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A Cladding Failure Monitor for Liquid Metal-Cooled Reactor Systems

Description: A cladding failure monitor using the principle of detection of rare gas fission products in the cover gas system of liquid metal-cooled reactors, was developed which efficiently discriminates against AR41. This discrimination is accomplished by electrostatic precipitation of the rare gas daughter nuclides; since K41, the daughter of Ar41, is not radioactive, the activity of the precipitation is chiefly due to decay of various Rb and Cs fission products. The monitor equipment is described. Results of monitor testing in EBR-1 are reported; a simulated fuel road failure experiment was made which shows that the charged-wire cover gas monitoring principle should be useful in other sodium-cooled fast reactors systems.
Date: October 1963
Creator: Smith, R. R. & Doe, B. B.
Partner: UNT Libraries Government Documents Department
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A Controlled-Environment Steam Corrosion Facility

Description: Abstract; Technical report describing a low-flow autoclave system developed for out-of-pile corrosion testing of materials in controlled environment steam up to 500 C. The system has been set up in triplicate to provide for the exposure of various zirconium alloys to steam at 300, 400, and 500 C. The oxygen and hydrogen of the steam were controlled at 25 ppm and 3 ppm, respectively, to simulate the gas conditions from radiolytic water decomposition found in a boiling water reactor. The autoclave internals were so designed to result in a temperature variation between specimens under test of less than 2C.
Date: October 1963
Creator: Nelson, W. B.
Partner: UNT Libraries Government Documents Department
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Design Review and Safety Analysis of Experiments in ORNL Research Reactors

Description: Since 1943 when the oak Ridge Graphite Reactor was put into operation, literally thousands of individual irradiation have been performed in this reactor and in other ORNL research reactors. Over the years there have been many minor incidents caused by experiments. Such incidents have provided a basis cor continued improvement in experiment design review and safety-analysis procedures. The reports lists the Design Review and Safety Analysis (1) objectives, (2) principles and rules of design, and (3) limits of application of the review process. The report includes a review of 19 incidents at ORNL research reactors.
Date: October 1963
Creator: Stanford, L. E. & Costner, R. A., Jr.
Partner: UNT Libraries Government Documents Department
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Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 4

Description: The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.
Date: October 1, 1963
Creator: Sorlie, T.
Partner: UNT Libraries Government Documents Department
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An Electrochemical pH-STAT

Description: An apparatus has been developed which controls automatically the acidity of unbuffered solutions in the region from pH 4 to 10. A potentiostat is used to control the potential of an inert electrode on which the hydrogen gas-hydrogen ion reaction occurs in a solution saturated with hydrogen gas. The inert electrode acts as both a sensing element and a regulating electrode for the control of acidity. Current from the potentiostat passes through the inert electrode and an auxiliary polarizing electrode in an external compartment separated from the main cell by a salt bridge or porous plate. Transients which occur during the regulating action are presented and analyzed. The electrochemical pH-stat may be used to measure corrosion rates. Limitations of the device are discussed and a modification is proposed which makes use of a differential amplifier instead of a potentiostat.
Date: October 1963
Creator: Posey, F. A. (Franz Adrian), 1930-; Morozumi, T. & Kelly, E. J.
Partner: UNT Libraries Government Documents Department
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Final Summary Report of the Gas-Cooled Reactor Experiment-1

Description: Report describing the Gas-Cooled Reactor test facility, its ongoing testing and evaluations of a test reactor, and its operating conditions and characteristics.
Date: October 1963
Creator: Chesworth, R. H.
Partner: UNT Libraries Government Documents Department
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Fuel Cycle Program Design and Fabrication of Special Assembly 10-L : Compacted Powder Fuel Rods Clad With 0.127-MM Wall Stainless Steel

