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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 1,765 records · Page 98
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Overview of Hydraulic Fracturing Test Site 2 in the Permian Delaware Basin (HFTS-2)

Here, the Hydraulic Fracturing Test Site 2 (HFTS-2) is a large collaborative field-based R&D program in the Permian Delaware Basin, funded by the US Department of Energy through the National Energy Technology Laboratory (NETL) and the E&P industry, with support from academia. The projects' main objective is to improve the understating of the hydraulic fracturing process through utilization of advanced diagnostics and collection of through-fracture cores to provide undisputable evidence and attributes of the created hydraulic fractures. At the HFTS-2, in excess of $30 million was used to perform hydraulic fracturing research focusing on the Wolfcamp formation at a field site hosted and operated by Occidental. In addition to the research data collected by the project, Occidental provided a significant amount of background data for about a dozen existing wells in the test area as well as access to previously collected core. Additional technical and laboratory support was provided by the program members. Building on learnings and unanswered questions from HFTS-1 in the Permian Midland basin, the HFTS-2 used eight new producing wells and two existing (parent) wells to perform hydraulic fracturing research. Multiple science wells were drilled to sample and characterize the subsurface, including the collection of 540 feet of core in a vertical pilot hole and 948 feet of high-angle through-fracture core. The project installed permanent fiber optic cables in 3 wells to monitor near wellbore signals during fracturing and to collect cross-well strain measurements. Additional advanced diagnostics included a significant formation evaluation program on the vertical whole core, multi array moment tensor inversion capable microseismic survey, multi-well time-lapse geochemistry analysis, analysis of proppant distribution in producing child and slant core well, and others. We will provide an overview of the HFTS-2 project, including list of the consortium members, details of the test site, experiments performed, and technologies tested.

58 GEOSCIENCES↗
DOE OSTI2019

Fast Flux Test Facility Benchmark Status for FY19

The Department of Energy’s Office of Nuclear Energy Advanced Reactor Technologies (ART) Program works on promising Generation-IV R&D. The Program focus is on reactor concepts that could dramatically improve performance in sustainability, safety, economics, security, and proliferation resistance. The near-term ART Program goal is to reduce technical barriers (e.g., by resolving key feasibility and performance challenges) to improve the prospects of advanced technology reactor systems. In the long-term, integrated advanced reactor system designs will be developed that implement the chosen missions and complete the testing and demonstration necessary for licensing. The Fast Flux Test Facility (FFTF) is the most recent Liquid Metal Reactor (LMR) to operate in the United States, from 1982 to 1992. The plant description and plant data from one specific passive safety test is now an International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP). This report is one in a series documenting the PNNL efforts to support this FFTF loss of flow benchmark for the IAEA.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Performance Assessment of High Efficiency Variable Speed Air-Source Heat Pump in Cold Climate Applications

This project was part of an effort by ComEd's emerging technology program to evaluate the energy saving potential of new energy efficiency technologies. The focus of this technology assessment was to determine energy and peak demand savings potentials of a high efficiency variable speed, air-source, split system heat pump designed for cold climate applications. The results of this technology assessment will be used by CLEAResult to develop a new energy efficiency measure for Commonwealth Edison Company's incentive programs. The project utilized the National Renewable Energy Laboratory's (NREL) Thermal Test Facility to experimentally characterize cooling and heating performance of a high efficiency heat pump split system under varying outdoor climate conditions. The selected climate conditions represented summer and low temperature winter conditions in ComEd's service territory. The laboratory experimentation results were used to develop equipment performance curves required by EnergyPlus hourly simulation engine. Using typical meteorological year 3 weather data for the Chicago O'Hare airport, hourly building simulations (using the EnergyPlus engine) was utilized to estimate the annual energy savings of the high efficiency heat pump in comparison to a standard efficiency unit in the following U.S. Department of Energy (DOE) building codes program energy prototypes: Single-family residence; Strip mall; and Low-rise office.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2019

Scalable Polymer Precursor Based LiPON Type Li-Metal Interface Enabled High Performance Li-Sulfur Battery

