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THERMONUCLEAR DIVISION PROGRESS REPORT, FEBRUARY 1, 1961-OCTOBER 31, 1961
None
Date:
January 24, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
REACTIVITY CALIBRATIONS AND FISSION-RATE DISTRIBUTIONS IN AN UNMODERATED, UNREFLECTED URANIUM-MOLYBDENUM ALLOY RESEARCH PROGRAM
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.
Object Type:
Report
System:
The UNT Digital Library
Effects of Seismic Vibrations on the Experimental Gas-Cooled Reactor
The effects of seismic vibrations on the dynamic behavior of a composite system were analyzed. The equations of motion were derived and soIved with special emphasis on determining the resulting stresses. The method of analysis thus developed was applied to the composite structure consisting of the core, pressure vessel, and supporting skirt of the Experimental Gas-Cooled Reactor (EGCR). A system with three degrees of freedom was considered in order to determine the effects of an earthquake of the maximum intensity expected in the area surrounding Oak Ridge, Tennessee. The system of equations of motion was solved both numerically and analytically, and the resonant frequencies were determined. The seismic effect was shown to be small when the frequency of the seismic disturbance coincided with a natural frequency of the system. In particular, the shear stresses in the graphite core were shown to be negligible. (auth)
Date:
June 22, 1962
Creator:
Witt, F.J.; Carver, D.R. & Maxwell, R.L.
Object Type:
Report
System:
The UNT Digital Library
Radiation Problems Associated With the Handling of the Actinide Elements
The hazards connected with the handling of actinide elements are surveyed. Emphasis is placed on Th, U, Np, and Pu. It is pointed out that the chemical toxicity of the actinides is usually minor when compared with radiochemi cal toxicity. Inhalation and ingestion are the important routes of entry but direct injection into the blood stream through wounds also requires consideration. Special enclosures, such as glove boxes, function primarily to minimize the risk of inhalation and aid in confinement. The external hazard from actinide elements, primarily due to gamma and fast neutron emission, varies considerably with the element and its source. Irradiated actinides, such as Th and Pu, usually show an increase in the external hazard from gamma radiation with extent of irradiation. (C.H.)
Date:
April 1, 1962
Creator:
Steindler, M.J.
Object Type:
Report
System:
The UNT Digital Library
Size effects and interstitial impurities in Nb/sub 3/Zr superconductors. Superconducting solenoids with metal insulation. [Oxygen and carbon]
None
Date:
September 11, 1962
Creator:
Betterton, Jr, J O; Kneip, Jr, G D; Easton, D S & Scarbrough, J O
Object Type:
Report
System:
The UNT Digital Library
Multigroup Calculations of Effective Delayed-Neutron Fractions, Prompt- Neutron Lifetime, and Related Kinetics Parameters. IBM-704 Program 1188/RE
In the course of a recent study of kinetics parameters, it was found that RE-185, because of its limitations, was inadequate to handle the problems involved with the reactors considered. Program ll88/RE was written entirely independent of and is intended to make RE-185 obsolete. The multigroup calculations of primary interest performed by the code are as follows: the total effective delayed-neutron fraction and its distribution into six ' families'' which are characterized by their periods; the prompt-neutron lifetime; and the worth of delayed and prompt neutrons in the core and blanket regions for all fissionable materials. Fluxes used in these calculations are taken directly from the real and adjoint solutions as obtained by RE-l22. (auth)
Date:
March 1962
Creator:
Kvitek, L. C.
Object Type:
Report
System:
The UNT Digital Library
MOLTEN-SALT REACTOR PROGRAM PROGRESS REPORT FOR PERIOD FROM MARCH 1 TO AUGUST 31, 1961
The Molten Salt Reactor Experiment design, component development, and engineering analysis is discussed. Materials studies for the Molten-Salt Reactor Program including metallurgy, in-pile tests, chemistry, engineering research, and fuel processing are described. (M.C.G.)
Date:
January 19, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
A FORTRAN Program for Calculating Single Crystal Absorption Corrections
A Fortran program is described for calculating the absorption correction for application to single crystal x ray or neutron intensity measurements. The program is available in the form of symbolic punched cards. InstrucPions are included. (J.R.D.)
