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ORGDP Container Test and Development Program Fire Tests of UF6-Filled Cylinders

Description: Fire tests of bare, UF{sub 6}-filled shipping cylinders were conducted at the ORGDP Rifle Range during October 1965 as part of the AEC-ORO Container Test and Development Program presently under way at the ORGDP. The multi purpose effort was to determine if the cylinders would hydrostatically or explosively rupture; the time available for fire fighting before either incident occurred; and the degree of contamination as related to the type of UF{sub 6} release, wind velocity, and terrain. In addition to the cylinder fire tests, other tests were made for further evaluation of the fire-resistant BOX foam plastic. These included a newly designed shipping drum for 5-in.-diam cylinders, and 15B-type wood shipping boxes for small containers. In one case, the latter contained a UF{sub 6}-filled Harshaw cylinder. The test times ranged from 45 to 95 min. In no instance did temperatures exceed 200 F These tests are discussed under Part B. Our Nuclear Engineering Department was responsible for site preparation and the test program. The Safety and Health Physics Departments Mr. A. F. Becher, head, provided primary assistance in the conductance of the tests and was additionally responsible for the environmental monitoring and sampling. Personnel of the Plant Shift Operations and Security, Fabrication and Maintenance, and Technical Divisions provided further support in the various operations. Mr. J. E. Wescott of the AEC-ORO and Mr. J. W. Edwards, ORGDP, were in charge of the motion and still photography. Two each of the following types of cylinders were tested: 3.5 in. diam x 7.5 in. Monel Harshaw, 5.0 in, diam x 30 in. Monel, and 8 in. diam x 48 in. nickel. Fill limits were 5, 55, and 250 lb of UF{sub 6} respectively, at an enrichment level of 0.22%. The larger cylinders were tested individually, with and without their metal valve covers. …
Date: January 12, 1966
Creator: A.J., Mallett
Partner: UNT Libraries Government Documents Department
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STUDY OF RESONANCES IN THE Σ-π SYSTEM

Description: In order to study resonances in the {Sigma}-{pi} system, we have analyzed reactions in which a {Sigma} hyperon and two or three pions are produced in K{sup -}-p interactions at 1.22 {+-} 0.040 and 1.51 {+-} 0.050 GeV/c incident K{sup -} momentum (i. e., 1895 and 2025 MeV center-of-mass energy), using the Lawrence Radiation Laboratory's 72-in. hydrogen bubble chamber.
Date: June 12, 1962
Creator: Alston, Margaret H.; Alvarez, Luis W.; Ferro-Luzzi, Massimiliano; Rosenfeld, Arthur H..; Ticho, Harold K. & Wojcicki, Stanley G.
Partner: UNT Libraries Government Documents Department
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PARTICLE SIZE DISTRIBUTION FOUND IN AIR AT HANFORD PLUTONIUM FABRICATION FACILITIES

Description: The conventional sampling methods of filtration and impaction are generally utilized in the collection of air samples at Hanford. In addition, a Goetz aerosol spectrometer is available for special studies. Particle-size determinations are made by autoradiographic studies making use of both optical and electron microscopes. A supplementary tool sometimes utilized is the Royco Particle Counter. The study included an analysis of the potential usefulness and the limitations of these air sample study methods for plutonium aerosol particle size measurement. Preliminary results, from a continuing study of the particle size distribution in normal and abnormal air, show particle sizes ranging from hundredths to tens of microns. The studies include considerations of the usefulness of particle size data in planning biological experirnents and in the application of these data to human internal deposition cases. (auth)
Date: September 12, 1963
Creator: Andersen, B.V.
Partner: UNT Libraries Government Documents Department
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STAINLESS STEEL WASTES. III. LABORATORY STUDIES OF THE RATE OF REMOVAL OF STAINLESS STEEL IONS BY MERCURY CATHODE ELECTROLYSIS

Description: ABS> The removal rates of iron, nickel, and chromium from synthetic stainless steel waste solutions during electrolysis over a mercury cathode were studied. The loading capacity of the mercury for the stainless steel metals was estimated on the basis of laboratory experiments to be about two% by weight. The laboratory data indicated that, at an electrode potential of --1.80 voits vs S.C.E., 85 ampere-hours per liter of waste removed essentially all of the stainless steel ions from a sulfuric acid solution containing 0.13M metal ions at 35 deg C. (auth)
Date: February 12, 1962
Creator: Anderson, D. R. & Rhodes, D. W.
Partner: UNT Libraries Government Documents Department
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Radioactive contamination in liquid wastes discharged to ground at the separations facilities through December 1962

