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At least 1,639 records · Page 91
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Enhanced Component Performance Study: Turbine-Driven Pumps 1998–2014

This report presents an enhanced performance evaluation of turbine-driven pumps (TDPs) at U.S. commercial nuclear power plants. The data used in this study are based on the operating experience failure reports from fiscal year 1998 through 2014 for the component reliability as reported in the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The TDP failure modes considered are failure to start (FTS), failure to run less than or equal to one hour (FTR=1H), failure to run more than one hour (FTR>1H), and normally running systems FTS and failure to run (FTR). The component reliability estimates and the reliability data are trended for the most recent 10-year period while yearly estimates for reliability are provided for the entire active period. Statistically significant increasing trends were identified for TDP unavailability, for frequency of start demands for standby TDPs, and for run hours in the first hour after start. Statistically significant decreasing trends were identified for start demands for normally running TDPs, and for run hours per reactor critical year for normally running TDPs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x10 25 n/m 2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x10 25 n/m 2 .

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Air Dispersion Modeling for the INL Application for a Synthetic Minor Sitewide Air Quality Permit to Construct with a Facility Emission Cap Component

The Department of Energy Idaho Operations Office (DOE-ID) is applying for a synthetic minor, Sitewide, air quality permit to construct (PTC) with a facility emission cap (FEC) component from the Idaho Department of Environmental Quality (DEQ) for Idaho National Laboratory (INL) to limit its potential to emit to less than major facility limits for criteria air pollutants (CAPs) and hazardous air pollutants (HAPs) regulated under the Clean Air Act. This document is supplied as an appendix to the application, Idaho National Laboratory Application for a Synthetic Minor Sitewide Air Quality Permit to Construct with a Facility Emissions Cap Component, hereafter referred to as “permit application” (DOE-ID 2015). Air dispersion modeling was performed as part of the permit application process to demonstrate pollutant emissions from the INL will not cause a violation of any ambient air quality standards. This report documents the modeling methodology and results for the air dispersion impact analysis. All CAPs regulated under Section 109 of the Clean Air Act were modeled with the exception of lead (Pb) and ozone, which are not required to be modeled by DEQ. Modeling was not performed for toxic air pollutants (TAPs) as uncontrolled emissions did not exceed screening emission levels for carcinogenic and non-carcinogenic TAPs. Modeling for CAPs was performed with the EPA approved AERMOD dispersion modeling system (Version 14134) (EPA 2004a) and five years (2000-2004) of meteorological data. The meteorological data set was produced with the companion AERMET model (Version 14134) (EPA 2004b) using surface data from the Idaho Falls airport, and upper-air data from Boise International Airport supplied by DEQ. Onsite meteorological data from the Grid 3 Mesonet tower located near the center of the INL (north of INTEC) and supplied by the local National Oceanic and Atmospheric Administration (NOAA) office was used for surface wind directions and wind speeds. Surface data (i.e., land use data that defines roughness, albedo, Bowen ratio, and other parameters) were processed using the AERSURFACE utility (Version 13016) (EPA 2013). Emission sources were modeled as point sources using actual stack locations and dimensions. Emissions, flow rates and exit temperatures were based on the design operating capacity of each source. All structures close enough to produce an area of wake effect were included for all sources. For multi-tiered structures, the heights of the tiers were included or the entire building height was assumed to be equal to the height of the tallest tier. Concentrations were calculated at 1,352 receptor locations provided by DEQ. All receptors were considered for each pollutant and averaging period. Maximum modeled CAP concentrations summed with average background concentration values were presented and compared to National Ambient Air Quality Standards (NAAQS). The background concentration values used were obtained using the Washington State University’s Laboratory for Atmospheric Research North West Airquest web-based retrieval tool (http://lar.wsu.edu/nw airquest/lookup.html). The air dispersion modeling results show the maximum impacts for CAPs are less than applicable standards and demonstrate the INL will not cause a violation of any ambient air quality standards.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2016

AGR-3/4 Irradiation Test Train Disassembly and Component Metrology First Look Report

