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At least 19 records
All sourcesNASA NTRSDOE OSTI
NASA NTRS2017

Experimental Observations of Nuclear Activity in Deuterated Materials Subjected to a Low-Energy Photon Beam

Exposure of highly deuterated materials to a low-energy (nom. 2 MeV) photon beam resulted in nuclear activity of both the parent metals of hafnium and erbium and a witness material (molybdenum) mixed with the reactants. Gamma spectral analysis of all deuterated materials, ErD2.8+C36D74+Mo and HfD2+C36D74+Mo, showed that nuclear processes had occurred as shown by unique gamma signatures. For the deuterated erbium specimens, posttest gamma spectra showed evidence of radioisotopes of erbium ((163)Er and (171)Er) and of molybdenum ((99)Mo and (101)Mo) and by beta decay, technetium (99mTc and 101Tc). For the deuterated hafnium specimens, posttest gamma spectra showed evidence of radioisotopes of hafnium (180mHf and 181Hf) and molybdenum ((99)Mo and (101)Mo), and by beta decay, technetium ((99m)Tc and (101)Tc). In contrast, when either the hydrogenated or non-gas-loaded erbium or hafnium materials were exposed to the gamma flux, the gamma spectra revealed no new isotopes. Neutron activation materials showed evidence of thermal and epithermal neutrons. CR-39 solid-state nuclear track detectors showed evidence of fast neutrons with energies between 1.4 and 2.5 MeV and several instances of triple tracks, indicating (is) greater than 10 MeV neutrons. Further study is required to determine the mechanism causing the nuclear activity.

Steinetz, Bruce M.↗
DOE OSTI2021

Corrosion of 347 stainless steel in the presence of uranyl sulfate solution and radiation

The U.S. molybdenum 99 (Mo-99) industry is pursuing production of fission-made Mo-99 using a uranium solution such as uranyl sulfate. In this process, uranyl sulfate solution containing low-enriched uranium will be bombarded by neutrons creating Mo-99 and other fission products. During the production, the uranyl sulfate solution will be irradiated until an acceptable activity level of Mo-99 is produced. The uranyl sulfate solution containing Mo-99 and other fission products will then undergo a series of separation steps. First, uranyl sulfate can be separated from Mo-99 using a primary titania column to recover Mo-99, with the uranyl sulfate solution to be used for another irradiation cycle. Then, raffinate from a primary titania column containing Mo-99 can be concentrated and purified using a LEU modified Cintichem process developed by Argonne National Laboratory. During irradiation, the temperature of the uranyl sulfate solution can reach near boiling (up to ~80° C assumed), causing radiolysis of water and the resultant formation of hydrogen peroxide. Because high-radiation fields will be present during each irradiation cycle, it is important to determine the corrosion rates of SS-347 under such conditions to estimate the life cycle of the target solution vessel. The buildup of corrosion products from the SS components in the uranyl sulfate solution also needs to be well understood because potential accumulation of iron, nickel, and other corrosion products may affect the Mo-99 recovery and purification process. To study the corrosion rates of SS-347 material under conditions relevant to future Mo-99 production facility, SS-347 coupons in uranyl sulfate solution at ~80° C were irradiated using Argonne’s Van de Graaff generator, which can generate high-radiation fields without fissioning of uranium or production of activation products.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Niowave Neutron Source Converter: Lead-Bismuth-Eutectic (LBE) Windowless Target Design and Evaluation

