SEARCH · Engineering Papers
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
This technical document (TECH) contains an analysis of several conditions deemed credible for plutonium oxide buildup in high-efficiency air particulate (HEPA) filters to determine the effect of holdup inside the filter on reactivity. This document is performed in accordance with NCS-AP-005, R3 [Ref. (1)].
An Addendum was written to the Model 9977 Safety Analysis Report for Packaging adding a new content consisting of DOE-STD-3013 stabilized plutonium dioxide materials to the authorized Model 9977 contents. The new Plutonium Oxide Content (PuO{sub 2}) Envelope will support the Department of Energy shipment of materials between Los Alamos National Laboratory and Savannah River Site facilities. The new content extended the current content envelope boundaries for radioactive material mass and for decay heat load and required a revision to the 9977 Certificate of Compliance prior to shipment. The Addendum documented how the new contents/configurations do not compromise the safety basis presented in the 9977 SARP Revision 2. The changes from the certified package baseline and the changes to the package required to safely transport this material is discussed.
Recent measurements with coded-aperture imagers demonstrate material mass determination in a holdup setting to an accuracy within a few percent. This capability is of particular interest to the Surplus Plutonium Disposition (SPD) project, which aims to dilute and dispose of surplus plutonium oxide. Gamma-ray imagers can be used to determine holdup without interrupting normal operations. In this work, we examine techniques for 3D source localization and mass determination using gamma-ray imagers. Coded-aperture imagers provide excellent source localization within the 2D image plane; however, multiple imagers operating in tandem are necessary to identify source location in 3D space. A Maximum Likelihood Expectation-Maximization (MLEM) method for fitting detector mappings is a powerful tool for accomplishing this task. MLEM allows 3D source localization to be simultaneously constrained using multiple gamma-ray imagers by constructing the basis for the MLEM fit using detector mappings from different detector locations stitched together. Each of these basis points represents a singular response from a source in 3D space and is generated using Monte Carlo simulations of sources placed individually at different locations throughout the imager’s field of view. Additionally, implementing knowledge of the physical equipment in the simulations of the glovebox used for the SPD project incorporates attenuation effects that are needed to calculate material holdup.
Shielded gloveboxes are currently being constructed to facilitate dilution and disposal of many tons of excess plutonium oxide. Measuring holdup in these gloveboxes is expected to be challenging because of the limited available lines of sight through the glovebox shielding. A system of gamma-ray imagers is being developed to provide localization and quantification for holdup. The system will be mounted above the glovebox, where there is minimal shielding. Multiple imagers with overlapping coded-aperture fields of view are employed to enable three-dimensional reconstruction. Compton reconstruction is also available to localize sources outside the coded-aperture field of view. Improved uncertainties with respect to current techniques are expected by virtue of the fixed installation, spectroscopic performance of the detectors, and iterative image reconstruction techniques. Measurement campaigns have been undertaken at an active glovebox at Savannah River Site to test the gamma-ray imaging system in an operational environment, providing a simple test of a single imager that mimics the geometry of the installation proposed for future shielded gloveboxes. Data taken during quiescent periods in the glovebox were used to measure the buildup of material on an outlet filter and record a trend over time. Calibration data was taken with known sources to simplify analysis and provide a reliable assay of the filter. Resulting images make it possible to isolate the filter from other sources and recognize compromised data. This paper will present details of the measurement and analysis methods.