The AGR-3/4 experiment was designed to study fission product transport within graphitic matrix material and nuclear-grade graphite. To this end, this experiment consisted of 12 capsules, each fueled with 4 compacts containing UCO TRISO particles as driver fuel and 20 UCO designed-to-fail (DTF) fuel particles in each compact. The DTF fuel was fabricated with a thin pyrocarbon layer which was intended to fail during irradiation and provide a source of fission products. These fission products could then migrate through the compact and into the surrounding concentric rings of graphitic matrix material and/or nuclear graphite. Through post-irradiation examination (PIE) of the rings (including physical sampling and gamma scanning) fission product concentration profiles within the rings can be determined. These data can be used to elucidate fission product transport parameters (e.g. diffusion coefficients within the test materials) which will be used to inform and refine models of fission product transport. After irradiation in the Advanced Test Reactor (ATR) had been completed in April 2014, the AGR-3/4 experiment was shipped to the Hot Fuel Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) for inspection, disassembly, and metrology. The AGR-3/4 test train was received at MFC in two separate shipments between February and April 2015. Visual examinations of the test train exterior did not indicate dimensional distortion, and only two small discolored areas were observed at the bottom of Capsules 8 and 9. No corresponding discoloration was found on the inside of these capsules, however. Prior to disassembly, the two test train sections were subject to analysis via the Precision Gamma Scanner (PGS), which did not indicate that any gross fuel relocation had occurred. A series of specialized tools (including clamps, cutters, and drills) had been designed and fabricated in order to carry out test train disassembly and recovery of capsule components (graphite rings and fuel compacts). This equipment performed well for separating each capsule in the test train and extracting the capsule components. Only a few problems were encountered. In one case, the outermost ring (the sink ring) was cracked during removal of the capsule through tubes. Although the sink ring will be analyzed in order to obtain a mass balance of fission products in the experiment, these cracks do not pose a major concern because the sink ring will not be analyzed in detail to obtain the spatial distribution of fission products. In Capsules 4 and 5, the compacts could not be removed from the inner rings. Strategies for removing the compacts are being evaluated. Sampling the inner rings with the compacts in-place is also an option. Dimensional measurements were made on the compacts, inner rings, outer rings, and sink rings. The diameters of all compacts decreased by 0.5 to 2.0 %. Generally, the extent of diametric shrinkage increased linearly with increasing neutron fluence. Most compact lengths also decreased. Compact lengths decreased with increasing fluence, reaching maximum shrinkage of about 0.9 % at a fast fluence of 4.0x1025 n/m2 E > 0.18 MeV. Above this fluence, the extent of length shrinkage appeared to decrease with fluence, and two compacts from Capsule 7 were found to have slightly increased in length (< 0.1 %) after a fluence of 5.2x1025 n/m2.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

9975 Shipping package component long-term degradation rates

Special nuclear materials are being stored in the K-Area Complex using 3013 containers that are held within Model 9975 shipping packages. The service life for these packages in storage was recently increased from 15 to 20 years, since some of these packages have been stored for nearly 15 years. A strategy is also being developed whereby such storage might be extended beyond 20 years. This strategy is based on recent calculations that support acceptable 9975 package performance for 20 years with internal heat loads up to 19 watts, and identifies a lower heat load limit for which the package components should degrade at half the bounding rate or less, thus doubling the effective storage life for these lower wattage packages. The components of the 9975 package that are sensitive to aging under storage conditions are the fiberboard overpack and the O-ring seals, although some degradation of the lead shield and outer drum are also possible. This report summarizes degradation rates applicable to lower heat load storage conditions. In particular, the O-ring seals should provide leak-tight performance for more than 40 years in packages for which their maximum temperature is ≤135 °F. Similarly, the fiberboard should remain acceptable in performance of its required safety functions for up to 40 years in packages with a maximum fiberboard temperature ≤125 °F.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Additively Manufactured IN718 Components with Wirelessly Powered and Interrogated Embedded Sensing

A methodology is described for embedding commercial-off-the-shelf sensors together with wireless communication and power circuit elements using direct laser metal sintered additively manufactured components. Physics based models of the additive manufacturing processes and sensor/wireless level performance models guided the design and embedment processes. A combination of cold spray deposition and laser engineered net shaping was used to fashion the transmitter/receiving elements and embed the sensors, thereby providing environmental protection and component robustness/survivability for harsh conditions. By design, this complement of analog and digital sensors were wirelessly powered and interrogated using a health and utilization monitoring system; enabling real-time, in situ prognostics and diagnostics.

embedded sensing↗
DOE OSTI2018

Thermal Analysis of Fermilab Mu2e Beamstop and Structural Analysis of Beamline Components

