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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 1,387 records · Page 77
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

SU-F-T-180: Evaluation of a Scintillating Screen Detector for Proton Beam QA and Acceptance Testing

Purpose: To test the performance of a commercial scintillating screen detector for acceptance testing and Quality Assurance of a proton pencil beam scanning system. Method: The detector (Lexitek DRD 400) has 40cm × 40cm field, uses a thin scintillator imaged onto a 16-bit scientific CCD with ∼0.5mm resolution. A grid target and LED illuminators are provided for spatial calibration and relative gain correction. The detector mounts to the nozzle with micron precision. Tools are provided for image processing and analysis of single or multiple Gaussian spots. Results: The bias and gain of the detector were studied to measure repeatability and accuracy. Gain measurements were taken with the LED illuminators to measure repeatability and variation of the lens-CCD pair as a function with f-stop. Overall system gain was measured with a passive scattering (broad) beam whose shape is calibrated with EDR film placed in front of the scintillator. To create a large uniform field, overlapping small fields were recorded with the detector translated laterally and stitched together to cover the full field. Due to the long exposures required to obtain multiple spills of the synchrotron and very high detector sensitivity, borated polyethylene shielding was added to reduce direct radiation events hitting the CCD. Measurements with a micro ion chamber were compared to the detector’s spot profile. Software was developed to process arrays of Gaussian spots and to correct for radiation events. Conclusion: The detector background has a fixed bias, a small component linear in time, and is easily corrected. The gain correction method was validated with 2% accuracy. The detector spot profile matches the micro ion chamber data over 4 orders of magnitude. The multiple spot analyses can be easily used with plan data for measuring pencil beam uniformity and for regular QA comparison.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

TEAMER: Pitching Foil Crossflow Turbine Efficiency Data

This dataset documents the efficiency testing of a pitching foil crossflow turbine, conducted at the University of New Hampshire's (UNH) Chase Ocean Engineering Laboratory tow tank facility. The tests explored various pitch phases and amplitudes, ranging from 0 to 18 degrees, across different flow speeds and turbine RPMs. Specifically, testing was performed at pitch amplitudes of 0,3,6,9,12,15 and 18 degrees. The results indicate that turbine efficiency improves with increased pitch amplitude, reaching optimal performance at 12 degrees. Modifications were made to the test frame by UNH to enhance the stiffness of the support structure. The testing specifically mapped the efficiency of a single quadrant of the Ocean Renewable Power Company (ORPC) Autonomous Turbine Generator Unit (ATGU), covering the full range of allowable pitching amplitudes and phases. The dataset provides both raw and processed data, including drag and torque measurements, and contains Python scripts used for data processing and visualization. This research was supported by funding from TEAMER RFTS 1 (Request for Technical Support).

16 TIDAL AND WAVE POWER↗
DOE OSTI2018

Production Facility Prototype Blower 1000 Hour Test Results III

Long duration operations tests of the Aerzen GM 12.4 roots style blower in a closed loop configuration provides valuable data and lessons learned for long-term operation at the Mo-99 production facility. The blower was operated in a closed loop configuration with the flow conditions anticipated in plant operation with a Mo-100 target inline. The additional thermal energy generated from beam heating of the Mo-100 disks was not included in these tests. Eight long duration 500 and 1000 hour tests have been completed since the first test was performed in January of 2016. All eight long duration tests have proven successful in exposing preventable issues related to oil and helium leaks.

