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Dombrowski, David E.

Publications and source records attributed to Dombrowski, David E..

16 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Texture Evolution in U-10Mo Nuclear Fuel Foils during Plasma Spray Coating with Zr

A uranium-molybdenum alloy clad in 6061 aluminum has the potential to lead to a wide application of low-enriched uranium fuels, replacing highly enriched uranium for research reactors. A Zr coating acts as a diffusion barrier between the fuel and the aluminum cladding. In this study, U-10Mo (mass %) was coated with Zr using a plasma spray technique recognized as a fast and economical coating method. Neutron time-of-flight diffraction was used to study the microstructure evolution by quantifying the phase fractions of involved phases as well as the texture evolution of U-10Mo and Zr during plasma spray coating with Zr. Quantitative texture analysis revealed that the texture was drastically changed for high coating temperatures, likely due to selective grain growth. Furthermore, the Zr coating showed a preferential orientation, which could be correlated with the initial texture of the uncoated U-10Mo. This could be explained by the epitaxial growth of the Zr on the U-10Mo substrate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Understanding Manufacturing Scale-Up of Plasma Sprayed Zr Barrier Coatings on U-10Mo Fuels

The United States government is committed to further strengthening nuclear security and nonproliferation in order to reduce the threat of terrorists acquiring nuclear material. To meet this important mission, the US Department of Energy/National Nuclear Security Agency's Office of Material Management and Minimization (M{sup 3}) Reactor Conversion Program minimizes the use of highly enriched uranium (HEU) in civilian applications by working with governments and facilities around the world to convert research reactors to the use of non-weapons-usable low-enriched uranium (LEU) fuels. Each reactor converted not only eliminates the need for HEU at civilian sites, but also reduces the amount of HEU being manufactured, stored, and transported, where it is at its most vulnerable. In instances where suitable LEU fuels do not exist for particular reactors to convert, the M{sup 3} Reactor Conversion Program (CONVERT) contributes to the development of new LEU fuels. Currently, the M{sup 3} Reactor Conversion Program is committed to developing a high-density U-Mo LEU fuel that will be able to convert high performance research reactors while establishing an efficient and economic fabrication capability to make the fuel. To that end, Los Alamos National Laboratory (LANL) has been tasked with developing a method to affix a Zirconium fuel diffusion barrier via plasma spraying that will prevent a reaction between the U-Mo fuel and the aluminum cladding. This report discusses the transitions from development of the plasma spray deposition process to a scientific understanding of pseudo-production conditions and quantification of coating layer properties, characterization, reproducibility, etc. Understanding quality assurance and its applications to experiment work is non-trivial; statistical analysis of various measurement techniques not only improve quality, but will allow for scale up of manufacturing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Feasibility Study for Electrical Discharge Machining of Large DU-Mo Castings

U-10 wt. % Mo (U-10Mo) alloys are being developed as low enrichment monolithic fuel for the CONVERT program. Optimization of processing for the monolithic fuel is being pursued with the use of electrical discharge machining (EDM) under CONVERT HPRR WBS 1.2.4.5 Optimization of Coupon Preparation. The process is applicable to manufacturing experimental fuel plate specimens for the Mini-Plate-1 (MP-1) irradiation campaign. The benefits of EDM are reduced machining costs, ability to achieve higher tolerances, stress-free, burr-free surfaces eliminating the need for milling, and the ability to machine complex shapes. Kerf losses are much smaller with EDM (tenths of mm) compared to conventional machining (mm). Reliable repeatability is achievable with EDM due to its computer-generated machining programs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Non-destructive Quantitative Phase Analysis and Microstructural Characterization of Zirconium Coated U-10Mo Fuel Foils via Neutron Diffraction

This report uses neutron diffraction to investigate the crystal phase composition of uranium-molybdenum alloy foils (U-10Mo) for the CONVERT MP-1 Reactor Conversion Project, and determines the effect on alpha-uranium contamination following the deposition of a Zr metal diffusion layer by various methods: plasma spray deposition of Zr powders at LANL and hot co-rolling with Zr foils at BWXT. In summary, there is minimal decomposition of the gamma phase U-10Mo foil to alpha phase contamination following both plasma spraying and hot co-rolling. The average unit cell volume, i.e. lattice spacing, of the Zr layer can be mathematically extracted from the diffraction data; co-rolled Zr matches well with literature values of bulk Zr, while plasma sprayed Zr shows a slight increase in the lattice spacing, indicative of interstitial oxygen in the lattice. Neutron diffraction is a beneficial alternative to conventional methods of phase composition, i.e. x ray diffraction (XRD) and destructive metallography. XRD has minimal penetration depth in high atomic number materials, particularly uranium, and can only probe the first few microns of the fuel plate; neutrons pass completely through the foil, allowing for bulk analysis of the foil composition and no issues with addition of cladding layers, as in the final, aluminum-clad reactor fuel plates. Destructive metallography requires skilled technicians, cutting of the foil into small sections, hazardous etching conditions, long polishing and microscopy times, etc.; the neutron diffraction system has an automated sample loader and can fit larger foils, so there is minimal analysis preparation; the total spectrum acquisition time is ~ 1 hour per sample. The neutron diffraction results are limited by spectra refinement/calculation times and the availability of the neutron beam source. In the case of LANSCE at Los Alamos, the beam operates ~50% of the year. Following the lessons learned from these preliminary results, optimizations to the process and analysis can be made, and neutron diffraction can become a viable and efficient technique for gamma/alpha phase composition determination for nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Processing and consolidation of copper/tungsten

