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At least 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Non-Destructive Plutonium Assay in Pyroprocessing Bulk Materials with a 3D Boron-Coated-Straw Detector Array

Assessment of plutonium content through all the processing steps is needed and is a challenging task. While several destructive assay methods have been developed for nuclear material accountability, a nondestructive assay (NDA) system for the assessment of plutonium in bulk materials is still needed. This system should withstand pyroprocessing harsh environments and have consistent sensitivity and accuracy despite different fuel form factors. We aim to enable the accurate assessment of the plutonium content of nuclear material during pyroprocessing to improve the separation process and enhance its proliferation resistance. We plan to achieve this goal by developing and demonstrating a new 3D boron-coated-straw neutron detector array (3D-BCSDA) with high efficiency and spatial resolution.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Complex Method of Non-Destructive Measurements of Plutonium Composition and Mass

A complex method of performing non-destructive measurements of the isotopic composition and mass of plutonium that does not require external data on {sup 242}Pu content is described. The method includes a set of measurements performed by means of γ-spectrometry, calorimetry, and neutron multiplicity. Determinations of the metrological characteristics showed that in terms of accuracy it greatly surpasses the existing analogues, approaching the coulometric and isotopic dilution methods. The method is recommended for use in the industrial production of mixed uranium-plutonium fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Utilization of non-weapons-grade plutonium and highly enriched uranium with breeding of the {sup 233}U isotope in the VVER reactors using thorium and heavy water

A method for joint utilization of non-weapons-grade plutonium and highly enriched uranium in the thorium–uranium—plutonium oxide fuel of a water-moderated reactor with a varying water composition (D{sub 2}O, H{sub 2}O) is proposed. The method is characterized by efficient breeding of the {sup 233}U isotope and safe reactor operation and is comparatively simple to implement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Natural Transmutation of Actinides via the Fission Reaction in the Closed Thorium–Uranium–Plutonium Fuel Cycle

It is shown for a closed thorium–uranium–plutonium fuel cycle that, upon processing of one metric ton of irradiated fuel after each four-year campaign, the radioactive wastes contain ~54 kg of fission products, ~0.8 kg of thorium, ~0.10 kg of uranium isotopes, ~0.005 kg of plutonium isotopes, ~0.002 kg of neptunium, and “trace” amounts of americium and curium isotopes. This qualitatively simplifies the handling of high-level wastes in nuclear power engineering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

The Potential of Plutonium as a Fuel in Near-Thermal Burner Reactors

Here, the influence of plutonium as a fuel constituent on burner reactor behavior is examined. Both the SM-1 and the Shippingport seed lattices are investigated. It is found that the decrease in the effective absorption cross section of Pu 240 with fuel burnup gives the plutonium the characteristic of a self-compensating fuel, thereby leading to the extension of core life with minimum shim control requirements. The illustrated behavior is similar to that which was previously shown to be possible for converter-type reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

HB-Line Anion Exchange Purification of AFS-2 Plutonium for MOX

Non-radioactive cerium (Ce) and radioactive plutonium (Pu) anion exchange column experiments using scaled HB-Line designs were performed to investigate the feasibility of using either gadolinium nitrate (Gd) or boric acid (B as H 3 BO 3 ) as a neutron poison in the H-Canyon dissolution process. Expected typical concentrations of probable impurities were tested and the removal of these impurities by a decontamination wash was measured. Impurity concentrations are compared to two specifications - designated as Column A or Column B (most restrictive) - proposed for plutonium oxide (PuO 2 ) product shipped to the Mixed Oxide (MOX) Fuel Fabrication Facility (MFFF). Use of Gd as a neutron poison requires a larger volume of wash for the proposed Column A specification. Since boron (B) has a higher proposed specification and is more easily removed by washing, it appears to be the better candidate for use in the H-Canyon dissolution process. Some difficulty was observed in achieving the Column A specification due to the limited effectiveness that the wash step has in removing the residual B after~4 BV's wash. However a combination of the experimental 10 BV's wash results and a calculated DF from the oxalate precipitation process yields an overall DF sufficient to meet the Column A specification. For those impurities (other than B) not removed by 10 BV's of wash, the impurity is either not expected to be present in the feedstock or process, or recommendations have been provided for improvement in the analytical detection/method or validation of calculated results. In summary, boron is recommended as the appropriate neutron poison for H-Canyon dissolution and impurities are expected to meet the Column A specification limits for oxide production in HB-Line.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2012

Technical Basis for DOE Standard 3013 Equivalency Supporting Reduced Temperature Stabilization of Oxalate-Derived Plutonium Oxide Produced by the HB-Line Facility at Savannah River Site

