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At least 847 records · Page 47
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Progress Report on Mockup Irradiation Capsule Fuel Measurements at LANSCE

Irradiation tests are a key component of nuclear fuel development and identifying typical and atypical regions in the irradiated fuel volume relies on very few characterization techniques. The goal of the effort reported here is to provide complementary measurements adding to the available parameter space for post irradiation examination as well as to inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Pulsed neutrons, enabling diffraction as well as energy-resolved neutron imaging and neutron absorption resonance spectroscopy, offer unique capabilities for this purpose. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel specimen with spatial resolution of 1 mm 3 to 1 cm 3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm 3 was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions, lattice strains (indicative of residual stresses or chemistry variations) and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotope densities of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Physical and Mechano-Electrochemical Phenomena of Thin Film Lithium-Ceramic Electrolyte Constructs

While a small number of solid electrolytes exhibit high ionic conductivity (~1mS cm at 298K), few are stable against Li metal. The garnet-type solid electrolyte, based on the nominal formula Li 7 La 3 Zr 2 O 12 (LLZO), is unique in that it is a fast ion conductor and – as we demonstrated in our recent project (DE-EE-00006821) – is stable against Li. Moreover, our former project successfully demonstrated a decrease in Li-LLZO interface resistance from 12,000 to 2 Ohms cm 2 and stable cycling at 1 mA cm 2 for 100 cycles (+/- 15 µm Li per cycle). Although the past project (DE-EE-00006821) demonstrated LLZO is a viable solid electrolyte for enabling batteries using metallic Li, the studies used thick pellets (1mm) and thick anodes (~500 µm). We believe that to achieve a step increase in technology readiness level (TRL), the same performance characteristics previously shown should be demonstrated in technologically relevant cells, e.g. thin LLZO and thin Li. The goal of this project is to acquire a deep fundamental understanding of the physical and mechano-electrochemical phenomena that control the performance of cells consisting of thin LLZO (~10 µm), thin Li anodes (~20 µm) and thin solid-state composite cathodes. If successful, the knowledge gained in this project will guide closely related commercialization efforts to scale the production of LLZO-based solid-state batteries.

25 ENERGY STORAGE↗
DOE OSTI2020

Design of Experiments Involving Molybdenum in SNL SPRF/CX Facility

This report is part of the collaboration with the Nuclear Criticality Safety Program and constitutes the “CED-2 report” of the experiments proposed in the SPRF/CX facility of SANDIA National Laboratories to contribute to the validation of Molybdenum in thermal energy spectrum. It supplements the study carried out in the CED-1 report by providing a more precise design of the proposed experiments and presents a complete evaluation of the experimental uncertainties. In the CED-1 report, corresponding to the preliminary design of the experiments, IRSN identified several promising configurations that could be implemented in the SPRF/CX facility of SNL. They use metallic molybdenum either in the form of foils between the UO 2 pellets inside BUCCX rods, or in the form of sleeves surrounding 7uPCX rods. In this CED-2 report, a new design using molybdenum sleeves surrounding BUCCX rods is also analyzed and the experimental uncertainties of all the proposed configurations are estimated. Given the constraints to carry out the experiments and the costs associated with the manufacturing of the foils, the experiments using molybdenum foils are discarded. Similarly, the design using molybdenum sleeves around BUCCX rods leads to significant manufacturing costs which could certainly be partially offset by use of existing grids but which would lead to slightly lower values of the integral sensitivities to the cross sections of molybdenum. Consequently, the design using molybdenum sleeves around 7uPCX rods is favored insofar as it presents higher $k_{eff}$ sensitivities to the molybdenum cross sections and that there is still possibility to reduce the number of sleeves to reduce the cost of experiments. Regarding feedback on nuclear data in support of criticality safety assessment, one of the advantages of the work described here is that the proposed experiments are not correlated with the MIRTE and FP experiments using molybdenum. In addition, they lead to higher sensitivity values in the thermal energy range than the existing experiments and partly “cover” the first resonance of the 95 Mo, which is a real advantage for improving knowledge on molybdenum cross sections.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Development of Novel Materials for Direct Air Capture of CO 2 : MIL-101(Cr)-Amine Sorbents Evaluation Under Realistic Direct Air Capture Conditions (Final Report)