Los Alamos National Laboratory (LANL) is working with Niowave on the design and evaluation of their lead-bismuth-eutectic (LBE) windowless target (i.e., neutron source converter). Niowave plans to use 200 kW electron beam at 40 MeV beam energy to produce neutrons by photonuclear reaction with LBE. Then, the neutrons undergo fission at the surrounding uranium target assembly (UTA) to produce Molybdenum 99 (Mo-99) as a fission product, which eventually decays to Technetium-99 (Tc-99m). Tc-99m is one of important radioisotopes that is used for medical diagnostics. LANL conducted 3D multiphysics analysis for the Niowave neutron converter design and provided design assessment in thermohydraulic aspects. LANL conducted radiation transport calculations using Monte-Carlo N-Particle (MCNP) code with unstructured meshing scheme. The 3D volumetric heating profiles in the LBE and Stainless-Steel (SS) housing were imported into multiphase computational fluid dynamics (CFD) to obtain 3D temperature profiles of LBE and SS through conjugate heat transfer (CHT) analysis. The key findings are: LBE film thickness at the center of the beam is approximately 1.6 cm with a maximum LBE velocity of approximately 1.8 m/s, which is below a 2 m/s limit to avoid erosion issues on supporting structures; Heat deposition in the LBE peaks at ~1 cm depth from the LBE free-surface because of the forward interactions of electron, photon, and neutron with LBE; LBE maximum temperature is ~360 °C, which is below LBE evaporation initiative temperature, ~450 °C; LBE-SS interface temperature is ~350 °C, which is below the safety thermal limit to prevent severe corrosion on SS. The results indicate that Niowave’s neutron converter design satisfies both hydraulic and thermal criteria for safe operation. By virtue of such computational analysis, Niowave can move toward establishing an experimental setup to experimentally test their LBE neutron converter. The following sections describe the detailed work done by LANL.

43 PARTICLE ACCELERATORS↗
DOE OSTI2021

Fission Induced Radiolysis of Uranyl Sulfate Solutions

At Argonne National Laboratory, an experimental pilot facility is being tested for the production of the medical isotope molybdenum-99 ( 99 Mo). AMORE (Argonne Molybdenum Research Experiment) is an experimental pilot facility in which a 40 MeV electron beam from the LINAC strikes a depleted uranium (DU) target. The target generates high energy neutrons, which are thermalized in an 18-liter uranyl sulfate solution of low enriched uranium (LEU). Thermalized neutrons fission the uranium-235 ( 235 U) to produce 99 Mo as well as many other fission products. The 99 Mo is recovered from the uranium solution and purified. The goal of our work is to reduce the use of highly enriched uranium (HEU) for the production of this isotope.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2023

Mo-99 Concentration and processing by Solvent Extraction and Ion Exchange

Normally derived from its transient-equilibrium parent molybdenum-99 ( 99 Mo) (t½ = 66 h), the short-lived isotope technetium-99m ( 99m Tc) (t ½ = 6.0 h) is the most widely used medical isotope in diagnostic imaging today. The National Nuclear Security Administration’s (NNSA’s) Material Management and Minimization (M3) program—established under the auspices of the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not use highly enriched uranium (HEU). Superconducting electron linear accelerators that employ high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and uranium (U) fission. Argonne, in collaboration with industrial partners, has been developing the process chemistry for superconducting linear accelerator (LINAC)-irradiated triuranium octoxide (U 3 O 8 ) targets to produce 99 Mo. In short, the process involves dissolving the irradiated targets in nitric acid, which simultaneously releases valuable fission products (xenon and iodine). After adjusting the acid concentration, producers process the feed using tri-n-butyl phosphate (TBP), then mix the raffinate derived from this extraction cycle with a phosphinic acid such as Cyanex 272 or di(2-ethylhexyl) phosphoric acid (HDEHP), which is selective for Mo (molybdenum liquid liquid extraction [MoLLE]). In the next step, a concentration column comprising an anion exchange platform is used to decontaminate the remaining fission products, generating a pure 99 Mo stream in sodium hydroxide (NaOH)/sodium chloride (NaCl).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Evaluation of Niowave's Proposed Solvent Washing Approach

Niowave, Inc., is a domestic supplier of medical and industrial isotopes from uranium (U) and radium (Ra). The company has recently entered into a cooperative agreement with the U.S. Department of Energy’s National Nuclear Security Administration (NNSA) and plans to deploy a superconducting electron accelerator (LINAC) to fission U for molybdenum-99 ( 99 Mo) production without the need for a nuclear reactor or highly enriched uranium (HEU). NNSA provided funding to the Savannah River National Laboratory (SRNL) to support Niowave in this effort. SRNL evaluated the application of the solvent washing process for Niowave.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Potential of electrolytic processes for recovery of molybdenum from molten salts for 99 Mo production

