The Bubble Investigation Loop (BIL) facility is a flow loop designed to allow particle image velocimetry measurements to be take on the flow around bubbles. Particle image velocimetry (PIV) is a method of analyzing a flows velocity field by seeding it with micron size particles. The particles are selected, by size and density, such that they are Stokesian in their behavior and, therefore, follow the path of the fluid closely. A laser sheet is passed though the fluid perpendicular to a high speed camera. The light produced by the laser illuminates the seed particles allowing their position to be captured. The images captured by the camera can be analyzed to determine the velocity field inside the field of view. This is done by measuring the statistical change in the position of a small group of particles from one frame to another through a correlation calculation. Measurement of the velocity field provides the raw data for the computation of material derivative and viscous term in the Navier-Stokes equations, creating an expression for the pressure gradient. This expression can then be integrated allowing for determination of the corresponding pressure field. The specific objectives of the data produced by the BIL facility are a reevaluation of bubble drag and virtual mass coefficients as well as an investigation into the physics which govern larger bubble geometries.
The camera-observed fuel fragmentation & expulsion experiment (COFFEE) facility is a benchtop-scale flow loop and static fluid tank housed in the LIFT lab in the radiation center, which is investigating the explosive dispersion of fragmented nuclear fuel which may occur during a loss-of-coolant accident (LOCA), one of the most limiting design-basis accidents in a nuclear reactor. This phenomena is especially important in modern operation of nuclear reactors as the push for higher burnup fuel necessitates re-evaluating the LOCA safety criteria, and the performance of fuel under such an accident.
The experimental facility houses 1:1 ratio geometries of standard pressurized water reactor (PWR) fuel rods in a square array. The single test rod in the experiment is filled with a surrogate nuclear fuel, and the remaining rods in the array are comprised of solid quartz rods. The working fluid in the system is a light mineral oil, which provides index of refraction matching to the quartz rods, rendering them transparent. This allows for optical access to the test rod, while maintaining the appropriate boundary conditions in the experiment. The test rod is connected to a charge line via a solenoid valve which allows for rapid blowdown of pressures up to 12 MPa. The subsequent dispersion of fuel is captured via two high-speed cameras in order to analyze the dispersion at both a high speed and resolution. Based on measurements from the high-speed particle data, drag coefficients of the fragmented fuel are extracted and input into a finite-element model which has been developed specifically for this application. The ultimate goal is to estimate fuel dispersion on a large scale to predict the movement and deposition location of fuel fragments post-rupture.
The Camera-Observed and Instrumented Loop (COIL) Facility uses a high speed camera and pressure transducers to analyze cross-flow induced vibrations of a test element over a range of flow regimes. A clear glass cylinder allows for visual access to the element under analysis and an angled mirror provides a simultaneous second view for three-dimensional reconstruction. The initial test element is an aluminum wire wound into a helical coil and attached to an inner cylinder. Piping connects the test section to a fluid reservoir with flow provided by a circulating pump. Flow control meters provide manual control of flow rate. Pressure transducers, before and after the test section, meter gauge and differential pressure across the helical coil.
The purpose of this facility is to fully characterize the motion of a helical coil over a range of flow rates. This will provide insight into how steady flow affects flexible structural elements under similar conditions. A similar geometry is found in current integral Pressurized Water Reactor (iPWR) steam generators. It will also provide empirical data on the pressure drop across a coil under fluid flow when suspended in an annulus.
Flow in this facility has a range between 0.4 and 14 gpm with a fine adjustment resolution of 0.2 gpm. Internal gauge pressure in the test section varies based on the flow conditions, ranging from 0.0 to 12 psig. Pressure drop across the test section can be detected with 0.015 psig accuracy. Adjustment of camera placement gives a range of pixel resolution, from 0.002 to 0.006 inches per pixel.
