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All sources NASA NTRS DOE OSTI

DOE OSTI2017

High-Pressure, High-Temperature Thermal Hydraulic Test Facility for Nuclear-Renewable Hybrid Energy System Studies; Facility Design Description and Status Report

A high-pressure, high-temperature water flow loop has been designed at Idaho National Laboratory (INL) for deployment in the INL Dynamic Energy Transport and Integration Laboratory (DETAIL) within the Energy Systems Laboratory (ESL) D100 northwest high bay. This loop is the first of three thermally coupled flow loops that will comprise the Advanced Reactor Technology Integral System Test (ARTIST) facility. The ARTIST facility will ultimately include a high temperature helium loop and a liquid salt loop, in addition to the PWR loop. The high-pressure water loop will operate at Pressurized Water Reactor (PWR) conditions. It will support experimental research on nuclear-renewable hybrid energy systems as well as advanced reactor technology topics. Within DETAIL, the water flow loop will also serve as a thermal energy source, emulating a reactor system. The loop will be thermally integrated with co-located energy systems including a thermal energy transport loop, thermal energy storage system, and a 25 kW high-temperature electrolysis system for hydrogen production. DETAIL will include additional energy systems such as an integrated microgrid and renewable energy systems including solar photovoltaics. The electrically heated flow loop will be dynamically controlled to simulate nuclear fuel behavior under normal and off-normal operating conditions. DETAIL will be designed for characterization of complex system dynamics. The various components will be interfaced with Digital Real-Time Simulators for supervisory control and dynamic simulation of complete systems. The design and assembly of the high-pressure water flow loop and supporting systems has been subdivided into two Phases. Phase 1, which will be completed by the end of FY17, includes installation of cable trays and supports, electrical conduit, DI water system, drain lift station and the concrete pad for the chiller located outside of the laboratory. Phase 2 includes the detailed design and final assembly of the flow loop, supporting structures and personnel platforms, the chiller and associated piping. As of September, 2017, the final detailed design of the flow loop is complete. Funding for Phase 2 procurement, assembly, system operability checkout, shakedown testing and initial operation of the flow loop is requested for FY18.

42 ENGINEERING↗

DOE OSTI2023

HIGH-LOW FIDELITY THERMAL HYDRAULIC COUPLING USING AI/MACHINE LEARNING ALGORITHMS

The primary goal of the US Department of Energy (DOE) office of Nuclear Energy Integrated Energy Systems (IES) program is to develop the tools and framework for coupling multi-scale and multi-physical thermal and electrical energy usage and storage systems. High- and low-fidelity (high–low) coupling is a key feature of multi-scale, multi-component systems and has been an important focus of research in the nuclear energy community for the past two decades. An essential feature of demonstrating the capability to couple high-fidelity and low-fidelity systems for real-time applications are surrogate/reduced order models (ROM). For the purposes of this study, surrogate models are essentially Blackbox models, typically developed using supervised Machine learning (ML) algorithms. The surrogate models can be used to mimic the response of high-fidelity models to represent large historical datasets and coupled with more general low-fidelity system models distributed as Functional Mock-up Interface (FMI) or Functional Mock-up Units (FMU) modules. The example is demonstrated with Spallation Neutron Source (SNS) First Target Station flow loop data. The flow loop is a liquid mercury loop with a pump, piping, heat exchange, and internal heat generation in the target window. This work elucidates some of the potential benefits and future needs of developing tools for high–low system coupling of energy systems.

Williams, Wesley↗

DOE OSTI2015

In-situ Condition Monitoring of Components in Small Modular Reactors Using Process and Electrical Signature Analysis. Final report, volume 1. Development of experimental flow control loop, data analysis and plant monitoring

