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  Partner: UNT Libraries Government Documents Department
 Resource Type: Report
 Decade: 1950-1959
CHEMICAL TECHNOLOGY DIVISION, UNIT OPERATIONS SECTION MONTHLY PROGRESS REPORT FOR AUGUST 1959

CHEMICAL TECHNOLOGY DIVISION, UNIT OPERATIONS SECTION MONTHLY PROGRESS REPORT FOR AUGUST 1959

Date: December 31, 1959
Creator: Bresee, J.C.; Haas, P.A.; Horton, R.W.; Watson, C.D. & Whatley, M.E.
Description: The concentration gradients of uranyl ion in aqueous and organic solutions were analyzed by taking a macro photograph of the desired gradient by monochromatic (436 m mu ) light transmitted by the solution normal to the gradient in an appropriate diffusion cell. Two Druhm runs were terminated due to malfunction of the sodium metering system and the third run was terminated when the UF/sub 6/ nozzle ruptured. Calculations of particle temperature versus time relations for the flame denitration-calcination method of preparing metallic oxide from nitrate solutions indicate that the times required for heat transfer are controlled by the rate of radiant heat transfer to particle surfaces instead of by conductive heat transfer within the particles. A completed experimental study indicated that electrolysis in a cell with a mercury cathode and a platinum anode is a practical process for removing nickel from HRT fuel solution. The apparent diffusion coefficient of uranium loading on Dowex 21K was shown to be directly related to the resin size. An explosion of sufficient violence to blow apart the Pyrex pipe dissolver occurred during the fifth Darex dissolution of simulated SRE fuel probably from a rapid gas phase reaction between hydrogen and oxidizing gases such as ...
Contributing Partner: UNT Libraries Government Documents Department
SER Temperature Coefficient

SER Temperature Coefficient

Date: December 31, 1959
Creator: Johnson, J.L.
Description: Experimentally determine the overall isothermal temperature coefficient of the SER up to the design operating temperatures.
Contributing Partner: UNT Libraries Government Documents Department
THERMONUCLEAR PROJECT SEMIANNUAL REPORT FOR PERIOD ENDING JULY 31, 1959

THERMONUCLEAR PROJECT SEMIANNUAL REPORT FOR PERIOD ENDING JULY 31, 1959

Date: December 30, 1959
Creator: unknown
Description: None
Contributing Partner: UNT Libraries Government Documents Department
EFFECT OF ADDITIONS TO ZIRCALOY ON HYDROGEN PICKUP DURING AQUEOUS CORROSION

EFFECT OF ADDITIONS TO ZIRCALOY ON HYDROGEN PICKUP DURING AQUEOUS CORROSION

Date: December 29, 1959
Creator: Berry, W.E.; White, E.L. & Fink, F.W.
Description: An investigation was conducted into the possibility of alloy additions to Zircaloy-2 to diminish hydrogen absorption during aqueous corrosion. The nickel in Zircaloy-2 is believed to be the major constituent responsible for the relatively high hydrogen absorption. Additions of up to 0.5 wt.% antimony, arsenic, bismuth, or tellurium were selected on the basis of their known ability to poison the catalytic effects of nickel in hydrogenation reactions of other systems. Results of tests conducted for a total of 224 days in 600 and 680 deg F water and 750 deg F steam revealed no decrease in hydrogen absorption in modified Zircaloy-2 containing the aforementioned alloy additions. Hydrogen absorption increased when these alloying elements were present in the range of 0.1 to 0.2 wt.%. Corrosion resistance also decreased with alloy additions in these ranges. A 2-atm. partial pressure of hydrogen in the steam or above the water did not affect hydrogen absorption in the alloys appreciably. The hydrogen partial pressure did not affect time to transition in corrosion rates, but did appear to produce higher weight gains than degassed water. (auth)
Contributing Partner: UNT Libraries Government Documents Department
PROTOTYPE FREEZE TRAP TEST