The Glenn T. Seaborg Institute at Idaho National Laboratory (INL-GTSI) focuses on advancing fundamental research in the actinide sciences by providing unique opportunities to early career scientists and engineers to gain experience studying the actinide elements and their associated systems. The INL-GTSI is built from three focus areas that are based on the expertise and supporting infrastructure at INL and include solid state chemistry and physics, solution phase chemistry and physics, and forensic and isotope science. INL is the lead Laboratory for nuclear energy research and development in the U. S. and the research on nuclear fuels performed under the INL-GTSI gives good examples of solid state studies. Uranium-Molybdenum (U-Mo) alloys are leading fuel candidates for conversion of high performance research and test reactors to low-enriched fuels. During irradiation, generated fission gas accumulates into bubbles and self-organizes into a gas bubble superlattice (GBS) that effectively stores fission gases and inhibits fuel swelling. A study on the early self- organizing behavior of the GBS shows that not only grain boundaries but the interfaces between the U-Mo matrix and uranium carbide (UC) impurities are important to GBS formation.[1] In solution, understanding the complex redox behavior of plutonium in aqueous environments is critical for establishing optimized nuclear waste reprocessing solvent systems and storage tank environments. INL-GTSI researchers have produced an experimentally validated multi-scale model of the gamma radiation induced behavior of plutonium ions in concentrated aqueous HNO3 solutions.[2] Here, gamma radiation effected only minimal steady state changes in the redox distribution of the plutonium oxidation states. The redox cycling between Pu(IV) and Pu(III) is demonstrated to be mediated by the •OH/NO3• radical oxidation of Pu(III) and the H2O2/HNO3 driven reduction of Pu(IV). The INL-GTSI offers young researchers the unique chance to work directly with actinide bearing materials in a U. S. National Laboratory environment. Further topical areas of interest to the INL-GTSI include, but are not limited to, fundamental actinide properties, structure/property (electronic, magnetic, thermal) relations, actinide quantum criticality, f- electron interactions, electron correlations, computational studies, new phases, defect effects, interface interactions, isotope production and separation, forensic analytical chemistry, structure and dynamic properties of actinides in non-aqueous media, separations chemistry and kinetics for advanced nuclear fuel cycles, radiation effects, and innovative and advanced ligand design for complexation of the actinides.
The analysis of particulates from environmental sampling is routinely performed for nuclear forensics applications. In order to increase the tools available for nuclear forensics applications, capability development materials (CDMs), or reference particulates, are necessary for developing and benchmarking new analytical methods. Historically, these CDM particulates have been produced with highly controlled and characterized isotopic and size parameters for applications in developing and benchmarking particle sizers and mass spectrometry analytical methods [1- 4]. However, the development of CDMs with highly characterized particle morphology and phase of the particles is vital for aiding in the development and benchmarking of particle analytical methods for those parameters. In many cases, key properties such as crystallographic phase, morphology, and microstructure, may be correlated to processing history of environmental sampling particulates [5]. To that end, this work investigates determining the structural fingerprint of produced Pu-doped uranium oxide CDMs using transmission electron microscopy (TEM). The particulates, one of which shown in Figure 1, were synthesized to Fig. 1. Single particulate of uranium oxide fabricated using the THESEUS technique. develop capabilities for environmental sampling investigations. These particles are monodisperse at diameter of 1 μm and a range of plutonium concentrations from 0- 1,000 ppm. Given their small diameter, TEM analysis was ideal for studying the particulate structural fingerprint in detail. Previous investigations confirmed the external homogeneity of the particulates at different plutonium concentrations, however this analysis focuses on determining the grain structure, phase distribution, and porosity of the particulates.
Explore the source record for details and available documents.
Auger electron spectroscopy (AES) was used to characterize the oxide scale formed on an ~0.5 wt. % Ga plutonium metal alloy. Analysis of changes to the characteristic Auger transitions for Pu and O during sputter depth profiles were used in combination with previously measured sputter yields to determine the thickness of the oxide scale and provide insight into its chemistry. From this data, the parabolic kinetics for the Pu-Ga alloy were determined as a function of relative humidity, RH = 14–92%, and oxygen partial pressure, p(O 2 ) = 31–157 Torr, at 25°C. For each relative humidity investigated, the parabolic rate constant was found to be independent of oxygen partial pressure. Furthermore, power-law fits to the relative humidity effect on the rate constant was found to have a power of n = 1.265, indicating the diffusion species which rate-limits the reaction isn’t molecular water, which subsequent dissociates at the oxide-metal interface, or hydroxyl from dissociation at the gas-surface interface.