The Battery 500 program objective is to research, develop, and demonstrate lithium battery technologies capable of achieving 500 Wh/kg with stable lifetime of 1000 cycles. This Battery 500 seedlings project has demonstrated a new approach and a pathway toward the program’s goals in just one awarded and executed phase. Cells have been assembled, tested, and evaluated. This report details efforts to develop polymer precursor electrolytes that offer properties anticipated to be similar or superior to LiPON glasses. Such precursors offer the potential to be used to process LiPON-like thin glass/ceramic coatings for use in lithium batteries. LiPON glasses provide a design basis for the synthesis of sets of oligomers/polymers by lithiation of OP(NH 2 ) 3-x (NH) x [from OP(NH) 3 ], OP(NH 2 ) 3-x (NHSiMe 3 ) x and [P=N] 3 (NHSiMe 3 ) 6-x (NH) x . Polymer electrolytes impregnated in/on celgard exhibit Li+ conductivities up to ~1×10 -5 S/cm at room temperature and are thermally stable to ≈150 °C. A Li-S battery assembled using a Li 6 SiPON composition polymer electrolyte exhibits an initial reversible capacity of 1500 mAh/g sulfur and excellent cycle performance at 0.25 and 0.5 C rate over 120 cycles, declining to 750 mAh/g at after a 220 h test period. At the same time, computational studies were performed to understand the conduction mechanism of LIPON and its interaction with Li metal. These initial project results provide a foundation for further development of the system to achieve the program goal.

25 ENERGY STORAGE↗
DOE OSTI2015

FX2 advanced fogging technology - 15711

Problem statement: The need for an advanced technology that can better address potential airborne contaminants at DOE Savannah River Site 235-F. Current methods of knocking down or fixing this contamination are labor intensive, costly, and sometimes ineffective. Thus, to make exposure ALARA, the advanced fogging effort is being tested using Idaho National Laboratory's FX2 fixative agent. The FX2 fogging agent is a proprietary mixture of water, latex paint (LTX), glycerin (GLY) and sodium lauryl sulfate (SLS). Initial development began in 2006 when a DOE Small Business Innovative Research (SBIR) grant resulted in the development of a conceptual advanced capture coating fog system. This coating is shown to be excellent at reducing airborne contamination and 'fixing' particulates in place as demonstrated by Idaho National Lab. Controlling Potential Airborne Particulates: Testing capacity to knockdown airborne particulates. Evaluating ability to fix to different surfaces (e.g., glass, wood, steel, concrete, brick). Line of sight as well as non-line of sight (shadowed) locations will be tested for coverage. Adhesiveness and Coverage of Surface Area: Contact area will be placed under air pressure to attempt dislodging of dried mixture. Particle analysis utilizing ImageJ will be conducted to determine quantifiable coverage. Ignitability: Observation of the FX2's reactivity to flame and heat sources as it is being applied. Results will provide useful information for the prevention of potential hazards. Determine Shielding Properties Against Alpha Sources: Determine the shielding abilities of the FX2. Fogging agent against an alpha particle emitting point source. Evaluating Critical Properties: These properties will be evaluated to better characterize the FX2 and allow for optimal understanding of results: ASTM E1354 - Burn Rate; NFPA 296-Flashover; ASTM D2196 - Viscosity; ASTM D1331 - Surface Tension; ASTM D792 - Density. Testing Innovations: Developed by the National Institutes of Health, ImageJ is a public domain image processing program that allows for the analysis that this effort requires. Coverage: Through contrast analysis, ImageJ can be used to determine the coverage of the agent upon any of the testing surfaces. Correlations: With the feature of coverage evaluation, a relationship, and thus a graphical representation, can be made between surface coverage and radiation shielding. Future work: Engineering and manufacturing of mobile robotic fogging device. ASTM WK5394: establishing performance specifications for stabilizers to minimize migration of dispersible radioactive contamination.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Development of a vacuum compatible rotary dynamic seal for cryogenic liquids