47 OTHER INSTRUMENTATION↗
DOE OSTI2020

Evaluation of short-term initial performance of a “Molekule Air” portable air cleaner

The aim of this project was to evaluate and characterize the performance of a MOLEKULE AIR air cleaner unit. The testing protocol comprised the evaluation of an unused Molekule Air unit operating at maximum (“boost” setting) fan speed setting. The device was tested over initial short-term periods of 70-80 h of continuous operation. The tested air cleaner was challenged in separate experiments with a well-characterized mixture of volatile organic compounds (VOCs), and with ozone. All tests were performed at realistic pollutant levels in a 20-m 3 environmental stainless-steel chamber. Variables controlled and measured in this study included the chamber temperature, relative humidity, the composition and concentration of the challenge VOCs, and the concentrations of ozone. Potential formation of byproducts was also investigated to assess the overall performance.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Extraction, Scrub, and Strip Test Results for the IBC BOBCalix Solvent Samples

A second vendor for solvent extraction components has been chosen for use at the Salt Waste Processing Facility (SWPF). IBC Advanced Technologies (IBC) has provided samples of their BOBCalix, as well as complete solvent. These samples were tested for quality assurance purposes to ensure their use at SWPF is acceptable. Two Extraction, Scrub, and Strip (ESS) tests were performed on solvent samples from IBC. The purpose of testing these samples was to determine if the solvents are displaying the correct extraction and stripping behavior with cesium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Test Beam and Irradiation Facilities

Progress in particle physics depends on a multitude of unique facilities and capabilities that enable to advance detector technologies. Among others, key facilities involve test beams and irradiation facilities, which allow users to test the performance and lifetime of their detectors under realistic conditions. Test beam facilities are particularly important for collider and neutrino detector applications, while irradiation facilities are crucial for collider as well as some space-based astro particle detectors. This contributed Snowmass paper aims to summarize existing test beam and irradiation facilities as well as develop the need and proposals for future facilities.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

A Perspective on the Successes of the NNSS Underground Test Area (UGTA) Activity - 20221

Between 1951 and 1992, 828 underground nuclear tests were performed at the Nevada National Security Site (NNSS). Underground testing was done in five major testing areas, which included (1) Frenchman Flat, (2) Rainier Mesa/Shoshone Mountain, (3) Yucca Flat/Climax Mine (4) Central Pahute Mesa, and (5) Western Pahute Mesa. About one third of the underground tests were detonated near, at or below the water table, and thus radioactive contamination was introduced to the groundwater system. The U.S. Department of Energy's Underground Test Area (UGTA) Activity was established in the late 1990's to address the fate and movement of residual radioactivity in groundwater, and characterize the risk that it may pose to NNSS workers and the offsite public. It has accomplished this goal through a process of iterative groundwater sampling and numerical groundwater flow and transport modeling, and long-term monitoring. The UGTA Activity draws on the expertise of scientists in the fields of geology, hydrology, radiochemistry, and risk assessment from the U.S. DOE staff, the lead contractor (currently Navarro) and many other organizations, in cooperation with the governing regulatory body, the State of Nevada's Division of Environmental Protection (NDEP). Los Alamos National Laboratory (LANL) is one of the participating research organizations involved with the UGTA Activity. Under the direction of DOE and the lead contractor, LANL's role in UGTA has evolved over time from an initial focus on geologic characterization, to conducting field and laboratory experiments, and finally to its current role of providing modeling and geochemistry expertise to characterize the rates and directions of groundwater and radionuclide movement. In its current role, LANL has either developed or provided the numerical tools for developing flow and transport models in each of the four major testing areas. In addition, LANL has provided an independent assessment of future groundwater flow and transport behavior through the analysis of naturally-occurring geochemical and isotopic tracers in groundwater. The five major testing areas are now in different stages of investigation or closure: (1) Frenchman Flat is in its fifth year of post closure modeling; (2) The Rainier Mesa/Shoshone Mountain Closure Report has been submitted to NDEP and is awaiting approval; (3) The Yucca Flat/Climax Mine Closure Report is under development; and (4) Central and (5) Western Pahute Mesa is still undergoing investigation and flow and transport model development. The success of the UGTA Activity in reaching or approaching closure in three of the five major testing areas is primarily due to the DoE's focus on regulatory closure requirements, but also is in part attributable to a rigorous internal review process and the involvement of NDEP and Nye County water professionals as participants and observers in the reviews. County and state involvement in the internal review process has promoted trust that the U.S. DOE is deeply committed to ensuring the present and future safely of NNSS workers and the surrounding communities. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

