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Army Gas-Cooled Reactor Systems Program Monthly Progress Report: April 1959

Description: Abstract: This monthly progress report covers the activities of the Army Gas-Cooled Reactor System Program for April 1959. The program includes a water-moderated heterogeneous reactor (Gas-Cooled Reactor Experiment I), a graphite-moderated homogeneous reactor (Gas-Cooled Reactor Experiment II), a mobile gas-cooled reactor (ML-1), and the coordination of the Gas Turbine Test Facility. [It reports] the progress of each project, the associated tests and data evaluation, the applicable design criteria, and the fabrication of reactor components" (p. 1).
Date: May 25, 1959
Creator: Aerojet-General Corporation
Partner: UNT Libraries Government Documents Department
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Brief Review of Heat Transfer Problems Encountered in the Production of Magnetic Fields

Description: The design of internally cooled electrical coils for the production of high intensity magnetic fields presents many new aspects and combinations of the familiar modes of heat transfer. However, the customary methodology appears to be sufficient for preliminary analysis and understanding of those problems. This methodology comprises the derivation of a qualitative, approximate equation expressing the relative performance of the various parts of a system, followed by an examination of this equation in order to locate the limiting features of the system. These features are then investigated by more powerful methods, which in turn provide guidance for development research in the laboratory. (auth)
Date: May 25, 1959
Creator: Alexander, L G
Partner: UNT Libraries Government Documents Department
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PROBLEMS IN ACCOUNTABILITY MEASUREMENTS ASSOCIATED WITH THE INTERIM CHEMICAL PROCESSING PROGRAM

Description: Available knowledge of precision limits in S.S. accountability measurements and/or calculations by reactor and chemical processing groups is surveyed and summarized. Experienee in comparisons of reactor (production and research) calculations vs. chemical plant accountability measurements is also reported. A general tentative conclusion is that available precisions ( plus or minus 0.54 to plus or minus 0.78%) in chemical plant measurements (bulk and analytical) for fissionable material accountability is superior to the variable precision ( plus or minus 1.0 to 1l.0%) possible by calculations (nuclear and/or engineering) of power reactor systems; however, with operation and empirical experience (e.g., after two or three core loadings), it is believed that calculations for given reactors can attain acceptable precisions, e,g., less than plus or minus 1.0%. It may be proposed that fuel payments be made as follows: 90% of fuel value based on reactor calculations, an additional 5% based on dissolver analyses, and final settlement based on chemical plant material balance (product plus loss analyses). (auth)
Date: May 28, 1959
Creator: Arnold, E D & Gresky, A T
Partner: UNT Libraries Government Documents Department
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Liquid Metal Fuel Reactor Experiment: Dynamic Utility Test Loop

Description: This report provides an overview of the creation of the Liquid Metal Fuel Reactor Experiment program. It furthers the work by constructing a single loop to test all the components required for the 16 loop reactor. This utility loop was also constructed to provide a facility for testing various components such as valves and flow meters.
Date: May 5, 1959
Creator: Baker, O. H.
Partner: UNT Libraries Government Documents Department
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CALCULATION OF GROUP CROSS SECTIONS FOR HOT MONOATOMIC MODERATOR WITH VARIABLE FLUX WEIGHTING WITHIN GROUPS, 704 CODE 521/RE 145

Description: This code finds inelastic cross-section matrix elements (transfer matrix) for hot monatomic moderator for multigroup calculations by numeric- analytic double integration of Cohen's formula. Several approximations to the actual neutron density ean be used as weight functions over the velocities of the initial groups. Modified and supplemented results are presented on binary cards and/or tape for direct input into the Argonne Transport Theory Codes or the SNG Code, or for offline output. (auth)
Date: May 1, 1959
Creator: Bareiss, E.H.; Denes, J.E. & Jankus, V.Z.
Partner: UNT Libraries Government Documents Department
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An Experimental Evaluation of the Radiation Protection Afforded by a Large Modern Concrete Office Building

Description: Abstract: "An experimental study was made to determine the effective shielding provided by a modern reinforced-concrete office building (AEC Headquarters building) from nuclear fallout. Pocket ionization chambers were used for measurement of the radiation-field strength. Fallout was simulated with distributed and point-source configurations of Co-60 and Ir-192 sources. Four typical sections were selected for study, and experiments were performed on each. These included an external wing with exposed basement walls and an external wing with a buried basement. Roof studies were made on an internal wing with a full basement and on the east end of wing A, which has a thin-roof construction. The thick-roof construction of 8 in. of concrete and 2 in. of rigid insulation covers all the building except the east end of wing A, which has 4 in. of concrete and 2 in. of insulation."
Date: May 1, 1959
Creator: Batter, J. F., Jr.; Kaplan, A. L. & Clarke, Eric Thacher
Partner: UNT Libraries Government Documents Department
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An Experimental Evaluation of the Radiation Protection Afforded by a Large Modern Concrete Office Building

