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Intermediate Heat Exchanger Preliminary Design. Vol. 1, IHX Preliminary Design

Description: Preface: The intermediate heat exchanger is designed for operation in a nuclear power plant using liquid sodium as the primary and secondary coolant. Since the primary fluid coming from the reactor is radioactive, the purpose of the IHX is to transfer heat to a nonradioactive fluid which then goes to a steam generator. Because of this activity the until will be enclosed in a concrete pit and will not be accessible during periods of operation. Immediately after shut down it will be necessary to allow time for radioactive decay before the unit will be accessible to personnel. Because of inaccessibility and possible long periods allowed for decay time, it is imperative that the unit give trouble free operation. During periods of shut down, the internals should have easy access for inspection and repair if necessary so that down time is held to a minimum. The general arrangement of the heat exchanger described in this report presents a conventional design utilizing known materials and existing methods of fabrication. In further consideration of all concepts, designs and analyses developed during this period of the program, it is felt that this preliminary design will provide an intermediate sodium heat exchanger of lower cost and more reliable operation.
Date: February 28, 1959
Creator: Alco Products (Firm)
Partner: UNT Libraries Government Documents Department
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STEAM GENERATOR PRELIMINARY DESIGN

Description: A conceptual study on design of sodium-cooled reactor steam generators was conducted. Included is a detailed description of the preliminary design and analysis, based on the use of known materials and existing methods of fabrication. (See also APAE-41 Vols. I and III.) (J.R.D.)
Date: February 28, 1959
Partner: UNT Libraries Government Documents Department
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BOUNDARY CONDITIONS AND CONSERVATION PROPERTIES OF FOPP, A PLASMA FOKKER- PLANCK CODE

Description: The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)
Date: February 27, 1959
Creator: Fowler, T.K.; Rankin, F.M. & Simon, A.
Partner: UNT Libraries Government Documents Department
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Fuel Cycle Costs in a Graphite Moderated U$sup 235$-Th Fueled Fused Salt Reactor

Description: A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)
Date: February 27, 1959
Creator: Guthrie, C. E.
Partner: UNT Libraries Government Documents Department
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SM-2 VAULT CRITICALITY

Description: To determine the safety of the array in the storage vault for the SM-2 experimental fuel plates, two criticality criteria were applied. A maximum of 18 fuel plates was stored in sthainless steel tubes and the tubes belted to a frame on the wall to prevent movement. No tube could go critical by itseIf. The vauit was then assumed completely flooded by water. In the first calculation, the fuel array was assumed to be distributed uniformly over the wall forming a large slab. This method indicated the array might be critical if the steel tube and cadmium lining were neglected. In the second method, a conservative calculation, wnich included the steel tube and cadmium lining was made. This method indicataed the array was subcritical. Calculations were then made of the criticalty of the SM-2 vault without the steel--cadmium tubes and wcoden blocks. The multiplication factor of the vault was also calculated. In order to determine the accuracy of these calculations, an ORNL critical experimental array was calculated applying the same analytical techniques. (M.C.G.)
Date: February 27, 1959
Creator: Fried, B.E.
Partner: UNT Libraries Government Documents Department
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SUMMARY OF RUNS D-77 THROUGH D-93. RATE OF OXIDATION OF CHROMIUM(III) IN DILUTE URANYL SULFATE SOLUTION IN THE PRESENCE OF RUTHENIUM

Description: The rate of oxidation of chromium(III) to chromium(VI), catalyzed by ruthenium, was determined at various temperatures and oxygen concentrations. The rate at 300 deg C was too rapid for measurement by aliquot sampling. In the temperature range of 225 to 275 deg C, oxidation was rapid and the rate increased with oxygen concentration. A linear dependence of initial oxidation rate on the reciprocal of chromium(VI) concentration suggested that a rate-controlling step in the reaction mechanism may be desorption of chromium(VI) from the ruthenium catalyst. The activation energy calculated for the reaction is 19 kcal/mole. (auth)
Date: February 27, 1959
Creator: Snavely, E.S.; Greeley, R.S. & Buxton, S.R.
Partner: UNT Libraries Government Documents Department
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THE CRITICAL CURRENT IN THE CASE OF NEUTRAL AND PLASMA BREAKUP

Description: An estimate of the critical current for the case of breakup by an arc has been obtained. It is shown that the same technique allows a quick estimate of the critical current in the case of no arc, using neutral and plasma breakup instead. (auth)
Date: February 26, 1959
Creator: Simon, A.
Partner: UNT Libraries Government Documents Department
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FOREIGN RESEARCH AND POWER REACTOR PRELIMINARY LIST

Description: Foreign research and power reactors are tabulated. Nuclear power buildup goals are given for each nation on which information is available. (J.H.D.)
Date: February 26, 1959
Creator: Ullmann, J.W.
Partner: UNT Libraries Government Documents Department
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NUCLEAR EXCURSION ANALYSIS ON THE IBM 650 (EARLY CODE)

Description: The Early Program is designed to compute the peak power and total energy release vhich will occur in a reactor due to a step increase of reactivity. These excursions may be terminated by one of several automatic shutoff mechanisms caused by the power excursion itself. The calculation assumes no thermodynanaic delays and assumes void formation is due to heat conduction from the cladding to the moderator. (W.D.M.)
Date: February 26, 1959
Creator: Reiback, E.M. & Stueck, C.J.
Partner: UNT Libraries Government Documents Department
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NUCLEAR EXCURSION ANALYSIS ON THE IBM 650 (EARLY CODE)