AGR-2 Compact 6-4-1 Post-Irradiation Examination Results

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established to perform research and development on tristructural isotropic (TRISO)-coated particle fuel to support deployment of a high-temperature gas-cooled reactor (HTGR). The work continues as part of the Advanced Reactor Technologies (ART) TRISO Fuel Program. The overarching goal of the program is to provide a baseline fuel qualification data set to support licensing and operation of an HTGR. To achieve these goals, the program includes the elements of fuel fabrication, irradiation, post-irradiation examination (PIE) and safety/heating testing, fuel performance modeling, and fission product transport (INL 2017). Several fuel irradiation experiments have been performed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), and additional irradiations are planned. These experiments are intended to provide data on fuel performance under irradiation, support fuel fabrication process development, qualify fuel for operating and accident conditions, provide irradiated fuel for accident testing, and support development of fuel performance and fission product transport models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Restart of the Transient Reactor Test (TREAT) Facility Neutron Radiography Program

The TREAT radiography system is used to perform neutron radiography of fuels, experiments, and other specimens before and after irradiation within the TREAT reactor. The TREAT neutron radiography facility performed approximately 5,000 radiographs by the spring of 1977. Originally built in 1958, the TREAT Facility was in operation until it was placed in a shutdown status in 1994. Following the Fukishima disaster and seeing a need for enhanced accident tolerant fuels, the United Stated Department of Energy decided to restart the TREAT facility and resume transient operations. In November 2017, the TREAT reactor was successfully restarted and is currently performing operational testing in preparation for initial experiment irradiations and transient testing. This paper discusses efforts to reactivate the TREAT neutron radiography facility. To characterize the neutron beam, gold foil activation measurements were made to determine an average neutron flux and flux profile. An open beam image provides the information about variations in the beam profile. In conclusion, a series of system qualification radiographs have been acquired to determine the effective image acquisition parameters, resulting image quality, and the relationships between the two.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Extraction, scrub, and strip test results for the solvent transfer to salt waste processing facility

The Savannah River National Laboratory (SRNL) prepared approximately 240 gallons of Caustic-Side Solvent Extraction (CSSX) solvent for use at the Salt Waste Processing Facility (SWPF). An Extraction, Scrub, and Strip (ESS) test was performed on a sample of the prepared solvent using a salt solution prepared by Parsons to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams. This data will be used by Parsons to help qualify the solvent for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 15.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Generation IV Benchmarking of TRISO Fuel Performance Models Under Accident Conditions Final Report

