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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 55 records · Page 3

Modeling Radionuclide Inventories in MSR Off-Gas Systems with Radiochemical Transport Analysis

Molten salt reactors (MSR) contain unique characteristics that may require enhancements to modeling tools to accurately predict phenomena. One characteristic that may be advantageous to leverage during normal operation is on-line processing of the circulating fuel salt, such as an off-gas system (OGS) to remove volatile fission products. Therefore, new modeling tools must be developed to integrate spatial resolution and chemistry effects into fuel depletion tools to be able to account for these non-core sources of radioactivity. Such types of radiochemical transport analysis tools were used to estimate the removal rates for 12 elements within a flow model of the Molten Salt Reactor Experiment (MSRE) by optimizing against legacy experimental data of the gas-borne (GB) percentages of 12 nuclides. The removal rates were used in a depletion model to calculate the FP inventory that enters the MSRE OGS. Calculations are in good agreement with the empirical GB percentages reported for the 12 nuclides, which validates the approach and verifies each tool’s treatment of the radiochemical flow effects. The OGS inventory is discussed in terms of the largest nuclide contributors to activity, dose consequence, decay heat, and elemental composition. Finally, insights from the study allow recommendations to be made for future code development activities.

Shahbazi, Shayan↗

Two-stage formation-energy correction (NbZr, TaZr, VZr)

This bundle contains the scripts, the raw and corrected per-structure data, and the manuscript plots for the NbZr / TaZr / VZr BCC binary formation energies and the associated RMSDs. Why a two-stage correction is necessary: The "raw" formation energy of every relaxed VASP configuration is computed in the usual way, FE_raw(c) = E_alloy(c) - sum_i x_i * E_pure_i , where E_pure_i are the per-atom total energies of the pure-element reference structures (Nb, Ta, V, Zr in the same BCC supercell, with identical INCAR / KPOINTS / PAW choices). With perfectly consistent reference runs the raw FE should vanish at the two pure-element endpoints (x = 0 and x = 1) by construction. In practice this does not hold for two reasons that are present in our dataset: 1. Reference-energy inconsistency (composition-dependent bias). Even with identical input parameters, the pure-element runs (stored in `corrected_DFT_pure_element_runs/`) differ slightly from the values that would be implied by the alloy runs at near-pure compositions (a few meV/atom). This bias is approximately linear in concentration, because the residual error in E_pure_Nb (or E_pure_Ta / E_pure_V) propagates into FE_raw(c) as (1 - x) * dE_pure_1, and the corresponding error in E_pure_Zr propagates as x * dE_pure_2. Left uncorrected, this produces a non-physical "tilt" of FE_raw(x) and shifts the entire FE-vs-x cloud away from zero at the endpoints. 2. Endpoint anchoring against the audited true endpoints. The strict endpoint values (FE_x0_meVatom, FE_x1_meVatom in `corrected_fe_strict_endpoints_20260518/strict_endpoint_check_20260518.csv`) were re-derived from an independent cross-check of the pure-element runs. After stage 1 removes the linear bias, the near-pure compositions in the alloy dataset still extrapolate to values that differ slightly from these audited endpoints — because stage 1 is fit from a few near-end alloy bins, not from the audited pure-element references themselves. The README.txt file discusses how these issues are addressed by the two-stage correction, and describes folder layout, pipeline summary, and how to re-run.

36 MATERIALS SCIENCE↗

Catalytic hot-gas filtration of biomass pyrolysis vapors

The present disclosure relates to a device that includes a filter element and a catalyst, where the filter element is configured to remove particulate from a stream that includes at least one of a gas and/or a vapor to form a filtered stream of the gas and/or the vapor, the catalyst is configured to receive the filtered stream and react a compound in the filtered stream to form an upgraded stream of the gas and/or the vapor, further including an upgraded compound, and both the filter element and the catalyst are configured to be substantially stable at temperatures up to about 500° C.

