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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Materials Data on U3O8 by Materials Project

U3O8 crystallizes in the trigonal P-31m space group. The structure is three-dimensional. there are two inequivalent U+5.33+ sites. In the first U+5.33+ site, U+5.33+ is bonded in a distorted hexagonal planar geometry to six equivalent O2- atoms. All U–O bond lengths are 2.17 Å. In the second U+5.33+ site, U+5.33+ is bonded in a body-centered cubic geometry to eight O2- atoms. There are two shorter (2.09 Å) and six longer (2.30 Å) U–O bond lengths. There are two inequivalent O2- sites. In the first O2- site, O2- is bonded in a trigonal non-coplanar geometry to three U+5.33+ atoms. In the second O2- site, O2- is bonded in a linear geometry to two equivalent U+5.33+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on U3O8 by Materials Project

U3O8 crystallizes in the monoclinic P2_1/m space group. The structure is three-dimensional. there are three inequivalent U+5.33+ sites. In the first U+5.33+ site, U+5.33+ is bonded to seven O2- atoms to form UO7 pentagonal bipyramids that share corners with two equivalent UO6 octahedra, corners with three equivalent UO7 pentagonal bipyramids, an edgeedge with one UO6 octahedra, and edges with two equivalent UO7 pentagonal bipyramids. The corner-sharing octahedra tilt angles range from 37–49°. There are a spread of U–O bond distances ranging from 2.08–2.49 Å. In the second U+5.33+ site, U+5.33+ is bonded to six O2- atoms to form UO6 octahedra that share corners with two equivalent UO6 octahedra, corners with four UO7 pentagonal bipyramids, and edges with two UO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 0°. There are a spread of U–O bond distances ranging from 2.08–2.17 Å. In the third U+5.33+ site, U+5.33+ is bonded to seven O2- atoms to form UO7 pentagonal bipyramids that share corners with two equivalent UO6 octahedra, corners with three equivalent UO7 pentagonal bipyramids, an edgeedge with one UO6 octahedra, and edges with two equivalent UO7 pentagonal bipyramids. The corner-sharing octahedra tilt angles range from 37–49°. There are a spread of U–O bond distances ranging from 2.08–2.48 Å. There are seven inequivalent O2- sites. In the first O2- site, O2- is bonded in a linear geometry to two U+5.33+ atoms. In the second O2- site, O2- is bonded in a linear geometry to two U+5.33+ atoms. In the third O2- site, O2- is bonded in a linear geometry to two equivalent U+5.33+ atoms. In the fourth O2- site, O2- is bonded in a trigonal planar geometry to three U+5.33+ atoms. In the fifth O2- site, O2- is bonded in a trigonal planar geometry to three U+5.33+ atoms. In the sixth O2- site, O2- is bonded in a 3-coordinate geometry to three U+5.33+ atoms. In the seventh O2- site, O2- is bonded in a 3-coordinate geometry to three U+5.33+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on U3O8 by Materials Project

U3O8 crystallizes in the orthorhombic Cmmm space group. The structure is three-dimensional. there are two inequivalent U+5.33+ sites. In the first U+5.33+ site, U+5.33+ is bonded to seven O2- atoms to form UO7 pentagonal bipyramids that share corners with two equivalent UO6 octahedra, corners with three equivalent UO7 pentagonal bipyramids, an edgeedge with one UO6 octahedra, and edges with two equivalent UO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 43°. There are a spread of U–O bond distances ranging from 2.08–2.38 Å. In the second U+5.33+ site, U+5.33+ is bonded to six O2- atoms to form UO6 octahedra that share corners with two equivalent UO6 octahedra, corners with four equivalent UO7 pentagonal bipyramids, and edges with two equivalent UO7 pentagonal bipyramids. The corner-sharing octahedral tilt angles are 0°. There are two shorter (2.08 Å) and four longer (2.14 Å) U–O bond lengths. There are four inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted trigonal planar geometry to three U+5.33+ atoms. In the second O2- site, O2- is bonded in a linear geometry to two equivalent U+5.33+ atoms. In the third O2- site, O2- is bonded in a linear geometry to two equivalent U+5.33+ atoms. In the fourth O2- site, O2- is bonded in a linear geometry to two equivalent U+5.33+ atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Shinkolobweite, from the Shinkolobwe Mine, Democratic Republic of Congo: A New Mineral Containing Uranium in the Rare Pentavalent Oxidation State

