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Strontium-90 Fueled Thermoelectric Generator Power Source for Five-Watt U.S. Coast Guard Light Buoy. Final Report
The objectives of the SNAP 7A program were to design, manufacture, test, and deliver a five-watt electric generation system for a U. S. Coast Guard 8 x 26E light buoy. The 10-watt Sr/sup 90/ thermoelectric generator, the d-c-to-d-c converter, batteries and the method of installation in the light buoy are describcd. The SNAP 7A generator was fueled with four capsules containing a total of 40,800 curies of Sr/sup 90/ titanate. After fueling and testing, the SNAP 7A electric generating system was installed in the Coast Guard light buoy at Baltimore, Maryland, on December 15, 1961. Operation of the buoy lamp is continuous. (auth)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
GAS-COOLED REACTOR PROGRAM QUARTERLY PROGRESS REPORT FOR PERIOD ENDING SEPTEMBER 30, 1961
Progress is reported on investigations in support of the Experimental Gas-Cooled Reactor, the Pebble-Bed Reactor Experiment, Advanced reactor design and development, test facilities, components, and materials. Topics covered include EGCR physics, EGCR performance analyses, structural investigations, EGCR component and materials development and testing, EGCR experimental facilities, PBRE physics and design studies, fueled-graphite investigations, clad fuel development, design studies of advanced power plants, experimental investigations of heat transfer and fluid flow, development of equipment anmd test facilities. and fabrication studies. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Examination of Uranium-2 w/o Zirconium Experimental Fuel Slugs Irradiated in EBR-I. Final Report-Program 6.1.11
Six groups of U-2 wt% Zr fuel slugs were irradiated in the first core of the EBR-I to burnups of 0.080 to 0.189 at.% at calculated temperatures of 307 to 353 deg C. Two groups of cast specimens were found to be more dimensionally stable than four groups of wrought slugs. Of the wrought slungs, the as quenched group showed less tendency to grow than the three groups which had some annealing after quenching. Specimens at burnups of about 0.189 at.% and at 383 deg C showed the onset of swelling as indicated by density measurements. The hardnesses of these specimens seemed but little affected by radiation, but there was an indication of softening with increasing irradiation temperature. (auth)
Date:
February 1, 1962
Creator:
Murphy, W. F.; Klank, A. C. & Paine, S. H.
Object Type:
Report
System:
The UNT Digital Library
KINETIC EXPERIMENTS ON WATER BOILERS-"A" CORE REPORT-PART II. ANALYSIS OF RESULTS
The status of the analytic portion of the KEWB program at the time of completion of the spherical core experiments is summarized. Three computer programs were developed for use in this analytic effort. The first reassembles and smooths three decades of reactor power data read separately from oscillogram records of reactor excursions. It then computes the logarithmic derivative of the power, energy release, fuel solution temperature, and temperature compensated reactivity. The second program utilizes the space-independent neutron kinetics equations with any number of delayed neutron groups to determine the reactivity in the reactor from the power and its derivative. The third program solves the space-independent kinetics equations for the neutron flux from an input reactivity or initial period. Up to 50 reactivity feedback equations includirg delayed neutrons are provided for in this program. A mathematical model of the reactor investigated extensively was one containing six delayed neutron groups, conventional treatment of temperature reactivity compensation, and void compensation of reactivity induced by radiolytic gas void growth proportional to the product of reactor power and energy release. Partial mathematical solutions to the kinetic equations were derived for reactivity feedback proportional to prompt temperature and void growth according to the product of power …
Date:
February 1, 1962
Creator:
Dunenfeld, M. comp.
Object Type:
Report
System:
The UNT Digital Library
THE DISTORTED-WAVE THEORY OF DIRECT NUCLEAR REACTIONS. I. "ZERO-RANGE" FORMALISM WITHOUT SPIN-ORBIT COUPLING, AND THE CODE SALLY
The distorted-wave theory of direct nuclear reactions is presented in a unified manner, in which the effects of assuming various reaction mechanisms and nuclear models appear only in certain radial form factors. The zero-range approximation is used, and spin-orbit coupling is neglected in the distorted waves. Formulas are given for transition amplitudes, cross sections, and polarizations. A description is given of the IBM-704 computer code SALLY that is based on these formulas. (auth)
Date:
February 1, 1962
Creator:
Bassel, R.H.; Drisko, R.M. & Satchler, G.R.