The overarching goal of this project is to evaluate the CO 2 adsorption properties of a small family of metal-organic framework (MOFs) materials functionalized with amines at sub-ambient conditions. Our goal is to develop capabilities to measure CO 2 adsorption at conditions more relevant to the weather of the planet. For this purpose, Georgia Tech is constructing a “sub-ambient adsorption facility” in partnership with ZCP Sorbent Development, LLC, aimed specifically at rapidly and deeply characterizing the performance of DAC candidate materials in this important operational range (adsorption at -20 to 20 °C and RH of 0-100%). Here, we use the sub-ambient lab instrumentation designed or adapted to study the behavior of the pristine metal organic framework (MOF) MIL-101(Cr) and the MOF in the presence of amines ranging from small molecules (e.g. TREN, tris(2-aminoethylamine)) to oligomers (e.g. PEI, poly(ethyleneimine)). Any DAC sorbent must be amenable to deployment in practical contactors for gas-solid contacting (traditional pellet-based fixed beds are impossible at scale). To this end, we developed and tested these DAC materials in the forms of composite polymer/MOF fibers and custom 3D-printed monolith structures containing MOF DAC sorbents. The proposed studies advance these materials from technology readiness level (TRL) 2 to TRL 3.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2020

NDCS Group Meeting Work overview: NCSP Task ORNL IP&D-5 [Slides]

The presentation discusses the primary goal of the NCSP Task IP&D-5 which is to use available data from HPRR operation to create a benchmark report for inclusion in the ICSBEP, as a Criticality Accident Alarm System (CAAS) shielding benchmark. The presentation elaborates on the HPRR operation and how it relates to the KENO VI model, sulfur pellets measurements, and calculation methodology between KENO and MAVRIC FW-CADIS. The presentation concludes by stating that some work is still needed to evaluate uncertainties and to explain the discrepancy computation/experiment.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Mesoscale modeling to inform Bison models of accident tolerant fuel concepts

U 3 Si 2 and doped UO 2 have been under investigation in recent years as potential accident-tolerant fuel concepts. In this report, lower-length scale studies of these fuel concepts carried out during Fiscal Year 2020 are detailed. U 3 Si 2 is a potential accident-tolerant fuel that shows promise due to its high thermal conductivity and higher uranium density relative to UO 2 . However, its swelling and fission gas release behavior in light water reactor (LWR) conditions is relatively unknown. To provide mechanistic insight and determine parameters for engineering-scale fuel performance modeling of pellet-form U 3 Si 2 , phase-field simulations of the growth, interconnection, and venting of intergranular fission gas bubbles were performed. The fractional coverage of the grain boundary and the fraction of bubble area that is vented were calculated as a function of time. From the simulation data, the fractional grain boundary coverage at saturation, an important parameter needed in engineering-scale modeling of swelling and fission gas release, was determined. Multiple simulations were run to determine the uncertainty in the calculated value. The effect of model assumptions and input parameters that are not well known was evaluated. Simulation results are compared to related theoretical and computational work. Based on the simulation results, a value of 0.60 for the fractional grain boundary coverage at saturation is recommended for U 3 Si 2 fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Health Physics Research Reactor Criticality Accident Alarm System Benchmark Overview [Slides]

This presentation discusses the Health Physics Research Reactor (HPRR) or Fast Burst Reactor (FBR) and how available data from Health Physics Research Reactor (HPRR) operation to can be used to create a benchmark report for inclusion in the ICSBEP, as a Criticality Accident Alarm System (CAAS) shielding benchmark. The evaluation of experimental data is also presented, specifically four experimental candidates are considered of potential value for the benchmark: 1. Neutron source estimation from a HPRR pulse (energy spectrum, fission yield), 2. Threshold Detector Unit (TDU) measurements at different distances from a HPRR pulse, shielded and unshielded, 3. Sulfur pellet activation at different distances from a HPRR pulse, shielded and unshielded, and 4. Total neutron fluence from a HPRR pulse measured by Bonner Sphere Spectrometry, shielded and unshielded. Additionally, a benchmark model overview is discussed as are sample calculation results.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

