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1A HEAT EXCHANGER LEAK TEST. CORE I, SEED 2. Test Evaluation. Section 2

Description: An investigation was conducted to determine which tubes of the 1A loop heat exchanger are leaking. Air pressure and probing tests are inconclusive and cannot be used to verify chemical sampling. (J.R.D.)
Date: July 24, 1961
Partner: UNT Libraries Government Documents Department
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Applied Health Physics Annual Report for 1960

Description: Although there were the ususl fluctuations in background at certain of the monitoring stations, the contamination levels recorded do not differ significantly from those of the previous year except that there appears to be a slight trend downward. Data are tabulated. This downward trend can be attributed to a curtailment of operations at the Laboratory, the gradual implementation of the containment program, and a curtailment in world-wide weapons testing. Tall-out data are included. Two personnel exposures were recorded which have been reported elsewhere. One emPloyee received a relatively high exposure to the left hand which consisted primarily of soft radiation. A second employee apparently has accumulated a sizeable fraction of a permissible body burden of Pu/sup 239/. The number of unusual occurrences increased over the previous year. However, in general, these events posed only routine problems and it is probable that the noted increase in such occurrences is due primarily to a more complete reporting system which was inaugurated early in the year. (auth)
Date: July 27, 1961
Partner: UNT Libraries Government Documents Department
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Aqueous Processes for Dissolution of Uranium-Molybdenum Alloy Reactor Fuel Elements

Description: Methods for dissolving unirradiated uranium-molybdenum alloy reactcr fuels in nitric acid, nitric acid--ferric nitrate, and nitric acid-- phosphoric acid solutions were studied on a laboratory scale. Flowsheets based on the results propose dissolution of alloys containing 3% molybdenum in boiling 6 M HNO/ sub 3/ to yield stsble solutions that are 0.6 M in uranium and 3 to 4 M in nitric acid. The uranium can then be easily decontaminated and recovered in a conventional Purex-type tributyl phosphate solvent extraction process. Alloys containing 10% molybdenum would be dissolved in boiling 11 M HNO/sub 3/, allowing molybdic oxide to precipitate. The molybdic oxide, which carries 5-10% of the uranium, is removed by centrifugation and the acidity of the supernatant solution adjusted tc allow recovery of the uranium by Purex-type solvent extraction procedures. The uranium carried by the molybdic oxide is recovered after the MoO/ sub 3/ is dissolved in warm 5 M NaOH. Less than 0.1% of the uranium is solubilized during the caustic dissolution. Alternative methods investigated involve dissolution in nitric acid containing 0.5 to 1 M ferric nitrate to complex the molybdenum. These techniques lead to undesirably large volumes of high-level solvent extraction waste solutions. Phosphate ion is also effective in complexing molybdenum; however, its use in the dissolvent would be purposeless since it must be complexed with iron during solvent extraction. Rates of reaction of the various alloys and the solubility of molybdic oxide were determined in nitric acid, nitric acid-- ferric nitrate, and nitric acid-- phosphonic acid solutions. (auth)
Date: July 14, 1961
Creator: Ferris, L. M.
Partner: UNT Libraries Government Documents Department
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Barytes Concrete for Radiation Shielding: Mix Criteria and Attenuation Characteristics

Description: Concrete mix design criteria, based on existing theories of proportioning and specifically oriented toward the solution of radiation shielding problems, were developed. Effects of aggregate gradation, cement-to- aggregate ratio, and water content were examined. A barytes concrete, designed according to these criteria, was thoroughly investigated in the Lid Tank Shielding Facility. Relative effectivenesses of dry aggregates, aggregates plus cement, and cured concrete were compared through thermal-neutron flux, fast- neutron dose, and gamma-ray dose measurements behind slab configurations. Attenuation was measured for the aggregate, the aggregate plus cement, and for the barytes concrete. Comparison with attenuations calculated on the basis of removal cross sections for the measured chemical compositions showed satisfactory agreement. (auth)
Date: July 25, 1961
Creator: Grantham, W.J. Jr.
Partner: UNT Libraries Government Documents Department
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Capacities of Stacks in Sanitary Drainage Systems for Buildings

Description: "Some of the important results obtained in investigations of capacities of plumbing stacks in test systems at the National Bureau of Standards and elsewhere are discussed. Data are shown from experiments on the flow of water and air in such systems, and analyses of certain flow phenomena are given. Methods are shown for applying the results of research in hydraulics and pneumatics to the preparation of loading tables (for drainage and vent stacks) suitable for use in plumbing codes" (p. 1).
Date: July 3, 1961
Creator: Wyly, Robert S. & Eaton, Herbert N.
Partner: UNT Libraries Government Documents Department
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CHARACTERIZATION OF UO$sup 2$ POWDERS. Progress Report No. 8, May and June 1961

