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The Thermal Conductivity of Uranium Monocarbide

Description: Uranium carbide shows promise as a fuel material for reactors operating at relatively high temperatures based on its high melting point, high uranium density and high thermal conductivity. Before refined reactor designs can be made, however, good quantitative data on the thermal conductivity at temperatures in excess of 1000C is required. This technical report presents data gathered as part of a continuing study aimed at determining the thermal conductivity of refractory uranium fuels as a function of temperature, density and composition over the temperature range 1000-2200C. At the inception of this program it was felt that an absolute method capable of achieving high temperatures was necessary and that the difficulties encountered in fabricating the large complex specimens needed were justified. The steady state radial heat flow method and apparatus of Rasor and McClelland were therefore chosen. The technical report discusses the experimental equipment and presents results of measurements on three specimens of UC over a temperature range 900 to 1600C. An analysis of the data is made with respect to other physical properties of the material and the measured conductivities are compared with the work of other investigators.
Date: April 2, 1964
Creator: Sobon, J. T.; Miller, A. D. & DeCrescente, M. A. (Michael A.)
Partner: UNT Libraries Government Documents Department
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Removal of Radioisotopes From Solution by Earth Materials From Eastern Idaho

Description: Abstract: Naturally occurring earth materials from Idaho, primarily from localities near the National Reactor Testing Station (NRTS), were used in laboratory tests for the removal of radioisotopes from aqueous solutions. These earth materials included lignitic deposits, clay-like materials, and specific minerals; ion exchange resins were also considered for a specific application. The aqueous solutions were low-level radioactive cooling water or synthetic solutions made up to represent low-level radioactive wastes at the NRTS. Cation exchange capacities and other properties which affect the removal of radioisotopes from solution were determined the cation exchange capacities varied from 0.006 to 1.0 meq/g of solid. Earth materials with cation exchange capacities greater than 0.3 meq/g, in general, had distribution coefficients in excess of 1000. The highest distribution coefficients for cesium and strontium occurred in the pH range from 6.0 to 9.0 The possible use of these materials for decontaminating low-level radioactive waste at the NRTS is discussed. The result of laboratory studies using these materials and an organic ion exchange resign for decontaminating a specific NRTS waste are given. A material high in clinoptilolite from a location near the NRTS was considered to be the most promising material for use in large beds or ion exchange-type columns.
Date: April 1964
Creator: Wilding, M. W. & Rhodes, D. W. (Donald Walter), 1919-
Partner: UNT Libraries Government Documents Department
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High Power Density Development Project: Sixteenth Quarterly Progress Report, January-March 1964

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, (2) Task 1B-Fuel Fabrication Development. Assembly, (3) Task II-Stability, Heat Transfer and Fluid Flow, (4) Task III-Physics Development, and (5) Task IV-Co-Ordination and Test Planning.
Date: April 1, 1964
Creator: Holladay, R. L.
Partner: UNT Libraries Government Documents Department
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Fabrication of fuel Cladding From Incoloy Alloy 800 : an Evaluation of Methods