Description: This document summarizes the amounts of radioactive contamination discharged to ground from separations facilities through December 1962. Detailed data for individual disposal sites are presented on a month-to-month basis for the period of January through December 1962. Previous publications of this series are listed in the bibliography and may be referred to for specific information on measurements and radioactivity totals prior to December 1962. Tables list the major disposal sites in the separation facilities, total volume of waste discharged to each location, and the gross amounts of plutonium and beta particle emitters discharged to ground since startup. This same data is presented on a monthly basis for cribs still in use. Information is presented on the source of the waste stream and the settling facility if used. Isotopic data are included for disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to swamps are also included.
Date: March 12, 1963
Creator: Backman, G. E.
Partner: UNT Libraries Government Documents Department
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The Design Study of Fluid Engine Power Systems

Description: From abstract: This report presents information generated during a six month feasibility study of an engine which uses a supercritical working fluid as the secondary portion of nuclear powered electric generating system.
Date: April 12, 1963
Creator: Baker, C. H.; Hunter, T. A.; Pauliukonis, R. S. & Pradhan, A. V.
Partner: UNT Libraries Government Documents Department
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B-Plant fission product flowsheets. Part 1

Description: B-Plant is currently being evaluated for use as an integrated fission product plant operating in conjunction with the Purex Plant and a waste calcination system. If the forecasted demands for fission products should increase to rates exceeding present capabilities and if private enterprise continues to remain outside the recovery field, present budget plans are to develop the use of B-Plant in three phases. In Phase 1, the B-Plant canyon would be activated and provisions made for preparing and storing fission product concentrates. In Phase 2, additional equipment would be installed to provide a single-line demonstration system for purifying and packaging fission products. In Phase 3, the plant would be converted to a double-line production system for recovering, segregating and storing, purifying and packaging fission products. The purpose of this document is to present the technical bases for B-Plant project scoping studies, including: Design flowsheets for the preparation and storage of fission product concentrates in the scope design of Phase 1 activities; and conceptual flowsheets for the purification of stored concentrates in the engineering studies of Phase 2 activities.
Date: January 12, 1961
Creator: Beard, S. J. & Judson, B. F.
Partner: UNT Libraries Government Documents Department
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Microwave Impedance Measurements and Standards

Description: Report presenting a survey and discussion of well-known microwave impedance measurement techniques. The discussion includes an introduction which emphasizes basic concepts and reflection coefficient-voltage standing-wave ratio relationships. Sources of error in the various measurement techniques are discussed and methods to reduce errors are presented. Methods using rotating loops and resonance lines are included and a brief discussion of microwave impedance standards is given.
Date: August 12, 1965
Creator: Beatty, Robert W.
Partner: UNT Libraries Government Documents Department
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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
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An Investigation of the Corrosion Resistance of Brazing Alloys for Austenitic Stainless Steel Fuel Elements for Service in 565 F Pressurized Water

Description: Since brazing was the method selected for joining the stainless steel SM- l reactor fuel element, corrosion studies were conducted on various potential brazing alloys to evaluate their resistance under the approximate pressurized- water conditions of the SM-1. The program consisted mainly of testing type 304L stainless steel T'' joints brazed with selected alloys in quiescent, degassed, and deionized autoclaved water at 565 deg F under 1200-psi pressure. In the initial phase of the investigation, tests were limited in duration to l000 hr in order to quickly screen some 18 potential alloys for longer time testing. Based on weight-change data and the metallographic examinations, five of the 18 alloys exhibited sufficient corrosion resistance to warrant further investigation. These alloys were subjected to autoclave tests of 12 and 16 months. In these extended tests, 1 cc O/sub 2/liter and a mixture of 1 cc O/sub 2/liter plus 50 cc H/sub 2/liter, respectively, were added to the water to more closely simulate SM- 1 reactor water conditions and to evaluate the effect of different gaseous additions on the corrosion behavior of the alloys. On the basis of weight-change data and metallographic examination after long-term exposure of the tested stainless steel-base joint; these alloys were considered to have acceptable corrosion resistance. No significant differences in the corrosion behavior of these alloys were noted between testing in oxygenated water and water containing the oxygen-hydrogen mixture. (auth)
Date: April 12, 1962
Creator: Beaver, R. J.; Leitten, C. F. Jr. & English, J. L.
Partner: UNT Libraries Government Documents Department
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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
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Maritime Gas-Cooled Reactor Program. Reactor Materials Compatibility With Impurities in Helium

Description: Results are presented of an investigation to determine the maximum permissible partial pressure of carbon dioxide that could be tolerated in the Maritime Gas Cooled Reactor system without oxidizing and embrittling niobium and its alloys, and to screen the heat-resistant nickel- and iron-base alloys for resistance to oxidation and carburization by hydrogen, carbon dioxide, and carbon monoxide. The effect of hydrogen additions on carburization of heatresisting alloys and on the decomposition of carbon monoxide was also studied along with the cataltic effects of heat-resisting alloys on disproportionation of carbon monoxide. (J.R.D.)
Date: January 12, 1961
Creator: Bokros, J. C. & Shoemaker, H. E.
Partner: UNT Libraries Government Documents Department
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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
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Residual Radiation of the LRL 184-Inch Cyclotron

