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At least 1,765 records · Page 98
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Development of a U.S. Rail Transport Capability for Spent Nuclear Fuel and High-Level Waste - 16623

The U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) is laying the groundwork for implementing an integrated nuclear waste management system. This includes preparing for future large-scale transport of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) since transport will be a necessary component of an integrated waste management system. DOE continues to plan and develop options for the design of an integrated nuclear waste management disposition system. A significant destination-independent component of this integrated system is a rail transport capability for SNF and HLW. This paper provides an overview of the current DOE-NE effort related to designing and fabricating prototype rail cars. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2005

HSCT Sector Combustor Hardware Modifications for Improved Combustor Design

An alternative to the stepped-dome design for the lean premixed prevaporized (LPP) combustor has been developed. The new design uses the same premixer types as the stepped-dome design: integrated mixer flameholder (IMFH) tubes and a cyclone swirler pilot. The IMFH fuel system has been taken to a new level of development. Although the IMFH fuel system design developed in this Task is not intended to be engine-like hardware, it does have certain characteristics of engine hardware, including separate fuel circuits for each of the fuel stages. The four main stage fuel circuits are integrated into a single system which can be withdrawn from the combustor as a unit. Additionally, two new types of liner cooling have been designed. The resulting lean blowout data was found to correlate well with the Lefebvre parameter. As expected, CO and unburned hydrocarbons emissions were shown to have an approximately linear relationship, even though some scatter was present in the data, and the CO versus flame temperature data showed the typical cupped shape. Finally, the NOx emissions data was shown to agree well with a previously developed correlation based on emissions data from Configuration 3 tests performed at GEAE. The design variations of the cyclone swirler pilot that were investigated in this study did not significantly change the NOx emissions from the baseline design (GEAE Configuration 3) at supersonic cruise conditions.

Greenfield, Stuart C.↗
NASA NTRS2011

Hyperthermal Environments Simulator for Nuclear Rocket Engine Development

An arc-heater driven hyperthermal convective environments simulator was recently developed and commissioned for long duration hot hydrogen exposure of nuclear thermal rocket materials. This newly established non-nuclear testing capability uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce hightemperature pressurized hydrogen flows representative of nuclear reactor core environments, excepting radiation effects, and is intended to serve as a low-cost facility for supporting non-nuclear developmental testing of hightemperature fissile fuels and structural materials. The resulting reactor environments simulator represents a valuable addition to the available inventory of non-nuclear test facilities and is uniquely capable of investigating and characterizing candidate fuel/structural materials, improving associated processing/fabrication techniques, and simulating reactor thermal hydraulics. This paper summarizes facility design and engineering development efforts and reports baseline operational characteristics as determined from a series of performance mapping and long duration capability demonstration tests. Potential follow-on developmental strategies are also suggested in view of the technical and policy challenges ahead. Keywords: Nuclear Rocket Engine, Reactor Environments, Non-Nuclear Testing, Fissile Fuel Development.

Litchford, Ron J.↗
DOE OSTI2017

Analysis of maximum allowable fragment heights during dissolution of high flux isotope reactor fuel in an h-canyon dissolver

As part of the Spent Nuclear Fuel (SNF) processing campaign, H-Canyon is planning to begin dissolving High Flux Isotope Reactor (HFIR) fuel in late FY17 or early FY18. Each HFIR fuel core contains inner and outer fuel elements which were fabricated from uranium oxide (U 3 O 8 ) dispersed in a continuous Al phase using traditional powder metallurgy techniques. Fuels fabricated in this manner, like other SNF’s processed in H-Canyon, dissolve by the same general mechanisms with similar gas generation rates and the production of H 2 . The HFIR fuel cores will be dissolved using a flowsheet developed by the Savannah River National Laboratory (SRNL) in either the 6.4D or 6.1D dissolver using a unique insert. Multiple cores will be charged to the same dissolver solution maximizing the concentration of dissolved Al. The recovered U will be down-blended into low-enriched U for subsequent use as commercial reactor fuel. During the development of the HFIR fuel dissolution flowsheet, the cycle time for the initial core was estimated at 28 to 40 h. Once the cycle is complete, H-Canyon personnel will open the dissolver and probe the HFIR insert wells to determine the height of any fuel fragments which did not dissolve. Before the next core can be charged to the dissolver, an analysis of the potential for H 2 gas generation must show that the combined surface area of the fuel fragments and the subsequent core will not generate H 2 concentrations in the dissolver offgas which exceeds 60% of the lower flammability limit (LFL) of H 2 at 200 °C. The objective of this study is to identify the maximum fuel fragment height as a function of the Al concentration in the dissolving solution which will provide criteria for charging successive HFIR cores to an H-Canyon dissolver.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