Description: Technical report describing sixteen fuel rods clad with thin type 304 stainless steel and filled with vibratory compact powder UO2 that were fabricated and incorporated into a bundle for irradiation testing in the VBWR. The UO2 powders were tested for gas content. N2, CO, and H2 were the principal gases evolved by both type of UO2, but the arc-fused UO2 released about ten times as much gas as the Dyna Pak UO2. The amount of gas released was also a function of particle size and temperature. The gas evolution data were used to design the gas plenum to accommodate the absorbed gases along with the fission gases.
Date: October 1963
Creator: Ogawa, S. Y. & Williamson, N. E.
Partner: UNT Libraries Government Documents Department
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Fundamentals in the Operation of Nuclear Test Reactors: Volume 2, Materials Testing Reactor Design and Operation

Description: Second volume of reports on the operation of nuclear test reactors. It includes six chapters: engineering description fo the materials testing reactor, reactor control components and instrumentation, reactor control circuitry, reactor operation, reactor shutdown and tank work, and supplemental facilities at the materials testing reactor.
Date: October 1963
Creator: Phillips Petroleum Company
Partner: UNT Libraries Government Documents Department
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Health and Safety Laboratory Fallout Program Quarterly Summary Report: June 1, 1963 - September 1, 1963

Description: Report that summarizes multiple laboratories' reports on global fallout deposition. Reports include data on Strontium-90 deposition recorded by the Health and Safety Laboratory, data from other laboratories, related interpretive reports, and recent publications related to fallout.
Date: October 1, 1963
Creator: Hardy, Edward P., Jr.; Rivera, Joseph & Collins, William R., Jr.
Partner: UNT Libraries Government Documents Department
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High Power Density Development Project: Fourteenth Quarterly Progress Report, July-September 1963

Description: Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC. Five tasks are in progress: (1) Task 1A-High Power Density Fuel Development. The number of assemblies has been reduced to seven as a result of the failure of two pellet fuel assemblies. The average burnup of the group operating as of September 1 is 7500 MWD/T. (2) Task 1B-Fuel Fabrication Development. Assembly. Assembly 12S gave positive signals of being a leaker under the multi-type in-core sampler and was declared failed based on the in-core results and visual observation of a cracked rod. Modifications to the instrumented fuel assembly probes were made by removing the failed flow meter rotors to allow continued use of the flux detectors and thermocouples. Flux detectors and thermocouples performed properly after reactor start up. Flux wire tubes were found to be kinked such that their use was prohibited. (3) Task II-Stability, Heat Transfer and Fluid Flow. A series of noise recordings of fluxes, flows, and temperatures has been made at 91 MWt at the Big Rock Point plant. Preliminary analyses of some of the these records were made to obtain noise amplitude as a function of frequency. Thermocouple response tests were performed to verify the temperature measurement obtained during the steady-state noise tests at Big Rock. (4) Task III-Physics Development. Plans for achieving optimum performance from the Big Rock plant are being based on the concept of maintaining a fixed power shape throughout each operating cycle. The desired shape for the present cycle has been computed. Methods of selecting control rod patterns to maintain this shape are being investigated for use in the on-line computer. The computer was put on line during plant startup in August, and is presently performing …
Date: October 1963
Creator: Holladay, R. L.
Partner: UNT Libraries Government Documents Department
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In-Core Instrumentation Development Program Quarterly Progress Report June - September 1963

Description: Introduction: The objective of Project Agreement 22 is to determine the feasibility of covering the complete reactor neutron flux start range from 10(3) - 5 x 10(13) nv by using in-core chambers. The counting mode of operating will be used at low neutron fluxes and the root mean square voltage fluctuation mode will be used at high neutron flux levels. Experiments have been run utilizing various ion chambers, gases, gas pressures, voltage, and cables to measure sensitivities and range operating in the counting and RMS voltage modes. Theoretical discussions are presented showing how the RMS voltage is related to individual pulse at both amplifier input and output. Noise is also compared at amplifier output so that the optimum bandwidth can be selected. Spectral shifts with changes in applied voltage causing signal variations have been examined and can be eliminated by appropriate selection of amplifier bandwidth. In the counting mode, all experiments have been conducted with unterminated cable. The chamber has been designed with geometry, gas, and pressure to completely stop fission fragments in the gas and hence maximize the charge generated in the chamber. Cables have been selected to minimize capacity. Various gases, pressures, and voltages have been used to determine that an optimum design has been achieved.
Date: October 1963
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department
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Operational Control Rod Reactivity Worths From Observed Heat Generation Rates