Liquid nitrogen is commonly circulating through radiative panels which cover satellite supporting structure inside thermal vacuum test chamber. In the Large Space Simulator (LSS) located in the ESA* ESTEC** Test Centre, there is the need to move the satellite specimen inside the vacuum chamber to modify its relative orientation vis-à-vis the artificial sun beam. Re-orientation of the radiative panels is then required. ESA is currently developing a New Motion System (NMS) allowing dynamic motion from Gimbal to the Yoke Stand without any reconfiguration of the test. To do so, a nitrogen supply is required which can follow the rotation of the thermal shrouds under vacuum. Nitrogen flexible hoses are not anymore suitable due to their volume and mass constraints and due to the limitation in the rotation range. This paper describes the design, manufacturing and tests of a new compact multi-turn Rotary Nitrogen Joint (RNJ) compliant with thermal vacuum conditions used for spacecraft thermal testing. This prototype development is funded by the ESA Technology and Research Program. *European Space Agency **European Space Research and Technology Centre.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

FY19 Transient Prescriptions in TREAT Full-Slotted Core

The transient reactor test facility (TREAT) is a graphite-based transient test reactor at the Idaho National Laboratory. TREAT started operation in 1959 and performed thousands of transient irradiations for the primary purpose of nuclear fuel safety research. TREAT operations were suspended in 1994 and then, several years later, resumed reactor operations in 2017 to support development of advanced nuclear fuels. During TREAT’s first era of operation the transient rod drive system and eventual Automatic Reactor Control System (ARCS) were upgraded and improved to create a transient reactor capable of unparalleled power shaping capabilities. TREAT’s first year of modern experiment work was performed in 2018 and saw the demonstration of new kinds of transients focused on modern experiment needs; especially focused on light water reactor (LWR) category transients including reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA). These transients were performed using a non-fueled neutronic mockup of a historical sodium loop which was not a perfect representation of future core configurations, but was adequate for demonstrating capabilities, successfully troubleshooting transient programming parameters, and calibrating nuclear models of the core. These physics tests were termed “transient prescription tests”. Later in 2018 the core was reconfigured to a modern fueled experiment arrangement and TREAT began performing irradiations on LWR rodlets where crucial fuel specimen nuclear heating data were obtained, and several supporting systems were commissioned. This successful effort gave way to additional fueled experiments which were performed throughout 2019.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Uncertainty Quantification of Calculated Temperatures for the AGR 5/6/7 Experiment

This report documents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation experiment conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technologies? research and development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR 5/6/7 capsules, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The calculated fuel temperatures for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. This report covers temperature uncertainty results for each of the five AGR 5/6/7 capsules. To quantify the uncertainty of calculated temperatures determined using the ABAQUS finite element heat transfer code, this study identifies and analyzes model parameters of potential importance to the calculated temperatures of fuel compacts and thermocouples. The selection of input parameters for uncertainty quantification is based on the ranking of their influences upon temperature predictions. Thus, selected input parameters include those with high sensitivity and those with the largest uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Measurement uncertainty, analysis of modeling assumptions, and expert judgment are used as the basis to quantify the uncertainty range for selected input parameters. The input uncertainties are dynamic, accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperatures and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish sufficiency of the time dependent first order (linear) expansion terms in constructing the temperature response surface. To achieve completeness, uncertainty propagation made use of pairwise noise correlations of model parameters. Furthermore, using an interpolation scheme over the input parameter domain, the analysis obtains time dependent sensitivity over the test campaign duration. This allows computation of uncertainty for the calculated fuel temperatures and the calculated graphite temperatures at thermocouple locations during the entire irradiation period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

An Architectural Survey of the Area 25 Nuclear Rocket Development Station, Nevada National Security Site, Nye County, Nevada, Volume 3 Of 3

The purpose of the current architectural survey is to record the Nuclear Rocket Development Station (NRDS) district, which is recommended eligible for listing to the NRHP under all four criteria as the testing center for the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) program from the initial boundary survey in 1956 to the end of nuclear rocket development on January 5, 1973. Resources in the NRDS district were evaluated to determine if they are eligible for listing to the National Register as district contributors. The current survey identified the Reactor Control Point subdistrict to provide a temporal, spatial, and functional framework for organizing the resources involved in the NRDS operations. However, the subdistrict was not evaluated as a potential historic district against the Secretary of Interior’s Significance Criteria.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2022

Large-scale parametric modeling of spent nuclear fuel dynamics in the 30 cm package drop scenario