Magrini, Kimberly A.↗

Data for Identifying the best high-biomass sorghum hybrids based on biomass yield potential and feedstock quality affected by nitrogen fertility management under various environments

Data were collected from agronomy fields in Urbana and Ewing, IL, during the 2022 and 2023 growing seasons. The dataset includes dry biomass yield, nitrogen, phosphorus, and potassium concentrations and removals, and chemical composition elements (cellulose, hemicellulose, lignin, and soluble fractions) for 13 high-biomass sorghum hybrids. data_sharing.xlsx contains 20 columns and 104 rows. Below is the explanation of all variables in the file: Year: 2022; 2023 Location: Urbana, IL; Ewing, IL N rate (kg-N/ha): 0; 112 Hybrid #: H1-H13 Pedigree: Pedigree for 13 hybrids Dry biomass yield (Mg/ha): Aboveground dry biomass yield N (g/kg): Nitrogen concentration in plant tissue P (g/kg): Phosphorus concentration in plant tissue K (g/kg): Potassium concentration in plant tissue N (kg/ha): Nitrogen removal by aboveground biomass P (kg/ha): Phosphorus removal by aboveground biomass K (kg/ha): Potassium removal by aboveground biomass Cellulose (g/kg): Cellulose concentration in plant tissue Hemicellulose (g/kg): Hemicellulose concentration in plant tissue Lignin (g/kg): Lignin concentration in plant tissue Soluble (g/kg): Soluble concentration in plant tissue Cellulose (Mg/ha): Cellulose content in aboveground biomass Hemicellulose (Mg/ha): Hemicellulose content in aboveground biomass Lignin (Mg/ha): Lignin content in aboveground biomass Soluble (Mg/ha): Soluble content in aboveground biomass

environmental adaptability↗

Sludge Removal: Success and Partnership with T Plant - 20510

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) has safely and efficiently managed the removal of highly radioactive sludge from the 105-K West Reactor fuel storage basin (105- KW Basin) near the Columbia River. In mid-September 2019, sludge retrieval was completed after 21 storage containers were safely loaded with highly radioactive sludge and transported approximately 13 miles from the 105-KW Basin to T Plant, located in the 200 West Area on the Central Plateau of the Hanford Site. Sludge removal represents the last major source term reduction necessary before the K West Reactor and fuel storage basin can transition to closure activities in preparation for final deactivation and demolition. The technical complexity of the sludge retrieval process, coupled with the challenging physical and radiological characteristics of the waste, necessitated a methodical and deliberate approach using unique design and operational solutions to safely conduct the work. This challenge was further complicated by the need to plan and conduct the work in two separate facilities, each subject to the controls and requirements of its own nuclear facility safety basis. The project overcame these challenges through creating an integrated team that actively engaged, communicated and coordinated each phase of the project to ensure successful completion. This paper will cover how these key elements of the Sludge Removal Project led to its success: - Integrating project planning and management; - Lessons learned from design and construction; - Preparing for a successful campaign (lessons learned from testing and start-up); - Lessons learned from operating in two separate facilities, transporting sludge containers on the Hanford Site and ensuring compliance with a range of regulatory and safety basis requirements. The establishment of an integrated group spanning two facilities and multiple organizations created a team capable of overcoming the challenges necessary to successfully plan and execute the sludge retrieval mission. Sharing lessons learned from this successful project can enhance the ability of teams across the DOE complex to successfully plan and execute complex projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Trace Chemical Detection Using Intercalated MXenes as a Signal Enhancing Substrate in Optical Probes