Between 1944 and 1989, 54.5 metric tons of the United States’ weapons-grade plutonium and an additional 12.9 metric tons of fuel-grade plutonium were produced and separated from irradiated fuel at the Hanford Site. Acidic high-activity wastes containing around 600 kg of plutonium were made alkaline and discharged to underground storage tanks from separations, isolation, and recycle processes to yield average plutonium concentration of about 0.003 grams per liter (or ~0.0002 wt%) in the ~200 million liter tank waste volume. The plutonium is largely associated with low-solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO2∙xH2O, could undergo sufficient crystal growth through dissolution and reprecipitation in the alkaline tank waste to potentially become separable from neutron absorbing constituents by settling or sedimentation. Thermodynamic considerations and laboratory studies of systems chemically analogous to tank waste show that the plutonium formed in the alkaline tank waste by precipitation through neutralization from acid solution probably entered as 2–4-nm PuO2∙xH2O crystallite particles that, because of their low solubility and opposition from radiolytic processes, grow from that point at exceedingly slow rates, thus posing no risk of physical segregation.
The invention provides a method for extracting plutonium from spent nuclear fuel, the method comprising supplying plutonium in a first aqueous phase; contacting the plutonium aqueous phase with a mixture of a dielectric and a moiety having a first acidity so as to allow the plutonium to substantially extract into the mixture; and contacting the extracted plutonium with second a aqueous phase, wherein the second aqueous phase has a second acidity higher than the first acidity, so as to allow the extracted plutonium to extract into the second aqueous phase. The invented method facilitates isolation of plutonium polymer without the formation of crud or unwanted emulsions.
High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.
The measurement of uranium and plutonium holdup in plants during process activity and for decommissioning is important for nuclear safeguards and material control. The amount of plutonium and uranium holdup in glove-boxes, pipes, ducts, and other containers has been measured for several decades using both neutron and gamma-ray techniques. For the larger containers such as hot cells and glove-boxes that contain processing equipment, the gamma-ray techniques are limited by self-shielding in the sample as well as gamma absorption in the equipment and associated shielding. The neutron emission is more penetrating and has been used extensively to measure the holdup for the large facilities such as the MOX processing and fabrication facilities in Japan and Europe. In some case the totals neutron emission rates are used to determine the holdup mass and in other cases the coincidence rates are used such as at the PFPF MOX fabrication plant in Japan. The neutron emission from plutonium and MOX has 3 primary source terms: 1) Spontaneous fission (SF) from the plutonium isotopes, 2) The (α,n) reactions from the plutonium alpha particle emission reacting with the oxygen and other impurities, and 3) Neutron multiplication (M) in the plutonium and uranium as a result of neutrons created by the first two sources. The spontaneous fission yield per gram is independent of thickness, whereas, the above sources 2) and 3) are very dependent on the thickness of the deposit. As the effective thickness of the deposit becomes thin relative to the alpha particle range, the (α,n) reactions and neutrons from multiplication (M) approach zero. In any glove-box, there will always be two primary modes of holdup accumulation, namely direct powder contact and non-contact by air dispersal. These regimes correspond to surfaces in the glove-box that have come into direct contact with the process MOX powder versus surface areas that have not had direct contact with the powder. The air dispersal of PuO{sub 2} particles has been studied for several decades by health physicists, because the primary health hazard of plutonium is breathing the airborne particles. The air dispersal mechanism results from the smaller particles in the top layer of powder that are lifted into the air by the electrostatic charge buildup from the alpha decay process, and the air convection carries the particles to new more distant locations. If there is open plutonium powder in a glove-box, the surfaces at more distant locations will become contaminated over time. The range of an alpha particle in a solid or powder is a function of the particle energy, the material density, and the atomic number A of the material. The average energy of a plutonium alpha particle is ∼5.2 MeV and the range in air is ∼37 mm. The range in other materials can be estimated via the Bragg-Kleenman equation. For plutonium, A is 94, and the typical density for a single particle is ∼11.5 g/cm{sup 3}, but for a powder, the density would be less because of the air packing fraction. The significance of the small diameter is that the range of the alpha particle is ∼50 μm for powder density 2.5 and significantly less for a single particle with density 11.5, so the thin deposit of separate small particles will have a greatly reduced (α,n) yield. The average alpha transit length to the surface in the isolated MOX particle would be < 2.5 