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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 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Modeling diurnal and annual ethylene generation from solar-driven electrochemical CO 2 reduction devices

Integrated solar fuels devices for CO 2 reduction (CO 2 R) are a promising technology class towards reducing carbon emissions. Designing integrated CO 2 R solar fuels devices requires careful co-design of electrochemical and photovoltaic components as well as consideration of the diurnal and seasonal effects of solar irradiance, temperature, and other meteorological factors expected for ‘on-sun’ deployment. Using a photovoltaic-electrochemical (PV-EC) platform, we developed a temperature and potential-dependent diurnal and annual model using experimentally-determined CO 2 R performance of Cu-based electrocatalysts, local meteorological data from the National Solar Radiation Database (NSRD), and modeled performance of commercial c-Si PVs. Here we simulated gaseous diurnal product outputs with and without the effects of ambient temperature. From these outputs, we observed seasonal variation in gaseous product generation, with up to two-fold increases in ethylene productivity between the Winter and Summer, analyzed the consequences of dynamic cloud coverage, and identified periods where device cooling/heating mechanisms could be implemented to maximize ethylene generation. Finally, we modeled the annual ethylene generation for a scaled 1 MW solar farm at three different locations (Beijing, CN; Sydney, AUS; Barstow, CA) to determine the consequences of local meteorological climates on PV-EC CO 2 R product output, recording a maximum ethylene output of 18.5 tonne per year at Barstow. Overall, this model presents a critical tool for streamlining the translation of experimental solar-driven electrochemical research to real-world implementation.

14 SOLAR ENERGY↗
DOE OSTI2022

A Structural Design Approach Tailored for the Rapid Preliminary Design of Microreactor Components

High-temperature microreactors can play a role in developing reliable, portable energy sources for off-grid remote locations, microgrid concepts, and industrial process heat. Portability and passive safety criteria tend to skew microreactor structural component designs toward complex geometries, high thermal stresses, and design bases with large numbers of startup/shutdown cycles. Current design rules, as typified by Section III of the American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel Code, are less than optimal for these conditions, particularly for preliminary component designs where developers need to rapidly consider a large number of potential component configurations. This paper presents a design method targeted toward rapid, efficient evaluation of preliminary component designs using modern finite element analysis. The new method retains key connections with the ASME Code rules and design data while streamlining the design approach. This paper presents the design method, several verification examples illustrating the similarities and differences between the new method and the current ASME rules, and the application of the new approach to the evaluation of a test article mimicking key features of a heat pipe–cooled microreactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Rapid prediction of fuel research octane number and octane sensitivity using the AFIDA constant-volume combustion chamber

Current research octane number (RON) and motor octane number (MON) gasoline performance characterization techniques use dated, complex engine testing methodology and limit researchers’ ability to easily characterize small volumes of experimental fuels. A novel methodology is presented that correlates measured ignition delay (ID) time to RON in an Advanced Fuel Ignition Delay Analyzer (AFIDA) constant-volume combustion chamber device at a single pressure/temperature condition, with an r2 of 0.99 and standard error (SE) of 1.0. The correlation of the slope of the ID time between two additional temperature points to octane sensitivity (S) produces an r2 of 0.97 and SE of 0.69; however, fuels with S>12 are indistinguishable. These results are based on methodology calibration using 31 primary and toluene reference fuels containing 0%-40% ethanol with RON values ranging from 85 to 113. Validation of these methods using a 102-sample fuel matrix spanning an array of base fuels and additive chemistry designed to test the robust applicability of the method, along with pump gasoline and high-octane surrogate blend samples, demonstrates an r2 of 0.94 and SE of 1.3 for the RON correlation over all samples, whereas the equivalent S correlation produces an r2 of 0.78 and SE of 1.2 by excluding two additives, 3-pentanone and diisobutylene, which displayed poor S correlation results. This novel AFIDA analysis method can be performed in 1 h and with 40 mL of fuel, offering significant improvements in time and volume requirements over traditional techniques.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2024

Micro-tensile Properties of Fueled Irradiated AGR-2 TRISO-coated Particle Buffer, IPyC, and SiC Interlayer Regions