Object Type:
Report
System:
The UNT Digital Library
HYDROLOGIC AND GEOLOGIC STUDIES FOR PROJECT GNOME. Preliminary Report
Geologic lnformation required to define the pre- and post-shot physical and chemical characteristics of the sait and other rocks affected by the exploslon was gathered. Information on the pre- and post-shot hydrologic condltlons at the site and the surrounding area was also obtalned. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
A Theoretical Study of Destructive Nuclear Bursts in Fast Power Reactors
The calculation of destructive nuclear bursts in fast reactors by an improved Bethe-Tait method, which, for purposes of calculation, neglects propagation of the pressure wave is described. TMen exact numerical calculations for hydrodynamic and neutronic conditions during the power burst are performed in order to assess the importance of this neglect. (auth)
Date:
February 1, 1962
Creator:
Jankus, V. Z.
Object Type:
Report
System:
The UNT Digital Library
SALT PHASE CHLORINATION OF REACTOR FUELS. III. CATALYZED DISSOLUTION OF URANIUM DIOXIDE IN LEAD CHLORIDE-CHLORIDE SYSTEMS
The rapid dissolution of uranium dioxide is described, wherein copper is added to molten lead chloride at 550 deg C. and chlorine is passed through the melt. The integral dissolution of zirconium-clad uranium dioxide fuels is also described. The dissolution rate of uranium dioxide is directly proportional to the concentration of cuprous chloride if an excess of chlorine is used; the value for the rate constant is approximately 100 mg (UO/sub 2/) cm/sup -2/ min/sup -1/ (CuCl molality). The uranium dioxide is converted to water soluble uranyl chloride. The dissolution rate can be controlled by three factors: the copper concentration, the flow rate of chlorine, and the surface area of uranium dioxide. Scoping work indicates that iron chloride may be as effective catalytically as copper chloride for uranium dioxide dissolution in the lead chloride -chlorine system at 550 deg C; thallium chloride is not as effective. (auth)
Date:
February 1, 1962
Creator:
Vander Wall, E. M.; Bauer, D. L. & Hahn, H. T.
Object Type:
Report
System:
The UNT Digital Library
ANALYSES OF THE RADIAL SEPARATOR
A mathematical analysis is presented concerning the performance of the radial or vane type steam separator. The analyses are concerned with the spread of flow on the vanes because of centrifugal force (which requires that the vane height be greater than the nozzle), and with separation of the vapor from the liquid. (J.R.D.)
Date:
February 1, 1962
Creator:
Robbins, C.H.
Object Type:
Report
System:
The UNT Digital Library
SOME CALCULATIONS PERTAINING TO FAST REACTOR SAFETY
A program on fast reactor safety is described in which the theoretical aspects, techniques and accuracy of calculation, and associated safety predictions are considered. One topic deals with the success obtained in analyzing complicated meltdown configurations using two-dimensional, twogroup diffusion theory. Another topic deals with the dependence of energy yield in a fast reactor explosion on the shape of the power distribution in the core. The results of a brief survey are included dealing with the change of dynamic reactor behavior during various startup accidents as a consequence of a reduction in the delayed-neutron fraction. (J.R.D.)
Date:
February 1, 1962
Creator:
O'Shea, D.M.; Okrent, D. & Chaumont, J.M.
Object Type:
Report
System:
The UNT Digital Library
DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Monthly Progress Letter for Month of January 1962
Development work leading to comparison of co-precipitated and mechanically blended UO/sub 2/-- PuO/sub 2? powders is reported in which special emphasis was placed on blending trials, pellet sintering studies, and subsequent evaluation of pellets made with blended material. Homogeneity studies indicate that currently used procedures are unsatisfactory because particle buildup occurs during blending. Powder preparation via the oxalate route was continued along with PuO/sub 2/ moisture pickup studies. Homogeneous precipitation studies on UO/ sub 2/ were continued to determine feasibility of direct preparation of dense UO/ sub 2/-- PuO/sub 2/ feed materials. Plasma-torch-produced PuO/sub 2/ spheres are being evaluated. (J.R.D.)
Date:
February 1, 1962
Creator:
Puechl, K.H.
Object Type:
Report
System:
The UNT Digital Library
THE CHEMISTRY OF NIOBIUM IN PROCESSING OF NUCLEAR FUELS
The chemical and industrial literature and laboratory work concerned with processing of niobium-containing nuclear fuels are reviewed. 57 references. (auth)
Date:
February 1, 1962
Creator:
Gens, T.A.