Description: Summary: On the basis of its high temperature, physical and corrosion properties, Incoloy Alloy 800 was selected as a candidate for fuel cladding nuclear superheat applications. At the time of its selection, there was little information or experience with Incoloy 800 in the production of thin-walled, small diameter tubing suitable for nuclear fuel cladding. As a result, special purchasing efforts were required for the procurement of initial tubing used in fuel fabrication. As-received welded and drawn tubing proved to be generally good but showed some conditions which were undesirable, the major one being lack of complete recrystallization and homogenization of the weld zone. The possible effect of this condition upon the fuel performance was not immediately known; however, subsequent development work indicated that the non-homogeneity of the weld could affect adversely its mechanical and corrosion properties in relation to the parent metal. A development program was initiated to determine treatment sequences suitable for the fabrication of welded and drawn tubing with a fully recrystallized and homogenized weld structure. This was accomplished by butt welding lengths of Incoloy strip which were subsequently cold rolled and annealed to simulate tube fabrication steps. Requirements imposed on this work were that all processes developed must be amenable to normal production equipment and procedures used in commercial tube manufacturing. As a result of the experiments undertaken, a sequence of cold drawing and annealing steps was established suitable for recrystallization and homogenization of weld zones in welded-and-drawn tubing. In order to obtain complete sequential chemical and metallurgical history of nuclear grade Incoloy 800 tubing, a 3000-pound ingot was purchased to the required chemical specifications and reduced to tubing through both seamless and welded drawn routes. The management of the sequential metallurgical steps needed for tubing fabrication resulted in establishing a workable process through a series of …
Date: April 1964
Creator: Kirby, R. F.; MacMillan, D. F. & Punches, J. R.
Partner: UNT Libraries Government Documents Department
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The Effects of Non-Uniform Flow and Concentration Distributions and the Effect of the Local Relative Velocity on the Average Volumetric Concentration in Two-Phase Flow

Description: Abstract: A general expression which can be used either for predicting the average volumetric concentration or for analyzing and interpreting experimental data is derived. The analysis takes into account both the effect of non-uniform flow and concentration profiles as well as the effect of the local relative velocity between phases. The first effect is taken into account by a distribution parameter, whereas the latter is accounted for by the weighted average drift velocity.
Date: April 1964
Creator: Zuber, N. & Findlay, J. A.
Partner: UNT Libraries Government Documents Department
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Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Tenth Quarterly Report, January-March 1964

Description: Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, transient testing of fuel, fuel performance evaluation, fast-flux irradiation of fuel, and reactor physics and core analysis.
Date: April 1964
Creator: Breizy, C. E.
Partner: UNT Libraries Government Documents Department
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Program for the Development of Plutonium Recycle for Use in Light Water Moderated Reactors Quarterly Progress Report: January 1 - March 31, 1964

Description: A research program is being conducted to obtain experimental data in the irradiation of plutonium-enriched fuel to confirm a theoretical model for predicting isotopic composition and reactivity changes in plutonium-enriched, light-water-moderated reactors. Quarterly progress: Work has begun in the Radioactive Materials Laboratory to sample the project fuel from the pins irradiated to 1800 and 5000 MWT/T. Some delay has been experienced due to preemption of the hot cells by priority work. Examination of the autoradiographs of the un-irradiated project fuel showed that in a volume of fuel approximately equivalent to a pellet there were 13 hot spots larger than 15 mils. Evaluation of these spots with the fuel analyzer showed that they contained about 14 mg of PuO2 or about 9% of the total present. The EPITHERMOS code is being modified to automatically normalize the epithermal scattering to the correct value for all moderators. Calibration of the flux wires has been made and the reduction of the data from the VBWR irradiation is nearly complete. A similar resonance activation was made in the water reflector of the Stanford Pool Reactor to obtain the relative activity in a well-defined pure water spectrum. Reduction of these data is also in progress.
Date: April 15, 1964
Creator: Robkin, M. A.
Partner: UNT Libraries Government Documents Department
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In-Core Instrumentation Development Program Quarterly Progress Report January - March 1964

Description: The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. This technical report discusses the following topics: low versus high cable termination impedance, amplifier considerations, noise considerations, gas and pressure selection, cable selection, effect of gamma, effect of temperature, and remaining problems.
Date: April 1964
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department
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Development of Pulsed Neutron Application to Power Reactor Start-Up Procedures. Seventh and Eighth Quarterly Progress Report, October 1, 1963-March 31, 1964