New NDA Methods for Thorium Fuel Cycle Safeguards (Final Report)

This project developed portable Neutron Resonance Transmission Analysis (pNRTA) as a new non-destructive assay (NDA) method for thorium fuel cycles safeguards and other applications where multiple isotopes must be measured when present together. pNRTA leverages epithermal neutron resonances to assay multiple safeguards-relevant isotopes (e.g., 233 U and 235 U) when they are present together in a sample. Existing techniques are challenged by this task, driving the need for new active interrogation methods. With selected detectors, pNRTA works in high gamma-ray backgrounds from fission and activation products and 232 U progeny expected in thorium fuel cycle samples. This project leveraged a pNRTA system developed at Pacific Northwest National Laboratory (PNNL) and collaboration with the Massachusetts Institute of Technology (MIT). The system uses a commercially available deuterium-tritium (DT) neutron generator at short standoff (2 m). Key achievements in this project included: first-of-a-kind pNRTA quantitative measurements of 233 U oxide samples, an assessment of neutron detector technologies suitable for pNRTA in high gamma-ray background environments, experimentally demonstrating quantitative assay of samples containing 233 U and 235 U, and modeling studies showing the applicability of pNRTA to a wide range of material forms. Further, a custom algorithm was developed at MIT, which provided mean bias of 9% and relative standard deviation of 36% in assaying 233 U, 235 U, 238 U, and 232 Th content in eight measured samples. These outcomes form a solid technical basis for pNRTA as a new promising capability for international safeguards verification that is portable, non-destructive, quantitative, and isotopic specific.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Analytical local stress model for UMo/Al dispersion fuel

The stress evolution occurring in UMo/Al dispersion fuel is important since it affects the fuel performance during irradiation. In this study, a new analytical model was developed to predict the local stresses in UMo/Al dispersion fuel. In the model, a hypothetical unit sphere composed of a UMo fuel particle, interaction layer (IL), and Al matrix was considered, and the governing equations for the stress-strain relationship, strain-displacement, and mechanical equilibrium were established using a spherical coordinate system. The mathematical derivations were obtained for local stresses in the radial and circumferential direction using a thick-walled sphere model. Furthermore, this analytical model employed the stress distribution as the boundary condition, which is calculated using a finite element model with homogenized fuel meat. The developed model’s solution scheme was verified against Abaqus solutions obtained for two irradiated plates with heterogeneous meat. The model calculated consistent results for interfacial stresses in the IL and Al matrix, indicating that the newly developed model was reliable when simultaneously calculating fission gas pressurization on the UMo/IL/Al composite, and the use of the stress distribution in the homogenized fuel meat as the boundary condition for the analytical model was acceptable

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1983

The study of integrated coal-gasifier molten carbonate fuel cell systems

A novel integration concept for a coal-fueled coal gasifier-molten carbonate fuel cell power plant was studied. Effort focused on determining the efficiency potential of the concept, design, and development requirements of the processes in order to achieve the efficiency. The concept incorporates a methane producing catalytic gasifier of the type previously under development by Exxon Research and Development Corp., a reforming molten carbonate fuel cell power section of the type currently under development by United Technologies Corp., and a gasifier-fuel cell recycle loop. The concept utilizes the fuel cell waste heat, in the form of hydrogen and carbon monoxide, to generate additional fuel in the coal gasifier, thereby eliminating the use of both an O2 plant and a stream bottoming cycle from the power plant. The concept has the potential for achieving coal-pile-to-busbar efficiencies of 50-59%, depending on the process configuration and degree of process configuration and degree of process development requirements. This is significantly higher than any previously reported gasifier-molten carbonate fuel cell system.

