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Unit Operations Section Monthly Progress Report February 1960 - Chemical Technology Division

Description: Summaries of progress are presented for Fuel cycle development, GCR coolant clean up studies, HR Thorium blanket studies, Ion exchange, Power reactor fuel processing, Solvent extraction studies, Volatility, and Waste processing
Date: May 18, 1960
Creator: Whatley, M. E. (Marvin E.), 1926-; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Study of the Fuel Value of Plutonium

Description: The fuel value of mixtures of plutonium isotopes has been calculated for five thermal reactors (Yankee, Dresden, Hallam, GCR-II, Carolinas-Virginia)
Date: February 11, 1960
Creator: Jaye, S.; Bennett, L. L. & Lietzke, M. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Assay Methods Used In The ORNL Radioisotope Program

Description: A brief description is given of methods used in the measurement of radioactive content of radioisotopes distributed by the Oak Ridge National Laboratory. A tabulation of the techniques employed for seventy eight nuclear species is presented.
Date: February 1, 1960
Creator: Reynolds, S. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Eurochemic Assistance Program: Progress Report for October Through December, 1959

Description: The status of the Eurochemic (European Company for the Chemical Processing of Irradiated Fuels) Organization and the preliminary plant design are summarized. This report covers the progress during the period October through December 1959, gives the status of the program as of January 1, 1960, and outlines the anticipated assistance from the United States for the next 6 months.
Date: March 31, 1960
Creator: Nicholson, E. L. & Shank, E. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Observed Heat Transfer In HRT Fuel And Blanket Heat Exchangers

Description: Data obtained during power operation of the HRT indicates that during the circulation of fuel solution at high temperatures for more than 260 days there was a small increase in heat transfer resistance in the heat exchangers. The change in heat transfer correlates with the increase of the amount of corrosion product solids in the reactor fuel system.
Date: February 8, 1960
Creator: Haubenreich, P. N.; Piper, H. B.; Buxton, S. R. & Engel, J. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Numerical Three Dimensional Temperature Analysis Of The EGCR Fuel Rod

Description: The temperature field in an outer rod of the EGCR fuel assembly is presented. The numerical analysis has taken account of the three dimensional spatial dependence of the thermal neutron flux and has considered the four regions involved, that is, fuel pellets, ceramic disc, steel end caps, and cladding. The convective heat transfer coefficient and fluid temperature were considered constant over the 2.185 in. of rod length studied.
Date: March 21, 1960
Creator: Epel, L. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Natural Circulation Burn Out Heat Flux For The ORR

Description: Natural circulation flow rate is calculated for the ORR core. From this flow rate, the burn out heat flux is found to be between 22,000 and 35,000 Btu/ft2-hr. It is concluded that under present conditions a positive acting after heat removal system is necessary for 30 mw operation.
Date: February 15, 1960
Creator: Wett, J. F.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Concentration Of Nitrogen 15 By Chemical Exchange Between NO2 And NO

Description: This program was planned to investigate the concentration of Nitrogen 15 by chemical exchange between liquid nitrogen tetroxide and gaseous nitrous oxide. The method used by Grey and coworkers was to be attempted.
Date: 1953
Creator: Spindel, William, 1922-; MacLaren, K. H.; Johns, R. N.; Lenni, R. P.; Reynolds, G. W. & Lanford, O. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Cladding Survey For The Enrico Fermi Reactor U-15 Wt % Mo Base Dispersion-Type Fuel Element

Description: Potential cladding materials for a flat-plate fuel element containing a dispersion of UC or UO2 in U-15 wt % Mo alloy were surveyed on the bases of compatibility with the fissile compounds, matrix material, protective cover materials, and liquid sodium as well as the feasibility of fabricating fuel plates by roll cladding.
Date: April 29, 1960
Creator: Martin, M.M. & Beaver, R. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Galvanic Corrosion Problem Associated With The Preparation Of Multimetallic Beryllium Samples

Description: Artifacts have been noticed in the microstructure of multimetallic beryllium samples that were polished on the Syntron vibratory polisher. These artifacts were at first erroneously attributed to diffusion of beryllium into the dissimilar metal. The real cause of the effect is thought to be due to a galvanic corrosion which takes place during polishing. Small additions of sodium nitrate to the polishing slurry have been found to eliminate the corrosion.
Date: April 20, 1960
Creator: Hewette, D. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Gamma And Beta Heat Generation Rates In The HFIR Core

Description: A calculation has been made to determine the fuel plate heat fluxes resulting from after shutdown fission product heating. Fission product source strengths were obtained via the IBM Internuc code. Slab geometry was assumed. The core coolant gamma heating rate during reactor operation has also been calculated using the same techniques, but including the fission and capture gamma sources.
Date: April 29, 1960
Creator: Hilvety, Neil
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Corrosion In The Oak Ridge Research Reactor Core-Cooling System

