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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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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Testing of PIP-II pre-production 650 MHz couplers in warm test stand and in cryomodule

650 MHz fundamental power couplers were developed for PIP-II project to deliver RF power for low-beta and high-beta elliptical cavities. Few prototypes were built and tested and after some modification we built 8 pre-production couplers (with few spares for vacuum side) for HB650 cryomodule. All couplers were successfully tested in pulse mode (up to 100kW) and in CW mode (up to 50kW) in test stand at full reflection at 8 phases. In baseline configuration with DC bias we do not see any multipactoring activity after short processing. We also tested power processing without bias for uncoated and TiN coated ceramic window. Results of these studies presented in this paper. One of the coupler was assembled on LB650 cavity and tested at cryogenic environment in STC cryostat at ~20kW power with full reflection at different reflection phase. We also demonstrated good result from power processing without bias for warm and cold cavity. Six couplers were assembled on HB650 cavities in pre-production cryomodule. Test results from cryomodule qualification is discussing in this paper.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Gear Test Assembly – Experimental Testing and Analysis of Gears and Bearings – FY2024

The Gear Test Assembly (GTA) is an experimental test apparatus built and installed in the Mechanisms Engineering Test Loop (METL) at Argonne National Laboratory (ANL). The focus of GTA has been to test larger radial spur gear sets machined from Inconel 718 along with a variety of bearing components that are to be used in the fuel handling machines built for use in pool-type sodium fast reactors. The testing in GTA also informs the design and material choices for the full-scale Gripper Test Assembly (GrTA) being developed by Argonne.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Thermal Hydraulic Experimental Test Article: Second Year of Testing with Secondary Sodium System (Fiscal Year 2025 Final Report)

The Thermal Hydraulic Experimental Test Article (THETA) is currently installed in the Mechanisms Engineering Test Loop (METL) 28” test vessel #4. Both the primary and secondary sodium systems remain online to facilitate continued testing. This fiscal year, work was performed using a COMSOL Multiphysics magnetohydrodynamic model to characterize flow more accurately in the secondary electromagnetic flowmeters. Experimental campaigns were then performed to study the thermal hydraulic differences between sodium and water as a surrogate fluid in the THETA geometry as well as a study to better characterize and understand temperature oscillations that exist at the outlet of the core to the hot pool. A peer-reviewed article was published in the ASME Journal of Nuclear Engineering and Radiation Science detailing the THETA facility and providing an overview of a test that was performed with the primary and secondary system online [1]. Work continues to develop a database to house experimental THETA data to better facilitate collaboration with industry and laboratory partners for their use of the data for code benchmarking/validation. THETA remains fully operational and is positioned for continued testing in fiscal year 2026.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Design of Modular Test Facility for Particle/sCO2 Heat Exchanger Evaluation

Heliogen has designed a 1.3 MW particle and supercritical carbon dioxide (sCO2) test loop to retire technical and manufacturing risks. The data from this facility will be used to validate near commercial-scale particle heat exchanger modules. The heat exchanger is scaled to capture all features of the full-scale modules. Testing will be conducted with CARBOBEAD HSP 16/30, a larger particle size than other recent testing. The system was designed to ASME codes with the constraint that sCO2 piping and other major components are stainless steels.

14 SOLAR ENERGY↗
DOE OSTI2020

Mechanical Characterization of Fe-Co-2V (Hiperco): Fatigue/Monotonic Testing Hardness Testing and Fractography

Fe-Co-2V is a popular metallic alloy used in electromagnetic applications. However, there is a lack of mechanical fatigue characterization of this alloy in the literature. In this work, Fe-Co-2V specimens with rectangular cross-sections were carefully prepared in accordance with standards. They were measured for surface roughness and then subjected to quasi-static monotonic testing, as well as fatigue testing at both 0.5 Hz and 1 Hz frequencies. Both Rockwell hardness and Vickers micro- hardness testing were performed. Additionally, scanning electron microscopy imaging of the fractured surfaces was done. The quasi-static testing revealed a flat yield region characteristic of Laders bands. Here, the fatigue results did not show significant differences or sensitivity to change in frequency, although the fatigue life was higher on average for the 0.5 Hz. However, the fatigue results differed from published work at 0.33 Hz. The fractography revealed purely brittle fracture, with clear chevron marks and fracture initiation always starting at the surface. Lastly, it was identified that the C, D, and F Rockwell hardness scales were appropriate for testing this material and that the grain size necessitated the use of the upper end of indentation force for Vickers micro-hardness testing.

