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DMM: A Multigroup, Multiregion, One-Space-Dimensional Computer Program Using Neutron Diffusion Theory. Part 1 - The Theory

Description: DMM is a program using one-space-dimensional multigroup diffusion theory to calculate the reactivity or critical conditions and flux distribution of a multiregion reactor. Calculations of fission-produced xenon and samarium and time variation due to production and depletion of isotopes are an essential part of this program. The adjoint fluxes may also be computed, and the program includes the calculation of the nuclear constaants from fairly simple input combined with a library of cross sections. The present code is written for the Remington Rand 1103A. Operating instructions are presented in Part II. (auth)
Date: December 31, 1960
Creator: Leshan, Edward J. & Kavanagh, Deveroux L.
Partner: UNT Libraries Government Documents Department
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DMM: A MULTIGROUP, MULTIREGION ONE-SPACE-DIMENSIONAL COMPUTER PROGRAM USING NEUTRON DIFFUSION THEORY. PART II. DMM PROGRAM DESCRIPTION

Description: Operating instructions are presented for DMM, a Remington Rand 1103A program using one-space-dimensional multigroup diffusion theory to calculate the reactivity or critical conditions and flux distribution of a multiregion reactor. Complete descriptions of the routines and problem input and output specifications are also included. (D.L.C.)
Date: December 31, 1960
Creator: Kavanagh, D.L.; Antchagno, M.J. & Egawa, E.K.
Partner: UNT Libraries Government Documents Department
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[Hanford weekly teletype report]: Supplement report for week ending June 12

Description: This document contains information about flooding of the Columbia River. It focuses attention on the following; increased elevation due to rainfall, seepage which destabilized the constructed dike, flooding of cellars, evacuation of people to emergency shelters, tug boat collision damage to power lines, and the washout of the Van Giesen Street Bridge on Yakima River.
Date: December 31, 1960
Partner: UNT Libraries Government Documents Department
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Applied Mathematics Division Summary Report for July 1, 1958 Through June 30, 1959

Description: The objective of the Applied Mathematics Division is to provide mathematical assistance to other scientists in the Lab. This goal is achieved by (1) conducting research in numerical analysis and other branches of mathematics, (2) providing mathematical consultation, and (3) operating a computational service, using both digital and analog machines. Publications, papers, seminars, lectures, and courses are listed. A summary listing of computer programs developed or in progress is given. (For preceding period see AHL-5954.) (W.D.M)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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Applied Mathematics Division Summary Report for July 1, 1959 Through June 30, 1960

Description: A summary of each computer program initiated during the report period together with code symbols indicating the extent to which information concerning the program is readily available are given. Programs previously reported are included if changes were made or additional information concerning them was placed in the program library. Abstracts of 704 newsletters and GEORGE bulletins are presented. (For preceding period see ANL-6089.) (W.D.M.)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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CHEMICAL ENGINEERING DIVISION SUMMARY REPORT FOR JANUARY, FEBRUARY, AND MARCH 1957

Description: Fluoride Volatilization Separations Process. Development of a fused fluoride process for dissolution of uranium-- zirconium fuel alloys continued. In corrosion tests to find a suitable container material, Ni was found to be susceptible to a sulfur-type attack. Hastelloy B showed promise, and graphite offers excellent chemical resistance but poor mechanical strength. The dissolution rate of Zr in NaF-- ZrF as affected by impingement of the HF sparge was studied. Production of UF/sub 6/ by fluidized bed fluorination of UF/sub 4/ from ore concentrates was studied. The preparation, melting point, vapor pressure, and vapor density of UF/sub 5/ are given. Preliminary dissolution and recovery runs in semi-works equipment are discussed. Fluidization. Fluidized- bed techniques have been applied to conversion of UO/sub 2/(NO/sub 3/)/sub 2/ to UF/sub 4/ and to calcination of radioactive liquid wastes. Activities of the Green Salt Pilot Plant and shakedown runs of the shielded waste calciner are described. Reactor Chemistry. Studies continued on the kinetics and mechanism of oxidation of U, Th, and Zr. Data are given for oxidation of U in oxygen from 125 to 295 tained C and 20 to 800 mm pressure, and for Zr from 400 to 900 tained C and 200 nan O/sub 2/ pressure. The ratio of capture to fission cross sections for U/sup 233/ and U/sup 238/ in EBR-I have been determined as a function of position. ChemicalMetallurgical Separations Processes. Development of pyrometullurgical processing of spent reactor fuels continued. Work is repcrted on: melt refining and casting of U--Pu; iodine volatility problem; the system U--B-- Ta; the distribution coefficients for Pu between U--Cr and Mg and U and Mg; extraction of Pu from U by liquid Mg; Ce removal by dross refining; adsorption of volatilized metuls on surface active materials; and fractional crystallization of U with Zn. Analytical Research. A study …
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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CHEMICAL ENGINEERING DIVISION SUMMARY REPORT FOR JANUARY, FEBRUARY, MARCH 1960

