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Analysis of the Reactivity Characteristics of Yankee Core I

Description: The reactivity characteristics of the operating Yankee Core I are analyzed. Calculations of kinetic parameters, kinetic coefficients, control rod and boron worth, core lifetime and burnup rate, and fission product poisoning, are described. A large amount of experimental data obtained during Core I operation is included and comparisons are made between prediction and experiments. (auth)
Date: January 1, 1963
Creator: Poncelet, C. G.
Partner: UNT Libraries Government Documents Department

An Annotated Bibliography of Analytical Methods for Alkali Metals

Description: A total of 107 abstracts is presented on analytical methods for alkali metals, as a part of a program for the evaluation of the performance of the primary cold trap from the Enrico Fermi Reactor. The abstracts are arranged into sections dealing with general aspects; sampling and dissolution techniques; and determination of uncombined alkali metal, oxygen, carbon, hydrogen, nitrogen, and other impurities. (D.L.C.)
Date: March 1, 1964
Creator: Garcia, E. E. & LaMont, B. D.
Partner: UNT Libraries Government Documents Department

An Evaluation of the Sulfuric Acid-Sodium Nitrite Etch for Zircaloy-2

Description: Preliminary experiments indicate that there are no significant differences in the corrosion rates of zirconium or Zircaloy-2 after etching with the nitric--hydrofluoric solution or the sulfuric--nitrite solution, provided proper etching and washing techniques are followed. Incomplete removal of the residual etchant is deleterious to the corrosion resistance; however, this effect in the case of the sulfuric--nitrite solution is not as pronounced as in the case of the nitric--hydrofluoric acid solution. The anticipated advantages in the new etch were not completely realized. Additional development aimed at modifying the sulfuric--nitrite etch would have to be performed in order to overcome the disadvantages before recommendation for the adoption of the etch could be made. (auth)
Date: February 17, 1954
Creator: Kass, S.
Partner: UNT Libraries Government Documents Department

Pressurized Water Reactor Program Technical Progress Report for the Period July 15, 1954 to August 26, 1954

Description: This progress report has 2 parts. Part 1 PWR Engineering covers the following topics: (1) power plant analysis and systems; (2) power plant components and component materials and tests; and (3) reactor and auxiliaries. Part 2 PWR Development covers: (1) fuel element development; (2) metallurgy of core materials; (3) materials application development; (4) chemistry development; (5) irradiation effects; and (6) reactor physics.
Date: October 31, 1957
Partner: UNT Libraries Government Documents Department

Pressurized Water Reactor Program Technical Progress Report for the Period May 5, 1955 to June 16, 1955

Description: The current PWR plant and core parameters are listed. Resign requirements are briefly summarized for a radiation monitoring system, a fuel handling water system, a coolant purification system, an electrical power distribution system, and component shielding. Results of studies on thermal bowing and stressing of UO/sub 2/ are reported. A graph is presented of reactor power vs. reactor flow for various hot channel conditions. Development of U-- Mo and U-Nb alloys has been stopped because of the recent selection of UO/sub 2/ fuel material for the PWR core and blanket. The fabrication characteristics of UO/sub 2/ powders are being studied. Seamless Zircaloy-2 tubing has been tested to determine elastic limits, bursting pressures, and corrosion resistance. Fabrication techniques and tests for corrosion and defects in Zircaloy-clad U-Mo and UO/sub 2/ fuel rods are described. The preparation of UO/sub 2/ by various methods is being studied to determine which method produces a material most suitable for PWR fuel elements. The stability of UO/sub 2/ compacts in high temperature water and steam is being determined. Surface area and density measurements have been performed on samples of UO/sub 2/ powder prepared by various methods. Revelopment work on U-- Mo and U--Nb alloys has included studies of the effect on corrosion behavior of additions to the test water, additions to the alloys, homogenization of the alloys, annealing times, cladding, and fabrication techniques. Data are presented on relaxation in spring materials after exposure to a corrosive environment. Results are reported from loop and autoclave tests on fission product and crud deposition. Results of irradiation and corrosion testing of clad and unclad U--Mo and U-Nh alloys are described. The UO/sub 2/ irradiation program has included studies of dimensional changes, release of fission gases, and activity in the water surrounding the samples. A review of the methods of ...
Date: October 31, 1958
Partner: UNT Libraries Government Documents Department

Pressurized Water Reactor Program Technical Progress Report for the Period September 9 to October 20, 1955

Description: Progress in the design, development, and construction of PWR power plant systems and components and PWR core and auxiliaries is summarized. The blanket assembly design is described and illustrated. Results of MTR evaluation of fuel element failure instrumentation are reported. Development of fabrication and testing tochaiques for clad fuel elements, fuel rods, plates, and assemblies is described. Investigations of fuel and cladding alloys include crystal structure and thermal stability determinations on U--Mo alloys, studies on the nature of the hydride phase formed during corrosion of gamma -phase alloys in high- temperature water, and specific heat, resistivity, and phase diagram studies of U- -Mo and U--Nb alloys. The equilibrium and kinetics in the system UO/sub 2/--O/ sub 2/ are being studied to gain information on the structure and stability of UO/ sub 2/ under various conditions. Results of irradiation tests on UO/sub 2/ samples and of thermal cycling tests of Zircaloy-2 clad UO/sub 2/ rods are reported. Corrosion test resuIts are summarized for unclad and Zircaloy-2 clad U- - Mo and U--Nb samples. The radiation induced volume change of prototype fuel reds has been investigated. Studies of the fabrication cladding, tensile properties, and corrosion of U-- Si systems are described. Corrosion tests are continuing on Zircaloy-2 clad U-- Zr fuel elements and on various experimental Al alloys for cladding applications. Work was continued on the preparation, corrosion and sinterability of pure UO/sub 2/ and UO/sub 2/ containing additives. Operation and chemical analysis of in-pile loop experiments are described. Results are reported from studies of the erosion of UO/sub 2/ in high-velocity coolant, decontamination of water by ion exchange resins, sorption of radioisotopes on stainless steel, and decontamination of corrosion loops. Work in reactor physics has included PWR control calculations using a 2-dimensional UNIVAC code, calculation of fission product activity in the ...
Date: October 31, 1958
Partner: UNT Libraries Government Documents Department