Date:
April 1, 1962
Creator:
Wehe, D. J.; Busing, W. R. & Levy, H. A.
Object Type:
Report
System:
The UNT Digital Library
Periodic Characterization of Radioactive Waste Disposal Effluents. Core I Seed 3. Test Evaluation
Data are given on the radioactive nuclides present in waste disposal effluents during Nov. 1961. The concentrations of all activities was well within established limits for discharge to the environment. (C.H.)
Date:
June 13, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Calculation of Distance Factors for Power and Test Reactor Sites
>Reference information and guidance on procedures and basic assumptions related to reactor siting are presented which can be used to calculate distance requirements consistent with current siting practices. A single method of competing distances and exposure for a general class of reactors is giveo. (J.R.D)
Date:
January 1, 1962
Creator:
DiNunno, J. J.; Anderson, F. D.; Baker, R. E. & Waterfield, R. L.
Object Type:
Report
System:
The UNT Digital Library
Delayed Failure Hydrogen Embrittlement of Zirconium. Quarterly Report No. 2, December 15, 1961 to March 14, 1962
The purpose of this investigation is to determine the extent to which Zr and Zr alloys exhibit delayed failure (static fatigune) as caused by a combination of absorbed H and applied stress. Both notched and unnotched specimens of unalloyed Zr and Zircaloy-2 were hydrogenated to 200 ppm and 500 ppm by means of a modified Sieverts apparatus; specimens were evaluated at room temperature. Thus far, no time-dependent fracture was observed which can be attributed to the delayed failure phenomenon; it appears that these materials are relatively insensitive to static fatigune. The effects of grain size, temperature, cold deformation, and superheated water and steam corrosion on susceptibility to delayed failure are being determined. (auth)
Date:
March 30, 1962
Creator:
Weinstein, D. & Holtz, F. C.
Object Type:
Report
System:
The UNT Digital Library
GRAPHITE-STAINLESS STEEL COMPATIBILITY STUDIES
S>The compatibility of type 304L stainless steel in intimate contact with graphite is being studied as a function of temperature and contact pressure. This study is an outgrowth of materials compatibility problems in present and advanced gas-cooled reactors, where structural members in direct contact with graphite provide the possibility of both carburization and self-welding. Initial studies were concerned with surface reactions in the absence of gaseous contaminants under a vacuum of 10/sup -6/ mm Hg at 540 to 705 deg C. Stainless steel specimens are pretreated to provide three surface conditions: H/sub 2/- fired, preoxidized, and Cu-plated. Surface contact pressures ranged from 0 to 10,000 psi. Test results are presented which establish the lower temperature limit for significant diffusion between graphite and stainless steel at approximately 60O deg C. Above this temperature, diffusion between untreated or H2-fired stainless steel surfaces was found to effect complete bonding of the two materials at contact pressures as low as 500 psi. Bonding was effectively prevented by the presence of either an oxide film or a Cu plate at temperatures up to 700 deg C. Where bonding occurred, diffusion rates measured for C in stainless steel were comparable with those reported for stainless steel …
Date:
September 25, 1962
Creator:
Fleischer, B.; DeVan, J. H. & Coobs, J. H.
Object Type:
Report
System:
The UNT Digital Library
LOADING AND ELUTION CHARACTERISTICS OF SOME NATURAL AND SYNTHETIC ZEOLITES
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 from 0.2 to 0.00001, while the total normality of the system constituents was held constant. Results were plotted as equivalent fraction of the traced cation on the zeolite at equilibrium vs. the fraction of traced cation remaining in the equilibrium solution. If an influent composition is known, and falls in the system listed above, zeolite column loading may be read directly in univalent-univalent or divalent-divalent systems. However, in the strontium-sodium or strontium-- hydrogen systems, a k value must be used because squared terms appear in the equilibrium expression as follows: 2 Na- zeolite plus or minus Sr-solution in equilibrium 2 Nasolution + Sr-zeolite, or k = (Sr-zeolite)(Na-solution)/ (Na-zeolite)2(Sr-solution). The k for a given cation is a function of zeolite loading with that cation. Several column 50% loading capacities were determined to randomly test the validity of the above curves to predict column capacities, Several columns …
Date:
August 10, 1962
Creator:
Ames, L.L. Jr. & Knoll, K.C.
