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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Development of spectroscopic tools for nuclear forensics signatures development

The Savannah River National Laboratory (SRNL) has been developing advanced spectroscopic tools for the characterization of plutonium-bearing compounds with the intent to develop material and process signatures for nuclear forensics. Plutonium in a production, refining, or finishing facility will exist in many forms including oxide precursors (PuF4, PuF3, Pu oxalate, etc.), oxide, and metal. The ability to identify plutonium in each of these chemical forms and determine their processing history is crucial for the development of spectroscopic signatures. This presentation will focus on our work to develop and apply spectroscopic tools at SRNL using doubled-walled cells (DWC) to characterize the thermal decomposition of oxalates, calcination chemistry, alpha-decay-induced chemistry, age dating since last calcination, and other signatures related to plutonium processing.

Villa-Aleman, Eliel↗
DOE OSTI2023

Development of Spectroscopic Tools for Nuclear Forensics Signatures Development

Diversion of nuclear materialsfrom legitimate commercial processes for use by nefarious actors is a frightening and surreptitious problem. Smuggled nuclear material can be used to manufacture a simple radiological dispersion device (RDD) and potentially incorporated into the development of a nuclear weapon. With some frequency, smuggled nuclear material is interdicted in foreign countries as actors cross border checkpoints. Nuclear material interdicted in this way can be analyzed using a variety of tools. Plutonium-239 is commonly used as the fissile material in nuclear weapon production as an alternative to uranium-235. Plutonium in a production, refining, or finishing facility will exist in many forms including oxide precursors (PuF4, PuF3, Pu oxalate, etc.), oxide, and metal. The ability to identify plutonium in each of these forms and determine properties, characteristics, or history of these materials is crucial for determining if the processes occurring in a nuclear facility are legitimate or not. This presentation will focus on the development and application ofspectroscopic tools developed at SRNL using doubled-walled cells (DWC) to characterize Pu-bearing compounds. Advances in the use of vibrational spectroscopy techniques (Raman and IR), UV-VIS-SWIR diffuse reflectance spectroscopy, gamma emission spectroscopy, luminescence spectroscopy, laser-induced breakdown spectroscopy in conjunction with XRD and SEM in the characterization of Pu-bearing compounds will be presented with emphasis in material processing history.

Villa-Aleman, Eliel↗
DOE OSTI2025

Direct Extraction of Uranium-Lanthanide Oxides in Tributyl Phosphate

Abstract Direct extraction of used nuclear fuel (UNF) in an organic solution could be more efficient than the previous practice with aqueous solutions. However, the UNF would need to be treated via voloxidiation before being processed using solvent extraction. The voloxidation process can form oxide and/or nitrate compounds. This work investigated the dissolution of uranium/lanthanide oxides in 30 vol % TBP diluted in dodecane (pre-equilibrated with 4 M nitric acid) in a glass reactor with air sparging to ascertain the uranium/lanthanide oxide dissolution behavior prior to scaling the process. The uranium/lanthanide oxides were prepared by co-precipitating uranium/lanthanide nitrates with hydroxide and then calcined to form a mixed oxide. While the relative concentrations of the lanthanides are not representative of used nuclear fuel, the neodymium and erbium allowed ease of tracking dissolution with visible spectroscopy. Cerium was used as a surrogate for plutonium. The dissolution rate of the oxides was similar but incomplete. A miniscule amount of cerium, as cerium oxide, took several months to slowly dissolve; however, when co-precipitated with uranium and other lanthanides, a significant amount of cerium dissolved readily.

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

Effect of ionizing radiation on PuO2 and surrogate oxide materials under storage conditions

The stockpile of plutonium separated from used nuclear fuel amounts to many 100s of tonnes. The majority of this plutonium is stored as PuO2 powder packaged in multi-can containers. The conditions during the packaging of reprocessed PuO2 for long term storage are closely controlled to limit water uptake as radiolytic decomposition of any adsorbed water will lead to the formation of both a potentially flammable atmosphere containing molecular hydrogen as well as reactive oxygen species which may be incorporated into the oxide phase. Consequently, the safety case for the long-term storage of PuO2 requires a complete understanding of the fundamental physical, chemical and materials degradation processes occurring at the water-oxide interface inside the storage canisters. This study investigates the effect of ionizing radiation on PuO2 and various surrogate oxides under atmospheres potentially encountered in storage canisters.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Technical Basis for Extending Storage Life of DOE-STD-3013 Containers

The U.S. Department of Energy (DOE) Technical Standard for Stabilization, Packaging, and Storage of Plutonium-Bearing Materials, DOE-STD-3013 [USDOE 2018], provides requirements for packaging stabilized metals, alloys, and oxides containing at least 30 weight percent (wt%) plutonium (Pu) plus uranium for storage for up to 50 years. This document provides the technical basis for extending that storage duration to up to 100 years for metals and alloys, and oxides that do not present a corrosion risk to the integrity of the container.

