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PLUTONIUM OXALATE DISK FILTER AND FILTER MEDIA STUDIES

Description: for filtration of plutonium oxalate slurries. A scalpel produces a slit in the filter precoat, leading to increased filtration in this slit, and the oxalate is removed by a doctor knife; this technique results in prolonged blowback cycles and more uniform delivery of filtered oxalate to subsequent processing steps. Several types of filter media were tested, and rigid porous aluminum oxide was found to be the best one. (D.L.C.)
Date: October 19, 1959
Creator: Rey, G.
Partner: UNT Libraries Government Documents Department
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Neutron flux in K Reactor discharge area during operation

Description: Based on the activation of gold foils in an hydrogeneous medium, the neutron flux incident on the rear wall of the discharge area of the KE reactor is estimated to be 6000 M/cm{sup 2} sec. The effective energy of the neutrons is estimated to be approximately 4 Mev. Neither of these values confirm order-of-magnitude estimates of the neutron flux and neutron energy expected to exist in the discharge area.
Date: June 19, 1959
Creator: Bunch, W. L.
Partner: UNT Libraries Government Documents Department
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Historical record of data on flood control

Description: Last year (1948) during the flood period the flow at Grand Coulee fluctuated widely. 2 PM, June 8, 543000 c.f.s.; 4 AM, June 9, 568000 c.f s.; 2 PM, June 9, 543000 c.f.s.; 2 AM, June 10, 573000 c.f.s. A total instantaneous fluctuations of 37,500 c.f.s. was reported. Now there is installed a new control. This control can keep downstream variation within 500 c.f.s. By lowering the lake level prior to the crest period, the drum gates could be used as flood control (1948 high water basis) the drum … more
Date: May 19, 1959
Creator: Kramer, H. A.
Partner: UNT Libraries Government Documents Department
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Production test IP-229-A evaluation of the uranium-Al-Si bond at high temperature

Description: The objective of this production test is to determine the changes that occur in the uranium-Al-Si bond during irradiation at bond temperatures between 255 and 285 C. Twenty-five M-388 jacketed dip canned depleted uranium solid fuel elements will be irradiated to an exposure of 500 MWD/T in high temperature water. The location and size of unbonded areas on the fuel elements will be measured by ultrasonic mapping before and after irradiation to show the changes in bonding resulting from irradiati… more
Date: January 19, 1959
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Post irradiation examination of KER-1-3 seven rod cluster fuel elements (RM-277)

Description: Two coextruded, Zr-2 clad, natural uranium, seven rod cluster fuel elements were irradiated to a calculated exposure of 1250 MWD/T in the KER Facility and discharged 1-16-59. The fuel elements were NPR candidate fuel and examination was requested to determine the behavior of coextruded, Zr-2 clad, natural uranium irradiated at core temperatures of approximtely 425{degree}C. The elements were transferred to the Radiometallurgy Laboratory 2-25-59. The elements demonstrated excellent in reactor pe… more
Date: October 19, 1959
Creator: Gruber, W. J.
Partner: UNT Libraries Government Documents Department
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Supplement C to Production Test IP-250-A, Irradiation of Zircaloy-2 jacketed tube and tube elements in the KER loop

Description: The objective of this Supplement described in this report to Pt-IP-250-A is to d enriched tube-and-tube elements will develop pitting corrosion on the Zircaloy-2 jackets when irradiated in pH 10 water. The measurement of dimensional changes in the fuel elements and the observation of the effect of irradiation on the uranium and bond area are also objectives of the test, but secondary in importance to identifying a pitting corrosion problem in NPR quality water, if one exists.
Date: October 19, 1959
Creator: Kratzer, W. K.
Partner: UNT Libraries Government Documents Department
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Existing reactor water plant study -- B, C, D, DR, F and H reactors interim report

Description: The five year forecast for operation of the HAPO reactors calls for the achievement of increased process water flows in B, C, D, DR, F and H reactors. The Process Design Operation has initiated a study in support of this forecast whose objectives are: to determine present water plant and effluent system flow capabilities; to provide basic data for determining the ultimate economic optimum flow capability of these plants; and-to provide a basis for scope and development work preliminary to the i… more
Date: January 19, 1959
Creator: Watson, D. F.
Partner: UNT Libraries Government Documents Department
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K Water Plant improvements