Description: Chemical-Metalluaical Processing. A direct-cycle fuelreprocessing plant using pyrometallurgical procedures is being designed as part of the Experimental Breeder Reactor No. II project. The reduction of uranium oxide was investigated, using pure Mg and solntions of Mg in Zn and Cd. Cadinium solntions of U were shown to be stable in Types 405 and 410 stainless steel containers at temperatures up to 550 deg C. The liquid metal corrosion loop in which a U-Mg--Cd alloy is being circulated at 550 deg C has been in trouble-free operation for 3000 hrs. Recovery of Pu from Mg solution by distillation of Mg was demonstrated on 1-g Pu scale. The solubility of Th in liquid Cd was measured over the temperature range from 1.9 x 10/sup -3/ per cent at 348 deg C to 1.8 x 10/sup -2/ per cent at 658 deg C. The solubility of Mn in liquid Cd was found to range from 0.27% at 414 deg C to 1.43% at 661 deg C. The solubility of Ni in liquid Cd was measured. The partition of U between liquid Al and liquid Cd was studied as a function of U concentration. The reaction of Al with a liquid Cd solution containing U, Zr, and Ce was studied. The free energy of formation of the U--Pb intermetallic compound UPb/sup 3/ was measured between 374 and 846 deg C by means of a galvanic cell method. Magnetic susceptibility measurements on the intermetallic comPound CeCd/sub 11/ were made over a range of temperature from 4 to 295 deg K. Fuel Cycle Applications of Volatility and Fluidization Techniques. The Direct Fluorination Process is currently aimed toward the processing of the Zircaloyclad, UO/sub 2/ fuel typical of the Dresden Reactor. The direct fluorination of dense UO/sub 2/ pellets submerged in an inert fluidized medium was carried …
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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DEVELOPMENT OF PLUTONIUM BEARING FUEL MATERIALS. Progress Report for January 1 through March 31, 1960

Description: Construction of the NUMEC Plutonium Facility was essentially completed. Methods for the preparation of Pu, U, and Th oxides of high purity and the fabrication of these materials into fuel shapes is discussed. (For preceding period see NUMEC-P-10.) (W.L.H.)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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Elk River Reactor Quarterly Project Report for September-October-November 1959

Description: With the project design work virtually complete, progress during the quarter consisted primarily of prccurement, fabrication of components, and construction at the reactor site. Developments are briefly summarized in the fuel element program, core physics vessel and internal components, control rods and rod drives, shielding, process, instrumentation, building and facilities, and construction. (For preceding period see ACNP-ERR-5.) (W.D.M.)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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Fuel Element Catastrophe Studies Hazards of Fission Product Release From Irradiated Uranium

Description: The rate of reaction of highly irradiated U with air, CO/sub 2/, and steam was studied in sn investigation of the fission product release potential in a loss-of-coolant type accident postulated for Pu-producing reactors. Highly irradiated U was found to be more reactive, probably because of the defects in the oxide coating formed by the inclusion of fission products. Complete oxidation or melting was found to release rare gases, I, and Te semi- quantitatively in most atmospheres. Other fission products (Ru, Cs, and Sr) were released to a lesser extent and apparently in proportion to the amount of self- heating induced. In order of their relative tendency to release fission products, the atmospheric conditions investigated were rated in the order: air > CO/sub 2/ > steam. (auth)
Date: October 31, 1960
Creator: Parker, G. W.; Creek, G. E.; Martin, W. J. & Barton, C. J.
Partner: UNT Libraries Government Documents Department
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FUEL ELEMENT DEVELOPMENT PROGRAM FOR THE PEBBLE BED REACTOR. Quarterly Progress Report for November 1, 1959 to January 31, 1960