Description: Correlation of a number of physical and chemical properties of 12 different UO/sub 2/ powders was continued. The UO/sub 2/ powders were studied by means of infrared absorption measurements, oxidation temperatures as determined by hot stage microscopy techniques, and B. E. T. surface area measurements. Additional pellets were prepared to study ceramic performance. (M.C.G.)
Date: July 19, 1961
Creator: Carpenter, J.F.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department Monthly Report: June 1961

Description: This report, for June 1961 from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; employee relations; weapons manufacturing operation; and safety and security.
Date: July 21, 1961
Creator: Hanford Atomic Products Operation. Chemical Processing Department.
Partner: UNT Libraries Government Documents Department
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The Closest Packing of Spheres (A Unifying Basis for Crystal Structures)

Description: "An intuitive approach to the understanding of crystal structures is presented in terms of the concept of the closest packing of spheres. The qualitative features of the concept are sorted out and correlated by successively treating single, double, triple, and multiple layered arrays of closest packed spheres" (p. ix).
Date: July 30, 1961
Creator: Gehman, William G.
Partner: UNT Libraries Government Documents Department
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Conceptual Design of an In-Pile Package Loop for Fast Reactor Fuel Testing

Description: Report issued by the APDA over a design study conducted on an "in-pile package loop for use as a reactor fuel test facility" (p. 5). The results are presented and discussed. This report includes tables, and illustrations.
Date: July 28, 1961
Creator: Blessing, W. G.; Balsbaugh, R. R.; Bloomfield, D. E.; Busch, J. S.; Hennig, R. J.; Jens, W. H. et al.
Partner: UNT Libraries Government Documents Department
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THE CORROSION OF ALUMINUM ALLOYS IN THE OAK RIDGE RESEARCH REACTOR

Description: A corrosion testing program designed to estimate the potential service life of aluminum alloys used in the construction of the Oak Ridge Research Reactor (ORR) cooling systems has been in progress for over two years. The five alloys (1100, 3003, 5052, 5154, and 6061) used to the greatest extent in the reactor exhibited continuously decreasing corrosion rates since the first 500-hr inspection. Samples exposed in the core-cooling loop have shown a decrease in corrosion rate from a 2.6 mpy maximum for one group during the first 500 hr to an over-all average of less than 0.1 mpy for another group after a full year in test, with the maximum metal loss less than 0.1 mils. Results indicate that with suitable water treatment the aluminum alloys used in the ORR may be expected to give satisfactory performance for many years. Based on the generalized corrosion rates alone, 40 to 50 years of service life may be expected. However, since occasional localized corrosion has been observed (rarely), minor repairs will almost certainly be required before that time. (auth)
Date: July 1, 1961
Creator: Neumann, P.D.
Partner: UNT Libraries Government Documents Department
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Corrosion Studies of Ternary Zirconium Alloys in High-Temperature Water and Steam

Description: The alloying of zirconium to improve corrosion resistance has an empirical basis, and satisfactory explanations for the alloying effects are not available. A theory of compensating valencies in the corrosion oxide is proposed, in which cations of lower and higher valence than zirconium (+4) are present in ratios such that electrostatic neutrality is ensured. An example is an alloy containing equimolar amounts of scandium (+3) and niobium (+5). A number of zirconium alloys were prepared in which scandium or yttrium were paired with elements capable of a +5 or +6 valence. The ternary alloys containing scanadium were superior to the alloys combining yttrium. The alloys containing scandium plus molybdenum, tantalum, or tungsten had relatively long lifetimes in steam at 540 deg C and 600 psi as compared with other alloy combinations, including Zircaloy-2. A quenched alloy containing 0.025 wt% Sc and 0.053 at.% Mlo, that is, 0.05 mol.% of each additive, corroded approximately according to a cubic law up to 758 hr, at which potnt the rate suddenly increased in a manner suggesting hydrogen damage. Examination of the oxide film from alloys containing scandium and molybdenum showed only monoclinic ZrO/sub 2/. It is believed that stabilization of this form of ZrO/sub 2/ instead of the cubic or tetragonal forms is a factor in promoting corrosion resistance. In this way the protective character of the film can be improved independently of the addition of cathodes. (auth)
Date: July 1, 1961
Creator: Misch, R.D. & Van Drunen, C.
Partner: UNT Libraries Government Documents Department
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CURRENT STATUS OF THE AC IONIZATION CHAMBER