Description: An experimental study was made to determine the effective shielding provided by a modern reinforced-concrete office building (AEC Headquarters building) from nuclear fall-out. Pocket ionization chambers were used for measurement of the radiation-field strength. Fall-out was simulated with distributed and point-source configurations of Co/sup 60/ and Ir/sup 192/ sources. Four typical sections were selected for study, and experiments were performed on each. These included an external wing with exposed basement walls and an external wing with a buried basement. Roof studies were made on an internal wing with a full basement and on the east end of wing A, which has a thin-roof construction. The thick-roof construction of 8 in. of concrete and 2 in. of rigid insulation covers all the building except the east end of wing A, which has 4 in. of concrete and 2 in. of insulation. (auth)
Date: May 1, 1959
Creator: Batter, Jr., J. F.; Kaplan, A. L. & Clarke, E. T.
Partner: UNT Libraries Government Documents Department
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Preparation of Pitch-Soluble Uranyl-Organic Compounds

Description: Batch processes on a scale of 250 to 300 g of uranium were developed for the production of uranyl oxinate (8quinolinate) and uranyl malonate. Both compounds are insoluble in water and were found to be suitably soluble in pitch. Uranyl oxinate was prepared by the reaction of an aqueous uranyl nitrate solution with an acetic acid solution of oxine (8-quinolirol) at about 80 deg C. Complete precipitation was accomplished by the addition of ammonium hydroxide. Yields of better than 99.5% were obtained. Uranyl malonate was prepared by the reaction of aqueous solutions of sodium malonate and uranyl nitrate at about 80 deg C in 97 to 98% yield. Uranyl 2-ethylhexanoate was prepared by a transesterification reaction from uranyl acetate and 2-ethylhexanoic acid. Yields of 90% were obtained but the process was quite laborious ard time consuming. A metathesis method of preparation was not successful. (auth)
Date: May 1, 1959
Creator: Baxman, H. R.; Jackson, D. D.; Williams, D. L. & Bard, R. J.
Partner: UNT Libraries Government Documents Department
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NONDESTRUCTIVE TESTING OF EBR-I MARK III FUEL ELEMENTS AND COMPONENTS

Description: Ultrasonic and eddy current methods were used to inspect EBR-I Mark III fuel elements and componentsUltrasonic techniques were used to inspect for homogeneity of the casting, bonding of the core to the clad on the extruded rod, bonding of the Zircaloy spacer disk to the uranium, and cracks in the Zircaloy rod used for end caps. Eddy current techniques were used to measure the cladding thickness on the extruded rods and to inspect the zirconium wire used for spacers on the completed fuel element. (auth)
Date: May 1, 1959
Creator: Beck, W.N.; Renken, C.J.; Myers, R.G. & McGonnagle, W.J.
Partner: UNT Libraries Government Documents Department
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Scram transient tests PT-IP-249-C

Description: The purpose of this production test is to provide a standard method of obtaining scram transient reactivity information at the eight reactors, under conditions conducive to valid data. These conditions include the bypassing of the Panellit system at a low power level for a short, controlled period of time during May 1959.
Date: May 25, 1959
Creator: Bowers, C.E.
Partner: UNT Libraries Government Documents Department
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THE DEVELOPMENT OF A FLUIDIZED BED REACTOR FOR THE FLUOROX PROCESS: UNIT OPERATIONS MONTHLY STATUS REPORTS FOR THE PERIOD NOVEMBER 1958 THROUGH MAY 1959

Description: Results of four experimentul runs in the Fluorox fluidized bed reactor system are reported. The engineering feasibility of UF/sub 6/ production from UF/ sub 4/ by use of dry air of O/sub 2/, 2UF/sub 4/ + O/sub 2/ = UF/sub 6/+ UO/sub 2/ F/sub 2/, in an Inconel fluidized bed reactor at 800 to 850 deg C was demonstrated in two experimental tests in which greater than 90% of the theoretical amount of UF/sub 6/ was collected or measured. Two runs made with crude UF/sub 4/ (produced from unpurified mill concentrate) as the feed material, showed thnt UF/sub 6/ could be produced at 700 to 725 deg C but corrosion on Inconel was prohibitive. (auth)
Date: May 26, 1959
Creator: Bresee, J C; Horton, R W & Scott, C D
Partner: UNT Libraries Government Documents Department
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REACTOR CONTAINMENT (INCLUDING A TECHNICAL PROGRESS REVIEW)