Description: A variety of spherical uranium--graphite fuel elements for the Pebble Bed Reactor (PBR) was fabricated. Poor results with sintered alumina coated UO/ sub 2/ particles led to the development of slumina coating by vspor deposition, for which good results have been obtained. A variety of sub-surface metal, metal carbide, and ceramic coatings located between an unfueled graphite shell and the fueled core of a PBR fuel element was prepared and examined. Most of the materials and processes showed poor results. Excellent metal recoveries were achieved from the metal oxide--graphite system using the grind-leach technique of reprocessing. Test results on Si-- SiC coated fuel elements showed good fission product retention in neutron activation tests, a self-welding tendency between adjacent spheres at 2500 deg F surface temperature, and no evidence of failure when an interanl gas pressure of 300 psi was applied. Fission product release rates from a pyrolytic carbon coated specimen under low-level irradiation were obtained at 150 to 1900 deg F. The design of the in-pile loop to study the behavior of fission products escaping from PBR fuel elements wss established. (C.J.G.)
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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Improved coolant backup 100-B, D, F, DR, H and C areas design study

Description: Preliminary engineering studies have indicated the need for modifications and improvements to the reactor coolant backup systems of the old areas in order to provide adequate safety of operation at power levels programmed for the future. These evaluations of the coolant backup systems were based on the recently adopted reactor cooling safety criteria. It was concluded that the secondary coolant systems would be adequate in capacity and reliability for the proposed future operating conditions except for certain cases of natural disaster such as earthquake damage. It was concluded that the last ditch coolant systems would be inadequate for the proposed future reactor operating conditions. The purpose of this report is to define the scope of modifications and improvements required to provide adequate last ditch systems in the old areas for future operating conditions as proposed by the Reactor Modification Program. Irradiation Processing Department, Fiscal Years 1961 through 1966. Adequate last ditch cooling will be provided for the 100-K Areas under Project CGI-844 which is currently in progress. The results of this study provide a basis for future budgeting action and project planning.
Date: October 31, 1960
Creator: Schack, M. H. & Tupper, W. J.
Partner: UNT Libraries Government Documents Department
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AN IMPROVED NUCLEAR DENSITY GAUGE. Period covered: September 2 to October 1, 1959

Description: Progress in development of a nuclear density gage for use in thickness and density measurements is reported. A Ross filter system for energy discrimination in the alpha ray region was constructed. Soller slits for collimating the radiation entering and leaving the filters so that a constant absorber thickness is seen by the radiation beam is under construction. Comments by visiting Russian scientists on the density gage and a discussion of their work in this area are included. (J.R.D.)
Date: October 31, 1960
Creator: Burgwald, G.M.
Partner: UNT Libraries Government Documents Department
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An Improved Nuclear Density Gauge Progress Report: October 1959

Description: Progress is reported on development of a scintillation detector which alternately views the radiation transmitted through a sample and through a calibrated wedge. From this information density and thickness data can be obtained. Long term stability measurements are being made on the commutating system and the causes of fluctuations are being investigated. Information concerning procurement and fund expenditure are given. (J.R.D.)
Date: October 31, 1960
Creator: Burgwald, G. M.
Partner: UNT Libraries Government Documents Department
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An Improved Nuclear Density Gauge. Quarterly Report No. 2 Covering Period September 1 to December 1, 1959