Object Type:
Report
System:
The UNT Digital Library
RADIOACTIVE FALLOUT FROM NUCLEAR WEAPONS TESTS. Proceedings of a Conference Held in Germantown, Maryland, November 15-17, 1961
Thirty papers are presented reviewing AEC research projects related to fall-out from weapons tests. Reviews of specific related programs by representatives from Canada and the UK are also included. The scope of the conference includes characteristics of fall-out, atmospheric factors affecting deposition, distribution in the environment, and distribution in the food chain and man. Separate abstracts have been prepared for each paper. (C.H.)
Date:
February 1, 1962
Creator:
Klement, A.W. Jr. ed.
Object Type:
Report
System:
The UNT Digital Library
Periodic Characterization of Radioactive Waste Disposal System Effluents. Test Evaluation
>The major nuclides present in the Radioactive Waste Disposal System effluent water and gas were characterized. The major gamma activity in all of the liquid samples analyzed was Co/sup 60/. The ion exchangers and gas stripper removed the major activities in both particulate and filterable form and reduced the gross activity to well within limits for discharge to the environment. Sr/ sup 90/ was not detectable in the reactor plant effluent. The major activity in the Vent Gas Systems was Xe/sup 133/. Mn/sup 54/ activity was present in the Decontamination Room effluents in a ratio of 1 to 3 with Co/sup 60/. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Plutonium Fuel Processing and Fabrication for Fast Ceramic Reactors
>A study was made of the processes available for fabrication of plutonium-containing fuel from a fast ceramic reacter, and for chemical reprocessing of irradiated fuel. Radiations from recycled plutonium are evaluated. Adaptation of conventional glove-box handling procedures to the fabrication of recycle plutonium appears practical. It is concluded that acceptable costs are obtainable using moderate extensions of conventional glove- box fabrication methods and wet processing techniques, provided a significant volume of production is available. The minimum economic scale for the preferred chemical reprocessing method, anion exchange, is about 500 Mw(e) of reactor capacity. The minimum scale of economic operation for the fuel refabrication facility corresponds to three 500 Mw(e) reactors, if only steady-state refueling provides the fabrication load. The minimum volume required falls to one 500 Mw(e) reactor, if the continued growth of capacity provides fabrication volume equal to that for refueling. The chemical reprocessing costs obtained range from 0.27 mills/kwh for 1500 Mw(e) of reactor capacity, to 0.10 mills/kwh for 3000 Mw(e) of capacity. The estimated fuel fabrication cost is l/kg of uranium and plutonium in the core region (excluding axial and radial blankets) or .06/ g of plutonium content, When axial blankets, fabricated in the same rods, are …
Date:
February 1, 1962
Creator:
Zebroski, E.L.; Alter, H.W. & Collins, G.D.
Object Type:
Report
System:
The UNT Digital Library
HAZARD SUMMARY REPORT FOR THE ARGONNE AGN-201 REACTOR
Essential characteristics of the AGN-201 reactor facility are described. Information on personnel responsibilities and requirements is included along with an evaluation of reactor safety. (J.R.D.)
Date:
February 1, 1962
Creator:
Ruzich, K.C. & Sturm, W.J. eds.
Object Type:
Report
System:
The UNT Digital Library
The Uranium-Rich End of the Uranium-Zirconium System. Final Report- Metallurgy Program 3.1.3.
The uranium-rich end of the uranium-zirconium alloy system was reinvestigated. The solubilities of zirconium in alpha and beta uranium were found to be 0.21 wt% at 662 deg C and 0.41 wt% at 693 deg C, respectively. The monotectoid decomposition of gamma /sub 1/ at 693 deg C and the eutectoid decomposition of t 662 deg C were confirmed. For alloys containing less than 150 ppm of oxygen by weight, the gamma /sub 1/ plus gamma /sub 2/ phase region boundaries were located at 4.5 and 22.0 wt% zirconium at the monotectoid temperature. Data are given which indicate that oxygen concentrations ranging from 160 to 355 ppm by weight have a marked effect on phase relations in the area of the gamma /sub 1/ plus gamma /sub 2/ phase region. (auth)
Date:
February 1, 1962
Creator:
Zegler, S. T.
