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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

MIS High-Purity Plutonium Oxide Hydride Product 5501579 (SSR124): Final Report

A high-purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample 5501579 represents process plutonium oxides from hydride oxide from Rocky Flats that are currently stored in 3013 containers. After calcination to 950°C, the material contained 87.42% plutonium with no major impurities. This study followed over time, the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions were measured periodically over a six year period. The maximum observed gas pressure was 124 kPa. The increase over the initial pressure of 70 kPa was primarily due to generation of nitrogen and carbon dioxide gas. Hydrogen and oxygen were minor components of the headspace gas. At the completion of the study, the internal components of the sealed container showed signs of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

MIS High Purity Plutonium Oxide Metal Oxidation Product MT1490 (SSR125): Final Report

A high purity plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample MT1490 represents process plutonium oxides from several metal oxidation operations that are currently stored in 3013 containers. This study followed over time, the gas pressure of a sample with nominally 0.5 weight percent water in a sealed container with an internal volume scaled to 1/500 th of the volume of a 3013 container. The sample was a mixture of as-received material, material calcined at 800°C and material calcined at 950°C. This material contained 87% plutonium with no major impurities. Gas compositions were measured periodically over a nine year period. The maximum observed gas pressure of 118 kPa was reached during the ninth year. The increase over the initial pressure of 95 kPa was primarily due to generation of nitrogen and hydrogen gas. Oxygen was a minor component of the headspace gas. At the completion of the study, the inner bucket showed signs of corrosion, including pitting, in the container material contact region.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

MIS Shelf-Life Final Report for Plutonium Oxide Item 1000089 (SSR136 and SSR136A) from Rocky Flats Peroxide Precipitation and Calcination Process

A plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied to determine the gas generation and corrosion behavior in a storage environment. Sample 1000089 represents product-quality plutonium oxides produced in the Rocky Flats peroxide precipitation process currently stored in 3013 containers. This study followed over time the gas pressure and composition of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. Gas compositions had been measured periodically over almost 13 years. The maximum observed gas pressure was 204 kPa and was related to a temperature excursion. The maximum observed gas pressure from gas generation was 126 kPa. The increase over the initial pressure of 86.7 kPa was due to the generation of hydrogen, oxygen, and nitrogen. Carbon dioxide, methane and carbon monoxide were minor components of the headspace gas. The material exhibited unusual behavior in that the atmosphere reached flammable levels of hydrogen and oxygen within the first 30 days and remained flammable for the duration of the experiment. It has been determined that the sample inside the reactor was a mixture of the AR material, material calcined at 800 °C, and material calcined at 950 °C. The experiment was terminated in 2012 and unloaded in 2016. A new reactor, SSR136A, was loaded with the original sample freshly calcined at 950 °C. The total pressure inside SSR136A remained close to the initial value of 93 kPa. The total pressure is composed of He (steady levels after the reloading) and hydrogen which is 10% of the reactor’s total pressure. Nitrogen and oxygen compose trace amounts of the reactor headspace. The internal components of SSR136 were inspected in 2021, five years after it was removed from the array. The inside surfaces of the inner bucket exposed to the original sample for 4,576 days had an etched appearance, and pit-like features were observed in the microscopic analysis. No corrosion was observed on the SSR136A inner bucket that held the freshly calcined sample that was exposed for 1,303 days.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

MIS High-Purity Plutonium Oxide Metal Oxidation Product BLO-39-11-14-004 (SSR132 and 132A): Final Report

A high-purity, reactor-grade plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. BLO-39-11-14-004 is a high specific power oxide that represents Hanford process plutonium oxides from oxalate process product, oxalate precipitation of impure plutonium and pure mixed uranium/plutonium nitrate solutions, PFP NDA and impure and scrap plutonium oxides 80-85% from PFP and the 300 area. It also represents plutonium oxides from oxalate precipitation at LANL and Savannah River. After calcination to 950 °C, the material contained 80% plutonium and 7% americium with no major impurities. This study followed over time the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/500th of the volume of a 3013 container. A leak occurred in the original reactor (SSR132) which was terminated after 3 years. The sample was reloaded (SSR132A) and an additional set of data was collected. Results from both loadings are discussed. Gas compositions were measured periodically over three (SSR132) and four and a half (SSR132A) year periods. The maximum observed gas pressures were 93 kPa and 94 kPa, respectively. The increase over the initial pressure was primarily due to generation of nitrogen gas. For SSR132, the maximum observed gas pressure for hydrogen was 1.2 kPa and oxygen was 9.3 kPa (includes oxygen from leak). For SSR132A, the maximum observed gas pressure for hydrogen was 5.5 kPa and oxygen was 0.9 kPa. At the completion of the study, the internal components of the sealed container showed limited signs of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

