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At least 37 records · Page 2

Small Punch Testing of Molybdenum-99 Targets

Northstar Medical Radioisotopes is developing an accelerator-based method to produce 99 Mo, which is a parent isotope of the commonly used 99m Tc medical isotope. The Mo targets being designed for the accelerator will be produced from enriched 100 Mo, also known as aMo . Pressed and sintered powder feedstock is used to fabricate aMo targets, producing 29 mm disk-shaped targets. The targets are subjected to 1–6 days in line of an electron beam with subsequent dissolution of the disk to retain the 99 Mo, which decays to 99m Tc at radio-pharmacies. The press and sinter method is advantageous because the inherent porosity produced by this method enables increased surface area and therefore increased flow of dissolution media, decreasing the dissolution time and reducing the need for a highly acidic media. Although porosity aids in dissolution, it reduces the mechanical strength and ductility. Targets require good mechanical integrity when subjected to the conditions in the accelerator. Therefore, Northstar is seeking methods to rapidly test disk samples after fabrication to assure mechanical performance metrics are achieved. This report details the design and testing of a small punch test (SPT) that accommodates the 29 mm disk. Initial data were used to relate the SPT data to tensile properties, such as the yield strength (YS), ultimate tensile strength (UTS), and total elongation to failure. Although the SPT has been established as a somewhat reliable method for testing metallic materials, few studies have applied the SPT to refractory materials such as Mo. Based on tensile testing performed at Oak Ridge National Laboratory on different Mo samples, correlation between the Mo tensile and SPT properties could be performed, establishing a standard calibration that could be applied to other Mo samples. To test the efficacy of the SPT with Mo, multiple different disk batches were fabricated under different conditions (e.g., pressure, lubricant) with commercially available pure Mo powder. Generally, only a UTS could be well defined because the press and sinter disks failed under brittle fracture, making it difficult to determine the YS and elongation. Compared with disks fabricated with aMo powder, the aMo samples underperformed their pure nat Mo counterparts. This report summarizes the current status of the SPT, but further evaluation is needed before it can be applied as a reliable quality assurance tool.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

Purification and Concentration of 99 Mo from a UREX raffinate

The medical isotope 99m Tc ( t ½ = 6.0 h) accounts for over 80% of isotopes used in diagnostic nuclear medicine today and is normally derived from its transient equilibrium parent 99 Mo ( t ½ = 66 h). To reduce U.S. dependence on 99 Mo derived from HEU by foreign suppliers, the NNSA/M3 program—under the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not utilize HEU. Superconducting electron linear accelerators with high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and LEU fission. After sufficient production intervals, targets can be rotated out and processed while another batch is irradiated. The process flow chemistry is shown in Figure 1. The irradiated U 3 O 8 targets are retrieved and dissolved in HNO 3 ; the volatile fission products are expelled and captured during this step. The HNO 3 liquor bearing the uranyl, fission and activation products is injected into the UREX suite of liquid-liquid extraction banks. The uranyl is partitioned in typical fashion using tri-$\textit{n}$-butyl phosphate (TBP) in a hydrocarbon diluent. The raffinate—comprising mainly 99 Mo and fission products—is then injected into the MoLLE (Molybdenum Liquid-Liquid Extraction) flowsheet, where the Mo is selectively extracted by an organophosphorous acid extractant such as di-(2-ethylhexyl phosphoric acid) (HDEHP). Trace amounts of Nb, Zr, Np, I, and Te are co-extracted. The Mo is stripped using acetohydroxamic acid (AHA) and fed onto an anion exchange column. Following a series of hydroxide, HCl, and oxalic acid wash steps, the Mo can be recovered in NaOH/NaCl. A manuscript describing this process chemistry in more detail was recently published.

07 ISOTOPE AND RADIATION SOURCES↗

Measurements of excitation functions and photoneutron cross sections of 96,98,99,104 Ru and 100 Mo

Photoneutron cross sections were extracted for isotopes of ruthenium and molybdenum on natural abundance targets via a reduced chi-squared analysis on excitation functions, or activation yields, as function of electron beam energy. Bremsstrahlung photons were produced via an electron linear accelerator and a tungsten radiator at the Idaho Accelerator Center. A total of 16 irradiations were performed with electron beam energies ranging from approximately 8 MeV to 23 MeV. Induced radioactivity was measured using a high purity germanium detector. The cross sections were extracted assuming a three parameter Lorentzian fit with a smooth truncated rise at reaction threshold.

