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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 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Full Scale Engine Demonstration of Additively Manufactured High Gamma Prime Turbine Blade

Additive manufacturing (AM), also known as 3D printing, is a rapidly developing technology with tremendous potential in both developmental and production applications. Solar Turbines Incorporated is committed to AM technology development for gas turbine applications. The ability to metal 3D print novel designs of turbine blades capable of actual turbine engine operation would effectively reduce design validation cycle time and allow acquisition of key performance data early in a design campaign. In support of Solar’s advanced manufacturing development and ongoing engine efficiency improvement goals, ORNL and Solar conducted a project to 3D print, machine, inspect, and engine test a full set of Mercury TM 50 stage 2 turbine blades. This full-scale, hot-fired engine demonstration has identified both the capabilities and limitations of the current technologies required to produce turbine hot-section AM components.

36 MATERIALS SCIENCE↗
DOE OSTI2022

IER-517: Molybdenum Optimized Benchmark System Demonstrating Integral Correlations (MOBY DICK)

Nuclear criticality experiments are essential to the validation of nuclear data used in simulation software. The quality of nuclear data becomes paramount as simulation software becomes more relied upon for criticality safety studies and designs of nuclear systems. To improve the quality of nuclear data, experimenters can design critical experiments that are sensitive to isotope reaction pairs in materials of interest. The efforts conducted by the Organisation for Economic Co-operation and Development - Nuclear Energy Agency (OECD-NEA) Working Party on Nuclear Criticality Safety (WPNCS) Subgroup 8: Preservation of Expert Knowledge and Judgement Applied to Criticality Benchmarks (SG8) to categorize benchmarks according to their usefulness for nuclear data validation have been of great importance. Based on the OECD studies benchmark experiments included in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook are concisely used by nuclear data evaluators, criticality safety engineers and others to validate nuclear data and simulation results. A lack of benchmarks sensitive to molybdenum in the (ICSBEP), particularly in the intermediate range, was noted by Los Alamos National Laboratory (LANL), the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), and Y-12 National Security Site prompting them to submit a joint integral experiment request to the Nuclear Criticality Safety Program (NCSP) in 2019. The request included both HEU and Plutonium systems in order to validate differential nuclear data focusing on the intermediate energy range but also includes thermal and fast configurations. This document represents the preliminary design work for a series of molybdenum integral experiments known as Molybdenum Optimized Benchmark System Demonstrating Integral Correlations (MOBY DICK).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Radioactive Waste Sludge Washing and Demonstration of the Nitric-Glycolic Acid Flowsheet for Sludge Batch 10 Qualification

For each sludge batch that is processed in the Defense Waste Processing Facility (DWPF), the Savannah River National Laboratory (SRNL) performs qualification testing to demonstrate that the sludge batch (SB) is processible. During processing of SB9, DWPF will be transitioning from the Nitric-Formic Acid (NFA) flowsheet to the Nitric-Glycolic Acid (NGA) flowsheet. Thus, the qualification of SB10 was requested to only be performed using the NGA flowsheet. In order to qualify the batch for the NGA flowsheet, Sludge Receipt and Adjustment Tank (SRAT) and Slurry Mix Evaporator (SME) cycles, designated SC-19, were performed using SB10 Tank 51 sample material. SRNL received Tank 51 material in the midst of Tank Farm washing. SRNL continued the washing in the SRNL Shielded Cells. The SRNL process included the addition of Sodium Reactor Experiment (SRE) material from H Canyon, simulating the transfer of SRE from H Canyon to Tank 51 during Tank Farm washing. The washed SB10 Tank 51 material, with SRE, was characterized prior to flowsheet qualification testing.

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

Designing Resilient Communities: Hardware demonstration of resilience nodes concept

As part of the project “Designing Resilient Communities (DRC): A Consequence-Based Approach for Grid Investment,” funded by the United States (US) Department of Energy’s (DOE) Grid Modernization Laboratory Consortium (GMLC), Sandia National Laboratories (Sandia) is partnering with a variety of government, industry, and university participants to develop and test a framework for community resilience planning focused on modernization of the electric grid. This report provides a summary of the section of the project focused on hardware demonstration of “resilience nodes” concept.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2022

Report on Fuel Cycle Facility Requirements for Deployment of Demonstration Reactors and Potential Evolutionary Fuel Cycle Scenarios