The research and development under this project was focused on the following three major objectives: Objective 1: Identification of critical in-vessel SMR components for remote monitoring and development of their low-order dynamic models, along with a simulation model of an integral pressurized water reactor (iPWR). Objective 2: Development of an experimental flow control loop with motor-driven valves and pumps, incorporating data acquisition and on-line monitoring interface. Objective 3: Development of stationary and transient signal processing methods for electrical signatures, machinery vibration, and for characterizing process variables for equipment monitoring. This objective includes the development of a data analysis toolbox. The following is a summary of the technical accomplishments under this project: - A detailed literature review of various SMR types and electrical signature analysis of motor-driven systems was completed. A bibliography of literature is provided at the end of this report. Assistance was provided by ORNL in identifying some key references. - A review of literature on pump-motor modeling and digital signal processing methods was performed. - An existing flow control loop was upgraded with new instrumentation, data acquisition hardware and software. The upgrading of the experimental loop included the installation of a new submersible pump driven by a three-phase induction motor. All the sensors were calibrated before full-scale experimental runs were performed. - MATLAB-Simulink model of a three-phase induction motor and pump system was completed. The model was used to simulate normal operation and fault conditions in the motor-pump system, and to identify changes in the electrical signatures. - A simulation model of an integral PWR (iPWR) was updated and the MATLAB-Simulink model was validated for known transients. The pump-motor model was interfaced with the iPWR model for testing the impact of primary flow perturbations (upsets) on plant parameters and the pump electrical signatures. Additionally, the reactor simulation is being used to generate normal operation data and data with instrumentation faults and process anomalies. A frequency controller was interfaced with the motor power supply in order to vary the electrical supply frequency. The experimental flow control loop was used to generate operational data under varying motor performance characteristics. Coolant leakage events were simulated by varying the bypass loop flow rate. The accuracy of motor power calculation was improved by incorporating the power factor, computed from motor current and voltage in each phase of the induction motor.- A variety of experimental runs were made for steady-state and transient pump operating conditions. Process, vibration, and electrical signatures were measured using a submersible pump with variable supply frequency. High correlation was seen between motor current and pump discharge pressure signal; similar high correlation was exhibited between pump motor power and flow rate. Wide-band analysis indicated high coherence (in the frequency domain) between motor current and vibration signals. - Wide-band operational data from a PWR were acquired from AMS Corporation and used to develop time-series models, and to estimate signal spectrum and sensor time constant. All the data were from different pressure transmitters in the system, including primary and secondary loops. These signals were pre-processed using the wavelet transform for filtering both low-frequency and high-frequency bands. This technique of signal pre-processing provides minimum distortion of the data, and results in a more optimal estimation of time constants of plant sensors using time-series modeling techniques.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Off-Normal Scenarios Using Resistively Heated Targets

This report describes the continuation of thermal experiments using the resistively heated target assembly within the helium flow loop at LANSCE that is described in the report, Resistively Heated Target Tests Using Helium Flow Loop, by Alexander Wass et al. In order to replicate off-normal operating scenarios of the helium cooled Mo-100 plant design disks during beam heating, an array of resistive heaters were placed within the helium flow loop, cooled via helium, and modified slightly. The resistively heated test piece is comprised of seven ceramic electric heaters with embedded thermocouples allowing temperature measurements of each heater. A total of three scenarios were performed. One unmodified scenario, and two off-normal scenarios were conducted to simulate a blocked cooling channel due to debris within the flow loop and a broken disk during in-beam heating at the Mo-99 production facility. A removable insert was placed within a cooling channel to block the flow of helium, and a resistive heater was removed to simulate a broken disk that had been evacuated from the target assembly. The inlet and outlet cooling pipe employs rectangular tubing to make a 90° bend at a radius consistent with and practical for the actual plant design.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Developing Separate Effects Transient Test Experiments Using an Out-of-Pile Flowing Water Loop

In 2017 the Transient Reactor Test Facility was restarted after being placed in a standby state since 1994. This Test Reactor’s restart has since enabled the progressive development of new nuclear technologies within the U.S. that were previously required to outsource to other countries. While the reactor’s restart was a large feat worthy of recognition, the experimental use of its characteristics has required the further development of in-pile experimental infrastructure sufficient to support programmatic needs. This hardware have taken the form of capsule designs (compact and elongated) as well as loop concepts representing the phenomena of interest for a subset of separate effects tests desire for each respective testing campaign. The Transient Testing Program has a large-integrated effort that aligns with the U.S. Department of Energy’s current needs. This study compliments those programmatic elements by developing, fabricating and demonstrating a full-scale flowing water loop in an out-of-pile environment. The goal of this effort is to develop a pragmatic understanding of the engineering capabilities and limitations associated with geometric form-factors, metering technology, and controls logic under representative thermal hydraulic conditions that would be experienced within the TREAT reactor during an in-pile reactivity initiated accident test. The outcomes of this study result in an evaluation of the conceptual design of a comprehensive flowing water loop including objective figures of merit (FoMs) for comparing unique instrumentation and the basis for their selection during operations. These efforts directly contribute to and are required for the further advancement of transient testing capabilities within the U.S.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Application of MP-PIC on Dual Interconnected Chemical Looping Cold Flow System: Validation Process with Hydrodynamic Experiments

