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0-2 kv Flash Tube Supplies

Description: The power supplies designed and constructed to power high-intensity flash tubes are described. Three supplies, capable of charging 100 mfd to 2 kv with a repetltion rate of not less than 0.8 sec, are operated remotely from a control panel containing 3 powerstats. The power supplies are full wave center trapped rectifiers employing silicon rectifiers. (M.C.G.)
Date: March 15, 1962
Creator: Miller, D. M.
Partner: UNT Libraries Government Documents Department
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6 kv CAPACITOR CHARGING SUPPLY

Description: The power supplies designed and constructed to power high intensity flash tubes used in bubble chamber experiments are briefly described and are accompanied by a schematic diagram of the layout. (D.C.W.)
Date: March 15, 1962
Creator: Miller, D. M.
Partner: UNT Libraries Government Documents Department
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20-ton HE Cratering Experiments in Desert Alluvium: Final Report, May 1962

Description: From abstract and summary: Project Stagecoach consisted of the detonation of three 40,000-pound charges. Blocks of cast TNT were stacked to resemble a sphere and, the whole center-detonated.
Date: March 1960
Creator: Vortman, Luke J. & MacDougall, H. R.
Partner: UNT Libraries Government Documents Department
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105-C overboring thirteen tube outage, March 6, 1961--March 10, 1961

Description: C Reactor was shut down on a scheduled basis at 8:30 a.m. March 6, 1961 for the purpose of overboring 17 process channels. this report will cover that outage and discuss problems encountered in completing the tasks involved in overboring.
Date: March 24, 1961
Creator: Munro, C. A.
Partner: UNT Libraries Government Documents Department
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A 200-Watt Conduction-Cooled Reactor Power Supply for Space Application

Description: The limited supply of relatively long-half-life isotopes having a reasonably high power density and the low conversion efficiencies obtainable with thermoelectric devices have so far limited the power output of isotope-fueled sources of electric power to several tens of watts. In addition, the high cost of the available isotopes results in a very large expense for isotope-fueled generators producing several hundred watts. It appears that a small, minimumweight, conduction-cooled reactor is an attractive alternate to the isotope-fueled power supplies in the 200-w size range. The proposed reactor is a small, high-density fast core of U/sup 233/ surrounded by a beryllium reflector. This approach, generally speaking, gives a reactor that is more compact and of lighter weight than can be obtained with a moderated system having a softer neutron spectrum. In the reactor design, the path of heat flow is from the core to the inner reflector and then to the thermoelements in close contact with the inner reflector. The reject heat flowing from the thermoelement cold junctions enters the outer pontion of the reflector, which acts as the heat sink and conducts the reject heat to the large, circular, tapered-fin radiator which is attached to the reflector. Survey physics calculations for various reactor systems fueled with U/sup 235/, U/sup 233/, and Pu/sup 239/ are reported. Some limits imposed on the system design by the thermoelectric generator are discussed, and the problem of radiator design for the space environment is treated in some detail. No attempt is made to present a detailed final design of the power supply; rather, the report is restricted to a general delineation of the limits imposed by various parameters and a resulting final conclusion as to the performance limits of small conduction-cooled reactors in this size range. (auth)
Date: March 1, 1963
Creator: MacFarlane, D. R.
Partner: UNT Libraries Government Documents Department
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500-Thgc--a 500 Node Transient Heat Transfer Code

Description: This document is a user manual for those who are familiar with problems 500 Node Transient Heat Transfer Code.
Date: March 27, 1964
Creator: Blaine, R. A. & Berland, R. F.
Partner: UNT Libraries Government Documents Department
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710 reactor program, progress report No. 13

Description: Declassified 4 Sep 1973. Information on the development of the 710 Reactor is presented concerning the performance testing of refractory-metal fuel elements, critical experiment mockup of 710 Reactor, reactor component design and development, and test facilities and pilot loop design, (DCC)
Date: March 31, 1965
Partner: UNT Libraries Government Documents Department
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A 1000 HOUR TEST OF A CORROSION PRODUCT SEPARATOR UNIT IN A HAYNES ALLOY NO. 25 LOOP CONTAINING MERCURY--TEST G-16

Description: One of the major unsolved problems affecting the life of the SNAP II Power Conversion System is the mass transfer of corrosion products by mercury and subsequent deposition. It is feared that the corrosion products might tend to accumulate in critical areas such as orifices, bearings and so forth. Therefore, this test was conducted to evaluate a corrosion product separator and to determine the influence of corrosion product removal on corrosion rate. The corrosion product separator was successful in removing 85 percent of the elements corroded from the container walls. The loop and separator, both fabricated from Haynes alloy No. 25, operated for 1000 hours. The mercury was boiled and condensed at 1100 deg F, superheated to 1190 deg F and subcooled to 325 deg F. The flow rate in this loop was much higher than in previous loops, being approximately 37 pounds of mercury per hour as contrasted with approximately l2 pounds of raercury per hour. No increase in corrosion rate was noticed as a result of the higher flow rates and velocity or by the removal of corrosion products. If this type of separator or an improved type works equally as well in the final application, the danger of failure from corrosion products should be greatly reduced. (auth)
Date: March 1, 1962
Creator: Nejedlik, James F.
Partner: UNT Libraries Government Documents Department
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1000 Mwe Closed Cycle Water Reactor Study