A series of fuel cycle scenarios studies were performed to inform on fuel cycle capacities and facilities needed for large-scale deployment of the Advanced Reactor Demonstration Program (ARDP) reactors and potential future evolutionary fuel cycle scenarios. The reactor deployment and evolutionary fuel cycle scenarios from the present to 2100 were developed based on the following assumptions: 1) achievement of a net-zero emissions economy in the United States by 2050, which requires a nuclear energy generation capacity of ~250 GWe by 2050, 2) the U.S. economic growth of 1% per year from 2051 to 2100, which results in ~340 GWe of nuclear energy capacity in 2100, and 3) commercial-scale recycling and high burnup fuel technologies are available after 2050. Thus, evolutionary fuel cycles with those advanced nuclear technologies start after 2050. A single once-through fuel cycle scenario was assumed from the present to 2050 to achieve a net-zero emissions economy in the United States, and the following four evolutionary fuel cycle scenarios from 2051 to 2100 were considered, 1) Once through fuel cycle with ARDP reactors (Natrium and Xe-100), 2) Once-through fuel cycle with Breed-and-Burn (B&B) fast reactors, 3) Recycling fuel cycle of used metallic fuel in fast reactors, and 4) Recycle fuel cycle of both used uranium oxide and metallic fuels in fast reactors. The projected front-end and back-end fuel cycle capacity demands are compared with the current domestic and global (if needed) fuel cycle capacities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Demonstration of Improved CHP Systems utilizing Improved Gas Turbine and sCO2 Cycles Using Additive Manufacturing (AM) Components

This project seeks to integrate a sCO2 bottoming cycle with a 5.3 MW gas turbine to develop a CHP system that is able to transition rapidly between 50% and 100% load by engaging or bypassing the bottoming cycle while maintaining electrical system efficiency above 30% at all times. The project team will engineer an optimized design of the power systems. The needed new key components, such as advanced heat exchangers and a steam injection system for the gas turbine, will be developed and their performance demonstrated in rig tests. The feasibility of the whole integrated CHP system will be validated through a cyber-physical approach.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Demonstration of Electron Cooling using a Pulsed Beam from an Electrostatic Electron Cooler

Electron cooling continues to be an invaluable technique to reduce and maintain the emittance in hadron storage rings in cases where stochastic cooling is inefficient and radiative cooling is negligible. Extending the energy range of electron coolers beyond what is feasible with the conventional, electrostatic approach necessitates the use of RF fields for acceleration and, thus, a bunched electron beam. To experimentally investigate how the relative time structure of the two beams affects the cooling properties, we have set up a pulsed-beam cooling device by adding a synchronized pulsing circuit to the conventional electron source of the CSRm cooler at Institute of Modern Physics *. We show the effect of the electron bunch length and longitudinal ion focusing strength on the temporal evolution of the longitudinal and transverse ion beam profile and demonstrate the detrimental effect of timing jitter as predicted by theory and simulations. Compared to actual RF-based coolers, the simplicity and flexibility of our setup will facilitate further investigations of specific aspects of bunched cooling such as synchro-betatron coupling and phase dithering.

Bruker, M. W.↗
DOE OSTI2022

CFD-Chemical Model of One-Third Scale Demonstration of DOE Sealed Canister with ATR Fuel

Road-ready and final disposition packaging configurations for the advanced test reactor (ATR) fuel currently specifies storage within helium backfilled DOE sealed standard canisters. The aluminum cladding of the ATR fuel contains an oxyhydroxide layer of boehmite/bayerite that generates hydrogen when subjected to irradiation. Understanding the effect of this hydrogen buildup over time to important for long term storage considerations. Previous modeling efforts have built a coupled CFD-chemical model to simulate the temperature gas phase concentrations within the DOE sealed standard canisters. A demonstration case for these DOE sealed canisters will be eventually performed with a one-third scale mockup that has been instrumented with thermocouples and gas concentration probes. This study seeks to model the planned canister for validation of the previously developed model, such that confidence in its long-term prediction can be increased. The deployment of the instrumented lid for online monitoring in intended for a period of 10+ years based on previous monitoring of commercial fuel storage; however, the model is still run for the previously used 50-year storage periods. This case is modeled with a G-value using a bi-linear function such that it decreases at higher dose rates to be consistent with experiments. Validation efforts of the model would likely revolve around the results for the 1st year, so the results of this timeframe are also highlighted. For dried fuel of the nominal decay heat (18W), the predicted hydrogen concentration is 0.19% after 1 year, 0.91% after 10 years, and 2.8% after 50 years, with a maximum pressure of 1.26 atm. Consistent with previous modeling, the decay heat of the fuel is the main factor that influences the results. For undried fuel in pure helium, the hydrogen concentration ranges from 0.23 to 1.2% after 1 year, 1.25-6.2% after 10 years and 8.8 to 18.84% after 50 years. For dried fuel in pure helium the hydrogen concentration ranges from 0.06% to 0.34% after 1 year, 0.35-2.11% after 10 years and 0.95 to 6.6% after 50 years. While long-term results in the presence of residual air are mostly the same, early reactions with O 2 can delay significant production of H 2 until it is consumed to form more water vapor, lowering the range to 0.14 to 1.0% after 1 year. The maximum absolute pressure that is reached across any scenario is 2.06 atm. In the event of residual air, the presence of nitric acid is possible in the range of 18-131 ppm after 1 year, 174-1180 ppm after 10 years, and 586-3500 ppm after 50 years. As long as the fuel is sufficiently dried, or of nominal decay heat, a 4% lower flammability limit of hydrogen will not be reached within a 10-year monitoring period.

