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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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DOE OSTI2022

Testing, Characterization, and Modeling of the Resonant Plate Test Environment

The resonant plate shock test is a dynamic test of a mid-field pyroshock environment where a projectile is struck against a plate. The structure undergoing the simulated field shock is mounted to the plate. The plate resonates when struck and provides a two sided shock that is representative of the shock observed in the field. This test environment shock simulates a shock in a single coordinate direction for components looking to provide evidence that they will survive a similar or less shock when deployed in their operating environment. However, testing in one axis at a time provides many challenges. The true environment is a multi-axis environment. The test environment exhibits strong off-axis motion when only motion in one axis is desired. Multiple fixtures are needed for a single test series. It would be advantageous if a single test could be developed that tests the multi-axis environment simultaneously. In order to design such a test, a model must be developed and validated. The model can be iterated in design and configuration until the specified multi-axis environment is met. The test can then execute the model driven test design. This report discusses the resonant plate model needed to design future tests and the steps and methods used to obtain the model. This report also details aspects of the resonant plate test discovered during the process of model development that aids in our understanding of the test.

42 ENGINEERING↗
DOE OSTI2021

Fabrication and Testing of Two Passively Actuated Creep-Fatigue Surveillance Test Articles

This report describes the fabrication, testing, and analysis of two families of passively-actuated creep-fatigue test articles. These test articles induce cyclic mechanical load in a test section actuated solely through a mismatch in the thermal expansion coefficient of two materials and driven by a change in the specimen temperature. Test articles of this type could provide material degradation data for a materials surveillance program in future operating molten salt reactors, where variations in the reactor temperature caused by normal or upset cycles would provide the temperature change in the sample. The purpose of the current experimental campaign is to demonstrate the feasibility of fabricating realistically-sized test articles, test the articles under repeated thermal cycling to assess their robustness, particularly the reliability of the bimetallic welds, and collect strain data from instrumented samples to validate the methods used to size the test articles to match key features of the mechanical response of operating reactor components. The thermal cycling tests are ongoing but the test data to date suggests the current test article design is viable, robust, and can be successfully designed to match a target mechanical response. Finally, the report describes preliminary work on welding refractory alloys to 316H stainless steel. Replacing the test article driver material, currently Alloy 617, with a refractory would increase the coefficient of thermal expansion mismatch and support further miniaturization of the test articles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Submersible Multistage Centrifugal Pump for Versatile Test Reactor Cartridge Test Loop

Submersible multistage centrifugal pumps are ideal for pumping in narrow confined spaces and achieving necessary head pressures and flow rates. Once a diameter is determined then manipulation of the number of stages and motor speed are all that are required to meet desired flow conditions. The Versatile Test Reactor (VTR) closed loop cartridge systems will need forced convection cooling independent of the main reactor. A multistage centrifugal pump can meet the necessary flow rates and pumping pressures while minimizing space taken. The pump considered for this work was based off a deep well submersible pump, a variation of a multistage centrifugal pump. We experimented with two pump sizes, 5 cm (2 inch) and 7.5 cm (3 inch) diameters. These diameters were chosen to fit into the inner diameter of standard 5 and 7.5 (2 and 3 inch) Schedule 40 pipe, respectively. This made the design for the test loop both simpler and less expensive as the need for an engineered pump housing was eliminated. Initial test cartridge planning indicated space for only a 5 cm (2 inch) diameter pump, though early testing of this size showed the need for an abnormally high-speed and high-power motor. Fine tuning of the cartridge design allowed a pump size increase to 7.5 cm (3 inches), which was the pump size most extensively tested in this work. The test loop is composed of various sizes of PVC and aluminum piping components in a loop configuration. The pump is driven by a Pittman 250 W (1/3 horsepower) electric motor with maximum speed of 3,450 RPM. Testing consisted of running the pump at a constant motor speed while varying a control valve to restrict flow through the loop, with differential pressure and flow rate recorded. This was done for one and two stage configurations for the 5 cm (2 inch) diameter impeller design and one, two, and three stage configurations for the 7.5 cm (3 inch diameter) impeller design, respectively. Due to pumping power requirements, two and three stage 7.5 cm (3 inch) diameter impeller testing at higher flowrates lowered the motor speed substantially. In regions where motor speed could not be maintained constant, the data were discarded. The test loop was also reconfigured to allow for the pump to be tested for pressure drop in a stalled or inoperable (0 RPM) flow condition. Demonstration of adequate natural convection cooling of the test cartridge fuel type is necessary under accident conditions, and this will depend upon the flow resistance through the impeller assembly when the pump is not operating. Thus, accurate knowledge of the effective impeller assembly loss coefficient is important for safety evaluations. The test loop was modified to provide water inlet and outlets on either side of the pump impeller stack, and a metered flow of lab water was provided in order to measure the pressure drop across the cartridges as a function of flowrate. Data from the pump head curve testing developed as part of this work and supported by analysis using pump head affinity laws indicates that a three stage 7.5 cm (3 inch) pump impeller design will meet target requirements for coolant flow within the VTR cartridge sodium cartridge at full power conditions [1] with margin; this corresponds to a flowrate of 45 l/min (12 gpm) at a pressure drop of 6.1 m (20 feet) of water head. The results of the pressure loss measurements across the impeller assembly when the pump is stationary (i.e., at 0 RPM) indicate that the pressure loss coefficient is 0.921 for a two impeller stack configuration; this value is calculated based on the flow velocity through the minimum available flow area within a single stage of the impeller which corresponds to 1.4 cm2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program (Rev. 2)