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

Recommended Representative Isotopic Compositions for Potential VTR Pu Supplies

This document recommends isotopic compositions to represent selected Pu supplies that might be made available for use in VTR driver fuel. These compositions are intended to be conservative representations suitable for determination of environmental impacts associated with use of the selected Pu supplies, but not upper bounds of all possible compositions. The selected Pu supplies under consideration are themselves representations of a more complicated set of possibilities, defined in a manner to allow efficient but conservative analysis of impacts. The source definitions to be considered are as follows: 1) Foreign sources of reactor-grade plutonium. This material would be provided as polished material in an oxide form and is assumed to be transported from Europe (France or England); 2) Domestic, non-weapon sources of plutonium such as unused FFTF fuel, Hanford metal (stored at SRS), and various oxides; 3) Domestic sources of weapons-grade plutonium, including metal from dismantled pits and otherwise excess to stockpile needs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The TRANSCEND University Consortium: Theme 4: Nuclear Materials - 20431

The safe and secure management of Pu is a matter of international concern, with ∼250 t of separated Pu currently stockpiled worldwide. The UK's civil inventory of nuclear materials contains significant stocks of separated Pu from the reprocessing of Magnox and AGR spent fuels. The preferred option for the 138.5 tonnes of Pu is re-use as mixed oxide (MOx) fuel, although 5% is not suitable for re-use and is recommended for direct disposal. However, it will take more than 15 years to implement re-use, requiring that the Pu be kept in interim storage in its current state for that period, i.e. as PuO{sub 2} powder within inert steel storage cans at Sellafield. The focus of the work presented here is thus plutonium storage and the direct disposal of plutonium. The Research and Development needs of both are now pressing: in the case of storage due to it being the current default; in the case of immobilization and disposal because of a comparative lack of Research and Development on Pu conditioning and packaging due to policy uncertainty as to whether it would be disposed of in a Geologic Disposal Facility (GDF). Addressing these needs is complicated by Pu's high radioactivity, decay heat and radiotoxicity, criticality, nuclear safeguard requirements and, for some UK Pu contaminated materials targeted for disposal, poor inventory. Thus, there is also a critical requirement for underpinning research on Pu bearing materials in these two contexts. In response to these needs, the TRANSCEND Consortium (Transformative Science and Engineering for Nuclear Decommissioning, a multi-disciplinary collaboration of 11 universities and 8 key industry partners from across the UK's civil nuclear sector) is seeking to provide technical underpinning to ongoing option development for the UK's civil Pu stockpile. Whilst understanding the behaviour of plutonium during its re-use as MOx is beyond the scope of the TRANSCEND Consortium work plan, the main objectives of the work are: (1) For interim storage: to understand how the surface structure and properties of pristine and radiation damaged PuO{sub 2} change with time in the absence and presence of water; and (2) For immobilization and disposal: to understand the mechanisms of incorporation of Pu into ceramic and glass-ceramic waste-forms, as well as the effect on these of self-induced radiation damage. Each objective is being addressed through separate work packages, the details of which are discussed in this paper. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impurity Levels in Cerium Oxide Microspheres Prepared by Internal Gelation Sol-Gel Methods

Impurity levels were measured in cerium oxide microspheres fabricated by modified internal gelation sol-gel methods. A combination of glow discharge mass spectrometry, electron impact mass spectrometry, combustion gas analysis and instrumental gas analysis were used to assess a wide range of potential elemental impurities. Low concentrations of carbon, nitrogen, and hydrogen impurities in microspheres showed a dependence on how microspheres were washed. Silicon impurities are believed to derive from silicone oil used during sol-gel processing. Spheres washed by the preferred approach had impurity levels below 100 ppm for all elements tested. For applications such as nuclear fuels, sol-gel methods could meet purity specifications as long as metal nitrate feed solutions of sufficient purity are used and microspheres are washed appropriately. In this study, cerium was used as a surrogate for plutonium-238, which is used in radioisotope power systems as a heat source, to determine whether carbon or other impurities were concentrated during internal gelation processing and remained after heat treatments. Analyses indicated low concentrations of impurities in cerium oxide microspheres after sintering steps that were well below documented limits for plutonium-238 oxide fuels. Modified washing methods, combined with a pressurized water treatment, resulted in sintered cerium oxide sol-gel microspheres with low impurity levels.