Description: A Task Force was established in the Irradiation Processing Department to examine the K-Reactor Water Plant to (1) review the operating and maintenance experience with the water plant as improved since startup, (2) identify major plant additions which could further improve reliability, and (3) estimate the costs of any such additions. The K-Water Plant basically consists of the electrically driven primary cooling system with power supplied by the BPA system, electrically driven secondary or back… more
Date: March 19, 1959
Creator: Trumble, R. E.; Heacock, H. W.; Reinig, L. P.; Jones, S. S. & Mollerus, F. J.
Partner: UNT Libraries Government Documents Department
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Weldability of Hayes Alloy #25

Description: Technical report describing the process to determine the fusion welding characteristics of Haynes Alloy #25 as applied to TLJ-100530, Corrosion Loops. Hayes Stellite Alloy #25 is a cobalt-base alloy for corrosion resistant high temperature applications. This material, when welded by the inert gas shielded tungsten arc method, produces sound ductile joints. Material thicknesses greater than 12 gauge require standard joint preparations, a V joint being preferred up to 1/4 inch and a U joint for … more
Date: May 19, 1959
Creator: Rogers, S. L.
Partner: UNT Libraries Government Documents Department
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Dispersions of Uranium Carbides in Aluminum Plate-Type Research Reactor Fuel Elements

Description: The technical feasibility of employing uranium carbide aluminun dispersions in aluminum-base research reactor fuel elements was investigated This study was motivated by the need to obtain higher uranium loadings in these fuel elements. Although toe MTR-type unit, containing a 13 18 wt% U-Al alloy is a proven reactor component, fabrication problems of considerable magnitude arise when attempts are made to increase the uranium investment in the alloy to more than 25 wt.%. Au approach to these fab… more
Date: November 19, 1959
Creator: Thurber, W. C. & Beaver, R. J.
Partner: UNT Libraries Government Documents Department
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THEORETICAL STUDY OF SINGLE-TRANSFER LINE CONCATENATED PULSE DOLUMN SYSTEMS

Description: Calculations indicate that single-transfer line concatenated pulse column systems can be operated with static pressures that are not excessive if a sufficient number of vessels are employed in the system. The required number of vessels can be attained by using a series of short columns or by using holdup pots in conjunction with a limited number of columns. General equations for calculating pressure drops and power requirements are presented. (auth)
Date: June 19, 1959
Creator: Johnson, H F
Partner: UNT Libraries Government Documents Department
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Operation of the Hrt Mockup With Boiling Fuel in a Titanium Pressurizer, Run CS-23

Description: The 0.045 m UO/sub 2/SO/sub 4/, 0.036 m CuSO/sub 4/, 0.025 m H/sub 2/SO/ sub 4 solution (HRT fuel composition) was chemically stable during 1,866 hr of operation at 280 C and 1500 psi. The system was pressurized by boiling a 0.4 gpm stream of the fuel in a titanium heat exchanger at 313 C. During cursions were made to pressurizer temperatures above 330 C where two liquid phases were formed. These tests indicated that heavy phase began formation at 325 C (vapor pressure equilibrium temperature) … more
Date: May 19, 1959
Creator: Korsmeyer, R B & Harley, P H
Partner: UNT Libraries Government Documents Department
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Interim Memorandum Report on Filter Development and Discussion on Availability of Materials

Description: A preliminary report on development of filter paper for use in A.E.C. operations is presented. Filters for use in removing harmful dusts or radioactive matter from the air discharged from various operations or for other uses are described. The main fiber furnished for the paper is a specially treated wood pulp with an addition of asbestos. Filters of higher or lower efficiencies with corresponding changes in static resistance can readily be made by modifying the manufacturing formula. (J.R.D.)
Date: May 19, 1959
Partner: UNT Libraries Government Documents Department
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POWER REACTOR FUEL REPROCESSING PROCESS WASTES