Description: Techniques for improving stability of scintillation counter chopper systems and their development for use in industrial control applications are being investigated. Tests are being performed to determine the causes of fluctuations. An integration and frequency converter is being developed in which the anode current of the photomultiplier tube is alternately switched between two integrating condensers by action of a commutator switch. The voltages developed across the two condensers are proportional to the respective intensities of the two radiation beams seen by the scintillation counter. Slective sampling by the use of Loss filters is scheduled for future investigations and a literature search on counter stabilization is also planned. (J.R.D.)
Date: October 31, 1960
Creator: Burgwald, G. M.
Partner: UNT Libraries Government Documents Department
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ISOTOPIC METHOD FOR AGE DETERMINATION OF INDUSTRIAL PRODUCTS. Quarterly Progress Report, July-September 1959

Description: >The materials and methods used for the long-term nuclide pairs are essentially the same as those previously described. (April - June Report). The split planchet method was used to determine the resolving time of the tracerlab superscalar counting apparatus which is currently in use. A planchet was split in two and a sample of nuclide placed in each half and counted. From these results a formula for calculating true counting rates was evolved. Data on resolving time loss for the Superscalar are tabulated. Transmission values of Sr/ sup 89/- Ct/sup 14/ mixture were measured to determine the degree to which error in these values can be reduced by long counting periods. Data on P/sup 32/ and C/ sup 14/ tagged rubber age counts are tabulated. Also, rubber samples were prepared by adding Ca/sup 45/, Cl/sup 36/, Nb/sup 95/, and T c/sup 99/, mixing and vulcanizing and isotope pairs (Ca/sup 45/- Cl/sup 36/ and Nb/sup 99/-Tc/sup 99/) were similarly incorporated in other samples. Ground and suspended samples of these preparations were age counted. The results obtained by incorporating radioisotopes into rubber can not show close correlation between observed and calculated time. Probable reasons for errors and inconsistencies are discussed along with plans for future work. (J.R.D.)
Date: October 31, 1960
Creator: Gregson, T.C. & Waisbrot, S.W.
Partner: UNT Libraries Government Documents Department
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MAGNETIC RECORDER FOR NUCLEAR PULSE APPLICATION. Covering Period: August 6, 1959 to October 5, 1959

Description: Direct recording of nuclear pulse height data on magnetic tape is being investigated. The characteristics of various brands of commercial tape are being investigated and a waveform analysis is being performed in an attempt to determine the most favorable frequency range for available tapes. Use of the magnetic modulator head is being investigated to minimize variations due to short term tape speed variations. (W.L.H.)
Date: October 31, 1960
Creator: Burgwald, G. M. & Stone, C. A.
Partner: UNT Libraries Government Documents Department
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Maritime Gas-Cooled Reactor Program Quarterly Progress Report for the Period Ending March 31, 1959

Description: Turbomachinery considerations indicated that it would be desirable to reduce the cycle pressure from 1,000 to 800 psia. The problem of determining the temperature distribution and the resulting thermal stress pattern within the graphite was considered. Preliminary designs for a heterogeneous fuel element and a semihomogeneous fuel element were developed. Utilization of the Hanford in- pile gas loop for fuel element testing is discussed. Two-group PDQ calculations were run to estimate control rod worth for the preliminary design core under cold, clean conditions. Curves of rod worth versus position were developed for the hot, clean and the cold, clean preliminary design core. A detailed lifetime calculation was made for the preliminary design heterogeneous core. Fuel cycle costs were estimated on the basis of the effect of B in the fuel elements. The schedule and facilities for the critical experiments are discussed in some detail. The speed of the main turbine shaft was tentatively set at about 12,200 rpm. The design and fabrication of a test stand to evaluate shaft seals and seal systems were completed and trial runs were made. The effects of minor heat transfer due to heat leakage, fluid flow, and thermodynamic phenomena on MGCR full-load cycle performance were studied. Operating characteristics of the heat exchanger test facility are described. A critical review was conducted on the desirability of using concentric ducts and valves. Block diagrams outlining reactor power level, outlet temperature, and plant inventory control are presented. Equations which permit the dynamic analysis of a closed-cycle gas-turbine plant were programed for a digital computer. Descriptions were prepared for fluid-mechanical systems. Several methods of purifying He in both storage bank and main loop were investigated. Investigations into the maximum operating temperatures of the various electrical equipment indicate that temperatures up to 140 deg F can be tolerated. Preliminary …
Date: October 31, 1960
Partner: UNT Libraries Government Documents Department
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