36 MATERIALS SCIENCE↗
DOE OSTI2023

A standard capsule design for structural material testing in the Advanced Test Reactor

Nuclear materials testing is commonly completed in various material test and research reactors throughout the world, including the Advanced Test Reactor (ATR), but the current capsule design, analysis, and fabrication process can take years to complete. To decrease the costs and time associated with materials testing, a standard capsule has been designed which houses various specimen geometries and allows irradiation in virtually any ATR test position. The standard capsule features locking end caps which connect and lock together to form the capsule stack, eliminating the need for a basket and maximizing the quantity of specimens which are contained within the capsule. A customizable internal gas gap provides thermal resistance between the specimens and reactor coolant, making specimen temperatures from approximately 370 to 1000 K achievable. The flexibility of the capsule design allows experimenters to choose irradiation positions based off desired neutron flux, with typical fluences per cycle ranging from 8.8x10 19 to 2.3x10 21 n/cm 2 , depending on experiment position. Here this paper presents and discusses the standard capsule design and analysis.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Hardware-in-the-Loop Test Bed and Test Methodology for Microgrid Controller Evaluation

Increasing penetration levels of distributed energy resources are making the deployment of microgrids more feasible. Controllers that operate such microgrids are key to realizing the objectives of the microgrid owner or operator and there is a need to evaluate microgrid controller performance prior to field deployment. This paper describes a controller hardware-in-the-loop and power hardware-in-the-loop microgrid controller test bed that was designed and constructed to evaluate the capabilities of a microgrid controller for a proposed campus microgrid. This paper also presents a test methodology to evaluate microgrid controller functionality, and it describes how the controller was assessed through the application of different test scenarios. Results from the testing are presented to provide insight into the capabilities of the test bed.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2025

Experimental Validation of a High-Temperature Test Facility for Future Additive Manufactured Supercritical Carbon Dioxide Turbine Testing

For next-generation power plants to achieve high cycle efficiencies consistent with the Department of Energy's 65% efficiency target, turbomachinery capable of operating within high-temperature power cycles must be demonstrated. Pairing additively manufactured superalloy turbines with the supercritical carbon dioxide (sCO 2 ) power cycle could enable turbine inlet temperatures approaching 1300 °C while providing flexibility in turbine cooling strategies. Development of test facilities to characterize and validate such systems is crucial. In this study, a turbine test facility capable of achieving inlet conditions of 800 °C, 11 MPa, and 0.43 kg/s while accommodating complex auxiliary cooling flow requirements has been designed and constructed to support future testing of a Haynes 282 additively manufactured 30 kW turbine-generator system with advanced cooling channels. This facility enables characterization of aerodynamic performance, leakage, and windage losses. Details of the facility's construction and operation are presented, along with experimental validation tests using an orifice as an expansion device in place of the turbine. These tests confirm that the facility can reach the required conditions, distinguish regions of achievable steady and pseudo-steady conditions, and identify the heater power required for each point in the upcoming turbine campaign. The campaign confirmed that ISO 5167-2 can reliably predict orifice mass flow rates in extreme supercritical carbon dioxide conditions with deviations of 0.5–7.5%.

42 ENGINEERING↗
DOE OSTI2024

3013 Fire Test 2020: Post-Test Measurements and Analysis

Sandia National Laboratory conducted fire tests on ten 3013 containers with varying amounts of water. The containers with attached manifolds were sent to Los Alamos National Laboratory for post-test measurements of the plastic deformation, the amount of water remaining on the material post-test and the mass loss. The measurements of the plastic deformation were used to determine the free gas volume in each container and manifold post-test. The disassembly of each container was documented by photographs. The amount of pre-test gas constituents is also reported using data in the Sandia Final Report.

3013 fire test↗
DOE OSTI2021

Thermal Hydraulic Experimental Test Article - Report of THETA Commissioning for METL Testing

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in a sodium-cooled fast reactor (SFR). High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. THETA is comprised of a traditional primary and secondary system. The primary system is submerged in the pool of sodium and consists of a pump, electrically heated core, intermediate heat exchanger, and connected piping and thermal barriers (redan). The secondary system, located outside of the sodium pool, consists of a pump, sodium to air heat exchanger, and connected piping and valves. Figure 1 illustrates the main components of the primary system. THETA has been installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, Figure 2 and Figure 3. Since the FY20 THETA status report, [1], further water shakedown testing was performed, some final welding modifications were performed, all system components were sanitized, dry assembled, and commissioned before installing into METL test vessel #4. The THETA submersible electromagnetic flowmeter was welded closed to hermetically seal internal components from sodium followed by a helium leak check for the final time

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Out-of-Pile Test of LVDT-Based Creep Test Rig at PWR Prototypical Conditions