Chemical looping with oxygen uncoupling (CLOU) is a promising carbon capture technologies, that utilizes a dual fluidized bed interconnected by loopseals to selectively separate oxygen from ambient air. This study validated Barracuda-VR® hydrodynamic simulations of dual circulating fluidized bed. The experiments were run on a scaled-down cold-flow unit of a 100 kW dual reactor to find the quantities of interest: circulation rate, bed mass location distribution, and pressure profile. The numerical parameters of time step and grid resolution were studied to find the numeric uncertainty for the quantities of interest. The scenario parameters of flowrates, particle densities, and particle voidage were studied. The uncertainty from the model parameters of tangent-to-wall, drag multiplier, and stress model were studied. Simulation and experimental results matched over some of the parameters studied. These can be used to scale the system back to the 100 kW dual reactor and to a larger system.

Hamilton, Matthew A↗

DOE OSTI2015

Neutron Imaging of a Two-Phase Refrigerant Flow

Void fraction remains a crucial parameter in understanding and characterizing two-phase flow. It appears as a key variable in both heat transfer and pressure drop correlations of two-phase flows, from the macro to micro- channel scale. Void fraction estimation dictates the sizing of both evaporating and condensing phase change heat exchangers, for example. In order to measure void fraction some invasive approach is necessary. Typically, visualization is achieved either downstream of the test section or on top by machining to expose the channel. Both approaches can lead to inaccuracies. The former assumes the flow will not be affected moving from the heat exchanger surface to the transparent section. The latter distorts the heat flow path. Neutron Imaging can provide a non-invasive measurement because metals such as Aluminum are essentially transparent to neutrons. Hence, if a refrigerant is selected that provides suitable neutron attenuation; steady-state void fraction measurements in two-phase flow are attainable in-situ without disturbing the fluid flow or heat flow path. Neutron Imaging has been used in the past to qualitatively describe the flow in heat exchangers in terms of maldistributions without providing void fraction data. This work is distinguished from previous efforts because the heat exchanger has been designed and the refrigerant selected to avail of neutron imaging. This work describes the experimental flow loop that enables a boiling two-phase flow; the heat exchanger test section and downstream transparent section are described. The flow loop controls the degree of subcooling and the refrigerant flowrate. Heating cartridges embedded in the test section are employed to control the heat input. Neutron-imaged steady-state void fraction measurements are captured and compared to representative high-speed videography captured at the visualization section. This allows a qualitative comparison between neutron imaged and traditional techniques. The measurements are also compared to correlations in the literature. Preliminary void fraction images from a macro-channel flow are presented, consisting of 1 channel, 4mm wide, 4mm high and 83.32mm long. Flow regime identification is examined. The experiments were conducted at the High Flux Isotope Reactor (HFIR) Cold Guide 1D neutron imaging facility at Oak Ridge National Laboratory, Oak Ridge, TN, USA.

Geoghegan, Patrick J↗

DOE OSTI2017

Heating and cooling system for an on-board gas adsorbent storage vessel

In one aspect, a system for controlling the temperature within a gas adsorbent storage vessel of a vehicle may include an air conditioning system forming a continuous flow loop of heat exchange fluid that is cycled between a heated flow and a cooled flow. The system may also include at least one fluid by-pass line extending at least partially within the gas adsorbent storage vessel. The fluid by-pass line(s) may be configured to receive a by-pass flow including at least a portion of the heated flow or the cooled flow of the heat exchange fluid at one or more input locations and expel the by-pass flow back into the continuous flow loop at one or more output locations, wherein the by-pass flow is directed through the gas adsorbent storage vessel via the by-pass line(s) so as to adjust an internal temperature within the gas adsorbent storage vessel.