Modeling of Reactor Design and Optimization for Scale-Up of the Catalyxx Process for Ethanol Conversion to Higher Alcohol Biofuels

This report summarizes the results of a collaborative efforts between Oak Ridge National Laboratory (ORNL) and Catalyxx Inc. to investigate scale-up of Catalyxx’s Ethanol upgrading to higher alcohols process. The study was funded by the U.S. Department of Energy (DOE) Bioenergy Technologies Office (BETO) under CRADA (Cooperative Research and Development Agreement) No: NFE-20-08396. The project is part of the Direct Funding Opportunity (DFO) for Computational Science to Enable Bioenergy program which utilized computational toolsets developed by the Consortium for Computational Physics and Chemistry, a multi-laboratory consortium in BETO. The report here summarizes a packed-bed reactor modeling effort spanning the range from lab to industrial scale (from 4 gram to 5-ton catalyst beds), and examining reactor design, process optimization strategies, and suggested design and operating conditions for Catalyxx’s ethanol upgrading plants. The results in this report have been shared in monthly steering meetings and presentations are available in the shared data house owned by Catalyxx. The modeling effort helped to define optimum operation conditions for maximum alcohol selectivity and yield: i.e., temperature control scenarios ranging from adiabatic to isothermal, feed rate, pressure, and inlet H 2 /Ethanol ratio. The modeling results were verified at lab-(4 gram) and pre-pilot (4 kg) scales and has been used to evaluate a 5-ton packed-bed reactor and identify operating conditions to maximize the butanol yield. Special focus was given to understanding mass-transfer effects in the pre-pilot and pilot-scale reactors, over the domain of flow rate, pressure, feed composition, pellet size, shape, porosity, bed voidage, and reactor dimensions (i.e., length/diameter). Modeling was also used to evaluate innovative reactor design concepts such as water removal to improve alcohol selectivity and yield, and a reactor with an additional side inlet to facilitate quenching. These concepts were thoroughly explored, and potential benefits were disclosed. The results in this report are summarized and described qualitatively to protect the IP rights of Catalyxx. The details have been shared with the Catalyxx team in the regular steering meetings. At the end of the project, Catalyxx Inc. announced a successful demonstration of pilot scale operation in Seville, Spain.

02 PETROLEUM↗
DOE OSTI2023

Production of Radioisotopes of Scandium for Imaging and Therapy (Final Report)

The theranostic approach is a powerful technique that enables scientists and physicians to diagnose and treat diseases using the same targeting moiety labeled with radionuclides for imaging and therapy. A challenge behind synthesizing and evaluating these theranostic radiopharmaceuticals lies in the production of suitable radionuclides. This study aimed to investigate the production of high radionuclidic purity 43 Sc from enriched [ 46 Ti]TiO 2 for Positron Emission Tomography (PET) and 47 Sc from enriched [ 50 Ti]TiO 2 for targeted radiotherapy, as well as establish a target recycling technique for continued production of these radionuclides. Methods: Targets were prepared with 100 mg of either enriched [ 46 Ti]TiO 2 or [ 50 Ti]TiO 2 , pressed into pellets, and placed into Ta target holders with a 0.127 mm natTi cover foil. The 46 Ti targets were bombarded with 18 MeV protons at 20 µA for 1.5 h, while 50Ti targets were bombarded with 24 MeV protons at 20 µA for 4 or 8 h. The radioscandium nuclides were purified using branched N, N, N’, N’ -tetra-2-ethylhexyldiglycolamide (DGA) resin. The titanium targets were recycled by precipitating the dissolved Ti collected from the column effluent and wash streams, pH adjustment, vacuum filtration, and then drying the solid at 250 °C prior to reforming the target for the subsequent bombardment. The apparent molar activity of the isolated radioscandiums was analyzed by titration with 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA). The relative imaging performance of 18 F, 43 Sc, and 68 Ga were analyzed with PET images acquired with 30-min static scans. Results: The average radioscandium recovery for 43Sc was 91.7 ± 7.4 %, with a radionuclidic purity of 98.8 %. The average recovery of 47Sc was 89.9 ± 3.9 %, with a radionuclidic purity of 90.8%. The average recovery of enriched TiO2 was 96 ± 4.0 %. The highest apparent molar activity for [43Sc]Sc-DOTA was 23.2 GBq/µmol and 3.39 GBq/µmol for [ 47 Sc]Sc-DOTA. The order of PET resolution from highest to lowest was 18 F > 43 Sc > 68 Ga. Conclusions: This work demonstrates the feasibility of using enriched recycled [ 46 Ti]TiO 2 and [ 50 Ti]TiO 2 targets for the production of the 43 Sc and 47 Sc theranostic isotope pair in high purity.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Health Hazards of Exposures to Radioiodine