Source record↗
DOE OSTI2016

CAMPUS: Fully Coupled Multiphysics Modeling and Simulation of LWR and FBR Nuclear Fuel Behavior

Nuclear fuels operate in an environment that involves complex multiphysics phenomena, and the multiphysics behavior is often tightly coupled. The prediction of fuel behavior under irradiation involves very complex material behavior that is not seen in any other engineering discipline. Understanding and predicting the evolving material properties and phenomenal behaviors of nuclear fuels are major challenges for fuel design and behavior prediction. The multiphysics behavior is often a tightly coupled thermal, mechanical and chemical phenomenon associated with a large number of control parameters and uncertainties, which requires an advanced fuel performance modeling and simulation capability to improve the existing reactors performance, to address environmental and reactor safety concerns. Nuclear fuel performance modeling and simulation capability is necessary as a result of the high cost and difficulty in experimental measurements. So, a fuel performance code called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed to predict both light water reactor and fast reactor fuel behavior. An effective approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with light water reactor and fast reactor fuels performance analysis. The modelling and simulations approach is of particular importance since it demonstrates highly interrelated and nonlinear fuel performance models can be developed and implemented into a multiphysics platform, and thus more efforts can be put into physics model development and improvement. The modelling results can be used to identify the important phenomena and parameters, to direct and design experiments and to assist in post analysis of expensive nuclear experiments for light water reactor and fast reactor nuclear fuel performance analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Lightweight Solid Oxide Cell Design for Ammonia-fed SOFC Systems

A lightweight solid oxide fuel cell (SOFC) was developed by OxEon Energy that demonstrates the potential for increased specific power (kW/g) suitable for the requirements of electric Vertical Take-off and Landing (eVTOL) applications. Fuel flexibility was demonstrated in button cells and stacks using both legacy electrolyte-supported SOFC structure and the lightweight structure. Button cell and stack performance was nearly identical among hydrogen, ammonia, and reformed natural gas. Durability testing shows similarly stable performance in each fuel. A button cell test in ammonia at 800 °C using the lightweight cell structure measured area specific resistance (ASR) of 0.25 Ω-cm2 in H2/N2 and NH3 fuel feeds and showed a peak power density of 0.98 W/ cm2 in NH3 fuel. Short term hold at 0.7 V showed stable performance.

Pike, Jenna↗
DOE OSTI2022

Reversible Fuel Cell Stacks with Integrated Water Management

Regenerative fuel cell/electrolyzers have long held promise as energy storage devices. Due to their small footprint and their ability to separately size the power and storage functions, they have many advantages over batteries, particularly for remote or deployable systems. A unitized regenerative fuel cell is proposed that takes advantage of three new technologies. The first is a H 2 /O 2 unitized regenerative stack that Giner, Inc. (Giner) has developed that combines a fuel cell and electrolyzer stack, eliminating one of the electrochemical stacks, the most expensive component in regenerative fuel cell systems. The second is an oxygen catalyst with low PGM loadings generated through Reactive Spray Deposition Technology (RSDT). The third is an amphiphilic microporous layer (AMPL) structure to greatly aid water management in both operational modes.