The Global Threat Reduction Initiative (GTRI) is a National Nuclear Security Administration (NNSA) nonproliferation program which aims to convert research reactors from the use of highly enriched uranium (HEU) to low enriched Uranium (LEU). Oregon State University is contracted by the Idaho National Laboratory to aid in the qualification of an ultra-high density Uranium-Molybdenum fuel for use in these reactors. As part of this project, the Endurance Flow Loop will study the long-term effects of the in-pile conditions on the fuel elements over the cycle length of the reactor, which is about one year
The Endurance Flow Loop (EFL) is a parallel system welded stainless steel flow loop designed to force a controlled volumetric rate of water over mock-up reactor fuel plates to simulate in-pile hydraulic conditions. Although compact in size, the EFL is outfitted with a number of different flow paths balanced by strategically placed valves to divert flow in the necessary manner to achieve desired flow and pressure conditions over the fuel elements. All hardware and sensors on the EFL are controlled or monitored through a Data Acquisition System (DAS) and displayed in a Graphical User Interface (GUI) control panel on the computer.
The driving force for the flow in the EFL is the twin-stage 10-hp vertical in-line pump that can deliver 35-225 GPM and can provide a differential pressure of up to 200 ft. dynamic head. This pump is controlled via a Variable Speed Drive (VSD) which adjusts the rotational speed of the pump. Flow through the EFL is measured from three vortex-style Rosemount flowmeters, which cover the whole operational range of the system. Pressure transmitters are placed around the facility to monitor the gage pressure of the inlet test sections, and the differential pressure across the fuel elements, and across the pump. Three thermocouples are located on the facility to measure the bulk coolant temperature and the differential temperature across the pump. Water quality is measured from the PH and Conductivity probes.
The restart of the TRansient REactor Test (TREAT) facility combined with the shutdown of the Halden reactor in Norway lead to an increased interest in mimicking Reactivity Initiated Accident (RIA) events conditions for Light Water Reactor within TREAT itself. One of the engineering challenges for this task is to increase the energy deposition rate on the tested fuel element by shortening the width of the power burst from its current 89 msec down to 40 msec. One of the options proposed by Idaho National Laboratory consists of pressurizing rapidly a specially designed hollow control rod with Helium-3. The insertion of negative reactivity in the form of Helium-3, a strong neutron absorber, must be well predicted and repeatable. In this prospect, an out-of-pile prototype, the Helium Negative Reactivity Insertion (HENRI) facility, has been designed and built at Oregon State University to assess the feasibility, the repeatability and the control of such process. Tests have been performed to assess the effect of the driving pressure and the minimum opening area. These tests, although performed with rupture disks, helped shed light on the physics occurring in the system and gain confidence in the predictability of the helium density evolution. The latter is critical to characterize properly the effect of the Helium-3 injection on the overall core reactivity.
The Hydro-Mechanical Fuel Test Facility (HMFTF) is a large-scale thermal-hydraulic separate effects test facility located in the Advanced Nuclear Systems Engineering Laboratory (ANSEL) at OSU. The facility operates under an NQA-1 compliant quality assurance program and is currently listed on the Idaho National Laboratory Quality Supplier List as a level 1 supplier. The facility is designed such that any element which can fit within the inner vertical height of the test section region may be tested. This is limited to a component of 15 foot total length (shown in first picture below).
OSU has been tasked by the United States Department of Energy Global Threat Reduction Initiative Fuels Development Program to design, construct, and utilize a thermal hydraulic experimental test facility. The primary objective is to produce a database of information to support the qualification of the new prototypic Uranium-Molybdenum, low enrichment fuel forms to be utilized in high performance research reactors to allow conversion from high enrichment fuels currently in use. This data will also be used to validate computational tools used to model fluid-structure interactions. This database of information is to include fuel plate and element plastic and elastic deformation and vibration as a function of operating system pressure, temperature, and flow rate.
The HMFTF was designed to cover the operating envelope of all high performance research reactors in the US while operating under subcooled conditions. The primary loop is rated to 600 psig and 460°F and has the capability to operate with net flow rates ranging from 100 gpm to 1600 gpm. Operators are able to maintain conditions within ± 5 °F, psig, and gpm during testing. In order to recreate the thermal-hydraulic conditions in reactors, the loop can be configured for up or down-flow through the test section
Plate vibration and deformation is measured through the use of accelerometers and strain gages strategically placed on test elements which are connected to a National Instruments PXIe chassis for data acquisition. Pitot tube assemblies are used to measure the static and total pressure within each subchannel of test elements to allow for characterization of flow bias within assembles under test. This system allows for data collection at rates up to 5 kHz for short periods of time over all connected instruments to allow for characterization of the frequency of test element vibrations.
