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Alloys of Uranium with Zirconium, Chromium, Columbium, Vanadium, and Molybdenum
Abstract: Information on five uranium binary alloy systems has been surveyed. These systems are the alloys of uranium with zirconium, chromium, columbium, vanadium, and molybdenum. The equilibrium diagrams are discussed, and where available, data have been included on diffusion studies, cladding experiments, and mechanical properties.
Aluminum Cladding Long Uranium Plates by Solid-State Bonding
From introduction: This report covers an investigation of solid-state bonding as a technique for aluminum cladding uranium plates of 3 by 0.180-in. cross section in lengths up to 14 ft.
The Ammonium Carbonate Pressure Leaching of Uranium Ores Proposed as Feed to the Pilot Plant at Grand Junction, Colorado : Progress Report
From introduction: This is Progress report BMI-282, the first of a series covering the operation of the pressure-leaching towers for the extraction of uranium by an ammonium carbonate leach...This report discusses the data obtained from tests run to show the effects of such variables as the amount of air and carbon dioxide passed through the pulp, temperature, pressure, and concentration of ammonium carbonate and ammonium bicarbonate in the leach solution on the extraction of uranium.
Annealing of crystal distortion in irradiated graphite
From introduction: "As part of the program for improvement of graphite, the structural characteristics of graphite have been studied to determine the relation between physical and structural changes during irradiation."
Apparatus for visual study of corrosion by hot water
Report describing an apparatus for visual study of corrosion of uranium by hot water. A windowed autoclave was designed and constructed to permit visual study of the reaction under conditions simulating a cladding failure in service.
The Application of Ceramics to Hanford Fuel Elements
From introduction: This work is a survey of possible uses of ceramic materials in Hanford-type fuel elements.
An Application of the Concepts of Particle Packing to the Consolication of Silicon Carbide Powders
From introduction: Silicon carbide is being considered as a basis material for nonmetallic fuel elements because of its high thermal conductivity, low nuclear cross section, high resistance to thermal rupture, and high degree of stability at high temperature in air. A requirement of the fuel elements is that they be thin and have as low porosity as possible. One shape of element under consideration is 0.050 to 0.070 inch thick by a few inches in width and breadth.
Aqueous Corrosion of Uranium Fuel-Element Cores Containing 0 to 20 w/o Zirconium
Abstract: A description is given of the design and operation of a windowed autoclave system employed in the study of corrosion by pressurized hot water. The device has been used to obtain time-lapse motion pictures of the swelling and rupture of deliberately defected zirconium-clad uranium specimens. A method is described by which corrosion rates were calculated from pressure and temperature measurements. A typical set of pictures taken during a test is presented, and corrosion rates are reported for uranium-0, 5, 10, 15, and 20 w/o zirconium alloys subjected subjected to 600 F water.
Beryllia and Beryllia-Beryllium Bodies for Moderating Cores in Fuel Elements
From introduction: This report presents the results of the two phases of the work: study of the compatibility of beryllia bodies with a coating material, molybdenum; and studies of beryllia-beryllium body fabrication.
Beta-Counting Methods Applied to the Determination of Uranium in Low-Grade Ores
The following report covers investigations on the application of the beta counter as an analytical tool for use in experiments on the extraction and concentration of uranium from low-grade (.005-.02 % uranium) shale and phosphate rock ores.
Bonding of uranium and zirconium alloys
Report describing a study of fabricating conditions for bonding zirconium and uranium alloys by roll cladding.
Canning graphite for gas-cooled reactors
From abstract: "A preliminary investigation was made of techniques and materials for canning graphite to protect it for use at high temperatures in a nitrogen-oxygen atmosphere"
Carbide Coatings on Graphite
From abstract: "A Method has been developed for the uniform coating of graphite tubes with carbides of niobium, tantalum, and zirconium by thermal composition of their respective halide vapors."