42 ENGINEERING↗

DOE OSTI2019

Heating and cooling system for an on-board gas adsorbent storage vessel

42 ENGINEERING↗

DOE OSTI2016

Advanced Test Reactor Mini-Plate Hydraulic Testing

Currently, flow testing is being performed at Oregon State University (OSU) on the hydromechanical Fuel Test Facility (HMFTF) and the Endurance Flow Loop (EFL) to aid in the qualification of low enriched plate-type Uranium fuels for use in high performance research reactors in the US, under the Global Threat Reduction Initiative (GTRI). Under this project, varying geometries of small capsules proposed for irradiation testing in the Advanced Test Reactor (ATR) are being hydraulically testing prior to being placed in the ATR. The premise of this study was to conduct a number of flow tests on these capsules to characterize the pressure loss through each as a function of flow rate and compare the results against predicted theoretical values. Two categories of flow tests were performed in the Endurance Flow Loop to supplement flow testing of proposed LEU plate-type fuel for conversion of High Performance Research Reactors: Capsule Comparison Tests and Orifice Characterization Tests. The purpose of these tests were to determine the pressure drop across each experiment as a function of flow rate and to compare the results against a simple analytical model of the pressure loss through a pipe, to support the safety basis for the in-pile experiments. The capsule comparison tests compared the pressure loss through three types of capsules: RERTR, MP-1, and EMPIRE. In this category, nine flow tests were performed to determine the effect of having one, two, and three of each type of capsule placed in series in a test section. The analytical model accurately predicted, for each case, an increase in hydraulic resistance for each capsule that was added, though the values were conservative. The orifice characterization tests compared the pressure loss through various configurations of small orifice spacers in conjunction with blank spacers and the MP-1 capsules. The analytical model predicted that adding blank spacers would slightly increase the pressure drop due to added frictional effects, but indicated no change in the pressure drop based on the orientation of the capsules. The measured results indicated that the addition of the blank spacers had no measurable effect on the pressure drop, while changing the orientation of the capsules did. In all 15 flow tests, the theoretical model provides a conservative estimate of the pressure drop through each configuration for an equivalent flow rate, over that of the measured results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

Validation of Pronghorn for Natural-Circulation Molten Salt Loops

This paper presents the development and validation of a high-fidelity thermal-hydraulic model of a molten salt natural circulation flow loop, designed for integration within a digital twin framework. The study evaluates the performance of Idaho National Laboratory’s Pronghorn against experimental data from Texas A&M University Molten Salt Flow Loop (MSFL) four Hitec-salt test benchmark data. Natural circulation of high-Prandtl-number fluids exhibits complex, counter-intuitive flow patterns that make pointwise thermocouple readings unreliable. Experimental work at TAMU’s MSFL provides benchmark data, including flow visualization at a test-section and centerline steady-state temperature measurements along the loop. Validation includes four single-phase natural circulation test cases with Hitec salt. Key metrics include flow profile agreement and steady-state temperature accuracy. Pronghorn results for two-dimensional single-phase agree qualitatively with the experimental flow profile. This paper illustrates the importance of Computational Fluid Dynamics (CFD) in elucidating the behavior of high-Prandtl-number thermal-hydraulics, along with how misleading centerline temperature measurements can be. Pronghorn reproduces the axial and radial stratification that makes single thermocouple readings unreliable. Future research will focus on reduced-order modeling techniques to enable rapid simulation suitable for real-time digital twin applications. The validated cases provide a basis for developing reduced-order surrogates aimed at real-time digital-twin applications.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