MXenes are 2D materials composed of layered transition metal nitrides or carbides. These materials are synthesized by HF exfoliation from MAX phases (Ti{sub 3}AlC{sub 2}). The 2D nanomaterial was synthesized by the removal of the 'A' element, resulting in a Mxene product (Ti{sub 3}C{sub 2}). MXenes have the general formula M{sub n+1}X{sub n}T{sub x}, where M is an early transition metal, X is Carbon and/or Nitrogen, and T accounts for surface terminated functional groups such as Fluoride, hydroxyl, and oxygen. These materials have very unique properties, similar to graphene, that allows them to be applied in a variety of trace detection techniques including surface-enhanced Raman spectroscopy (SERS). MXenes have also been demonstrated to selectively uptake uranyl ion, UO{sub 2}{sup 2+}. If this property can be combined with SERS or fluorescence detection, it may be possible to use MXenes as the basis for an alternative method to kinetic phosphorescence analysis (KPA) for trace uranyl measurements. Objectives: To confirm that MXene Nano materials are suitable substrates for SERS and sensor development by enhancing Raman signaling. To determine if certain MXene preparation methods yield materials that are more suitable for trace sensing methods. To determine uranyl uptake properties of these MXene materials and test them for analytical signals. Sample Preparation: Preparation of Ti{sub 3}C{sub 2}MXene (at FSU). MXenes were prepared by etching Al from Ti{sub 3}AlC{sub 2} (MAX phase)material. Two etching techniques yield different MXene products: LiF/HCl: Milder reaction, larger MXene flakes. HF: Harsher reaction, smaller flakes, larger layer separation. Products washed to remove etchant, vacuum filtered, and dried. Dried MXene flakes are air-stable. Film preparation for sensor testing (at SRNL): Suspend powder in diH{sub 2}O, purge with Ar, sonicate for 30 min. Centrifuge and collect supernate with suspended particles. Observed LiF-etched Mxene yielded a higher density of particles and darker collected solution. Drop-cast (4 ml) supernate onto slides and dried with Ar. For Rhodamine B (RhB) testing, drop-cast 4 ml drops onto Mxene spots and dried with Ar. Scanning Electron Microscopy conditions: 10 kV Beam energy, high vacuum; Working distance of 8 mm; beam penetration depth appx. 4 microns, beam spot size appx. 2 nanometers. Results: Detection of aluminum correlates with bright spots on image. Presence of aluminum shows that LiF/HCl etching was less thorough than HF etching. Trace Cl detection in LiF images suggests incomplete rinsing. HF has smaller feature size, more layer structure, and increased homogeneity, consistent with expectations. Macroscopic Raman spectroscopy measurements: 532 nm excitation, ∼50 mW with a ∼100 micron spot size (InPhotonics RPB probe). Kaiser Optical Holospec f/1.8 spectrometer with cooled (-60 deg.C) Andor iDus OE420 CCD. LiF 1x supernate showed good signal for trace measurements of Rhodamine B. HF and 1/4x LiF supernates showed little Mxene or Rhodamine B signal. Low deposition densities led to excess background signal from glass slides. For LiF film, response is linear with Rhodamine B concentration over range tested. Will retest with Raman microscope (∼1 micron spot size) to characterize SERS of more dilute LiF and HF etched Mxenes. Conclusions: The LiF etched material was more suitable for macroscopic SERS measurements because it was more concentrated, resulting in a thicker film than the HF etched Mxene and diluted LiF sample. However, the other materials may give greater SERS enhancements, which we hope to determine from measurements with the Raman microscope. From characterization with SEM we concluded that the HF etched Mxene is more uniform/homogenous and has smaller particle size than the LiF etched Mxene. There is still aluminum present in both samples indicating that etching wasn't complete, but the removal of the aluminum was more efficient in the HF method. Path Forward: Observe SERS with Raman microscopy, to obtain better signals for the more diluted samples and be able to compare enhancement effects for the different MXenes. Characterize uranyl sorption into MXene films and test Raman and fluorescence signals. Revisit the etching conditions to improve removal of aluminum. FSU and SRNL will continue to collaborate to create and characterize different Mxene materials and test their usefulness for sensor applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Improving manganin gauge measurements near radiographic fiducials

