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Anion Exchange Recovery of Plutonium From Reduction Residues

Description: Technical report. From Abstract: "An anion exchange process was demonstrated for the recovery of plutonium from waste produced in the reduction of plutonium salts to the metal. Plutonium in a highly salted 6M nitric acid solution derived from the dissolution of slag and crucible waste was separated from impurities by absorbing the Pu (IV) nitrate complex on the anion exchange resin and subsequently eluting with nitric acid. A flowsheet for plant operation is presented."
Date: February 1960
Creator: Russell, Edwin R.
Partner: UNT Libraries Government Documents Department
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Anion Exchange Recovery of Plutonium From Reduction Residues

Description: An anion exchange process was demonstrated for the recovery of Pu from waste produced in the reduction of Pu salts to the metal. Pu in a highiy salted 6M nitric acid solution, derived from the dissolution of slag and crucible waste, was separated from impurities by absorbing the Pu(IV) nitrate complex on the anion exchange resin and subsequentiy eluting with dilute nitric acid. A flowsheet for plant operation is presented. (auth)
Date: February 1, 1960
Creator: Russell, E. R.
Partner: UNT Libraries Government Documents Department
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Boiling Water Reactor Study [Part 2, Supplement]: Separate Studies

Description: From summary and conclusions: This study has been made to determine the feasibility of installing reheat steam facilities in conjunction with the selected 306-mw boiling water nuclear power plant.
Date: February 1960
Creator: Ebasco Services Incorporated
Partner: UNT Libraries Government Documents Department
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The Catalysis of the Hydrogen-Oxygen Reaction by Aqueous Slurries of Thorium Oxide and Thorium-Uranium Oxide

Description: Aqueous slurries of thorium oxide and thorium oxide containing urarium were investigated for their catalytic activity for the reaction of hydrogen and oxygen to form water. Pure thorium oxide. thorium-uranium oxide mixed crystals prepared by calcining coprecipitated oxalates, and thorium oxide with uranium oxide sorbed on the surface were used after calcining at 650, 800, and 1000 deg . The reaction rates were found to be first order with respect to hydrogen pressure and zero order with respect to oxygen pressure in all cases at temperatures from 230 to 300 deg - and total gas pressures from 100 to 2000 psi. For the pure thorium oxide an average activation energy of 41 kcal/mole and an average frequency factor of 4.6 x lO/sup 8/ moles/psi H/sub 2/hr-g of ThO/sub 2/ were found. Addition of uranium lowered both factors, the maximum effect giving a DELTA E/sub a/ of approximately 14 kcal with an A of approximately 10/sup -2/. Actual rates for all catalysts were within one order of magnitude when compared on a unit surface area basis. This compensation effect was explained on the basis of a two-site process, one site being related to the uranium concentration on the catalyst surface and the other characteristic of pure thorium oxide. A few tests on uranium trioxide slurries gave initial fast rates followed by slow ones, the change being accompanied by reduction of the surface uranium under the experimental conditions. The apparent activation energy for both surface conditions was 26 kcal/mole based on first order rate constants with frequency factors of 2.2 x 10/sup 4/ and 2.5 x 10/sup 3/ moles/psi H/sub 2/hr-g for the initial and final rates, respectively. (auth)
Date: February 1, 1960
Creator: Krohn, N. A.
Partner: UNT Libraries Government Documents Department
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CHANGE OF KEWB REACTOR CORES-EVALUATION OF SIGNIFICANCE WITH REGARD TO ASSOCIATED HAZARDS

Description: The KEWB Facility is described, and an over-all technical evaluation is made of the hazards associated with changing from a spherical to a cylindrical core. The characteristics of the two systems, the operation and emergency procedures, a brief history of the program, a summary of the data obtained, and maximum accident analyses are given. The conclusion is that the core change does not represent an adverse change with respect to associated hazards. (T.R.H.)
Date: February 1, 1960
Partner: UNT Libraries Government Documents Department
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CHEMICAL ENGINEERING DIVISION SUMMARY REPORT FOR OCTOBER, NOVEMBER, DECEMBER 1959