Here, we developed a copper/tungsten (Cu/W) composite for mesoscale Materials Science applications using the novel High-Energy Diffraction Microscopy (HEDM) technique. Argon-atomized copper powder was selected as the starting raw powder and screened to remove the extremely large particle fraction. Tungsten particles were collected by milling and screening the -325 mesh tungsten powder between 500 and 635 mesh sieves. Hot pressing of screened Cu powder was performed at 900 °C in Ar/4 %H 2 atmosphere. XRD and ICP results show that the hot-pressed Cu sample consists of about 5 vol% Cu 2 O, which is caused by the presence of oxygen on the surface of the starting Cu powder. Hot pressing the copper powder in a pure hydrogen atmosphere was successful in removing most of the surface oxygen. Our process was also implemented for hot pressing the Cu/W composite. The density of the Cu/W composites hot pressed at 950 °C in pure hydrogen was about 94 % of the theoretical density (TD). The hot-pressed Cu/W composites were further hot isostatic pressed at 1050 °C in argon atmosphere, which results in 99.6 % of the TD with the designed Cu grain size and W particle distribution. Tensile specimens with D-notch were machined using the wire EDM method. Furthermore, the processing and consolidation of these materials will be discussed in detail. The HEDM images are also showed and discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Bond Strength of Plasma Sprayed Zr Diffusion Barrier on U-Mo Fuels

The United States government is continually committed to strengthening nuclear security and nonproliferation for reducing the threat of radioactive materials acquisition by terrorists. To meet this important mission, the US Department of Energy and National Nuclear Security Agency's (NNSA) Office of Material Management and Minimization (M{sup 3}) Reactor Conversion Program (CONVERT) seeks to minimize the use of highly enriched uranium (HEU) in civilian applications through work with governments and facilities around the world to convert research reactors to the use of non-weapons-usable low enriched uranium (LEU) fuels. Each reactor converted not only eliminates the need for HEU at civilian sites, but also reduces the amount of HEU being manufactured, stored, and transported, where the nuclear material is most vulnerable. In instances where suitable LEU fuels do not exist for the conversion of particular reactors, CONVERT contributes to development of new LEU fuels. Currently, the CONVERT Program is committed to development of a high-density U-Mo fuel that will be able to convert high performance research reactors, while establishing an efficient and economic fabrication capability to manufacture the fuel. To that end, Los Alamos National Laboratory (LANL) has been tasked with development of a method to affix zirconium (Zr) metal fuel diffusion barrier via plasma spraying, which will prevent a reaction between the U-Mo fuel and the 6061 aluminum cladding. A successful diffusion barrier requires adequate Zr layer thickness, uniform coverage, sufficient bonding at the Zr/fuel and Zr/aluminum cladding interface, and maintenance of the original nuclear fuel properties i.e. physical dimensions, alloy composition and dispersion, crystallographic phase concentration, neutronics, etc. This work outlines development of the Zr plasma spray processing techniques, parameters affecting the diffusion barrier/nuclear fuel bond strength, and characterization of the Zr diffusion barrier. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Fuel Fabrication Capability WBS 01.02.01.05 - HIP Bonding Experiments (Final Report)

The goals of this project were to demonstrate reliable, reproducible solid state bonding of aluminum 6061 alloy plates together to encapsulate DU-10 wt% Mo surrogate fuel foils. This was done as part of the CONVERT Fuel Fabrication Capability effort in Process Baseline Development . Bonding was done using Hot Isotatic Pressing (HIP) of evacuated stainless steel cans (a.k.a HIP cans) containing fuel plate components and strongbacks. Gross macroscopic measurements of HIP cans prior to HIP and after HIP were used as part of this demonstration, and were used to determine the accuracy of a finitie element model of the HIP bonding process. The quality of the bonding was measured by controlled miniature bulge testing for Al-Al, Al-Zr, and Zr-DU bonds. A special objective was to determine if the HIP process consistently produces good quality bonding and to determine the best characterization techniques for technology transfer.