In dynamic experiments diagnosed with radiography and stress gauges, the gauges are often placed adjacent to radio-graphic fiducials, such as Ta foil, which perturbs the stress measurement. We conducted experiments to test the performance of different arrangements of manganin stress gauges and Ta foils. Three different gauge/foil configurations were tested, as well as a solitary gauge as control. Gauge/foil layers were placed between high-density polyethylene (HDPE) discs and impacted by a projectile in a gas gun, a simplification of a dynamic, high explosive experiment configuration. The stress measurements were minimally perturbed when a section of foil was removed near the active element of the gauge, as expected. In conclusion, a follow-up experiment showed that the modification to the foil has minimal impact on the radiographic measurement of the dynamic flow.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Stochastic Approaches for Calculating and Aggregating Detection Probabilities for Nuclear Material Diversion

The authors built and tested a stochastic simulation to estimate achieved detection probabilities (DPs) on a stratum basis, over a tailorable range of diverted amounts from 0 to 2 significant quantities (SQ), using typical International Atomic Energy Agency (IAEA) inspection data: i.e., SQ in stratum, number of items, number of gross/partial/bias defect measurements conducted, and realistic relative standard deviation (RSD) values for typical IAEA verification measurements. For bulk strata, the model calculates achieved DP at 0.01 SQ diversion increments; for item strata, the model calculates DP using the smallest realistic diversion increment (e.g., a plate, pin, or coupon). After successfully benchmarking against IAEA deterministic models, the simulation was used to test the sensitivity of DP to certain standard assumptions and selected input parameters. First, the equal defect assumption was tested; the results suggest significant complexity in the effectiveness of partial defect measurements. Next, the authors explored the sensitivity of DP to the assumed RSD of attribute tests. Then, the authors compared non-normal models for instrument performance (e.g., logistic, step, or arbitrary functions) to the typical results from a normal distribution (characterized by RSD). This last comparison was supplemented with experimentally derived performance data for an HM-5 gamma spectrometer. The HM-5 was used to make enrichment measurements on both LEU and HEU MTR fuel elements as plates were removed, and the results fit with logistic and step curves and applied in the simulation. These stochastic DP results were compared to DP estimates from a deterministic model assuming a normal curve and typical RSD, yielding insights that could improve effectiveness in the field. These early results illustrate the potential of stochastic models to better understand achieved DP and to improve safeguards effectiveness.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Evaluating Safeguards Statistical Assumptions via Stochastic Simulation

Herein, the authors built and tested a stochastic simulation to estimate achieved detection probabilities (DPs) on a stratum basis, over a tailorable range of diverted amounts from 0 to 2 SQ, using typical IAEA inspection data: i.e., SQ in stratum, number of items, number of gross/partial/bias defect measurements conducted, and realistic relative standard deviation (RSD) values for typical IAEA verification measurements. For bulk strata, the model calculates achieved DP at 0.01 SQ diversion increments; for item strata, the model calculates DP using the smallest realistic diversion increment (e.g., a plate, pin, or coupon). After successfully benchmarking against IAEA deterministic models, the simulation was used to test the sensitivity of DP to certain standard assumptions and selected input parameters. First, the equal defect assumption was tested; the results suggest significant complexity in the effectiveness of partial defect measurements. Next, the authors explored the sensitivity of DP to the assumed RSD of attribute tests. Then, the authors compared non-normal models for instrument performance (e.g., logistic, step, or arbitrary functions) to the typical results from a normal distribution (characterized by RSD). This last comparison was supplemented with experimentally derived performance data for an HM-5. The HM-5 was used to make enrichment measurements on both LEU and HEU MTR fuel elements as plates were removed, and the results fit with logistic and step curves and applied in the simulation. These stochastic DP results were compared to DP estimates from a deterministic model assuming a normal curve and typical RSD, yielding insights that could improve effectiveness in the field. These early results illustrate the potential of stochastic models to better understand achieved DP and to improve safeguards effectiveness.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗

Removal and Deactivation of Bond Sodium from Fast Reactor Blanket Materials

The disposition of sodium-bonded spent nuclear fuel and blanket materials in a repository is complicated by the presence of sodium metal that is used as a thermal bond between the uranium metal fuel and blanket slugs and their cladding. The concern is that the metallic sodium could react with water, producing explosive hydrogen gas, or could exhibit a pyrophoric character. Thus, experimental studies were performed to investigate and demonstrate the removal and deactivation of bond sodium from blanket material in a dry environment. Specifically, bond sodium was removed from unirradiated Fermi-1 blanket elements and an assembly via a melt-drain-evaporate process using elevated temperature and reduced pressure. The effectiveness of sodium metal removal from the blanket materials and their associated cladding was =99.9998%, based on post-test quantitative analyses. The separated sodium metal was collected and subsequently deactivated by reacting it in a molten state with a controlled addition of ammonium chloride particles atop a molten salt medium. In this process sodium chloride is formed and assimilates into the salt pool. The subsequent deactivation of the bond sodium produced a solid ingot of sodium chloride, potassium chloride, lithium chloride, and cesium chloride that was devoid of sodium metal based on post-test analyses of the salt product. Both the sodium removal and deactivation operations were conducted within a dry inert atmosphere enclosure. The results of this study substantiate a path forward for the disposition of sodium-bonded blanket materials, including 34 metric tons heavy metal in irradiated Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Conceptual Fuel Element Design Candidates for Conversion of High Flux Isotope Reactor with Low-Enriched Uranium Silicide Dispersion Fuel

Engineering design studies are underway to assess the feasibility of converting the High Flux Isotope Reactor (HFIR) to operate with low-enriched uranium (LEU) silicide dispersion (U3Si2-Al) fuel. These studies are supported by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. A systematic approach employing neutronic and thermal-hydraulic analyses have been performed with the ORNL Shift and HFIR Steady State Heat Transfer Code tools, respectively, to predict reactor performance and thermal safety margins for proposed LEU3Si2-Al fuel designs. The design process was initiated by generating an optimized design with fabrication features identified from previous studies that result in excellent performance and safety metrics. The approach continued by substituting a single fabrication feature anticipated to be difficult to manufacture with another feature expected to perform an analogous function to that of the removed feature. Four conceptual fuel element design candidates, with various fabrication features, for conversion of HFIR with 4.8 gU/cm3 LEU3Si2-Al fuel have been generated and shown to meet pre-defined performance and safety metrics. Results to date indicate that HFIR could convert with the subject fuel system and meet performance and safety requirements if, among other considerations, fabrication of the specific design features are demonstrated and qualification of the fuel is complete under HFIR-specific conditions.

Chandler, David↗

The Architecture of Area 12 Camp - Nevada's Atomic Ghost Town. An Architectural Survey of Area 12 Camp, Nevada National Security Site, Nye County, Nevada