Description: Introduction. The reactivity difference associated with a reactor change can be simply related to the coincident changes in the neutron loss and generation rates. Unfortunately, in many instances these rates are difficult to measure directly during high-level operation; thus relativity values are normally found by other methods such as buckling calculations or low-level rising period measurements. However, with certain applicable control rod systems, it may be feasible to use heat generation rate in the rods as a measure of the reactivity-compensation effect. The neutron absorption rate in the Hanford reactor control rods can be determined under equilibrium conditions (and without disturbing these conditions) from the heat transfer rate to the control rod coolant. This information, when combined with a measurement of the change in reactor leakage caused by rod insertion, allows the calculation of control rod strength.
Date: October 1963
Creator: Fredsall, J. R. & Bowers, C. E.
Partner: UNT Libraries Government Documents Department
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Prediction of Two-Phase Critical Flow Rate

Description: Technical report of a proposal of an analytical model to predict two-phase critical flow rate. The model is based upon thermal equilibrium, a "lumped" treatment of the two-phase velocity (each phase is represented by a single mean velocity), and upon the neglect of frictional and hydrostatic pressure losses. A comparison, of the proposed predictions with available test results and previous analyses shows that: (1) The present model agrees very well with the published test data. (2) In contrast to all other analyses, the model requires no assumption about the gas void fraction.
Date: October 1963
Creator: Levy, S.
Partner: UNT Libraries Government Documents Department
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Reactivity Control Problems in the Plutonium-Fueled EBR-1 Reactor

Description: In contrast with previous EBR-1 reactor cores which were fueled with enriched uranium, the current Mark IV core is a 7 x 8-in. cylinder made of delta-stabilized plutonium alloyed with 1 1/4 w/o aluminum. The reactivity of the reactor is largely controlled by the movement of the entire outer blanket mounted on a hydraulic elevator with a travel of 80 inches. Partial meltdown occurred in November, 1955 during the last experiment scheduled for the core which was directed toward identifying the time constants associated with the components of the over-all reactivity coefficients. The incident is reported elsewhere. The changes in reactivity apparently come about as a result of changing from operation on the high temperature system to the low temperature system or vice versa. Continuous operation from day to day on either system does not effect any significant change in reactivity. Reactivity is not dependent on the duration of a run, but rather on the maximum power attained during that run.
Date: October 1963
Creator: Haroldsen, R. O. (Ray Ottley), 1928-; McGinnis, F. D. & Smith, R. R.
Partner: UNT Libraries Government Documents Department
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Sorption Measurements in Ultrahigh Vacuum at Constant Pressure

Description: Introduction. Sorption of gases by surfaces is the primary step in many heterogeneous processes. Because sorption processes are in general pressure-dependent, and very fast at normal pressures, low pressure system are mandatory. In gas-metal interaction studies the flash filament technique with ribbons and filaments and Wagener's technique with evaporated films allow observations at the requisite low pressures. However, these method have two major drawbacks: (1) Since the pressure changes during the sorption experiment, pressure-dependent parameters can only be obtained from indirect evidence; (2) The pressure change in the sorption cell during the experiment may cause significant interaction between the sample and the rest of the system and thus experimental data must be corrected for this effect. These drawbacks have been eliminated in a new approach based on a flow system in which the sample is exposed to constant pressure. This new technique has been employed for the study of adsorption and absorption of gases by filaments and evaporated films. This method can also be used for studying gas-metal solution equilibria.
Date: October 1963
Creator: Gibson, Richard.; Bergsnov-Hansen, B.; Endow, Noboru. & Pasternak, R. A.
Partner: UNT Libraries Government Documents Department
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Specific Zirconium Alloy Design Program Quarterly Progress Report: Sixth Quarter, July - September, 1963

