Two important correlations have been obtained to calculate the maximum coolant temperature and the Nusselt number in the wake (the flow recirculating zone) downstream of blockages in 19-pin sodium-cooled bundles. These correlations can be applied to predict the maximum temperature rise and the average heat transfer coefficient behind small non-heat-generating blockages in the fuel assemblies of the FFTF and the CRBR for the wide range of flow and power conditions. Experiments with partial blockages in simulated LMFBR fuel assemblies have been performed at the THORS facility (formerly called FFM) of the Oak Ridge National Laboratory. Nineteen-pin sodium-cooled bundles were used, which had the same pin diameter and pitch as the CRBR and the F...
Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local s...
Thermal-hydraulic analysis was conducted of a 7-pin fuel bundle by using CFD. Currently, wire-wrappe...
Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled ...
To assess the effect of partial flow blockages on the local temperature distributions in LMFBR fuel ...
Experimental and analytical investigations performed in the United States, Germany, Great Britain, a...
Heat transfer coefficients were experimentally determined for 31-rod sodium-cooled bundle with a 6-s...
A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coo...
The recent modification of the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility at ORNL will...
Fuel pins in Sodium cooled Fast Reactors (SFR) are arranged in a tightly packed triangular pitch wit...
Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partial...
The power density in the core of the block next generation nuclear power plant (NGNP) will not be un...
In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the tempera...
The thermal-hydraulic code TRACE is currently being extended at the Paul Scherrer Institute (PSI) fo...
An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle...
In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the tempera...
Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local s...
Thermal-hydraulic analysis was conducted of a 7-pin fuel bundle by using CFD. Currently, wire-wrappe...
Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled ...
To assess the effect of partial flow blockages on the local temperature distributions in LMFBR fuel ...
Experimental and analytical investigations performed in the United States, Germany, Great Britain, a...
Heat transfer coefficients were experimentally determined for 31-rod sodium-cooled bundle with a 6-s...
A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coo...
The recent modification of the Thermal-Hydraulic Out-of-Reactor Safety (THORS) facility at ORNL will...
Fuel pins in Sodium cooled Fast Reactors (SFR) are arranged in a tightly packed triangular pitch wit...
Three-dimensional structures of a vortical flow field and heat transfer characteristics in a partial...
The power density in the core of the block next generation nuclear power plant (NGNP) will not be un...
In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the tempera...
The thermal-hydraulic code TRACE is currently being extended at the Paul Scherrer Institute (PSI) fo...
An experimental campaign was performed on a non-uniformly heated 19-pins wire-spaced fuel pin bundle...
In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the tempera...
Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local s...
Thermal-hydraulic analysis was conducted of a 7-pin fuel bundle by using CFD. Currently, wire-wrappe...
Two three-pin cluster tests simulating the Unprotected Loss-of-Flow (ULOF) accident of Sodiumcooled ...