ROBUST ONLINE MONITORING FOR CALIBRATION ASSESSMENT OF TRANSMITTERS AND INSTRUMENTATION

Robust online monitoring (OLM) technologies are expected to enable the extension or elimination of periodic sensor calibration intervals in operating and new reactors. Specifically, the next generation of OLM technology is expected to include newly developed advanced algorithms that improve monitoring of sensor/system performance and enable the use of plant data to derive information that currently cannot be measured. These advances in OLM technologies will improve the safety and reliability of current and planned nuclear power systems through improved accuracy and increased reliability of sensors used to monitor key parameters. In this paper, we discuss an overview of research being performed within the Nuclear Energy Enabling Technologies (NEET)/Advanced Sensors and Instrumentation (ASI) program, for the development of OLM algorithms to use sensor outputs and, in combination with other available information, 1) determine whether one or more sensors are out of calibration or failing and 2) replace a failing sensor with reliable, accurate sensor outputs. Algorithm development is focused on the following OLM functions: • Signal validation – fault detection and selection of acceptance criteria • Virtual sensing – signal value prediction and acceptance criteria • Response-time assessment – fault detection and acceptance criteria selection A GP-based uncertainty quantification (UQ) method previously developed for UQ in OLM, was adapted for use in sensor-fault detection and virtual sensing. For signal validation, the various components to the OLM residual (which is computed using an AAKR model) were explicitly defined and modeled using a GP. Evaluation was conducted using flow loop data from multiple sources. Results using experimental data from laboratory-scale flow loops indicate that the approach, while capable of detecting sensor drift, may be incapable of discriminating between sensor drift and model inadequacy. This may be due to a simplification applied in the initial modeling, where the sensor degradation is assumed to be stationary. In the case of virtual sensors, the GP model was used in a predictive mode to estimate the correct sensor reading for sensors that may have failed. Results have indicated the viability of using this approach for virtual sensing. However, the GP model has proven to be computationally expensive, and so alternative algorithms for virtual sensing are being evaluated. Finally, automated approaches to performing noise analysis for extracting sensor response time were developed. Evaluation of this technique using laboratory-scale data indicates that it compares well with manual techniques previously used for noise analysis. Moreover, the automated and manual approaches for noise analysis also compare well with the current “gold standard”, hydraulic ramp testing, for response time monitoring. Ongoing research in this project is focused on further evaluation of the algorithms, optimization for accuracy and computational efficiency, and integration into a suite of tools for robust OLM that are applicable to monitoring sensor calibration state in nuclear power plants.

calibration interval extension↗

DOE OSTI2017

Modeling the Transient Reactor Test Loop in TRACE for PWR Critical Heat Flux Benchmarking

Understanding transient critical heat flux (CHF) resulting from a Reactivity Insertion Accident (RIA) in nuclear systems is an extremely undeveloped field. Though many experiments have been done for CHF, remarkably few have been done for rods at PWR operating temperature and pressure (300 deg. C, 2200 psia) with a short time constant ramp in power. This leads to many concerns regarding understanding how RIA scenarios develop and evolve in commercial nuclear reactors. With a limited understanding of the phenomenology governing transient CHF, there is little hope in being able to accurately model these scenarios using today's thermal hydraulic codes. The Accident Tolerant Fuels (ATF) Program established by the Department of Energy (DOE) is seeking out new fuel and cladding systems that can be used to reduce the probability and consequences of severe accidents. To better understand this phenomenon, the DOE is restarting the Transient Reactor Test (TREAT) Facility at the Idaho National Laboratory (INL). TREAT will be able to provide representative neutronic boundary conditions for testing coupled thermal hydraulics on the Transient Water Environment Recirculating Loop (TWERL) to study transient CHF with the best representation possible. To support this task, Oregon State University (OSU) is designing and building a representative flow loop that will be able to mimic the conditions of TWERL (sans the neutronic boundary conditions) to create a benchmark study for existing lumped parameter thermal-hydraulic codes such as RELAP5 and TRACE. This flow loop, the Transient Reactor Test Facility (TRTL), will support the ATF program by providing much needed NQA-1 data for transient CHF at operating temperature and pressure.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2019