Iodine is a chemical element with atomic number 53. Iodine-127 is stable (non-radioactive) and commonly found in nature. Elemental iodine is a purple-colored solid at room temperature and pressure, but spontaneously sublimates (turns into vapor). Iodine is an essential element for life, and is required for proper functioning of the thyroid. Iodine is present in many foods, and is readily absorbed by the body and concentrated in the thyroid gland. A fraction of iodine ingested or inhaled is rapidly removed by the kidneys. The rest of the inhaled or ingested iodine is absorbed the by thyroid and retained for many months. Iodine has a biological half-life of approximately 120 days in health individuals. The biological half-life can be shorter in individuals with hyperthyroidism, and longer in individuals with hypothyroidism. Iodine has a number of radioactive isotopes, most of which have relatively short half-lives (days or weeks). Short half-life iodine isotopes are useful for a variety of medical applications, including imaging and cancer therapy. For example, Iodine-123 (half-life 13 hours) is commonly used for medical imaging of the thyroid, while iodine-131 (half-life 8 days) is used for suppressing thyroid function in individuals with hyperthyroidism or ablating (killing) thyroid cells to treat thyroid cancer. Iodine-125 (half-life 59 days) is produced in nuclear reactors, and has medical uses. Although iodine-125 can be used for thyroid imaging, Iodine-123 is more commonly used for that purpose because of its shorter half-life and higher-energy emissions. Iodine-125 is more commonly used for cancer treatment, and can be processed into small metal pellets (seeds) inserted directly into a tumor. Iodine-125 emits low-energy x-rays which can kill tumor cells and generally cannot escape the tumor, sparing other tissues. Medical iodine for imaging or treatment is typically administered orally in the form a pill or liquid solution. A typical adult thyroid scan using iodine-123 involves having the patient swallow between one and four 0.1 millicuries pills, with the exact dose dependent on the patient’s weight. This results in a whole-body committed effective dose of 80 – 320 mrem, and a thyroid equivalent dose of 1443 – 5772 mrem. Note that the whole-body effective dose relates to the overall cancer risk, while the larger equivalent dose to the thyroid only indicates that most of this risk is the result of exposure to the thyroid. These doses are considered safe, although the procedure is not recommended for pregnant or breastfeeding women. In contrast, the quantities of iodine-131 used for treatment of hyperthyroidism and thyroid cancers are much higher. For treatment of hyperthyroidism, 4 – 10 millicuries are administered, while for thyroid cancer the administration can range from 50 – 150 millicuries of I-131. In addition to medical exposures, large populations were exposed to radioiodine as result of the atomic bombings of Hiroshima and Nagasaki in Japan, and the Chernobyl nuclear accident. These populations have been carefully followed for many years to assess the effect of their radiation exposures on cancer risk. As a result, a great deal is known about the cancer risks associated with radioiodine exposure. Because iodine is concentrated in the thyroid, the principal risk of exposure to radioiodine is thyroid cancer. Children have the highest risk of thyroid cancer after exposure to radioactive iodine. According to a large study of Japanese atomic bomb survivors, an effective dose of one Sievert (100,000 mrem) has been observed to increase the risk of thyroid cancer by a factor of 9.5 in children aged zero to nine years old, by a factor of 3 in children aged 10 to 19 years old, and by barely detectable amounts in adults. Another way of quantifying the risk from radioiodine exposure is from risk coefficients, which provide the risk per unit intake of radionuclides in terms of both morbidity (any cancer) and mortality (death). Both morbidity (risk of cancer) and mortality (death) risks are shown in the table below. Note that because thyroid cancer is almost never fatal, the morbidity coefficients are much larger than the mortality coefficients.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2023