PROTOTYPE FREEZE TRAP TEST

Date: December 29, 1959
Creator: Cygan, R.
Description: A performance evaluation was made of a prototype liquid cooled freeze trap with sodium at 350 and 1000 deg F. The sodium freeze-off function was adequate for all test conditions encountered. The freeze-off occurred satisfactorily with the larger clearance provided by a test modification to provide 0.030 eccentricity to the rotating shaft. Turning the freeze-trap handle was successful in opening the unit for gas venting when 350 deg F sodium was used. For a seal formed with 1000 deg F sodium, 16 turns of the trap handle gave no measurable gas venting at pressures up to 30 psi. Melting out the seal opened the vent satisfactorily. All the major problems encountered during the test were mechanical and associated with the rotating feature of the unit. (M.C.G.)
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CONTROL AND DYNAMICS PERFORMANCE OF A SODIUM COOLED REACTOR POWER SYSTEM. Report No. 171

CONTROL AND DYNAMICS PERFORMANCE OF A SODIUM COOLED REACTOR POWER SYSTEM. Report No. 171

Date: December 28, 1959
Creator: Hansen, P.D. & Eaton, J.H.
Description: None
Contributing Partner: UNT Libraries Government Documents Department
INTERACTION OF TWO METAL SLABS OF PLUTONIUM IN PLEXIGLAS

INTERACTION OF TWO METAL SLABS OF PLUTONIUM IN PLEXIGLAS

Date: December 28, 1959
Creator: Schuske, C.L.; Goodwin, A. Jr.; Bidinger, G.H. & Smith, D.F.
Description: Neutron multiplication measurements were performed on two identical finite Pu-metal slab assemblies separated and reflected by plexiglas. (auth)
Contributing Partner: UNT Libraries Government Documents Department
SM-2 REACTOR CORE AND VESSEL REVIEW REPORT FOR AUGUST 25, 1959 TO DECEMBER 14, 1959

SM-2 REACTOR CORE AND VESSEL REVIEW REPORT FOR AUGUST 25, 1959 TO DECEMBER 14, 1959

Date: December 24, 1959
Creator: unknown
Description: The most adverse power distribution was revised based on a comparison of PDQ calculations and measurements made during the SM-2 flexible experiments. A review of the basic nuclear data and calculational models employed in the SM-2 nuclear analysis was rnade. A comparison between initial reactivily, hot-to-cold reactivity change, and xenon reactivity with experiment was rnade. Based on a revised power distribution, the core flow requirement was reestimated to be 7800 gpm. Tentative designs of the core support and fuel element structure were prepared and evaluated for pressure drop and flow distribu-tion. The ETR and MTR irradiation programs are suramarized. The TIG process for welding elements is discussed. Specimens of Eu/sub 2/O/sub 3/ dispersions in stainless steel were autoclave tested. Static deflection messurements indicated that a fuel element with cold rolled plates will have a deflection aproximately 18% lower than annealed plates. measurement of plate collapse on two elements indicated possible collapse in the range 140 to 164% of rated flow. Flow distribution and pressure drop tests were made for several core support structure configurations. Mockup experiments on the SM-2 initial cold, clean and SM-2 mid-life cores were completed. Limited power distribution and flux distributions were performed in the clean mockup. ...
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PRESSURE DROP MEASUREMENTS ACROSS A MOCKUP OF AN SRE 7-ROD FUEL ELEMENT WITH AN ORIFICE PLATE AT THE TOP

PRESSURE DROP MEASUREMENTS ACROSS A MOCKUP OF AN SRE 7-ROD FUEL ELEMENT WITH AN ORIFICE PLATE AT THE TOP

Date: December 23, 1959
Creator: Begley, R.J.
Description: None
Contributing Partner: UNT Libraries Government Documents Department
MODIFIED ZIRFLEX PROCESS FOR DISSOLUTION OF ZIRCONIUM-AND NIOBIUM-BEARING NUCLEAR FUELS IN AQUEOUS FLUORIDE SOLUTIONS: LABORATORY DEVELOPMENT

MODIFIED ZIRFLEX PROCESS FOR DISSOLUTION OF ZIRCONIUM-AND NIOBIUM-BEARING NUCLEAR FUELS IN AQUEOUS FLUORIDE SOLUTIONS: LABORATORY DEVELOPMENT