Description: Activities in a program to develop techniques in the use of pulsed neutron sources to measure shutdown parameters related to large thermal power reactors are reported. In the course of this program, a new theory was suggested and an experimental apparatus was designed and built. Experiments were carried out to test the new model. This present report contains additional data and information extracted from the experiments at PG&E Humboldt Bay Power Reactor at Eureka, California. During the last days of 1963 a number of control rod and fuel bundle worth measurements were made in the ESADA Vallecitos Experimental Superheat Reactor (EVESR) using the (k[beta]/[script l] technique. A description of the experiments is given in the text of the report and some results are reported. A computer program was written to perform the data analysis of the pulsed neutron experiments and the code is discussed in the Appendix.
Date: April 24, 1964
Creator: Garelis, Edward & Meyer, P.
Partner: UNT Libraries Government Documents Department
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Development Program for Increased Output in the Garigliano Nuclear Reactor. Quarterly Report No. 6

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: April 1, 1964
Creator: Howard, C. L.
Partner: UNT Libraries Government Documents Department
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Transition Boiling Heat Transfer Program; Fifth Quarterly Progress Report, January - March 1964

Description: Summary: Tests with the two-rod assembly were performed with liquid film trippers attached to the unheated wall, and a variation in rod spacing. Experimental data and improved high-speed motion pictures have been obtained of transition boiling behavior. The changes of the local heat transfer process between nucleate and film boiling can be readily distinguished i the motion pictures. Observational test performed with very short fins on the heated surface resulted in essentially eliminating transition boiling temperature fluctuations and doubling the film boiling coefficient. These gains were attained without reduction of the critical heat flux
Date: April 1, 1964
Creator: Quinn, E. P.
Partner: UNT Libraries Government Documents Department
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100-KEW coolant backup adequacy

Description: No Description Available.
Date: April 29, 1964
Creator: Heacock, H. W.
Partner: UNT Libraries Government Documents Department
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Supplement B to production test IP-648-AC evaluation of thorium oxide as a fringe loading

Description: Irradiation of thorium oxide elements of the original test began in January 1964. Twenty columns of thorium oxide elements located in the outermost processing channels of F Reactor, accompanied by 40 columns of enriched, 0.947 w/o U-235, uranium elements in two adjacent lattice unit channels, were charged to permit evaluation of the reactivity effects of the loading and to obtain data on product and contamination build-up rates. Of the two loadings of thoria elements authorized by the initial test, the first has been discharged. Supplement A to Production Test IP-648-AC authorized two of the columns of thorium oxide elements to be irradiated through both cycles of the fringe loading. This supplement, Supplement B, to the test involves modification of the discharge schedule of the second loading of thoria elements and the two columns of elements authorized by Supplement a to permit discharge of the thoria columns at goal exposure of the supporting enriched uranium fuel columns.
Date: April 21, 1964
Creator: Hladek, K. L.
Partner: UNT Libraries Government Documents Department
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Budget for FY 1966 and revised budget for FY 1965 separations research and development program: (02 50 10 15 1)

Description: This report presents additional justification for operating costs of the Separations Research and Development Program which consists of eleven identifiable activities in four general areas of effort. The activities and their proposed levels of effort are listed. For each of the identifiable activities, the major accomplishments of the last year are listed, together with the proposed goals for the next two years, in the following pages. Significant trends in the program are noted.
Date: April 1, 1964
Creator: Frank, W. S.
Partner: UNT Libraries Government Documents Department
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N-Reactor co-product target element rupture detection study

Description: At present there seem to be some uncertainties concerning the needs of rupture detection for the co-producer programs Also in question is the method by which ruptures of the target elements should be detected and located. The purpose of this report is to discuss the waste and hazard control of tritium (H{sup 3}) (the product of the co-producer program), which will determine rupture detection needs, and to discuss methods by which rupture detection and location may be accomplished. The scope of the report considers first, adaptation of the present rupture monitoring system, and second, monitoring systems using H{sup 3} analyzers, together with the costs and time required to develop and use each method of rupture detection.
Date: April 27, 1964
Creator: Allred, D. O.
Partner: UNT Libraries Government Documents Department
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Thulium Oxide Irradiation