07 - ISOTOPES AND RADIATION SOURCES↗

Development and Optimization of a Purification Process to Recover 99 Mo from Low-enriched Uranium

Supported by the DOE-NNSA Office of Material Management & Minimization (M3), a number of domestic entities are pursuing non-highly enriched uranium (non-HEU) production of 99 Mo. As the production technologies of 99Mo pivot toward low-enriched uranium (LEU) or molybdenum targets, new reaction channels and accelerators are being evaluated. Superconducting electron linear accelerators (LINACs) with high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that are capable of inducing photonuclear reactions and LEU fission. A particular advantage of a LINAC is that it does not rely on HEU-fueled reactor cores (which are currently slated for LEU conversion) and can operate on an almost continuous basis. Regarding the chemical purification of 99 Mo from irradiated uranium targets under acidic digestion, there exists a procedure known as Cintichem or modifications thereof with respect to LEU (LEU Modified Cintichem process, LMC). The process relies on a number of selective precipitation steps and column chromatography to purify Mo. It is important to note that LMC prescribes the addition of stable Mo to carry 99 Mo on alpha benzoin oxime, which reduces the specific activity of 99 Mo. This is especially important for processing 99 Mo batches with lower activities (~33 Ci of 99 Mo per batch).

07 ISOTOPE AND RADIATION SOURCES↗

Effect of Prestraining of Recrystallization Temperature and Mechanical Properties of Commercial, Sintered, Wrought Molybdenum

Given three presumably identical lots of commercial, sintered, wrought molybdenum, the 1-hour recrystallization temperature of one lot remained above 2900 F by limiting the amount of effective restraining to 35 percent or less. Different recrystallization temperatures were obtained in various atmospheres, the highest in argon and the lowest in hydrogen. Metal thus fabricated and then stress-relieved possessed an ultimate tensile strength at room temperature within 10 percent of metal swaged 99 percent and also possessed equivalent ductility. At 1800 F, equivalent strength and ductility was obtained irrespective of the amount of swaging over the range of 10 to 99 percent. The amount of swaging greatly influenced the recrystallized grain size but the difference in grain size is not the major controlling factor which determines whether recrystallized molybdenum is ductile or brittle at room temperature.

MATERIALS, PROPERTIES - TENSILE↗

Separation and determination of ultratrace rhenium quantities in molybdenum matrix

The production of very pure molybdenum, free of Re even at ultratrace levels, is extremely important in meeting target purity specifications during the production of 99 Mo for medical applications. Here we propose two methods for separating ultratrace levels of Re from a bulk Mo matrix using either solvent extraction or extraction chromatography, combined with inductively coupled plasma quadrupole mass spectrometry (ICP-MS) detection. Re is extracted (D > 70) from solutions in sulfuric acid as a complex with tri-n-butyl phosphate. Scrubbing and washing steps result in total decontamination factors for Mo up to 1,700 for solvent extraction and 3,500 for extraction chromatography. Re is stripped into a 3 M NH 4 OH matrix and analyzed by ICP-MS. Detection (L D ) and quantification (L Q ) limits were optimized by matrix-matching to allow detection of Re in strip solutions at levels of L D = 0.1 ng Re/g-Mo and L Q = 0.3 ng Re/g-Mo in Mo powder samples. By employing these two extraction methods, excellent recoveries of Re from bulk Mo are achieved, and the L Q is improved by a factor of 5 x 10 4 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Sorption of Anionic Iodine and Molybdenum to Multi-sorbing Barrier (MSB) Materials - 20030

A multi-sorbing barrier (MSB) material containing sodium bentonite (NaB), organoclay, and anion resin is being evaluated for containment of anionic radionuclides. Batch tests on the MSB were conducted using two non-radioactive analog isotopes ({sup 127}I, {sup 95}Mo) as surrogates for {sup 129}I and {sup 99}Tc due to their similar valence and molecular weight relative to the radionuclides. Tests were conducted on the organoclay, anion resin, and NaB individually and on NaB-resin blends with resin contents of 1% and 5% by mass. The anion resin exhibits very strong sorption of {sup 127}I and {sup 95}Mo relative to the other materials. The organoclay exhibited modest sorption of {sup 95}Mo and {sup 127}I, and NaB little sorption. The sorption isotherms follow a linear model for NaB, a Freundlich model for the organoclay, and a Langmuir model for the anion resin and NaB-resin mixtures. Combining resin with NaB significantly enhances sorption relative to NaB alone, with a larger resin fraction yielding greater sorption. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Recovery of high specific activity molybdenum-99 from accelerator-induced fission on low-enriched uranium for technetium-99m generators