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 and SHRT-6 were protected loss of flow tests where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R, SHRT-39, and SHRT-43R were unprotected loss of flow tests where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. PICT-5/6 demonstrated how core power could be increased and decreased via changes in the intermediate sodium and feedwater flow rates. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for all tests simulated there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. Thirteen additional tests would be performed in the CABRI reactor over the next decade on fuel rods with higher burnups [1]. Additionally, a larger testing program at the NSRR reactor in Japan with high burnup fuel would uncover a similar trend of pre-DNB ruptures in high burnup test rods at lower-than-expected peak enthalpies [2]. Numerous out of pile testing programs involving a variety of innovative mechanical testing techniques have been employed in an attempt to better quantify the failure thresholds of corroded zirconium alloy cladding in these rapid heating and loading conditions [3][4][5]. While previously interim guidance had been issued, in June of 2020 the NRC officially published updated regulatory guidance to account of these pre-DNB failures in Regulatory Guide 1.236 [6]. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Optimal Nested Test Plan for Combinatorial Quantitative Group Testing

We consider the quantitative group testing problem where the objective is to identify defective items in a given population based on results of tests performed on subsets of the population. Under the quantitative group testing model, the result of each test reveals the number of defective items in the tested group. The minimum number of tests achievable by nested test plans was established by Aigner and Schughart in 1985 within a minimax framework. Furthermore, the optimal nested test plan offering this performance, however, was not obtained. In this paper, we establish the optimal nested test plan in a closed form. This optimal nested test plan is also order optimal among all test plans as the population size approaches infinity. Using heavy-hitter detection as a case study, we show via simulation examples orders of magnitude improvement of the group testing approach over two prevailing sampling-based approaches in detection accuracy and counter consumption. Other applications include anomaly detection and wideband spectrum sensing in cognitive radio systems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Test Methodologies for Hydrogen Sensor Performance Assessment: Chamber vs. Flow Through Test Apparatus: Preprint

