Development of a dispersion strengthened nickel base alloy using the high intensity arc process Summary report
Dispersion strengthened nickel base alloy using high intensity arc process
NICKEL ALLOY↗
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Containerless high temperature property measurements
Containerless processing in the low gravity environment of space provides the opportunity to increase the temperature at which well controlled processing of and property measurements on materials is possible. This project was directed towards advancing containerless processing and property measurement techniques for application to materials research at high temperatures in space. Containerless high temperature material property studies include measurements of the vapor pressure, melting temperature, optical properties, and spectral emissivities of solid boron. The reaction of boron with nitrogen was also studied by laser polarimetric measurement of boron nitride film growth. The optical properties and spectral emissivities were measured for solid and liquid silicon, niobium, and zirconium; liquid aluminum and titanium; and liquid Ti-Al alloys of 5 to 60 atomic pct. titanium. Alternative means for noncontact temperature measurement in the absence of material emissivity data were evaluated. Also, the application of laser induced fluorescence for component activity measurements in electromagnetic levitated liquids was studied, along with the feasibility of a hybrid aerodynamic electromagnetic levitation technique.
Nordine, Paul C.↗
Compatibility of refractory materials for nuclear reactor poison control systems
Metal-clad poison rods have been considered for the control system of an advanced space power reactor concept studied at the NASA Lewis Research Center. Such control rods may be required to operate at temperatures of about 140O C. Selected poison materials (including boron carbide and the diborides of zirconium, hafnium, and tantalum) were subjected to 1000-hour screening tests in contact with candidate refractory metal cladding materials (including tungsten and alloys of tantalum, niobium, and molybdenum) to assess the compatibility of these materials combinations at the temperatures of interest. Zirconium and hafnium diborides were compatible with refractory metals at 1400 C, but boron carbide and tantalum diboride reacted with the refractory metals at this temperature. Zirconium diboride also showed promise as a reaction barrier between boron carbide and tungsten.
Sinclair, J. H.↗
Advanced cast aluminum alloys for automotive engine application with superior high-temperature properties
A high fatigue strength aluminum alloy comprises in weight percent copper 3.0-3.5%, iron 0-1.3%, magnesium 0.24-0.35%, manganese 0-0.8%, silicon 6.5-12.0%, strontium 0-0.025%, titanium 0.05-0.2%, vanadium 0.20-0.35%, zinc 0-3.0%, zirconium 0.2-0.4%, a maximum of 0.5% other elements and balance aluminum plus impurities. The alloy defines a microstructure having an aluminum matrix with the Zr and the V in solid solution after solidification. The matrix has solid solution Zr of at least 0.16% after heat treatment and solid solution V of at least 0.20% after heat treatment, and both Cu and Mg are dissolved into the aluminum matrix during the heat treatment and subsequently precipitated during the heat treatment. A process for heat treating an Al—Si—Cu—Mg—Fe—Zn—Mn—Sr-TMs alloy comprises heat treating the alloy to produce a microstructure having a matrix with Zr and V in solid solution after solidification.
36 MATERIALS SCIENCE↗
Castable hot corrosion resistant alloy
Some 10 wt percent nickel is added to an Fe-base alloy which has a ferrite microstructure to improve the high temperature castability and crack resistance while about 0.2 wt percent zirconium is added for improved high temperatur cyclic oxidation and corrosion resistance. The basic material is a high temperature FeCrAl heater alloy, and the addition provides a material suitable for burner rig nozzles.
Charles A. Barrett↗
Aluminum-Silicon Alloy Having Improved Properties At Elevated Temperatures and Process for Producing Cast Articles Therefrom
A process for making a cast article from an aluminum alloy includes first casting an article from an alloy having the following composition, in weight percent: Silicon 11.0-14.0, Copper 5.6-8.0, Iron 0-0.8, Magnesium 0.5-1.5, Nickel 0.05-0.9, Manganese 0-1.0, Titanium 0.05-1.2, Zirconium 0.12-1.2, Vanadium 0.05-1.2, Zinc 0.05-0.9, Strontium 0.001-0.1, Aluminum balance . In this alloy the ratio of silicon to magnesium is 10 to 25, and the ratio of copper to magnesium is 4 to 15. After an article is cast from the alloy, the cast article is aged at a temperature within the range of 400F to 500F for a time period within the range of four to 16 hours. It has been found especially advantageous if the cast article is first exposed to a solutionizing step prior to the aging step. This solutionizing step is carried out by exposing the cast article to a temperature within the range of 900F to 1000F for a time period of fifteen minutes to four hours. It has also been found to be especially advantageous if the solutionizing step is followed directly with a quenching step, wherein the cast article is quenched in a quenching medium such as water at a temperature within the range of 120F to 300F. The resulting cast article is suitable in a number of high temperature applications, such as heavy-duty pistons for internal combustion engines.