Abstract A new process was developed to recover high specific activity (no carrier added) 99 Mo from electron-accelerator irradiated U 3 O 8 or uranyl sulfate targets. The process leverages a novel solvent extraction scheme to recover Mo using di(2-ethylhexyl) phosphoric acid following uranium and transuranics removal with tri- n -butyl phosphate. An anion-exchange concentration column step provides a final purification, generating pure 99 Mo intended for making 99 Mo/ 99m Tc generators. The process was demonstrated with irradiated uranium targets resulting in more than 95% 99 Mo recovery and without presence of fission products or actinides in the product.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Control of iodine and tellurium for use in domestic molybdenum-99 production efforts

Three initial proof-of-concept experiments were performed to test the removal of I and Te from solutions containing Mo for potential abatement during chemical processing to purify 99 Mo. All three proof-of-concept experiments produced results indicating that further optimization could provide methods for I and/or Te abatement from multiple matrices. Decamp and Happel (2013) describe the use of CL resin from Eichrom Technologies loaded with Ag + combined with Amberlite XAD-4 in the chromatographic separation of irradiated fission products in 1 M HNO 3 to separate iodine species where iodide and iodate react with Ag + and I2 adheres to the Amberlite resin. The method worked well under processing conditions in a medical isotope facility although many separation methods use higher concentrations of HNO 3 which could adversely impact the Ag + loading. The experiment performed here used a roughly 2 mL column of CL resin loaded with Ag + and Amberlite XAD-4 to separate µg quantities of I, Te, and Mo in 3 M HNO 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Field-Scale Lysimeter Studies of Glass and Cementitious Waste Forms at the Hanford Site - 20392

The Hanford site Integrated Disposal Facility (IDF) will receive waste forms from vitrification activities at the Hanford Waste Treatment and Immobilization Plant (WTP). The waste form inventory to be disposed of at the IDF will consist of vitrified low-activity waste (LAW) in glass forms and solidified (or encapsulated) secondary wastes in cementitious forms. The IDF is a near-surface burial facility located near Hanford's Waste Treatment and Immobilization Plant in the 200 East Area of the site's Central Plateau. An extensive set of laboratory experimental data has been collected to support IDF performance assessment calculations of the eventual degradation of waste forms and mobility of contaminants. This paper describes the start of a field experiment to generate waste-form performance data on a larger scale (tens of centimeters), over a longer duration (five years or more), and under field conditions representative of the IDF. Results from this study are expected to improve model descriptions of contaminant mobility, reduce uncertainties about the representativeness of laboratory results in the IDF performance assessment, improve stakeholder confidence in the safe disposal of treated waste at the IDF, and facilitate the adoption of informed facility designs with the potential to reduce operational costs. A lysimeter test facility near Hanford's 200 West Area has been repurposed to carry out long-term assessments of waste forms buried in sediments excavated from the IDF. The waste forms are buried in large drainage lysimeters, open caissons that are two meters in diameter and three meters deep. Waste forms are buried in up to six locations in each lysimeter. Samplers collect pore water and drainage water beneath the waste forms and air from the pore space adjacent to waste forms. Sensors measure temperature, soil water content, soil water tension, and drainage flux. One lysimeter containing cementitious waste forms and one containing glass waste forms have been completed, along with a control lysimeter (without waste forms). The cementitious waste forms are Cast Stone grout formulations using a liquid secondary waste simulant and a generic Hanford high-salt simulant, as well as a hydrated lime-based grout formulation using a liquid secondary waste simulant. Technetium-99 and iodine-127 were added as tracers to monitor contaminant mobility. Glass waste forms were fabricated in two formulations (LAWA44 and ORLEC28), with rhenium and molybdenum tracers to monitor waste form degradation. The predicted performance of the waste forms in the lysimeters was modeled prior to installation to determine waste form size, tracer concentration/activity, and locations of sensors and samplers. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3

Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.

