodiu...
The 10 MWth high-temperature test reactor (HTR-10) is a proof of concept reactor prototype for the h...
The design of passive heat removal systems is one of the main concerns for the modular Very High Tem...
The present paper deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the co...
The present document deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the...
One of the candidates for advanced reactor designs identified for the Next Generation Nuclear Plant ...
International audienceFormation and destruction of thermal stratification can occur under certain fl...
International audienceThe paper presents the numerical analysis of core thermal-hydraulic performed ...
National audienceFormation and destruction of thermal stratification can occur under certain flow co...
Graduation date: 2006The Very High Temperature Reactor (VHTR) design was one of six designs chosen\u...
In the development and safety evaluation process of nuclear reactors, the thermal hydraulic analysis...
In the U.S., the helium-cooled VHTR has become the centerpiece of DOE NGNP program. Among the VHTR d...
The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHT...
Graduation date: 2006Among the Generation IV reactors, the Very High Temperature Reactor (VHTR) is b...
Graduation date: 2017The Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system ...
© 2020 American Nuclear Society. Investigating thermal stratification in the upper plenum of a sodiu...
The 10 MWth high-temperature test reactor (HTR-10) is a proof of concept reactor prototype for the h...
The design of passive heat removal systems is one of the main concerns for the modular Very High Tem...
The present paper deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the co...
The present document deals with scaling in nuclear-system thermal-hydraulics (SYS TH), including the...
One of the candidates for advanced reactor designs identified for the Next Generation Nuclear Plant ...
International audienceFormation and destruction of thermal stratification can occur under certain fl...
International audienceThe paper presents the numerical analysis of core thermal-hydraulic performed ...
National audienceFormation and destruction of thermal stratification can occur under certain flow co...
Graduation date: 2006The Very High Temperature Reactor (VHTR) design was one of six designs chosen\u...
In the development and safety evaluation process of nuclear reactors, the thermal hydraulic analysis...
In the U.S., the helium-cooled VHTR has become the centerpiece of DOE NGNP program. Among the VHTR d...
The Very High Temperature Reactor (VHTR) is a proposed next generation nuclear power plant. The VHT...
Graduation date: 2006Among the Generation IV reactors, the Very High Temperature Reactor (VHTR) is b...
Graduation date: 2017The Direct Reactor Auxiliary Cooling System (DRACS) is a passive safety system ...
© 2020 American Nuclear Society. Investigating thermal stratification in the upper plenum of a sodiu...
The 10 MWth high-temperature test reactor (HTR-10) is a proof of concept reactor prototype for the h...