Packages used to transport spent nuclear fuel (SNF) are required by the U.S. Code of Federal Regulations 10 CFR 71.71 to demonstrate satisfactory performance during a drop scenario. While the CFR is meant to ensure safe package function, it does not evaluate survival of the SNF within. The U.S. Department of Energy Spent Fuel and Waste Science and Technology program is working on closing the knowledge gap related to the response of SNF to external mechanical loads, including the hypothetical 30 cm package drop scenario in the CFR. In support of this effort, LS-DYNA finite element simulations were developed by Pacific Northwest National Laboratory (PNNL) to model generic drop scenarios at both the package and fuel assembly level. The models were validated against one-third scale package and full scale fuel assembly drop test data and were exercised to predict fuel cladding strains in a narrow range of model configurations. This work describes a large-scale parametric study conducted by PNNL using the previously developed and validated PWR finite element model, with the addition of a new generic BWR assembly model. The motivation for the parametric study was to characterize the broad range of SNF responses in the 30 cm package drop scenario. This was accomplished by varying the drop orientation, fuel assembly type (17x17 PWR and 10x10 BWR), burnup, cladding temperature, spacer grid buckling load, package mass, impact limiter stiffness, and mechanical gap conditions within the basket. A MATLAB framework was developed to automate LS-DYNA model generation and execution on PNNL institutional computing resources. In total, over 2000 simulations were performed. For each simulation, the SNF response was quantified in terms of permanent grid deformation, fuel rod contact pressure, and strains within the fuel rods, guide tubes, and water rods. The results provide valuable insight into the range of responses that could be reasonably expected from SNF in the hypothetical drop scenario, as well as the sensitivity to each input parameter. The results of this parametric study are a key component of the testing and modeling strategy the Spent Fuel and Waste Science and Technology program is using to close the external loads knowledge gap.

Kadooka, Kevin↗
DOE OSTI2021

Ground Motion Inputs for the Seismic Shake Table Test

Currently, spent nuclear fuel (SNF) is stored in on-site independent spent-fuel storage installations (ISFSIs) at seventythree (73) nuclear power plants (NPPs) in the US. Because a site for geologic repository for permanent disposal of SNF has not been constructed, the SNF will remain in dry storage significantly longer than planned. During this time, the ISFSIs, and potentially consolidated storage facilities, will experience earthquakes of different magnitudes. The dry storage systems are designed and licensed to withstand large seismic loads. When dry storage systems experience seismic loads, there are little data on the response of SNF assemblies contained within them. The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test to close the gap related to the seismic loads on the fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. The main component of the test unit will be the full-scale NUHOMS 32 PTH2 dry storage canister. The canister will be loaded with three surrogate fuel assemblies and twenty-nine dummy assemblies. Two dry storage configurations will be tested – horizontal and vertical above-ground concrete overpacks. These configurations cover 91% of the current dry storage configurations. The major input into the shake table test are the seismic excitations or the earthquake ground motions – acceleration time histories in two horizontal and one vertical direction that will be applied to the shake table surface during the tests. The shake table surface represents the top of the concrete pad on which a dry storage system is placed. The goal of the ground motion task is to develop the ground motions that would be representative of the range of seismotectonic and other conditions that any site in the Western US (WUS) or Central Eastern US (CEUS) might entail. This task is challenging because of the large number of the ISFSI sites, variety of seismotectonic and site conditions, and effects that soil amplification, soil-structure interaction, and pad flexibility may have on the ground motions.

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

A Nuclearly Safe Triangular-Slab Tank for Handling Enriched Uranium

In this study, an improved abrasive slurry tank for use with conventional liquid-honing machines was built and tested. Its shape, a 1.7-in.- (4.3-cm)-thick triangular slab, promotes good mixing by improving recirculation and is nuclearly safe for uranium enriched to 93% 235 U.