Date: December 22, 1959
Creator: Gens, T.A. & Baird, F.G.
Description: Modified Zirflex process flowsheets were developed for recovering uranium from the newer power reactor fuel alloys after discharge from the reactor. The STR (1% U97% Zr-2% Sn) and EBWR Core-1 (93.5% U-5% Zr-1.5% Nb clad in Zircaloy-2) fuels are used as examples of low- and high-uranium fuels, respectively. A dissolvent of 6 M NH/sub 4/F yields a solution of zirconium and a precipitate of ammonium uranous fluoride. In one process, ammonium hydroxide is added to produce insoluble hydrous oxides of uranium, zirconium and niobium. The NH/sub 4/F-NH/sub 4/OH supernatant is removed by filtration, partially evaporated, and recycled as dissolvent. The uranium and zirconium oxides are dissolved in nitric acid to yield a solvent extraction feed solution of low fluoride content. In an alternative process nitric acid and aluminum nitrate are added to the ammonium fluoride fuel solution to oxidize U(IV) to soluble V(VI) and prepare a stable solution suitable for solvent extraction. Chromic acid is also added in the case of the STR fuel. In a variation of this flowsheet for the EBWR fuel, only- enough 6 M NH/sub 4/F is added to dissolve the cladding. Nitric acid and aluminum nitrite are then added io dissolve the core. Insoluble niobic ...
Contributing Partner: UNT Libraries Government Documents Department
SURVEY OF THE RADIATION LEVELS IN THE CONTAINMENT VESSEL OF THE ENRICO FERMI ATOMIC POWER PLANT. PART V. GAMMA RADIATION LEVELS ON THE OPERATING FLOOR OF THE CONTAINMENT BUILDING. a. LEVELS ABOVE THE EQUIPMENT COMPARTMENT. Technical Memorandum No. 16

SURVEY OF THE RADIATION LEVELS IN THE CONTAINMENT VESSEL OF THE ENRICO FERMI ATOMIC POWER PLANT. PART V. GAMMA RADIATION LEVELS ON THE OPERATING FLOOR OF THE CONTAINMENT BUILDING. a. LEVELS ABOVE THE EQUIPMENT COMPARTMENT. Technical Memorandum No. 16

Date: December 22, 1959
Creator: Chaltron, W.F. & Hungerford, H.E.
Description: The results are presented of a survey of calculated gamma-ray levels at many points on the surface of the operating floor of the containment building for the Enrico Fermi reactor. That portion of the floor surveyed lies directly above the equipment compartment. The calculations were made with the aid of an IBM-650 electronic computer. The main source of radioactivity which gives rise to gamma radiation above the floor is the radioactive sodium-24 in the primary coolant system. This system was considered to be completely filled with sodium, and activated to an equilibrium activity of 0.05 curies/cc, which corresponds to infinite reactor operation at 500 megawatts power. No fission product contamination was considered for these calculations. The operating floor is 5 feet thick and of concrete and steel. The results of the survey indicate that above the equipment compartment the surface dose on the operating floor will in no case exceed 0.9 mr/hr at the expected full operating power of 430 megawatts. Included as appendices are derivations and methods of corrections from one set of concrete and steel thicknesses to another. (auth)
Contributing Partner: UNT Libraries Government Documents Department
Report of the Forty-Fourth National Conference on Weights and Measures, 1959

Report of the Forty-Fourth National Conference on Weights and Measures, 1959

Date: December 18, 1959
Creator: United States. National Bureau of Standards.
Description: Report of the annual conference on weights and measures, hosted by the U.S. National Bureau of Standards in Washington D.C. It includes conference proceedings, a list of attendees, information about committees and officers, and other reports or commentaries discussed at the meetings.
Contributing Partner: UNT Libraries Government Documents Department
PRELIMINARY STUDIES OF SCAVENGING SYSTEMS RELATED TO RADIOACTIVE FALLOUT. Letter Report No. 10 for October 1 to December 1, 1959

PRELIMINARY STUDIES OF SCAVENGING SYSTEMS RELATED TO RADIOACTIVE FALLOUT. Letter Report No. 10 for October 1 to December 1, 1959

Date: December 18, 1959
Creator: Stockham, J. & Rosinski, J.
Description: Progress is reported in the development of scavenging systems for the collection of fall-out. Data are included from tests of two cyclone separators for the collection of air samples. Results are included from laboratory studies on the scavenging of aerosol particles by evaporating and condensing water droplets. (C.H.)
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METALLURGY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 1, 1959

METALLURGY DIVISION ANNUAL PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 1, 1959