Description: Corrosion specimens of the five major aluminum alloys used in the construction of the Oak Ridge Research Reactor have been exposed to the high-purity primary cooling water in the ORR core and in the external portion of the primary cooling loop to determine their corrosion rates under actual operating conditions. These alloys, 1100, 3003, 5052, 5154, and 6061, exhibited average corrosion rates of less than 2.6 mpy during the first 500-hr test period and less than 0.5 mpy for a 4032-hr test. Only… more
Date: April 25, 1960
Creator: Neumann, P. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Some Experiments On The Accuracy Of Thoria Slurry Samples

Description: Tests were performed on a thoria slurry flowing in a 3 inch diameter pipe to determine the magnitude of the possible error involved in the sampling process. Evidence indicates that a correct sample is obtained by withdrawing the sample isokinetically (i.e. by facing the sampler into the flow and adjusting the sampler velocity to match the ambient velocity) provided that the sampler is larger than some minimum diameter that is dependent on the mean eddy length and/or the mean particle size.
Date: April 21, 1960
Creator: Wichner, R. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Standard Operating Procedure For TSR-II

Description: The daily check list, the startup procedure both for in-pool and in-air operation, and the normal shutdown procedure for the Tower Shielding Reactor-II are outlined. The sequence of events which leads to each annunciator warning and the automatic corrective action which is then initiated by the control circuits are described.
Date: April 20, 1960
Creator: Lewin, J. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Study Of The Fuel Value of U233

Description: The fuel value of U233 was calculated for five thermal reactors (Dresden, Yankee, Carolinas-Virginia, Hallam, OCR-II). Relative to a U235 value of $17 per gram, pure U233 had a value that varied from $18.2 to $20.2 per gram. U233 contained in once- and twice-recycle uranium from an initial U233-Th cycle value of U233 in recycle uranium from an initial U235-Th cycle was less than that for pure U233 and decreased with each succeeding cycle.
Date: April 11, 1960
Creator: Jaye, S.; Bennett, L. L. & Lietske, M. P.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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An IBM 704 Subroutine For Trapezoidal Integration With Controlled Error (TRICE)

Description: An IBM-704 subroutine, TRICE, has been written in FORTRAN which evaluates a single integral by means of the trapezoidal rule approximation. In the calculational procedure, the integration step is halved until the desired degree of convergence is reached or until a preset maximum number of steps is exceeded. A maximum of ten different integrals may be used in any one calling program.
Date: April 19, 1960
Creator: Nestor, C. S.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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GAP: The IBM 704 Grid Analysis Program

Description: GAP is an IBM 704 program for analyzing simply-supported rectangular beam grillages in which the beams intersect at right angles. The grillage need not be symmetrical and may have both uniformly distributed and concentrated loads. Also, the individual beams may have arbitrary cross sections. Strain energy methods are used to determine the bending, shearing, and torsional effects in a given grid structure. Deflections and moments can be found directly by the program, and series coefficients are … more
Date: May 2, 1960
Creator: Witt, F. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Desirability Of Chemical Decladding From The Waste Disposal Viewpoint

Description: Chemical decladding of stainless steel and Zircaloy jacketed fuel is discussed with respect to cladding activation, capacity for storage of decladding waste in existing concrete tanks at ORNL, and waste disposal cost as compared with a total dissolution process.
Date: April 1, 1960
Creator: Irvine, A. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Unit Operations Section Monthly Progress Report - April 1960, Chemical Technology Division

Description: Progress is reported on chemical engineering research, solvent extraction, fuel cycle development, thorium studies, ion exchange, reactor fuel processing, solvent extraction, and waste processing.
Date: July 28, 1960
Creator: Whatley, M. E.; Haas, P. A.; Horton, R. W.; Ryon, A. D.; Suddath, J. C. & Watson, C. D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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VPP -- Design Criteria For An Installation To Remove Hydrogen Fluoride And Fluorine From The Cells 1 And 2 Ventilation Gases Prior To Filtration

Description: Criteria are presented for a horizontal concurrent spray nozzle scrubbing system designed to remove fluorine and hydrogen fluoride from the 3000 cfm of ventilation air passing through the Volatility Pilot Plant located in cells 1 and 2, Bldg. 3019. A reduction of fluorine concentration from 1520 to <2 ppm during a total release of 68 lbs, and a reduction of hydrogen fluoride concentration from 4090 to <1 ppm during a total release of 200 lbs, will adequately protect the Fiberglass media filters… more
Date: April 11, 1960
Creator: Ruch, J. B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Studies Of Improvement Of Power Density In ORR Loops

Description: Using a simplified model, calculations have been made concerning the possible effects of voids upon the power density in ORR loop experiments. It is concluded that the power density may be markedly increased if voids and channels are plugged with moderator material such as graphite or beryllium.
Date: April 11, 1960
Creator: Tobias, M. L. & Vondy, D. R.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Simple Formula For Computing Fission Product Thermal Cross Sections And Resonance Integrals

Description: This memorandum gives a simple formula (and the necessary constants) for computing fission product thermal cross sections and resonance integrals as a function of exposure. The simple formula was used in calculating the effective cross section and resonance integral as functions of fuel-processing cycle time in fluid-fueled reactors.
Date: April 7, 1960
Creator: Nestor, C. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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