Description: This report has two volumes, volume 1 contains the summary and detailed description of plant design, volume 2 contains a comprehensive nuclear evaluation of the reactor core.
Date: March 1, 1963
Creator: Westinghouse Electric Corp., Pittsburgh, Pa. Atomic Power Div.
Partner: UNT Libraries Government Documents Department
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1000 MWE Closed Cycle Water Reactor Study Volume II

Description: This report includes the nuclear evaluation that has been conducted for the purpse of studying those problem areas which are expected to increase in severity as the core size is increased to produce 1000 MWE.
Date: March 1, 1963
Partner: UNT Libraries Government Documents Department
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14CO2 INCORPORATION INTO THE NUCLEIC ACIDS OF SYNCHRONOUSLYGROWING CHLORELLA CELLS

Description: A study of the incorporation of {sup 14}CO{sub 2} into cell components of synchronously growing Chlorella pyrenoidosa has shown that DNA is synthesized primarily during the latter stages of the cell cycle prior to cell division. RNA was synthesized at an approximately equal rate during each of the three phases of the cell growth studied. No major differences were noted in the incorporation of {sup 14}CO{sub 2} into the soluble cell components in these long-term incorporation studies.
Date: March 8, 1962
Creator: Stange, Luise; Kirk, Martha; Bennett, Edward L. & Calvin, M.
Partner: UNT Libraries Government Documents Department
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2000 KILOWATT SODIUM TEST FACILITY. Project Description and Progress Reports, June 30, 1958 through September 30, 1959

Description: The design and construction work completed on the 2000-kw Sodium Teat Facility during the period from April 1958 to Oct. 1959 is described. The purpose of the facility project in to test models of equipment components which are to be used in a moltsn plutonium, sodiumcooled fast reactor. (D. L.C.)
Date: March 1, 1961
Creator: Whinery, L.A.
Partner: UNT Libraries Government Documents Department
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630A MARITIME NUCLEAR STEAM GENERATOR. Progress Report No. 7

Description: This progress report covers the period from November 15, 1963 to February 15, 1964. A study indicated that the most desirable type of blower drive turbine is one using main turbine throttie steam conditions and exhausting to the main turbine cross-over line. Preliminary planning for the initiation of a dynamic structural analyses of the overall steam generator was completed. External pressure loading and thermal stress calculations show that the calandria has a suitable design margin. A revised fuel latch operable from the rear face of the core was designed. A study was initiated to determine the feasibility of substituting Zircaloy for the stainless steel tubing within the active core. Preliminary sizing of control rod extensions and gang plates was completed. Initial loading of the second configuration of the 630A critical experiment reactor was completed. Detailed power distributions were measured in the 11 typical positions. Subcritical and critical rod worth curves were obtained in the critical experiment with up to 132 shim rods in the core. Moderator temperature coefficient measurements were made and agreed well with analytical data. Critical experiment correlation of fine radial power calculations in the revised mock-up showed good agreement. Performance specifications were prepared for a 1-Mw(e) power plant. Parametric thermal analyses, based on various tube sheet thicknesses and heating rates, were completed for the top and bottom tube sheets. The shield plug was redesigned to accommodate additional shim rods and to facilitate fabrication and moderator flow adjustments. Calculations of the operating and shutdown dose at the nuclear sensor location was started. Studies were performed to determine the size of the port openings needed to prevent buckling in the containmert vessel in the event of ship sinking. A blower shaft static seal was built and tested with satisfactory results. Manufacture and procurement of all parts for two developmental …
Date: March 16, 1964
Partner: UNT Libraries Government Documents Department
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ABRAC--AN IBM-704 THREE DIMENSIONAL NUCLEAR-THERMAL DEPLETION PROGRAM WITH DISTRIBUTED VOID EFFECTS

Description: ABRAC is a three dimensional nuclear thermal depletion program to study the effects of water moderator density changes, resulting from flow variations and boiling, on neutron flux distribution and depletion. The program requires an IBM-704 with a memory of 32.768 words and ten tape units. (auth)
Date: March 1, 1960
Creator: Jacobi, W. M.; Lawton, T. J.; Meanor, S. H. & Parrette, J. R.
Partner: UNT Libraries Government Documents Department
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THE ABSOLUTE ABUNDANCE OF THE CHROMIUM ISOTOPES IN SOME SECONDARY MINERALS

