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.

This technical report describes development work done on zonal centrifuge systems at the Oak Ridge National Laboratory and the Oak Ridge Gaseous Diffusion Plant during the period July 1 to December 31, 1962 under the Joint National Institute for Health-Atomic Energy Commission Zonal Centrifuge Development Program. A basic purpose of this project is to develop new methods for isolating virus particles associated with cancerous cells and tissues. Three classes of rotor systems capable of separating particles ranging in size from whole animal or plant cells to protein or nucleic acid molecules on the basis of either sedimentation rate or density alone have been developed. Experiments with phage particles indicate the feasibility of large-scale virus isolation by continuous-flow centrifugation, followed by isopycnic banding in cesium chloride and velocity sedimentation in sucrose - all steps being carried out sequentially in the same rotor. Zonal rotors using the reorienting gradient principle for molecular separations have been tested to 141,000 rpm (formula). Previous work on zonal centrifugation and future plans for this program are discussed.

As part of the thermonuclear effort at Oak Ridge National Laboratory, large volume high-beta plasmas have recently been produced by microwave power at the electron-cyclotron frequency. The original work was done with 13-cm microwave radiation in a magnetic mirror and was reported at Salzburg. Since that time, higher frequency and higher power radiation has been used in a succession of experiments. The current experiments involve heating a plasma in the EPA Facility which has a 3:1 magnetic mirror ratio. Up to 50-kw CW of 3-cm microwave power are applied to the plasma in a large volume cavity.

Calculated data and graphs describing the effects of batch thermal-neutron irradiations on the buildup of fission products in natural uranium, enriched uranium, and thorium are presented together with empirical equations and plots correlating total fission product activities and/or decontamination factors. Fluxes of 1012-1015 are considered.

An evaluation of the relative health hazards of radioisotopes produced in nuclear reactors is reported. The most important hazards were indicated to be I131, the Sr90 - Y90 chain, the Ce144 -Pr144 chain, Sr 89, the Ba140-La40 chain, Y91, the Zr95-Nb95 chain, Pr143, La140 , and Pa233. The most critical body organs affected by air-borne contamination are the thyroid gland, the bone marrow, the lungs, and the gastrointestinal tract. Where possible, continuous daily removal of gaseous and solid fission products from the reactor environment can be shown to permit very significant reductions in the total hazards. Homogeneous reactors, such as the Thermal Breeder Reactor and the Homogeneous Plutonium Producer Reactor, specifically studied in this report, are designed with daily removal cycles and may be considered potentially safer than heterogeneous reactors.

Activity and ionic impurity specifications are presented for Thorex thorium nitrate products. Two sets of specifications are given, one set for direct handling during refabrication of production reactor thorium metal slugs and the second for refabrication of future power reactor thorium metal elements by semi-remote technics. Consideration was given to the health hazard problems associated with each process step between the Thorex process and final refabricated source material in order to arrive at these specifications.

A fifth in-pile loop experiment, L-4-8, was completed. The loop operated in-pile for a total of 1637 hr, during which time the LITR energy output was 4377 Mwhr. The average fission power in the loop based o cesium analyses was 622 w when the LITR was at full power (3 Mw). Based on oxygen data, the generalized corrosion rate for the first 300 hr was 4.0 mpy; the rate for the remaining 1357 hr was 0.7 mpy. The nickel data gave parallel results. The corrosion of the type 347 stainless steel, Zircaloy-2, and Ti-55AX [unintelligible] exposed in the core and in in-line holders was generally consistent with that observed in previous in-pile loop experiments. Some differences with steel were attributed to the fact that this was the first loop containing steel specimens operated with 0.04 m H2SO4 present in the uranyl sulfate charge solution (0.17 m UO2SO4, 0.03 m CuSO4). Stress specimens, made from the alloys Zircaloy-2, type 17-4 PH stainless steel, and Ti-C-130-AM, were exposed in care, in-line, and pressurizer locations. Microscopic examination and average weight loss gave no indication of effects attributable to the stressed condition of the specimens.

The subject which I will discuss is the distorted - waves theory of direct reactions and its application to the study of nuclei. I shall mainly be concerned with a study of the reaction theory, its range of validity, and its ability to give quantitative information about nuclear structure.