The casting of radiator-type fuel elements of the uranium-chromium eutectic alloy
From abstract: "The feasibility of casting a bare, radiator-type fuel element of the uranium-chromium eutectic alloy (5 w/o chromium) was investigated."
Centrifugal Casting of Aluminum-Uranium Alloys
"Centrifugal-casting techniques were investigated as a method of producing hollow cylindrical extrusion billets of aluminum-35 w/o uranium. Among the variables evaluated were melt temperature, mold and pouring-spout configurations, mold speed, and method of pouring. With the equipment employed it was found that the best castings were produced stilizing a pouring temperature of 2400 F, a heavy-walled steel cylinder rotating between 700 to 900 rpm for the mold and bottom-pouring technique employing a retractable pouring spout. Sound, nonporous billets 26 in. long and 5 in. in diameter were produced with a yield after machining of over 75 per cent of the original charge. The major losses occurred in the pouring spout-and-cup assembly. This loss is relatively unaffected by the casting length; and, therefore, coatings of greater length than 26 in. should results in even greater recoveries.
Centrifugal casting of plate-type fuel elements
From introduction: "Centrifugal casting appeared to be the only method by which the intricate shape could be produced in a cold mold. With this in mind, a program was initiated for the purpose of determining the feasibility of centrifugally casting plates of uranium-5 w.o chromium and uranium-2 w/o zirconium."
Ceramic Investigations of UOâ‚‚
This report covers the progress made on an intensive program to develop and evaluate UO2 as a possible fuel element for the PWR.
The Characteristics of the Bond Interface Formed Between Zircaloy 2 and Uranium-12 w/o Molybdenum
The following report analyzes the results taken from studies on the characteristics of the bond interface formed between zircaloy 2 and uranium-12 w/o molybdenum alloys.
Characterization of Inclusion in Dingot Uranium
Abstract: The nonmetallic inclusions in both as-reduced and fabricated dingot uranium have been studied for comparison with those in ingot uranium. Special attention was paid to the hydride for the purpose of determining the amount and distribution in the various types of uranium. The types and distribution of other inclusions were also studied. It was found that the dingot uranium was of a higher quality than ingot uranium and was comparable to as-reduced derby uranium on the basis of over-all inclusion count. The hydrogen content in dingot uranium, however, was found to be appreciably higher than in either ingot or derby uranium.
The Cladding of Delta-Phase Zirconium Hydride
Abstract: A study has been made of the cladding of solid and powdered delta-phase zirconium hydride is both red and flat shapes with stainless steel. The program included investigations of metallurgical bonding, both with and without the sore of metallic barrier materials. Types 304 and 347 stainless steel were used for cladding material. The intermediate barrier-layer materials used were niobium, molybdenum, a combination of copper and molybdenum, and a combination of copper and niobium. The pressure-bonding techniques, involving the use of gas pressure at elevated temperatures, was employed in this study. Variable times and temperatures with a constant pressure of 10,000 poi were utilized by produce bonding. In this study, the best results were archived is cladding delta-phase zirconium hydride directly with Types 304 or 347 stainless steel. Good bonds were obtained by pressure bonding at 1600 F for 3 or 4 hr subsequent to pressure bonding at 1900 F for 1 to 2 hr at a pressure of 10,000 poi. Partial bonding was achieved between niobium and zirconium hydride and molybdeum and girconium hydride.
Composition of vapors from boiling nitric acid solutions
From abstract: "The composition of vapors from aqueous nitric acid solutions boiling at 200 mm mercury total pressure is established for solutions containing between - and 67.5 w/o nitric acid. The volatility characteristics of low concentrations of chloride in the same concentration range of nitric acid have been measured in solutions boiling at 200 mm mercury. The effects of chloride concentration and pressure of boiling are evaluated. A spectrophotometric method of the determination of chloride in nitric acid solutions is described."
