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100-Watt Curium-242 Fueled Thermoelectric Generator--Conceptual Design. SNAP Subtask 5.7 Final Report

Description: A thermoelectric generator which produces 100 watts of electrical power continuously over a six-month operational life in a space environment was designed. It employs the heat produced by the decay of Cm/sup 24/ as the source of power. Uniform output over the operational life of the generator is accomplished by means of a thermally actuated shutter which maintains the hot junction temperature of the thermoelectric conventer at a constunt figure by varying the amount of surplus heat which is radiated directly to space from the heat source. The isotopic heat source is designed to safely contain the Cm/sup 242/ under conditions of launch pad abont and rocket failure, but to burn up upon re-entry to the earth's atmosphere from orbital velocity. (W.L.H.)
Date: May 1, 1960
Creator: Weddell, J. B. & Bloom, J.
Partner: UNT Libraries Government Documents Department
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Active Metal Reduction of Plutonium Trichloride

Description: The reduction characteristics of plutonium trichloride are investigated. A flowsheet for batch reduction with Ca is included. (J.R.D.)
Date: May 1, 1960
Creator: Soine, T. S. & Hopkins, H. H., Jr.
Partner: UNT Libraries Government Documents Department
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Analog Computer Study of the MSR-ORR in-Pile Pressurized Water Loop No. 1

Description: A study of the dynamic behavior of the Merchant Ship Reactor Pressurized Water Loop was made using the Reactor Controls Analog Facility. Computer curves show the predicted response of the loop temperatures to normal load changes and component failure accidents. Except for complete flow stoppage, which was not investigated here, the safety system was shown to be adequate in curbing loop temperature excursions due to postulated accidents. (auth)
Date: May 1, 1960
Creator: Ball, S. J.
Partner: UNT Libraries Government Documents Department
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Aqueous Corrosion of Magnesium Alloys

Description: BS>The aqueous corrosion of Mg alloys was investigated at 53 to 150 deg C. Corrosion rates rose rapidly with temperature, reaching about 3 mils per day at 150 deg C for AZ-31 STAMg-2.5 to 3.5 wt.% Al-0.7 to 1.3 wt.% Zn-0.2 wt.% Mn!. Additions of small amounts of Cu and/or Ni to the basic AZ-31 composition reduced the corrosion rate at 150 deg by a factor of about two. Sn may be advantageously substituted for Zn in AZ-31. Control of the pH in the range between 6 and 7 and maintenance of a fluoride concentration in the range between 1 and 10 ppm reduced the corrosion rate of AZ-31 to about 0.1 mil per day at 150 deg C. (auth)
Date: May 1, 1960
Creator: Greenberg, S. & Ruther, W. E.
Partner: UNT Libraries Government Documents Department
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ARMY GAS-COOLED REACTOR SYSTEMS PROGRAM. GCRE-I HAZARDS SUMMARY REPORT. ADDENDUM III

Description: The hazards evaluation was modified to reflect certain changes made to the equipment as a result of operating experience. These changes included: the addition of a startup interlock circuit; the modification of a startup interlock circuit; several minor modifications to the control rod actuators; and the addition of the tube-sheet cooling system. (M.C.G.)
Date: May 1, 1960
Partner: UNT Libraries Government Documents Department
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Bone Marrow Activity in Vitro Under the Influence of Anemic Serum and Human Erythropoietin

Description: A method is described for observing the uptake of Fe by rat bone marrow cells in vitro. Results of experiments on effects of anemic serum and human erythropoietin are presented with a brief discussion. It is concluded that the differences in uptake of Fe/sup 59/ are the result of isotope dilution. (auth)
Date: May 1, 1960
Creator: Beck, J. S.
Partner: UNT Libraries Government Documents Department
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CHARACTERISTICS OF ANODIC AND CORROSION FILMS ON ZIRCONIUM