The Mu2e project at Fermilab National Accelerator Laboratory aims to observe the unique conversion of muons to electrons. The success or failure of the experiment to observe this conversion will further the understanding of the standard model of physics. Using the particle accelerator, protons will be accelerated and sent to the Mu2e experiment, which will separate the muons from the beam. The muons will then be observed to determine their momentum and the particle interactions occur. At the end of the Detector Solenoid, the internal components will need to absorb the remaining particles of the experiment using polymer absorbers. Because the internal structure of the beamline is in a vacuum, the heat transfer mechanisms that can disperse the energy generated by the particle absorption is limited to conduction and radiation. To determine the extent that the absorbers will heat up over one year of operation, a transient thermal finite element analysis has been performed on the Muon Beam Stop. The levels of energy absorption were adjusted to determine the thermal limit for the current design. Structural finite element analysis has also been performed to determine the safety factors of the Axial Coupler, which connect and move segments of the beamline. The safety factor of the trunnion of the Instrument Feed Through Bulk Head has also been determined for when it is supporting the Muon Beam Stop. The results of the analysis further refine the design of the beamline components prior to testing, fabrication, and installation.

Narug, Colin S.↗
DOE OSTI2018

Embedded Sensors and Controls to Improve Component Performance and Reliability - Final Report

Oak Ridge National Laboratory has researched embedded instrumentation and controls (I&C) for extreme environments under the Department of Energy, Office of Nuclear Energy's Nuclear Energy Enabling Technology program. Advanced sensors and controls have the potential to dramatically increase the performance and efficiency of nuclear reactor components, however, the extreme environments of exiting and advanced reactor designs limit the options for sensors, actuators, and control designs. To address two of the extreme environment challenges to embedded I&C, the focus of this research is to fill technology gaps needed for electromagnetic systems that can operate at temperatures greater than 700 C in corrosive environments. A magnetically suspended, canned-rotor pump was chosen for the embedded I&C development platform because it is a system with a practical need in industry and because the sensors, actuators, and embedded I&C technologies needed to successfully build this platform will be widely applicable to any high-temperature components that utilize electromagnetic interactions. This design also eliminates the need for pump seals and mechanical bearings which are the main sources of pump failures and maintenance costs. During the project a conceptual design for the high-temperature magnetically suspended pump was developed and analyzed for manufacturability and reliability. Specifications for the materials, control system, sensors, mechanical performance, high-temperature wiring, and electromagnetic properties were developed for the conceptual design. The canned rotor design of the high-temperature pump presents unique fluid force characteristics arising from the thin fluid channel between the rotor and stator. A model of the fluid forces was developed and the impact of these forces on the stability and performance of magnetic bearing controllers was studied. Then a room temperature bench-scale magnetic testbed was designed and fabricated for testing the performance of sensors and electromagnetic actuators, modern control design techniques, closed-loop system identification, state estimation, and model uncertainty. Finally, a loop-scale testbed with submerged magnetic bearings was designed and fabricated. This testbed will be used to analyze the stability and performance of control system designs in response to the nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics, gyroscopic effects, and impeller disturbances.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Report on the completion of long-lead components procurement of new sodium materials test loop