42 ENGINEERING↗
DOE OSTI2016

Hyper-Heuristics to Automatically Target Code to Computer Architectures

With each new generation of High Performance Computing (HPC) architecture, the gap between peak theoretical performance and the observed performance is growing. The goal of this research is to develop a tool to utilize hyper-heuristics to target code to a computational environment. In order to test this, sorting algorithms will be evolved on several different architectures. The final solutions will then migrate to all other architectures and their fitnesses compared. If the natively-evolved algorithms out-perform all others, then it can be concluded that the tool successfully targeted its solutions to the architecture of origin. This is the first step towards creating a program-agnostic tool for optimizing code to the native environment. The results are pending testing on the high performance cluster. If it can be shown that the tool is able to optimize solutions for the environment, then the door opens to automatically optimizing entire programs.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Use of EORTC Target Definition Guidelines for Dose-Intensified Salvage Radiation Therapy for Recurrent Prostate Cancer: Results of the Quality Assurance Program of the Randomized Trial SAKK 09/10

Purpose: Different international target volume delineation guidelines exist and different treatment techniques are available for salvage radiation therapy (RT) for recurrent prostate cancer, but less is known regarding their respective applicability in clinical practice. Methods and Materials: A randomized phase III trial testing 64 Gy vs 70 Gy salvage RT was accompanied by an intense quality assurance program including a site-specific and study-specific questionnaire and a dummy run (DR). Target volume delineation was performed according to the European Organisation for the Research and Treatment of Cancer guidelines, and a DR-based treatment plan was established for 70 Gy. Major and minor protocol deviations were noted, interobserver agreement of delineated target contours was assessed, and dose-volume histogram (DVH) parameters of different treatment techniques were compared. Results: Thirty European centers participated, 43% of which were using 3-dimensional conformal RT (3D-CRT), with the remaining centers using intensity modulated RT (IMRT) or volumetric modulated arc technique (VMAT). The first submitted version of the DR contained major deviations in 21 of 30 (70%) centers, mostly caused by inappropriately defined or lack of prostate bed (PB). All but 5 centers completed the DR successfully with their second submitted version. The interobserver agreement of the PB was moderate and was improved by the DR review, as indicated by an increased κ value (0.59 vs 0.55), mean sensitivity (0.64 vs 0.58), volume of total agreement (3.9 vs 3.3 cm{sup 3}), and decrease in the union volume (79.3 vs 84.2 cm{sup 3}). Rectal and bladder wall DVH parameters of IMRT and VMAT vs 3D-CRT plans were not significantly different. Conclusions: The interobserver agreement of PB delineation was moderate but was improved by the DR. Major deviations could be identified for the majority of centers. The DR has improved the acquaintance of the participating centers with the trial protocol.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2023

An Architectural Survey of the Area 25 Nuclear Rocket Development Station, Nevada National Security Site, Nye County, Nevada, Volume 2 Of 3

The purpose of the current architectural survey is to record the Nuclear Rocket Development Station (NRDS) district, which is recommended eligible for listing to the NRHP under all four criteria as the testing center for the Rover/NERVA (Nuclear Engine for Rocket Vehicle Application) program from the initial boundary survey in 1956 to the end of nuclear rocket development on January 5, 1973. Resources within the boundary of the NRDS district were evaluated to determine if they are eligible for listing to the National Register as district contributors. Subdistricts within the larger NRDS district were defined to provide a temporal, spatial, and functional framework for organizing the resources involved in the NRDS operations. However, developing contexts for the individual subdistricts was beyond the scope of the NRDS survey. Therefore, the potentially eligible subdistricts were not individually evaluated under any of the Secretary of Interior’s Significance Criteria for eligibility to the National Register.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2023

FAST-1.2 User Installation and Verification Guide: Developed Under NQA-1-2017

The purpose of this document is to provide the user information about the installation of Fuel Analysis for Steady state and Transient (FAST)-1.2 on their computers or servers. General information about the code and supported operating systems is described in Section 1.0. Self-service oriented FAST-1.2 software licensing steps are described in Section 2.0. An installation verification test suite is provided with FAST-1.2 and described in Section 3.0. A convenience script for converting Fuel Rod Analysis Program – Constant (FRAPCON) to FAST inputs is discussed in Section 4.0. FAST-1.2 was developed and released under a software quality assurance program based upon NQA-1-2017 at Pacific Northwest National Laboratory (PNNL). FAST-1.2 is the latest baseline code. The installation verification test suite contains both steady state and transient Anticipated Operation Occurrences (AOOs), accident conditions, such as Reactivity Initiated Accidents (RIAs) and Loss of Coolant Accidents (LOCAs).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