Description: Zirconium anodizes similarly to tungsten in respect to the change of interference colors with applied voltage. However, the oxide layer on tungsten cannot reach as great a thickness. Hafnium does not anodize in the same way as zirconium but is similar to tantalum. By measuring the interference color and capacitative thicknesses on zirconium (Grades I and III) and a 2.5 wt.% tin ailoy, the film was found to grow less rapidly in terms of capacitance than in terms of iaterference colors. This was interpreted to mean that cracks develop in the oxide as it thickens. The effect was most pronounced on Grade III zirconium and least pronounced on the tin alloy. The reduction in capacitative thickness was especially noticeable when white oxide appeared. Comparative measurements on Grade I zirconium and 2.5 wt.% tin alloy indicated that the thickness of the oxide film on the tin alloy (after 16 hours in water) increased more rapidly with temperature than the film on zirconium. Tin is believed to act in ways to counteract the tendency of the oxide to form cracks, and to produce vacancies which promote ionic diffusion. (auth)
Date: May 1960
Creator: Misch, R. D.
Partner: UNT Libraries Government Documents Department
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DECONTAMINATION TESTING OF HIGHLY CONTAMINATED PROTECTIVE COATINGS

Description: Decontaminability measurements of 34 protective coatings (primarily vinyls and epoxies) and 10 comparative control surfaces (metals, glass, and plastics) highly contaminated with either mixed fission products from typical Purex waste or Oak Ridge National Laboratory Thorex waste did not indicate any unusual improvement over coatings tested in 1950 (AECD-2296). These measurements and chemical rcsistance tests were made of the protective coatings in common acids, alkalis, and deionized water at 81 and 120 deg F as a guide in the selection of coatings for radiochemical plant applications and to determine if coatings containing water vehicles could be substituted for coatings containing more hazardous organic solvent vehicles. Vinyl base coatings were superior in both decontaminability and chemical resistance. The two watervehicle coatings evaluated were markedly inferior to organic-vehicle vinyls. A vinyl wallpaper'' was decontaminated poorly with the reagents used. Water and 3M HNO/sub 3/ removed Purex contaminants relatively efficiently but were relatively ineffective for the removal of Thorex contaminants. Coatings determined to be sufficiently decontaminated from Purex waste by a water flush and acid scrub and to be of superior resistance to common acids and alkalis are given. (auth)
Date: May 1, 1960
Creator: West, G. A. & Watson, C. D.
Partner: UNT Libraries Government Documents Department
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The Diffusion of Gold in Gamma Uranium

Description: The diffusion coefficient of tracer amounts of Au/sup 198/ in gamma U is described by an Arrhenius-type equation: D = 4.86 x 10 cm/sup -3/sec exp(-30,400 cal/mol/RT). The values of D/sub 0/ and the activation energy are close to those for self-diffusion in gamma U, indicating that the low activation energy for self- diffusion is due to a general weakness of the lattice rather than to easy compressibility of the U atom. (auth)
Date: May 1, 1960
Creator: Rothman, S. J.
Partner: UNT Libraries Government Documents Department
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Effects of Irradiation on Some Corrosion-Resistant Fuel Alloys

Description: An investigation was made of the behavior under irradiation of uranium- rich corrosion-resistant alloys with the following nominal compositions: U--3 wt.% Nb, U--3 wt.% Nb--0.5 wt.% Sn, U/sub 3/Si (U--3.8 wt.% Si), and U--2 wt.% Zr (diffusion heat treated). The U--3 wt.% Nb alloy in the rolled and gamma- quenched condition was highly unstable dimensionnlly under irradiation. The U--3 wt,% Nb--0.5 wt.% Sn alloy, in the cast and gamma-quenched condition, was only moderately stable dimensionally. The U/sub 3/Si in the cast condition showed good dimensional stability but in the extruded condition developed moderate anisotropic growth. Clad Nb--0.5 wt.% Sn, UaSi (U--3.8 wt.% Si), and U--2 wt.% Zr (diffusion heat treated). The U--3 wt.% Nb alloy in the rolled and gamma- quenched condition was highly unstable dimensionally under irradiation. The U--3 wt.% Nb--0.5 wt.% Sn alloy, in the cast and gamma-quenched condition, was only moderately stable dimensionally. The U/sub 3/Si in the cast condition showed good dimensional stability but in the extruded condition developed moderate anisotropic growth. Clad rods of diffusion heat treated U--2 wt.% Zr alloy were generally highly stable. Only the U/sub 3/Si and U--2 wt.% Zr alloy specimens retained a significant degree of their preirradiation corrosion resistance. (auth)
Date: May 1, 1960
Creator: Kittel, J. H. & Smith, K. F.
Partner: UNT Libraries Government Documents Department
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Electron-Electron Coincidence Spectrometer BRS-IV

