The Recovery of Uranium from Chattanooga Shales (open access)

The Recovery of Uranium from Chattanooga Shales

Recovery of uranium from Chattanooga shales.
Date: January 14, 1954
Creator: Wilson, R. Q.; Ewing, R. A.; Hanway, J. E., Jr.; Meeley, W. A.; Bearse, A. E.; Filbert, R. B., Jr. et al.
System: The UNT Digital Library
Numerical Results for EGCR Moderator-Element Stress Problems (open access)

Numerical Results for EGCR Moderator-Element Stress Problems

From introduction: "A recent report describes the development of a general program for the IBM Type 7090 electronic computer for calculating plane thermal stresses."
Date: July 3, 1961
Creator: Hulbert, Lewis E. & Redmond, Robert F.
System: The UNT Digital Library
Hazards Summary Report for the Battelle Research Reactor (open access)

Hazards Summary Report for the Battelle Research Reactor

From summary: "This report was prepared for the Advisory Committee on Reactor Safeguards for their review and consideration of the potential hazards concerning the Battelle Memorial Institute Research Reactor."
Date: April 4, 1955
Creator: Chastain, J. W.; Redmond, R. F.; Klickman, A.; Anno, J. & Fawcett, S. L.
System: The UNT Digital Library
Mechanical Properties of Zirconium and Zirconium-Uranium Alloys Containing Tin (open access)

Mechanical Properties of Zirconium and Zirconium-Uranium Alloys Containing Tin

Abstract: "The mechanical properties of a large number of zirconium alloys have been investigated. These alloys include zirconium-tin and zirconium-tin-uranium. Tensile tests have been run on these alloys at room temperature, 500 F, and 600 F. Hot-hardness data have been obtained from room temperature to 1600 F, with particular emphasis on the hot-rolling range. Creep tests and fatigue tests have been run on a number of zirconium-tin alloys at 500 F."
Date: September 29, 1952
Creator: Schwope, A. D.; Muehlenkamp, G. T. & Chubb, Walston
System: The UNT Digital Library
Mechanical Properties of Zirconium and Zirconium-Uranium Alloys Containing Tin (open access)

Mechanical Properties of Zirconium and Zirconium-Uranium Alloys Containing Tin

Abstract: "The mechanical properties of a large number of zirconium alloys have been investigated. These alloys include zirconium-tin and zirconium-tin-uranium. Tensile tests have been run on these alloys at room temperature, 500 F, and 600 F. Hot-hardness data have been obtained from room temperature to 1600 F, with particular emphasis on the hot-rolling range. Creep tests and fatigue tests have been run on a number of zirconium-tin alloys at 500 F."
Date: September 29, 1952
Creator: Schwope, A. D.; Muehlenkamp, G. T. & Chubb, Walston
System: The UNT Digital Library
Mechanical Properties of Iodide Zirconium Alloys (open access)

Mechanical Properties of Iodide Zirconium Alloys

Introduction: "The tensile properties, hot hardness, impact strength; and corrosion resistance of some arc-melted, iodide zirconium alloys have been determined. The alloys investigated include binary alloys of zirconium containing zero to five per cent tin, binary alloys of zirconium containing zero to 0.14 per cent nitrogen, and ternary alloys of zirconium containing tin and nitrogen, tin and uranium, and uranium and beryllium."
Date: November 1, 1952
Creator: Schwope, A. D. & Chubb, Walston
System: The UNT Digital Library
The Preparation of Beta Silicon Carbide (open access)

The Preparation of Beta Silicon Carbide

This report explores the production of beta silicon carbide, a low-temperature, cubic modification of cubic silicon carbide. The temperature, heating time, type of starting material, and proportion of starting materials were all varied in order to test yields of beta silicon carbide.
Date: December 29, 1952
Creator: Gambino, J. R.; Mixer, W. G., Jr.; Wagner, H. E. & Harman, Cameron G.
System: The UNT Digital Library
Evaluation of Oxidation-Resistant Ceramics for High-Temperature Reactor Elements (open access)