MIS High-Purity Plutonium Oxide Oxalate Precipitation/Calcination Product PBO-47-09-012-023 (Final Report)

A high-purity fuel-grade plutonium dioxide material from the Material Identification and Surveillance (MIS) Program inventory has been studied with regard to gas generation and corrosion in a storage environment. Sample PBO-47-09-012-023 originated from the Hanford continuous oxalate precipitation and calcination process and represents plutonium oxides from multiple processes currently stored in 3013 containers. After calcination to 950°C, the material contained 87.5% plutonium/americium with no major impurities. This study followed over time the gas pressure of a sample with nominally 0.5 wt% water in a sealed container with an internal volume scaled to 1/400 th of the volume of a 3013 container. Gas compositions were measured periodically over a three year period. The maximum observed gas pressure was 96 kPa. The increase over the initial pressure was primarily due to generation of hydrogen (5.7 kPa) and nitrogen (2 kPa) gas. Oxygen was a minor component of the initial headspace gas and was depleted. At the completion of the study, the internal components of the sealed container did not show significant signs of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Conceptual Design for the Pilot-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

This report describes a conceptual design for a pilot-scale capability to produce plutonium oxide for use as exercise and reference materials, and for use in identifying and validating nuclear forensics signatures associated with plutonium production. This capability is referred to as the Pilot-scale Plutonium oxide Processing Unit (P3U), and it will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including plutonium dioxide (PuO2) dissolution, purification of the Pu by ion exchange, precipitation, and conversion to oxide by calcination.

plutonium oxide↗
DOE OSTI2019

Probing Orbital Bonding and Surface Interactions of Hydrolyzed Actinide Oxides using Synchrotron X-ray, Vibrational Spectroscopy, and Electron Microscopy

Actinide dioxides (MO 2 , M = Th, U, Np, Pu, Am, Cm) are the simplest actinide compounds and often used as long term storage materials. They can provide fundamental insights into 5f/6d orbital bonding and electronic structure. Scientists have been trying to understand the role of 4f, 5f, and 6d electrons in covalent bonding for decades. Most hydrolysis studies of actinide materials involve uranium and uranium oxides due to their relevance for nuclear fuel and waste. Very few hydrolysis studies, if any, involve thorium, neptunium, and trans-plutonium dioxides or mixed actinides. Actinide surface speciation can be investigated in great detail with spectroscopic techniques such as EXAFS, vibrational spectroscopy, or electron microscopy. Actinide mineral-water interface reactions have been thoroughly summarized for non-oxide actinides; however, the reaction mechanisms provide useful ideas into the incorporation of water into the actinide lattice as a result of hydrolysis. Hydrolysis will have an incremental effect on the crystalline lattice allowing for models to predict the next changes to the electronic states.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Investigation of process history and underlying phenomena associated with the synthesis of plutonium oxides using Vector Quantizing Variational Autoencoder

Accurate, high throughput, and unbiased analysis of plutonium oxide particles is needed for analysis of the phenomenology associated with process parameters in their synthesis. Compared to qualitative and taxonomic descriptors, quantitative descriptors of particle morphology through scanning electron microscopy (SEM) have shown success in analyzing process parameters of uranium oxides. Among other candidates, a neural network called a Vector Quantizing Variational Autoencoder (VQ-VAE) has shown the ability to quantitatively describe particle morphology to attain >85% accuracy in identifying uranium oxide processing routes. We utilize a VQ-VAE to quantitatively describe plutonium dioxide (PuO 2 ) particles created in a designed experiment and investigate their phenomenology and prediction of their process parameters. PuO 2 was calcined from Pu(III) oxalates that were precipitated under varying synthetic conditions that related to concentrations, temperature, addition and digestion times, precipitant feed, and strike order; the surface morphology of the resulting PuO 2 powders were analyzed by SEM. A pipeline was developed to extract and quantify useful image representations for individual particles with the VQ-VAE, then further reduce the dimensionality of the feature space using a bottlenecking neural network fit to perform multiple classification tasks simultaneously. The reduced feature space could predict process parameters with greater than 80% accuracies for some parameters with a single particle. They also showed utility for grouping particles with similar surface morphology characteristics together. Both the clustering and classification results reveal valuable information regarding which chemical process parameters chiefly influence the PuO 2 particle morphologies: strike order and oxalic acid feedstock. Doing the same analysis with multiple particles was shown to improve the classification accuracy on each process parameter over the use of a single particle, with statistically significant results generally seen with as few as four particles in a sample.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Interaction of Molecular Hydrogen on α-Pu at 300 K: Nucleation of Hydrides and Pu-Catalyzed Carbides