The Generation IV International Forum (GIF) is a co-operative international endeavor of fourteen members organized to carry out the research and development needed to establish the feasibility and performance capabilities of the next generation nuclear energy systems. GIF selected six reactor technologies, amongst which is the Very High Temperature Reactor (VHTR) that is primarily dedicated to the cogeneration of electricity and hydrogen. The technical basis for VHTR is the tristructural isotropic (TRISO)-coated particle fuel, the graphite as the core structure, helium coolant, as well as the dedicated core layout and lower power density to removal decay heat in a natural way. At the heart of safety features of the VHTR concept lie the TRISO fuel particles that are designed to keep their structural integrity and retain fission products at temperatures up to 1600°C. As part as the design and future operation of VHTRs, a key aspect is the accurate prediction of fuel performance under irradiation and accident conditions. Modeling and simulation allow prediction of TRISO fuel behavior when subject to neutron flux and in high temperature accident scenarios. The refinement of the fuel performance models and codes is performed by comparison to in-pile and out-of-pile experimental data that reproduce the expected irradiation conditions in high temperature gas-cooled reactors (HTGRs). Historically, the International Atomic Energy Agency (IAEA) developed a benchmark dedicated to the validation of predictive methods for fuel and fission product behavior through the Coordinated Research Program CRP-2 (IAEA, 1997). CRP-2 was later updated to cover fuel fabrication, quality assurance, irradiation performance, safety testing, and spent fuel. The scope of the resulting CRP-6 benchmarks focused on HTGR fuel performance and fission product release (IAEA, 2012). Taking advantage of additional TRISO fuel fabrication, irradiation, and safety testing campaigns, GIF launched a Generation IV Benchmarking of TRISO Fuel Performance Models under Accident Conditions in late 2015. This GIF benchmark is a three-year program steered by Idaho National Laboratory (INL, USA). The other participants include the Japan Atomic Energy Agency (JAEA, Japan) and the Korea Atomic Energy Research Institute (KAERI, Korea). The objectives of the benchmark are to: follow on the IAEA CRP benchmarks, (2) model fission product release under accident conditions, (3) compare results obtained by the fuel performance modeling codes of the benchmark participants, and (4) compare these code predictions to experimental data. Safety tests chosen for modeling include the first and second experiments of the Advanced Gas Reactor program (AGR-1 and AGR-2) and the High Flux Reactor (HFR) EU1bis experiment. This report presents the results obtained by the three research institutions using their respective fuel performance modeling codes. Comparisons of the corresponding fission product release predictions are made with experimental data from AGR-1, AGR-2, and HFR-EU1bis. The benchmark results show good agreements between all participants but also show a general trend of over-prediction of the experimental release data, which is mainly attributed to the use of over-estimated diffusion coefficients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Drop Testing of a Container for the Storage, Transport and Disposal of Intermediate Level Waste - 18078

Impact tests were performed, as part of a corresponding container's substantiation, during design development of a shielded container proposed for use at Sellafield for waste retrieval from the First Generation Magnox Storage Pond (FGMSP) in the UK. The mechanical test program comprised a 9 m free drop test onto an unyielding target in a container long lid edge down orientation at ambient conditions. Further, a 0,5 m free drop test onto a punch target was performed. Here, the container was orientated with the lid downwards in a way that the punch, a mild steel bar, impacts a filter lid. The test specimen was instrumented with strain gauges and accelerometers for the drop tests. Transient strains at selected points of the inner and outer container walls, at the shielding lid, as well as at the lid bolts were measured during the container's impact. Furthermore, decelerations of the container body, container lid, and the skip were measured. The complex geometrical changes of the container due to impact were determined by optical 3d- deformation measures using the projected fringes method in combination with multi-image photogrammetry. This paper summarizes the performance of the drop tests and various drop test results in context with the design development as well as aspects regarding the associated Finite Element (FE) analyses and post-test evaluation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

PLATFORM for Product Launch

Pecan Street Inc. (PSI) and its partners set out in 2017 to validate the PLATFORM for Product Launch, a new model for improving the effectiveness of investments in clean energy technologies and accelerating market entry of disruptive technologies. In partnership with a newly formed Innovation Advisory Council, PSI recruited fourteen companies with new hardware-based technologies that enable or create clean energy opportunities. Pecan Street carried out product evaluations and conducted market research with stakeholders and target customers to develop recommendations on product optimization opportunities for all companies, and created third party performance evaluation reports that include environmental metrics, such as water and energy savings potential, to assist in creating investment confidence and to unlock additional sources of funding, such as impact investments. Pecan Street set-up the product testing environment, performed small-batch field testing, and provided a verified third-party performance report on the product of twelve companies. PSI Street and its partners leveraged the performance evaluations and insights from the newly formed PRI toolkit to raise over $1.5M across five companies, and garner interest for private sector adoption of the PLATFORM for Product Launch.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2021

FY21 Status Report on Creep Rupture Testing at ANL to Support the Development of Alloy 709 Code Case