Evaluation of Oxidation-Resistant Ceramics for High-Temperature Reactor Elements

Abstract: "As a possible aid in the selection of ceramic materials for use in the fabrication of high-temperature reactor elements, the reported properties of 23 refractory oxidation-resistance ceramic compounds are tabulated. The thermal stresses and heat throughout capacities for nine of these compounds are estimated for conditions of steady heat flow and uniform heat generation in the temperature range of 1500 to 2500 F. The compounds studied have macroscopic thermal-neutron-absorption cross sections lower than 0.2 per centimeter. Data concerning the high-temperature stability of uranium compounds are tabulated for those with melting points above 2450 F."
Date: December 1, 1952
Creator: Harman, Cameron G.; Ferrell, Edward Francis.; Wagner, H. E. & Quirk, J. F.
System: The UNT Digital Library
Mechanical Properties of Zirconium-Tin Alloys (open access)

Mechanical Properties of Zirconium-Tin Alloys

Abstract: "The mechanical properties and corrosion resistance of induction-melted, zirconium-tin alloys have been determined. The alloys investigated contained from zero to five per cent tin and up to 0.3% carbon. The zirconium used was United States Bureau of Mines sponge zirconium. The mechanical properties investigated include the tensile, hot-hardness, and impact properties."
Date: December 22, 1952
Creator: Schwope, A. D. & Chubb, Walston
System: The UNT Digital Library
The Rolling of Uranium (open access)

The Rolling of Uranium

The following report discusses the method of rolling uranium rods and strips under various rolling conditions and the roll-separating force measured.
Date: December 30, 1952
Creator: Saller, Henry A.; Keeler, J. R. & Donley, R. J.
System: The UNT Digital Library
Electroplates on Thorium (open access)

Electroplates on Thorium

Abstract: "The protection of thorium in hot water was studied. Copper electrocladding on thorium offers promising protection against hot water. The corrosion of thorium in 95 C water was reduced appreciably by a thorium phosphate film. The results of this abbreviated investigation offer a basis for the development of a method of protecting thorium at elevated temperatures in corrosive media."
Date: January 9, 1953
Creator: Beach, John G.; Schickner, W. C.; Vaaler, L. E. & Faust, Charles L.
System: The UNT Digital Library
The Fabrication of Thorium Tubes (open access)

The Fabrication of Thorium Tubes

Abstract: "Thorium tubes were cold drawn from extruded tube stock and also from drilled hot-rolled rod. Total reductions up to approximately 85 per cent were accomplished by conventional tube-drawing methods, without an intermediate anneal. The hard-drawn tubes had high tensile strength, but the ductility, as indicated by flaring and flattening tests, was low."
Date: January 16, 1953
Creator: Saller, Henry A. & Keeler, J. R.
System: The UNT Digital Library
Protective Coatings for Molybdenum (open access)

Protective Coatings for Molybdenum

Report discussing studies on various types of protective coatings for a molybdenum sheet. Procedures, materials, and results of each coating are included.
Date: February 17, 1953
Creator: Powell, Carroll F.; Jones, R. P.; Secrest, A. C.; Sherwood, E. M. & Campbell, Ivor E.
System: The UNT Digital Library
Creep Strength of Uranium Alloys at 1500 and 1800 F (open access)

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."
Date: May 28, 1953
Creator: Saller, Henry A.; Stacy, J. T.; Eddy, N. S. & Klebanow, H. L.
System: The UNT Digital Library
The Solubility of Oxygen in Uranyl Sulphate Solutions at Elevated Temperatures (open access)

The Solubility of Oxygen in Uranyl Sulphate Solutions at Elevated Temperatures

Abstract: "The solubility of oxygen in uranyl sulphate solutions and in water at 212, 275, and 325 F, and at oxygen partial pressures up to 1500 psia was investigated. The results are presented in tabular and graphic form. The solubility is proportional to the partial pressure of oxygen. The ratio of solubility in the uranyl sulphate solution to solubility in water at the same temperature and pressure is a function of the concentration of the salt in solution."
Date: June 10, 1953
Creator: Pray, H. A. & Stephan, Elmer F.
System: The UNT Digital Library
The Ductility of Brazed Stainless Steel Joints (open access)