An α-Pu sample was studied in its “as-received” and “sputtered” state by both X-ray photoelectron spectroscopy (XPS) and time-of-flight secondary ion mass spectroscopy (ToF-SIMS). Analysis of the “as-received” surface indicated significant carbon and oxygen, with oxygen signal pertaining to plutonium dioxide (PuO 2 ) and sesquioxide (Pu 2 O 3 ). The “sputtered” clean metal surface was then dosed with hydrogen (H 2 ) gas at 300 K and found to nucleate hydride species (PuH x ). Interestingly, H 2 was observed to be critical for nucleation of carbides (Pu x C y ) by acting as a catalyst for Pu’s reaction with background gases. Because Pu metal has a high affinity for oxygen, reoxidation occurred under UHV, forming Pu x O y and disrupting both PuH x and Pu x C y growth. Observations from the experiments were modeled for the α-Pu(020) surface, illustrating the ability of H 2 to readily dissociate onto sputtered and speciated surfaces at 300 K. The projected density of states from these calculations were compared to experiments, showing good agreement between the Pu 5f, 6d, and 7s states and the H 1s, O 2p, and C 2p valence states. Overall, the above results indicated that the formed hydride layer is unstable at 300 K in UHV conditions.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Improving the Estimates of Waste from the Recycling of Used Nuclear Fuel - 13410

Estimates are presented of wastes arising from the reprocessing of 50 GWD/tonne, 5 year and 50 year cooled used nuclear fuel (UNF) from Light Water Reactors (LWRs), using the 'NUEX' solvent extraction process. NUEX is a fourth generation aqueous based reprocessing system, comprising shearing and dissolution in nitric acid of the UNF, separation of uranium and mixed uranium-plutonium using solvent extraction in a development of the PUREX process using tri-n-butyl phosphate in a kerosene diluent, purification of the plutonium and uranium-plutonium products, and conversion of them to uranium trioxide and mixed uranium-plutonium dioxides respectively. These products are suitable for use as new LWR uranium oxide and mixed oxide fuel, respectively. Each unit process is described and the wastes that it produces are identified and quantified. Quantification of the process wastes was achieved by use of a detailed process model developed using the Aspen Custom Modeler suite of software and based on both first principles equilibrium and rate data, plus practical experience and data from the industrial scale Thermal Oxide Reprocessing Plant (THORP) at the Sellafield nuclear site in the United Kingdom. By feeding this model with the known concentrations of all species in the incoming UNF, the species and their concentrations in all product and waste streams were produced as the output. By using these data, along with a defined set of assumptions, including regulatory requirements, it was possible to calculate the waste forms, their radioactivities, volumes and quantities. Quantification of secondary wastes, such as plant maintenance, housekeeping and clean-up wastes, was achieved by reviewing actual operating experience from THORP during its hot operation from 1994 to the present time. This work was carried out under a contract from the United States Department of Energy (DOE) and, so as to enable DOE to make valid comparisons with other similar work, a number of assumptions were agreed. These include an assumed reprocessing capacity of 800 tonnes per year, the requirement to remove as waste forms the volatile fission products carbon-14, iodine-129, krypton-85, tritium and ruthenium-106, the restriction of discharge of any water from the facility unless it meets US Environmental Protection Agency drinking water standards, no intentional blending of wastes to lower their classification, and the requirement for the recovered uranium to be sufficiently free from fission products and neutron-absorbing species to allow it to be re-enriched and recycled as nuclear fuel. The results from this work showed that over 99.9% of the radioactivity in the UNF can be concentrated via reprocessing into a fission-product-containing vitrified product, bottles of compressed krypton storage and a cement grout containing the tritium, that together have a volume of only about one eighth the volume of the original UNF. The other waste forms have larger volumes than the original UNF but contain only the remaining 0.1% of the radioactivity. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Laser-induced annealing of aged PuO 2