36 MATERIALS SCIENCE↗
DOE OSTI2015

LANL Experience Rolling Zr-Clad LEU-10Mo Foils for AFIP-7

The cleaning, canning, rolling and final trimming of Low Enriched Uranium-10 wt. pct. Molybdenum (LEU-10Mo) foils for ATR (Advanced Test Reactor) fuel plates to be used in the AFIP-7 (ATR Full Size Plate In Center Flux Trap Position) experiments are summarized. Six Zr-clad foils were produced from two LEU-10Mo castings supplied to Los Alamos National Laboratory (LANL) by Y-12 National Security Complex. Details of cleaning and canning procedures are provided. Hot- and cold-rolling results are presented, including rolling schedules, images of foils in-process, metallography and local compositions of regions of interest, and details of final foil dimensions and process yield. This report was compiled from the slides for the presentation of the same name given by Duncan Hammon on May 12, 2011 at the AFIP-7 Lessons Learned meeting in Salt Lake City, UT, with Los Alamos National Laboratory document number LA-UR 11-02898.

Materials Science(36)↗
DOE OSTI2014

Evaluation of LANL Capabilities for Fabrication of TREAT Conversion Fuel

This report estimates costs and schedule associated with scale up and fabrication of a low-enriched uranium (LEU) core for the Transient Reactor Test Facility (TREAT) reactor. This study considers facilities available at Los Alamos National Laboratory, facility upgrades, equipment, installation and staffing costs. Not included are costs associated with raw materials and off-site shipping. These estimates are considered a rough of magnitude. At this time, no specifications for the LEU core have been made and the final schedule needed by the national program. The estimate range (+/-100%) reflects this large uncertainty and is subject to change as the project scope becomes more defined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Update on Fabrication of Extrusions for TREAT Trade Study

This supplemental report describes fuel fabrication efforts conducted for the Idaho National Laboratory Trade Study for the TREAT Conversion project that is exploring the replacement of the HEU (Highly Enriched Uranium) fuel core of the TREAT reactor with LEU (Low Enriched Uranium) fuel. Previous reports have documented fabrication of fuel by the “upgrade” process developed at Los Alamos National Laboratory. These experiments supplement an earlier report that describes efforts to increase the graphite content of extruded fuel and minimize cracking.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Fabrication of TREAT Fuel with Increased Graphite Loading

As part of the feasibility study exploring the replacement of the HEU fuel core of the TREAT reactor at Idaho National Laboratory with LEU fuel, this study demonstrates that it is possible to increase the graphite content of extruded fuel by reformulation. The extrusion process was use to fabricate the “upgrade” core1 for the TREAT reactor. The graphite content achieved is determined by calculation and has not been measured by any analytical method. In conjunction, a technique, Raman Spectroscopy, has been investigated for measuring the graphite content. This method shows some promise in differentiating between carbon and graphite; however, standards that would allow the technique to be calibrated to quantify the graphite concentration have yet to be fabricated. Continued research into Raman Spectroscopy is on going. As part of this study, cracking of graphite extrusions due to volatile evolution during heat treatment has been largely eliminated. Continued research to optimize this extrusion method is required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Hot Isostatic Press Can Optimization for Aluminum Cladding of U-10Mo Reactor Fuel Plates: FY12 Final Report and FY13 Update

Currently, the proposed processing path for low enriched uranium – 10 wt. pct. molybdenum alloy (LEU-10Mo) monolithic fuel plates for high power research and test reactors includes hot isostatic pressing (HIP) to bond the aluminum cladding that encapsulates the fuel foil. Initial HIP experiments were performed at Idaho National Laboratory (INL) on approximately ¼ scale “mini” fuel plate samples using a HIP can design intended for these smaller experimental trials. These experiments showed that, with the addition of a co-rolled zirconium diffusion barrier on the LEU-10Mo alloy fuel foil, the HIP bonding process is a viable method for producing monolithic fuel plates. Further experimental trials at Los Alamos National Laboratory (LANL) effectively scaled-up the “mini” can design to produce full-size fuel prototypic plates. This report summarizes current efforts at LANL to produce a HIP can design that is further optimized for higher volume production runs. The production-optimized HIP can design goals were determined by LANL and Babcock & Wilcox (B&W) to include maintaining or improving the quality of the fuel plates produced with the baseline scaled-up mini can design, while minimizing material usage, improving dimensional stability, easing assembly and disassembly, eliminating machining, and significantly reducing welding. The initial small-scale experiments described in this report show that a formed-can approach can achieve the goals described above. Future work includes scaling the formed-can approach to full-size fuel plates, and current progress toward this goal is also summarized here.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