This document summarizes the amounts of radioactive contamination discharged to the ground from separation facilities through June 1956. Detailed data for individual disposal sites is presented on a month to month basis for the period July 1955 through June 1956. The major disposal sites in separation facilities, total volume of waste discharged at each location, and the gross amounts of plutonium and fission products discharged to the ground since startup are listed. This same data are presented on a month to month basis and also include information on the source of the waste stream and the settling facility used. Isotopic data are included for all disposal sites from which the waste was analyzed for specific contaminants. Estimates of contamination and volumes discharged to the swamps are also included.

Six 1.6% enriched Zircaloy-2 jacketed seven-rod cluster elements with modified end supports and one Zircaloy-2 jacketed Pu-A1 seven-rod cluster element will be irradiated in KER Loop 1 to an exposure of about 4500 MWD/T.

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Design and construction schedules and forecast monetary allowances for projects in the 100 Areas, 200 Areas, 300 Areas, and 700 Area of Savannah River Plant have recently been reviewed. There has been no change in the overall funds now forecast for the program, although there has been changes in the distribution of the total between various areas.

A tabulation (too large to be included in the memorandum) of the hypergeometric probability distribution for lot sizes up to and including 50 is described. Uses of the tabulation in sampling inspection and reliability are described. The tabulation provides a simple way of testing the equality of two proportions when the total number of observations on the proportions does not exceed 50. (auth)

A comparison between the performance of the d-c unit as a pump and as a generator is presented.

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After almost four months elapsed time since product introduction into the RMA Line, it was pertinent to evaluate both process and equipment performance and arrive at a program for future improvement. The suggestions and recommendations provided are those of the Technical Section. In those passages made to acknowledge the views of the appropriate groups.

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The testing and evaluation of three types of mechanical joints in a circulating molten fluoride salt system, at temperatures up to 1600 F, were accomplished. The feasibility of these joints for use in a large scale molten salt system is discussed. Design criteria and operating techniques are described. Measurements were made of the leakage rates of helium through the Joints. The effects of thermal cycling, atmospheric oxidation, salt corrosion, and thermal stresses were noted. Disassembly and reassembly procedures are described. ( auth)

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Current operation of the Redox plant dissolvers results in evolution of ammonia during the slug jacket removal step. Staggering the schedules of the dissolving cycles of the three dissolvers is being practiced at present to prevent the formation of ammonium nitrate crystals in the stack gases, but this procedure limits the dissolver capacity. Steps are being taken to provide ammonia scrubbers for the dissolvers to eliminate this need for close scheduling of dissolving operations. It has been proposed (HW-30838) that mercury-catalyzed nitric acid dissolving of the jackets be adapted to the Redox plant. Such a process would eliminate ammonia from the off-gases and would offer several other advantages as well. A flowsheet was proposed based on work in the Process Chemistry laboratory and developed by runs in the 321 Building 1/20-scale dissolver. However, the off-gases resulting from this flowsheet were found to contain hydrogen, and further laboratory work was thus undertaken to provide a solution to this problem. It is the purpose of this paper to summarize the recent development work on the mercury-catalyzed nitric acid slug de- jacketing and dissolving procedure, and to show how these findings may be applied to the Redox operation by the use of alternate …

Activation of indium foils by thermal neutrons has been applied to measurement of fast-neutron fluxes. Foils are encased in paraffin spheres placed in cadmium boxes. The high-energy neutrons that penetrate the cadmium become thermal neutrons; the thermal-neutron flux is proportional to the incident fast-neutron flux over a range of about 20 kev to 20 Mev. The foils are removed from the boxes and counted on a methane-flow proportional counter. High instantaneous neutron fluxes are easily detected and counted by use of these foils. Many simultaneous measurements have been made easily by this method.

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The feasibility study of the Purex flowsheet in pulse column equipment has been completed in the ORNL Purex Pilot Plant. Eleven runs have been made in the pilot plant at a throughput of 75 kg to 125 kg of uranium per day.

Since the presence of neptunium-239 in disassembly basin water had been suggested, analysis of the water was undertaken. The occurrence of Np-239 was thought to be due to its diffusion through the slugs. Samples of water from the D and E Canals in K and R-Areas were analyzed to determine the presence of Np-239. Samples from and K and R Areas both showed Np-239 to be present in quantities greater than 50% of the initial total activity.

This report is a bibliography of studies on the toxicology of polonium with focus on issues addressed by the Atomic Energy Commission, but also includes published literature. Some of the citations contain abstracts and the reports are annotated as to contemporary classification status of reports.