Description: Data on waste volumes and heat generation of several reactor fuels which may be reprocessed in the Power Reactor Fuel Reprocessing Pilot Plant at ORNL are tabulated. (auth) l6876 A tabulation containing information on the power of existing and proposed U. S. and U. S.-built reactors of 10 kw or greater thermal power is presented. Estimated fuel reprocessing loads for irradiated fuels are also iucluded. (auth)
Date: June 19, 1959
Creator: Conger, W L
Partner: UNT Libraries Government Documents Department
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SOLUBILITY OF LITHIUM HYDROXIDE IN WATER AND VAPOR PRESSURE OF SOLUTIONS OF LITHIUM HYDROXIDE ABOVE 220 F

Description: The solubility of lithium hydroxide in water was determined at 220 to 650 F. The literature furnished data for temperatures below 200 F. A maximum in the curve was found at about 240 and a minimum at 480 F. The variations in solubility, however, were relatlvely small. At 40, the solubility is 12.7 g LiOH per 100 g H/sub 2/O, while at 240, it is 17.7, and at 650 F, it is 16.5. The vapor pressures of 4.76 wt. % (2.09 molal), 8.59 wt.% (3.92 molal), and saturated (approximately 6.25 molal) lithium… more
Date: March 19, 1959
Creator: Stephen, E.F. & Miller, P.D.
Partner: UNT Libraries Government Documents Department
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REACTIVITY OF SUBSTITUTION ELEMENTS

Description: A method is sought for predicting reactivity differences between fuel elements attached to control absorbers and fixed fuel elements. The approach described is based on a logic that is approximate, and which should be subjected to experimental check. (auth)
Date: January 19, 1959
Creator: Murray, R.L.
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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Losses associated with the interim purification processing of neptunium

Description: This report discusses the interim program for the production of neptunium oxide at HAPO which applies the following processing steps: isolation of neptunium from the Purex process streams, using Purex flow sheets specially adapted for this purpose; purification of the neptunium nitrate by an ion exchange process carried out in one of the Redox laboratory (222-S) multi-curie cells; and precipitation of neptunium oxalate and conversion of the oxalate to oxide in laboratory-type equipment. The pro… more
Date: May 19, 1959
Creator: Harmon, K. M.
Partner: UNT Libraries Government Documents Department
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Determination of Thickness of Oxide Film on Phosphor Bronze

Description: The thickness of an oxide film on phosphor bronze helices was determined by first establishing the oxygen content of the helix "as received" and after cleansing with nitric acid. Based on the assumption that the difference between the two values was the oxygen in the film, and that the film consisted entirely of cupric oxide, the thickness of the film was calculated from the density of cupric oxide, weight of film, and surface area of film. A value of 1080 A was calculated as the thickness by t… more
Date: May 19, 1959
Creator: White, J. C.
Partner: UNT Libraries Government Documents Department
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HNPF PROCESS TUBE-GRID SEAL

Description: A maximum leak rate of 0.08% was measured for a piston ring seal assembly which was evaluated for use as the Hallain Power Reactor process tube- grid plate seal. A maximum leak rate of 0.14% was observed after subjection to 10,000 cycles (560 hr) in 625 deg F Na. The maximum leak rate was 0.07% after 25 cycle exposure in 1000' Na. Vertical scoring of both the rings and bore tube was observed. Sodium was observed to remain behind the rings after washing. (C.J.G.)
Date: August 19, 1959
Creator: Charles, J.
Partner: UNT Libraries Government Documents Department
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THERMODYNAMICS IN THE FUSED SALT DISSOLUTION PROCESS FOR ZIRCONIUM FUEL

Description: A discussion is given of the role of thermodynamics in the fused-salt volatility process, particularly as it applies to oxidation-reduction reactions affecting zirconium, uranium, nickel, chromium, ruthenium, and other elements present in the hydrofluorination head-end step. (auth)
Date: November 19, 1959
Creator: Cathers, G.I.
Partner: UNT Libraries Government Documents Department
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TRANSIENT TESTS OF HNPF PROTOTYPE SODIUM PUMP DRIVES

Description: The objectives of this study were to demonstrate that the pump speed control system will respond as defined in the equipment specifications and to determine optimum values of controlling variables that will minimize the oscillations that occur in the Na flow rate when transient signals are imposed on the pump speed control system. (W.L.H.)
Date: October 19, 1959
Creator: Atz, R. W.
Partner: UNT Libraries Government Documents Department
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