The Recovery of Medical Isotope 188 W from Irradiated W Metal Target - A New Approach

Tungsten-188 is in widespread use in 188 W(t 1/2 = 69 d )/ 188 Re(t 1/2 = 16.9 h ) biomedical generators. Oak Ridge National Laboratory has been providing this product to the world since 1999. At ORNL, 188 W is produced via irradiation in ORNL’s High Flux Isotope Reactor (HFIR). Enriched 186 W targets in the form of sintered metallic pellets or rings achieve a compact loading in the irradiation vessel, providing a high yield per unit target. The enrichment of the target is >90% 186 W, and this isotope undergoes double neutron capture to produce the desired 188 W product. While 188 W is produced by neutron bombardment, 191 Os(t 1/2 = 15.4 d ) is simultaneously produced as a by-product and expected to be separated from 188 W by postirradiation treatment.In the current processing pathway, the irradiated W metal rings are first converted into an oxide form of WO 3 by heating the irradiated W metal target at 750°C in a quartz reaction vessel inside a vertical furnace under a constant flow of air. During heating, W metal reacts with oxygen in the air to produce WO 3 , which is soluble in 6 M NaOH for preparation of 188 W product. This oxidation process also converts 188 Os (the decay daughter of 188 W) and 191 Os (15.4 d , the irradiation produced byproduct) into OsO 4 , a highly volatile and toxic gas. The gaseous effluents driven from the quartz reaction vessel are passed through a scrubbing array to remove OsO 4 before the air is discharged from the process. This heterogeneous oxidation method simultaneously achieves goals of (1) converting metal target to a soluble oxide form and (2) removing volatile OsO 4 away from the solid WO 3 product by air flow and absorbing the harmful Os species by the scrubbing array. But this method has two potential problems as well: (1) O 2 reacts with only W metal at high temperatures, not with W alloyed with other elements. The O 2 –W reaction will be retarded when formation of WRe or WC occurs, or even when a layer of non-W materials on the surface of the irradiated W rings.; (2) 100% absorption of OsO 4 of high yield (>90%) from the reaction of Os + O 2 is a strict requirement to the OsO 4 scrubbing system--so NaOH scrubbers of a redundant size (2x 1.5 L) are in use for safety reasons.To resolve above two potential problems, direct dissolution of the irradiated W metal target by a selected reagent is a preferred pathway to avoid heating step with generation of tremendous amount of volatile OsO 4 . Hydrogen peroxide (H 2 O 2 ) is such a candidate to dissolve W in forms of either metal or alloys, although literature lacks information of solubilities of Re or Os in H 2 O 2 . With experimental results of dissolving non-radioactive W, Re and Os in H 2 O 2 under various conditions, this report illustrates a method of H 2 O 2 dissolution for irradiated W target, with a complete dissolution of W and Re, but ≤10% dissolution of Os (converted into gaseous OsO 4 and carried out into a scrubbing for absorption) during processing irradiated W target. The portion of undissolved Os can be separated from W solution by a follow up filtration step. Solubilities of W, Re and Os in H 2 O 2 at a temperature range from 14° to 50°C are presented. And a dissolution rate of W metal per surface area of W metal in H 2 O 2 is calculated based on results of dissolving a W metal cylinder of known surface area in H 2 O 2 at room temperature without stirring.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Wood Heater Design Challenge Workshop Report

