This technical report describes development work done on method of particle separation by the Biology Division of the Oak Ridge National Laboratory and the Oak Ridge Gaseous Diffusion Plant during the period January 1 to June 30, 1963, under the Joint National Institute for Health-Atomic Energy Commission Zonal Centrifuge Development Program. The central effort has been to develop zonal centrifuge systems for the separation of cells and sub-cellular particles, including viruses, and bio-colloids, including proteins and nucleic acids.

A program has been coded for the ORACLE which will solve, using Monte Carlo technique, the transport problem for monodirectional, monoenergetic gamma radiation incident at an angle Θ, upon an infinite laminated slab of finite thickness. Each of the laminations (or regions) is itself an infinite, homogeneous slab of finite thickness. The code is designed to give estimates of energy deposition, energy flux, tissue dose rate, reflected and transmitted energy current, and the angular and energy distribution of the reflected and transmitted energy current. All the answers except for energy deposition and reflected and transmitted energy current are optional.

This manual represents the recommendations of the Instrumentation and Controls Division committee on Electrical and Electronic Symbols and Drawing that have been issued to date, and supersedes the previously issued ORNL Electrical Symbols List and CF-58-12-141, Electrical and Electronic Drawing Standards for Wiring and Device Coding and Applications.

In order that fairly accurate thermal stress calculations can be made on the ART, it is necessary to have a reasonable picture of the temperature distribution in the reactor. To get the temperature distributions, and to determine cooling requirements in various parts of the reactor, one must know the heat deposition rates due to alpha particles, beta rays, gamma rays, and neutrons in all parts of the reactor. The present report contains only the basic physical data necessary to determine the heat deposition rates due to gamma rays. Neutron fluxes in the core and reflector regions of the ART are to be obtained from two-dimensional multigroup calculations (performed by the Curtiss-Wright Corporation). These fluxes, in conjunction with the neutron absorption cross sections, determine the neutron capture and inelastic scattering rates in the core and in the reflector. The data in this report permit the calculation of the number of gamma rays originating at various energies at every point in the core and reflector.

The proposed ORNL Research Reactor is designed to serve as a general purpose research tool delivering a maximum thermal flux of 8x10^13 n/cm2-sec at the initial power level of five megawatts. Operation at power levels up to ten megawatts is proposed for such items as sufficient cooling capacity is available to handle the increased heat load. The reactor will use MTR-type fuel elements and beryllium reflector pieces in a 7 x 9 grid with moderation and cooling provided by forced circulation of demineralized water. The reactor tanks are submerged in a barytes concrete pool, filled with water, which serves as a biological shield. Experimental facilities include two 18" diameter "Engineering Test Facilities" and six 6" diameter beam holes. In addition, access to the core is available through the water of the pool. The result on the surrounding population of release to the atmosphere of a large fraction of the radioactive material in the core has been computed by two methods. It is shown that under certain conditions off-area personnel could be subjected to greater than the maximum permissible exposure. An analysis of the maximum hazard caused by the release of the entire contents of the core to the local watershed …

This memorandum sets forth a recommended uniform basis for designing the ORN shield.This includes design values for power level and emergent radiation, standards values for various material properties, and basic radiation intensities.

A summary of the laboratory development program on aqueous uranium slurry fuels for the Homogenous Reactor Project during the period April 1951 through March 1953 is presented. These investigations were devoted primarily to a study of the uranium oxides in aqueous suspensions. It was concluded that U(VI) was most likely to be the stable valence state in such slurry fuels and it was shown that β-UO3·H2O platelet crystals were the stable modification at 250°C. Very pure slurries of β-UO3·H2O platelets, uranium concentration of 250g/liter and average particle size of about 10 μ, had favorable settling rates and could be easily redispersed. Their viscosity and corrosion rate in stainless steel were comparable with those in water. Exposure of these slurries to pile radiation disclosed that radiolytic hydrogen and oxygen gas pressure comparable in magnitude to those of uncatalyzed uranyl sulfate solutions could be expected. Fission products in the irradiated slurries were predominantly associated with the solids. Radiation also tended to promote caking of these solids on the walls of the radiation bombs. Uranyl phosphate and the magnesium uranates were briefly investigated as alternate system but were not found satisfactory. The program was discontinued before the feasibility of uranium slurries for reactor …

A Lucite model of a multi-stage countercurrent hydroclone solvent extraction apparatus has been constructed and tested with Amsco-water. The diffusivity of Cs 134 tracer in aqueous chloride solution was measured to check the performance of the capillary diffusivity measuring system. The experimental data from four Druhm runs showed that 1/8in. thick graphite liners are usable for reactor temperatures above the boiling point of sodium.