This report provides an update on the completion of long-lead components procurement of new sodium materials test loop. The loop, designated as SMT-3, is a forced-convection sodium loop to be used for the evaluation of sodium compatibility of advanced Alloy 709 with emphasis on long term exposures of tensile, creep, fatigue, creep fatigue, and fracture toughness ASTM-size specimens in support of ASME Code qualification and NRC licensing. The report is a deliverable (Level 4) in FY18 (M4NT-18AN050503013), under the Work Package NT-18AN05050301, “A709 Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of the Advanced Materials Program for the Advanced Reactor Technology. This work package enables the development of advanced structural materials by providing corrosion, microstructure, and mechanical property data from the standpoint of sodium compatibility of advanced structural alloys. The first sodium loop (SMT-1) with a single tank was constructed in 2011 at ANL and has been in operation for exposure of subsize sheet specimens of advanced alloys at a single temperature. The second sodium loop with dual tanks (SMT-2) was constructed in 2013 and has been in operation for the exposure of subsize sheet specimens of advanced alloys at two different temperatures. The current loop (SMT-3) has been designed to incorporate sufficient chamber capacity to expose a large number of ASTM-size specimens to evaluate the sodium effects on tensile, creep, fatigue, creep-fatigue, and fracture toughness properties, in support of ASME Code Qualification and USNRC Licensing. Three major long lead items were identified: 1) Selection and approval of location for the new loop and installation of the necessary infrastructure, such as transformer, power supply panels, and wiring, that is adequate for the long term operation of the loop, 2) Perform a detailed piping analysis that meets the ANL QA and Safety guidelines as well as complying with ASME Code requirements, and 3) Design and fabrication of the specimen exposure vessels that are capable of at least 42 tensile, creep, and fatigue specimens in each vessel and is compliant and compatible with ASME Code and the piping design analysis. The SMT-3 loop will be located in Building 206 adjacent to the currently operating SMT-2 loop. In addition, we have removed the aged power supply system in Building 206 and installed a new transformer, wiring, and power panels for the new loop. The preliminary loop design was sent to an outside vendor for a detailed piping analysis, to ensure compliance with the ASME Code guidelines. The piping design was completed in FY2017. The construction of components such as cold trap, economizers, piping arrangement etc. will be performed in the central shops at ANL. About 150 liters of sodium for the loop will be procured in early FY2018. The loop system is designed to circulate sodium through the sample tanks and he associated loop without an operator for an extended period of time. With the three sodium loops (i.e., single-tank, dual-tank and four–tank), materials can be tested at different sodium temperatures, and large tensile, creep, fatigue, creep-fatigue, and fracture toughness specimens can be exposed to sodium for extended periods of time and generate data on mechanical properties in support of ASME Code Qualification and USNRC Licensing of advanced Alloy 709 for use as a structural material in SFRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Microlfuidic Component Development

908 Devices Inc. (908 Devices) had proposed to meet Defense Threat Reduction Agency (DTRA) requirements for a portable chemical vapor detector system, by coupling their high-pressure ion trap mass spectrometer to a microfabricated gas chromatography column (μGC). Sandia National Laboratories (SNL’s) role under this Cooperative Research and Development Agreement (CRADA) Project Task Statement (PTS) was to provide silicon μGC die (of two design variants), and μPC die (also of two variants) for testing and integration by the CRADA partners. These design variants were fabricated and delivered by SNL in Project Year 1. In Project Year 2, SNL personnel have advised 908 Devices personnel regarding operation and test of μGC and μPC components, as well as integration of these components into a complete chemical analysis system.

47 OTHER INSTRUMENTATION↗
DOE OSTI2018

New Mechanistic Models of Creep-Fatigue Interactions for Gas Turbine Components (Final Report)

This report documents activities for DE-FE0011796 New Mechanistic Models of Creep-Fatigue Interactions for Gas Turbine Components. As part of the computational mechanics work two rate dependent strain gradient plasticity models are developed. One model is an extension of the Norton model and the other model an extension of the Kocks-Mecking model of unified viscoplasticity. Both models are used to compute results for time dependent deformation problems (gradient creep, void growth, a rotating turbine disk). Detailed numerical verification studies are reported. A plasticity model appointing for the strength differential effect in IN 718 is implemented and employed to analyze indentation experiments. Crack growth simulations are conducted with a cohesive zone model approach combining a rate independent fatigue damage component with a time dependent damage model. Crack growth simulations reported for several combinations of the cohesive zone model with the strain gradient plasticity formulations. Plastic strain gradients lead to local suppression of plastic strain which alters the fatigue crack growth response and crack tip fields locally. It also contributes to the condition of crack deflection and bifurcation at interfaces and interlayers as relevant for the understanding of the crack growth response in certain conditions of IN 718 manufactured additively. IN 718 from a conventional and additive manufacturing route (including a HIP process) are considered. Detailed microstructure characterization is reported. Fatigue crack growth at room temperature and at 650°C are reported for all three material variants, and microscale mechanisms are described.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Preliminary Activation Analysis of Accelerator Components and Beam Stops at the NSLS-II

Bremsstrahlung is generated at the accelerator systems of the synchrotron radiation facilities by the radiative interaction of circulating electron beam with the accelerator components and with the residual gas molecules in the vacuum chamber. The photoneutron interaction of bremsstrahlung with the materials leads to the radio activation of the accelerator components through neutron emission and the production of radioisotopes. However this activation interaction is a second order effect because the intermediate process of bremsstrahlung production is involved. The photoneutron interaction takes place above the threshold bremsstrahlung energy of 7-20 MeV and the cross section for these interactions are in millibarns.