To support the United States Department of Energy (DOE) Bioenergy Technologies Office’s (BETO) goal of advancing residential wood heating technologies, Brookhaven National Laboratory (BNL), Lawrence Berkeley National Laboratory (LBNL), and the Alliance for Green Heat (AGH) hosted the fifth Wood Heater Design Challenge (WHDC). The Challenge consists of three virtual workshops and a Technology Slam aimed at inspiring and advancing the next generation of residential wood heaters. With an estimated 12.5 million homes in the United States using wood or pellets for space heating, this challenge stands to create a significant impact on the environmental footprint of domestic heating. This report provides a summary of the activities and outcomes from the three virtual workshops held in January, March, and April of 2022. The goal of the workshops was to bring together stakeholders from across the world to discuss research and development (R&D) challenges and opportunities for advancing residential wood heaters in the United States. The workshops focused on three topics: 1. Advances in wood heater design and technology; 2. Advances in instrumentation used for wood heater testing and field data collection; and 3. Adoption of new wood heater technology and integration with other renewables. Each workshop had more than 100 participants each day which included wood heater manufacturers, researchers, academics, government agencies, and policy makers. Overall, a clear message was presented that collaborative input from experts at air quality agencies, industry, academic institutions, and non-profits is necessary to ensure that wood heating technology continues as a mainstream renewable energy option in our changing energy landscape and to facilitate innovative technology solutions. Participants also provided insights and recommendations that may help guide future R&D investments.

09 BIOMASS FUELS↗
DOE OSTI2021

Tritium Diffusion in Fe-Al Aluminide Coating Bulk Phases

The theoretical investigations of interstitial tritium diffusion pathways in various Al-rich (>60% at. Al) iron aluminide coating phases found that in FeNiAl 5 tritium prefers to diffuse in between layers, while in Fe 4 Al 13 , tritium prefers to diffuse across layers. In the case of Fe 2 Al x phases, the simulations show that the variable Al content in the channels affects the diffusion behavior of tritium in the channels. For 100% of vacancy, tritium prefers to diffuse at the center of the channel, while for 0% Al vacancy, tritium prefers to diffuse along the channel’s wall. In the case where 50% Al vacancy are present in the channels, tritium can diffuse in a mixed way, with positions oscillating between the channel’s center and wall. However, the large energy barrier involved in this case suggests that tritium diffusion could be correlated with Al diffusion in the channels. While we found that at 600 K interstitial tritium diffuses faster in Fe 2 Al x phases (D Τ ≤10 -11 m 2 .s -1 ), then in Fe 4 Al 13 (D Τ ≈10 -12 m 2 .s -1 ), and finally in FeNiAl 5 (D Τ ≈10 -13 m 2 .s -1 ), we also find that interstitial tritium generally diffuses faster in Fe-Al coating phases than in γ-LiAlO 2 pellets (D Τ ≈10 -14 m 2 .s -1 ).

36 MATERIALS SCIENCE↗
DOE OSTI2023

Characterization of Aluminum Metal Powders for Plutonium-238 Program

As part of radioisotope thermoelectric generator (RTG) production for power sources in deep space exploration, Oak Ridge National Laboratory (ORNL) produces plutonium-238 ( 238 Pu). To produce 238 Pu, neptunium-237 ( 237 Np) targets are fabricated at ORNL and subsequently irradiated at the High Flux Isotope Reactor (HFIR) at ORNL and the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Specifically, current flowsheets utilize neptunium dioxide ( 237 NpO 2 ) targets. Aluminum (Al) powder is blended with 237 NpO 2 target material to improve thermal conductivity during irradiation. The current composition of 237 NpO 2 /Al pellets (i.e., cermets, or ceramic–metallics) is 20% 237 NpO 2 , 70% Al, and 10% void space. Al powders utilized by the 238 Pu program are high-fired under vacuum before blending into targets. The primary objective of this work is to measure the physical, chemical, and thermal properties of Al before and after the high-firing process. Properties of interest include the morphology, surface texture, particle size, crystal phase(s), and thermal conductivity of the material. These properties are measured with existing materials characterization equipment at ORNL, including powder x-ray diffraction (pXRD), scanning electron microscopy (SEM) with energy-dispersive spectroscopy (EDS), laser flash analysis (LFA), and thermogravimetric analysis (TGA). A secondary objective of this work is to compare the Al powders before and after the high-firing procedure to determine the effects of high-firing on chemical, physical, and thermal properties and to determine the efficacy of the high-firing process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Biomass Gasification for Chemicals Production Using Chemical Looping Techniques (Final Report)