This quarterly progress details ongoing research at the Oak Ridge National Laboratory as a part of the Aircraft Nuclear Propulsion Project. Topics discussed include reactor theory and design, [part two is not included], materials research, and appendixes with information on analytical chemical studies.

Report issued by the Oak Ridge National Laboratory discussing quarterly progress made by the Homogeneous Reactor Program. Descriptions of progress and studies conducted are presented. This report includes tables, illustrations, and photographs.

A preliminary chemical flowsheet is presented of a fluoride volatility process for recovering and decontaminating uranium from heterogeneous reactor fuels after dissolution in a fused salt. In laboratory work, a gross β decontamination factor of > 10 4 was obtained in the fluorination of a UF4-NaF-ZrF4 melt by passing the product UF6 through NaF at 650°C. The solubility of UF6 in molten NaF-ZrF4 was shown in kinetic studies to cause a lag in the evolution of UF6 from the fluorinator. Corrosion of nickel in the fluorination step appeared to be 2-4 mils/hr during the time that uranium was present. The average corrosion rate over the process as a whole was less than O.4 mil/hr. Earlier studies were reported in ORNL-1709 and 1877.

A conductometric titration method described by Goldstein was adapted for use in a remote analytical facility. The results obtained by mean of experiments made prior to this adaptation indicated that methanol is the most satisfactory medium in which to determine excess sulfuric acid in uranyl sulfate solutions that stimulate Homogeneous Reactor type fuel. When methanol is used, the complexation of hydrolyzable ions with sodium fluoride, as described by Pepkowitz, Sabol, and Dustin, is not required.

The attached memo by G. N. Krouse of Sturm-Krouse, Inc. gives results of a preliminary analysis of the deflections of beam holes due to thermal expansion and internal pressure in the vessel. A partial solution of the problem is suggested. Based on preliminary pressure-temperature data the following deflections were derived: Movement of horizontal beam tubes = 0.046 in. Movement of Engineering facility tubes = 0.117 in. Vertical motion of the vessel at the horizontal beam tubes due to thermal expansion may be eliminated by locating the supports in that plane. That also will reduce the expansion at the point where the slant tubes pierce the vessel wall.

An extension of the concepts advanced by Langmuir regarding the nature of the platinum catalyzed oxidation of hydrogen and the application of the resulting theory to the experimental data observed by Ranschoff and Spiewak for an HRE type recombiner indicates that their data are corrected by the dimensionless equation (see report) equally well, with a mean deviation of 3.8 percent. This expression is recommended as a basis for the design of catalytic recombiners. The catalytic combinations is pictured as consisting of two surface chemical mechanisms, one of which is oxygen diffusion controlled, the other hydrogen diffusion regulated, the mechanism "change-over" occurring at that point in the recombiner where the components are arriving at the catalyst surface by diffusion in stoichiometric proportions. The catalyst volume requirements for three two portions of the bed are shown to be (see report). The hydrogen mole fraction at the mechanism "change-over" point is (see report). And the relationship between the two mass transfer coefficients is (see report). Methods for evaluating the necessary transport properties of the ternary system steam-hydrogen-oxygen for carrying out design calculations are summarized, and the new significant parameters are tabulated and plotted to facilitate these calculations. The question of non-uniform velocity profiles …

Report issued by the Oak Ridge National Laboratory discussing thermodynamic calculations of nuclear fuels. As stated in the introduction, "the equations developed in this report could be used to calculate the capacity of chlorine for transporting plutonium tetrachloride" (p. 2). This report includes tables, and illustrations.