Description: An attempt is made to present available information pentinent to reactor containment. This is done directly, by summary and reference, or by reference alone. To provide a reference framework, the first review document must necessarily be handled differently from supplemental periodic reviews. The plan is to: (3) provide a detailed account of the problem and suggestions for work needed to yield adequate solutions; (2) present the accumulated knowledge and accomplishments; (3) give an account of experience in applying the containment concept; and (4) append extensive bibliographical material. An attempt is made in each case to indicate the significance of the information and its relation to the problems outlined. (A.C.)
Date: May 1, 1959
Creator: Brittan, R.O.
Partner: UNT Libraries Government Documents Department
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Fringe isotope production

Description: The Purpose of the work described in this report has been to determine experimentally the rate of production of tritiun in fringe lithium-aluminum alloy loadings with the degree of precision necessary for economic analyses of such a method of isotope production. These results are provided for use in such an analysis.
Date: May 6, 1959
Creator: Bunch, W. L.
Partner: UNT Libraries Government Documents Department
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Adsorption of Xenon in an Activated Charcoal Column

Description: Performance characteristics of two activated charcoal columns at room temperature in separating fission-product xenon from an air stream were investigated by installing each column in the exhaust from an enclosure in which irradiated slugs were dissolved. Breakthrough curves are presented and the variation in xenon concentration within the columns is examined. Theoretical treatments of adsorption columns in the literature are found to agree well with the experimental data. Performance of the colunms is evaluated in terms of concentration factor'' and number of effective theoretical plates. (auth)
Date: May 11, 1959
Creator: Cantelow, H. P.
Partner: UNT Libraries Government Documents Department
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NUMERICAL SOLUTION OF TRANSIENT AND STEADY-STATE NEUTRON TRANSPORT PROBLEMS

Description: A general numerical procedure, called the discrete S/sub n/ method, for solving the neutron transport equation is described. The main topics relate to the derivation of suitable difference equations, and to the problem of solving these, while maintaining generality, accuracy, and reasonable computing speed. A few comparisons with other methods are made. (auth)
Date: May 16, 1959
Creator: Carlson, B.
Partner: UNT Libraries Government Documents Department
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Increased production from deliberate discharge cycling

Description: Considerable production gains might be attained if each reactor discharged its entire flattened region during one scheduled outage instead of utilizing several outages for this purpose. Several of the older reactors are now discharging a high percentage of their flattened zones in a single outage and could be put into this type of operation with relatively little difficulty. Production gains may be possible through better flattening efficiency, a more favorable rupture rate effect, fewer non-equilibrium losses, higher conversion ratio, and more efficient usage of outage work. Since this document is written Primarily from the Operational Physics standpoint, some gains and pitfalls which must be evaluated by other affected groups will only be mentioned here as possibilities. The purpose of this document is simply to point out the potential gains in flattening efficiency from this method. Potential gains from improved fuel performance have been described in another document.
Date: May 28, 1959
Creator: Carter, R. D.
Partner: UNT Libraries Government Documents Department
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Irradiation performance of coextruded enriched uranium fuel rod PT-IP-A172-A: Final report

Description: The proposed operating conditions for fuel elements to be charged into the NPR require the fuel to be of an extended surface geometry and maintain adequate strength and corrosion resistance in 300 C water. A contract was let to Nuclear Metals Inc. to produce by co-extrusion lengths of fuel rod containing both natural and 1.6% enriched uranium of irradiation quality for fabrication into fuel elements. The fuel rods used in the irradiation test represent the first enriched uranium rods coextruded in 0.030 inches of Zircaloy-2 to be irradiated and examined at Hanford. The rods used for this test were fabricated into four, 4 rod cluster fuel elements thus allowing adequate space between individual rods for expansion in the case of a fuel rod failure. This rod was of particular interest since it contained an irregular uranium-Zircaloy-2 interface. The purpose of the irradiation was to determine the dimensional stability of coextruded fuel rods and to determine whether the irregularity in the bond interface had any effect upon the irradiation performance of the fuel. Fuel elements were irradiated in 200 C water in the KER Loop 2 facility to an exposure of 0.28 a/o burnup (2,200 MWD/T). Post irradiation examination showed that each rod had increased an average of 0.008 inches in outside diameter and that macrocracks had formed throughout the uranium core. The uranium had also increased in length to fill 0.050 inch of space left between the end cap and uranium for thermal expansion and uranium growth. A metallurgical bond between the end cap and the uranium had been formed during the irradiation. There was no effect of the irregular interface on the dimensional stability of the fuel rods.
Date: May 26, 1959
Creator: Claudson, T. T.
Partner: UNT Libraries Government Documents Department
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