Corrosion in 650 F Degassed Water of Uranium-Molybdenum Alloys Containing Impurity Additions
From introduction: "At the request of WAPD, a study has been made of the effect of minor compositional variables on the corrosion behavior of uranium-molybdenum alloys in 650 F degassed water."
Corrosion of thorium and uranium during long-term storage
From introduction: "The over-all objective of the present study was to determine the nature and extent of the present study was to determine the nature and extent of the corrosion of these materials under a variety of such storage conditions and to determine the ability of several potential protective coatings to retard this corrosion."
Corrosion of Type 347 Stainless Steel in the Uranium-Derby Pickle Bath
Abstract: In one of the final stages of the process at the Mallinckrodt Uranium Refining Center, a 45 per cent nitric acid solution at about 170 F is used to pickle the calcined magnesium fluoride scale off the uranium derbies. The increase in the fluoride-ion content of the bath tends to promote corrosion of the Type 347 stainless tanks. This attack becomes excessive if 0.3 g/liter of fluoride ion or more is present. It was found that if aluminum ion is added to the solution the corrosiveness of the bath can be controlled. Two or three times as much aluminum ion as fluoride ion present is found satisfactory at 170 F. Indications are that the tying up of the fluoride-ion by the complexion [AlF6]8 is responsible for the corrosion control.
Corrosion of Uranium Alloys in High-Temperature Water
This report is one of a series of five, dealing with alloys of the uranium-zirconium series. Particularly, this report focuses on the corrosion properties of uranium alloys, with zirconium as the major alloying agent, in high-temperature water.
Corrosion-Resistant Materials for Hydrofluoric Acid : Progress Report
Introduction: Among the many corrosion problems encountered in the production of uranium tetrafluoride and uranium metal from ores and concentrates, some of the most serious occur where hydrofluoric acid must be handled.
Creep of 2S-O Aluminum Sheet at 500 and 550 C
Abstract: "Creep and creep-rupture tests were made on 25-O aluminum sheet at temperatures of 500 and 550 C. The estimated stresses that will produce 0.5 per cent deformation and rupture in 10,000 hours at 400, 450, 500, and 550 C (data at 400 and 450 C are from BMI-T-29, dated June 9, 1950) are presented."
Creep Strength of Uranium Alloys at 1500 and 1800 F
Abstract:"The creep resistance of various uranium binary alloys was investigated at 1500 and 1800 F in vacuum. Tests were made on alloys of uranium with beryllium, columbium, molybdenum, tantalum, titanium, and zirconium and on molybdenum-UO2 composites. Of the alloys examined, those of the uranium-molybdenum system exhibited the best creep resistance. At 1500 F, creep rates of about 0.005%/hr were produced in uranium-molybdenum alloys by a stress of 2500 psi and, at 1800 F, similar creep rates were obtained in composites of 90 wt % molybdenum-10 wt % UO2 by a stress of 12,000 psi."
Critical-Assembly Studies on an Intermediate Reactor for Aircraft Propulsion
The following report studies an intermediate solid-fuel reactor system for aircraft propulsion.
Degradation Products of Tributyl Phosphate
Abstract: "A method for determination of dibutyl phosphate in solvent streams containing tributyl phosphate is outlined. The method is based on analysis for total uranium of total phosphate after removal of uranium and phosphate in excess of that present as uranyl dibutyl phosphate. Results on plant solvent samples are presented. Difficulties with precipitation of [...] dibutyl phosphate, when fresh solvents were employed, are discussed. Small amounts of nonphosphate uranium-complexing agents were found in plant solvents. Diluent degradation is postulated as the source of these components. Emulsion-formation tendencies appeared to correlate better with concentrations of these contaminants than with dibutyl phosphate content."