cerium oxide, sol-gel, internal gelation, microsph↗
DOE OSTI2026

Accessibility and Reactivity of Pentavalent Plutonium under Alkaline Conditions

Plutonium exhibits a particularly complex and rich aqueous chemistry due to a dynamic redox equilibrium that can result in the coexistence of four different oxidation states in aqueous solution. Within this equilibrium, pentavalent plutonium is among the most soluble and dominant oxidation states of Pu in neutral to alkaline aqueous solutions of relevance to environmental chemistries and the extreme chemical environments encountered in high-level radioactive wastes. Despite this, the fundamental chemistry of Pu­(V) under such conditions remains poorly understood. Here, in this study, we demonstrate the redox accessibility of Pu­(V) in alkaline media based on the observations of an electrochemically reversible Pu­(VI/V) couple at E 1/2 = 0.19 V vs Hg/HgO (0.29 V vs NHE), allowing us to electrolytically generate Pu­(V) in alkaline solution. These Pu­(V) solutions remained stable under ambient conditions for at least 9 days before precipitation of crystalline Na 2 [Pu V O 2 (OH) 3 ]·2H 2 O, the first structurally characterized Pu­(V) hydroxide. The modest potential of the Pu­(VI/V) redox couple allowed us to reproduce Na 2 [Pu V O 2 (OH) 3 ]·2H 2 O synthetically as well as isolate two additional novel Pu­(V)-bearing hydroxide phases Na­[Pu V O 2 (OH) 2 (H 2 O)]·2.5H 2 O and K 2 [Pu V O 2 (OH) 3 ]·2H 2 O as single crystals from alkaline reactions under mild hydrothermal conditions (<200 °C) using Na 2 O 2 and KO 2 as reductants. This reactivity is rationalized and informed by our electrochemical measurements. The properties of the Pu­(V) solids were further characterized by using vibrational and optical spectroscopy.

Kravchuk, Dmytro V. [Argonne National Laboratory (↗
DOE OSTI2026

Mark-18A Target Material Recovery Program: Initial Hot Startup

The Savannah River National Laboratory (SRNL) has been tasked by the National Nuclear Security Administration to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The Mk-18A targets were manufactured with Pu-242 then irradiated under high neutron flux in K-Reactor at the SRS from 1968 to 1978. The sixty-five Mk-18A targets are currently stored in the SRS L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The original targets were manufactured with varying quantities of Pu-242, ranging from 5.2 grams to 121.5 grams. Taking a graded approach to process start up, the lowest loaded target (FT-80-03) was selected as the first target to be received and processed at SRNL. As operational experience and knowledge is gained from processing targets, higher loaded targets will be selected for processing to increase the quantities of valuable isotopes recovered. Due to dose concerns, the receipt and processing of the Mk-18A targets is performed in the SRNL Shielded Cells Facility. The targets are stored in a double J-can confinement in the L-Area Basin. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets are loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 length) are then processed one at a time through the following processes: caustic dissolution and filtration, acidic dissolution and filtration, elutable Reillex anion exchange, diglycolamide (DGA) resin extraction, and DGA calcination. This processing results in two product streams. The first is an aqueous plutonium solution which is removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide product containing the Am and Cm as well as other lanthanide fission products which is removed from the shielded cells using a bagless transfer system. Both materials are packaged for shipment to Oak Ridge National Laboratory (ORNL). This paper will discuss the operating experience, lessons learned, and results from initial process hot startup.