Lee, Jonathan A.↗
Fluid-bed fluoride volatility process recovers uranium from spent uranium alloy fuels
Fluid-bed fluoride volatility process recovers uranium from uranium fuels containing either zirconium or aluminum. The uranium is recovered as uranium hexafluoride. The process requires few operations in simple, compact equipment, and eliminates aqueous radioactive wastes.
Barghusen, J. J.↗
Process for making a high toughness-high strength iron alloy
A steel alloy is produced by a process which includes using cold rolling at room temperature and subsequent heat treatment at temperatures ranging from 500 C to 650 C. The resulting alloys exhibits excellent strength and toughness characteristics at cryogenic temperatures. This alloy consists essentially of about 10 to 16 percent by weight nickel, to about 1.0 percent by weight aluminum, and 0 to about 3 percent by weight of at least one of the following additional elements: copper, lanthanum, niobium, tantalum, titanium, vanadium, yttrium, zirconium and the rare earth metals, with the balance being essentially iron. The improved alloy possesses a fracture toughness ranging from 200 to 230 ksi sq in. and yield strengths up to 230 ksi.
Stephens, J. R.↗
Alternate Methods for Cleaning Zirconium Plate
The United States High Performance Research Reactor Conversion Program identified the need to transition from highly enriched uranium fuel to lowly enriched uranium fuel for high power research and civil reactors. One of the fuel configurations selected for these reactors was a uranium (U) foil alloyed with 10% molybdenum (Mo). The U-10Mo fuel foil was then covered with a zirconium (Zr) interlayer and pressed into an aluminum cladding. Idaho National Laboratory prescribed a need to remove surface oxidation on the Zr prior to bonding it with U-10Mo fuel foil. To determine the best method of oxidation removal, a variety of polishing methods, including hand polishing with a diamond paste or Scotch-Brite™ pads, mechanically wet-polishing with abrasive belts, and chemical etching with hydrofluoric/nitric acid mixtures, were investigated on pre-polished Zircaloy 702 sheets purchased directly from the manufacturer. No change in surface roughness was detected using any of these methods relative to the as-received material. In fact, scanning electron microscopy detected grit from hand polishing with Amplex Grade 30 waste soluble diamond paste and Scotch-Brite™ pads. Visible staining was found, and fluorine detected via X-ray photoelectron spectroscopy on the flash etching samples. These contaminants were not seen on the as-received material. In addition to the need to remove the oxide layer on the Zr prior to co-rolling, there are concerns that the Neolube debonding agent on the co-rolling can may contaminate the Zr surface after co-rolling. The experimental results found that zirconium in contact with Neolube coated steel plates was easily cleaning with ethanol wiping after heat treated and no carbide inclusions remained on the surface. Overall, all cleaning methods tested did not significantly change the surface roughness of the plates or reduce the oxide layer present on the zirconium. Carbon was found on all the samples, before and after cleaning, except for the flash etch samples which detected fluorine instead. The insignificant changes seen in the cleaned plates relative to the as received plates question the need for cleaning prior to rolling. Further investigation in the requirements for cleaning at this step is recommended. Current methods for cleaning the surface after hot rolling with alcohol appear sufficient. After hot rolling, if Neolube does peel off the can and onto the zirconium coated foil surface, a dry abrasive or diamond paste polish is not recommended due to the chance of abrasive being embedded into the surface. However, the ability of an etchant to lift the lubricant off the zirconium surface indicates that a spot etch method could work to clean specific spots where Neolube is present, if the need arises.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Development of Refractory Alloys and Refractory Coatings for Advanced Nuclear Reactors
The next generation of nuclear reactors will benefit from materials that enable operation at higher temperatures (>500°C), higher irradiation doses (up to 200 displacements per atom (dpa)), and the use of more corrosive and reactive coolants. This work package represents the first experimental steps towards a longer-term effort to develop refractory materials for nuclear energy applications which will enable operation under these conditions. Specifically, this work package focuses on additive manufacturing of refractories as both a refractory liner coating deposited onto the interior surface of a metallic tubular backbone and as bulk refractory alloys. During fiscal year 23 (FY23), several refractory metal coating systems and bulk alloys were examined and selected using a decision criteria matrix. The refractory metal coating systems included molybdenum, tungsten, and zirconium as refractory coatings on backbones of either carbon-carbon (C/C) or silicon carbide-silicon carbide (SiC/SiC) ceramic matrix composites. The bulk refractory alloys included C-103, WTa, and WNiFe as bulk alloys. During FY24 additional bulk refractory alloys and metallic backbones were evaluated using the decision criteria matrix based on input from the AMMT leadership team. These included 316 SS and 316H SS for the metallic backbones and Mo-La, Ta, and Nb1Zr as bulk refractory alloys. The primary focus for the FY24 effort was placed on establishing the capabilities to deposit refractory coatings based on the results of the scoring in the decision criteria matrix and finalizing the additively manufactured TZM studies which were incorporated into the AMMT program from the microreactor program. Further efforts were dedicated to establishing the capabilities to additively manufacture down-selected bulk refractory alloys.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