Experimental determination of heat-transfer coefficients, burnout heat fluxes, and friction factors have been made for swirl flow of low-and moderate-pressure water through electrically heated aluminum, nickel, and copper tubes containing full-length Inconel twisted tapes. For nonboiling conditions, swirl-flow heat-transfer coefficient were successfully correlated with both the Froude modulus (the ratio of inertial to centrifugal forces) and a grouping of the Grashof and Reynolds moduli (ratio of buoyant to inertial forces).

A portable shield has been designed, developed, fabricated and shop tested to provide the HRT with a facility for direct dry maintenance operations. It provides temporary replacement for any one of the lower roof plugs and should permit many operations to be performed without flooding the reactor cell with water.

An ion-exchange -- polarographic method was developed for the determination of iron(III) in Homogeneous Reactor Fuels. Copper, which interferes, is removed from the fuel by plating it onto a cadmium coil. Iron is oxidized to iron(III) by potassium permanganate, and the iron(III) is separated from interfering metal ions by ion exchange on a Dowex 1 resin column that is in the sulfate form. The iron(III) in the effluent is determined polarographically in 0.5 M sodium citrate solution as supporting electrolyte. A fairly well defined polarographic wave is obtained for the iron(III) → iron(II) reduction at a half-wave potential of approximately -0.15 v. vs. the S.C.E. The relative standard deviation of the data for 2 µg of iron(III) per ml of solution in the polarographic cell was 6.5%; for 10 µg of iron(III) per ml it was 0.6%.

EXPIRE is a calculation which predicts the reactivity-lifetime, instantaneous and integrated effective multiplication constants and instantaneous and integrated effective multiplication constants and instantaneous conversion ratio for heterogeneous reactors. The concentration of all the isotopes of interest from Th232 to Am243 are calculated as a function of time using the average reactor power density and a uniform flux distribution. The equations have been programmed for the IBM-704 computer and the average running time is approximately two minutes per reactor lifetime.

The effect of adding small quantities of fuel or poison to the HRT has been estimated using perturbation theory. The results have been reduced to a single relation and a set of graphs which make the estimation of added reactivity relatively simple. The perturbation theory results are compared with multigroup results and reasonable agreement is demonstrated; however, there is some question concerning the prompt neutron lifetime.

Technical report covering experiments happening on the Analytical Chemistry Division's sites at the Oak Ridge National Laboratory. Includes information on ionic analyses, radio-chemical analyses, spectrochemical analyses, service analyses, inorganic preparations, analytical chemical control of homogeneous reactor solution, optical and electron microscopy, and service analyses for the period ending March 26, 1951. [From Abstract]

This quarterly progress report discusses in detail the work done by the Analytical Chemistry Division at Oak Ridge National Laboratory. In particular, this report discusses the research and development work ongoing at the X-10 Site and the Y-12 Site.

The continuous dissolution of 304 stainless steel and stainless steel-UO2 alloy in dilute aqua regia was studied with subsequent stripping of the dissolver product to remove chloride ion. The process has the advantage of producing, by means of a simple head end treatment, a solvent extract feed in a conventional nitric acid medium so that existing solvent extraction processes, materials of construction and waste disposal methods can be used. The purposes of this study were to investigate the variables affecting the dissolution process and to obtain dissolver scale-up data, and to investigate the removal of chloride from the dissolver product and the variables affecting the stripping operation. A continuous flooded pot dissolver was used. It has the advantages of stability of operation and ease of control in comparison with column dissolvers and requires a minimum of mechanical processing prior to dissolution. Stripping of the dissolver product to remove chloride ion was studied in a 4-in. diameter Pyrex bubble-cap column containing 12 single bubble cap plates. Continuous dissolution rates and dissolver product stainless steel loading were correlated with aqua regia feed composition, acid feed rate and surface area exposed to reaction. Profiles of chloride concentration down the stripping column were obtained …

Report containing the description and design of U₃O₈-Aluminum Dispersion Fuel Elements for Core I of the Puerto Rico Research Reactor. Topics include the requirements of core materials, materials specifications, manufacturing procedures, and shipping preparation.