Delta-Phase Zirconium Hydride as a Solid Moderator
Abstract: "In a study of the preparation and properties of delta-phase zirconium hydride it was found that large, sound bodies of the hydride can be prepared by direct combination of the elements if the rate of the reaction is retarded by limiting the supply of available hydrogen. Specimens up to 1-in. diameter were prepared using this technique. Because delta phase zirconium hydride does not readily form eutectics with iron-and nickel-base alloys below 1800 F these materials may be utilized for clodding the hydride. Delta-phase zirconium hydride is unaffected by exposure to liquid NaK or to nitrogen gas at temperatures below 1000 F. The hot hardness of delta-phase zirconium hydrid is about 130 kg per mm-2 at room temperature and 40 kg per mm-2 at 1500 F. The mean coefficient of thermal expansion (68 to 1337 F) is 6.5 x 10^-6 per deg F. The thermal conductivity varies from 5.7 Btu/(ft)(hr)(F) at 300 F to 5.1 Btu/(ft)(hr)(F) at 1300 F."
Determination of Rare Earths in Refined Uranium : Topical Report
Abstract: "Rare-earth analyses were performed on several samples of refined uranium oxide and nitrate, using a modified cellulose column procedure. Rare earths were not detectable in samples of 0.03 shotgun of less. Detection limits were from 0.0005 to 0.015 ppm. Negligible quantities of low-cross-section rare earths were found in less pure uranium samples, obtained during start-up operations of the pilot plant at the Feed Materials Production Center, primarily those rare earths of greater than average abundance in nature. No correlation was found between rare-earth content and shotgun values. Rare earths do not appear to be contributing significantly to the neutron absorption of refined uranium. A modified cellulose column procedure, which includes a preconcentration solvent-extraction step, is described. Radioactive tracer tests indicating rare-earth recoveries by the method of >98 per cent are reported."
The Determination of Rare Earths in Thorium
Abstract: "A quantitative method for the determination of individual rare earths in thorium down to a level of 0.05 ppm, is described. The procedure consists of a chromatographic cellulose-columns separation followed by a solution-type spectrographic determination. Values are given for the recovery of a number of rare earths using this combined procedure."
Development of Cermet Fuel Elements
Abstract: "Fabrication techniques for making metal-ceramic fuel elements containing 60 to 90 volume per cent of UN or UO2 in a Type 302B stainless steel matrix was investigated. A hot press-forging procedure was most successful for fabricating fuel cores with a density of 90 per cent of theoretical or better. This procedure consisted of sealing the cold-pressed core compacts in stainless steel picture-frame packs, heating to 1900 F, and pressing to a total reduction in thickness of 35 per cent. A pressure of approximately 50 tsi ores used. Specimens produced by this method were evaluated on the basis of their microstructure, modules if rupture, electrical conductivity, and resistance to thermal shock. Microscopic and macroscopic examination showed the presence of a continuous metal skeleton even in specimens containing 90 volume per cent fuel. The modulus of rupture at room temperature varied from 22,500 psi for a specimen cnotaining 63 volume per cent UO2. Both the electrical conductivity and resistance to thermal shock of UO2 were improved by the addition of a small volume of metal. Gas-pressure-bonding techniques appear promising for clodding these cores into composite elements."
Development of Methods for End Capping PWR Fuel Elements
The following report discusses the development of methods for sealing the ends of metallic-cored fuel rods.
Differential Thermal Analysis of Uranium Tetrafluoride-Uranium Dioxide Mixtures
Abstract: "Approximate melting points have been determined for five samples of uranium tetrafluoride representing incompletely converted uranium dioxide and covering the range from about 2 to 20 w/o UO2, using the method of differential thermal analysis. The results indicate the melting temperatures are in the range of 920 to 980 C. No significant correlation between melting point and UO2 content was observed, possibly because of calcite formation. Similar results were obtained on synthetic mistakes of UF4 containing from 10 to 40 w/o UO2."