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

Demonstration and Evaluation of the Human-Technology Integration Guidance for Plant Modernization

The significance of nuclear power in its role producing carbon-free electricity to the U.S. cannot be overstated. However, with changes in the energy market coupled with changes in incentives given to certain resources like solar and wind, the operating and maintenance costs for these sources have seen a significant reduction, which has consequently negatively impacted the economic viability of the existing U.S. nuclear power plant fleet. Digital instrumentation and control (I&C) and control room modernization is a major critical work domain to reduce operating and maintenance costs. Existing nuclear power plants are commonly configured with mostly legacy analog I&C as well as isolated pockets of digital I&C (a plant process computer, digital recorders, etc.). One challenge with this analog I&C is that replacement parts are becoming prohibitively more expensive and difficult to obtain. Moreover, a significant challenge with the existing analog I&C is that the way in which plants are currently operated and maintained is no longer competitive with other electricity generating sources, like natural gas, where advanced digital I&C technologies are commonplace. This gap between the waynuclear power plants are operated compared to other electricity generating sources significantly challenges the economic viability of the nuclear industry. Indeed, digital I&C systems can fundamentally change the way the plant is operated (i.e., the concept of operation). The introduction of digital I&C technologies offers a wealth of benefits to the nuclear industry. However, it is important to emphasize that, to realize these benefits, a careful understanding of how to integrate technology in a collaborative way that leverages the capabilities of people and technologies is necessary. Human-technology integration applies human factors engineering methods and tools to ensure the safe and reliable use of these technologies while ensuring that the inherent features of the technologies that provide economic value are not missed. The scope of this work documents the demonstration of the recently developed human and technology integration methodology, as applied to developing a new vision and concept of operations for a major U.S. nuclear power plant fleet. This report focuses on the methodological aspects of developing a vision and concept of operations. This report shares the tools, activities, and lessons learned during a modernization currently underway for industry as a whole to consider when planning any significant digital modification and developing a new vision and concept of operations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Demonstration of Niobium Tin in 218 MHz Low-beta Quarter Wave Accelerator C

A 218 MHz quarter wave niobium cavity has been fabricated for the purpose of demonstrating Nb3Sn technology on a low-beta accelerator cavity. Niobium-tin has been established as a promising next generation SRF material, but development has focused primarily in high- beta elliptical cell cavities. This material has a significantly higher TC than niobium, allowing for design of higher frequency quarter wave cavities (that are subsequently smaller) as well as for significantly lowered cooling requirements (possibly leading to cryocooler based designs). The fabrication, initial cold testing, and Nb3Sn coating are discussed as well as test plans and details of future applications.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Demonstrating and Expanding DRiFT’s Helium-3 Detector Response Capabilities [Slides]

Helium-3 tubes are an efficient way to detect neutrons after they have thermalized by polyethylene. They are often arranged in rings to perform coincidence counting. Applications include nuclear safeguards, treaty verification, and emergency response. DRiFT has the ability to simulate pile-up, room return, and end effects. A new feature determines real and accidental correlated counts. The next steps include DRiFT 2.0 released as executable, continued demonstration of gas module capabilities, and semiconductor module under a development.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Advanced Systems Level Technology Development, Integration, and Demonstration for Efficient Class 8 Trucks (SuperTruck II)

The Cummins SuperTruck II program has completed the planned technical work on the project. This work includes Demonstrating 50% BTE on the engine-only, 55% BTE for an engine with waste heat recovery, and 125% improvement in vehicle freight efficiency. The engine system included a high efficiency diesel engine integrated with a state-of-the-art waste heat recovery (WHR) system and an advanced aftertreatment system capable of meeting the current emissions standards.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Modeling Enhancements and Demonstration of Shift Capabilities for PBRs and MSRs