ABSTRACT Shinkolobweite, Pb1.333[U5+O(OH)(UO2)5O4.67(OH)5.33](H2O)5, is a new lead uranyl oxide-hydroxide hydrate mineral containing hexavalent and pentavalent uranium from the Shinkolobwe mine, Democratic Republic of Congo. Crystals of shinkolobweite are dark reddish-brown prisms up to 0.5 mm in length, occurring on a matrix of massive uraninite associated with fourmarierite, rutherfordine, soddyite, and sklodowskite. Crystals are translucent with subadamantine luster and light bronze-yellow streak, are flattened on {010}, are elongated on [001], and exhibit the forms {100}, {010}, {101}, and . The mineral is non-fluorescent under both longwave and shortwave ultraviolet illumination. It has a Mohs hardness of ∼2 and exhibits brittle tenacity with perfect cleavage on {010}, imperfect cleavage on {100}, and even fracture. The calculated density is 5.853 g/cm3 based on the empirical formula. Electron probe microanalysis provided the empirical formula Pb1.290U6+4.876U5+1.166O27H16.633 based on 27 O apfu and U5+:U6+ determined by X-ray photoelectron spectroscopy. Shinkolobweite is orthorhombic, superspace group Pnnm(0b0)000, a = 14.4808(4), b = 7.0681(8), c = 11.9423(3) Å, V = 1222.32(15) Å3, modulation wave vector [0 1/3 0], and Z = 2. The structure was refined from 8959 reflections to a final R1 = 0.0736 for all reflections. Uranyl oxide-hydroxide sheets in shinkolobweite adopt the β-U3O8 topology and possess (3 + 1) commensurately modulated ordering that results from the long-range ordering of U5+ and U6+ in the sheet, as well as the position and occupancy of interlayer Pb2+ cations. Observations of a topological transition between α-U3O8 and β-U3O8 type sheets in shinkolobweite supplements our understanding of U5+ mineral oxidation and stability.

Mineralogy↗
DOE OSTI2021

Effect of Added Gadolinium Oxide on the Thermal Air Oxidation of Uranium Dioxide

To develop a more reliable and stable UO2-based nuclear fuel, the Pacific Northwest National Laboratory (PNNL) investigated modifying fuel with several soluble lanthanides and zirconium. This article provides the results of these studies investigating gadolinium doping at levels up to 10 mass%. The authors characterized and compared commercially- and PNNL-prepared gadolinium-doped UO2 to determine the oxygen-to-metal ratio, elemental distribution, chemical composition, physical appearance, lattice parameters, and grain structure using atomic force microscopy, scanning electron microscopy coupled with energy dispersive spectroscopy, and X-ray diffractometry. After confirming PNNL-prepared UO2 and (Gd,U)O2 were similar to commercially prepared UO2 and (Gd,U)O2, we measured the thermal behavior of these gadolinium-doped UO2 materials to air oxidation using differential scanning calorimetry and thermogravimetric analysis. Addition of gadolinium stabilized the first oxidation product U4¬O9/U3O7 and slowed the subsequent oxidation to U3O8. Comparison of our measured two-step oxidation of UO2 to U3O8 at 325°C to common gas/solid kinetic reaction models found that each oxidation step is best described as a convolution of kinetic behaviors; the gadolinium insertion into the UO2 lattice enforces a significant alteration in mechanism and oxidation rate. Noticeable changes in mechanism become apparent between 1% and 3% gadolinium content.

