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REDUCTION OF CUPRIC OXIDE BY HYDROGEN. II. CONVERSION OF HYDROGEN TO WATER OVER FIXED BEDS

Description: The conditions under which hydrogen could be quantitatively recovered from mixtures of gases by oxidation over fixed beds of CuO were investigated. The conversion of H/sub 2/ to H/sub 2/O by reduction of CuO in fixed beds increased with in- creasing bed length, temperature, hydrogen/argon ratio, and decreasing mesh size of CuO. Residence times required for 99% conversion in a 1- in.-diam. bed were 0.6 and 1.2 sec for 30% hydrogen-70% argon and 10% hydrogen90% argon mixtures, respectively, at a total gas flow of 1 l/min. The CuO used was 25-mil-diam. wires with a surface area of 0.019 m/sup 2//g. The residence time required for a given value of conversion decreased about 10% when the total flow rate was increased from 1 to 1.7 liters/min, which indicates that the reduction is mass-transfer controlled to a slight extent under the experimental conditions used. (auth)
Date: February 12, 1960
Creator: Bond, W.D. & Clark, W.E.
Partner: UNT Libraries Government Documents Department

Discussion on Optimization of Large Oil-Pumped Ultra-High Vacuum Systems

Description: Discussion is directed toward eventual optimization of the largest diffusion-pump systems. Less than 100 diffusion pump fluid molecules per cm/sup 2//sec are possible to detect using an accumulation method. Optimization discussed demands highest possible system speed compatible with the above contamination rate. Bakeable oil diffusion pump systems, with equal orifice valve-trap units, without conventioual baffling, permit theoretical throughput speeds up to 0.3 of the system orifice. The average backstreaming rate of oil in two unconventionally baffled commercial pumps of 6- and 10-in. size is reduced to <3 x lO/sup -4/ g/cm/sup 2//24 hr, including heating and cooling the pump boiler. A Ho facter of> 0.4 is maintained. Results on trapping with activated alumiua, retained by a wire mesh, at both liquid-nitrogen and room temperature are included. A variety of pumping speed measurements and bakeable combination valve-trap units is discussed. (auth)
Date: September 12, 1960
Creator: Milleron, N. & Levenson, L. L.
Partner: UNT Libraries Government Documents Department

MEASURING THE RELEASE OF SHORT-LIVED FISSION GASES DURING CAPSULE IRRADIATIONS

Description: A technique is described for the determination of the release of short- lived fission gases during capsule fuel irradiations. Sweep helium passes over the fuel specimen and carries fission gases to a delay trap located close to the capsule. Short-lived fission gases decay in the trap and deposit long-lived daughters which are assayed by radiochemical methods after trap removal. In two demonstrations of the technique the radiochemical analyses have provided information on the rate of release of 1.7-sec Xe/sup 141/, 16-sec Xe/sup 140/, 3.9 min Xe/sup 137/, and 3.2-min Kr/sup 89/. The technique should find considerable application in the irradiation evaluation of fuel materials when it is necessary to determine the contribution shont-lived fission gases make to the coolant-contamination problem. (auth)
Date: September 12, 1960
Creator: Townley, C.W.; Raines, G.E.; Diethorn, W.S. & Sunderman, D.N.
Partner: UNT Libraries Government Documents Department

Processing of High-Fired Uranium Dioxide Fuels by a Reduction-Mercury Extraction-Oxidation Process

Description: A preliminary flowsheet for the purification of uranium dioxide fuels by a magnesium reduction-- mercury extraction-- steam oxidation process is proposed. Feasibility was indicated by laboratory-scale scouting experiments. Data evaluation indicated 100% reduction of uranium dioxide by magnesium although this figure was not demonstrated, chiefly because of poor choice of materials and design of equipment. Steam oxidation of uranlum tetramercuride produced an oxide with an O/U ratio of 2.43. This ratio was decreased to 2.09 by heating the oxide in a hydrogen atmosphere at 900 deg C for 1 hr. The final product had a surface area of 3.5 m/sup 2//g, and 18% of the panticles were &lt; 1 mu diam. A pellet of the oxide sintered at 1750 deg C had a density of 9.76 g/cc, 89% of theoretical. Decontamination factors demonstrated for ruthenium, cesium, and samarium, when present originally in amounts equivalent to 30,000 Mwd/ton fuel burnup and 60 days&#x27; decay, were
Date: August 12, 1960
Creator: Messing, A. F. & Dean, O. C.
Partner: UNT Libraries Government Documents Department