This technical report documents the modeling enhancements and demonstrations with the Shift Monte Carlo (MC) code targeted at pebble-bed reactors (PBRs) and molten salt reactors (MSRs) under the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in fiscal year (FY) 2023. The work performed included several enhancements, such as improvements for multigroup cross section generation, a new eigenvalue mode considering only prompt fission neutrons, and enhancements to the Titan frontend for Shift to allow for new geometry types and tally functionality. Additionally, new PBR equilibrium core search reference calculations were generated with Shift and compared to Serpent calculations provided by Idaho National Laboratory (INL). These enhancements provide a robust foundation for applying Shift for both reference and two-step neutronics analysis for advanced reactor simulation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Demonstrating Safeguards Applications of DRiFT Gas Detector Capabilities [Slides]

DRiFT, a Detector Response Function Toolkit, post-processes MCNP output to model detector response. Three primary components: scintillator (focus of FY 21 TED work), gas detectors (focus of this FY), and semiconductor detectors. Capability to model scintillator detectors has previously been demonstrated.

47 OTHER INSTRUMENTATION↗
DOE OSTI2023

Thermal and Deposition Modeling Update for the Canister Deposition Field Demonstration

This report provides an update on the thermal and deposition modeling being performed in support of the Canister Deposition Field Demonstration (CDFD). The goal of the CDFD testing is to collect deposition measurements on the surface of the dry shielded canister (DSC) to aid in chloride-induced stress corrosion cracking (CISCC) research. Ideally, this testing will occur in a marine coastal environment, but at the time of this report, no official site has been selected. The CDFD plans to use the NUHOMS® Advanced Horizontal Storage Module – High Seismic (AHSM-HS) with a 32PTH2 canister. The spent nuclear fuel (SNF) assemblies will be mimicked with electrical heaters. Canisters are currently being outfitted with these heaters and tested at Sandia National Laboratories (SNL). Thermal modeling was conducted to ensure the electrical heaters are representative of the temperature distribution found within a storage system with SNF. Deposition models were also built to evaluate particulate deposition on SNF canisters. A preliminary site model was constructed to support a planned upcoming ambient aerosol characterization campaign.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

EXPERIMENTAL DEMONSTRATION OF A STRAIGHT-MERGER BEAMLINE

Beamlines capable of merging beams with different energies are critical to many applications related to advanced accelerator concepts and energy recovery linacs (ERLs). A straight merger system composed of a transverse deflecting cavity with a superimposed dipole magnet was proposed and recently tested at the Argonne Wakefield Accelerator. This paper reports on the experimental demonstration of the concept along with detailed beam dynamics investigations of the merger concept and its ability to conserve the beam brightness associated with the fresh bunch. Possible applications to ERL and beam-driven wakefield accelerators are also discussed.

Phillips, C.↗
DOE OSTI2023

Experimental Demonstration of Alkalinity Concentration Swing for Direct Air Capture of CO 2

This work constitutes the first lab-scale experimental demonstration of the alkalinity concentration swing (ACS) for direct air capture (DAC) of CO 2 . ACS can be implemented by taking a dilute alkaline solution that has contacted ambient air, and then concentrating it using commercially available desalination technologies. The more concentrated solution has a higher partial pressure of CO 2 relative to the initial solution, allowing the CO 2 to be separated and stored. The ACS approach is based on simple thermodynamics but had never been verified experimentally in terms of extracting CO 2 from air until now. We tested two distinct methods for concentrating the alkaline liquid: reverse osmosis (RO) and capacitive deionization (CDI). We found that to complete the DAC cycle in feasible operating conditions, a bicarbonate-enrichment step needed to be introduced, which selects on bicarbonate ions over carbonate ions. Two approaches for CO 2 extraction to measure the CO 2 yield per amount of water processed by the ACS method were tested. The energy required per amount of CO 2 captured was quantified. We further applied an energy model to study how operating the ACS cycle under different conditions would affect energy requirements.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Non-LWR Fuel Cycle Scenarios for SCALE and MELCOR Modeling Capability Demonstration

To assess the modeling and simulation capabilities for radionuclide characterization, criticality, and shielding in the nuclear fuel cycle of non-light–water reactor (LWR) technologies, the US Nuclear Regulatory Commission (NRC) initiated a collaborative project between the NRC, Sandia National Laboratories (SNL), and Oak Ridge National Laboratory (ORNL) with the goal to demonstrate capabilities of MELCOR and SCALE to calculate accident scenarios in all stages of the nuclear fuel cycle for relevant non-LWRs. The first project task was to develop representative nuclear fuel cycles and identify potential hazards and accident scenarios in the individual fuel cycle stages based on publicly available information. Because the nuclear fuel cycle is not established for any non-LWR concept, many assumptions were made, and it is anticipated that the details of the fuel cycles will eventually look different.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