Sodium Pump Development and Pump Test Facility Design

Description: The study defines a program for the development of large sodium pamps for use with sodium cooled reactor systems of 1000 to 1500 Mw(e) capability, and the functionai requirements of a related sodium pump test facility for testing large pumps. The future pump requirements of large power systems have been estimated, a type of sodium pump recommended for farther development, the development problems identified and a program research and developmert prepared to resolve these problems. The functional requirements of a sodium pump test facility for testing pumps for large reactor use have been established. (auth)
Date: August 1, 1963
Partner: UNT Libraries Government Documents Department

Survey of Sodium Pump Technology

Description: A review is presented of the current status of sodium pump development as related to nuclear power applications. A description is given of the design features and performance characteristics of the more important types of sodium and sodium-- potassium alloy (NaK) pumps. Some requirements for sodium pumps for future large liquid metal reactor systems are presented with some preliminary consideration of the potential of various pump types to meet these requirements. (auth)
Date: June 1, 1963
Creator: Nixon, D. R.
Partner: UNT Libraries Government Documents Department

Upflow Burnout Data for Water at 2000, 1200, 800, and 600 Psia in Vertical 0.070 In. X 2.25 In. X 72 In. Long Stainless Steel Rectangular Channels, Corrections

Description: Upflow burnout data for water at 2000, 1200, 800, and 600 psia in vertical 0.070 in. by 2.25 in. by 72 in. long stainless steel rectangular channels are presented. The burnout heat fluxes for these channels depend upon the same parameters as did previously tested channels with much smaller L/D ratios. Bettis burnout design equations were used in calculations. (J.R.D.)
Date: July 1, 1958
Creator: Troy, N.
Partner: UNT Libraries Government Documents Department

Valve Operating System System Description No. 10

Description: A description is given of the valve operating system for the PWR. The valves served by this system are a component pant of the following systems: reactor coolant system; pressurizer and pressure relief system; coolant discharge and vent system; and failed element detection and location system. (W.L.H.)
Date: May 1, 1957
Partner: UNT Libraries Government Documents Department

Variable Flow Resistance With Adjustable Multipenetration Orifice Plates in Series

Description: Full-scale water flow tests were performed to determine the feasibility of a variable resistance orificing scheme for PWR Core 2. A low temperature, low pressure loop was used, and three, two, and one multi-hole and multi-slot orifice plates in series were tested. For the three and two plates in series, thc range in flow repenetrations were aligned to the condition of maximum misaligament. Flow characteristics of a single multihole and multi-slut plate were also studied. Tests performed on the multi-hole and multi- slut orifices in series showed that a considerable renge of flow resisttions is varied. For the area ratio, which gave the 6 in the case of three multi-hole plates and about 1 to 4 for three multislot plates. Fur the multi-hole geometry tested, the incrcase in flow resistance with angular misalignment was very high and occurred over an angular rotation of the center plate of 15 deg . Employing angular slots allowed greater freedom in the choice of the range of angular rotation to go from the minimum to the maximum flow resistance conditions. For this reason the angular slot configuration will be considered for further study for application in PWR Core 2. (auth)
Date: October 17, 1958
Creator: Grochowski, F. A.
Partner: UNT Libraries Government Documents Department

Waste Disposal Treatment of Pwr Hot Laundry and Decontamination Room Wastes. Appendix 1: Survey of Application of Standard Water Clarification Procedures to Pwr Laundry and Decontamination Room Wastes. Appendix Ii: Conference Between R. Lloyd and j.r. Pointe to Establish Tentative Procedures and Determine Equipment for Applying Adsorption-Flocculation Treatment to Pwr Laundry W

Description: This report and three appendixes were issued separately, but are cataloged as a unit. The necessity for treatment of hot laundry and decontamination room wastes prior to disposal at the out, and means for accomplishing this are discussed. A feasible procedure suggested consists of an adsorptionflocculation treatment with supernate disposal by dilution, pass through an evaporator, transfer to surge and decay tanks, with final sludge concentration in drums for retention and burial at sea. (T.R.H.)
Date: March 23, 1956
Creator: Cohen, P.; Lloyd, R.; LaPointe, J.R. & Abrams, C.S.
Partner: UNT Libraries Government Documents Department

Manufacture of Uranium-Niobium Fuel Rods for Irradiation in the Materials Testing Reactor

Description: Zircaloy-2 clad fuel rods whose fuel was U/sup 238/ -10 wt.% Nb-5 wt.% U/ /sup 235/ were fabricated for testing in the Materials Testing Reactor. The fuel was duplex melted and given a homogenizing heat treatment prior to assembly into extrusion billets. The rods were prepared by co-extruding the fuel alloy in Zircaloy-2 cups, then drawing the rods to size. (auth)
Date: September 16, 1955
Creator: Haynes, W. B.
Partner: UNT Libraries Government Documents Department

Mechanical and Thermal Problems of Water-Cooled Nuclear Power Reactors

Description: Some of the principal problems faced in the mechanical and thermal design of the Shippingport Pressorized Water Reactor core are discussed. The interplay of these problems with the requirements of other technologies is discussed, and areas which need more work are outlined. (M.H.R.)
Date: January 1, 1955
Creator: Palladino, N. J. & Sherman, J.
Partner: UNT Libraries Government Documents Department