Uranium Dioxide, Gadolinium Oxide, Thermal Air Oxi↗
DOE OSTI2026

Single particle – MC-ICP-MS for isotopic analysis of uranium particles

Single particle – multi-collector – inductively coupled plasma – mass spectrometry (SP-MC-ICP-MS) was employed to measure a suspension of 1 µm U3O8 particles (∼1.3 pg total U/particle) to determine their individual isotopic compositions of 234U/238U, 235U/238U, and 236U/238U. The effects of different detector combinations for 235U and 238U, including secondary electron multipliers (SEM) and Faraday detectors (1011 and 1013 Ω amplifiers), were explored for accuracy and precision optimization on the observed 235U/238U. The minor isotopic ratios (i.e., 234U/238U and 236U/238U) were analyzed such that the 234U and 236U were monitored on SEM detectors and the 238U was monitored on a Faraday (1011 Ω) detector. Various integration times (5, 10, 25, and 50 ms) were investigated in all detector configurations to gain a better understanding of their impact on sensitivity, accuracy, and precision. For 235U/238U ratios of 1 µm U3O8 particles, a dual Faraday detector measurement with 1011 Ω was the optimal choice; measurement of 1021 particles yielded an average 235U/238U ratio of 0.00170 (14), a −1.8% relative difference (% RD) from the reference value. The minor isotopic compositions were determined to be 0.0000070 (14) and 0.0000758 (48) for the 234U/238U and 236U/238U, respectively. These measurements correspond to <8% and <1% RD from their reference value for the 234U/238U and 236U/238U, respectively. SP-MC-ICP-MS was also able to provide insight into measurement sensitivity. In these individual particles, merely 15 and 165 atto-grams (ag) of 234U and 236U were present (calculated). Initial limits of detection for SP-MC-ICP-MS were determined to be ∼1.0 ag (when measured via SEM detectors). This valuable approach is applicable to areas including nuclear forensics, nuclear safeguards, and geochemical analysis, which require high-precision measurements of uranium within micron-sized particles.

Manard, Benjamin [ORNL] (ORCID:0000000207400627)↗
DOE OSTI2026

Analytical workflow dependence of experimental observables for uranium chemistries

In this work, we investigated how the sequencing of laboratory analytical methods used for chemical and morphological characterization influences analytical findings for particulate materials relevant to the nuclear fuel cycle, including UO2, U3O8, studtite (UO2O2·4H2O), and β-UO3, in the context of nuclear forensic analysis. Particles of each chemistry obtained from consistent production batches were exposed to Raman spectroscopy and scanning electron microscopy in varying orders to elucidate how the order in which the techniques are applied influences morphological and chemical observations as a function of particle size. The results indicate that particles from all four chemistries exposed to high-resolution electron imaging before Raman spectral analysis demonstrate optical vibrational spectral changes that reduce accurate interpretation of the underlying chemistry via Raman spectral analysis. We hypothesize that these changes are due to the thermal load of the electron beam imparted to the sample being unable to be dissipated by materials with poor thermal conduction properties. Results from this study will aid in determining best practices for forensic analysis procedures to reduce uncertainty in chemical determination of unknown particulate samples.

Manns, Rebecca [University of Nevada, Las Vegas]↗
DOE OSTI2023

Advanced Fuel Cycle Cost Basis Report: Supporting Document 7 Presentation: Du and RU Disposal Costs [Slides]

Worldwide DU and RU in various chemical forms are some of the largest legacy products of the nuclear industry (defense & power) in both mass and volume. Chemical forms include U metal or alloy, UF6, UO2, UO3, U3O8, and UF4. Most of this material is now in above ground storage. Due to chemical stability and low water solubility oxides are the preferred form for safe storage and ultimate disposal. Long term U disposal in large amounts presents more of a potential radioactivity problem than its conversion and temporary storage. Now a near term problem with freshly-mined uranium ore, radon emanation will eventually be a long-term problem for both DU and RU dispositioned in large quantities at a specific location.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Conceptual Design of a Scaled Voloxidation Apparatus for Testing with Used Nuclear Fuel