Date: December 16, 1959
Creator: unknown
Description: 7 = 9 9 9 9 7 7 7 = 9 9 9 95 : > @ 9 ; 5 8 @ = K : . ighpurity Nb deformed by impact or slow compression at - 196 deg C. An apparent phase transformation was detected in high- purity Ga deformed at 4.2 deg K. The specific heat of the group IV-A metals and alloys of Zr-In and Zr-Sn were measured from 1.2 to 4.5 deg K. In the Zr-rich portion of the Zr-Ga phase diagram, the alpha / beta phase boundaries of Zr are depressed by additions of Ga and the beta phase decomposes by a eutectoid reaction. The Cd pressures of alpha - and beta - Zr alloys containing 1 to 11% Cd were measured between 1090 and 1325 deg K. Crystal structures of several unreported transition-metal fluorides, rare-earth hydrides and nitrides were determined. Progress in the study of phase transitions in beta -quenched Zr-Nb alloys aged below the eutectoid temperature is reported. A high-temperature investigation of the order-disorder phase transition of a Cu31 at.% Au alloy has revealed an intermediate periodic antiphase condition. A previously described x- raydiffraction technique for the measurement of the thickness and strain of ...
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Metallurgy Division Annual Progress Report, September 1, 1959

Metallurgy Division Annual Progress Report, September 1, 1959

Date: December 16, 1959
Creator: Oak Ridge National Laboratory. Metallurgy Division.
Description: Report documenting ongoing research and development undertaken by the Metallurgy Division of the Oak Ridge National Laboratory.
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THE EARLY ANTIPROTON WORK [Nobel Lecture]

THE EARLY ANTIPROTON WORK [Nobel Lecture]

Date: December 15, 1959
Creator: Chamberlain, O.
Description: Early work on the antiproton, particularly that part which led to the first paper on the subject, is described. Conclusions that can be drawn purely from the existence of the antiproton are discussed. (W.D.M.)
Contributing Partner: UNT Libraries Government Documents Department
Hazards Analysis of the Organic Moderated Reactor Experiment

Hazards Analysis of the Organic Moderated Reactor Experiment

Date: December 15, 1959
Creator: Williams, R. O. & Allen, W. O.
Description: Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.
Contributing Partner: UNT Libraries Government Documents Department
HRT PROCESS FLOWSHEETS--REVISED EDITION

HRT PROCESS FLOWSHEETS--REVISED EDITION

Date: December 15, 1959
Creator: Robertson, R C & Jones, J E
Description: Revised HRT flowsheets are presented. These revisions cover such items as relocation of freezer units on the lines, corrections to the numbering of lines, valves or instruments, and the addition of a few lines in the service areas. The waste and vent system flowsheet was redrawn as two sheets. (C.J.G.)
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IONIUM (THORIUM-230) FOR RADIOISOTOPE PREPARATION--STATUS REPORT

IONIUM (THORIUM-230) FOR RADIOISOTOPE PREPARATION--STATUS REPORT

Date: December 15, 1959
Creator: Coppinger, E.A. & Rohrmann, C.A.
Description: The general prospects of several radioisotopes are reviewed; the special properties of U/sup 232/ and Th/sup 228/ are poi nted out; and ionium (Th/sup 230/ ) and protactinium target materials are discussed from the sthndpoint of availability and chemical separations processes required for the preparation of U/ sup 232/ and Th/sup 228/. Outlines are given for potential schem es for the separation of U/sup 232/ and Th/sup 228/ from uranium milling pr ocess waste streams and from the irradiation products of Th/sup 230/--Th/sup 232/ mixtures. The high heat generating rates of these potent alpha emitters make them especially suitable for primary consideration as heat sources for small thermoelectric generators. The exceptionally high alpha activity suggests their use in special neutron sources as Ra-Be sources, and they may have sufficiently high neutron generating rates to be in contention with some of the smaller research reactors and experimental neutron producers. (B.O.G.)
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IRRADIATION EFFECTS ON THE SURFACE REACTIONS OF METALS. Summary Report for October 1, 1958 to November 1, 1959

IRRADIATION EFFECTS ON THE SURFACE REACTIONS OF METALS. Summary Report for October 1, 1958 to November 1, 1959

