Investigation of Transients Important for Heat Pipes in Microreactors
Heat pipe microreactors (HPMRs) are compact, mobile nuclear reactors that are designed to provide carbon-free power in remote applications such as space, military, mining, disaster relief and backup power. An extensive experimental database is required to aid and facilitate novel HPMR designs by improving modeling techniques, and developing new mechanistic models.
Heat pipes in microreactor applications contain liquid metal working fluids for high temperature operation, which makes it challenging and costly to conduct experiments in laboratory. In addition, working with liquid metal heat pipes requires extensive safety precautions and working within an isolated environment. The Low Temperature Heat Pipe Test Facility (LTHPF) addresses these difficulties and allows internal flow temperature and pressure measurements along with flow visualization capabilities. The choice of the low-temperature working fluids at low pressures is based on a detailed scaling analysis on the phenomena of interest.
Modeling and Prediction of Cryogenic Propellant Behavior in Microgravity
NASA's future missions require the storage of cryogenic propellant in space for a long duration. Over time, heat leaks into the storage tank and pressurizes the tank due to the vaporization of the propellant. To conserve the mass, venting is undesirable. Cost savings associated with Zero Boil Off (ZBO) has been identified as a major objective. Therefore, it is vital to simulate the thermal and fluid dynamics inside the tank accurately for the design optimization of cryogenic fluid management technologies associated with the depressurization process of the tank. Current high fidelity CFD codes cannot be directly used to simulate long transients (order of months) due to computational costs. To overcome this challenge, we are developing a nodal model to simulate the long transient process by incorporating physics-informed correlations developed from CFD simulations and leveraging nodal model's computational efficiency.
Annular Two-Phase Flow Measurements and Modeling
Annular two-phase flows, where the liquid and gas flow co-currently or counter-currently, extensively exist during phase change heat transfer in nuclear reactors and thermal management solutions. Annular flows typically feature high gas quality and relatively thin liquid film on the wall. The interface between the gas core and liquid film is wavy and sometimes unstable, which could lead to liquid droplets entrainment into the gas core and potential heat transfer deterioration or dryout under certain scenarios. An annular flow test facility is designed and constructed at THL to perform air-water annular two-phase flow experiments. The test facility consists of three major subsystems: the water supply system, the air supply system, and the test section equipped with various two-phase flow instruments. In addition, the test facility is designed to conduct both co-current and counter-current annular flow experiments by switching the inlet and extraction ports in the test section.
Modeling and Simulation of Fuel Dispersion Behavior
With transition to high-burnup fuels for cost competitiveness, the investigation of fuel behavior becomes increasingly crucial in assuring the safe operation of Light Water Reactors (LWRs). Fuel fragmentation, relocation, and dispersal (FFRD) studies are critical for ensuring post-accident nuclear fuel assembly management and limiting radiation spread beyond the containment boundary. During a design-based accident scenario such as a loss of coolant accident (LOCA), the fuel cladding temperature may rise over the design limit, causing the cladding to rupture and the radioactive fuel fragments to be released into the coolant system. This represents a complex three-phase flow phenomenon where molten or solid fuel fragments carried by high-pressure fission gasses immediately interact with the surrounding fluid media. The current study focusses on developing a simulation framework to capture the trajectory and settlement behavior of the fuel fragments in the primary system. From a modeling and simulation point of view, the key aspect is to establish a set of closure relations for the interfacial property exchange terms that govern the particle trajectories after the fuel dispersion. Moreover, the developed modeling framework can be implemented to similar safety analyses in other reactor types such as Sodium-cooled Fast Reactors.
Thermal Power Dispatch System Design for Nuclear Flexible Plant Operation and Generation
Hydrogen could become a widely used energy source in the future. One method of producing hydrogen is by high temperature steam electrolysis (HTSE). The necessary thermal energy for the HTSE process could be provided by nuclear power plants, which produce a large amount of thermal power in the form of steam that is delivered to turbines for electricity generation. By diverting a fraction of this steam and some of the generated electricity, a supply of deionized water can be heated to be used in the HTSE plant. This method would be able to efficiently produce hydrogen at a larger scale without producing undesired byproducts like carbon dioxide.
The goal of this project is to design a system that extracts thermal power from a nuclear power plant with a thermal power extraction (TPE) system, and deliver it to HTSE facilities through a thermal power delivery (TPD) system. The challenge of this project is to ensure the systems can deliver a varying amount of heat extracted from the plant, while ensuring the impact on the nuclear power plant operation is minimized. To do this, thermal hydraulic analysis can be conducted on these systems using a system analysis code such as RELAP-5. In addition, simulations of power plant operations with the integrated TPE/TPD systems can be conducted to ensure they are safe, optimized, and can be deployed into existing nuclear power plants.
Photron Mini UX100 High-speed camera
LUNA ODiSI 6000 series sensing instruments
Cluster access, including the state-of-the-art AiMOS supercomputer
Low-Temperature Heat Pipe Test Facility (LTHPF)
Annular Two-Phase Flow Test Facility