Description: This report summarizes the original construction and recent modifications of the electron-electron coincidence spectrometer BRS-IV located in Building 70 of the Lawrence Radiation Laboratory. In addition to new information on the modified spectrometer, this report also contains a large amount of material taken from two previous UCRL reports.
Date: May 1960
Creator: Unik, John P.
Partner: UNT Libraries Government Documents Department
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Fabrication of Yankee Core I Prototype Fuel Element

Description: Introduction: Development of a process for manufacturing fuel elements for the Yankee reactor and fabrication and testing of a prototype assembly is reported.
Date: May 1960
Creator: D'Amore, M.
Partner: UNT Libraries Government Documents Department
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Flux Program for the IBM 704

Description: Monitor activation theory leading to the basic relation between the monitor weight and activity and the neutron flux is presented. Manipulations to convent the theory into a form suitable for machine computation and details concerning programing, accuracy, range, usage, and costs are included along with a sample case. (auth)
Date: May 1, 1960
Creator: Haynes, V. O. & Williams, B. D.
Partner: UNT Libraries Government Documents Department
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Fortran Programming Techniques for Graph Plotting on the IBM-704 Computer

Description: The numerical results of the IBM-704 computer can be put into graphieal form with the aid of the IBM-717 peripheral line printer, and programming techniques using FORTRAN II are described for the production of several kinds of graphs. Different symbols can be used, and a total of 100 horizontal spaces are available, giving a resolution of 1%. (D.L.C.)
Date: May 1, 1960
Creator: Cohn, C. E.
Partner: UNT Libraries Government Documents Department
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GAP: THE IBM 704 GRID ANALYSIS PROGRAM

Description: A description is given of GAP, an IBM 704 program for analyzing simply- supported rectangular beam grillages in which the beams intersect at right angles. The grillage need not be symmetrical and may have both uniformly distributed and concentrated loads. The individual beams may have arbitrary cross sections. Strain energy methods are used to determine the bending, shearing, and torsional effects in a given grid structare. Deflections and moments can be found directly by the program. Series coefficients are provided for additional computation. (auth)
Date: May 1, 1960
Creator: Witt, F J
Partner: UNT Libraries Government Documents Department
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Geologic Reconnaissance of the Topopah Spring and Timber Mountain Quandrangles, Nye County, Nevada

Description: The Topopah Spring and Timber Mountain 15-minute quadrangles comprising about 500 square miles in southern Nye County, Nev., are bounded by longitudes 116°15' W. and 116°30' W. and by latitudes 36°45' N. and 37°15' N. (fig. 1) The quadrangles are mostly within the Nevada Test Site and are approximately 90 miles northwest of Las Vegas (fig. 1). Lathrop Wells, the nearest town, is about 10 miles south of the southern boundary of the Topopah Spring quadrangle.
Date: May 1960
Creator: Orkild, Paul P. & Pomeroy, John S.
Partner: UNT Libraries Government Documents Department
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Hazards and Experimental Procedure Evaluation for Studies on the Polymerization and Hydrolysis of Plutonium in Uranyl Nitrate and Nitric Acid Solution at Elevated Temperatures

Description: Because of the danger of precipitation of polymerized Pu/sup 4+/, a research program was initiated for studying the polymerization of Pu/sup 4+/ in the UO/sub 2/(NO/sub 3/)/sub 2/--HNO/sub 3/ system. Th e present state of knowledge concerning the polymerization of Pu/sup 4+/ is reviewed, and the program is discussed in detail, including equipment and procedures. The program is evaluated from the viewpoint of hazards, health physics procedures, and personnel safety. (D.L.C.)
Date: May 1, 1960
Creator: Biggers, R E & Costanzo, D A
Partner: UNT Libraries Government Documents Department
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HAZARDS SUMMARY REPORT FOR THE ARMY PACKAGE POWER REACTOR SM-1, TASK XVII