A critical review of the history of fabricating monolithic U-Mo fuel plates
The fabrication of monolithic uranium-molybdenum alloy fuels, specifically those developed for high-performance research and test reactors, began in the early 2000s. The primary fuel form consists of uranium alloyed with 10-wt% molybdenum in a thin foil coated with zirconium and encapsulated in aluminum-6061 cladding. Over the years, the process has evolved with different types of casting, heat treatments, rolling schedules, and cladding applications. This review examines the history of these fabrication processes and their impact on microstructure and fuel-swelling performance. Even though various fabrication methods were used, we found little correlation between fabrication variation and fuel swelling. This insensitive relationship between fabrication variation and fuel swelling is primarily due to inhomogeneous microstructures that formed during casting and grain refinement that occurred during rolling. We conclude that the fabrication processes we examined produced similar microstructures, indicating that the fuel microstructure is somewhat insensitive to the fabrication parameters evaluated. However, the relatively small amount of historic data, such as those for grain sizes, limited this analysis. More recently fabricated materials, such as those from ongoing irradiation experiment, Mini-Plate-1 and Mini-Plate-2, were also excluded from this analysis and are intended to be reviewed separately. The findings, that fuel microstructure is somewhat insensitive to the fabrication parameters, do not imply that any fabrication method is acceptable, given the uncertainties in data and fuel-swelling observations. For example, only arc melting and vacuum induction melting casting processes were previously explored in the historic fabrication efforts. In conclusion, the findings should not be extrapolated to other casting processes.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Critical Review on the History of Fabricating Monolithic U-Mo Fuel Plates
Fabrication of uranium-molybdenum alloy fuels has been occurring since the early 2000s in support of the development of a high-uranium-density low-enrichment fuel for use in high-performance research and test reactors which operate at relatively low temperatures. The primary fuel form—a thin foil of uranium, alloyed with 10wt% molybdenum, which is coated in a layer of zirconium and then encapsulated with aluminum 6061 as a cladding material—was developed over a number of years.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Assessment of Cladding-Cladding Delamination Failures in U-Mo Fuel Plates
Monolithic fuel is a candidate fuel form for the conversion of high-performance research reactors. This plate-type fuel consists of a high-density, U-Mo alloy-based fuel in a monolithic form that is sandwiched between zirconium diffusion barriers and encapsulated in an aluminum cladding. To qualify this fuel system, the program must demonstrate that the fuel meets the safety standards, has a predictable behavior, and performs well in a reactor. Although many plates have shown satisfactory thermo-mechanical performance in reactor, a small number of experimental plates exhibited failures, rendering performance of these plates unacceptable. Post-irradiation examinations of failed plates have revealed that various failures modes are present, including pillowing, blistering, buckling, warping, delamination, fracture, rupture, oxide spallation, etc. Previous studies have implied that such various operational and loading conditions could generate a complex stress-strain state, facilitating distinct failure modes that could impact the structural integrity. To understand such various failure modes, a comprehensive set of studies focusing on failure modes is being performed. As a part of these studies, this work studies the cladding-cladding separation failures that were observed in some of previous irradiation tests. To understand the root cause of cladding separation failures, selected plates from previous irradiation experiments were simulated under comparable operational conditions. Stress-strain state of the plates at the interface were examined. The results were comparatively evaluated to understand possible effects of stress state of the bond-line on delamination failures. This paper presents these findings, discusses such failure modes, and examines the influence of various factors on previously observed cladding delamination failures.