The Ohio State University (OSU) is investigating the Biomass to Syngas (BTS) chemical looping technology to produce syngas for chemical production applications from biomass under US Department of Energy (DOE) Award #DE-EE0007530. The BTS process aligns with the programmatic area of interest of “Conversion, via biological, thermal, catalytic or chemical means, of acceptable feedstocks into advanced biofuels and/or biobased products including intermediate and end-use products”. Compared to conventional biomass gasification processes, the BTS process eliminates the need for air separation units and tar reforming reactors, which leads to energy efficiency improvement and capital cost reduction. The overall objective is to ascertain the potential of biomass gasification based on the chemical looping technique through mitigation of the possible techno-economic challenges in the steps of scale up for commercialization. The scope of work consists of 1) designing, constructing and operating a 10 kWth commercially scalable sub-pilot BTS system and; 2) completing a comprehensive techno-economic analysis (TEA) of the BTS process using methanol production as an example. Over the course of the project, the project team completed the design, fabrication, and operation of a 10 kWth sub-pilot scale test unit for the BTS process. Corn cob and wood pellets were successfully tested in the unit for high purity syngas generation in extended test campaigns that totals over 200 hours. Syngas purity (H2 and CO) of >70% was achieved with a CH 4 concentration of <6%. The H 2 /CO ratio was greater than 1.8. A comprehensive techno-economic analysis was performed to compare the BTS process and a reference indirectly heated gasification process for methanol synthesis. The result, updated with experimental results for BTS process performance, shows a methanol required selling price (MSP) of $\$ $1.15/gal, compared to $1.28/gal for the reference case.

08 HYDROGEN↗
DOE OSTI2023

Disruption Mitigation Solutions for Long Pulse Tokamaks (Final Technical Report)

The project "Disruption Mitigation Solutions for Long Pulse Tokamaks" was a collaborative effort between ORNL, GA, and UCSD. The main goal of the project was to study the injection of shattered cryogenic pellets (SPI) as a method of reducing wall damage from global tokamak plasma instabilities called disruptions. This work was primarily carried out in the JET tokamak in the UK. The team already had extensive experience studying PSI in the medium sized DIII-D tokamak in the US, but adding data from the larger JET tokamak enabled more confident extrapolation to the even larger planned ITER tokamak in France. Each of the collaborating institutions focused on different aspects of SPI disruption mitigation. The UCSD collaboration focused on the volume recombination of post-disruption runaway electron (RE) beams by massive hydrogen injection. Disruptions can sometimes result in the formation of toroidal (ring shaped) beams of highly energetic (relativistic) electron beams, which can cause serious local wall damage, even in present-day tokamaks. A promising method to reduce wall damage from the RE beams is to inject massive amounts of H 2 or D 2 into the beams using SPI. although the precise method by which the injected H 2 reducses RE wall damage is still being investigated, it seems clear that a necessary and required step for the reduction to occur is for the H 2 injection to cause volume recombination (free electron density reduction) of the background cold plasma surrounding the fast electron beam. The UCSD collaboration focused on answering the question of if this recombination could be achievable in the future reactor-size tokamak ITER using realistic levels of H 2 injection. A model capturing the most essential physics was developed and was constrained using both DIII-D and JET data. This model was then used to extrapolate to ITER. It was predicted that RE beam recombination should be achievable in ITER, although in the worst-case scenari (a 10 MA Re beam in ITER), recombination is barely achieved within the maximum allowed levels of H 2 injection.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2023

LLNL FY23 Aging and Lifetimes Exit Criteria: Milestone 8650, GC #4

Silicone rubbers are important materials within the complex and are used in various applications due to wide operating temperatures, favorable aging characteristics, and chemical inertness. However, silicone rubbers are susceptible to degradation in the presence of acid or base. The goal of the project is to design outgassing vessels to probe material degradation byproducts from additively manufactured (AM) silicones compressed against Arrhenius base under dry (sub 100 ppm moisture) environments. The overarching goal of this task is to develop a mechanistic understanding of degradation from experimental data that will be used to inform predictive lifetime and aging models. An outgassing chamber for multi-material compressive aging was designed. The chamber consists of a stainless-steel enclosure, which is hermetically sealed once AM silicone samples are compressed against Arrhenius base pellets in compression rigs (Fig. 1). Outgassing chambers, compression rigs, and accessories were procured, arrived onsite, and received surface treatments. Outgassing of degradation byproducts will be monitored over 12 weeks at 35 °C and 70 °C.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Bench-Scale Testing of a High Efficiency, Ultra-Compact Process for Pre-Combustion CO 2 Capture (Final Report)