Computational and Experimental Benchmarking for Transient Fuel Testing

In 2018 the Transient Reactor Test (TREAT) Facility was restarted. This now-operational facility brings back world-class transient fuel testing capabilities to the U.S. Among a number of parallel efforts prior to the TREAT Facility’s restart a Department of Energy (DoE) Integrated Research Project (IRP) was funded for the purpose of benchmarking the TREAT with several goals in mind: To produce a comprehensive evaluation of existing TREAT Facility neutronics data using the next generation reactor core neutronics codes in accordance with established guidelines per the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). To perform a benchmark using ready-use computational fluid dynamics tools against that of a historically collected sodium loop experiment collected in the TREAT. To design, develop and test a new experimental thermal hydraulic system of a representative TREAT Facility water flow loop. And benchmark the data collected against that produced via RELAP5-3D and TRACE. To develop a comprehensive instrumentation plan for the TREAT Facility by benchmarking performing of advanced in-pile instruments in the Massachusetts Institute of Technology Research Reactor (MITR) and the TREAT. The integrated team brought together collaborating members from three universities including Oregon State University (OSU), the University of Michigan (UM), and Massachusetts Institute of Technology (MIT); three national laboratories, Idaho National Laboratory, Argonne National Laboratory (ANL), and Oak Ridge National Laboratory (ORNL); and two industry partners, TerraPower (TP) and Harris Thermal Transfer Products (HTTP). A brief summary of accomplishments resulting from this IRP includes two filed patents, more than 40 conference papers, two published benchmark reports through IRPhEP, a new representative thermal hydraulic test facility alike the in-pile PWR water flow loop for TREAT and the first advanced instruments tested within the TREAT core upon its restart. The papers presented herein were developed and provide a summary of the significant technical accomplishments which align with the four project’s objectives. The collaborators has significant support from countless parties including the operational and technical staff at TREAT to ensure that their outcomes most impactfully aligned with program needs. Lastly the collaborators would like to acknowledge the fiscal support of the DoE’s Office of Nuclear Energy under work package identification number NU-15-OR-OSU- 0701-01. This report contains selected papers from the Computational and Experimental Benchmarking for Transient Fuel Testing Integrated Research Project.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2022

Comparison of animal and human blood for in vitro dynamic thrombogenicity testing of biomaterials

Abstract Background To determine suitable alternatives to human blood for in vitro dynamic thrombogenicity testing of biomaterials, four different animal blood sources (ovine, bovine, and porcine blood from live donors, and abattoir porcine blood) were compared to fresh human blood. Methods To account for blood coagulability differences between individual donors and species, each blood pool was heparinized to a donor‐specific concentration immediately before testing in a dynamic flow loop system. The target heparin level was established using a static thrombosis pre‐test. For dynamic testing, whole blood was recirculated at room temperature for 1 h at 200 ml/min through a flow loop containing a single test material. Four materials with varying thrombotic potentials were investigated: latex (positive control), polytetrafluoroethylene (PTFE) (negative control), silicone (intermediate thrombotic potential), and high‐density polyethylene (HDPE) (historically thromboresistant). Thrombus weight and surface area coverage on the test materials were quantified, along with platelet count reduction in the blood. Results While donor‐specific heparin levels varied substantially from 0.6 U/ml to 7.0 U/ml among the different blood sources, each source was able to differentiate between the thrombogenic latex and the thromboresistant PTFE and HDPE materials ( p < 0.05). However, only donor ovine and bovine blood were sensitive enough to differentiate an increased response for the intermediate thrombotic silicone material compared to PTFE and HDPE. Conclusions These results demonstrated that multiple animal blood sources (particularly donor ovine and bovine blood) may be suitable alternatives to fresh human blood for dynamic thrombogenicity testing when appropriate control materials and donor‐specific anticoagulation levels are used.

Engineering↗

DOE OSTI2013

High-Temperature-High-Volume Lifting for Enhanced Geothermal Systems

This report summarizes the progress made during the April 01, 2010 – December 30, 2013 period under Cooperative Agreement DE-EE0002752 for the U.S. Department of Energy entitled “High-Temperature-High-Volume Lifting for Enhanced Geothermal Systems.” The overall objective of this program is to advance the technology for well fluids lifting systems to meet the foreseeable pressure, temperature, and longevity needs of the Enhanced Geothermal Systems (EGS) industry for the coming ten years. In this program, lifting system requirements for EGS wells were established via consultation with industry experts and site visits. A number of artificial lift technologies were evaluated with regard to their applicability to EGS applications; it was determined that a system based on electric submersible pump (ESP) technology was best suited to EGS. Technical barriers were identified and a component-level technology development program was undertaken to address each barrier, with the most challenging being the development of a power-dense, small diameter motor that can operate reliably in a 300°C environment for up to three years. Some of the targeted individual component technologies include permanent magnet motor construction, high-temperature insulation, dielectrics, bearings, seals, thrust washers, and pump impellers/diffusers. Advances were also made in thermal management of electric motors. In addition to the overall system design for a full-scale EGS application, a subscale prototype was designed and fabricated. Like the full-scale design, the subscale prototype features a novel “flow-through-the-bore” permanent magnet electric motor that combines the use of high temperature materials with an internal cooling scheme that limits peak internal temperatures to <330°C. While the full-scale high-volume multi-stage pump is designed to lift up to 80 kg/s of process water, the subscale prototype is based on a production design that can pump 20 kg/s and has been modified for high-temperature operation. In parallel with the design and fabrication of the subscale prototype ESP system, a subscale test facility consisting of a high-temperature-high-pressure flow loop was designed, fabricated, and installed at GE Global Research in Niskayuna, NY. A test plan for the prototype system was also established. The original plan of testing the prototype hardware in the flow loop was delayed until a future date.

