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Supporting Analysis for Thermal Suitability of Fuel Elements for SM-1A Core I Loading

Description: A recommended SM-1A Core I loading chart was derived from available, metallurgically acceptable elements at the SM-1A and SM-1 sites. The derivation was based on local thermal and hydraulic considerations of minimum elementto- element coolant channel clearances. These clearances were determined from field inspection measurements of outer fuel plate spacing, as modified by analytical calculations of plate ripple growth during exposure to reactor operating thermal stresses. (auth)
Date: January 10, 1962
Creator: Brondel, J. O.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Delayed Failure Hydrogen Embrittlement of Zirconium. Summary Report, September 15, 1961 to September 14, 1962

Description: The extent to which zirconium and zirconium alloys exhibit delayed failure (static fatigue) as caused by a combination of absorbed hydrogen and applied stress was investigated. Susceptibility to time-dependent fracture was evaluated for unalloyed zirconium and Zircaloy-2 with 200 and 500 ppm hydrogen as well as for an experimental Zr Al-Sn-Mo alloy and the Canadian Zr-2.5Nb cladding material. For unalloyed zirconium and Zircaloy-2 containing up to 500 ppm hydrogen, no room-temperature, timedepe… more
Date: October 10, 1962
Creator: Weinstein, D. & Holtz, F. C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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REACTIVITY CALIBRATIONS AND FISSION-RATE DISTRIBUTIONS IN AN UNMODERATED, UNREFLECTED URANIUM-MOLYBDENUM ALLOY RESEARCH PROGRAM

Description: Completion of zero-power critical experiments with the ORNL Health Physics Research Reactor is reported. A description is given concerning these experiments which were used to determine the critical size, fission-rate distributions, reactivity calibrations of its movable parts, the temperature coefficient of reactivity, and the reactivity effects of the presence of neutron- reflecting materials adjacent to the reactor. (J.R.D.)
Date: May 10, 1962
Creator: Mihalczo, J.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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LOADING AND ELUTION CHARACTERISTICS OF SOME NATURAL AND SYNTHETIC ZEOLITES

Description: The cation exchange characteristics of several commercially available zeolites were evaluated, Equilibrium loading characteristics and exchange capacities were determined experimentally in the binary systems cesium-lithium, cesium-- sodium, cesium--potassium, cesium-Tubidium, cesium -- hydrogen, cesium-- ammonium, strontium-- magnesium, strontium-- calcium, strontium-- barium, and strontium-- hydrogen for several zeolites. Exchangeable cation fractions in the equilibrium solutions were varied f… more
Date: August 10, 1962
Creator: Ames, L.L. Jr. & Knoll, K.C.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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TWO AUXILIARY CODES FOR USE WITH RENUPAK

Description: Two IBM-7090 codes were written to aid the user of the neutron moments method code RENUPAK. One code computes and punches response function input cards for RENUPAK or NIOBE (another neutron transport code). The second code reads RENUPAK flux tapes and prints out a compact edit including dose rate as a function of distance. (auth)
Date: August 10, 1962
Creator: Trubey, D.K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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NEUTRON THERMALIZATION MEASUREMENTS USING AN ELECTRON LINEAR ACCELERATOR

Description: Discussions are presented concerning integral experiments primarily on water and graphite systems. A description of experimental facilities is included along with details of die-away experiments, infinite-medium spectrum measurements, spatially dependent spectrum measurements, spectrum measurements for graphite systems, and development of a subcritical assembly program. (J.R.D.)
Date: April 10, 1962
Creator: Beyster, J R
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Ornl Analytical Chemistry Division's 150-Kv Cockcroft-Walton Generator

Description: A general description of the facilities used to house a small Cockcroft- Walton generator is presented. Preliminary irformation and data are given as to the operational performance of the generator, the radiation safety controls involved, and the expected use of the device, In addition, an automatic-manual device for control of tritium target usage is described. (auth)
Date: September 10, 1962
Creator: Strain, J. E.; Hampton, W. J. & Leddicotte, G. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Numerical Analysis of Neutron Resonances