Plutonium dioxide (PuO 2 ) is an important compound used in nuclear fuel, irradiation targets, and heat sources. As such, improved understanding of its structural and spectroscopic properties has numerous applications. Alpha particle-induced damage of a PuO 2 crystal lattice modifies several properties of its Raman spectrum, including band intensities, positions, and widths. The decay also induces growth of new bands and creates electron-trapped defects with luminescent properties. Herein, we show for the first time that laser-induced heating can reverse damage to an aged and damaged PuO 2 lattice. Using automated instrumentation to heat a single 6–10 μm spot to temperatures above 1,300°C, we show that laser-induced annealing of aged PuO 2 results in restoration of T2g band intensity with a concomitant decrease in defect band intensity, a result that indicates laser annealing can be used to reverse age-damage in PuO 2 . This laser annealing approach permitted in situ observation of temperature-dependent Raman spectral changes, thereby providing insight into the thermodynamics of structural alterations in a radiolytically damaged PuO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Interactions of HCl and H2O with the surface of PuO2

In order to explore the potential of heat treatment to decontaminate chloride-contaminated legacy plutonium dioxide (PuO2) powders from the UK stockpile, samples retrieved from storage have been heated in air from 400 to 950 C. These samples also contain high levels of other adsorbed gases from the atmosphere, including water. The amounts of chloride remaining on the PuO2 particles after heat treatment (measured by a caustic leaching process) decrease whilst the amounts of volatilised chloride increase with increasing heat treatment temperature. Clear evidence for a non-leachable (strongly bound) chloride species on the PuO2 surface is found from the thermal treatments. The lattice parameter decreases with increasing heat treatment temperature, reflecting annealing of structural defects caused by over 35 years of radiation damage, with no change in the fcc Fm3m crystal structure. Heating chloride-contaminated PuO2 powder to ~230 C and cooling back to ambient temperature in a sealed vessel reveals the production of H2, He, NO and CO gases. Water adsorption/desorption behaviour with the untreated PuO2 powder is remarkably different than PuO2 which had previously been heat treated at 700 C. From thermal treatments in open and sealed systems it is concluded that water and chloride coadsorb and interact on the PuO2 surface and each affects the adsorption/desorption behaviour of the other. These data also support practical considerations for repackaging of chloride-contaminated PuO2 for long term safe and secure storage at Sellafield.

Sutherland-Harper, Sophie↗
DOE OSTI2023

Probing the thermal decomposition of plutonium (III) oxalate with IR and Raman spectroscopy, X-ray diffraction, and electron microscopy

The thermal decomposition of Pu(III) oxalate was analyzed by Raman microspectroscopy, infrared spectroscopy, scanning electron microscopy, and powder X-ray diffraction. These data show that crystalline Pu 2 (C 2 O 4 ) 3 •9H 2 O progressively loses water and oxalate ligands as it is heated, which leads to a decrease in long-range lattice ordering, though minimal changes are observed in gross crystalline morphology. The onset of PuO 2 formation was observed between 200 - 250 °C. Thermal decomposition of oxalate ligands leads to the formation of CO 2 and plutonium oxalate-carbonate moieties, which had not been observed in previously published thermogravimetric measurements of Pu(III) oxalate. Formation of plutonium oxalate-carbonate moieties is believed to be associated with a change in the plutonium oxidation state from 3+ to 4+, which occurs prior to PuO 2 formation. The data provided herein demonstrate the rich spectroscopic nature of a rather underexplored, and technologically relevant, plutonium system. Ideally these results will further future investigations into the Pu(III) oxalate system both experimentally and computationally.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Nuclear Considerations for the Application of Lanthanum Telluride in Future Radioisotope Power Systems