Description: ABS>The design concept of an a-c ionization chamber and its supporting electronics is described. Several designs are possible and the sensors can be tailored to specific requirements when necessary. Mode of operation, signal voltage development, and switching frequency are discussed. High-sensitivity operation is described. Requirements for high-temperature, power-level operation are outlined. (M.C.G.)
Date: July 1, 1961
Creator: Rusch, G.K.
Partner: UNT Libraries Government Documents Department
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A DESCRIPTION OF INTEGRAL PHYSICS DATA FOR FAST REACTOR DESIGN

Description: Integral physics data for fast reactor design are discussed. The measurements needed include those of critical mass, shape factor, detector ratios, neutron spectra, material replacement experiments, reflector savings, neutron lifetime, Rossi- alpha , and similar quantities. Topics covered include Pu- and U/sup 233/-fueled systems, highly enriched U/sup 235/ systems in optimum geometry, uranium cores of various enrichments and dilutions, extreme geometry critical experiments, specific reactor systems, core mockup inhomogeneities, spectral studies and detector ratios, uranium equilibrium spectrum data, materialreplacement measurements, fast reactor dynamics, and suggested future experiments and experimental programs. (M.C.G.)
Date: July 1, 1961
Creator: Loewenstein, W.B. & Meneghetti, D.
Partner: UNT Libraries Government Documents Department
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Design Criteria for Steel in Nuclear Reactors

Description: S>Criteria for stress analysis and structural design with steel for the critical components of nuclear plants are presented. An effort was made to integrate the effects on the strength of steel of the coexisting phenomena, such as mechanical and thermal loads, stress cycling and fatigue, creep and creep rupture, irradiation, and loss of ductility. Extensive use of the plastic region of steel was made for the accommodation of thermal stresses. The concept of cumulative damage in the plastic region was expounded for thermal fatigue and creep. A short description is given of the five avenues followed for the development of a theory governing the strength of materials. An approach was taken up that attempts to establish a "theory of fatigue" based on experiments. (auth)
Date: July 1, 1961
Creator: Fistedis, S.H.
Partner: UNT Libraries Government Documents Department
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Design of production test IP-409-A-FP, pilot test of self-supported fuel elements in K size smoothbore zirconium process tubes

Description: In the Plant Improvement Program, it is proposed to retube the K-Reactors with standard size Zircaloy-2 smooth-bore process tubes and to charge self supported fuel elements starting March 15, 1964. The first step in support of this transition program is to confirm compatibility of the fuel-tube geometry and secondly to obtain fuel-tube performance information prior to full scale commitment of the K-Reactors to this design. In view of the testing of self-supported fuel which has been accomplished to date and that which is planned, there is little incentive to install more tubes in a K-Reactor than are required to make the fuel-tube geometry check. To accomplish this, ten tubes are viewed as the maximum number that would be required. This report presents the design of a test to fabricate and irradiate ``K`` self-supported fuel elements in limited quantities.
Date: July 26, 1961
Creator: Clinton, M. A. & Hodgson, W. H.
Partner: UNT Libraries Government Documents Department
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Design of supplement A to production test IP-183-A-98-FP evaluation of projection fuel elements in K process tubes

Description: Fuel element misalignment is an apparent cause of ``hot-spot`` ruptures. Several methods of eliminating hot spots have been tried, however, none appear to have completely solved the hot-spot problem. Attachment of projections to the side of the fuel elements appears to offer a means of minimizing misalignment since they act as bumpers against the side of the process tubes. Preliminary data from tests now in progress or recently completed indicate excessive corrosion rates are not to be expected on fuel elements with projection attached. In fact, no hot spots were found on 39 columns of normal self-supported I&E fuel elements discharged at normal goal or on four columns of enriched self-supported I&E elements irradiated to 850 mwd/t exposure.
Date: July 27, 1961
Creator: Hodgson, W. H. & Clinton, M. A.
Partner: UNT Libraries Government Documents Department
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Design Study of Portable Thermoelectric Nuclear Systems

Description: Design studies were performed and costs were estimated for an air transportable, 10 Mw(t), pressurized light water thermal circulation reactor, combined with a direct conversion thermoelectric generator and static electrical inversion equipment. This TCR-TE'' concept appears to have potential for ultimate use as a remote unmanned power station. Based on an extrapolation of present reactor technology and on assumed thermoelectric materials properties forecasted to January 1, 1963, a net a-c electrical output of 315 Kw is estimated, assuming the use of 80 deg F local water for cooling purposes. An alternate concept using 80 deg F air for cooling produces 271 Kw, net. These electrical output figures can be improved significantly through a recommended research and development effort. The design and construction of a prototype plant is also recommended. (auth)
Date: July 1, 1961
Creator: Chajson, L.; DelCampo, A. R. & Kellogg, H.B. et al
Partner: UNT Libraries Government Documents Department
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