Tristructural isotropic (TRISO) coated nuclear fuel particles are proving to be a versatile fuel form for new reactor designs. Understanding the bounding strength and failure mode of each coating interface is important to both fuel quality evaluation and failure prediction. A mechanism of key significance is failure of the silicon carbide (SiC) layer to retain fission products due to incomplete tearing of the buffer layer. This is a two-step mechanism involving both mechanical failure in the buffer and inner pyrolytic carbon (IPyC) layers and degradation of the SiC layer through palladium silicides at the IPyC-SiC interface. However, the mechanical properties of TRISO particle coating layers have yet to be fully characterized due to the small dimension of TRISO fuel particles and high radioactivity. To investigate this mechanism, in situ micro-tensile properties of the buffer, IPyC, SiC, buffer-IPyC, and IPyC-SiC interlayer regions of fueled TRISO particles have been tested at both as-fabricated and irradiated conditions. Determination of the mechanical properties of these TRISO particle regions will lead to a better understanding of the SiC layer failure mechanism and enable progress towards TRISO fuel qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Rational design of septenary high-entropy alloy for direct ethanol fuel cells

Promoting C–C bond cleavage of ethanol by a complete 12-electron (12e) ethanol oxidation reaction (EOR) is a grand challenge for the development of highly efficient direct ethanol fuel cells (DEFCs). Most state-of-the-art catalysts only implement the EOR in an incomplete 4e or 2e pathway because of the poisoning of the catalysts by strongly adsorbed CO, leading to the poor output performance and low cell efficiency of DEFCs. Herein, a septenary PtPdFeCoNiSnMn high-entropy alloy (PtPd HEA) with a PtPd-rich surface but super-low platinum group metals loading was developed. We identified and proved the functions of each element in the PtPd HEA. The DEFCs assembled with the PtPd HEA (0.12 mg PtPd ∙cm -2 ) achieved a maximum power density of 0.72 W cm -2 and a durable operation for 1,200 h, which outperforms state-of-the-art catalysts for DEFCs. This work will be a design principle for nanostructured alloy development for renewable energy and sustainability applications.

25 ENERGY STORAGE↗
DOE OSTI2025

Modeling Continuous Online Refueling with SCALE 6.3.1 and Serpent-2 in the EIRENE Novel Molten Salt Reactor Design

The Molten Salt Reactor (MSR) is a Generation IV advanced fission reactor design in which the coolant, and in some cases the fuel itself, is in the form of liquid molten alkali-halide salts with a fluoride or chloride ionic base. In liquid-fueled MSRs, reactor refueling may be performed online, where fresh fuel salt is added to the core during operation without the need for shutdown periods. Refueling for these reactor designs is typically modeled using either a batch or continuous refueling approach, both of which can be simulated with the SCALE 6.3.1 and Serpent-2 code systems. However, the specific manner in which they are implemented can vary depending upon the desired rate of refueling, whether the refueling rate is constant or variable, and consideration of salt drainage for systems where the in-core salt volume is kept constant. In a novel MSR fuel cycle concept termed the “Sourdough” fuel cycle, fuel salt is allowed to “grow” within the core, with excess salt either being transferred to an external holding tank to maintain a constant core volume or diverted to an upper plenum within the core to allow for volume growth. In this work, continuous refueling was modeled in a thermal-spectrum, LEU-fueled, small MSR design operating with the Sourdough fuel cycle using the SCALE 6.3.1 and Serpent-2 codes. Two different continuous refueling approaches were simulated, with the performance of each being compared to determine which is most suitable for use with the Sourdough fuel cycle concept.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

The development of yield-based sooting tendency measurements and modeling to enable advanced combustion fuels

This project produced clear evidence that biofuels can reduce particulate emissions from road vehicles compared with petroleum-derived fuels. It also produced fundamental knowledge that can be used to select fuel compositions and engine designs that will maximize this benefit. This information has been disseminated through more than 20 peer-reviewed publications, more than 30 conference presentations, and 5 freely available internet databases and web tools. This project also provided educational opportunities to 17 students at the high school, undergraduate, and graduate levels.