Armstrong, Christopher [Savannah River National La↗
DOE OSTI2024

Mark-18A Cold Runs

The Savannah River National Laboratory (SRNL) is tasked by the National Nuclear Security Administration (NNSA) to recover highly valued isotopes from irradiated Mark-18A (Mk-18A) targets. The Savannah River Site (SRS) has sixty-five Mk-18A targets available for the recovery of the high valued materials. The sixty-five Mk-18A targets are currently stored in the L-Area Basin and will be removed one at a time and individually transported to SRNL. Upon receipt at SRNL, the Mk-18A target material will be removed from the confinement, dissolved, chemically separated, and calcined to a stable oxide. The flowsheet is designed to recover the plutonium as well as the trivalent actinides. The remaining unrecovered material will be discarded to the high activity drain (HAD) system in SRNL. A specially designed cask was procured for transport of the targets from L-Area to SRNL. Once received at SRNL, the targets will be loaded into the back of Cell 7 and resized as they enter the cell. The resized targets (1/4 lengths) will then be processed one at a time through the following processes: caustic dissolution and filtration; acidic dissolution and filtration, Reillex anion exchange, diglycolamide (DGA) cation exchange; and DGA calcination. This processing will result in two product streams. The first is an aqueous plutonium solution which will be removed from the shielded cells and taken to a glovebox for further purification and conversion to an oxide. The second is a calcined oxide containing the Am and Cm as well as other lanthanide fission products which will be removed from the shielded cells using a bagless transfer system. Both materials will be packaged for shipment to Oak Ridge National Laboratory (ORNL). All equipment to carry out this process was designed, procured or fabricated, and installed in a mock-up facility (716-4A) at SRS to allow for simulation testing in a non-radioactive area. This equipment was then dismantled and transferred from 716-4A to 773-A and installed in the SRNL Shielded Cells. After installation in the shielded cells facility testing was performed using water followed by surrogates and cold chemicals. Issues were identified during these evaluations, including equipment issues as well as technical challenges. Many of the issues were rectified during performance of the cold runs, and the remaining have a resolution identified. Table ES-1 provides a summary of all issues identified during the cold run operations, as well as the status and identified resolutions to outstanding issues.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Examination of actinide chemistry at solid-water interfaces to support advanced actinide separations (Final Report)

In this work we have examined the thermodynamics of actinide interactions with solid interfaces in the presence and absence of multi-functional complexing ligands. The term “interactions” is a global term to capture many potential reactions including aqueous complexation, sorption/surface complexation, surface-mediated redox reactions, electrostatic attraction and repulsion, surface passivation, surface precipitation, and counter-ion competition. Each of these potential reactions may control actinide interactions with a solid surface. This work is helping to develop a more fundamental understanding of actinide bonding at solid:water interfaces and examine unique surface mediated oxidation/reduction reactions which are not expected based on the reduction potentials of aqueous species. We place a large emphasis on understanding both sorption and desorption processes as a means of understanding the sorption mechanisms. This work is directly aligned with the DOE BES Heavy Element Chemistry programs interests in understanding “the chemical and physical properties of these elements to determine solution, interfacial and solid-state bonding and reactivity”.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Relative humidity threshold for oxygen generation by high-purity plutonium dioxides in 3013 containers

High-purity plutonium dioxide with adsorbed water in a sealed container can generate oxygen, but oxygen will not be generated in low water loading conditions. A threshold for oxygen generation has been observed in the hydrogen generation rate normalized by the specific power for high-purity plutonium dioxide. Normalized hydrogen generation rates above this threshold generate oxygen and below the threshold consume oxygen. The threshold rate is determined by the amount of water adsorbed onto the surface of the oxide. The relative humidity of the atmosphere that the oxide is in equilibrium with determines the amount of physisorbed water adsorbed on the surface. We calculate the relative humidity corresponding to the threshold rate as a function of mass and specific surface area of material stored in 3013 containers. Exceeding the calculated relative humidity may result in flammable gas mixtures of hydrogen and oxygen at some time during storage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Small-scale production of bespoke accelerated aging plutonium alloy

Here, this study demonstrates the 100 g scale manufacture of a plutonium alloy that ages at an accelerated rate. The resulting alloy ages six times faster than typical weapons-grade plutonium due to the addition of 238 Pu. As a major innovation, the process involved using a partial direct oxide reduction technique. This method was achieved by developing a new, complex geometry stirrer using additive manufacturing to reduce the 238Pu oxide and efficiently incorporate it into weapons-grade plutonium metal. The material was then purified by molten salt extraction and electrorefining before being alloyed with gallium. The alloy was then cold-rolled and annealed in a homogenization heat treatment. The resulting disk was characterized by metallography and differential scanning calorimetry, and the impurity content was determined using analytical chemistry techniques. The results show that a homogeneous delta phase plutonium alloy was achieved with expected microstructure and minimal impurities. This study was also successful in changing the plutonium isotopic composition by incorporating additional 238 Pu to accelerate the effects of radiation damage. This enables researchers to study long-term aging phenomena in a reduced time frame, thus avoiding the need for large-scale material production and circumventing the limitations of using naturally aged, archived plutonium.