Description: Thulium oxide can be irradiated to meet the desired specific heat generation of 1.5 watts per gram as part of a flux amplification test, which will be conducted the early part of FY-1965. To obtain the heat generation level, is calculated that a 120-day irradiation will be required at a flux of 1.1 {times} 10{sup 14} neutrons/cm{sup 2}-sec. if the proposed target element is modified by reducing the thulium oxide wafer thickness from 0.080 to 0.050 inch. Also, the length of the target element should be reduced to not more than 20 inches to minimize the possibility of sticking in the process tube. This would require four targets rather than a single element as proposed. Prior to undertaking this irradiation, it would be desirable to place a prototype target element in the Hanford Test Reactor to determine the approximate rise factor (change in neutron flux level between the target element and adjacent tubes of fuel elements) in order to more precise data for planning the loading pattern in the production test.
Date: April 17, 1964
Creator: Kusler, L. E.
Partner: UNT Libraries Government Documents Department
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DT-IP-434-D: Irradiation effect on ceramic materials

Description: The purpose of the test described in this report was to determine the effect of production reactor flux on ceramic materials being considered for lining the vertical safety rod channels at the production reactors. This document is being written to describe the test, the results to date, and the work which remains to be done. The materials being evaluated in the test are: Aluminum silicate (Maryland Lava Co.); silicon nitride bonded silicon carbide (Norton Co. ``crystalon``); and aluminum oxide 85% (Norton Co. ``Alundum RA-98``).
Date: April 21, 1964
Creator: Cooley, D. E.
Partner: UNT Libraries Government Documents Department
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Hanford Laboratories monthly activities report, March 1964

Description: The monthly report for the Hanford Laboratories Operation, March 1964. Reactor fuels, chemistry, dosimetry, separation processes, reactor technology, financial activities, biology operation, and physics and instrumentation research, and applied mathematics operation, and programming operations are discussed.
Date: April 15, 1964
Partner: UNT Libraries Government Documents Department
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Production Test IP-584-D, supplement B, irradiation of target assemblies in the KER Loops

Description: The objective of this supplement to Production Test IP-584-D is to authorize low exposure (5--8 GVR) irradiation tests of the following lithium-containing target materials: Magnesium -- 14 percent lithium alloy, Lithium-aluminate (LiA1O2), Lithium-silicate (Li2SiO), Lithium-aluminate -- aluminum cermet, Lithium-silicate -- aluminum cermet. Lithium target materials will be contained in 4.50 inch long aluminum cans which are clad with 35 mil Zircaloy-2 alloy. The target elements will be contained in 1.9 inch OD, 1.5 inch ID Zircaloy-2 flow distributing sleeves. The target element assemblies and N Reactor inner-fuel elements (NIE-1) in Zircaloy-2 sleeves, will be irradiated in KER-1 or KER-2 at operating conditions approximating N Reactor operation.
Date: April 28, 1964
Creator: Deobald, T. L.
Partner: UNT Libraries Government Documents Department
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PT-IP-659-AC, Supplement A transition to normal discharge plan using striped target columns

Description: The scheduled termination of PT IP-659-AC, presently being irradiated in F Reactor, will result in discharge of 105 columns of enriched uranium (0.947 w/o U-235) at less than 50 per cent of the goal exposure. The test block is currently scheduled to be replaced with natural uranium columns. Since F Reactor is on a semiblock discharge plan (alternate rows), subsequent operating plans would require that 64 of these replacement columns of natural uranium be likewise discharged during the scheduled outage in May of 1964 at less than 50 per cent of goal. It appears desirable to minimize the economic costs of the production test by an alternative discharge scheme (e.g., interim poison irradiation). The objective of this supplement is to soften the economic impact of low exposure fuel discharge scheduled by IP-659-AC and simultaneously to obtain a useful alternate product by irradiating nine columns of Li-Al and Bismuth in a ``striped`` charge.
Date: April 8, 1964
Creator: Masche, G. C.
Partner: UNT Libraries Government Documents Department
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