Certification of hydrogen sensors to standards often prescribes using large-volume test chambers [1, 2]. However, feedback from stakeholders such as sensor manufacturers and end-users indicate that chamber test methods are often viewed as too slow and expensive for routine assessment. Flow through test methods potentially are an efficient, cost-effective alternative for sensor performance assessment. A large number of sensors can be simultaneously tested, in series or in parallel, with an appropriate flow through test fixture. The recent development of sensors with response times of less than 1s mandates improvements in equipment and methodology to properly capture the performance of this new generation of fast sensors; flow methods are a viable approach for accurate response and recovery time determinations, but there are potential drawbacks. According to ISO 26142 [1], flow through test methods may not properly simulate ambient applications. In chamber test methods, gas transport to the sensor can be dominated by diffusion which is viewed by some users as mimicking deployment in rooms and other confined spaces. Alternatively, in flow through methods, forced flow transports the gas to the sensing element. The advective flow dynamics may induce changes in the sensor behaviour relative to the quasi-quiescent condition that may prevail in chamber test methods. One goal of the current activity in the JRC and NREL sensor laboratories [3, 4] is to develop a validated flow through apparatus and methods for hydrogen sensor performance testing. In addition to minimizing the impact on sensor behaviour induced by differences in flow dynamics, challenges associated with flow through methods include the ability to control environmental parameters (humidity, pressure and temperature) during the test and changes in the test gas composition induced by chemical reactions with upstream sensors. Guidelines on flow through test apparatus design and protocols for the evaluation of hydrogen sensor performance are being developed. Various commercial sensor platforms (e.g., thermal conductivity, catalytic and metal semiconductor) were used to demonstrate the advantages and issues with the flow through methodology.

14 SOLAR ENERGY↗
DOE OSTI2025

Test Results of the LQXFA/B02 and LQXFA/B03 Cryo-Assemblies for the High Luminosity LHC Upgrade

Here, the US High-Luminosity LHC Accelerator Upgrade Project (AUP) is responsible for delivering cryo-assemblies for the Q1/Q3 quadrupole optical components of the High Luminosity LHC upgrade at CERN. Total of 10 cryo-assemblies containing two Nb3Sn quadrupole magnets per cold mass will be delivered within this program. After the successful test of the first pre-series cryo-assembly in 2023, two more cryo-assemblies were tested at Fermilab's horizontal test facility. Production overview and the test results of the LQXFA/B02 and LQXFA/B03 cryo-assemblies are summarized in this paper. After the first test, to increase the capability of the horizontal test facility, various improvements have been made. These improvements are also described in this paper.

Chlachidze, Guram [Fermi National Accelerator Labo↗
DOE OSTI2021

Controlled Pyroshock Test Transients Can Be Used To Better Match Operational Shock Transients and Reduce Artificial Pyroshock Test Failures [Slides]

Shock analysis is typically done via SRS (shock response spectrum). Oftentimes shock analysis is not performed because most loads specifications only supply SRS requirements with no representative shock transients that can be used for analysis. SRS analysis is too conservative. There is no structural dynamics involved as all modes are assumed to have peak responses simultaneously and in-phase with each other. It is possible that all other loads and combination of loads show the design to be good, whereas SRS analysis (1000 g’s are involved) results may require redesign. Multiple shock transients can satisfy a SRS, yet most loads documents don’t provide a representative shock transient that could be used for shock analysis. Shock testing is less common than harmonic and random vibration testing. Engineers typically go into shock tests with very little knowledge of what to expect. A pass/fail type of approach with very little engineering done to prepare for the test or to control the type of shock transient is used. Most engineers will not model the shock test fixture due to lack of knowledge about the details of the test setup. Shock analysis and testing is often an open loop type situation with no test setup modeling, no test correlation, and no control over the test shock transient. In this work, if we can create applied test force transient that comes close to matching calculated force transient then a more reasonable acceleration transient will be created at the test reference accelerometer that better matches operational shocks.