The HB-Line (HBL) facility at the Savannah River Site (SRS) is designed to produce high-purity plutonium dioxide (PuO 2 ) which is suitable for future use in production of Mixed Oxide (MOX) fuel. The MOX Fuel Fabrication Facility (MFFF) requires PuO 2 feed to be packaged per the U.S. Department of Energy (DOE) Standard 3013 (DOE-STD-3013) to comply with the facility's safety basis. The stabilization conditions imposed by DOE-STD-3013 for PuO 2 (i.e., 950 °C for 2 hours) preclude use of the HBL PuO 2 in direct fuel fabrication and reduce the value of the HBL product as MFFF feedstock. Consequently, HBL initiated a technical evaluation to define acceptable operating conditions for production of high-purity PuO 2 that fulfills the DOE-STD-3013 criteria for safe storage. The purpose of this document is to demonstrate that within the defined operating conditions, the HBL process will be equivalent for meeting the requirements of the DOE-STD-3013 stabilization process for plutonium-bearing materials from the DOE complex. The proposed 3013 equivalency reduces the prescribed stabilization temperature for high-purity PuO 2 from oxalate precipitation processes from 950 °C to 640 °C and places a limit of 60% on the relative humidity (RH) at the lowest material temperature. The equivalency is limited to material produced using the HBL established flow sheet, for example, nitric acid anion exchange and Pu(IV) direct strike oxalate precipitation with stabilization at a minimum temperature of 640 °C for four hours (h). The product purity must meet the MFFF acceptance criteria of 23,600 µg/g Pu (i.e., 2.1 wt %) total impurities and chloride content less than 250 µg/g of Pu. All other stabilization and packaging criteria identified by DOE-STD-3013-2012 or earlier revisions of the standard apply. Based on the evaluation of test data discussed in this document, the expert judgment of the authors supports packaging the HBL product under a 3013 equivalency. Under the defined process conditions and associated material specifications, the high-purity PuO 2 produced in HBL presents no unique safety concerns for packaging or storage in the 3013 required configuration. The PuO 2 produced using the HBL flow sheet conditions will have a higher specific surface area (SSA) than PuO 2 stabilized at 950 °C and, consequently, under identical conditions will adsorb more water from the atmosphere. The greatest challenge to HBL operators will be controlling moisture content below 0.5 wt %. However, even at the 0.5 wt % moisture limit, the maximum acceptable pressure of a stoichiometric mixture of hydrogen and oxygen in the 3013 container is greater than the maximum possible pressure for the HBL PuO 2 product.

08 HYDROGEN↗
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 OSTI2016

Closure Report for Corrective Action Unit 412: Clean Slate I Plutonium Dispersion (TTR) Tonopah Test Range, Nevada, Revision 0

This Closure Report (CR) presents information supporting the clean closure of Corrective Action Unit (CAU) 412: Clean Slate I Plutonium Dispersion (TTR), located on the Tonopah Test Range, Nevada. CAU 412 consists of a release of radionuclides to the surrounding soil from a storage–transportation test conducted on May 25, 1963. Corrective action investigation (CAI) activities were performed in April and May 2015, as set forth in the Streamlined Approach for Environmental Restoration (SAFER) Plan for Corrective Action Unit 412: Clean Slate I Plutonium Dispersion (TTR), Tonopah Test Range, Nevada; and in accordance with the Soils Activity Quality Assurance Plan. The purpose of the CAI was to fulfill data needs as defined during the data quality objectives process. The CAU 412 dataset of investigation results was evaluated based on a data quality assessment. This assessment demonstrated the dataset is complete and acceptable for use in fulfilling the data needs identified by the data quality objectives process. This CR provides documentation and justification for the clean closure of CAU 412 under the FFACO without further corrective action. This justification is based on historical knowledge of the site, previous site investigations, implementation of the 1997 interim corrective action, and the results of the CAI. The corrective action of clean closure was confirmed as appropriate for closure of CAU 412 based on achievement of the following closure objectives: Radiological contamination at the site is less than the final action level using the ground troops exposure scenario (i.e., the radiological dose is less than the final action level): Removable alpha contamination is less than the high contamination area criterion: No potential source material is present at the site, and any impacted soil associated with potential source material has been removed so that remaining soil contains contaminants at concentrations less than the final action levels: and There is sufficient information to characterize investigation and remediation waste for disposal.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2012

Treatment of Uranium and Plutonium Solutions Generated in the Atalante Facility, France - 12004

The Atalante complex operated by the French Alternative Energies and Atomic Energy Commission (CEA) at the Rhone Valley Research Center consolidates research programs on actinide chemistry, especially separation chemistry, processing for recycling spent fuel, and fabrication of actinide targets for innovative concepts in future nuclear systems. The design of future systems (Generation IV reactors, material recycling) will increase the uranium and plutonium flows in the facility, making it important to anticipate the stepped-up activity and provide Atalante with equipment dedicated to processing these solutions to obtain a mixed uranium-plutonium oxide that will be stored pending reuse. Ongoing studies for integral recycling of the actinides have highlighted the need for reserving equipment to produce actinides mixed oxide powder and also minor actinides bearing oxide for R and D purpose. To meet this double objective a new shielded line should be built in the facility and should be operational 6 years after go decision. The main functions of the new unit would be to receive, concentrate and store solutions, purify them, ensure group conversion of actinides and conversion of excess uranium. This new unit will be constructed in a completely refurbished building devoted to subcritical and safe geometry of the process equipments. (author)

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

Advanced Waste Management Concepts for Uranium and Plutonium Multi-Recycling - 18533