Description: Chemical-Metallurgical Processing. A direct-cycle pyrometallurgical fuel-processing plant is being constructed in conjunction with EBR-II. The gamma- irradiation testing of the 175-watt white fluorescent mercury vapor lamp was continued to an integrated exposure of 2 x 10/sup 9/ rad. Irradiation tests of Shell APL grease were completed, and the estimated useful life of this grease in the Air and Argon Cells is 2 and 3 years, respectively. Tests of the three - types of d-c motors used in the operating manipulator of the Argon Cell indicate an expected useful cell operating life of 2 to 3 years. Continued irradiation tests of mineralinsulated cable show no catastrophic breakdown of the electric insulation even after an accumulated gamma dose of 8.4 x 10/sup 9/ rad. The scheme currently under consideration for processing melt-refining residues involves a reduction of skull oxides by a solution of Mg in liquid Cd. Molten salt fluxes had variable effects on the rate of oxide reduction in dilute Mg systems. Work was continued on development of processes for EBR-II blanket materials. A large-scale metal-distillation unit to demonstrate metal distillation at rates up to 100 kg/hour is under construction. Two medium carbon steel thermal-convection loops were built and operated to ascertain the extent of corrosion under adverse thermal conditions. Data were obtained for the solubilities of V, Fe, and Ni in liquid Cd at temperatures from 400 to 650 deg C. Corrections were made to data on the solubility of Th in liquid Zn. Solubilities of U in Si-free Zn--Mg solutions were found to be significantly higher than when silicon was present. The peritectic temperature in the U--Cd system was found to be 474 plus or minus l deg C. The activity coefficient of U in Al at4.8% U and 686 deg C is estimated to be 1.25 x 10/sup …
Date: February 1, 1960
Partner: UNT Libraries Government Documents Department
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Circuit Dynamics of the Pinch

Description: The following document aims to analyze the dynamics of a pinch tube, including the reaction back on the energy source.
Date: February 1960
Creator: Killeen, John & Lippmann, B. A.
Partner: UNT Libraries Government Documents Department
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A Compendium of Information for Use in Controlling Radiation Emergencies: Including Lecture Notes from a Training Session at Idaho Falls, Idaho, February 12-14, 1958

Description: From introduction: This report is a summary of the lecture material in the training course held at Idaho Falls, Idaho, to familiarize members of radiological assistance teams with information helpful in the response to an unusual accident including release of radioactive materials to a populated environment.
Date: February 1960
Creator: Brodsky, Allen & Beard, G. Victor
Partner: UNT Libraries Government Documents Department
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Description of Facilities and Mechanical Components: Medical Research Reactor (MRR)

Description: Report issued by the Brookhaven National Laboratory discussing the Medical Research Reactor at the Medical Research Center. Design, tests, and operations of the reactor are presented. This report includes tables, illustrations, and photographs.
Date: February 1960
Creator: Godel, Jules B.
Partner: UNT Libraries Government Documents Department
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THE DIFFUSION OF KRYPTON-85 FROM URANIUM DIOXIDE POWDER

Description: The diffusion of Kr/sup 85/ in two UO/sub 2/ powders was studied by performing a series of post-irradiation anneals on the powders. The emanation data were analyzed by considering the effect of sintering as well as the effect of a distribution of particle sizes within the sample. Measurements were made at 900 to 1500 deg C. The time at a temperature was between 8 and 24 hours. The diffusion coefficients for Kr/sup 85/ in the two powders are represented by the equations: D = 2.65 x 10/sup -4/ exp - 65,500/RT for UO/sub 2/ prepared from crushed UO/sub 2/ pellets and, for a chemically prepared UO/sub 2/ powder, D = 4.9 x 10/sup -4/ exp - 73,800/RT. (auth)
Date: February 1, 1960
Creator: Auskern, A.B.
Partner: UNT Libraries Government Documents Department
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The Drip Casting of Zirconium Metal. Work Completed: March 1951

Description: A drip casting process initiated to obtain zirconium castings uncontaminated by the melting process and to remove volatile impurities from the zirconium feed rod is described. A feed rod of zirconium is held above the mold, and the bottom of the rod is melted rapidly off into a mold to produce the casting. The melting process is carried out under high vacuum, so that very little atmospheric contamination can result, and some removal of volatile impurities is possible. Since no crucible is used to contain the molten metal, no contamination can result from this source. (auth)
Date: February 1, 1960
Creator: Dunworth, R. J. & Macherey, R. E.
Partner: UNT Libraries Government Documents Department
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EURIPUS-3 AND DAEDALUS--MONTE CARLO DENSITY CODES FOR THE IBM-704