09 BIOMASS FUELS↗
DOE OSTI2025

Characterization of MC&A for the Molten Salt Fuel Cycle

Advanced reactor developers are exploring diverse reactor designs, including molten salt reactors (MSRs). These advanced reactors are considered for wider applications and a range of deployment locations, including supporting the integration of renewable energy sources in the grid. There are three main types of MSRs: (1) reactors in which the fuel salt freely circulates within the core; (2) reactors with the fuel salt contained within vented fuel tubes; and (3) reactors that use molten salt solely as a coolant, with the fuel in a separate, solid form. In this document, the term MSR refers specifically to the first two types, which use fuel salt—special nuclear material (enriched uranium, plutonium, and 233 U) in chloride or fluoride form mixed with chloride- or fluoride-based carrier salt in a peritectic mixture—as the primary medium for fission. The composition of fuel salt, both at startup and for makeup or refueling, varies depending on the MSR design and the chosen fuel cycle approach, which can be either once-through or closed. For MSRs, a variety of fuel cycle approaches (e.g., U, U–Pu, U–Pu–TRU, U–Th, U–Pu–Th) are being considered. Fuel in MSRs is much different than traditional solid fuel, including its preparation. The uniqueness warrants investigation into characterizing fuel preparation processes, known as fuel salt synthesis . This effort characterized major fuel preparation and synthesis processes, identifying temperature, equipment, and environmental requirements for uranium-, plutonium-, and thorium-based fuel preparation and synthesis. Because MSR fuel salt synthesis facilities handle special nuclear material in loose, bulk form, a material control and accounting plan will be required for licensing from the US Nuclear Regulatory Commission or under the US Department of Energy authorization. This effort serves as a foundation to investigate material control and accounting approaches for synthesis facilities, including determining measurement points and techniques. Because several MSR developers are planning demonstration facilities in the coming years, this effort will support stakeholders with preparing or reviewing material control and accounting plans for providing assurance that all special nuclear material is accounted for at fuel salt synthesis facilities. This report was produced for Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy (DOE), Office of Nuclear Energy, Nuclear Fuel Cycle and Supply Chain.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Effect of ATF Cr-coated-Zircaloy on BWR In-vessel Accident Progression during a Station Blackout

The deposition of protective coatings on nuclear fuel cladding has been considered as a near-term Accident Tolerant Fuel (ATF) concept that will reduce the high-temperature oxidation rate and enhance accident tolerance of the cladding while providing additional benefits during normal and transient. In this study, the performance of the proposed ATF concept of Cr-coated-Zircaloy is assessed using a generic Boiling Water Reactor MELCOR plant model considering a Short-term Station Blackout (STSBO) scenario. Simulation results indicate that the use of Cr-coated-Zircaloy as cladding and canister material mitigates the core degradation process as compared to the traditional Zircaloy cladding and canister design. The onset of fuel degradation and collapse is delayed by over thirty minutes, and the extent of fuel degradation is reduced. Specifically, the gross in-vessel hydrogen generation decreased by almost a factor of three. Although the eutectic reaction between Cr-coating and Zircaloy could cause an early failure of the coating, the improvement in the delay of fuel degradation is still notable. In addition, a thicker coating is found helpful to obtain additional coping time and to decrease hydrogen generation. In addition to the eutectic formation that may compromise Cr-coated Zr, a different failure mode is identified for the Cr-coated-Zr when compared to Zircaloy; i.e., a complete melt of base material leads to component collapse before the coating is oxidized and consumed. These findings can help the industry focus on productive areas of research and development for accident-tolerant fuel concepts and enhancement of core safety margins.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Application of Proliferation Resistance Optimization (PRO-X) Methodology to a Generic Research Reactor

The Proliferation Resistance Optimization Program (PRO-X) has been established by the NNSA to provide a framework for evaluating and integrating proliferation resistance into nuclear reactor system (core, fuel, and auxiliary facilities) designs that also maintain the safety and peaceful use missions of those systems. The research reactor (PRO-RR) area is the component of PRO-X that supports the program objectives by applying state-of-the-art analysis methods to research reactor systems. This report details the results of using a defined set of analytical tools to evaluate the neutronics, thermal hydraulics and proliferation risk characteristics of a set of parametric cores based on a generic 10 MW materials test reactor that uses a plate-type low enriched uranium (LEU) fuel. The analysis shows that by suitable adjustment of the core size, reflector configuration and power level, mission performance and safety margins can be maintained or improved while reducing the potential for production of special nuclear material.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

The importance of delayed neutron precursors in gamma dose calculations for activated primary heat exchanger components in the Molten Salt Breeder Reactor