Description: Summary: Fundamental studies in support of the alloy design work are complete except for the experimental determination of the diffusion of oxygen in alloy-doped non-stoichiometric ZrO2. Over 100 oxidation runs have now been made on samples of ZrO2 doped with 1 mole percent of the oxides of Al, Y, Fe, Cr, and Ni. The first round testing of 31 alloys is now essentially complete. Analysis of the steam corrosion rate and hydriding raw data taken at 300, 400, and 500 degrees C indicates that Zr-Cr and Zr-Cu-Fe alloys show the most promise for development for service in steam over the entire temperature range 300-500 degrees C. Maximum resistance to corrosion hydrogen embrittlement requires high initial ductility and thus low, perhaps less than ~2.5 a/o total alloy content. For any composition, susceptibility to hydrogen embrittlement depends on crystallographic texture of the component; under certain circumstances hydrogen embrittlement may be high anisotropic. The second-round testing of 10 selected Zr-Cr and Zr-Cu base alloys is now about 50% complete. Three alternate fabrication schedules were evaluated; and the preliminary results indicate that the Zr-Cu alloy tested is less sensitive to heat treatment than is the Zr-Cr alloy tested. Raising the final alpha annealing temperature from 565 degrees C to 788 degrees C gives better over-all corrosion and hydrating performance for both the Zr-Cr and Zr-Cu alloy tested. Beryllium additions to Zr-Cr or Zr-Cu do not appear to be advantageous. Nickel additions to Zr-Cu do not give an over-all improvement. Nickel additions to Zr-Cu give about the same improvement over Zr-Cu as did iron additions to the Zr-Cu in the first-round test.
Date: October 1, 1963
Creator: Klepfer, H. H.; Jaech, John L.; Douglass, D. L. (David Leslie), 1931-; Blood, R. E. & Perrine, H. E.
Partner: UNT Libraries Government Documents Department
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A System for Measurement of Wall Thickness of Corroded Aluminum Reactor Process Tubes

Description: A sector gauge was developed for routine measurement and recording of wall thicknesses between the ribs and at the top of installed aluminum reactor power tubes. The basic criteria selected for the device were that it would measure and record wall thickness over the length of the tube with an accuracy of plus or minus 2 mils at an average rate of 3 min per tube. An eddy-current measuring system was used in the device.
Date: October 1963
Creator: Dulin, Ralph V.
Partner: UNT Libraries Government Documents Department
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Training of Consumers Public Power District Personnel for the Operation of the Hallam Nuclear Power Facility

Description: Training of Consumers Public Power District (CPPD) operating personnel for the Hallam Nuclear Power Facility (HNPF) was carried out under two formalized training programs. Both programs were organized and directed by Atomics International (AI). The first program was conducted in 1960 while he HNPF was under construction. The second program was begun in September 1961, prior to the initial HNPF dry critical loading experiment, and was completed in February 1963. The conventional portion of Sheldon Station has been a commercial power plant since July 1, 1961. Sheldon Station CPPD personnel were utilized extensively as instructors for the two formalized HNPF training programs and as responsible engineers for numerous test of the HNPF. CPPD shift personnel constituted the principal operating force throughout HNPF construction, reactor startup, and testing.
Date: October 1963
Creator: Loomis, J. S.
Partner: UNT Libraries Government Documents Department
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Transition Boiling Heat Transfer Program; Third Quarterly Progress Report, July - September 1963

Description: Summary: Initial critical heat flux, transition boiling temperature fluctuation, and film boiling coefficient data have been obtained on a two-rod cluster assembly at 1000 psia and 25 to 90 percent steam qualities. A representation showing the range of critical heat flux data is presented. Typical temperature recordings which indicate transition and film boiling behavior are shown. Fabrication of a new high pressure observational test section is nearly complete. An optical table and illumination system has been build and operationally tested for photographic use on the new observational section.
Date: October 1, 1963
Creator: Quinn, E. P.
Partner: UNT Libraries Government Documents Department
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