Description: Neutron resonances are analyzed numerically by determining the resonant energies, the total widths, and the neutron widths of the resonances. It was assumed that the total cross section can be represented by the sum of single- level Breit-Wigner formulas, with interference between resonance and potential scattering but no interference between resonances; that the Doppler broadening is represented by a Gaussian function; and that the instrument resolution is represented by a Gaussian function. G… more
Date: January 10, 1962
Creator: Atta, S.E. & Harvey, J.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

IN-PILE GAS-COOLED FUEL ELEMENT TEST FACILITY

Description: Paper presented at American Nuclear Society Meeting, June I8-21, 1962, Boston, Mass. Design and operating problems of unclad and ceramic gas-cooled reactor fuels in high temperature circulating gas systems will be studied using a test facility now nearing completion at the Oak Ridge Research Reactor. A shielded air-tight cell houses a closed circuit gas system equipped for dealing with fission products circulating in the gas. Experiments can be conducted on fuel element performance and stabilit… more
Date: July 10, 1962
Creator: Zasler, J.; Huntley, W. R.; Gnadt, P. A. & Kress, T. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

OGRE-P2, a Monte Carlo Program for Computing Gamma-Ray Leakage From Laminated Slabs With a Distributed Source

Description: A Monte Carlo program OGRE-P2 was written for the IBM-7090 to solve the problem of gamma radiation from a slab of laminated regions composed of various materials. The dose rate is calculated on one side of the slab. (J.R,D.)
Date: August 10, 1962
Creator: Trubey, D. K.; Penny, S. K. & Emmett, M. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

THE FABRICATION AND IRRADIATION TESTING OF DISPERSION-TYPE BeO-UO$sub 2$ CERAMIC FUEL MATERIALS

Description: Considerations affecting the design of dispersion-type ceramic fuel materials are discussed. The results of measurements made to study the effect of fuel loading on the properties of the fuel bodies and the results of irradiation tests of fuel pellets composed of 20 vol% UO/sub 2/ dispersed in BeO are presented. Experience in the fabrication of dispersion-type BeO --UO/sub 2/ ceramic fuel materials is described. From the results obtained, it is believed that BeO --UO/sub 2/ fuel materials of th… more
Date: October 10, 1962
Creator: Johnson, D. E. & Lofftus, F. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

Super-Prompt-Critical Behavior of an Unmoderated Unreflected Uranium- Molybdenum Alloy Assembly

Description: The time-dependent behavior was investigated of the neutron population in an unreflected unmoderated cylindrical assembly of 90 wt.% U (93.2 wt.% U/sup 235/), 10 wt.% Mo alloy following rapid establishment of a super-prompt critical c ondition with negligible initial neutron population. Reactivity increases up to 11 cents above prompt critical resulted in bursts yielding as many as 1.8 x 10/ sup 17/ fissions with reactor periods as short as 16 mu sec and temperature increases as large as 400 de… more
Date: May 10, 1962
Creator: Mihalczo, J.T.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

A Bench-Scale Natural-Recirculation Dissolver

Description: A natural-recirculation dissolver closely approaching a stirred-tank reactor in behavior was developed, and its feasibility was demonstrated in the mercury-catalyzed dissolution of aluminum in nitric acid, It was designed to utilize the heat of reaction and evolution of gaseous reaction products for recirculation and mixing. The dissolvent was 5-molar nitric acid, containing mercury in concentrations from 10/sup -4/ to 10/sup -3/ molar as a catalyst. Dissolution rates for 1/7 scale, simulated M… more
Date: January 10, 1962
Creator: Erickson, E. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
open access

DISCUSSION OF AN ACTIVATION METHOD FOR MEASURING THE RATIO OF U$sup 238$ ABSORPTION TO U$sup 235$ FISSION IN A REACTOR

Description: The theory is presented in detail of an activation method for measuring the ratio of U/sup 238/ absorptions to U/sup 235/ fissions in a nuclear reactor. The origin of the activities is discussed, the equations are given for interpretation of the measurements, and the expected counting rates are computed for a typical experiment in the EGCR. (auth)
Date: August 10, 1962
Creator: Preskitt, C.A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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