Description: Isotopic assays have been made on the Cr in samples from 14 different chrominiferous minerals from different geographic and meteoritic sources. The results of the assays indicate that it is not possible to state unequivocally that variations in isotopic composition have been observed. (auth)
Date: March 14, 1962
Creator: Svec, H.J.; Flesch, G.D. & Capellen, J.
Partner: UNT Libraries Government Documents Department
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Abstract -- A description of low temperature hydriding of zircaloy at Hanford

Description: The two K Reactors at the Hanford Plant of the Atomic Energy Commission were retubed in 1963 with ribless Zircaloy process tubes. These tubes, 1.81 inch OD x 0.037 inch wall, were made by tube reducing methods from extruded blooms and were installed without final pickling or autoclaving. These tubes are used with once-through filtered water with a maximum coolant temperature of approximately 110 C. After approximately two years` service, it was discovered that these tubes were absorbing hydrogen in a totally unexpected and alarming manner. It was found that the downstream few feet of these tubes had a layer of massive hydride platelets, several mils thick, on the inner surface, and that the hydrogen content of the tube wall beneath this layer had increased from the as-fabricated level of 5--15 ppm to 50--75 ppm. The portion of the tubes so affected was downstream of the fuel charge. Physical testing has indicated loss of ductility but no significant changes in strength, Cold burst tests resulted in a ductile failure mode with no tendency toward brittle fracture. A program to determine the basic mechanism responsible for this phenomenon as well as practical means for its elimination will be the subject of a companion paper.
Date: March 31, 1966
Creator: Alexander, W. K.
Partner: UNT Libraries Government Documents Department
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Accurate Nuclear Fuel Burnup Analyses; Ninth Quarterly Progress Report, (December 1963 - February 1964)

Description: The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Date: March 1, 1964
Creator: Rider, B. F.; Peterson, J. P., Jr.; Ruiz, C. P. & Smith, F. R.
Partner: UNT Libraries Government Documents Department
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Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 2. Plant Conceptual Studies

Description: Second part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. II-i).
Date: March 1962
Creator: Combustion Engineering, inc. Nuclear Division.
Partner: UNT Libraries Government Documents Department
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An Advanced Sodium-Graphite Reactor Nuclear Power Plant

Description: Abstract: This report describes an advanced sodium-cooled, graphite-moderated nuclear power plant which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency.
Date: March 15, 1960
Creator: Churchill, J. R. & Renard, J.
Partner: UNT Libraries Government Documents Department
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AN ADVANCED SODIUM-GRAPHITE REACTOR NUCLEAR POWER PLANT

Description: An advanced sodium-cooled, graphite-moderated nuclear power plant is described which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency. Steam is generated at 2400 psig, superheated to 1050 deg F and, after partial expansion in the turbine, reheated to 1000 deg F. Net thermal efficiency of the plant is 42.3%. In a plant sized to produce a net electrical output of 256 Mw, the estimated cost is 8232/kw. Estimated cost of power generation is 6.7 mills/kwh. In a similar plant with a net electrical output of 530 Mw, the estimated power generating cost is 5.4 mills/ kwh. Most of the components of the plant are within the capability of current technology. The major exception is the fuel material, uranium carbide. Preliminary results of the development work now in progress indicate that uranium carbide would be an excellent fuel for high-temperature reactors, but temperature and burnup limitation have yet to be firmly established. Additional development work is also required on the steam generators. These are the single-barrier type similar to those which will be used in the Enrico Fernri Fast Breeder Reactor plant but produce steam at higher pressure and temperature. Questions also remain regarding the use of nitrogen as a cover gas over sodium at 1200 deg F and compatibility of the materials used in the primary neutron shield. All of these questions are currently under investigation. (auth)
Date: March 15, 1960
Creator: Churchill, J. R. & Renard, J.
Partner: UNT Libraries Government Documents Department
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ADVENTURES IN NUCLEAR PHYSICS

Description: No Description Available.
Date: March 1, 1962
Creator: Alvarez, Luis W.
Partner: UNT Libraries Government Documents Department
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AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel

Description: Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter of the fuel rods is 1.308 cm (0.515 inch), and the clad wall thickness if 0.051 cm (0.020 inch). the cladding material is Type-304 stainless steel. The fuel pellets were all centerless ground to achieve a uniform outside diameter and thereby control the pellet-to-clad diametral clearance within a range of 0.076 to 0.102 mm (0.003 to 0.004 inch). Operation of the fuel rods will be at high specific power levels with surface heat fluxes of about 157 W/cm(2) (~500,000 Btu/h-ft(2)). The assembly was designed for a lifetime of 4.1 x 10(20) fission/cc (15,000 MWD/T) exposure.
Date: March 1964
Creator: Ogawa, S. Y.
Partner: UNT Libraries Government Documents Department
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