Object Type:
Report
System:
The UNT Digital Library
Progress Report on Neutron Radiography
BS> The potential advantages of neutron radiography as an inspection method are discussed along with a historical review and discussions of neutron sources and detectors. The results of the current investigation of neutron-image detectors are discussed in regard to photographic speed, relative neutron-gamma response, and resolution comparisons. Two neutron-image detecting methods are discussed. In one, the direct-exposure method, both the converter screens and the film are exposed to the neutron beam together. The other method, the transfer method, makes use of a radioactive, image-carrying screen, which is transferred to photographic film after the neutron exposure is completed. The direct-exposure method results in increased speed, but has the disadvantages that the film also responds to any gamma radiation in the imaging beam and that, in most cases, improved image resolution can be obtained with the transfer method. Reference is given to several application possibilities. (auth)
Date:
February 1, 1962
Creator:
Berger, H. & McGonnagle, W.J.
Object Type:
Report
System:
The UNT Digital Library
Nuclear Fuel Research Fuel Cycle Development Program Quarterly Progress Report, October 1-December 31, 1961
The irradiation testing program involving enriched UO/sub 2/ capsules fabricated by low-temperature sintering was continued. Flux monitor data indicate an average burnup level in the capsule of 5800 Mwd/t compared with 9100 Mwd/t calculated. Preparation of 5 different high-purity UC irradiation test specimens of 12% enrichment is reported. Bescriptions of UC preparation procedures by the propane reaction and by skull melting are included. (J.R.D.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
TRANSPORT OF NOBLE GASES IN GRAPHITES. Progress Report for the Period January 31, 1961 to January 31, 1962
A model was derived for the diffusion of gases in porous media in the absence of temperature and pressure gradients, in which portions of the medium are visualized as a collection of uniformly distributed dust'' particles (giant molecules) which are constrained to be stationary. Thus, it was possible to derive all the desired results from rigorous diffusion equations for multi- component mixtures. The results apply over the entire pressure range from the Knudsen region to the normal diffusion region. This model permits a satisfactory derivation of the fact that, at all pressures, the flux ratio of two counter- diffusion gases is (m/sub 2/mi/sub 1/)/sup 1/2/ in porous media under steady state and uniformpressure conditions. The effect of non-zero pressure gradients on the diffusion equations is to introduce into the fundamental kinetic theory equations both a pressure diffusion term and an external force term. There is a considerable cancellation of terms, and the final diffusion equation has the same form as in the uniform pressure case. No additional parameters beyond those necessary to define a diffusing system at uniform pressure are thus required to compute the diffusion rates when pressure gradients are present. A complete solution requires also a forced flow …
Date:
February 1, 1962
Creator:
Truitt, J.; Smith, N. V.; Watson, G. M.; Evans, R. B., Ill, & Mason, E. A.
Object Type:
Report
System:
The UNT Digital Library
PROCESSING OF POWER REACTOR FUELS SIXTEENTH QUARTERLY PROGRESS REPORT, JULY 1 TO OCTOBER 1, 1961
In the semiworks dissolver, average currents were about 2000 amp during dissolution of stainless steel and about 1200 amp during dissolution of Zircaloy. These currents were obtained when the current was transferred to a fuel assembly by direct contact with the columbium basket. High dissolution rates are anticipated when the dissolver is modified so that the electrolyte makes the only electrical contact with the charge. Laboratory experiments were made to determine the important characteristics of this type of dissolver. High corrosion rates of stainless steel equipment by boiling nitric acid-stainless steel solutions (under evaporator conditions) is caused by Cr/sup +6/ (auth)
Date:
February 1, 1962
Creator:
Rust, F.G. comp.
Object Type:
Report
System:
The UNT Digital Library
Reactor Development Program Progress Report, February 1962
Progress is reported on EBWR, BORAX-V, and development of liquid metal cooled reactors including EBR-I and -II. Developments in general reactor technology are reported in sections on physics, fuels, components, materials, engineering, and chemical separations. Other research and development is reported in advanced systems and nuclear ssfety. (J.R.D.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Salt Bath Heat Transfer Rates for Uranium Plate
Experiment to study temperature conditions during a five-minute heating cycle in the annealing of uranium sheets used as fissile material.
Date:
February 1, 1962
Creator:
Burditt, R. B.
Object Type:
Report
System:
The UNT Digital Library
[Photograph 2012.201.B0246.0428]
Photograph used for a story in the Oklahoma Times newspaper. Caption: "Old Mr. Ground Hog apparently is losing his place in the sun as a forecaster for weather for the six weeks following his special day. Seeing shadow, he returns to burrow."
Date:
February 1, 1962
Creator:
Lucas, Jim
Object Type:
Photograph
System:
The Gateway to Oklahoma History