The overarching objective of this bench-scale study is to field-validate the technical feasibility of a membrane- and adsorption-enhanced water gas shift (WGS) reaction process that employs a carbon molecular sieve (CMS) membrane reactor (MR) followed by an adsorption reactor (AR) for pre-combustion CO 2 capture while demonstrating progress towards achievement of the overall performance goals of CO 2 capture with 95% CO 2 purity at a cost of electricity (COE) of 30% less than the baseline capture approaches. The main goal of this bench-scale project is to advance the proposed MR-AR transformative CO 2 capture technology to a sufficient maturity level so that our present industrial partners (M&PT and Worley) can potentially adapt the technology for further small pilot-scale testing. The project began at TRL 4, as the system prototype had already been validated in the laboratory on simulated syngas (as part of project DE-FE0026423). The project ended at TRL 5, via scaling-up of the prototype system and testing it on actual syngas at a host site (CAER at the University of Kentucky (UKy)). The project was carried-out in two different phases: In Phase I (during BP1 of the project), the team designed, constructed, and assembled the bench-scale experimental MR-AR system, prepared the membranes, adsorbents, and catalysts, tested the bench-scale unit with simulated syngas to validate functionality, and prepared a preliminary TEA of the technology. In Phase II (during BP2 of the project), the team installed the unit at the test site at the UKy site and completed all utility connections and hookups, field-tested the novel MR-AR process in the bench-scale system using real syngas, collected and analyzed experimental data, and completed a detailed TEA of the technology. A key aim of the project was to identify and address the technical and process risks and to generate information to advance the technology to the next stage of development. All project milestones and success criteria were met. Specifically, the team: (i) designed, constructed and assembled a bench scale system that handled 1 – 5 scfm syngas and tested its functionality with simulated syngas; (ii) prepared high performance CMS membrane tubes (ID: 3.5 mm, OD: 5.7 mm, 30” long) that meet the target H 2 permeance ( > 1 m 3 /(m 2 .hr.bar) or (> 370.3 GPU)) and a target H 2 /CO selectivity >80 at the relevant temperature (up to 300oC) and pressure conditions (up to 25 bar) with a <10% decline in performance over each 250 hr. testing period (iii) procured commercial sour-shift catalyst in sufficient quantity and prepared up to 10 kg of pelletized adsorbent for use at relevant conditions (250oC 2.5wt.% and target sorbent attrition rate < 0.2; (iv) installed the bench-scale unit at the test site (UKy) and tested the MR-AR technology using real syngas for over 250 hr. for both static and flow experiments; (v) updated the TEA analysis based on bench-scale data and met CO 2 capture goals of 95% CO 2 purity at a cost of electricity 30% less than baseline capture approaches. The CMS membranes, HTC adsorbents and catalysts employed all exhibited very robust and stable performance during the long-term run (over a >250 hr live syngas run). Furthermore, the proposed MR-AR IGCC system achieved a LCOE with a N 2 sale price of $\$$ 30/ton, with N 2 compression (no N 2 compression) of 101.2 $\$$/MWh (95.3 $\$$/MWh), which represents a 29.8% (33.9%) LCOE reduction in the baseline IGCC with carbon capture of 144.2 $\$$/MWh. The proposed MR-AR IGCC delivers a CO 2 capture cost of 44.8 $\$$/tonne (39.9 $\$$/tonne) vs 98.1 $\$$/tonne of the baseline capture case and a net power production of 586 MWe (623 MWe) vs 556 MWe of the baseline capture case.

01 COAL, LIGNITE, AND PEAT↗