42 ENGINEERING↗
DOE OSTI2022

SAS4A/SASSYS-1 Validation with EBR-II Tests Performed During the SHRT Testing Program

The EBR-II sodium-cooled fast reactor supported a wide range of fast reactor research activities. The final fifteen years of operations at EBR-II were used for experiments and tests to demonstrate the importance of passive safety in liquid metal reactors and the capability of such systems to provide strong passive responses during off-normal conditions. The Shutdown Heat Removal Test program at EBR-II provided test data supporting the validation of computer codes for design, licensing, and operation of LMRs, among other objectives. The program included nearly sixty tests, most of which were protected and unprotected loss of flow and loss of heat sink tests run at various initial powers and flow rates. This report documents the results of modeling and simulation of tests from the Shutdown Heat Removal Test program to support validation of Argonne’s SAS4A/SASSYS-1 fast reactor safety analysis code. SHRT-17 was a protected loss of flow test where a loss of electrical power to all sodium coolant pumps was simulated to demonstrate the effectiveness of natural circulation cooling characteristics. SHRT-45R was an unprotected loss of flow test where the control rod scram function of the plant protection system was disabled to demonstrate the effectiveness of EBR-II’s passive reactivity feedbacks. BOP-301 and BOP-302R were unprotected loss of heat sink tests that demonstrated how reactivity feedback effects driven by changes to the core inlet temperature can shut down the fission process. This validation activity leverages modeling and simulation efforts that originated under an International Atomic Energy Agency Coordinated Research Project for benchmark analysis of the EBR-II Shutdown Heat Removal Tests. The SAS4A/SASSYS-1 models originally developed for the CRP have been modernized to utilize new code features and modeling practices. Differences between the predicted and measured test data were explored and the causes of these differences were identified as being due to modeling approximations, uncertainty in transient component behavior, and instrumentation uncertainties. Overall, it was concluded that for SHRT-17, SHRT-45R, BOP-301, and BOP-302R there was good agreement between the flow, power, and temperature predictions and the measured data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Flow Sensor Test Article (F-STAr) - Water Testing and Commissioning Report for FY2024

Argonne National Laboratory’s Mechanism Engineering Test Loop (METL) facility is developing new experimental testing capabilities with the Flow Sensor Test Article (F-STAr). F-STAr’s mission is to provide high flowrate, sodium submersible testing capabilities to support the development of sodium cooled nuclear fission reactors. Figure 1 shows a model of F-STAr (left) and a photo of the partially assembled test article (right) with several of the main components labeled. The system currently includes a large pump with desired flowrates of up to 100 GPM; an immersion heater with a maximum power of 5 kW; an immersion cooler with an estimated maximum cooling power of 2 kW; and an experimental test section to support sub-test articles. Initially, F-STAr will be setup for testing flow sensors, specifically Eddy Current Flow Sensors (ECFS). However, the system can be reconfigured to support a wide range of testing needs. For example, F-STAr could be setup to complete testing of hydraulic components, heat exchangers, hydrodynamic bearings, seals, and more. This report will provide an update on the development of F-STAr with a focus on the water testing and commissioning activities during Fiscal Year 2024 (FY24). First, the report will focus on describing the pump development. Specifically, updates will be provided on troubleshooting the pump’s excessive vibrational issues. In addition to updates on the pump, pressure-flowrate performance testing in water will be described. Furthermore, updates will be provided on a pump shaft bushing development as well as rigging procedures. Future work and paths forward for F-STAr will be described. Finally, a short summary on the ECFS development will be described at the end of the report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

The Uranium from Seawater Program at the Pacific Northwest National Laboratory: Overview of Marine Testing, Adsorbent Characterization, Adsorbent Durability, Adsorbent Toxicity, and Deployment Studies