PURIFICATION OF PROMETHIUM BY LIQUID-LIQUID EXTRACTION

Description: A process was developed for separating promethium from raixed fisBion product rare earths by continuous multistage conntercurrent extraction with 100% tri-nbutylphosphste from nitric acid of 12 N or higher concentration. Distribution coefficients at 12 N acidity for aecdamium. promethium. and samarium are 0.43. 0.82, and 1.55, respectively. Single-stage separation factors of 1.9 between successive elements can be maintained throughout the system to give separations dependent only on the number of stages. Extracted values can be recovered from the organic solution by stripping with a smaller volume of dilute nitric acid. A flowsheet for purification of promethium includes one cycle for separation of promethium from neodymum and lighter elements and a secondycle for removal of samarium and heavier elements. Each cycle consists of a series of countercurrent partitioning stages. followed by stripping stages and an evaporator. With 20 stages in the first cycle and 34 stages in the second, a 90% yield of promethium with a purity of 83% can be obtained from a typical mixture of fission product rare earths, assuming essentially perfect mechanical efficiency. An increase to 34 stages in the first cycle would permit a 93% yield of 99% promethium. (auth)
Date: February 12, 1960
Creator: Weaver, B. & Kappelmann, F.A.
Partner: UNT Libraries Government Documents Department

SPERT PROJECT. Quarterly Technical Report for April, May, June 1959

Description: SPERT I: The characteristics of the boiling process and its relatin to moderator expulsion in Spert I were investigated in a series of capsule type experiments. A fuel-bearing oapsule, instrumented to provide pressure, volume, and temperature data during transient power excursions, was placed in a high flux region of the Spert I P core. Five step-induced transients initiated from boiling indicate that the kinetic behavior of the stainless steel clad P-18/19 core is dependent on initial temperature in a manner similar to that of previously tested aluminum clad spert cores. Reactivity oscillator techniques were used in the P-18/19 core to determine the phase and magnitude of the reactivity-to-power transfer function from 0.01 to 18.4 cps at low power and at temperatures below boiling. Criticality data on relatively simple lattices, both rod-free and containing a single poison rod, were obtained from a series of clean critical experiments performed on a number of light water-moderated and - reflected slab configurations of Spert III fuel elements. Changes in water height during the critical water height experiment were measured to plus or minus 0.0013 inches by means of a simple remoteindicating system designed and built for this purpose. SPERT III: The operational loading was determined to be 44 fuel assemblies and 8 control rods. Experiments revealed that this loading provides about .50 excess at 2500 psi ard 650 deg C and a total rod worth above ambient critical of about .50. Measurements were made of the pressure and temperature coefficients of reactivity and the temperature defect. Apparatus for non-nuclear engineering tests was installed and the tests were initiated to obtain data on core hydraulics, system energy balance, and general plant equipment. A primary coolant sampling station has been added. Experimental apparatus for void experiments was designed and fabrication started. DATA REDUCTION AND INTERPRETATION: A ...
Date: April 12, 1960
Creator: Haire, J.C. ed.
Partner: UNT Libraries Government Documents Department