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."
Date: July 8, 1953
Creator: Saller, Henry A.; Stacy, J. T. & Eddy, N. S.
System: The UNT Digital Library
The Application of Ceramics to Hanford Fuel Elements (open access)

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.
Date: September 4, 1953
Creator: King, B. W.; Tripp, H. P. & Duckworth, W. H.
System: The UNT Digital Library
The Properties of High-Uranium Alloys Containing Zirconium or Chromium (open access)

The Properties of High-Uranium Alloys Containing Zirconium or Chromium

From abstract: "Various physical and mechanical properties have been determined for high- and low-carbon base uranium, uranium-chromium, and uranium-zirconium alloys. These data complement those obtained at ANL on the same materials."
Date: February 21, 1955
Creator: Saller, Henry A. & Rough, Frank A.
System: The UNT Digital Library
Extrusion Cladding of Fuel Element (open access)

Extrusion Cladding of Fuel Element

From introduction: "This investigation was undertaken with the general objective of determining whether uranium core could be sheathed with aluminum by an extrusion-cladding process. The ultimate objective of this work was to determine the feasibility of extrusion cladding flat plates of uranium."
Date: March 22, 1955
Creator: Slunder, C. J.; Fiorentino, R. J. & Hall, A. M.
System: The UNT Digital Library
Aluminum Cladding Long Uranium Plates by Solid-State Bonding (open access)

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.
Date: March 29, 1955
Creator: Saller, Henry A.; Dickerson, R. F. & Carlson, R. J.
System: The UNT Digital Library
Electrodeposition of Aluminum on Uranium (open access)

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."
Date: March 30, 1955
Creator: Beach, John G.; Schickner, William C.; Hopkinson, Dolores. & Faust, Charles L.
System: The UNT Digital Library
Electroplated Metals on Uranium for Aluminum Cladding (open access)

Electroplated Metals on Uranium for Aluminum Cladding

Abstract: "Aluminum-clad nickel-plated uranium is corrosion resistant in boiling water. Samples with intentional defects in the aluminum-nickel clodding layers, after testing for more than 300 hr. showed no sign of rapid failure. The study of electroplated metals on uranium for aluminum cladding has defined the effects of several processing variables on the corrosion resistance. The quality of the uranium and the quality of the aluminum-nickel bond were important. Heat treatment and vacuum outgassing of the nickel-plated uranium prior to aluminum cladding were beneficial, and minimizing the amount of air present in the hot-press assembly was desirable. Variation in the steps for electroplating on uranium did not affect the corrosion resistance of aluminum-clad uranium."
Date: May 4, 1955
Creator: Beach, John G.; Schickner, William C.; Hopkinson, Dolores. & Faust, Charles L.
System: The UNT Digital Library
Apparatus for visual study of corrosion by hot water (open access)

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.
Date: 1955
Creator: Grieser, Daniel R. & Simons, Eugene M.
System: The UNT Digital Library
Grain Refinement of the As-Cast Uranium-5 w/o Chromium Alloy by Heat Treatment (open access)

Grain Refinement of the As-Cast Uranium-5 w/o Chromium Alloy by Heat Treatment

Introduction: "The uranium-5 w/o chromium alloy has been intensively investigated for use in a power reactor in the as-cast condition. However, when cast sample fuel elements were subjected to irradiation, their stability was found to be unsatisfactory. In order to improve the behavior of the alloy, a program of grain refinement by heat treatment was begun. The effect of chill casting on the heat-treated grain size was also investigated."
Date: May 13, 1955
Creator: Saller, Henry A.; Rough, Frank A. & Bauer, Arthur A.
System: The UNT Digital Library