The Pacific Northwest National Laboratory’s (PNNL) Marine Science Laboratory (MSL) located along the coast of Washington State is evaluating the performance of uranium adsorption materials being developed for seawater extraction under realistic marine conditions with natural seawater. Two types of exposure systems were employed in this program: flow-through columns for testing of fixed beds of individual fibers and pellets and a recirculating water flume for testing of braided adsorbent material. Testing consists of measurements of the adsorption of uranium and other elements from seawater as a function of time, typically 42 to 56 day exposures, to determine the adsorbent capacity and adsorption rate (kinetics). Analysis of uranium and other trace elements collected by the adsorbents was conducted following strong acid digestion of the adsorbent with 50% aqua regia using either Inductively Coupled Plasma Optical Emission Spectrometry (ICP-OES) or Inductively Coupled Plasma Mass Spectrometer (ICP-MS). The ORNL 38H adsorbent had a 56 day adsorption capacity of 3.30 ± 0.68 g U/ kg adsorbent (normalized to a salinity of 35 psu), a saturation adsorption capacity of 4.89 ± 0.83 g U/kg of adsorbent material (normalized to a salinity of 35 psu) and a half-saturation time of 28 ± 10 days. The AF1 adsorbent material had a 56 day adsorption capacity of 3.9 ± 0.2 g U/kg adsorbent material (normalized to a salinity of 35 psu), a saturation capacity of 5.4 ± 0.2 g U/kg adsorbent material (normalized to a salinity of 35 psu) and a half saturation time of 23 ± 2 days. The ORNL amidoxime-based adsorbent materials are not specific for uranium, but also adsorb other elements from seawater. The major doubly charged cations in seawater (Ca and Mg) account for a majority of the cations adsorbed (61% by mass and 74% by molar percent). For the ORNL AF1 adsorbent material, U is the 4th most abundant element adsorbed by mass and 7th most abundant by molar percentage

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Analytic Considerations and Design Basis for the IEEE Distribution Test Feeders

One of the greatest challenges for the analysis of distribution systems is the high level of diversity across the world. For nearly 20 years the Test Feeder Working Group of the Distribution System Analysis Subcommittee has been developing distribution test feeders. The purpose of these test feeders is to provide openly-available models of distribution systems that reflect the diversity in systems and the analytic challenges they present. Because of their utility and accessibility, the test feeders have been used for a wide range of research, some of which they were not originally intended to cover. The purpose of this paper is to provide an overview of the variety of existing distribution systems, and to clarify the specific analytic challenges that each test feeder was originally designed to examine. Additionally, the paper will provide guidance on which feeders are best suited for various types of analysis. The purpose of this paper is to provide the original intent of the Working Group and to inform researchers so that they may make an informed decision on which of the test feeders are most appropriate for their work.

Schneider, Kevin P.↗
DOE OSTI2021

Characterization of Embedded Sensors in Stainless Steel Test Articles and Design/Planning for MAGNET Testing

The nuclear industry is pursuing microreactors that can be factory assembled and deployed to remote regions for reliable power generation. One class of microreactors uses a monolithic metal core block coupled to heat pipes for heat rejection, which results in significant thermal stresses in the monolithic structures. This work describes the initial characterization and test plan for evaluating stainless steel test articles fabricated with embedded sensors for measuring heat pipe performance limits, as well as spatially distributed temperatures and strains during electrically heated thermal testing. The electrically heated testing will be performed in the non-nuclear Microreactor Agile Non-Nuclear Testbed and Single Primary Heat Extraction and Removal Emulator facilities located at Idaho National Laboratory. The goals of these tests are to (1) accurately monitor temperature and strain distributions that result from differential thermal expansion in the test articles and (2) quantify heat rejection limits of heat pipes as a function of operating temperature and working fluid during steady-state and transient operations. More generally, the ability to monitor component and system health during microreactor operation is attractive for providing a high sensor density to inform a limited number of microreactor operators to ultimately reduce operation and maintenance costs and move toward semi-autonomous operation. This report discusses the characterization of embedded thermocouples and fiber optic sensors in relevant test articles, including cylindrical pipes and hexagonal monolithic test articles for heat pipe-based reactors. The sensors were embedded by placing them in machined channels and then building up additional material by using ultrasonic additive manufacturing (UAM). UAM is a solid-state welding process that uses downward pressure and a lateral scrubbing motion to bond thin metal foils to a base material layer by layer. The ultrasonic welding process relies on the plastic deformation of the metal—as opposed to typical melting and solidification—to break oxide scales and bond the metal layers. The characterization of these embedded sensors included evaluating fiber optic signal attenuation, observing residual strain in the fibers, investigating microstructural and mechanical aspects, and demonstrating the sensors under various thermal loads and acoustic vibrations. Post-embedding characterization showed a fine grain structure (<1 μm) near the interfaces of the bonded foils as a result of severe deformation from the welding process. A large increase in hardness was observed at the foil interfaces and the fiber/matrix interface compared with the bulk matrix. Even when compared with the SS304 interfaces, the higher hardness observed around the embedded fiber suggests a higher degree of deformation due to the soft metal coating around the silica fiber core. The distributed fiber-optic temperature sensors and embedded thermocouples reliably measured temperature distributions during steady-state and transient thermal testing. The embedded fiber-optic sensors reliably measured strain during both transient and steady-state testing and properly identified resonant frequencies during acoustic testing.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Report on Year-4 of Water NSTF Matrix Testing: Facility Maintenance and Accident Testing