Aging theory↗
DOE OSTI2022

Capability to Process and Characterize Uranium-Zirconium (U-Zr) and Uranium-Zirconium-Plutonium (U-Zr-Pu) Alloys

This work investigated co-reduction of anhydrous compounds of uranium, zirconium, and plutonium to produce uranium rich ternary nuclear fuel alloys. Metallothermic co-reduction is a novel method to produce all-metal nuclear fuels. Metal fuels offer thermal, fissility, compatibility, and security benefits over oxide fuels. Alloys of uranium–10% zirconium with plutonium contents of 0%, 2.5%, 5%, and 10% were produced, with yields of 60–85% of theoretical values in a traditional calciothermic bomb reduction apparatus. Microstructural analysis indicated transformation of uranium phase from gamma to beta and then alpha, in alternating lamellar plates typical of alpha phase uranium and delta phase uranium-zirconium, with zirconium and carbides at prior grain boundaries. Some of the analyses were inconclusive in their results and therefore require additional testing. Successful ternary co-reduction would simplify production of homogeneous feedstock and thereby streamline manufacture of homogeneous ternary metallic fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Two Neptunium(III) Mellitate Coordination Polymers: Completing the Series Np–Cf of Trans-Uranic An(III) Mellitates

For this work, two neptunium(III) mellitates, 237 Np 2 (mell)(H 2 O) 9 ·1.5H 2 O (Np-1α) and 237 Np 2 (mell)(H 2 O) 8 ·2H 2 O (Np-1β), have been synthesized from 237 NpCl 4 (dme) 2 by reduction with KC 8 and subsequent reaction with an aqueous solution of mellitic acid (H 6 mell). Characterization by single-crystal X-ray crystallography and UV–vis–NIR spectroscopy confirms that the neptunium is in its +3 oxidation state and both polymorphs are isostructural to the previously reported plutonium mellitates. Of the two morphologies, Np-1α is indefinitely stable in air, while Np-1β slowly oxidizes over several months. This is due to the change in the energy of the metal-ligand charge-transfer absorption exhibited by these compounds attributed to differing numbers of carboxylate bonds to Np(III), where in Np-1β the energy is low enough to result in spontaneous oxidation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

SCALE 6.2.4 Validation: Nuclear Criticality Safety

The computational bias of criticality safety computer codes must be established through the validation of the codes to critical experiments. A large collection of suitable experiments has been vetted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and made available in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). More than 600 cases from this handbook have been prepared and reviewed within the Verified, Archived Library of Inputs and Data (VALID), which is maintained by the Reactor and Nuclear Systems Division at Oak Ridge National Laboratory. The performance of the KENO V.a and KENO-VI Monte Carlo codes within the SCALE 6.2.4 code system is assessed using the VALID models of benchmark experiments. A range of nuclear cross section libraries based on Evaluated Nuclear Data File (ENDF)/B-VII.1 in both multigroup (MG) and continuous energy (CE) formats is considered. The critical experiments available to validate the KENO V.a code cover 15 broad categories of systems. These systems use a range of fissile materials, including a range of uranium enrichments, various plutonium isotopic vectors, and some mixed uranium/plutonium oxides. The physical forms of the fissile material also vary and are represented as metal, solutions, or arrays of rods or plates in a water moderator. The neutron energy spectra of the systems also vary and cover fast, intermediate, mixed, and thermal spectra. Over 550 of the total cases use the KENO V.a code for the four nuclear data libraries considered in this report.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Performance Characterization of FB-Line Neutron Multiplicity Counter and Large Neutron Multiplicity Counter

Savanah River National Laboratory’s (SRNL) Nuclear Measurements group was tasked with characterizing the performance of two neutron multiplicity counters located at SRNL. Characterization measurements were made to determine the gate width, pre-delay, deadtime parameters, triples and doubles gate fractions, detector efficiency, and operating high voltage for the Large Neutron Multiplicity Counter (LNMC) and the FB Line Neutron Multiplicity Counter (FBLNMC). The parameters were determined, shown below, and were, as to be expected, slightly different than the previous calibrations, which were performed over 20 years ago. Several Pu samples were measured to validate the characterizations of the FBLNMC and LNMC. The measurements determined the sample Pu-240 mass within <2% deviation for the pure plutonium samples and ~8% for the mixed oxide sample. The pure Pu samples had significantly better accuracy compared with the impure mixed oxide sample due to the lack of induced fission or alpha,n neutrons from impurities. Overall, the characterization of the neutron multiplicity counters, and the determination of their operability has been completed successfully.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