New and improved materials are being considered to support the existing nuclear reactors and future next-generation reactors. The materials can significantly degrade and limit their properties in harsh reactor environments. To accurately understand the material’s degradation, real-time data is required under prototypic-irradiation conditions. Additionally, understanding the creep behavior of materials under harsh environments is essential for an evaluation of safety concerns. To provide these capabilities, the Idaho National Laboratory’s (INL’s) High Temperature Test Laboratory (HTTL) has developed several instrumented test rigs to obtain real-time data from specimens in well-controlled pressurized water reactor (PWR) coolant conditions in the Materials and Test Reactors (MTRs). This technical report focuses on INL’s efforts to evaluate and enhance the former prototype creep test rig that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig is capable of detecting changes in the length of tensile specimen, which is useful for measuring thermal expansion and creep loading.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

VERAIn Software Requirements, Test Plan, and Test Report

This document describes the software test plan for VERAIn and provides appendices for the software requirements and software test report. In this document, the test platform hardware and software are described. Appendix A provides a list of the tests run and their acceptability as a test report, and Appendix B provides the list of low-level software requirements as a requirements traceability matrix.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Gear Test Assembly - Experimental Testing and Gear Analysis - FY2022

The Gear Test Assembly (GTA) is an experimental test article built for use in the Mechanisms Engineering Test Loop (METL) at Argonne National Laboratory (ANL). The GTA was the first test article to operate in METL, with the initial experimental campaign beginning in early 2019. A total of four experimental campaigns have been completed to-date, and the fifth is currently underway. This report will discuss the results of the four completed campaigns with gear analysis in the form of over-pin dimensions. Finally, information regarding Experimental Campaign #5 will be discussed. All over-pin measurements of the gears tested in sodium will be available in the Appendix.

42 ENGINEERING↗
DOE OSTI2022

MPACT Software Test Plan, Requirements, and Test Report (V.4.3)

This document presents the software test plan for MPACT. The software requirements and software test report are provided in the appendices. The test platform hardware and software are described herein. Appendix A provides a list of the tests that were run and their acceptability as a test report. Appendix B provides the list of low-level software requirements as a requirements traceability matrix.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2024

Assessment of Potential Dose and Environmental Impacts from Proposed Testing at the INL National Security Test Range

This assessment uses screening level models to calculate potential environmental impacts from proposed tests at two locations at the Idaho National Laboratory (INL) National Security Test Range (NSTR) site. Proposed tests could be conducted using 11 different radioactive material types that include K 2 O, LaBr 3 , KBr, Cu, Zr, F, Ga, Ga 2 O 3 , NaNO 2 , Ga-68, and Tc-99m. The tests could potentially release radioactive material to the atmosphere and radionuclides and other contaminants to the soil, which could leach into the unsaturated zone and migrate to the aquifer. Atmospheric transport of radionuclides to potential human receptors and time-integrated air concentrations were calculated with a Gaussian plume model and three years of hourly meteorological data. Potential surface soil impacts were calculated with the computer program Mixing-Cell Model (MCM). Groundwater impacts were calculated with the computer programs MCM and GWSCREEN. Radiological doses from potential atmospheric releases were calculated for public receptors off the INL Site and for workers at nearby INL facilities. Results were compared to regulatory dose limits. Maximum potential groundwater concentrations were estimated in the aquifer below the NSTR site and compared to drinking water standards or risk-based screening levels for resident tap water. Soil concentrations were calculated and compared to risk-based screening levels for workers and potential future residents. All impacts were estimated based on the assumption that 12 tests are conducted annually using all 11 material types for 15 years. This document provides the resources to enable a subject matter expert in the field of environmental assessments to replicate the modeling and calculations. The methodology and parameters are presented in the text. All electronic files, including computer code input, output, executable files, batch files, scripts, and spreadsheet files, are contained in a zip file that can be accessed by selecting “Additional Information” (select Native File) in the INL Electronic Document Management System (EDMS).

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

The Structural Evaluation Test Unit (SETU) Benchmark Test Results

A series of extensively instrumented tests was performed on the Structural Evaluation Test Unit in the early 1990s. The purpose of these tests was to determine the response of a minimally designed cask to impacts that were more severe than the design basis impact. This test series provides an excellent opportunity for benchmarking explicit dynamic finite element analysis programs for behaviors that may be experienced by casks during regulatory and extra-regulatory impact events. This report provides the results of the four tests that were conducted. It is meant to go along with a companion report that defines the benchmark problem and gives the locations for the instrumentation and inspection points.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Accident Tolerant Fuel Test 2B (ATF-2B) Irradiation Test Report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2 has completed 7 cycles of prototypic steady state irradiation irradiating 41 test pins to burnups as high as 30 MWd/kgU. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 41 test pins has been calculated using a combined MCNP, and ORIGIN methodology to provide accurate and reliable power histories on a per pin basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