The Molten Salt Reactor (MSR) Multiphysics Applications technical area in the Nuclear Energy Advanced Modeling and Simulation program has supported the development of 3D Monte Carlo models of the Molten Salt Breeder Reactor (MSBR) over the last couple of years. This MSBR model was previously run with the Shift Monte Carlo code to perform radiation shielding calculations in the reactor cell area. The MSBR is a 2250 MWth (1000 MWe) liquid-fueled molten salt reactor design developed at Oak Ridge National Laboratory in the 1970s. Determining the source terms from activated primary heat exchanger (HX) components is important because delayed and prompt neutron fluxes incident on these components affect the dose rate in the primary HX maintenance areas. This information can be used in the development of remote handling procedures required during shutdown for maintenance. A methodology has been developed and is proposed in this paper to quantify the activated source term from the primary HX components as a result of the movement of the delayed neutron precursors in flowing primary fuel salt through the primary HXs in the MSBR. The goal of this research is to evaluate the gamma dose rates in the maintenance hatches above the primary HX using the activated HX source terms. The study showed that the gamma dose rates are approximately two orders of magnitudes higher when accounting for the neutron activation from the movement of delayed neutron precursors through the HXs than when flowing fuel is not considered. Thus, the movement of delayed neutron precursors must be taken into account for accurately predicting the neutron activation of primary loop components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Criticality Experiments to Reduce Compensating Errors in Plutonium Nuclear Data

Compensating errors between nuclear data observables in a library can adversely impact application simulations. The primary goal of the EUCLID project (Experiments Underpinned by Computational Learning for Improvements in Nuclear Data) is to reduce compensating errors in nuclear data. A new criticality experiment, described in this work, was designed with the specific target nuclear data of 239 Pu fission, inelastic scattering, elastic scattering, capture, nu-bar, and prompt fission neutron spectrum (PFNS). This work will focus on the design and execution of the EUCLID experiment, performed on the Planet vertical lift critical assembly machine at the National Criticality Experiments Research Center (NCERC). The criticality experiment includes two different configurations with very different geometries: one is cube-like to minimize neutron leakage while the other is slab-like to maximize leakage. Having these two widely varying configurations allows the scattering sensitivities of 239 Pu to the neutron multiplication factor to be greatly changed while minimally impacting the other cross section sensitivities. Both configurations utilize the Pu ZPPR (Zero Power Physics Reactor) plates as fuel. The experiments were designed using a D-Optimality criteria, which is an optimization method minimizing the log-determinant of the adjusted nuclear data covariance for the target reactions. These experiments include not only inference of k eff , as done in all critical benchmark experiments, but several other responses as well, such as neutron multiplication measurements and reaction rate ratios. After analysis of the measured data is complete, adjustment of nuclear data will be performed to assess whether the new experimental data successfully reduced compensating errors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Cooled Gas Turbine and Combined Cycle Analysis for NH 3 -CH 4 Fuel Mixes (Up to 100% NH 3 )

In this study, a cooled gas turbine (GT) analysis was conducted for varying levels of ammonia (NH 3 ) blends with methane. The ultimate goal is to have a gas turbine design that can be used for all the fuel blends (including 100% NH 3 ) without any changes to the system. The technological developments in the cooling system, gas turbine design, and materials that will be required for NH 3 combustion were identified and analyzed in this study to develop an advanced gas turbine design for NH 3 fuels. The study includes a combined cycle performance analysis with the NH 3 fuel blends using the advanced gas turbine design developed in this study. A techno-economic analysis was conducted for analyzing the impact of the NH 3 fuels on the levelized cost of electricity and cost sensitivities to fuel price and capacity factor.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2025

Numerical investigation of magneto-inertial fusion targets magnetized by dynamic enforcement of helical current flow