THE BONDING OF MOLYBDENUM-AND NIOBIUM-CLAD FUEL ELEMENTS

Description: A solid-state bonding technique involving the use of gas pressure at elevated temperatures was utilized for the self-bonding of molybdenum and niobium. Bonding conditions and surface preparation as a function of the integrity of the bond achieved were evaluated for each material. Optimum self-bonding of niobium was achieved by bonding parameters of 2100 to 2300 deg F at 10,000 psi for 3 hr with surfaces which had been prepared by etching in a nitrichydrofluoric acid solution prior to bonding. The process as developed was used to prepare niobium- clad flat-plate- and rod-type fuel elements and flat-plate subassemblies. Niobium tubing was also fabricated by this technique. (Molyb denum self-bonding was most readily achieved by gaspressure bonding at temperatures of 2300 to 2600 deg F at 10,000 psi for periods of 3 hr. With these bonding conditions a number of different surface preparations were satisfactory. Directional ductility of the molybdenum was encountered after bonding and methods to eliminate this were evaluated. Cross rolling with respect to the original rolling direction was shown to improve the ductility of molybdenum-clad specimens. (auth)
Date: July 12, 1960
Creator: Paprocki, S. J.; Hodge, E. S. & Gripshover, P. J.
Partner: UNT Libraries Government Documents Department

URANIUM HEXAFLUORIDE: A SURVEY OF THE PHYSICOCHEMICAL PROPERTIES

Description: >A handbook containing all available current physicochemical data on UF/ sub 5/ is presented. Every effort was made to obtain and consider all reports of original data for incorporation in the compilation. One hundred and forty nine references are given. (J.R.D.)
Date: August 12, 1960
Creator: DeWitt, R.
Partner: UNT Libraries Government Documents Department

Progress Relating to Civilian Applications During May 1960

Description: The investigation of the stabilizing influence of oside additions to UO/ sub 2/ was continued. The effects of fast-neutron irradiation upon the mechanical properties of Type 347 stainless steel are being investigated. Tensile data are reported for alloys of Nb-Cr and Nb-Zr. The major effect of fast-neutron irradiation on the creep properties of Zircaloy-2 is being studied by comparing the in-reactor creep behavior of the alloy with out-of-reactor creep properties. New methods for the determination of low concentrations of oxygen in sodium are being investigated. The mechanisms of wear and friction of various metals in sliding contact in liquid sodium are being studied. Data are presented on corrosion properties, short-time tensile properties, stress-rupture properties, and electrical resistivity for Nb--U alloys. The development of Th and Th--U alloys with improved radiation resistance is being investigated. A program is reported for the determination of diffusion coefficients of Xe in single-crystal UO/sub 2/ specimens and the in-pile study of fission-product release from sintered UO/sub 2/. Thermal-conductivity, thermal-expansion, electrical-resistivity, and modulus-of-rupture measurements are being made on 80 vol.% UO/sub 2/ cermets. Techniques for the fabrication of compartmented Mo- and Nbclad UO/sub 2/ and cermet fuels are being investigated. Compaction studies of UO/sub 2/ were undertaken to provide core materials which will achieve a range of desired densities on pressure bonding. The fabrication of Type 304 stainiess steel-clad UO/sub 2/ rod, tube, and flat-plate elements and flat-plate assemblies was accomplished by gas pressure bonding. Various methods for economically producing dense UC components by powder-metallurgical techniques are being investigated. A study of the properties of UC is under way. Thermal-gradient experiments are being performed to establish a theoretical basis for the prediction of hydrogen migration in zirconium hydride. Experiments were continued in producing single crystals of UO/sub 2/. Portland cement samples were analyzed for sulfate content ...
Date: July 12, 1960
Creator: Dayton, Russell W. & Tipton, Clyde R., Jr.
Partner: UNT Libraries Government Documents Department

Evaluation of External Holdup of Circulating Fuel Thermal Breeders as Related to Cost and Feasibility

Description: The external holdup of expensive materials and associated capital costs for the heat removal systems of fluid fuel breeders were determined. The aqueous homogeneous and molten salt breeders were found to contain substantially less uranium holdup external to the core than the liquid metal fueled breeder. The cost of heat removal and turbogenerator plant equipment for the three systems was compared. (auth)
Date: May 12, 1960
Creator: Spiewak, I & Parsly, L F
Partner: UNT Libraries Government Documents Department

