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.

It is the purpose of the Fuel Element Technical Manual to Provide a single document describing the fabrication processes used in the manufacture of the fuel element as well as the technical bases for these processes. The manual will be instrumental in the indoctrination of personnel new to the field and will provide a single data reference for all personnel involved in the design or manufacture of the fuel element. The material contained in this manual was assembled by members of the Engineering Department and the Manufacturing Department at the Hanford Atomic Products Operation between the dates October, 1955 and June, 1956. Arrangement of the manual. The manual is divided into six parts: Part I--introduction; Part II--technical bases; Part III--process; Part IV--plant and equipment; Part V--process control and improvement; and VI--safety.

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.

This report provides a summary of the Hot Semiworks Unit Run PX-13. During Run PX-12 acceptable HA Column decontamination was demonstrated; however, generally under unstable column operating conditions. The principal objectives of PX-13 were: To duplicate the high PX-12 decontamination factors under stable HA Column operating conditions; to study the effects of fresh solvent, IO Column CP carbonate washed solvent, and IO Column Technical grade carbonate washed solvent on A Column decontamination factors; to study the effect of increasing the HA, HC, and IA uranium processing rates to 14 tons per day (plant equivalent); to study the effect of an intermediate 2A Column scrub on UX{sub 1} decontamination; and to study the effect of sulfamic acid addition to the HAF, HAS, IAF, and IAS and NaNO{sub 2} omission from the HAF and IAF.

The operating conditions to which the special Mark VIII quatrefoils will be exposed during the proposed piloting program have been compared with the conditions required to cause burnout, using an established method of calculating these conditions. The results of this comparison permit the following conclusions to be drawn: (1) With normal flow of coolant through the special elements the heat flux to be encountered in the R-8 cycle (1400 MW) will be 70% or that required to cause burnout (30% margin from burnout). (2) With a reduction of coolant flow to 82% of normal through one tube of a special element, burnout of that fuel column is possible in the R-8 cycle. (3) In the R-6 cycle (1280 MW), the margin from burnout in the special Mark VIII quatrefoils is 42% with full flow and 20% with the above reduced coolant flow. A similar comparison of operating conditions predicted for the L-3 cycle (full Mark VIII charge) shows that, even at the highest power level (1250 MW), the margin from burnout is greater than 55% with normal flow and 40% with reduced flow.

Sodium dichromate has been tested for use as a process water corrosion inhibitor at a concentration of 0.5 ppm. The test was conducted by operating one half of D Reactor with 0.5 ppm while the other half operated at the normal concentration of 2 ppm sodium dichromate. A concentration of 0.5 ppm was found to be as effective a corrosion inhibitor as the normal concentration of 2 ppm except under one very important set of circumstances. If the reactor process tube ribs were previously worn or corroded down, so that the water in the between-rib annulus became excessively hot; 0.5 ppm sodium dichromate was found to be non-effective in stopping corrosion at this location. From five to ten times as many slug ruptures and tube leaks occurred from this accelerated between-rib corrosion if the sodium dichromate concentration was 0.5 ppm rather than 2 ppm.

``Safe-failure`` is an important criterion for power reactor fuel elements designed to contain core materials which react rapidly with high-temperature coolant. Out-of-pile experiments indicate that, without a ``safe-failure`` fuel element, expensive and possibly catastrophic reactor shutdowns may occur if the fuel element jacket fails during irradiation in high temperature water. That jacket failures occasionally will occur, regardless of the material or fabrication process involved, seems to be generally agreed. Fuel elements involving cluster arrangements, oxide cores, and wafer assemblies with ``bulkheads`` for compartmentalization have been proposed as being relatively safe in event of jacket failure. A fourth possibility is a type of long cartridge fuel element.

In April of 1955 the Radiation Laboratory of the University of California sent out to 18 high schools in the Bay Area letters which said, in part, 'The object of this program is to help science teachers achieve some practical knowledge and experience in nuclear science by working with some of our basic research groups in physics and chemistry. In this way we believe that we can assist these teachers in their efforts to make their students more aware of the nature of atomic energy and help to prepare them for the impact of atomic energy in their future.'

When acid leach liquors obtained from U ores containing Mo are contacted with anion exchange resin, both U and Mo are adsorbed. The adsorbed Mo is not removed in the norraal U elution process, and when the U-barren resin is recycled, the resin sites already occupied by Mo are unavailable for U adsorption. In succeeding cycles, more and more sites become occupied by Mo, and the resin capacity for U is seriously lowered. This rapid accunmulation of an adsorbed species is known as resin poisoning.'' The many disadvantages of such a situation are self-evident. (auth)

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S>The Zircex Process for the recovery of zirconiuun from zirconium clad fuel elements is presented. Various types of apparatus were investigated for desublimation and collection of the solid zirconiuun tetrachloride. Of primary importance is the determination of particle size and distribution of the ZrCl/sub 4/ condensed for, in general, the larger and more uniform the particle size, the easier the collection. (W.L.H.)

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Feasibility studies of a high flux, solid fuel research reactor are presented. By means of a parameter study, a reactor consisting of a cylindricai fuel annulus submerged in heavy water is described. The thermal neutron flux peaks in the heavy water adjacent to the annulus and is a maximum in the region surrounded by the fuel annulus.. while the minimum thermal flux occurs in the annulus. The fast flux has the opposite shape. Calculations indicate that practical peaking factors, ratios of maximum thermal flux in the heavy water to average thermal flux in the fuel annulus, as high as eight can be obtained. A typical reactor in which the maximum thermal and fast fluxes are greater than 10/ sup 15/n/cm/sup 2/-sec is also described. (auth)

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