μm; whereas, the range of the alpha particle is much longer. Thus, most of the alpha particles would escape from the MOX particle and be absorbed by the walls and air. The air dispersal particles will have access to a large surface area that includes the walls, whereas, the powder contact surface area will be orders of magnitude smaller. Thus, the vast majority of the glove-box surface area does not produce the full (α,n) reaction neutron yield, even from the O{sub 2} in the PuO{sub 2} as well as any impurity contamination such as H{sub 2}O. To obtain a more quantitative estimate of the neutron (α,n) yields as a function of holdup deposit thickness, we have used MCNPX calculations to estimate the absorption of alpha particles in PuO{sub 2} holdup deposits. The powder thickness was varied from 0.1 μm to 5000 μm and the alpha particle escape probability was calculated. As would be expected, as the thickness approaches zero, the escape probability approaches 1.0, and as the thickness gets much greater than the alpha particle range (∼50 μm), the escape probability becomes small. Typically, the neutron holdup calibration measurement are performed using sealed containers of thick MOX that has all 3 sources of neutrons [SF, (α,n), and M], and no significant impurities. Thus, the calibration counting rates need to include corrections for M and (α,n) yields that are different for the holdup compared with the calibration samples. If totals neutron counting is used for the holdup measurements, the variability of the (α,n) term needs to be considered.
In the framework of the WPNCS Expert Group on Uncertainty Analysis for Criticality Safety Assessment, we propose a blind numerical benchmark. Since the WPNCS Expert Group on Experimental Needs had highlighted the poor database of representative experiments of low-moderated MOX powders, this blind benchmark is focused on MOX wet powders Three PuO{sub 2} contents are proposed: 100%, 30% and 12.5%. Two Pu isotopic vectors are considered: a conservative one (0%, 71%, 17%, 11%, 1% for {sup 238}Pu up to {sup 242}Pu) and a realistic one obtained from reprocessed LWR 30 GWd/T fuels. For these 6 cases the powder moisture rate is 3%. The MOX fissile medium is a sphere surrounded by 20 cm water reflector. The required results are the calculated k{sub eff} and the associated prior uncertainty. Using benchmarks experiments, contributors should provide the projected-k{sub eff} and the corresponding posterior uncertainty. CRISTAL-V2 results based on JEFF3.1.1 nuclear data are presented. As an example, in the 30% PuO{sub 2} challenging case, the calculation uncertainty is decreased from 1160 pcm (1 σ) down to 450 pcm using EOLE and MASURCA critical Pu experiments. This uncertainty depends slightly on 'conservative' or 'realistic' Pu because the representativity factor does not vary significantly with the Pu vector. (authors)
The ARIES Program at the Los Alamos National Laboratory removes plutonium metal from decommissioned nuclear weapons, and converts it to plutonium dioxide in a specially-designed Direct Metal Oxidation furnace. The plutonium dioxide is analyzed for specific surface area, particle size distribution, and moisture content. The purpose of these analyses is to certify that the plutonium dioxide powder meets or exceeds the specifications of the end-user, and the specifications for the packaging and transport of nuclear materials. Analytical results from plutonium dioxide from ARIES development activities, from ARIES production activities, from muffle furnace oxidation of metal, and from metal that was oxidized over a lengthy time interval in air at room temperature, are presented. The processes studied produce plutonium dioxide powder with distinct differences in measured properties, indicating the significant influence of oxidation conditions on physical properties.
A dual-particle imager with sensitivity to both neutrons and photons has been developed for the detection of special nuclear material. This system has the capability to provide spectral information on detected sources that is useful for safeguard applications. In nuclear facilities with high photon background rates, neutron detection can aid in identification and verification capabilities. The dual-particle imager creates a reconstructed neutron spectrum by correlating neutron counts in pairs of liquid scintillators. The reconstructed spectrum, a combination of energy deposited and time-of-flight, contains resolution effects that blur the true, emitted source-spectrum. To reduce the impact of resolution effects, the reconstructed spectrum can be unfolded to create an estimated source spectrum. A statistical technique, maximum-likelihood expectation-maximization, has been shown to produce an unfolded result for Cf-252 and Am-Be. This technique requires a system response-matrix that contains reconstructed spectra for single neutron energies. Since measuring mono-energetic neutrons is difficult, the system-response-matrix is simulated. Monte Carlo modeling with MCNPX-PoliMi has been show to produce an accurate system response. This work expands on previous neutron unfolding results by showing results for mixed-oxide fuel and plutonium metal measured with the dual-particle imager. Furthermore these results also demonstrate the ability of the dual-particle imager to discrimination between spontaneous fission neutrons and (α,n) reaction neutrons.