This report details the conceptual design overview for Department of Energy (DOE) Level 2 milestone M2FT-25IN030101021 titled “Complete conceptual design of a scaled voloxidation apparatus for testing with used nuclear fuel.” Advanced voloxidation uses high temperature nitrogen dioxide (NO2) gas to oxidize uranium dioxide (UO2) material to triuanium octoxide (U3O8) and ultimately to uranium trioxide (UO3). After consideration of multiple available options and inputs, a rotating advanced voloxidation design was determined to be the best fit for the needs of a full-scale apparatus. The system, which will be deployed in hot cell 4 of the Analytical Research Laboratory (ARL), has been designed to meet the space constraints of the facility while still accommodating at least 100g of used nuclear fuel (UNF). The apparatus consists of three main unit operations, including the gas panel, the custom furnace with reaction vessel, and the liquid scrubber.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Overview of Algorithms for Using Particle Morphology in Pre-Detonation Nuclear Forensics

A major goal in pre-detonation nuclear forensics is to infer the processing conditions and/or facility type that produced radiological material. This review paper focuses on analyses of particle size, shape, texture (“morphology”) signatures that could provide information on the provenance of interdicted materials. For example, uranium ore concentrates (UOC or yellowcake) include ammonium diuranate (ADU), ammonium uranyl carbonate (AUC), sodium diuranate (SDU), magnesium diuranate (MDU), and others, each prepared using different salts to precipitate U from solution. Once precipitated, UOCs are often dried and calcined to remove adsorbed water. The products can be allowed to react further, forming uranium oxides UO3, U3O8, or UO2 powders, whose surface morphology can be indicative of precipitation and/or calcination conditions used in their production. This review paper describes statistical issues and approaches in using quantitative analyses of measurements such as particle size and shape to infer production conditions. Statistical topics include multivariate t tests (Hotelling’s T 2 ), design of experiments, and several machine learning (ML) options including decision trees, learning vector quantization neural networks, mixture discriminant analysis, and approximate Bayesian computation (ABC). ABC is emphasized as an attractive option to include the effects of model uncertainty in the selected and fitted forward model used for inferring processing conditions.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2021

URANIUM HOLDUP MONITORING WITH COMPTON IMAGING AS FUNCTION OF DEPTH AND MASS

Mass quantification using standoff gamma detection is a logical progression for MC&A applications in uranium processing facilities. While obtaining a ground truth from physical collection and weighing of masses is ideal and routine, it poses a logistical problem from both a standpoint of ALARA programs as well as interruptions to operations. The standard approach is the Generalized Geometry in Holdup (GGH) method that can apply for many cases where geometries of the holdup match simulated models; however, there are many cases where a model may not exist or a feature of the holdup geometry is such that it does not match a model particularly well.1 Gamma imaging at standoff distances and outside process equipment has potential to address these shortcomings and improve mass estimations while reducing personnel exposure; however, algorithms still require a significant degree of benchmarking to build sufficient confidence in algorithms before displacing GGH in some procedures. This paper presents the creation of a set of laboratory standards of depleted U3O8 in fixed size containers covering a range of masses and effective thicknesses up to the infinite thickness for the 185-keV line. These blocks are identically sized aluminum boxes loaded such that a thin, flat face is placed towards the imager, while the back side contains closed cell foam to apply pressure to keep the powder in a fixed position at the front face as the sources are manipulated for measurements. Included is an analysis of the mass estimation using a commercially available Compton imaging instrument where multiple geometries are created showing limitations in current methodologies.

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

VERIFICATION MEASUREMENTS OF THE MASS AND ENRICHMENT OF URANIUM OXIDE CARD SOURCES

Measurements were performed to determine the mass and enrichment of 10 uranium oxide (U3O8) card sources. The measurements and analysis were completed as a verification of the card sources in support of Oak Ridge National Laboratory’s nuclear material control and accountability program. Although these cards are not nationally accredited as a nuclear standard, they are used as a working reference for measurements, such as holdup. The uranium card source measurements were taken with a broad energy germanium detector and the Genie 2000 Gamma Acquisition and Analysis software. A complete characterization of each of the 10 uranium card sources was performed using the 4 characteristic full energy peaks of 235U. Using the In Situ Object Counting System software to determine the mathematical efficiency of the measurement, the mass of 235U in each card was determined. The 235U mass in each card ranged from 10.42 to 12.63 g with a systematic error between 0.76 and 0.95 g and a random error of 0.01 g for each card source. The Multi-Group Analysis for Uranium (MGAU) software and the Fixed-Energy, Response Function Analysis with Multiple Efficiency (FRAM) isotopic analysis software were used to determine the isotopic composition of the uranium cards. The measured enrichment was compared to the declared enrichment for each card, with uncertainties ranging from 2.7% to 4.3% for the MGAU analysis and 2.3% to 4.1% for the FRAM analysis. This is a good example of how a well-benchmarked mathematical calibration method can be useful in characterizing uranium sources.