08 HYDROGEN↗
NASA NTRS1975

Hypersonic research engine project. Phase 2: Preliminary report on the performance of the HRE/AIM at Mach 6

Test results of the Aerothermodynamic Integration Model are presented. A program was initiated to develop a hydrogen-fueled research-oriented scramjet for operation between Mach 3 and 8. The primary objectives were to investigate the internal aerothermodynamic characteristics of the engine, to provide realistic design parameters for future hypersonic engine development as well as to evaluate the ground test facility and testing techniques. The engine was tested at the NASA hypersonic tunnel facility with synthetic air at Mach 5, 6, and 7. The hydrogen fuel was heated up to 1500 R prior to injection to simulate a regeneratively cooled system. The engine and component performance at Mach 6 is reported. Inlet performance compared very well both with theory and with subscale model tests. Combustor efficiencies up to 95 percent were attained at an equivalence ratio of unity. Nozzle performance was lower than expected. The overall engine performance was computed using two different methods. The performance was also compared with test data from other sources.

Sun, Y. H.↗
DOE OSTI2021

Development of MCNP Training Modules for Safeguards Practitioners [Abstract]

The Monte-Carlo N-Particle (MCNP) software developed at LANL is the most widely used neutron transport code in the world. It is an essential tool for a variety of applications including detector development and design, nuclear fuel burnup simulation, criticality safety, and nondestructive assay system optimization. For this reason, it is indispensable within the safeguards and materials control & accountability (MC&A) communities. Multiple MCNP training courses have been created and taught over the last several decades by the MCNP development team at LANL, however there are no existing courses that cover specialized topics considered fundamental to NDA and safeguards models. To fill this gap, the MCNP team and Safeguards Science and Technology group at LANL have co-created a set of training modules customized to meet the specialized needs of the safeguards and MC&A communities. The basic modules cover concepts such as NDA system optimization, He-specific and other capture tallies, and tools for improved theoretical understanding. An advanced module was also created to cover topics including variance reduction for active interrogation simulations, use of the LANL MCNPTools post-processor, PTRAC (particle tracking) and list-mode data simulations, and fuel burnup simulations. The training modules teach to the latest and most state-of-the-art MCNP features and tools released by the development team at LANL and are intended to be taught jointly by the developers and safeguards experts. Ultimately, we hope that creation of these modules will serve to capture and convey the safeguards modeling and MCNP expertise at LANL, and that we will be able to share the modules more broadly with the MC&A and safeguards communities.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Fuel Performance Modeling Plan to Support the Advance Gas Reactor Program

This report documents the current status of the fuel performance modeling initiative to support the Advanced Gas Reactor (AGR) program in the development of tristructural isotropic (TRISO) coated fuel particles. A brief summary of the modeling codes that have been developed to support TRISO modeling is included along with a summary of the behavior of fuel particles during irradiation and the modeling used to capture these effects. In addition, this report identifies further modeling and material property needs that have been identified for further development based on experience from previously performed AGR experiments. In general, the remaining activities to support fuel performance modeling for the AGR program include continued AGR experiment support for AGR-3/4 and AGR-5/6/7 as well as modeling improvements identified throughout the course of the program. These modeling improvements can be summarized as the following: Fission product transport Thermomechanical particle behavior Additional modeling needs may be identified during the processing the data collected during the AGR post-irradiation examination (PIE) campaign and may lead to further modeling improvements that are not included in this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Characterizing surrogates to develop an additive manufacturing process for U 3 Si 2 nuclear fuel

U 3 Si 2 is considered as an accident-tolerant nuclear fuel candidate due to its promising high uranium density along with its improved thermal properties. The current U 3 Si 2 powder metallurgical fabrication process encompasses numerous process steps with extensive fabrication times. Thus, an additive manufacturing process to fabricate U 3 Si 2 is currently under development to ultimately produce U 3 Si 2 fuel at a lower cost, in a timely and commercially-reliable manner. The current research work seeks to characterize U 3 Si 2 surrogate materials fabricated by a laser synthesis. The characterization results provided novel microstructural data that aided in the early development stages of an additive manufacturing process. For this study, cerium, zirconium, and hafnium, were selected as surrogates based on their thermodynamic properties and crystal structure. As a result, the laser-synthesized samples were characterized by means of scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS) along with X-ray diffraction (XRD), where the different phases of formation, morphology, and microstructural features have been analyzed. Laser synthesis showed suitability for alloying Zr-Si and Hf-Si to ultimately form silicide compounds (Zr 3 Si 2 and Hf 3 Si 2 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Design and Development of a High-Efficiency Single Cylinder Natural Gas-Fueled Jet Ignition Engine