Description: The early program is designed to compute the peak power and total energy release which will occur in a reactor due to a step increase of reactivity. The calculation assumes no thermodynamic delays and assumes void formation is due to heat conduction from the cladding to the moderator. For periods in the 1 to 10 msec. range, the calculation yields results which are in excellent agreement with SPERT experimental results. (W.D.M.)
Date: February 26, 1959
Creator: Stueck, C.J. & Reiback, E.M.
Partner: UNT Libraries Government Documents Department
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ANALYSIS OF NEUTRON PULSES IN A GODIVA-TYPE REACTOR

Description: Some calculations have been made to estimate the characteristics of a neutron-burst type fast reactor similar to Godiva but made up of relatively small component parts--the so-called "layered assembly." One spherical and three cylindrical assemblies have been considered. Critical masses, assuming 5% voids, range from 58 to 65 kg of 93.4% enriched U/sup 235/. For a reactivity addition of 0.33 dollars above prompt criticals bursts between 2 x 10/sup 17/ and 6.7 x 10/ sup 17/ fissions were computed with accompanying temperature rises varying from 514 to 1600 deg C. The burst width at half-maximum was about 12 microseconds. To obtain an idea of the possibilities of stress reduction which might be achieved by layerings an assembly made of small rings was considered. While the critical masses obtained here are believed to be fairly accurates the predictions concerning mechanical energy generated, total fissions, and burst width may be subject to sizeable error due to the many simplitications required to allow hand computations. Neverthelesss considerable improvement in safety and burst-size is indicated by the use of a "layered assembly" instead of an assembly composed of relatively thick parts. (auth)
Date: February 25, 1959
Creator: Nestor, C.W. & Tobias, M.
Partner: UNT Libraries Government Documents Department
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H Reactor enrichment

Description: No Description Available.
Date: February 24, 1959
Creator: Turner, R. L.
Partner: UNT Libraries Government Documents Department
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Strontium Program Quarterly Summary Report: November 1958 - January 1959

Description: From Abstract: "This report is one of a sequence of quarterly reports, each designed to up-date its predecessor beginning with HASL-42, "Environmental Contamination from Weapon Tests." Herein are reported data which have accrued since HASL-51. In particular, the levels of strontium 90 in fallout, milk, tap water, vegetation, and foods are given, based on data available from November 1, 1958 to January 30, 1959."
Date: February 24, 1959
Creator: Hardy, Edward P., Jr. & Klein, Stanley
Partner: UNT Libraries Government Documents Department
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Pilot plant denitration of Purex wastes with formaldehyde

Description: The reaction between formaldehyde and nitric acid, in which the acid is destroyed with the production of predominantly gaseous products, has been recognized as of great potential value in the processing of radioactive fuels, particularly during waste treatment. Laboratory studies of the reaction at Harwell and at Hanford have shown that a major fraction of the nitric acid can be readily removed from an acidic solution containing nitrates by the addition of formaldehyde. The process possesses the advantages of low chemical cost; recoverability of nitric acid; and, in the case of waste treatment, the production of a solution relatively low in inert salt concentration suitable for fission product recovery or ultimate disposal. The primary purpose of the study was to confirm and extend existing information on the formaldehyde reaction to the destruction of nitric acid in Purex type waste (1WW) through operation of pilot plant scale apparatus. Operational behavior, formaldehyde utilization efficiency, and safety considerations were particular subjects of study. In addition, destruction of nitric acid in a Darex-type dissolver solution was investigated.
Date: February 23, 1959
Creator: Evans, T. F.
Partner: UNT Libraries Government Documents Department
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Production Test IP-237-A, irradiation of enriched seven-rod cluster elements for ETR testing

Description: Two Zircaloy-2 jacketed seven-rod cluster elements will be irradiated in the 3674 KE front-to-rear test hole to an exposure of 1000 MWD/T and two elements will be irradiated in the 3674 KW front-to-rear test hole to an exposure of 2000 MWD/T. After irradiation, the elements will be sent to the ETR where they will be ruptured during reactor operation to determine the failure characteristics of co-extruded Zircaloy-2 jacketed cluster elements.
Date: February 23, 1959
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department Monthly Report: January 1959

Description: This report for January 1959, 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; and employee relations.
Date: February 20, 1959
Creator: Hanford Atomic Products Operation. Chemical Processing Department.
Partner: UNT Libraries Government Documents Department
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Irradiation Processing Department monthly record report, January 1959

Description: This document details activities of the irradiation processing department during the month of January 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.
Date: February 20, 1959
Creator: Greninger, A. B.
Partner: UNT Libraries Government Documents Department
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THE NUCLEAR FUEL CYCLE: PROSPECTS FOR REDUCING ITS COST

Description: Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)
Date: February 20, 1959
Creator: Albrecht, W.L.
Partner: UNT Libraries Government Documents Department
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Thermal Bowing of Core Subassemblies

Description: Report issued by the APDA over studies conducted on thermal bowing of core sub-assemblies at the Enrico Fermi Atomic Power Plant. As stated in the introduction and summary, "the present report deals with this problem in the design of the Enrico Fermi Reactor and with the method which has been adopted to avoid a positive temperature coefficient of reactivity" (p. 1). This report includes illustrations, and photographs.
Date: February 20, 1959
Creator: De Stordeur, Arnold
Partner: UNT Libraries Government Documents Department
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THE FABRICATION OF THE GRAPHITE-URANIA FUEL FOR THE TRANSIENT REACTOR TEST

Description: The two predominate methods of dispersing uranium in graphite are reviewed and evaluated. This study indicated that the most feasible method of dispersing uranium in graphite would be to fabricate a mixture of graphite and U/ sub 3/O/sub 8/ bonded with a thermosetting resin. A commercial type graphite was developed through independent research, and this fabrication procedure was adapted for the manufacture of the TREAT fuel matrix. (auth)
Date: February 19, 1959
Creator: Handwerk, J.H.; McCuaig, F.D. & Bean, C.H.
Partner: UNT Libraries Government Documents Department
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