Description: Residual radioactivity at the Lawrence Radiation Laboratory 184-Inch Cyclotron was measured during November 1960. The study was conducted along three principal lines: (1) general survey of radiation levels in the cyclotron vault, (2) activation of foils placed near the cyclotron, and (3) gamma-ray spectra of the cyclotron gap region, including dee structure. Initial radiation levels were less than 8 r/hr which dropped to abcut 10 mr/hr after 48 hr. The observed activities induced in copper foils were Cu/sup 64/ and Co/sup 58/; in iron foils, Mn/sup 52/, Mn/sup 54/, and Mn/sup 56/; in aluminum foils Na/sup / 2>s/sup 4/ The gamma-ray spectra from the gap region included two intense long-lived peaks, at 510 and 810 kev, caused principally by Co/sup 58/. (auth)
Date: July 12, 1961
Creator: Boom, R. W.; Toth, K. S. & Zucker, A.
Partner: UNT Libraries Government Documents Department
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Sample calculation -- GVR values for determining tritium separations costs

Description: This paper shows the calculations for the Gas Volume Ratio, defined as the Total Gas Volume/Target Volume. Using 20 tons of LiAlO{sub 2} as the target, 5028 cubic feet of tritium, protium, and helium are produced. The target volume equals 312.5 cubic feet of aluminate, so that the GVR = 5028/312.5 = 16.1.
Date: December 12, 1966
Creator: Bown, R. W.
Partner: UNT Libraries Government Documents Department
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Protection Against Fallout Radiation in a Simple Structure

Description: A reinforced Butler building was exposed to fallout from Shots Diablo and Shasta, and the resulting dose rates and fallout deposition inside and outside the structure were measured with various instruments and techniques. Protection factors and roof and ground contributions to the total dose rates at points within the structure were determined from the measurements. Comparisons were made with the results of theoretical and other experimental studies. Information obtained from this experiment should be of value as basic experimental data for fallout protection work, although it is recommended that additional substantive data obtained under more controlled conditions.
Date: August 12, 1963
Creator: Breslin, A. J.; Loysen, P. & Weinstein, M. S.
Partner: UNT Libraries Government Documents Department

Oral History Interview with Allan Shivers, April 12, 1965

Description: Interview with former state senator, lieutenant governor, and governor of Texas (1950-57), Allan Shivers from Lufkin, Texas. The interview includes Shivers' observations on his career in the Texas Senate (1934-45), liquor control issue, pari-mutuel betting, lobbyists, oil politics, labor relations, comments about Governors James Allred and W. Lee O'Daniel, race for lieutenant governor in 1946, views on the operation of state government, education, Truman presidential race of 1948, tidelands oil, death of Governor Jester in 1949 and succession to governorship, gubernatorial election in 1950 and the Stevenson-Eisenhower campaign of 1952.
Access: Restricted to UNT Community Members. Login required if off-campus.
Date: April 12, 1965
Creator: Brewer, Thomas B. & Shivers, Allan
Partner: UNT Oral History Program
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Design criteria for coolant backup three remaining smaller reactors

Description: This document defines the objectives, bases, and functional requirements that shall govern the preparation of design of the coolant backup system for three remaining smaller reactors. This project will increase the reliability of the coolant backup facility at B Area by providing an independent last-ditch coolant system to B and C Reactors. The reliability of the last-ditch system for D Reactor will also be improved in that the present F and H leg of the export system will no longer be a part of the new export system.
Date: August 12, 1964
Creator: Brinkman, L. B.
Partner: UNT Libraries Government Documents Department
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Emergency storage basin coolant: Design criteria for architect-engineer usage

Description: This document defines the objectives, bases, and functional requirements that shall govern the preparation of detail design of the gravity fed water supply to reactor storage basins for all eight reactors. In the event that appears advisable and feasible to discharge all metal from the reactors into the metal storage basins, it would be necessary to add water to the storage basins to prevent overheating of the fuel elements. Without the addition of cool water the storage basin water would soon start to boil and evaporate, eventually exposing the metal to the air. Existing facilities do not permit assurance that sufficient water can be added to the storage basins for the required period of time to protect a complete discharge of fuel elements in the storage basin if the area is left unattended and pumps are shut down. A positive gravity fed system to the metal storage basins from existing supplies of stored water shall be provided by this project. This new gravity fed system, once it is started, shall operate unattended and shall supply adequate water to the storage basins for the required period of time. It shall not be dependent on electric or steam-driven pumps for its continuous operation during the critical period; however, provision is to be made to utilize specific electric pumps to refill certain tanks and basins as long as electric power is available in order to maintain the original supply of water as backup for as long a time as possible.
Date: September 12, 1963
Creator: Brinkman, L. B.
Partner: UNT Libraries Government Documents Department
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