The oxidation of an ~1.5 at% Ga plutonium metal alloy in dry air as a function of time was studied using a combination of FIB-SEM, AES, and XRD. Analysis of SEM images of FIB cross-sections of the oxide allowed for direct measurement of the oxide thickness as well as determination of heterogeneity in the extent of the oxide. AES analysis performed on cross-sections of the oxide allowed for quantitative analysis of the atomic composition and semi-quantitative analysis of the chemical composition. Initial preparation of the sample was found to induce a martensitic transformation to α' - Pu with an average thickness of 233 nm. Electropolishing was found to be sufficient to fully remove the α' layer with creation of only a minimum oxide layer. Growth of the oxide in dry air was found to initially follow parabolic kinetics, with a parabolic rate constant orders of magnitude higher than expected for δ-Pu but in-line with values extrapolated to room temperature for α-Pu. Following complete oxidation of the initial α` layer, further regions of plutonium metal depleted in gallium were observed to form at the oxide / metal interface. However, gallium, as Ga 2 O 3 , was found to be enriched at the surface and present in the oxide in an amount comparable to the relative Ga:Pu ratio of the bulk metal alloy. Finally, the results presented here serve to bridge a diverse array of previous experimental and theoretical data and provide the first comprehensive view of the fate of the δ-stabilizer during formation of an oxide.
The investigation of physical properties of uranium and plutonium mixed oxide (MOX) fuels is important for the development of fast reactor fuels. It is well known that MOX is a nonstoichiometric oxide, and the physical properties change drastically with the Oxygen-to-Metal (O/M) ratio. A control technique for O/M ratio was established for measurements of high temperature properties of uranium and plutonium mixed oxide fuels. Sintering behavior, thermal expansion and O/M change of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} and (U{sub 0.7}Pu{sub 0.3})O{sub 1.99} were investigated in PO{sub 2}-controlled atmosphere which was controlled by H{sub 2}/H{sub 2}O gas system. Sintering behavior changed drastically with O/M ratio, and shrinkage of (U{sub 0.7}Pu{sub 0.3})O{sub 2.00} was faster and more advanced at lower temperatures as compared with (U{sub 0.7}Pu{sub 0.3})O{sub 1.99}. Thermal expansion was observed to be slightly increased with decreasing O/M ratio. (authors)
The safe and secure management of Pu is a matter of serious international concern, with ∼250 t of separated Pu currently stockpiled worldwide. Nearly 50% of this material is in long-term storage in the UK whilst the British government develops its options for final treatment and disposition. Options for this are: (i) reuse as a fuel in modern reactors (e.g. as MOX fuel or within a fast reactor) prior to disposal; (ii) prompt immobilisation for disposal as soon as is practicable; or, (iii) continued long term storage (prior to disposition). The focus of the work presented here is options (ii) and (iii). The R and D needs of both are now pressing: in option (iii) due to it being the current default; in option (ii) because of a comparative lack of R and D 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, 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 characterisation methods for Pu bearing materials. In response to these needs, the DISTINCTIVE Consortium (Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories, a multi-disciplinary collaboration of 11 universities and 3 key industry partners from across the UK's civil nuclear sector) has been seeking to: - Understand how the structure and properties of PuO{sub 2} change with time in the presence of H{sub 2}O. - Understand the roles these processes play in gaseous product evolution from PuO{sub 2} in storage. - Understand radiation-induced amorphization and dissolution kinetics of Pu waste forms. - Develop novel, fast neutron based radiometric methods for the quantification, isotopic composition assessment and remote imaging of Pu-bearing materials. This paper describes recent progress made towards achieving these objectives. (authors)