The following documents are progress reports that follow chemical research studies in subjects such as ruthenium tetroxile distillations, extraction-scrub studies of zirconium and niobium, and oxidation of plutonium and stabilization of plutonium(VI) during extraction of IAF solutions.

The following report was made to determine the effective clarity improvement, flow rate characteristics, and operational problems encountered on passing Redox uranium feed solution through sintered, modified Type 304 stainless steel filter elements.

The following report provides results from laboratory tests on the corrosion resistance of various types of stellite facings on astenitic stainless steel T-347 to Redox streams.

The following report describes a successful procedure used for the determination of americium in the presence of plutonium. Provided are articles listed that describe the research done in developing the method.

"An apparatus was designed and fabricated to obtain film samples from irradiated process slugs. This apparatus has been used to obtain film samples from slugs of the sodium dichromate elimination test."

From introduction: "Analytical procedures for the determination of plutonium samples from the extraction and concentration plants utilize the carrying of the hydroxylamine-reduced trivulent plutonium by coprecipitation on lanthanum fluoride. Most other plutonium procedures also take advantage of the efficient carrying on lanthanum fluoride."

The following report describes the process to determine the method for determination of americium in the presence of plutonium and uranium.

This report presents data recorded during a program made to improve uranium slug quality in January, 1952, to determine what improvements could be made in the welding process.

The following report observes infrared measurements in the region of OH bands, the C2O band vicinity, the low frequency end of the spectrum, and the region of the UO2 band on the system hexone - water - uranyl nitrate.

From introduction: "The prime motive behind the experiments described in this report and in those preceding it in this series is the desire to reduce the number of Redox cycles necessary for decontamination from fission products of the plutonium and uranium streams sufficient to allow final disposition of these two products...This report deals with the mechanics of manganese dioxide handing, laboratory development of manganese dioxide scavenging, and a series of experiments at multicurie level combining ruthenium volatilization and manganese dioxide scavenging in stainless steel equipment."

Summary: "This investigation showed that a sound , weather-resistant coating can be applied to uranium that the surface to be plated is cleaned and etched properly. Using the proposed two-bath plating method at 80 C., a four-mil nickel coat can be deposited in thirteen hours."

This report follows an investigation on how hydrolysis products of tributyl phosphate (such as dibutyl and monobutyl phosphoric acids, orthophosphoric acid and butanol) effect the tributyl phosphate process for uranium recovery.

The following report follows a test for remelting and casting uranium scrap specified that a pressure of 25 microns or less must be attained within the furnace before a charge is melted. The report was made to evaluate the effect of furnace pressures upon the quality if recast uranium metal.

The following document analyzes plutonium in urine which uses nuclear track film for evaluation of the quantity of separated plutonium.

The study of the flame photometric method for determining the concentration of tributylphosphate (TBP) in a hydrocarbon diluent was undertaken because it showed promise of being economically competitive and subject to fewer interferences than the existing methods. The determination of TBP in the hydrocarbon solutions throughout the range 0.01 to 80 per cent TBP may be performed using the flame photometer. The principal conclusions resulting from the study found a that there was higher precision in the one to two percent TBP rang than the 45 per cent TBP level due to error in reproducing narrow slit width in the higher range, 2) nitric acid, uranyl nitrate, and thorium nitrate all interfere if present but can be eliminated by contacting the solution with water or sodium carbonate solution. 3) a single determination may be made in 15 minutes, multiple determinations can be made at a rate of one to two minutes per determination, and 4) a modification of an existing method makes possible the application of this method to the determination of TBP dissolved in aqueous samples. A relative precision of plus of minus five per cent (95 per cent confidence level) is obtained.

A plutonium solution of known Pu-241 content was purified by solvent extraction techniques to remove uranium and americium impurities. Portions of this purified solution were then analyzed for U-237 as a function of time. Since the growth of U-237 occurs directly by the alpha decay of Pu-241, these measurements provided sufficient data for calculation of the Pu-241 alpha half-life, which was found to be 2.91 ± 0.50 x 10⁵ years.

The purpose of this report is to present a HAPO proposal for the design and construction of a high-temperature, high-pressure light water recirculation test facility for the Westinghouse Electric Corporation, Atomic Power Division This report does not attempt to justify the feasibility of site location nor does it attempt to elaborate on such technical considerations as permissible test specimen heat generation and enrichment requirements which have been covered in an earlier document. Instead, detail information is developed on the proposed design with major emphasis on detail descriptions of the proposed loop and auxiliaries, design and construction scheduling and costs, and operating costs.

A new type of flexible hose connector consisting of a Teflon tube enclosed in a stainless steel wire braid sheath was proposed for use on 105-C rear face in late 1954. A test was initiated to determine the suitability of Teflon, and specifically Fluoroflex Teflon compound for rear face application. Probable useful life limit is estimated at 10⁶ R, which would be attained in about three years in contact with fresh effluent.

A method for determining the appropriate isotopic content of uranium-235 in uranium slugs has been investigated. In the conduct of the experiment it was found that there was a background counting rate several times the counting rate of the phenomena under study. Methods are suggested for improving the counting rate and reducing the background. The conclusions at this time are that reactivity determinations will provide the most rapid and precise measure of U-235 content in uranium.

The primary purpose of this writeup is to present a description of the electrical and mechanical features of the developed system for use in maintenance and operation of a simultaneous scaler control unit.

A preliminary soil column experiment with a waste sample from the 207-S-11 well, which monitors the 216-S cribs, indicated that cesium was absorbed almost completely by the soil but that strontium breakthrough to ground water in the near future seemed likely. Accordingly, it was suggested that the 216-S wastes be discharged to a new disposal site, and that samples of the wastes which are currently being discharged to the 216-S cribs be obtained for soil absorption tests. The primary purpose of these tests was to obtain data which could be used to suggest ways and means of improving the absorption of strontium by soil from the process condensate stream (D-2_ and the cell drainage stream (D-1), both of which have been discharged to the 216-S cribs for the past several months.

As recommended in a previous report a study of the NO₂ and HNO₃ exposures of plant personnel in the 224 U Building has been made. Concentrations of NO₂ and HNO₃ from the 224 U roof stack, in the concentrator cells, and in the calcining pot rooms have been determined. The problems have been outlined and recommendations have been made for its alleviation.

The determination of various parameters such as thermal utilization, resonance escape probability, control rod calibrations and etc. is often accomplished by the measurement of the pile period due to some given perturbation. With the present thoughts being directed towards plutonium fueled reactors, it was felt worthwhile to discuss a few of the basic differences in pile neutron behavior due to the smaller fraction of delayed neutrons per fission of Pu239 and the larger fraction of delayed neutrons per fast fission of U238 than occurs from fission of U235 alone.

In the fall of 1952, a pilot plant batch dissolution study was carried out to obtain the data required for the design of the Purex Plant dissolvers. Particular emphasis was placed on the rate of dissolution and the efficiency of nitrogen oxide recovery from the dissolver off-gas. The results are presented in this report.