The U.S. Department of Energy, National Nuclear Security Administration Nevada Field Office (NNSA/NFO) plans to demolish 10 buildings and one structure in Area 12 Camp at the Nevada National Security Site (NNSS) in Nye County, Nevada, to meet current and future National Weapons Science, Global and Homeland Security Program, and Environmental Management mission requirements and to reduce dangers to site workers from some of these resources. These demolitions constitute an undertaking subject to review under Title 54 U.S.C. § 306108, commonly known as Section 106 of the National Historic Preservation Act, Title 54 U.S.C. § 300101, et seq., and its implementing regulations, 36 C.F.R. Part 800. These resources were recorded and evaluated as individually ineligible for the National Register of Historic Places (National Register or NRHP) by Menocal and Shaw (2019). Upon review by the Nevada State Historic Preservation Office (SHPO) and internal review by Desert Research Institute (DRI) and NNSA/NFO staff, it was determined that these resources needed to be rerecorded and evaluated in relation to what appeared to be a National Register-eligible Historic District made up of the entirety of Area 12 Camp. Based on preliminary information, it appeared that all 11 resources would likely be contributing elements of that historic district. With that assumption in mind, recording of the entire proposed Area 12 Camp Historic District was done for three purposes. First, it would result in a definitive significance evaluation of the 11 resources in question. Second, recording of the entire camp would be offered as an appropriate mitigation measure for the demolition of these resources. Third, it would present an essential background for future management of resources within the district boundaries. Prior to the present survey, Area 12 Camp had not yet been systematically recorded; therefore, DRI surveyed an area of approximately 229 acres for architectural resources. This effort resulted in the identification, recording, and evaluation of the Area 12 Camp Historic District (SHPO Resource No. D372), including the identification of its contributing components. This district is recommended eligible for the National Register of Historic Places under Criteria A and C. It is unevaluated under Criteria B and D. The Area 12 Historic District contains 71 Principal Resources, which include landscapes, buildings, and structures. Of these resources, 69 (including 10 of the 11 resources to be removed) are recommended as contributing elements of the historic district during its period of significance corresponding to nuclear testing from 1960 through 1992. During most of the Cold War, the government reservation now called the NNSS was named the Nevada Test Site (NTS).

54 ENVIRONMENTAL SCIENCES↗

An energy-based study of the embedded element method for explicit dynamics

Abstract The embedded finite element technique provides a unique approach for modeling of fiber-reinforced composites. Meshing fibers as distinct bundles represented by truss elements embedded in a matrix material mesh allows for the assignment of more specific material properties for each component rather than homogenization of all of the properties. However, the implementations of the embedded element technique available in commercial software do not replace the material of the matrix elements with the material of the embedded elements. This causes a redundancy in the volume calculation of the overlapping meshes leading to artificially increased stiffness and mass. This paper investigates the consequences in the energy calculations of an explicit dynamic model due to this redundancy. A method for the correction of the edundancy within a finite element code is suggested which removes extra energy and is shown to be effective at correcting the energy calculations for large amounts of redundant volume.

Martin, Valerie A.↗

R-Value Measurements Performed on Actinide Targets Irradiated using the GODIVA IV Critical Assembly in FY22

The separation and characterization of two irradiated uranium targets, a depleted uranium (DU) and a highly enriched uranium (HEU) target as well as a plutonium (Pu) target, was conducted in April of 2022. The three targets were assembled at Los Alamos National Laboratory (LANL) and irradiated using the GODIVA critical assembly at the National Criticality Experiments Research Center (NCERC). Splits of the dissolved targets were received by Pacific Northwest National Laboratory (PNNL) after which the PNNL and LANL teams chemically separated the solutions using independent separation schemes and analyzed the separated fractions for short lived actinides and fission products. Chemical separations were traced with stable or radioactive tracers to allow for the determination of chemical yields, analyzing using either inductively coupled plasma optical emission spectroscopy (ICP-OES), inductively coupled plasma mass spectrometry (ICP-MS) or gamma emission analysis (GEA) depending on the nature of the tracer. The Pu target solution was traced with stable elements at LANL to follow elemental fractionation during a Pu removal step. Many analytical techniques were used by PNNL including kinetic phosphorescence analysis (KPA), ICP-OES, ICP-MS, GEA, and thermal ionization mass spectrometry (TIMS) depending on the analyte’s need.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Actinide Elemental Ratios of Spent Nuclear Fuel Samples by Resonance Ionization Mass Spectrometry