Object Type:
Report
System:
The UNT Digital Library
THE DISTORTED-WAVE THEORY OF DIRECT NUCLEAR REACTIONS. I. "ZERO-RANGE" FORMALISM WITHOUT SPIN-ORBIT COUPLING, AND THE CODE SALLY
The distorted-wave theory of direct nuclear reactions is presented in a unified manner, in which the effects of assuming various reaction mechanisms and nuclear models appear only in certain radial form factors. The zero-range approximation is used, and spin-orbit coupling is neglected in the distorted waves. Formulas are given for transition amplitudes, cross sections, and polarizations. A description is given of the IBM-704 computer code SALLY that is based on these formulas. (auth)
Date:
February 1, 1962
Creator:
Bassel, R.H.; Drisko, R.M. & Satchler, G.R.
Object Type:
Report
System:
The UNT Digital Library
SALT PHASE CHLORINATION OF REACTOR FUELS. IV. NIOBIUM BEHAVIOR IN THE LEAD CHLORIDE AND CHLORINE-LEAD CHLORIDE SYSTEMS
Investigation of the behavior of Nb in PbCl/sub 2/ showed that the Nb dissolution rate is quite low. Since stirring increases the rate appreciably, it is likely that the initial reaction is diffusion controlled. A subsequent reaction is linear in rate with an activation energy of 23.4 kcal per mole. Incomplete volatilization of dissolved Nb from PbCl/sub 2/ is due to the presence of lower oxidation states of Nb. The addition of Cl to PbCl/sub 2/ increases the rate of dissolution of Nb. The rate is high enough at reasonable temperatures to be practical for fuel dissolution, e.g. 12 mg min/sup -1/ cm/sup -2/ at 550 deg C and 120 mg min/sup -1/ Cl. Volatilization of NbCl/sub 5/ is 99.9% complete at 550 deg C. (auth)
Date:
March 15, 1962
Creator:
Teague, J. L.; Hahn, H. T. & Vander Wall, E. M.
Object Type:
Report
System:
The UNT Digital Library
A SURVEY AND EVALUATION OF U$sup 233$ FISSION YIELD DATA
A survey of the pertinent literature was made to ascertain the status of data on U/sup 233/ fission-product yields. The various experimental determinations were evaluated, and the most recent mass-spectrometric results were used as a basis for deriving a set of preferred yields. These yields were compared with values reported in two other recent compilations, and for yields >1%, the three setrs agreed with each other to an average precision of <5%. It was concluded that recent measurements have somewhar improved the reliability of U/sup 233/ fission yield data, but some recommendations for additional experimental work were made (auth)
Date:
July 13, 1962
Creator:
Ferguson, R.L. & O'Kelley, G.D.
Object Type:
Report
System:
The UNT Digital Library
HEAD-END TREATMENT OF LOW LEVEL WASTES PRIOR TO FOAM SEPARATION
Calcium-magnesium precipitation apparatus was used to reduce the concentrations of these elements in ORNL tap water, used as a substitute for waste water of low level of radioactivity, prior to strontium removal by foam separation. With and without alkali and flocculator chambers and with a stirred sludge of ratio height to diameter equal to 1/1 to ~4/1, use of 5 x 10/sup -3/ M each of NaOH and Na/sub 2/CO/sub 3/ and 2 ppm Fe/sup 3+/ reduced the dissolved Ca + Mg concentrations to 1 to 2 ppm as calcium. Simultaneously, a strontium DF of 20 to 200 was achieved, and, by adding Grundite clay in the proportion ~0.5 1b/ 1000 gal, a cesium DF of 10 to 40 was achieved. (auth)
Date:
May 29, 1962
Creator:
Schonfeld, E. & Davis, W. Jr.