Differential Thermal-Expansion Effects on Brazed Joints
Abstract: "Differential thermal-expansion effects in brazed joints involving Type 310 stainless steel and GE-62 brazing alloy were investigated. The work included dilation and modulus-of-elasticity measurements using homogenous cast specimens and observations on bimetallic cantilevers made of the two constituents. No anomalies were found, although there were irregularities in the expansion of the brazing alloy which were ascribed to a solubility phenomenon. The elastic modulus of the brazing alloy was determined. Cantilever deflections with temperature and with load were measured, and the results were interpreted using equations which treat the specimens as true bimetals consisting of two homogeneous components. The difference in thermal-expansion coefficients obtained in this way from the temperature-deflection data was consistent with the dilation measurements. The load measurements yielded an average elastic modulus for the bimetal which was about two-thirds of what would have been expected from knowledge of the components. This discrepancy probably arose from porosity which was observed in the braze components."
Dissolution of Aluminum-Canned Thorium
The following report studies the dissolution of aluminum-canned thorium, providing results that suggest a dissolution cycle that permits the separation of the canned-slug components.
The Ductility of Brazed Stainless Steel Joints
Abstract: "The ductility of Type 310 stainless steel T-joints brazed with GE-62 brazing alloy was measured at room temperatures 1200, 1650, and 1800 F. The measure of ductility was taken as the plastic axial strain required to crack braze fillets in T-section tensile specimens. At elevated temperatures, the ductility of as-brazed joints approximated that of the stainless steel, but at room temperature the brazed joints had only one-tenth the ductility of the base metal. Annealing for 16 hr at 1800 F in air was found to triple the room-temperature ductility of the brazed joint."
Eddy-Current Inspection of a Possible PWR Fuel Element
The following report follows the inspections of a number of eddy-current instrument procedures. The objectives of these tests were to detect both weld defects and lack of integrity of the body of the fuel elements.
Effect of Ceramic or Metal Additives in High-UOâ‚‚ Bodies
The following report focuses on research made to determine whether the service performance of UOâ‚‚ fuel-elements cores for the PWR can be improved by certain ceramic or metal additions.
The Effect of Fabrication Variables on the Structure and Properties of UOâ‚‚-Stainless Steel Dispersion Fuel Plates
From introduction: "This report deals with a part of the research and development studies which preceded the manufacture of fuel elements for the Gas Cooled Reactor Experiment (GCRE)." The studies evaluate the effects of varying the type and size of UO2 particles, stainless steel matrix powders, blending procedures, compacting pressures, sintering times, temperatures and atmospheres, roll-cladding temperatures and reduction rates, total cold reduction, and heat-treating times and temperatures has been made for UO2 stainless steel dispersion fuel elements."
The Effectiveness of Spray Cooling
Abstract: "A possible method of cooling a liquid-fuel reactor is by spraying liquid metal through the liquid fuel, and then circulating the liquid metal through a heat exchanger. To evaluate the effectiveness of this cooling method, a few simple experiments were made with mercury sprayed through water. On the basis of the results, it was concluded that this method was intrinsically a low-power-density method, which could not find application except where a low fissionable-material inventory was the dominating requirement in a low-power reactor. Even there, it is thought that a boiling homogeneous reactor might be superior. The results are reported, in spite of their probably lack of value in the reactor program, simply to make the record complete."
The Effects of Chemical Impurities on the Quality of Rolled Uranium Rod
Abstract: "Thirty-four uranium ingots containing controlled amounts of carbon, nitrogen, and Mgl2 slag were cast, rolled, and examined to investigate the relation between these impurities and the quality of the rolled rod. Carbon in concentrations up to 1400 ppm and nitrogen up to 170 ppm, either singly or in combination, had no significant effect on the number of defects observed in the rolled rod. The quality of the rods, however decreased with increasing amount of slag necessary to cause observable differences in the rod could not be detected on analysis, but was visible in the microstructure."
The Effects of Chemical Impurities on the Quality of Rolled Uranium Rod
Abstract: "The effects of nitrogen, slag, and hydrogen additions on the surface quality of rolled uranium rod were studied. The addition of 400 ppm of nitrogen resulted in severe striations. Slag inclusions elongated during rolling and produced short open seams in the etched surface. The effects of 9 to 10 ppm of hydrogen were obscured by defects in the original costings. These defects were caused by oxide films entrapped in the ingot."
