Advanced reactor technologies have generated interest for their potential to reduce fossil fuel emissions, improve energy efficiency and cut down on nuclear waste. We have designed, constructed , and operated a specialized test facility to measure hydraulic parameters and validate computational … [Read more...]
The Water-Cooled Reactor Cavity Cooling System (RCCS)
The Very High Temperature Gas-Cooled Reactor (VHTR) is one of the six Gen IV reactor designs, which will generate electricity and high temperature heat for industrial applications. The design of the VHTR includes a new passive safety system that is designed to remove heat and maintain temperatures … [Read more...]
Pebble Bed Test Facility
The Advanced High-Temperature Reactor (AHTR) concept leverages a particle-based fuel format consisting of discrete spherical graphite pebbles arrayed in a packed bed architecture. Thermal regulation achieved via flow of gas (e.g., helium) or liquid (e.g., molten salt) coolants through the void … [Read more...]
Pressurized Thermal Shock (PTS) Facility: “The Cold Leg Mixing CFD-UQ benchmark”
The 5th benchmark of CFD applications to Nuclear reactor safety has been approved by CSNI. Its main objective is to go a step further in the application of single-phase CFD to nuclear safety issues with mixing problems possibly in presence of buoyancy effects. The Pressurized Thermal Shock (PTS) … [Read more...]
The Generic Safety Issue 191
Our research team, in collaboration with the South Texas Project Nuclear Operating Company (STPNOC), has solved the Generic Safety Issue (GSI) - 191, a problem resulting from a loss of coolant accident in a nuclear reactor, which can cause debris to be generated and potentially impact the … [Read more...]
Experimental Studies of PWR 5×5 Fuel Bundles
Fuel assemblies of pressurized water reactors (PWRs) are constructed from fuel rods and supports to create rod bundles. To support and maintain the lateral spatial gaps of the rods, support grids are positioned at regular intervals in the direction of axial flow. With the presence of the spacer … [Read more...]
Helical Coil Steam Generator Test Loop
The helical coil steam generator is a specific type of Tube and Shell Heat Exchanger known for having a higher heat transfer coefficient than many other designs. We have designed, constructed, and operated three test facilities with main focus on the shell side flow for a specific design of the … [Read more...]
The V&V Benchmark Problem: Twin Jet Experimental Facility
In early 2016, the ASME Verification and Validation in Computational Nuclear System Thermal Fluids Behavior Committee (ASME V&V30 Standard Committee) has selected the high-fidelity experimental data obtained in the Twin Jets experimental facility as the first V&V benchmark problem. The … [Read more...]
Subcooled Flow Boiling Experiments
Subcooled flow boiling experimental benchmarks in simple geometries are part of an ongoing two-phase flow fundamental research performed at the Thermal-hydraulics laboratory in the Nuclear Engineering Department. State-of-the-art visualization techniques are simultaneously implemented, … [Read more...]
GSI-191
Short Summary:The thermal-hydraulic team has actively contributed to the Generic Safety Issue (GSI) 191 resolution project, developing computational and experimental capabilities and supporting the nuclear power industry with thermal-hydraulic analyses of different aspects of the issue.Abstract:The … [Read more...]
Steam Condensation Test Facility
The steam-air condensation facility investigates the phenomena of direct contact condensation that occurs when steam comes in direct contact with water creating a violent and complicated exchange of mass and energy. The heat transfer coefficient between the steam and water is calculated, as well as … [Read more...]
Evaluation and Testing of HTGR Reactor Building Response to Depressurization Accidents
The analysis and understanding of air ingress events are an important aspect of the design of high-temperature gas-cooled reactor (HTGR). During accident conditions (depressurized loss of forced cooling, D-LOFC) air may enter into the reactor pressure vessel as a result of a break in the helium … [Read more...]
Critical Heat Flux Test Facility
The Critical Heat Flux / Departure from Nucleate Boiling test loop provides a means of evaluating the DNB performance of test geometries representative of commercial pressurized water reactor (PWR) nuclear fuel. The test facility has operated at the Product Engineering Development Laboratory, which … [Read more...]
HTGR Upper Plenum Natural Circulation Facility
One major accident of interest in the High Temperature Gas-Cooled Reactors is the Pressurized Conduction Cooldown (PCC) scenario. The PCC scenario results in a loss of forced convection to the core, while the loop stays pressurized since there is no breach in the boundary. The coolant flows through … [Read more...]