Hunneke, Rachel↗
DOE OSTI2022

A New Process for Small-Batch Purification of the Medical Isotope Molybdenum-99: Non-Technical Overview

The U.S. medical community depends on a reliable supply of the radioisotope molybdenum-99 (Mo-99) for nuclear medical diagnostic procedures. Mo-99's decay product, technetium-99m (Tc-99m), is used in over 40,000 medical procedures in the United States each day to diagnose heart disease and cancer, to study organ structure and function, and to perform other important medical applications. For example, patients undergoing a common procedure—the cardiac “stress test”—likely have benefited from Tc-99m. Historically, Mo-99 was primarily produced through the fission of uranium-235, in the form of highly enriched uranium (HEU) targets irradiated in research and test reactors. HEU is a proliferation-sensitive material that, if diverted or stolen, could be used as a component of a nuclear weapon. NNSA’s Office of Material Management and Minimization (M3) manages the Molybdenum-99 (Mo-99) Program as part of its mission to minimize the use of HEU in civilian applications. The Mo-99 Program assists global Mo-99 production facilities in converting to non-HEU processes and supports the establishment of domestic supplies of Mo-99 without the use of proliferation-sensitive HEU. As part of this program, M3 funds U.S. national laboratories to provide non-proprietary technical support to U.S. companies working to establish non-HEU-based Mo-99 production capabilities. The results of this research are published on OSTI.gov for the benefit of the Mo-99 community and the public. However, it can be difficult for readers without a scientific background to understand and interpret these publications. In order to increase public understanding of the work being done in M3’s Mo-99 Program, this paper aims to provide an overview of a key, recent national laboratory technical publication in terms that can be understood by readers without a technical background. To accomplish this, the paper first explains key scientific concepts—primarily related to chemistry—that provide a foundation for understanding research in this area. This includes chromatography, absorption vs. adsorption, dissolution and precipitation, and liquid-liquid extraction. Drawing on these concepts, the paper then provides an explanation for non-technical audiences of the Argonne National Laboratory publication entitled Recovery of High Specific Activity Molybdenum-99 from Accelerator-Induced Fission on Low-Enriched Uranium for Technetium-99m Generators (Brown, M.A. et al., 2021) and related article Separation and Purification of Mo-99 Produced from Natural U3O8 Targets via Photo-Fission (Brown, M.A. et al., 2021).

Stamler, Bradley↗
DOE OSTI2022

Benefits of using multiple Raman laser wavelengths for characterizing defects in a UO 2 matrix

Abstract Raman spectroscopy is one of the most useful techniques for studying the structure of UO 2 and changes due to specific defects caused by doping, changes in stoichiometry, irradiation, or heating under oxidizing conditions. In this paper, we illustrate several aspects of the application of Raman techniques to the study of UO 2 , including the use of wavelength‐dependent excitation (455, 532, and 785 nm) to assess the effects of doping (Nd, Th, and Zr), ion irradiation, and in situ heating and oxidation (UO 2 to U 3 O 8 ). Additionally, we show examples of how correlative microscopy is possible using electron backscatter diffraction combined with Raman maps of specific vibration bands or of laser‐induced luminescence generated by rare‐earth dopants in the matrix. For each of these applications, we suggest optimal excitation wavelengths that vary depending on the desired data. Blue (455 nm) excitation tends to promote oxidation even at low powers, but because Raman spectra change little with doping, irradiation‐induced changes are easier to observe. Green (532 nm) excitation is optimal for observing electron–phonon resonance effects in UO 2 and offers a good compromise for high‐temperature oxidation experiments, delivering high‐quality spectra for both UO 2 and U 3 O 8 . Infrared (785 nm) excitation is best for observing “defect” bands associated with doping in UO 2 , as changes with irradiation are small. Raman spectroscopy is particularly suited for studying the stability of UO 2 towards oxidation in the presence of dopants simulating fission products, where electron–phonon resonant effects, dopant ion luminescence, and mapping can be used together to investigate structural rearrangement as a function of temperature. These techniques can offer insight into microstructural changes in UO 2 fuels at higher burnups envisioned in future reactors.