07 ISOTOPE AND RADIATION SOURCES↗

Baseline Hypothetical Facility for the Production of 131 I, 99 Mo, and 133 Xe from HALEU Fission Targets

The report describes the development of a hypothetical facility to produce the pharmaceutical radioisotope 131 I at an amount of 60 curies per week via the fission of a High Assay Low Enriched Uranium (HALEU) target, along with the chemical and physical processes and equipment needed to separate the 131 I and co-produced 99 Mo and 133 Xe. The hypothetical design was carried out using a 10 MWt research reactor. The irradiation calculation determined that three HALEU targets with aluminum cladding can be used to produce 60 Ci/week of 131 I, 750 Ci/week of 99 Mo and 560 Ci/week of 133 Xe. The process selected for the baseline design uses caustic dissolution of the target material with ion exchange processes to separate and purify the iodine and molybdenum. The xenon is processed using a cryogenic carbon bed separation process. Target processing occurs in seven shielded hot cells with a total footprint of 16m 2 and waste management occurs in an eight hot cell with a 2.5 m 2 footprint. Waste generated from the processing of the three targets per week would generate less than 4 shielded drums of waste annually. These hot cells would need varying levels of shielding due to the amount of fission products being handled in the unit. Hot cell facilities also require support services including QA/QC, health physics, administrative staff, operator changing room, radiological buffer areas, and waste storage. The overall facility would require a footprint of 1050 m 2 with 20.25 m 2 of shielded hot cells.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

Efficient Hydrogen Delivery for Microbial Electrosynthesis via 3D-Printed Cathodes

The efficient delivery of electrochemically in situ produced H 2 can be a key advantage of microbial electrosynthesis over traditional gas fermentation. However, the technical details of how to supply large amounts of electric current per volume in a biocompatible manner remain unresolved. Here, we explored for the first time the flexibility of complex 3D-printed custom electrodes to fine tune H 2 delivery during microbial electrosynthesis. Using a model system for H 2 -mediated electromethanogenesis comprised of 3D fabricated carbon aerogel cathodes plated with nickel-molybdenum and Methanococcus maripaludis, we showed that novel 3D-printed cathodes facilitated sustained and efficient electromethanogenesis from electricity and CO 2 at an unprecedented volumetric production rate of 2.2 L CH4 /L catholyte /day and at a coulombic efficiency of 99%. Importantly, our experiments revealed that the efficiency of this process strongly depends on the current density. At identical total current supplied, larger surface area cathodes enabled higher methane production and minimized escape of H 2 . Specifically, low current density (<1 mA/cm 2 ) enabled by high surface area cathodes was found to be critical for fast start-up times of the microbial culture, stable steady state performance, and high coulombic efficiencies. Our data demonstrate that 3D-printing of electrodes presents a promising design tool to mitigate effects of bubble formation and local pH gradients within the boundary layer and, thus, resolve key critical limitations for in situ electron delivery in microbial electrosynthesis.

03 NATURAL GAS↗

Crystal structure and magnetic properties of CeFe 11 M (M = Mo, W)

Tetragonal CeFe 11 Mo (struct. type ThMn 12 , a = 8.529(1) Å, c = 4.766(1) Å, V = 346.68(1) Å 3 ) and novel isostructural CeFe 11 W (a = 8.539(1) Å, c = 4.754(1) Å, V = 346.62(1) Å 3 ) are obtained in bulk, as homogenized single-phase arc-melted alloys. Here, their crystal structures are evaluated by full profile Rietveld analysis of the powder X-ray diffraction patterns which confirm that Ce occupies the 2a atomic site in both cases, whereas atomic disorder is present in the Fe-sublattice, i.e., predominantly on the dumbbell 8i site and minorly on two other sites 8j and 8f. In CeFe 11 Mo the third element, M, is mixed with Fe on the 8i dumbbell site and 8f atomic position, whereas in CeFe 11 W, M is mixed on the 8i and 8j Fe sites. In these bulk CeFe 12-x M x alloys the tetragonal structure persists at lower M content, x < 1 here, in comparison to known similar alloys, with x > 1. Because of the low M content, comparably higher values of mass magnetization and Curie temperature are achieved i.e., CeFe 11 Mo: ϭ s = 104 emu/g (14.7 µ B /f.u.), T c = 420 K and CeFe 11 W: ϭ s = 99 emu/g (16.5 µ B /f.u.), T c = 450 K. Magnetocrystalline anisotropy is uniaxial with anisotropy fields H a ∼ 14.5 and 16.5 kOe, respectively. Based on Curie temperature and magnetic anisotropy enhancements, using tungsten as the stabilizing dopant is more beneficial than molybdenum.

36 MATERIALS SCIENCE↗