Chemical Technology Division, Chemical Development Section C Progress Report for April-May 1960

Description: An economical process was successfully demonstrated in bench-scale continuous equipment for stripping U from amines with ammonium carbonate solution. A continuous countercurrent mixer-settler extraction system was set up for further testing of the process for recovery of Te, Np, and U by tertiary amine extraction from UF/sub 6/ transfer cylinder was solutions. The effect of Purex aqueous feed adjustment procedures on Pu extraction by 1 M di-secbutyl phenylphosphonate (DSBPP) was studied. Work was continued on plutonium(IV) nitrate extraction with TBP and phenylphosphonate esters. The response of Ru/sup 106/ extraction to variations in the treatment of TBP-Amsco 125-82 solvent was tested. Two solvents have shown ability to extract cesium. (For preceding period see CF-80-3-136.) (W.L.H.)
Date: July 12, 1960
Creator: Brown, K B
Partner: UNT Libraries Government Documents Department

Chemical Technology Division, Chemical Development Section B Monthly Progress Report, June-July 1960

Description: The effect of two neutron poisons, baron and cadmium, on the rate of dissolution of high-density 95% ThO/sub 2/-5% UO/sub 2/ pellets in the Zirflex Process was determined. Dissolution of U-10% Mo alloy in boiling HNO/sub 3/ resulted in a precipitation of uranyl molybdates. Air caused greater uranium and thorium losses during decladding of ThO/sub 2/-UO/sub 2/ fuel than irradiation. Processing of U-Mo fuel by a Zircex type process is discussed. Two leaches of graphitized fuel with 90% HNO/sub 3/ recovered more than 99% of the uranium. Irradiation of synthetic ThO/sub 2/-UO/sub 2/ fuel solution to 5 and 10 watt-hr/l in a Co/sup 60/ source resulted in about a 50% decrease in decontamination factor using the acid-Thorex flowsheet. Corrosion of titanium, tantalum, and Ni-o-nel in Thorex solution and titanium corrosion in various molybdenum core alloy solutions were investigated. The solubilities of ferric mono- and dibutyl phosphates in HNO/sub 3/ and 30% TBP-Amsco-HNO/sub 3/ solutions were determined. Fission product concentrations expected in Purex waste from processing Yankee Atomic Reactor fuel were calculated. Chemical applications of nuclear explosions to H/sup 3/ exchange, reduction of CaSO/sub 4/, and Gnome sampling are discussed. (For preceding period see CF-60-6-108.) (M.C.G.)
Date: December 12, 1960
Creator: Blanco, R E
Partner: UNT Libraries Government Documents Department

Development of Corrosion-Resistant Niobium-Base Alloys

Description: The hot-water corrosion resistance and mechanical properties of niobium and a number of its alloys were evaluated relative to their usefulness in pressurized-water thermal reactors. Unalloyed niobium was found to be rapidly attacked by 600 and 680 deg F water and 750 F steam. A number of alloying additions were found which markedly improve the corrosion resistance of niobium. Of these, binary and ternary combinations of chromium, molybdenum, titanium, vanadium, and zirconium were among the most effective. Many of these alloys exhibited as low or lower weight gains than those obtained for Zircaloy-2 under similar test conditions. Most of the niobium-base alloys tested for strength exhibited excellent resistance to creep at temperatures up to 1200 deg F under stresses through 20,000 psi. (auth)
Date: May 12, 1960
Creator: Maykuth, D. J.; Klopp, W. D.; Jaffee, R. I.; Berry, W. E. & Fink, F. W.
Partner: UNT Libraries Government Documents Department

RAINOUT CONTAINMENT

Description: An evaluation of the efficiency of a sprinkler system as an airborne fission product containment mechanism Is presented. The primary inadequacies and objections to such a system are outlined and data on droplet efficiencies are included. (J.R.D.)
Date: January 12, 1960
Creator: Wegmann, G.L.
Partner: UNT Libraries Government Documents Department