Effects of Ternary Additions on Aluminum-35 w/o Uranium Alloys
Abstract: "The effects of a number of ternary additions on the constitution, casting, and fabricating characteristics and the physical properties of aluminum-35 w/o uranium were investigated. Initial investigations were concerned with the effects of 3 w/o ternary additions on the microstructure and press-forging characteristics of the alloy. It was found that additions of this magnitude often introduced extrinsic phases in the alloy. At the 3 w/o level, additions of germanium, silicon, tin or zirconium inhibited the formation of UAl4 and thereby increased the content of the aluminum matrix in the alloy. It was also noted that these additions decreased the pressures required for extruding, and the tin addition also improved the homogeneity of cast shapes. Lead and palladium also improved the homogeneity of the cast material; however, neither of these was an effective inhibitor of UAl4 and free lead was detected in the alloy to which lead had been added at the ternary. From these studies it appears that tin and zirconium are as effective as silicon in enhancing the fabricating characteristics of aluminum-35 w/o uranium alloys, and may prove superior when evaluated on the bases of casting qualities and recycling characteristics."
The Effects of Ternary Alloying Additions on the Corrosion Resistance of Epsilon-Phase Uranium-Zirconium Alloys
Abstract: "The corrosion rate in 680 F water of the uranium-50 w/o zirconium binary alloy was found to be -0.20 mg/(cm2)(hr), and that of the uranium-40 w/o zirconium binary alloy was -0.34 mg/(cm2)(hr). Both alloys correlated uniformly, with no evidence of discontinuous failure. Normal variations in interstitial content in either alloy had no significant effect on corrosion behavior. Tantalum additions, in the range of 0.2 to 5 w/o, improved the corrosion rate of the uranium-50 w/o zirconium base, with a minimum rate of -0.06 mg/(cm2)(hr) for the 5 w/o tantalum alloy. The 5 w/o tantalum addition to the uranium-40 w/o zirconium alloy. The 5 w/o tantalum addition to the uranium-40 w/o zirconium alloy decreased the corrosion rate of the base to -0.11 mg/(cm2)(hr) in the [...] condition only. In either conditions, the 5 w/o tantalum alloy failed discontinuously. All other additions to both bases either had no effect or decreased corrosion resistance. These included aluminum, chromium, iron, molybdenum, nickel, platinum, tin, titanium, tungsten, and vanadium additions."
The Electrical Properties of Uranium Oxides
From introduction: "The work described here was part of an integrated investigation of the fundamental properties of uranium oxides done for the Mallinckrodt Chemical Works. Electrical measurements were employed to characterize the oxides produced by various processes from different starting materials. The basic objective of the program was to determine those factors which affect the sintering characteristics of uranium dioxide."
Electrodeposition of Aluminum on Uranium
Abstract: "Aluminum electroplating was studied in a search for new methods of cladding uranium fuel elements. Uranium electroclad with 12 mils of aluminum over a 0.5-mil (or nickel plus copper) electroplate resisted corrosion for more than 100 hr in boiling water. This quality of protection was effected by hot pressing the electroclad uranium with 5.1 tons per sq in. for 5 min at 950 F. The electroclad uranium with hot-pressed samples paralleled those of later experiments with hot-pressed wrought aluminum claddings on uranium. In both cases, the uranium was electroplated with thin (0.5 mil) layers of metals to prevent aluminum-uranium diffusion, to aid bonding, and to assist in corrosion protection. This aluminum electroplating study helped to define the importance of the intermediate coating between the aluminum and the uranium, the effect of good bonds between the various layers, and the effects of hot pressing in protecting uranium with an aluminum cladding."
Electroplated Metals on Uranium
The following report follows the studies of electroplating on uranium and concurrent metallurgical clodding.
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