Molten salt reactor (MSR) technologies are receiving significant interest for commercialization because of their potential safety features and efficient energy production. In addition to producing reliable and clean energy, high-value radioisotopes could be harvested from MSRs as an additional revenue stream. Although significant quantities of 99 Mo will be produced in MSRs through fission, how accessible 99 Mo will be for direct recovery from a fuel salt is unclear. Electrolytic processes in fused salts have been used extensively for processing molten salts and have the potential to enable the extraction of 99 Mo directly from the fuel salt. This paper reviews the behavior of 99 Mo in the molten salt reactor experiment and summarizes theoretical aspects of the electrolytic process in high-temperature fused salt and the published experimental results on Mo extraction through electrolytic processes. Here, the general principle of fused-salt electrolytic processes is described, by electrodeposition of Mo with the following aspects: electrowinning, electrorefining, electroplating, and electroextraction with several experimental considerations. Finally, important considerations for recovery of 99 Mo from MSR relevant molten salts via electrodeposition are discussed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Technology Maturation of Hot Isostatic Pressing for Nuclear Waste Treatment - 20259

ANSTO is constructing a nuclear waste treatment facility for the immobilization of intermediate-level liquid waste (ILLW) from its nuclear medicine production [1]. This facility shall deploy ANSTO Synroc process technology that converts a liquid waste stream into a durable solid wasteform. Hot Isostatic pressing (HIP) shall be employed to consolidate the dry powdered waste effectively reducing the waste volume and immobilizing the waste components. This paper presents the technology maturation for HIP technology for utilisation in a nuclear waste treatment facility. The HIP has been designed to operate in a remote shielded enclosure, fully integrated with the front-end powder production. Attention has been given in the HIP system design to ensure that it can operate and be maintained to achieve the required production output of the facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Current Status of the Synroc Waste Treatment Facility - 20263

Construction of the Synroc Waste Treatment Facility is well underway at ANSTO. This plant has been designed to treat intermediate level liquid waste arising from the production of Mo-99 within the newly constructed ANSTO Nuclear Medicine (ANM) precinct [1]. The facility utilises ANSTO Synroc technology that has been tailored to the chemical, physical, and radiological properties of the waste. The result is a highly durable wasteform with a significant reduction in the final volume of the treated waste. This paper presents the construction status of the Synroc Waste Treatment Facility and the status of the associated technology maturation plan. Construction of the building is scheduled for completion in late 2020 followed by process installation and commissioning. A key component of the technology maturation plan has been the construction of an Inactive Engineering Demonstrator [2]. This has aided in the mitigation of risks with respect to technology selection, process integration, and process nuclearisation. Furthermore, the demonstration facility has provided an environment for the design and development of the instrumentation and control philosophy resulting in a seamlessly integrated plant control system. Successful demonstration of the technology by the inactive engineering demonstrator has significantly reduced the associated risks. Details of the Synroc Waste Treatment Facility project progress will also be presented. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Thermal Process Technology for Nuclear Applications - 20379

ANSTO's Synroc technology has been developed to provide a safe, secure matrix for the immobilization and final disposal of radioactive waste. Synroc technology will be used to manage radioactive wastes from the production of the radioisotope Molybdenum-99 (Mo-99). This paper shall outline various stages of the process development with specific reference to the thermal treatment technology of calcination. Calcination is a key step in the Synroc process [1-2].The rotary thermal processing system includes: an advanced heating element design for increased robustness and ease of remote operation and maintenance, an enhanced modular design of components for ease of remote maintenance in a hot cell and in compliance with hot cell radioactive environment requirements for safety, reliability and maintainability. In addition to thermal treatment of waste from nuclear medicine production, this technology provides solutions for a variety of nuclear materials processing applications including sintering UO{sub 2} pellets for reactor fuel rods, oxidation of UO{sub 2} pellets, swarf, and powder to U{sub 3}O{sub 8}, de-nitration of Uranyl nitrate and hydrofluorination of UO{sub 2} pellets. The paper will also discuss thermal processing solutions for a range of nuclear applications. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Crystallization of Ammonium Heptamolybdate for Reduction to Mo Metal