Under support from the Department of Energy (DOE) and the Office of Advanced Reactor Technologies (ART), a large-scale test facility has been constructed at Argonne National Laboratory to generate NQA-1 qualified validation data for passive decay heat removal systems in advanced reactors. The Natural convection Shutdown heat removal Test Facility (NSTF) reflects key features of a ½ scale, water-based, Reactor Cavity Cooling System (RCCS) and is intended to study the behavior, bound performance, and ultimately guide design decisions for passive decay heat removal systems for advanced reactors. In addition to the experimental activities detailed in this report, a supportive computational modeling effort is on-going which has been demonstrated to significantly strengthen the experimental program while also improving accuracy of the computer models. Together these create a mutually beneficial relationship integral to meeting the overall program objective of examining the heat removal performance of the RCCS concept. This report serves as a summary of maintenance and experimental activities during the program’s fourth year of water-based operation. A planned six-month maintenance period began in August 2021, during which major inspections, repairs, cleaning, and installation of new instrumentation and data acquisition hardware were conducted. Most significantly, two heaters that faulted during Year-3 were repaired, allowing the facility to resume use of the full heated section area and full range of available electric power. The remainder of the year consisted of eight months of test operations, during which the facility logged 211 hours of active heating across one bake-out (following the maintenance period) and seven matrix tests; five classifieds as Accepted per NQA-1, one as Trending, and one as Failed. Testing began by performing two repeat cases to confirm expected facility response and behavior during both single- and two-phase flow conditions, ensuring no changes were introduced during the maintenance period that might have altered the thermal-hydraulic characteristics of the facility. In continuation of the power parametric series initiated in previous years, a high-power test case was then performed examining heat removal performance at a decay heat load equivalent to 2.4 MWt, full-scale, a level exceeding maximum design targets. Additional testing then introduced various blockage scenarios along the network piping, examining the effects of partial and complete blockages of the flow paths on the system behavior and heat removal performance. A study of static boiling tests directed at understanding the geysering two-phase instability was also conducted. The loop was filled only to the bottom of the tank outlet, creating an open loop configuration that prevents any natural circulation flow from occurring, and the heaters were powered on until the facility reached saturation conditions. Following, a series of quasi-steady-state conditions were introduced by adjusting the inventory level in the adiabatic chimney piping at decreasingly lower elevations above the heated region. A strong correlation of geysering characteristics and loop level was observed, with flow and temperature excursions decreasing in intensity, but increasing in frequency, as the fill were reduced to lower elevations along the chimney piping. Once the level fell very low in the chimney, at points near the top of the heated section, the system reached a stable state of continuous boiling without any occurrence of geysering eruptions. A final significant testing accomplishment this year was successful completion of an “accident scenario” test, whose operating conditions were based on a prototypic decay heat curve provided by Framatome and scaled for the NSTF. This test began by establishing steady-state, single-phase “normal operation” conditions, before simulating an accident trip where the availability of active cooling systems was lost. Loop temperatures gradually increased until reaching saturation and subsequent two-phase boiling flow. Over the course of an extended operational period along the defined decay heat curve, steam boil-off caused gradual but continued depletion of liquid inventory until reaching a critically low level causing flow stagnation and cessation of natural circulation heat removal. At this point, after nearly 72 hours of continuous operation, a cold refill was performed to replenish the system inventory and allow the facility to re-establish closed loop natural circulation flow and return to a safe operational state.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Characterizing the Influence of Abstraction in Full-Scale Wind Turbine Nacelle Testing

