Tests were conducted on a laboratory and semi-plant scale to determine the effect of permitting scrap grossly contaminated with uranium to be used in steel manufacture. It was found the most of the uranium is removed with the slag. Steel made with this scrap would have a uranium constituent so little above that made with uncontaminated scrap as to be hardly significant. The slag itself would not present any hazard in handling or normal use. It is recommended, therefore, that in the future steel with only surface uranium contamination be released through normal scrap channels.

The relative physiological and toxicological properties of americium and plutonium have been studied following their intravenous administration to rats. The urinary and fecal excretion of americium was similar to that of plutonium administered as Pu(N0{sub 3}){sub 4}. The deposition of americium the tissues and organs of the rat was also similar to that observed for plutonium. The liver and the skeleton were the major sites of deposition. Zirconium citrate administered 15 minutes after injection of americium increased the urinary excretion of americium and decreased the amount found in the liver and the skeleton at 4 and 16 days. LD{sub 30}{sup 50} studies showed americium was slightly less toxic when given in the acute toxic range than was plutonium. The difference was, however, too slight to be important in establishing a larger tolerance does for americium. Survival studies, hematological observations, bone marrow observations, comparison of tumor incidence and the incidence of skeletal abnormalities indicated that americium and plutonium have essentially the same chronic toxicity when given on an equal {mu}c. basis. These studies support the conclusion that the tolerance values for americium should be essentially the same as those for Plutonium.

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Slag and crucible can be dissolved satisfactorily by the Los Alamos method of total dissolution with nitric acid in the presence of aluminum nitrate. Extraction of 99% of the plutonium from total dissolution salted with aluminum nitrate was achieved by three successive contactings with 1/10 volumes of 30% TBp- AMSCO 125-90W. The TBPAMSCO phases contacted with leaching solutions salted with calcium nitrate must be scrubbed to remove calcium. Stripping with three 1/10 volumes of 0.1M hydroxylamine sulfate removes 99% of the plutonium from the 30% TBPAMSCO, initially O.lM HNO/sub 3/. Plutonium(III) oxalate, (which could be blended into 234-5 operations) can be precipitated from the aqueous strip solution. It is chemically feasible to recycle slag and crucible solution to the Redox IIA column in amounts up to at least lO% of IIAF by volume. It is also chemically feasible to recycle nitric acid solution of slag crucible obtained in the presence of aluminum nitrate to the Product Precipitation step of the Second Decontamination Cycle of the Bismuth Phosphate Process. (auth)

The purpose of the research work presented in this volume is to develop a process for the separation and purification of actinium-227 produced by neutron bombardment of radium-226 and to develop methods by which uniform films of actinium metal may be deposited on metallic surfaces. The design work on the cave structure and mechanical equipment used in the actinium separation is proceeding on schedule. As the mechanical design phase is nearing completion the emphasis is being directed toward processing equipment. The process as well as the mechanical equipment has been adapted from the research work of F. T. Hagemann and the Remote Control Group at Argonne National Laboratory. Consequently, one of the first objectives is to become familiary with the chemistry of the process and the operation of the mechanical equipment. Cold runs have been made on the T.T.A. benzene extraction using lanthanum and barium in place of actinium and radium. No difficulty with the operation was observed. The formation of precipitates was one of the difficulties encountered with the process as the precipitates carry radium. It has been found that metals such as nickel cause these precipitates to form and should, therefore, be avoided in the construction of equipment. …

The purpose of the research work reported in this volume is the development of a process for the separation and subsequent purification of radium from the K-65 (pitchblende) residue. Except for the accumulation of additional experimental data the process is essentially complete. After a preliminary extraction of about 85% of the lead and 40% of the silica with a 40% sodium hydroxide solution, the residue is treated with a solution containing both sodium hydroxide and sodium carbonate to obtain additional lead removal and partial conversion of the radium and barium sulfates to carbonates. After leaching out the soluble carbonates, the residue containing unconverted sulfates is treated with a sodium carbonate solution at 170{sup o} under pressure. A study of the amount and concentration of sodium carbonate required to obtain a satisfactory conversion of the sulfates has shown that the total sodium carbonate may be reduced by 40% and that smaller volumes of solution may be employed without serious decrease in the efficiency of the conversion. In order to make calculations on the number of steps and tank sizes for the radium-barium separation by fractional precipitation, equations have been developed to calculate this information from the theoretical separation factors determined experimentally. …

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Experimental techniques are described for the spectral measurement of a collimated fast-neutron beam. A H/sub 2-/ filled cloud chamber, proton-recording nuclear plates, and threshold fission foils were used as neutron detectors in the measurements. As an application of these techniques, the energy distribution and absolute flux of the fast neutron beam emerging from the Los Alamos fast reactor was measured from 0.1 to 18 Mev. (D.E.B.)

This report addresses the corrosion of zirconium in 600°F water and in 750°F superheated steam.

On 2/8/1950, there was mined from process tube 2273 in D pile at Hanford a quantity of graphite power, which was expected to show the most extensive radiation damage of any graphite available at that time. A series of samples of this powder were annealed in 100 degrees increments from 100 degrees to 2000 degrees C at this labaoratory. There were returned to Hanford and shipped by them to the National Bureau of Standards for total stored energy measurements. The present memorandum is comprised of a description of the annealing procedure used here, curves giving the detailed annealing history of each sample, and various curves derived from data obtained from these samples at Hanford and at the National Bureau of Standards.

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A study was made of the stability of tributyl phosphate (TBP) to acids, bases, and UO/sub 2/(NO/sub 3/su)/sub 2/ solutions, and of the chemistry of its decomposition products. The effect of such decomposition products on the extraction behavior of ion of MBP and lines the RAW analyzer and a subsequent reduction of feed flow rate in each case resulted in immediate reduction in losses, as evidenced by the polarographic results. In many such instances, several hours of operation at high waste losses were avoided. A dropping mercury electrode was used for stream analysis, and the RAW solution was analyzed directiy without either dilution or addition of supporting eleclrolyte. Removal of dissolved oxygen and control of temperature were urnecessary. The precision of the method was better than plus or minus 5%, and the accuracy, which is dependent on calibration with a standard solution, was well within the precision value. The calibration curve is linear, for practical purposes, up to approximately 8 g/l UNH or 0.032 lb/gal uranium. (auth)

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