Report summarizing corrosion tests meant to determine the corrosion rates of bronze, mild steel, T-304 stainless steel and 2S aluminum in 10% potassium tetraborate solution. Tests were completed under the boiling point of each solution except for 2S aluminum, which was tested at 40C, the approximate temperature in a pile.

An ex-reactor test stimulating the charging of a tubular fuel element in a Zircaloy-2 process tube showed a problem of severe process tube scratching and galling may occur if metal sliding contact fuel element supports are used. Both the fuel supports and the process tubes were autoclaved. Although the conditions of this test were severe in that the sharp edges rather than a flat surface of the supports were in contact with the process tube, it pointed out the potential problem of galling by point on small area contact supports on heavy fuel elements. This difficulty had not been observed with the lighter tubular or cluster fuel elements used in KER irradiation. In addition to the loss of metal and thinning of the process tube after a number of charge-discharge operations, these scratches may lead to localized corrosion or act as points of stress concentration.

This report describes the work carried out in 321 Building semiworks equipment, to define the factors contributing to high corrosion-product contamination and presents recommendations for reducing the impurity level to meet current specifications (maximum of 200 parts total metals per million parts U).

The effects of oxidation and irradiation on the microsurface structure of pile graphite have been investigated by measuring the surface area and pore size distribution of several samples. The results obtained for both oxidized and irradiated graphite samples indicate that changes in surface characteristics which occur are determined by the flux, temperature of irradiation, and gaseous atmosphere in which the radiation takes place.

Unanticipated radioactive contamination of the 200 Areas environs was discovered in the latter part of 1947, about 2.5 years after the startup of the Bismuth Phosphate Separation Plants. It was subsequently established that this contamination was due to radioactive particles emanating from the Separation Plants' stacks, and the need for exhaust ventilation air filtration equipment was recognized. Sand filters were installed in the fall of 1948 in the ventilation systems of the B and T Bismuth Phosphate Plants. These filter units essentially eliminated this contamination problem and have performed satisfactorily to the present time.

Many procedures for the chemical isolation of the microscopic angular inclusions in rolled uranium have been investigated. Four were selected as most satisfactory for this work and will be described in detail. The procedures are: (1) the copper displacement of the uranium, (2) the electrolytic displacement of the uranium, (3) the chemical dissolution of the uranium in an anhydrous solution of hydrogen chloride in methanol, and (4) the chemical dissolution of the uranium in an aqueous solution of hydrogen chloride and hydrogen peroxide.

In support of an in-reactor thermocouple stability testing program now in progress, laboratory studies were made relative to the stability of thermocouple bundles similar to those projected for in-reactor use. Two sets of chromel-alumel, copper-constantan and iron-constantan thermocouples were held for extended periods at 300 C in successive atmospheres of 75% helium-25% carbon dioxide. 100% carbon dioxide and air. Thermocouple stability was verified by measuring emf's at the freezing point of lead. The results of these tests are to be used to evaluate data obtained from like thermocouples subjected to condition similar to the laboratory tests with the added facet of reactor irradiation.

Maximum safe cylinder diameters, slab thicknesses, and minimum critical masses in spherical geometries for fuel of < 5.0& U-235 enrichment have been evaluated. It is of interest to know the minimum critical masses as a function of cylinder diameter for vessels that are not safe by geometry. Curves will be presented of critical masses as a function of cylinder diameter and fuel rod size for 1.6%, 3.0%, and 5.0% U-235 enriched metal-water systems.

Neutron radiation can be greatly reduced by substituting plutonium trichloride for plutonium tetrafluoride in the dry chemistry step of metal preparation. Chloride processes are now common in the metal preparation industry, and the 234-5 Developmental Laboratory has shown plutonium trichloride can be satisfactorily prepared and reduced to the metal.

In recognition of the need for an integrated system plan for future development requirements for radiological instrumentation, this information report was written to provide the initial step concerning such a planning. The report is to be considered a basic information analysis of the present in-use plant radiological instrumentation, of the instrumentation now undergoing active development, and of the future development work contemplated.