The 100 Mo/ 98 Mo recycle process developed by Argonne National Laboratory in collaboration with Oak Ridge National Laboratory (ORNL) and supported by the U.S. Department of Energy (DOE) National Nuclear Security Agency’s (NNSA’s) Office of Material Management & Minimization (M3) is vital to sustaining the economic production of 99 Mo. The high-yield molybdenum solution extraction (MOEX) process recovers enriched Mo by acidifying spent generator solutions, extracting Mo using tri-n-butyl phosphate (TBP), and stripping Mo into ammonium hydroxide, where it is later converted to solid (NH 4 ) 6 M 07 O 24 (ammonium heptamolybdate or AHM) by crystallization. It is important to produce the AHM product with consistent particle size and morphology before its thermal treatment and reduction to Mo metal in a furnace. AHM particles that are too small (< 50 µm) result in Mo metal powder that is much too fine to properly fabricate into targets with optimum properties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

An engineering perspective on evaluating mechanisms governing ductility in pure molybdenum

The United States lacks a stable domestic supply of 99 mTc, a critical medical imaging isotope generated from 99 Mo. Accelerator-based production using 100 Mo targets (aMo) introduces mechanical concerns due to the ductile-to-brittle transition temperature inherent to refractory metals like Mo. This study evaluates the tensile behavior of aMo targets from 25 to 1000 °C, compared to powder-metallurgy-processed natural Mo (PMo) and cast-and-rolled Mo (RMo), both as-received and after 5 ppm O 2 exposure in flowing He. RMo showed superior ductility (7.3% at 25 °C, 48% at 1000 °C) and strength, attributed to its fine, elongated grains and high geometrically necessary dislocation density. PMo exhibited variable ductility (up to 33%), while aMo remained brittle, with a maximum elongation of 10.4% at 600 °C. EBSD analysis revealed weak texture in PMo and aMo, but high defect density in aMo limited dislocation mobility. This work links processing, microstructure, and deformation mechanisms to guide fabrication of ductile refractory targets.

Hyer, Holden C. [Oak Ridge National Laboratory (OR↗
DOE OSTI2023

Molybdenum-99 from Molten Salt Reactor as a Source of Technetium-99m for Nuclear Medicine: Past, Current, and Future of Molybdenum-99

Technitium-99m ( 99m Tc), a widely used radioisotope, is used in tens of millions of medical diagnostic procedures annually. However, it is hard to store and must be immediately used upon production due to its short half-life (i.e., 6 h); thus, it is currently produced from 99 Mo, which itself is a result of 235 U fission. The majority of 99 Mo supplies to U.S. patients are currently provided by foreign producers and produced using highly enriched uranium (HEU). In order to minimize the proliferation risks of HEU-based medical isotope production, the U.S. Department of Energy’s National Nuclear Security Administration has funded a program to accelerate the development of technologies to produce 99 Mo without the use of HEU. Today, the global supply of 99 Mo depends on a limited number of nuclear reactors, and production has been interrupted unexpectedly since 2009 due to the fleet’s advanced age. Herein, alternative options for 99 Mo production are discussed, and one potential option is to obtain 99m Tc from molten salt reactors (MSRs). A MSR is a nuclear fission reactor that can operate at or close to atmospheric pressure with liquid fuel, which allows for producing isotopes in a timely manner. In this paper, the past and current production of 99 Mo via nuclear reactors is described, and the future of 99 Mo production by MSRs is discussed. The behavior and chemical properties of molybdenum in fluoride salts in MSRs and the possible extraction methods are also examined in addition to the limitation of current studies.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Molybdenum Recovery from Filters Used in Large-scale Dissolution of Sintered Mo-disks

99 Mo production with linear accelerators can be achieved via the bremsstrahlung photonuclear reaction 100 Mo(γ,n) 99 Mo or the neutron capture reaction 98 Mo(n,γ) 99 Mo. For commercial producers, maximum recovery of enriched 98 Mo and 100 Mo target material is critical for sustaining an economic production cycle. During the peroxide dissolution of Mo metal disks and subsequent conversion to K 2 MoO 4 , several grams of Mo can be lost during filtration from solution when several hundred grams of sintered Mo disks are processed. This investigation shows that 5–8 g Mo is routinely retained on the filter units, but it can be almost fully recovered using aqueous washes. Washing can be done immediately and incorporated into the dissolution procedure, or it can occur several months after the initial filtration process to decrease processing time.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

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↗
DOE OSTI2022

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↗
DOE OSTI2026

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↗