Quarterly Health Physics Report. Through June 30, 1960. (Deleted Version)

Description: A resume of Health Physics activities for April, May, and June, 1960 is presented. Discussions and tabulations which summarize results of field surveys, bioassay, personnel monitoring, and environmental surveys are included.
Date: September 12, 1960
Creator: Meyer, H.E.
Partner: UNT Libraries Government Documents Department

HEAT-TRANSFER EXPERIMENTS ON A PROPOSED FUEL ASSEMBLY FOR THE EXPERIMENTAL GAS COOLED REACTOR. SECTION II FO FUEL-ASSEMBLY HEAT-TRANSFER AND CHANNEL PRESSURE-DROP EXPERIMENT FOR THE EGCR RESEARCH AND DEVELOPMENT PROGRAM

Description: Heat-transfer data are presented for the Experimental Gas Cooled Reactor Title I seven-rod fuel-assembly design. The effect on heat transfer of (1) the radial location of the outer six rods of the seven-fuel-rod cluster and of (2) the addition of helical-finned spacers at the midpoint of each of the seven fuel rods is discussed. The heattransfer data were obtained to verify preliminary general assumptions pertaining to the heat-transfer characteristics of the seven- rod fuel-assembly design and to obtain local heat-transfer correlations. The heat-transfer tests were performed at near-atmospheric pressure using air as the coolant medium. Plots and equations of heattransfer correlations over a Reynolds Number range from 12,000 to 80,000 are included. The test set-up and test procedure are also described. (auth)
Date: April 12, 1960
Creator: Beaudoin, C.L. & Higgins, R.M.
Partner: UNT Libraries Government Documents Department

Discussion on Optimization of Large Oil-Pumped Ultra-High Vacuum Systems

Description: Abstract: "Discussion is directed toward eventual optimization of the largest diffusion-pump systems. Less than 100 diffusion pump fluid molecules per cm 2/sec are possible to detect using an accumulation method. Optimization discussed demands highest possible system speed compatible with the above contamination rate. Bakeable oil diffusion pump systems, with equal orifice valve-tap units, without conventional baffling, permit theoretical through-put speeds up to 0.3 of the system orifice. The average backstreaming rate of oil in two unconventionally baffled commercial pumps of 6-in. and 10-in. size is reduced to < 3 x 10-(-4) g/cm-2/24 hr, including heating and cooling the pump boiler. A Ho factor of >0.4 is maintained. Results on trapping with activated alumina, retained by a wire mesh, at both liquid-nitrogen and room temperature are included. A variety of pumping speed measurements and bakeable combination valve-trap units are discussed."
Date: September 12, 1960
Creator: Milleron, Norman & Levenson, L. L.
Partner: UNT Libraries Government Documents Department

Preliminary design basis modifications for improved coolant backup 100-B, D, F, H, DR, and C areas

Description: The purpose of this document is to establish the design scope for the proposed modifications to the reactor ``last ditch`` cooling systems in the 100-B, D, F, H, DR, and C Areas. The objective in making these modifications is to provide adequate ``last-ditch`` reactor coolant flows for safety of operation at power levels currently programmed for the period CY 1964 when additional ``last-ditch`` cooling facilities are planned in connection with major plant modifications. Additional interim modifications may be required for the last ditch system at the 100-C and DR Areas and for the export water system prior to major plant modifications during CY 1964--1965.
Date: August 12, 1960
Creator: Schack, M. H. & Tupper, W. J.
Partner: UNT Libraries Government Documents Department

Energy release per fission in the Hanford reactors

Description: The average energy release per fission event in a reactor is dependent on the composition and arrangement of the lattice materials. In a study of heat generation in the NPR, Nilson developed expressions for calculating the average energy released in each material per fission event. These relationships have been used in the present calculations to obtain the energy release per fission in existing Hanford reactors.
Date: February 12, 1960
Creator: Morgan, W. C. & Bunch, W. L.
Partner: UNT Libraries Government Documents Department