Magnetized Liner Inertial Fusion (MagLIF) targets require premagnetization to reduce thermal conduction losses from the laser-heated fuel to the liner material during implosion. MagLIF targets are typically magnetized by external magnetic field coils which are technologically limited to providing <20 T axial magnetic fields in the fusion fuel. We present a novel target design, AutoMag-D, which employs dynamic enforcement of helical current in the liner, resulting in >20 T axial magnetic field in the fuel region prior to implosion without the need of external magnetic field coils. These self-magnetizing liners are made of helically oriented alternating electrically conductive material and electrically insulating material surrounded by a thin, conductive outer radial layer. As the liner is pulsed with a ∼ 20 MA, ∼ 100 ns rise time drive current from Z, the outer layer of conductive material is shocked and intensely Joule heated, causing it to melt, vaporize, and turn to plasma. This allows the magnetic drive field to diffuse radially inward which dynamically enforces current flow in the helical conduction paths in the liner and produces axial magnetic field in the fuel region prior to implosion of the inner liner surface. AutoMag-D liner designs do not require dielectric breakdown of electrically insulating material (as in traditional auto-magnetizing helical liners) and do not require helical return current geometries (as in dynamic screw pinches). We present results from three-dimensional radiation-magnetohydrodynamic simulations of MagLIF implosions employing AutoMag-D liner designs. Simulated AutoMag-D targets demonstrate improved fuel conditions and thermonuclear yield compared to traditional MagLIF.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2022

Ordered CoPt oxygen reduction catalyst with high performance and durability

A high-performance and durable polymer electrolyte membrane fuel cell cathode catalyst composed of an ordered L10–CoPt core and a thin Pt shell was designed and prepared. Under practical fuel cell membrane electrode assembly testing conditions, the cathode catalyst showed an outstanding initial mass activity of 0.6 A/mgPt, which satisfies the U.S. Department of Energy performance target while also meeting the durability target of less than 40% loss in mass activity after 30,000 accelerated stress test voltage cycles. The high structural stability of the L10–CoPt@Pt-shell catalyst was confirmed by postmortem materials characterization, and the origin of this robustness was revealed by density functional theory calculations, where the barriers for diffusion of Co atoms were observed to be significantly increased in the ordered intermetallic core.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Integral LOCA fragmentation test on high-burnup fuel

Increasing the fuel burnup limit in light water reactors is sought to enhance fuel cycle economics and requires establishing a technical basis. Experimental observations of severe fuel fragmentation under loss-of-coolant-accident conditions at Halden and Studsvik had raised the need for additional considerations during the development of this technical basis. These test data suggested that the burnup threshold for high burnup fuel fragmentation may be influenced by pre-transient power. Additional loss-of-coolant-accident test data is therefore valuable to complement these tests and enhance the current state of understating. Oak Ridge National Lab has developed the Severe Accident Test Station capable of examining the oxidation kinetics and accident response of irradiated fuel and cladding materials for design basis accident and beyond design basis accident scenarios. Severe Accident Test Station provides various temperature profiles, steam, and the thermal shock conditions necessary for integral loss of coolant accident testing, defueled oxidation quench testing, and high-temperature beyond design basis accident testing. Severe Accident Test Station has been successfully installed and demonstrated in the Irradiated Fuels Examination Laboratory at Oak Ridge National Lab. Furthermore, descriptions of the in-cell re-fabrication capabilities and assembly of the loss-of-coolant-accident test train are provided. Installation of the Severe Accident Test Station system and in-cell re-fabrication restores United States capability to examine postulated and extended loss-of-coolant-accident conditions on spent fuel and cladding and provides a platform for evaluating advanced fuel and accident-tolerant fuel cladding concepts. Lastly, three in-cell integral loss-of-coolant-accident test were performed in the Severe Accident Test Station and compared to the Nuclear Regulatory Commission sponsored loss-of-coolant-accident test as well as the original loss-of-coolant-accident test performed at Argonne National Lab. Finally, the results of these tests as well as all publicly available integral loss-of-coolant-accident test were used to validate a threshold for high burnup fuel fragmentation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Cooled Gas Turbine and Combined Cycle Analysis for H 2 -CH 4 Fuel Mixes (Up to 100% H 2 )

In this study, a cooled gas turbine analysis was conducted for varying levels of hydrogen (H 2 ) blends with the natural gas. The ultimate goal is to have a gas turbine design that can be used for all the fuel blends (including 100% H 2 ) without any changes to the system. The technological developments in the cooling system, gas turbine design and materials that will be required for H 2 combustion were identified and analyzed in this study to develop an advanced gas turbine design for H 2 fuels. The study includes a combined cycle performance analysis with the H 2 fuel blends using the advanced gas turbine design developed in this study. A techno-economic analysis was conducted for analyzing the impact of the H 2 fuels on the levelized cost of electricity and cost sensitivities to fuel price and capacity factor.

03 NATURAL GAS↗