In recent years, there has been a growing interest in full-scale wind turbine nacelle testing to complement individual component testing. As a result, several wind turbine nacelle test benches have been built to perform such testing with the intent of loading the integrated components as they are in the field. However, when mounted on a test bench the nacelle is not on the top of a tower and does not have blades attached to it - this is a form of abstraction. This paper aims to quantify the influence of such an abstraction on the dynamic response of the nacelle through a series of simulation case studies. The responses of several nacelle components are studied including the main bearing, main shaft, gearbox supports, generator, and yaw bearing interface. Results are presented to highlight the differences in the dynamic response of the nacelle caused by the abstraction. Additionally, the authors provide recommendations for mitigating the effects of the abstraction.

17 WIND ENERGY↗
DOE OSTI2018

Advanced Photovoltaic Inverter Control Development and Validation in a Controller-Hardware-in-the-Loop Test Bed

Penetration of solar photovoltaic (PV) installations into the grid has increased in recent years. In the past, photovoltaic installations have not included grid-support functionality. Grid support functions including Volt-VAR (VVAR), Frequency-Watt (FW), Volt-Watt (VW), frequency/voltage-ride through, as well as other inverter functions such as anti-islanding are recommended by standards such as IEEE 1547 to allow PV inverters to contribute to acceptable operation and stability of distribution systems with high penetrations of PV and to ensure predictable response of the PV inverters. First, low-level and high-level grid support inverter control functions were developed and tested in simulation. The developed controller is able to integrate advanced controller functions in a coordinated approach, avoiding any conflict between the different control objectives. The algorithms were then programmed on an off-the-shelf embedded controller that included a dual-core CPU and FPGA. The details of the high-level grid support function and the lower level switching control will be presented. Second, this programmed controller is tested using a controller-hardware-in-the-loop (CHIL) test bed setup using an FPGA-based real-time simulator. The CHIL setup has the capability to run at a 500 nano-second time-step to accommodate the 20 kHz switching frequency of the developed controller. The advanced control function and CHIL test bed details provided here will aide future researchers to design, implement, and test advanced inverter functions for PV inverters.

14 SOLAR ENERGY↗
DOE OSTI2017

Performance Testing of Tracer Gas and Tracer Aerosol Detectors for use in Radionuclide NESHAP Compliance Testing

The Rad-NESHAP program, part of the Air Quality Compliance team of LANL’s Compliance Programs group (EPC-CP), and the Radiation Instrumentation & Calibration team, part of the Radiation Protection Services group (RP-SVS), frequently partner on issues relating to characterizing air flow streams. This memo documents the most recent example of this partnership, involving performance testing of sulfur hexafluoride detectors for use in stack gas mixing tests. Additionally, members of the Rad-NESHAP program performed a functional trending test on a pair of optical particle counters, comparing results from a non-calibrated instrument to a calibrated instrument. Prior to commissioning a new stack sampling system, the ANSI Standard for stack sampling requires that the stack sample location must meet several criteria, including uniformity of tracer gas and aerosol mixing in the air stream. For these mix tests, tracer media (sulfur hexafluoride gas or liquid oil aerosol particles) are injected into the stack air stream and the resulting air concentrations are measured across the plane of the stack at the proposed sampling location. The coefficient of variation of these media concentrations must be under 20% when evaluated over the central 2/3 area of the stack or duct. The instruments which measure these air concentrations must be tested prior to the stack tests in order to ensure their linear response to varying air concentrations of either tracer gas or tracer aerosol. The instruments used in tracer gas and aerosol mix testing cannot be calibrated by the LANL Standards and Calibration Laboratory, so they would normally be sent off-site for factory calibration by the vendor. Operational requirements can prevent formal factory calibration of some instruments after they have been used in hazardous settings, e.g., within a radiological facility with potential airborne contamination. The performance tests described in this document are intended to demonstrate the reliable performance of the test instruments for the specific tests used in stack flow characterization.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