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. Creep tests were performed in air on the lever-arm load frames in the Mechanical and Environmental Testing Laboratory at Argonne National Laboratory (ANL). In FY21, the focus was on the age-hardened ESR-1150°C material. Twenty two tests were initiated, among which ten were completed. This report presents the data from the completed tests. Metallographic characterization of the fractured specimens were performed and the results are included. The creep data enabled the preliminary analysis of the creep behavior of this material. The characterization showed the effect of test temperature and stress on the deformation microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Measurements of Hydrogen Generation Rates using a Sealed Reactor System - 19428

Sealed reactor tests were performed to measure the hydrogen generation rate (HGR) from stored radioactive waste at the Savannah River Site. HGR may be very low, on the order of 7.5 x 10{sup -13} (m{sup 3}/h)/L [that is, cubic meters of H{sub 2} generated per hour per liter of radioactive liquid waste, or 10{sup -10} (ft{sup 3}/h)/gal]. Because the estimated HGR limit of detection for an open, continuous flow, system is approximately 7.5 x 10{sup -10} (m{sup 3}/h)/L [or 10{sup -7} (ft{sup 3}/h)/gal], a more sensitive method was needed. By accumulating generated gasses in a sealed system without a continuous purge gas, the resulting concentrations can increase the sensitivity of measurement, which is based on time of heating. In 2017, Savannah River Remediation declared a Potential Inadequacy in the Safety Analysis related to the potential thermal decomposition of organics that generate hydrogen in each of three Liquid Waste facilities: Concentration, Storage and Transfer Facility, Saltstone Production Facility, and the Defense Waste Processing Facility. The research for this study concerns low activity waste of Tank 50, which feeds the Saltstone Production Facility. This study evaluated the impact of organics including formate on the generation of H{sub 2}. Samples of SRS Tank 50 waste, which is a nominal 5 molar sodium salt solution, were sealed in stainless steel reactors and held at varying temperatures, e.g., 50 deg. C, 75 deg. C, 100 deg. C, and 120 deg. C, to determine if organic compounds are decomposing into H2. The experiments included trace solids used to stabilize waste in grout, e.g., grout-premix and Blast Furnace Slag, to determine if any solid phase reactions are present, which would accelerate or inhibit the generation of H{sub 2} at these temperatures. This paper describes the sealed reactors designed and constructed at Savannah River National Laboratory and the radioactive tests that were performed to measure HGR in both oxic and anoxic atmospheres. The shakedown tests showed the sealed reactors to be leak tight and hydrogen concentrations from calibrated standards could be accurately measured. Previous simulant tests helped to perfect the test methods and better design the radioactive tests. The radioactive results showed that the rate of hydrogen generated from Tank 50 waste, over a range of temperatures from 30 deg. C up to 120 deg. C, is bounded by an empirical relation (Hu Equation) that can be used to approximate HGRs in Tank 50 radioactive waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Extraction, scrub, and strip test results for the salt waste processing facility caustic side solvent extraction solvent example

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D(Cs)), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D(Cs) measured 12.9, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D( Cs )), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D( Cs ) measured 12.5, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges. This revision was created to correct an error. The previous revision used an incorrect set of temperature correction coefficients which resulted in slight deviations from the correct D( Cs ) results.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Extraction, Scrub, and Strip Test Results for the Salt Waste Processing Facility Caustic Side Solvent Extraction Solvent Sample

An Extraction, Scrub, and Strip (ESS) test was performed on a sample of Salt Waste Processing Facility (SWPF) Caustic-Side Solvent Extraction (CSSX) solvent and salt simulant to determine cesium distribution ratios (D (Cs) ), and cesium concentration in the strip effluent (SE) and decontaminated salt solution (DSS) streams; this data will be used by Parsons to help determine if the solvent is qualified for use at the SWPF. The ESS test showed acceptable performance of the solvent for extraction, scrub, and strip operations. The extraction D (Cs) measured 13.2, exceeding the required value of 8. This value is consistent with results from previous ESS tests using similar solvent formulations. Similarly, scrub and strip cesium distribution ratios fell within acceptable ranges

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