The Laser-Imaged Natural Circulation (LINC) facility is an experimental apparatus designed for the observation of natural circulation phenomena associated with vertical, heated cylinders. Continuous operation of natural circulation flow is made possible by a cooling plate that makes up the top of an acrylic tank, designed to allow heater rods to pass through in configurations that include two 3/8” rods with variable pitch or one rod with variable diameter. It leverages Particle Image Velocimetry (PIV) as well as an array of thermocouples to measure heat transfer and time-resolved fluid velocity. From these measurements, deductions can be made about boundary layer thickness, heat transfer rate, boundary layer transition location, time-resolved velocity profiles and more.
The purpose of the LINC facility is to provide valuable data in support of furthering understanding natural convection in this geometry. It is of particular interest to the nuclear engineering field as the majority of nuclear fuel in use in the world consists of long, slender cylinders which are immersed in water and cooled via natural convection for much of its life. The data acquired from the LINC facility can be applied directly to safety and performance assessment models used to maintain nuclear fuel at safe temperatures at all times.
The LINC facility allows for investigations using two rod configurations with variable pitch as well as one-rod configurations with variable rod diameter. The maximum power that can be supplied to each rod (and removed by the cooling plate) is 1,400 W, powered by a digital power supply with feedback of actual power used. Velocity field data can be collected using the PIV system at frequencies of 5,000 Hz. Temperature is monitored at many locations including inside the tank, internal to the heater rod(s) and at the inlet and outlet of the cooling plate. The chiller that circulates coolant through the cooling plate has a rating of 1,700 W at 20°C. The tank operates at atmospheric pressure.
The Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL) contains six pressurized water loops used for in-pile testing. Each loop contains its own in-pile tube, pump pressurizer, purification and heat exchanger for controlling pressure temperature and chemistry conditions. For four of the loops, pumps have been running over more than 45 years exceeding by far their original service and design life. To ensure continuity in the research program conducted at the ATR, the need to replace the pumps of the existing loop has been anticipated. In this context, the NRTL facility has been designed and built at Oregon State University. Its objectives are to test the reliability of the new designed pumps by operating them in series in a wide range of conditions and extensively monitoring all parameters associated with their operation (bearing and motor temperature, power input, vibration, head….). The NRTL facility has been operated since fall 2017 and is designed for pressure up to 2500 psi at 650F.
The Transient Reactor Test Loop facility is designed to perform operational and benchmarking experiments that investigate the properties of materials during transient power conditions. The primary objective of this facility is the collection of and benchmarking against new, NQA-1 compliant experimental thermal hydraulic data of a representative TREAT Facility water flow loop. The outcome of the initial test matrix will yield a documented water flow loop design and demonstration that is representative of a prototypic configuration for the TREAT Facility to provide operational information and benchmarking data; and a fully benchmarked thermal hydraulic model of the water flow loop that may be utilized for future TREAT Facility water flow loop safety analyses.
The TRTL facility is a first-of-its-kind experimental apparatus that operates within the same parameter space as the TREAT facility with regard to system pressure, mass flow, temperature, and power capabilities, resulting in the performance of like-for-like experiments without the radiation effects. A custom-built heater rod (300 mm in length, 9.525 mm OD) is designed to deliver ~500 kW of power to the system on an order-of-milliseconds timeframe. Flexible design of the controls system package allows for variable pulse length and maximum power inputs, providing for a wide range of tests to be performed.
The Traveling Wave Reactor Loop (TWRL) is a large-scale separate-effects test facility which was originally designed under the sponsorship of TerraPower, LLC and has been utilized for a verity of tests focused on improving the understanding of flow-induced vibration phenomena associated with wire-wrapped pins within a sodium fast reactor bundle. This water-filled loop boasts at 14 foot tall test section and is capable of circulating water through the test section over a range of 20 to 350 gallons per minute while maintaining pressures and temperatures of 0 to 200 psig and 70-150 F. The TWRL was utilized in the demonstration of a new measurement method developed by OSU in collaboration with TerraPower, LLC to characterize vibration of pins in real-time through use of fiber-optic sensors using a technique referred to as distributed strain sensing. This outcome of this project led to the first-ever successful measurement of pin-to-pin interaction and contact in real time under fluid-elastic conditions.