Description: EURIPUS-3 calculates the one-dimensional spatial density of neutrons slowing-down past a given energy in an infinite homogeneous medium consisting of hydrogen and one other isotope with arbitrary mass and energydependent differential-elastic and absorption cross sections. DAEDALUS determines the corresponding spatial distribution of angular integrals of an arbitrary function times the vector flux density. Spatial moments of all density functions are furnished directly. Although scattering angles are calculated by Monte Carlo, the spatial distributions and, in DAEDALUS, the energy distribution are obtained partly from an analytic treatment which, besides saving tinne, enables the output to be in the form of actual density functions at specified planes and energies, rather than histograms covering finite intervals. At certain steps in the computation of both the spatial and energy distributions, part of the analytic treatment is replaced by Monte Carlo in order either to maximize efficiency and/ or to avoid round-off error. The neutron source may be monoenergetic with either isotropic or monodirectional angular distributions, or else the source may be that from deuterons bombarding deuterons. The volume displaced by a cylindrical tube from an accelerator to the source can be accounted for in the neutron first flight but not thereafter. (auth)
Date: February 1, 1960
Creator: Amster, H. J.; Kuehn, H. G. & Spanier, J.
Partner: UNT Libraries Government Documents Department
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An Experiment to Measure Effective Delayed Neutron Fractions

Description: >An experimental measurement of the effective delayed neutron fraction ( beta -bar) was made for a clean critical assembly by determining the asymptotic period associated with introduction of a known amount of reactivity. The "known amount" of reactivity was obtained by replacing, uniformly throughout the reactor, a small quantity of U/sup 235/ with an alloy of B/sup 10/ and Hf designed to match the absorption properties of U/sup 235/. The replacement was thus equivalent to a uniform reduction in nu , the number of neutrons emitted per fission from the fuel. Such a reduction introduces a reactivity change equal exactly to delta nu / nu /sub 0/. Two analyses of the experiment were made using different high energy cross sections in conjunction with four group, two dimensional diffusion theory. The measured value of beta lay between the results of these computations, the error spread (an average rms error of plus or minus 5.2%) being too great to permit any conclusion regarding the significance of the comparison. (auth)
Date: February 1, 1960
Creator: Kaplan, S. & Henry, A.F.
Partner: UNT Libraries Government Documents Department
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Fuel element handling before irradiation

Description: This report on fuel element handling presents in some detail the current status of an engineering study which has been underway for some time, and which is continuing. The study was undertaken to determine if it is feasible, and if it is practicable, to revise the method and equipment used for fuel element handling with existing charging machines.
Date: February 1, 1960
Creator: Gilbert, R. D.
Partner: UNT Libraries Government Documents Department
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Heavy Water Moderated Power Reactor Plants [Part 1]: Design Study

Description: From preface: This report is one of a series covering a design study of heavy water moderated power reactor plants with the objective of selection and recommendation of a conceptual design for providing optimum economics with slight fuel enrichment.
Date: February 1960
Partner: UNT Libraries Government Documents Department
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IMPROVED METHOD FOR PRECIPITATING MANGANESE DIOXIDE

Description: An improved method for precipitating manganese dioxide was demonstrated that significantly increases the allowable feed rate of the Purex head-end centrifuge. The effects of several process variables are discussed. (auth)
Date: February 1, 1960
Creator: Clark, H.J. Jr.
Partner: UNT Libraries Government Documents Department
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THE INTERNAL FEEDBACK OF EBR-I MARK-III

Description: No Description Available.
Date: February 1, 1960
Creator: Carter, J. C.; Sparks, D. W. & Tessier, J. H.
Partner: UNT Libraries Government Documents Department
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The Liquid Metal Thorium Breeder Reactor

Description: From introduction: "Investigation of breeding potential and technical and economic feasibility of breeding in the Liquid Metal Fuel Reactor concept."
Date: February 1960
Partner: UNT Libraries Government Documents Department
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Material Balance Flowsheets

Description: Material balance flowsheets are presented for the dissolution of UO/sub 2/, UO/sub 2/-ThO/sub 2/, and U-Mo fuels clad in stainless steel or zirconium by the Sulfex, Darex, and Zinflex process. The mechanics of the three processes are discussed. Basic assumptions upon which the flowsheets are based are contained. (auth)
Date: February 1, 1960
Creator: Shappert, L.B.
Partner: UNT Libraries Government Documents Department
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