Technical report containing the general arrangement of the Low Intensity Training Reactor and outcomes of experiments that have been performed in the reactor to measure the rate of fission product heat dissipation from the fuel pieces after loss of water. [From Summary, Introduction]

The Homogeneous Reactor Test (HRT) is the experimental reactor facility (Frontispiece) being designed and constructed at ORNL as the next step in homogeneous reactor development between the 1-Mv HRE and a "full-scale" power station. The HRT will provide an integrated test at 5 to 10 Mv for the flowsheet and equipment designs on which the full-scale effort will be based. Furthermore, its design is such that several homogeneous systems which require essentially the same operating equipment may be tested with comparatively minor modifications of the original reactor installation. The reactor will be assembled in the building which housed the HRE, located in the experimental reactor exclusion area approximately one mile south of the oak ridge laboratory. (See figure 1) / It is the purpose of this report to provide information with which the hazardous aspects of this reactor may be evaluated. Briefly, it will be shown after a statement of purpose and a general description of the reactor that: 1. The design characteristics and equipment requirements are such that escape of highly reactive material from the reactor piping is unlikely. 2. Should the entire core and blanket contents suddenly escape from the reactor system, a seal-welded steel tank surrounding the …

Technical report covering classified work of the Critical Experiments Program in the Physics Division for the period December 20, 1950 to March 20, 1951. Report outlines investigations of properties of critical assemblies composed of uranium and various reflectors and moderators, preliminary measurements of the effective energy for fission in the assembly. [From Introduction and Summary]

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.

LITR Fluoride-Fuel Loop. — The inconel loop was dismantled for removal of the samples and for recovery of the uranium by using the remote cutting tools installed in a half cell of the Solid State Building. Disassembly proceeded without incident. An electric-arc cutting technique was developed for removal of the stainless steel enclosure around the pump bowl. Fission power and maximum flux were determined by irradiating a simulated loop, by heat-balance calculations, by radiochemical analyses for fission products in the fuel, by measuring the activation of cobalt foils attached to the loop, and by activation of the loop tubing itself. The determination of the power by these various methods gave 2.5 to 2.8 kw during operation of the loop, and the maximum power density was 0.4 kw/cc. Chemical analyses of the fuel were carried out to determine U, Zr, and the major constituents of inconel: Ni, Cr, and Fe.

This semiannual progress report and future reports will be published as two documents to permit a wider distribution of the unclassified material. The report numbers are assigned in sequence so that the two reports will fall together when filed by report number.

Technical report outlining the building status, crystal physics, solid state reactions, radiation metallurgy, engineering properties, and liquid fuels activities of the Physics of Solids Institute in period ending April 30, 1951. [From Introduction and Summary]

This quarterly progress report discusses the ongoing work within the Solid State Division at the Oak Ridge National Laboratory for the period ending August 10, 1952. Projects discussed include radiation metallurgy, engineering properties, fused salts, crystal physics, and solid state reactions.

From introduction: "This report covers work performed in the Solid State Division for the quarter ending May 10, 1952." Topics covered include radiation metallurgy, engineering properties, solid state reactions, crystal physics, and special projects.

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.

Bench scale studies have been made of the recovery of uranium from acid leach liquors (and slurries) by solvent extracting with di (2-ethylhexyl) phosphoric acid in an organic diluent. Uranium may be stripped from the organic solvent by either alkaline or acidic reagents, the former having been studied in greater detail. On the basis of these tests, a recovery process may be considered which shows promise both from the standpoint of operation and chemical costs. Under proper conditions, vanadium can also be extracted by the di (2-ethylhexyl) phosphoric acid and stripping again may be accomplished with either acidic or alkaline reagents. Preliminary studies have been made of these possibilities. In addition to di (2-ethylhexyl) phosphoric acid, some other organophosphorus acids, have been cursorily examined in respect to their extraction and/or stripping performance.

From abstract: "This report presents the laboratory development of the MTR-RaLa Process for the production of barium140 from Materials Testing Reactor assemblies." The laboratory studied four possible chemical schemes with the final process being chosen "on the basis of chemical yield and purity of product, ease of recovery of uranium235, simplicty of equipment, and overall time required for processing" (from introduction).

From abstract: "The methods of applying results of bulk-shielding measurements to the design of reactor shields is outlined. Geometrical transformations for the more common shapes are derived, as are approximate means of calculating leakage of radiation. As an illustration of the methods, the ORNL Lid Tanks and Bulk Shielding Facility data are transformed to a standard geometry and compared. Two direct calculations of water attenuation, using cross sections described in the previous article, are finally compared with the experimental data."

Technical report presenting a consideration of the characteristics of fused salt-ThF4 solutions suitable for use in homogeneous reactors. Presented in tandem with a brief survey of the literature pertaining to such solutions and a summary of the experimental work accomplished. [From Abstract]

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 …