42 ENGINEERING↗
Quantify Contaminant Partitioning in Advanced Chlorination Process
Chlorination of the zirconium in spent fuel cladding has been identified as a significant means of Zr recovery and purification, enabling a 25% (wt.) or greater reduction in waste requiring geologic disposition. Several chlorinating agents have been identified over the last few years, including gases and liquids. In particular, the interactions of sulfur-chloride solvents with zirconium may allow the separation of zirconium from uranium with the selective formation of ZrCl 4 . This report reviews what is known about reactions between chlorinating agents, uranium metal and oxide, Zircaloy and its alloying elements, and fission plus activation products. Plans for testing the feasibility of applying advanced chlorination methods developed for zirconium recycling to the front end of the recycling of spent nuclear fuel are also included.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Aluminum-Silicon Alloy Having Improved Properties at Elevated Temperatures and Articles Cast Therefrom
An aluminum alloy suitable for high temperature applications, such as heavy duty pistons and other internal combustion applications. having the following composition, by weight percent (wt %): Silicon: 11.0-14.0; Copper: 5.6-8.0; Iron: 0-0.8; Magnesium: 0.5-1.5; Nickel: 0.05-0.9; Manganese: 0.5-1.5; Titanium: 0.05-1.2; Zirconium: 0.12-1.2; Vanadium: 0.05-1.2; Zinc: 0.005-0.9; Strontium: 0.001-0.1; Aluminum: balance. In this alloy the ratio of silicon:magnesium is 10-25, and the ratio of copper:magnesium is 4-15. After an article is cast from this alloy, the article is treated in a solutionizing step which dissolves unwanted precipitates and reduces any segregation present in the original alloy. After this solutionizing step, the article is quenched, and is then aged at an elevated temperature for maximum strength.
Lee, Jonathan A.↗
Silicon carbide reinforced zirconium based cladding
A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding
Here, structure of the pellet/cladding-interaction (PCI) layer of a high-burnup nuclear fuel with niobium-bearing cladding was investigated using modern transmission electron microscopy (TEM) techniques. Morphology of the PCI consisted of uniform zirconia layer on the cladding side and finger-like features towards to the ceramic fuel. The PCI layer showed a complex microstructure that constitutes three distinct zirconia layers formed by two zirconia phases. From cladding to fuel, monoclinic, tetragonal, and another tetragonal phase were determined via TEM diffraction patterns and precession electron diffraction (PED) using ASTAR microscopy platform (AMP). Characterizations showed that monoclinic and tetragonal phases close to the cladding were crack-free, but the tetragonal layer adjacent to the fuel contained cracks and pores. Presence of the monoclinic layer suggested that its formation involved the gaseous diffusion of oxygen below threshold dose of fission product damage before the fuel/cladding contact occurred. Furthermore, high-resolution TEM revealed compound phase of (U,Zr)O 2 at fringes of zirconia fingers on the fuel side.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Impact Toughness and Heat Treatment for Cast Aluminum
A method for transforming a cast component made of modified aluminum alloy by increasing the impact toughness coefficient using minimal heat and energy. The aluminum alloy is modified to contain 0.55%-0.60% magnesium, 0.10%-0.15% titanium or zirconium, less than 0.07% iron, a silicon-tomagnesium product ratio of 4.0, and less than 0.15% total impurities. The shortened heat treatment requires an initial heating at 1,000deg F. for up to I hour followed by a water quench and a second heating at 350deg F. to 390deg F. for up to I hour. An optional short bake paint cycle or powder coating process further increase.
Lee, Jonathan A↗