15 GEOTHERMAL ENERGY↗

DOE OSTI2012

Continued Evaluation of the Pulse-Echo Ultrasonic Instrument for Critical Velocity Determination during Hanford Tank Waste Transfer Operations

Laboratory (PNNL) conducted an extensive evaluation of the ability of three ultrasonic instruments to detect critical velocity for a broad range of simulated Hanford nuclear waste streams containing particles with mean particle sizes of >50 microns. Evaluations were perform using the pipe loop at the Process Development Laboratory – East (PDL-E) at PNNL that was designed and built to evaluate the pipeline plugging issue during slurry transfer operations at the Hanford Waste Treatment Plant. In 2011 the ability of the ultrasonic PulseEcho system to detect critical velocity continued to be evaluated using the PDL-E flow loop and new simulants containing high-density particles with a mean particle size of < 15 microns. The PDL-E flow loop was modified for the 2011 testing to include a new test section that contained 5-MHz and 10-MHz ultrasonic transducers non-invasively mounted to schedule 40 pipe. The test section also contained reference instrumentation to facilitate direct comparison of the real-time PulseEcho transducer responses with experimentally observed critical velocities. This paper presents the results from the 2011 PulseEcho evaluation using a variety of simulated Hanford nuclear waste streams that were selected to encompass the expected high-level waste feed properties.

Critical Velocity↗

DOE OSTI2020

Deep Learning for Intelligent Bubble Size Detection in the Spallation Neutron Source Visual Target

The Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) will undergo proton power upgrade (PPU), increasing the proton beam power from 1.4 MW to 2.8 MW. From 2.8 MW, 2.0 MW will go to the current First Target Station and the rest will go to the future Second Target Station (STS). The First Target Station uses a liquid mercury target that is contained in a 316L stainless steel vessel. The proton beam is pulsed at 60 Hz, with a pulse of about 0.7μs. When the proton beam hits the target, the intense energy deposition leads to a rapid rise in temperature in the mercury. This temperature rise creates pressure waves that propagate through the mercury and cause cavitation erosion. The power upgrade will cause stronger pressure waves that will further increase damage because of cavitation. Injecting small helium bubbles in the mercury has been an efficient method of mitigating the pressure wave at 1.4 MW. However, at higher power, additional mitigation is necessary. Therefore, the 2 MW target vessel will be equipped with swirl bubblers and an additional gas injection port near the nose to inject more gas in the target. To develop a gas injection strategy and design, flow visualization in water with a transparent prototypical target (“visual target”) was performed. Bubble sizes and their spatial distribution in the flow loop are crucial to understanding the effectiveness of the bubbles in mitigating pressure waves. Bubbles were generated in the visual target under varied conditions of input pressures with helium and air. Images were captured using a high-speed camera at varied frame rates at different positions away from the swirl bubbler and different depths in the flow loop under varying lighting conditions. Initially, methods such as circular Hough transforms were applied after a series of images processing to obtain a general distribution of bubble sizes. Bubbles smaller than 500 μm are preferred to effectively mitigate the effect of pressure waves, which demands an accurate bubble detection and sizing system. Intelligent detection and identification of bubble sizes alleviate misdetection and improves accuracies. Employing neural networks, intelligent detection of bubble sizes and their distribution was developed and provides a robust alternative to traditional techniques. Human intervention was employed to label in-focus and out-of-focus bubbles in the set of training images. An object detection network using a pretrained convolutional neural network was created that extracted the features from the training images. Data augmentation was used to improve network accuracy through a random transformation of the original data.

Rasheed, Fayaz↗