36 MATERIALS SCIENCE↗
DOE OSTI2013

NSLS-II Storage Ring Layout and Components

The NSLS-II storage ring consists of 30 cells with alternating long (9.3m) and short (6.6m) straights. The layout of the magnets and the vacuum components has 15-fold symmetry, even though some of the components, such as scrapers, strip lines, are put in asymmetrically at only a few places. The insertion devices will likely be different for each straight; therefore the magnet strength will be slightly adjusted when IDs are installed.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Development of Agile and Cost Effective Routes for Manufacturing Reliable Ceramic Components for SOFC Systems

Additive manufacturing routes were explored and developed to enable the low to medium volume manufacturing of ceramic components with a short lead time, flexibility for iterative design, and without the high CAPEX and tooling costs associated with green machining or injection molding. Agile manufacturing route of gelcasting and 3D printing were selected and work focused on the production of ceramic manifold components utilized in the balance of plant for Saint-Gobain’s all ceramic SOFC stack. This work showed that both gelcasting and a specific version of 3D printing were able to create parts with the proper fidelity and at lower cost than traditional powder processing followed by green machining.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Development of Agile and Cost Effective Routes for Manufacturing Reliable Ceramic Components for SOFC Systems (Final Report)

Additive manufacturing routes were explored and developed to enable the low to medium volume manufacturing of ceramic components with a short lead time, flexibility for iterative design, and without the high CAPEX and tooling costs associated with green machining or injection molding. Agile manufacturing route of gelcasting and 3D printing were selected and work focused on the production of ceramic manifold components utilized in the balance of plant for Saint-Gobain’s all-ceramic SOFC stack. This work showed that both gelcasting and a specific version of 3D printing were able to create parts with the proper fidelity and at lower cost than traditional powder processing followed by green machining.

03 NATURAL GAS↗
DOE OSTI2019

Additive Manufacturing Fabrication of PEEK Counter-Current Centrifugal Contactor Components

Additive manufacturing (AM, also known as 3D printing) offers significant advantages for the fabrication of annular centrifugal contactor-separator (CCS) stages, particularly when complemented by post-processing via traditional machining. The shortcomings of 3D printing of CCSs are offset by its low cost, quick turnaround, and ability to manufacture enhanced mixing sections and multistage modules. High-density acrylic materials show promise for improved chemical resistance and mechanical properties. The demonstration of the MOEX process using acrylic AM CCS stages confirmed their full functionality and showed significant flexibility in geometry design while using relatively low-cost plastic materials. Over the past two years, AM has been used to produce low-cost centrifugal contactor components, ranging from bench-scale laboratory testing to a full-scale industrial process with on-hand availability of spare parts and great flexibility in design modifications and quick testing and implementation. However, the shortcomings of stereo lithography assembly (SLA)-printed acrylic components became apparent rather quickly. By the completion of the initial MOEX demonstration using the 9-stage 2-cm CCS bank, nearly half of the rotors had begun to fail from pitting and cracking. These problems were most likely caused by swelling of the rotor due to liquid absorption, resulting in improper excess run out, mechanical abrasion, and chemical exposure of the roughened surface. Additional deterioration of the acrylic stages was later discovered when the interiors of the acrylic housings began to experience flaking and delamination as the plastic dried out several weeks after the tests. All of these issues caused severe limitations to the lifespan of the CCS stages, which are acceptable for R&D but undesirable for commercial-scale production.

42 ENGINEERING↗
DOE OSTI2019

Development of Agile and Cost Effective Routes for Manufacturing Reliable Ceramic Components for SOFC Systems (Final Scientific Report Updated)

Additive manufacturing routes were explored and developed to enable the low to medium volume manufacturing of ceramic components with a short lead time, flexibility for iterative design, and without the high CAPEX and tooling costs associated with green machining or injection molding. Agile manufacturing route of gelcasting and 3D printing were selected and work focused on the production of ceramic manifold components utilized in the balance of plant for Saint-Gobain’s all-ceramic SOFC stack. This work showed that both gelcasting and a specific version of 3D printing were able to create parts with the proper fidelity and at lower cost than traditional powder processing followed by green machining.

36 MATERIALS SCIENCE↗