42 ENGINEERING↗
DOE OSTI2022

Qualitative assessment of uranium ore concentrates and related materials using scanning electron microscopy

Several studies have evaluated the morphology of uranium compounds produced under controlled conditions at the laboratory scale, but it is unclear whether the morphological characteristics of these materials persist in commercially produced uranium ore concentrates (UOCs). To assess the morphology in “real-world” UOCs, we qualitatively evaluated the morphological profile of secondary electron images from over 100 commercial UOCs using a previously published lexicon. We observe differences between samples with differing chemical composition and samples with similar chemical composition and differing provenance. Further, this work contextualizes morphology for commercially produced UOCs and will provide a basis for future machine learning efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Rehydration of metastudtite in the alteration kinetics of α– and β–U 3 O 8 in dilute aqueous solutions of hydrogen peroxide

The formation of alteration phases on uranium ore concentrates and used nuclear fuels under oxidizing conditions is key to understanding the potential mobility of radionuclides in the environment and designing optimal storage conditions of materials. However, the time-dependent distribution of alteration phases on α– and β–U 3 O 8 under oxidizing conditions has yet to be explored. Here, in this study, crystalline powders of α– and β–U 3 O 8 were submerged in aqueous solutions of hydrogen peroxide (1.6 × 10 −1 to 5.4 × 10 −5 M) with aliquots of solution and solid removed for analysis at 1, 8, 15, 22, 29, 36, 46, 58, 71, and 83 days. Within one day there is significant alteration of U 3 O 8 to the uranyl peroxide metastudtite, [(UO 2 )(O 2 )(H 2 O) 2 ], that is replaced by studtite, [(UO 2 )(O 2 )(H 2 O) 2 ]·2H 2 O, within a week regardless of the polymorph of U 3 O 8 or the initial concentration of H 2 O 2 in solution, as determined by partial least squares regression (PLSR) of Raman spectra collected from the solids. A dissolution/reprecipitation mechanism is proposed for both the alteration of U 3 O 8 to metastudtite and the subsequent alteration of both U 3 O 8 and metastudtite to studtite. The two polymorphs of U 3 O 8 exhibit similar rates and extents of alteration over time. The rehydration of metastudtite to studtite has not been previously reported and highlights the need for future work to determine the mechanism by which metastudtite is converted to studtite and what other conditions facilitate this rehydration.

Alteration of U3O8↗
DOE OSTI2024

Spark plasma sintering of fuel meats for U 3 O 8 based dispersion fuels

Research and test reactors often use dispersion-type fuel due to its increased thermal conductivity and burn-up capabilities compared to conventional fuel. Al-U 3 O 8 (aluminumtriuranium octaoxide) dispersion fuels have several advantages over their competitors, such as higher service temperature and better stability of oxygen stoichiometry. However, the two-step fabrication of dispersion fuel causes undesirable porosity in cold-pressed fuel meats that is preserved in co-extruded fuel plates. To combat this, spark plasma sintering (SPS) was used for the fabrication of Al-15, 20, and 30 vol% U 3 O 8 and 8 and 12 vol% Mo-U 3 O 8 fuel meats for the Al-U 3 O 8 time. The in situ SPS data was used to construct and validate Master Sintering Curves (MSCs) with accuracies in Al fuels at 0.02 g/cm 3 , and Mo fuels at 0.07 and 0.17 g/cm 3 . The as-sintered fuel meats were characterised using x-ray diffraction (XRD) and scanning electron microscopy (SEM) to understand chemical and physical changes following the SPS process. The pellets exhibited very high relative densities, the U 3 O 8 was observed to undergo reduction to UO 2 .

Aluminium↗
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↗