While resonance ionization mass spectrometry (RIMS) has demonstrated utility in measuring isotopic compositions of elements in complex matrices without the need for chemical separation to remove isobaric interferences, it has had limited application in measuring elemental compositions. The ability to determine elemental compositions via an in situ method like RIMS would be an exceptional asset in spent nuclear fuel analysis, where they are important in assessing reactor histories and whose chemical separation presents a radiological hazard. However, quantitative elemental analysis by RIMS requires special considerations because each element is ionized by its own set of lasers tuned to element specific resonant ionization wavelengths. We present the first comprehensive study of measuring elemental ratios by RIMS in spent nuclear fuel. All actinides produced by neutron capture are enhanced significantly radially from the center to the edge of a fuel pellet. This edge effect is not readily accessible by conventional bulk measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Tuple checkout with notify in coordination namespace system

A system and method for notifying a process about a creation or removal event of a named data element (NDE) in a coordination namespace distributed memory system. A controller runs methods to: generate a tuple corresponding to data generated by a requesting process, the tuple having a tuple name and data value; and generate a notification indicator in a pending notification list to indicate to one or more processes a notification of the creation or removal event associated with the corresponding tuple. Upon detecting the event performed on the tuple by a second process, the method further searches for NDEs in the distributed memory system having the same tuple name, and in response to determining an existence of an associated pending notification record in a pending notification list of records, notify each corresponding process of the one or more processes indicated in the list of the creation or removal event.

Jacob, Philip↗

Sampling and Characterization of Coal Basins as an Alternative Rare-Earth Element Resource

Conference presentation at TechConnect World Innovation Conference & Expo, Washington, D.C., June 13–15, 2022. The University of North Dakota (UND) through the Energy & Environmental Research Center (EERC) and the Institute for Energy Studies teamed with the University of Kentucky Center for Applied Energy Research, Microbeam Technologies Inc., the Kentucky Geological Survey (KGS), the North Dakota Geological Survey (NDGS), and the mining industry to execute a contract with the U.S. Department of Energy National Energy Technology Laboratory (NETL) to obtain and characterize samples of U.S. domestic precombustion coal and coal-related materials with a minimum concentration of 300 parts per million (ppm) total rare-earth elements (TREEs) as the material is removed from the ground, with no processing other than drying.

01 COAL, LIGNITE, AND PEAT↗

Removal of Iron from Pyrite-Rich Coal Refuse by Calcination and Magnetic Separation for Hydrometallurgical Extraction of Rare Earth Elements

In the metallurgical extraction of rare earth elements (REEs), the ratio of contaminant ions to REEs in the leachate dictates the cost and operational efficiency of the downstream processes. The current study investigated the potential iron contamination removal from the feed to the hydrometallurgical process by calcination followed by magnetic separation. The 2.20 specific gravity sink fraction of Baker coal seam coarse refuse was pulverized to finer than 180 μm, calcined at various temperatures, and separated into magnetic and non-magnetic fractions using a wet high-intensity magnetic separator at different field strengths. The untreated feed, calcined products, and their subsequent magnetic and non-magnetic fractions were subjected to acid leaching tests with 1.2 M sulfuric acid at 75 °C and 1% w/v solids concentration. The recovery of light and heavy rare earth elements (LREEs and HREEs, respectively) along with the concentration of common contaminant ions (Al, Ca, and Fe) were measured as output variables. The weight percent of magnetic material was maximized at approximately 29% by calcination at a temperature of 400 °C. Magnetic removal of this fraction using a field strength of 1.15 Tesla resulted in the rejection of 81% of the iron. Leaching of the magnetic fraction provided significantly higher Fe recovery relative to untreated feed material and the non-magnetic fraction. The non-magnetic fraction was subsequently calcined at 600 °C to dehydroxylate the clays and released the REE minerals in the same manner as the treatment of the original coarse refuse material. A comparison of the leachate elemental concentrations resulting from the leaching of both the calcined non-magnetic and original coarse refuse showed only a slight reduction in the iron content from the non-magnetic material. This finding combined with the REE loss in the magnetic fraction resulted in the conclusion that the magnetic removal step was unfavorable.

58 GEOSCIENCES↗