Date: December 15, 1959
Creator: Carpenter, F.D. & White, J.L.
Description: Weight increases during the oxidation of irradiated foils of pure copper were greater than for unirraaiated specimens. Enhanced reactivity appeared to be strongest in the thin-film region up to about 5 mu g/cm/sub 2/. Oxide film (Cu/ sub 2/O) thickness for both irradiated and unirradiated specimens was approximately 1200 A. Radiation did not affect the reduction of Cu/sub 2/O during the induction period (period in which the reduction proceeds very slowly or not at all). In later stages of the reduction process, a serious lack of reproducibility was observed. Radiation effects on films of Cu/sub 2/O formed by prior oxidation of the copper substrate decreased the kinetics of secondary oxidation. The secondary oxidation curve exhibited a large gap at the point of interrnption for irradiation. The development of an automatic recording microbalance of high sensitivity and a furnace for studies in reactor radiation fields is reported. Measurements were made of the electrode potentials of irradiated (5.5 x 10/sup 19/ neutrons cm/sup -2/) copper, aluminum, magnesium, and zirconium. Cell potentials were found to be dominated by the oxide films formed on the electrode surfaces. The results indicate that radiation does affect the local anode reaction potential. No significant difference between the ...
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Power Flattening in Sodium Graphite Reactors by Spatial Variation of Moderator Properties

Power Flattening in Sodium Graphite Reactors by Spatial Variation of Moderator Properties

Date: December 15, 1959
Creator: Connolly, T. J.
Description: Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.
Contributing Partner: UNT Libraries Government Documents Department
DEVELOPMENT OF METALLIC URANIUM FUEL ELEMENTS OF IMPROVED IRRADIATION STABILITY. Semiannual Progress Report No. 1

DEVELOPMENT OF METALLIC URANIUM FUEL ELEMENTS OF IMPROVED IRRADIATION STABILITY. Semiannual Progress Report No. 1

Date: December 14, 1959
Creator: unknown
Description: The feasibility of improving swelling resistance in metallic uranium by increasing the dislocation density is under investigation. Increasing the density of dislocations is expected to increase the number of sites at which fission product gas atoms are "pinned," increase the number of gas bubble nuclei, and increase mechanical strength. Dislocations are introduced by a treatment which involves deformation of metastable beta or gamma phase in uranium-rich alloys, followed by transformation. Activity has been concerned primarily with selection of uraniurnbase alloys for initial evaluation, procurement of materials, installation of specialized equipment, and development of techniques. A number of small ingots were produced, and screening tests were carried out on several analyses. These preliminary experiments were designed to show the deformation characteristics of the alloys in the metastable state as a function of rolling temperature and time. Initial results on gamma-stabilized binary alloys containing 2, 3.5, 5, and 7 wt. % Mo showed that large reductions are possible in 5 and 7 wt. % Mo alloys at450 deg C. The more dilute alloys are relatively difficult to roll, although small reductions were achieved. Aging experiments on deformed and undeformed 5 wt. % Mo alloys indicate that deformation accelerates the aging process and ...
Contributing Partner: UNT Libraries Government Documents Department
DESIGN STUDIES ON CESIUM-137 AS A SOURCE FOR HIGH LEVEL GAMMA IRRADIATORS. Quarterly Progress Report No. 1 covering the Period June 1 to August 31, 1959

DESIGN STUDIES ON CESIUM-137 AS A SOURCE FOR HIGH LEVEL GAMMA IRRADIATORS. Quarterly Progress Report No. 1 covering the Period June 1 to August 31, 1959

Date: December 11, 1959
Creator: Voyvodic, L. & Stone, C.A.
Description: A study was made of radiation physics problems involved in the design of high-level cesium-137 gamma sources. The radiation properties of cesium-137 sources are reviewed and design and dosimetry problems are discussed. The economics, efficiency, and dose distribution for material undergoing process irradiation were calculated. A comparison of cesium-137 with cobalt-60 gamma sources indicated that in the case of irradiators specifically designed for high efficiency of useful energy conversion, the performance of cesium-137 source material should be at least comparable to the performance of cobalt-60 source material. (C.H.)
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Health chemistry design recommendations for enclosed firing facility

Health chemistry design recommendations for enclosed firing facility

Date: December 11, 1959
Creator: Lindeken, C.L.
Description: No abstract available.
Contributing Partner: UNT Libraries Government Documents Department
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