Description: The SM-1 is described and the various hazards are reviewed. Because of the reactor's location near the nation's capital, containment is of the utmost importance. The maximum energy releace in any possible accident is 7.4 million Btu which is completely contained within a 7/8-in.-thick steel cylindrical shell with hemispherical ends. The vapor container is 60 ft high and 32 ft in diameter and is lined on the iside with 2 ft of reinforced concrete which provides missile protection and is part of the secondary shield. All possible nuclear excursions are reviewed. The energy from any of these is insignificant compared to the stored energy in the water. The maximum credible accident is caused by the reactor running constantly at its maximum power of 10 Mw and through an extremely uniikely sequence of failures, csusing the temperature of the water in the primary and secondary systems to rise to saturation, whereupon a ruptare occurs releasing the stored energy of 7.4 million Btu into the vapor container. If the reactor core melts during the accident, a maximum of 1.5 x lO/sup 7/ c of activity is released into the vapor container. While it is highly improbable for a ruptare of the vapor container to occur except by sabotage or bombing, the hazards to the surrounding area are discussed in the event of such a rupture occurring simultaneously with the maximum credible accident. The SM-1 has been in operation for about 3 years, during which time much operating experience was accumulated and nuclear and system data were obtained through both normal operations and the research and development programs conducted. These data are discussed in relation to their effect on the hazards evaluction of the SM-1 and include core nuclear data, primary and secondary system thermal data, radiological safety, dose rates at various positions …
Date: May 1, 1960
Creator: Rosen, S.S. ed.
Partner: UNT Libraries Government Documents Department
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Hazards Summary Report for the Army Package Power Reactor : SM-1, Task XVII

Description: Preface. This technical report is an updated and revised version of the original SM-1 (APPR-1) Hazards Summary Report. The original report was issued July 1955, almost two years before construction of the SM-1 plant was completed. During that time interval there were numerous design changes. Consequently, the original report does not accurately describe the plant as built. This revision is written after the SM-1 has been operated 3 years. It describes the as-build plant and includes plant modifications and experience obtained during the first 3 years of SM-1 operation.
Date: May 1960
Creator: Rosen, S. S.
Partner: UNT Libraries Government Documents Department
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HEAVY WATER MODERATED POWER REACTORS PROGRESS REPORT, APRIL 1960

Description: Safeguards analyses of the isolated coolant Icops of the HWCTR indicated that accidental loss of ccoling water from the Icop heat exchangers will not lead tc vaper binding of the Icop D/sub 2/O pumps, or to other detrimertal consequences, providing that the reactor is scrammed following the loss of coolant. Several static seals of the types that will be used in the HWCTR exhinited leakage rates that were well below design specifications during cyclic tests at peak conditions of 1500 psi and 260 deg C. From full-scale experiments with lattices of seven-rod clusters of natural uranium metal at various lattice spacings in D/sub 2/0, values for the bucklings, flux distributions, and microscopic parameters of the lattices were measured. (For preceding report see DP-480.) (auth)
Date: May 1, 1960
Creator: Hood, R.R. & Isakoff, L. comps.
Partner: UNT Libraries Government Documents Department
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The High Intensity Food Irradiator: Report 17

Description: From introduction: A thorough investigation, both experimental and theoretical was carried out on plaque type radiators. Five phases of the work are reported: preliminary design analysis, experimental measurements, computational program and analytical studies, followed by a general discussion of source design principles. The results are summarized in graphs and formulae which may be applied to other design requirements.
Date: May 1, 1960
Creator: Curtiss-Wright Corporation
Partner: UNT Libraries Government Documents Department
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Hindrance Factors For Alpha Decay

Description: The theoretical half lives for alpha emissions have been calculated for nearly all of the complex alpha spectra. The spin independent equations of Preston were used for the calculations. The nuclear radius for the even-even nuclei was determined with the assumption that the alpha transition to the ground state is unhindered. For odd mass nuclides the average of the nuclear radii of the adjacent even-even nuclides was used. For odd-odd nuclides the average of the nuclear radii of the adjacent odd mass nuclides of the same atomic number was used.
Date: May 1960
Creator: Michel, Helen V.
Partner: UNT Libraries Government Documents Department
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Irradiation of an Aluminum Alloy-Clad Ceramic Pellet-Fueled Plate

Description: An aluminum-nickel alloy-clad ceramic-fueled plate of the BORAX-IV type was examined destructively after irradiation to a maximum burn-up of 7.800 Mw days per ton in the circulation water loop in the MTR. Irradiation was at 600 psig and 465 deg F. with local boiling in the area of highest neutron flux and a maximum heat flax of 500.000 Btu/(hr) (ft/sub 2/). The specimen performed satisfactorily in spite of several factors which made conditions more severe than those expected to exist in a reactor fueled with this type of element. The corrosion rate of the aluminum-nickel cladding on the element was approximately 10 mils per year. Oniy slight breakup of the ceramic pellet fuel was experienced and there was no evidence that the degree of cracking was influenced to any appreciable extent by exposure to radiation. (auth)
Date: May 1960
Creator: Gavin, A. P.
Partner: UNT Libraries Government Documents Department
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