The multiple recycling of plutonium and uranium will play a major role in gradually achieving a fully closed nuclear fuel cycle in the future, once the first Generation IV fast reactors (FR) are deployed in the NPP fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, the development of innovative concepts are required to meet the scientific, technical and economic challenges imposed by the treatment of spent fuels with an increasingly higher Pu content. Based on the significant past experience acquired by France in LWR and FR MOX fuel treatment, the CEA is running R and D studies to develop new efficient processes for Pu recycling and for FR MOX fuel fabrication. More compact and economically competitive processes, ensuring the safety/security of the facilities (particularly as regards criticality and proliferation resistance) have to be developed taking into account that the amounts of plutonium involved will be one order of magnitude higher than in UOX fuels. New concepts of head-end treatment, dissolution, and separation have been proposed. Consequently, effluent and waste management must be adapted to these processes and to the characteristics of fast reactor MOX fuels. New waste treatment processes are suggested: 1) Cladding waste and undissolved metallic particles could be treated simultaneously in a melting process. The molten metal could be decontaminated by a specific slag and the oxides recovered, in particular PuO{sub 2}, would be recycled at the frond-end of the process, 2) Fission Products and Actinides (FPA) containment glass will also be optimized to accommodate significant amounts of iron, nickel and chromium coming from hull corrosion, and / or decontamination slag coming from the metal melting process. 3) New peraluminous-type glass formulations are under development to accept increased amounts of FPA, as well as iron, nickel, and chromium. The first trials to evaluate the melting process have indicated that metallic compounds from the dissolution residues could be easily incorporated into a molten metal at levels of 1, 2 and 3 w%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Dissolution of aerosol particles collected from nuclear facility plutonium production process

Here, a simple, robust analytical chemistry method has been developed to dissolve plutonium containing particles in a complex matrix. The aerosol particles collected on Marple cascade impactor substrates were shown to be dissolved completely with an acid mixture of 12 M HNO 3 and 0.1 M HF. A pressurized closed vessel acid digestion technique was utilized to heat the samples at 130 °C for 16 h to facilitate the digestion. The dissolution efficiency for plutonium particles was 99 %. The resulting particle digestate solution was suitable for trace elemental analysis and isotope composition determination, as well as radiochemistry measurements.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2015

Evaluating bis(2-ethylhexyl) methanediphosphonic acid (H 2 DEH[MDP]) based polymer ligand film (PLF) for plutonium and uranium extraction

This paper describes a new analyte extraction medium called polymer ligand film (PLF) that was developed to rapidly extract radionuclides. PLF is a polymer medium with ligands incorporated in its matrix that selectively and quickly extracts analytes. The main focus of the new technique is to shorten and simplify the procedure for chemically isolating radionuclides for determination through alpha spectroscopy. The PLF system was effective for plutonium and uranium extraction. The PLF was capable of co-extracting or selectively extracting plutonium over uranium depending on the PLF composition. As a result, the PLF and electrodeposited samples had similar alpha spectra resolutions.

36 MATERIALS SCIENCE↗
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 OSTI2021

Evidence of an oxidation induced phase transformation for a delta phase plutonium-gallium alloy

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.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Development of small-scale plutonium electrorefining in molten CaCl 2

Molten salt extraction and electrorefining of plutonium were successfully conducted with only a 200-g initial charge. Due to this small scale, the electrorefining cell components were redesigned relative to conventional, multi-kg standards. The efficiency and yield of this small-scale process are comparable to those of larger-scale exemplars. Chemical analysis of the plutonium product and the anode heel is presented. Here, the product was 99.8% pure. The anode heel was able to be depleted beyond the expected freezing point without significant impurity carryover.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Determination of total plutonium content in spent nuclear fuel assemblies with the differential die-away self-interrogation instrument

As a part of the Next Generation Safeguards Initiative Spent Fuel project, we simulate the response of the Differential Die-away Self-Interrogation (DDSI) instrument to determine total elemental plutonium content in an assayed spent nuclear fuel assembly (SFA). We apply recently developed concepts that relate total plutonium mass with SFA multiplication and passive neutron count rate. In this work, the multiplication of the SFA is determined from the die-away time in the early time domain of the Rossi-Alpha distributions measured directly by the DDSI instrument. Here, we utilize MCNP to test the method against 44 pressurized water reactor SFAs from a simulated spent fuel library with a wide dynamic range of characteristic parameters such as initial enrichment, burnup, and cooling time. Under ideal conditions, discounting possible errors of a real world measurement, a root mean square agreement between true and determined total Pu mass of 2.1% is achieved.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Superconductivity in plutonium compounds

Although the family of plutonium-based superconductors is relatively small, consisting of four compounds all of which crystallize in the tetragonal HoCoGa 5 structure, these materials serve as an important bridge between the known Ce- and U-based heavy fermion superconductors and the high-temperature cuprate superconductors. Additionally, the partial localization of 5f electrons that characterizes the novel electronic properties of elemental plutonium appears to be central to the relatively high superconducting transition temperatures that are observed in PuCoGa 5 , PuRhGa 5 , PuCoIn 5 , and PuRhIn 5 .

36 MATERIALS SCIENCE↗