The current energy climate has created a push toward reducing consumption of fossil fuels and lowering emissions output in power generation applications. Combined with the desire for a more distributed energy grid, there is currently a need for small displacement, high efficiency engines for use in stationary power generation. An enabling technology for achieving high efficiencies with spark ignited engines for such applications is the use of jet ignition which enables ultra-lean (λ > ~1.6) combustion via air dilution.This paper provides a comprehensive review of the development of a 390cc, high efficiency single cylinder natural gas-fueled jet ignition engine operating ultra-lean. The engine was developed as part of the Department of Energy’s Advanced Research Projects Agency–Energy (DOE ARPA-E) GENSETS program. Design choices for minimizing friction are highlighted as well as test results showing further friction reduction through downspeeding. Extensive hardware optimization of the combustion system has been performed and results are presented for air-flow path optimization and the jet igniter. The efficiency benefits related to enleanment and downspeeding are analyzed using an efficiency loss breakdown based on the First Law of Thermodynamics. Through optimization efforts a peak brake thermal efficiency in excess of 34% was achieved, representing an increase of greater than 20% over the current state-of-the-industry for comparably sized CNG engines.

Peters, Nathan↗
DOE OSTI2022

Grain growth kinetics of the gamma phase metallic uranium

We report metallic uranium is a leading fuel form for sodium cooled fast reactors as an enabling technology of future nuclear energy systems. Mechanistic understanding of fuel behaviors and kinetics under thermodynamic equilibrium and highly non-equilibrium conditions are essential for evaluating fuel performance. It is important to understand and predict the grain and pore evolutions of metallic fuels under thermal and irradiation conditions. However, very limited data are available on the grain growth kinetics and mechanisms of pure gamma phase uranium. In this paper, the pure gamma uranium pellets with different grain structures were fabricated by combining high-energy ball milling and spark plasma sintering. Isothermal annealing tests were performed to investigate the grain growth behavior of the pure gamma phase uranium with different initial grain sizes. A parabolic relationship in grain growth with time was identified for the submicron-sized (374 nm) sample. In contrast, for the nano-sized (137 nm) sample, the grain growth shows a linear relationship with time. The activation energies of grain growth were determined as 199.5 KJ/mol and 80.6 KJ/mol for nano-sized and submicron-sized grain structures, respectively. For the nano-sized sample, the rate-control step of grain growth is dominated by the triple-junction migration, in which the grain boundary triple junction drags the grain growth, leading to a higher activation energy than the bulk diffusion. The dominating mechanism for the submicron-sized sample is grain boundary diffusion. The mechanistic understanding and critical data obtained on the kinetics of pure uranium phases will be useful to evaluate fuel behavior under thermodynamic equilibrium conditions and develop a high fidelity model to predict fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Technical Assistance in Fabrication and Testing of MEAs: Cooperative Research and Development Final Report

Altergy's fuel cell cost is highly dependent on expensive purchased membrane electrode assemblies (MEAs), which amount to 42¢/watt in a 5kW system. In order to better compete with other power solutions, the company has identified an MEA breakthrough cost of 20¢/watt to unlock new, major markets for its fuel cell products. Since Altergy presently does not have an accurate cost model for the real costs of these MEAs, it is difficult to know whether the company goal is achievable. Altergy seeks to leverage NREL expertise to reduce cost and improve performance for Altergy's fuel cell products. The objective of this Small Business Voucher (SBV) project therefore will be for NREL and Altergy to work together in three major task areas: 1) demonstrating the fabrication of MEAs at NREL using NREL non-proprietary methods, to the intent of potentially fabricating MEAs at Altergy to reduce product costs; and, 2) creating a cost model to assist Altergy in manufacturing decisions.

30 DIRECT ENERGY CONVERSION↗