Object Type:
Report
System:
The UNT Digital Library
HOMOGENIZATION OF MOLTEN-SALT REACTOR PROJECT FUEL SAMPLES
A copper pulverizer-mixer was designed for homogenizing Molten-Salt Reactor Project (MSRP) fuel. The copper sampling ladle that contains the solidified fuel is placed in the pulverizer-mixer, which is agitated on a mixer mill. The fuel is fractured out of the ladle, pulverized into a homogeneous powder, and transferred to a storage bottle. The homogenized fuel sample is then available for analysis. (auth)
Date:
July 1, 1962
Creator:
Gaitanis, M.J.; Lamb, C.E. & Corbin, L.T.
Object Type:
Report
System:
The UNT Digital Library
CHEMICAL EFFECTS OF HIGH EXPLOSIVE SHOCK WAVES ON VARIOUS COMPOUNDS WHICH OCCUR IN THE GNOME CONTAINMENT MEDIUM
None
Date:
May 1, 1962
Creator:
Bond, W.D.
Object Type:
Report
System:
The UNT Digital Library
KINETIC EXPERIMENTS ON WATER BOILERS, "A" CORE REPORT. PART I. PROGRAM HISTORY, FACILITY DESCRIPTION, AND EXPERIMENTAL RESULTS
None
Date:
March 30, 1962
Creator:
Flora, J.W.; Gardner, E.L.; Greenfield, M.A.; Roecker, J.H. & Stitt, R.K.
Object Type:
Report
System:
The UNT Digital Library
Quarterly Technical Report on Spert Project, January-March 1962
None
Date:
August 10, 1962
Creator:
Schroeder, F.
Object Type:
Report
System:
The UNT Digital Library
HYDROLOGIC AND GEOLOGIC STUDIES FOR PROJECT GNOME. Preliminary Report
Geologic lnformation required to define the pre- and post-shot physical and chemical characteristics of the sait and other rocks affected by the exploslon was gathered. Information on the pre- and post-shot hydrologic condltlons at the site and the surrounding area was also obtalned. (M.C.G.)
Date:
February 1, 1962
Creator:
unknown
Object Type:
Report
System:
The UNT Digital Library
Calculation and Measurement of the Transient Temperature in a Low- Enrichment UO$sub 2$ Fuel Rod During Large Power Excursions
The results of Spert I in-pile transient tests of a rodtype, low- enrichment UO/sub 2/ fuel element are presented. The tests were performed to investigate the possibility of damage to such long thermal-time-constant fuel rods when subjected to short-period power excursions, and to test the effectiveness of an instrumentation technique for measurement of UO/sub 2/ fuel temperatures within the rods. In an initial series of power excursion tests, in which the range of reactor periods was from approximately 1 sec to 7.5 msec, simultaneous measurements were made of the transient temperature at the center of the fuel rod and at the outer cladding surface. Fuel rod rupture occurred during the exponential rise of the 7.5-msec excursion. Similar short-period tests performed on a second fuel rod contain ing no internal thermocouples did not result in cladding failure, supporting the postulation that rupture of the first rod was caused by waterlogging of the UO/sub 2/ as a result of the cladding penetrations made for installation of the internal thermocouples. Calculations of the transient temperature distribution in the fuel rod were made, and the results are found to be in good agreement with the experimental data obtained on the central-UO/sub 2/ and cladding-surface …
Date:
May 18, 1962
Creator:
Houghtaling, J. E.; Quigley, T. M. & Spano, A. H.
Object Type:
Report
System:
The UNT Digital Library
Thermal-Stress and Strain-Fatigue Analyses of the MSRE Fuel and Coolant Pump Tanks
Thermal-stress and strain-fatigue analyses of the MSRE fuel and coolant pump tanks were completed for determining the quantity of coollng air required to obtain the maximum life of the pump tanks and to determine the acceptability of the pump tanks for the intended service of 100 heating cycles from room temperature to 1200 deg F and 500 reactor power-change cycles from zero to 10 Mw. A cooling-air flow rate of 200 cfm for the fuel pump tank was found to be an optimum value that provided an ample margin of safety. The coolant pump tank was found to be capable of the required service without air cooling. (auth)
Date:
October 1, 1962
Creator:
Gabbard, C.G.
Object Type:
Report
System:
The UNT Digital Library