Thermoelectric-based radioisotope power systems (RPSs) produced in the United States convert the heat generated by the radioactive emission of alpha particles from plutonium dioxide ( 238 puO 2 ) into electricity by means of the Seebeck effect [1]. Certain designs for thermoelectric convertors propose the use of lanthanum telluride (La 3 Te 4 ) materials due to their significant conversion capabilities [2]. The generation of neutrons from spontaneous fission and alpha-neutron (α,n) reactions is also associated with the decay of 238 PuO 2 . A portion of these neutrons will interact with the thermoelectric materials and induce trace amounts of transmutation reactions in various lanthanum and tellurium isotopes. While very small quantities of several transmutation products are predicted, the most significant reaction channels are expected to produce trace amounts of iodine which will accumulate over time. Although iodine is classified as a halogen, it is the least reactive of the halogens, and as such, it is the most likely to be able to chemically convert back into the molecule I 2 . Since I 2 is a gas at RPS temperatures, it may be possible for iodine to attack other components in the thermoelectric cavity of an RPS system. Iodine reacts easily with metals to produce a wide variety of salts. This behavior could affect the performance of La 3 Te 4 thermoelectric devices, particularly the segmented architectures that include multiple sets of bonding and metallization layers. In this type of architecture, several segments of different thermoelectric materials are joined to increase the average thermoelectric figure of merit over a relatively large temperature gradient. It is plausible that sophisticated bonding/metallization layers could be required to join the segment interfaces to each other and to the cold- and hot-shoe materials. The long-term stability and performance of these segmented material combinations could degrade as a result of the potential formation and reactions of metal-iodide compounds at the segment interfaces. This paper (1) investigates the degree to which, if any, this process may threaten potential La 3 Te 4 thermoelectric technologies, (2) presents calculations of the amount of iodine generated over the operational life of a radioisotope thermoelectric generator design, and (3) discusses the potential effects of the resulting material's chemical reactions in a segmented couple-level architecture containing La 3 Te 4 . Conclusions drawn from combined particle transport, transmutation, and thermochemical calculations for La 3 Te 4 thermoelectric materials undergoing a notional 20-year mission scenario suggest that there is no significant potential for transmutation-induced thermoelectric (TE) performance degradation.

Smith, Michael↗
DOE OSTI2014

Nuclear risk assessment for the Mars 2020 mission environmental impact statement.

In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. An alternative option being considered is a set of solar panels for electrical power with up to 80 Light-Weight Radioisotope Heater Units (LWRHUs) for local component heating. Both the MMRTG and the LWRHUs use radioactive plutonium dioxide. NASA is preparing an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS will include information on the risks of mission accidents to the general public and on-site workers at the launch complex. This Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG or LWRHU options to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of both options for the EIS.

NESDPS Office of Nuclear Energy Space and Defense ↗
DOE OSTI2015

Design of the Laboratory-Scale Plutonium Oxide Processing Unit in the Radiochemical Processing Laboratory

This report describes a design for a laboratory-scale capability to produce plutonium oxide (PuO 2 ) for use in identifying and validating nuclear forensics signatures associated with plutonium production, as well as for use as exercise and reference materials. This capability will be located in the Radiochemical Processing Laboratory at the Pacific Northwest National Laboratory. The key unit operations are described, including PuO 2 dissolution, purification of the Pu by ion exchange, precipitation, and re-conversion to PuO 2 by calcination.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

MEMORANDUM OF UNDERSTANDING between Los Alamos National Laboratory (LANL) and Savannah River National Laboratory (SRNL) for Analytical Chemistry Support for Oxide Production Samples

This MOU establishes the responsibilities and requirements for the packaging and transport of plutonium dioxide (PuO 2 ) samples for shipment from LANL to SRNL. The scope includes the shipping, packaging, quality assurance (QA), inspection, and documentation requirements to successfully obtain the chemical and isotopic characteristics of the PuO 2 . The requirements in this document are necessary, but not sufficient to execute this work and do not imply exemption from contractual requirements at either site. This document is not intended to specify all of the processes and procedures necessary to execute this work. This MOU also establishes appropriate requirements, goals, and expectations. Each party will establish a technical point of contact (POC) who will be responsible for addressing issues as they arise

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Investigation into Practical Implementations of a Zero Knowledge Protocol.

In recent years, the concept of Zero Knowledge Protocols (ZKP) as a useful approach to nuclear warhead verification has become increasingly popular. Several implementations of ZKP have been proposed, driving technology development toward proof of concept demonstrations. Whereas proposed implementations seem to fall within the general class of template-based techniques, all physical implementations of ZKPs proposed to date have a complication: once the instrumentation is prepared, it is no longer authenticatable; the instrument physically contains sensitive information. In this work we explore three different concepts that may offer more authenticatable and practical ZKP implementations and evaluate the sensitive information that may be at risk when doing so: sharing a subset of detector counts in a preloaded image (with spatial information removed), real-time image subtraction, and a new concept, CONfirmation using a Fast-neutron Imaging Detector with Anti-image NULL-positive Time Encoding (CONFIDANTE). CONFIDANTE promises to offer an almost ideal implementation of ZKP: a positive result is indicated by a constant rate at all times enabling the monitoring party the possibility of full access to the instrument before, during, and after confirmation. A prototype of CONFIDANTE was designed, built, and its performance evaluated in a series of measurements of several objects including a set of plutonium dioxide Hemispheres. Very encouraging results proving feasibility are presented. 1 Rebecca is currently a graduate student in Nuclear Engineering at UC Berkeley

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