A report about aluminum-clad fuel elements, fuel-element development, zirconium uranium alloys, corrosion of zirconium, and reactor material development.

A report about elevated-temperature properties of dilute uranium-aluminum and uranium-zirconium alloys. Hot-hardness measurements show that small amounts of wither aluminum or zirconium increase the hardness of uranium at elevated temperatures.

A report about the properties of dilute uranium alloys. The mechanical properties of cold worked sirconium and zircaloy 2 at temperatures up to 500 C ore being determined.

A report which discusses the response of dilute uranium alloys to heat treatment with the ultimate objective of developing high-strength fuel elements.

A report based on a study about the factors which affect the amount of chemical reaction between water and Zircaloy 2 at high temperatures. Also, experimental programs for the measurement of radiation emissivity, chemical reaction rates, and diffusion rates have been completed.

A report about mechanical properties of dilute uranium alloys are being investigated in an effort to develop a high-strength alloy for fuel elements. Elevated temperature tensile tests were made on two dilute uranium-aluminum alloys, and several ternary alloys were arc melted.

A report about a study of the effect of irradiation on the thermal conductivity of uranium.

A report about the use of clad specimens to successfully measure the thermal conductivity of irradiated uranium.

A report which is about an investigation to determine the solubility of uranium in thorium.

An experiment designed to evaluate the in-pile performance of the U-Mg fuel material when irradiated to high burnups has been completed. Twelve specimens of the fuel material which contained uranium particles that packed 50 volume per cent, (91.5 weight per cent), uranium in a magnesium matrix were canned in Zircaloy cans and irradiated in the Materials Testing Reactor to 0.1 (1000 MWD/T), 0.3 (5000 MWD/T), 1.0 (10000 MWD/T) and 2.0 20000 MWD/T) per cent burnup of the total uranium atoms; more exactly, 1 MWD/T = 1.16 x 10⁻⁴ per cent burnup of the total uranium atoms. Irradiation of the twelve capsules began on August 1, 1954. The burnup figures used in this report are calculated values assuming a conversion ratio for the capsules of 1.0. Because of the lack of confirmed experimental burnup data for exposures of this magnitude, there is a possible error in the calculated values of about 20 per cent at 2.0 per cent burnup. However, recent results based on chemical analysis for cesium indicate that the calculated values of burnup agree quite closely for the higher exposures. Burnup estimates based on the results of the chemical analysis will be published when they become available. Six of …

A new gas-liquid countercurrent system (the ANCO system from Anisole-Complex) for the enrichment of boron isotopes has been developed. It is believed that use of this systems will result in a considerably lower unit cost for enriched boron-10 than was previously possible. The system utilizes the exchange reaction between BF3 (gas) and BF3·anisole (liquid) to concentrate boron-10 in the liquid phase. The single stage isotopic separation factor for this system has been found to vary from 1.039 at 0°C to 1.029 at 30°C. The isotopic exchange reaction has been shown to be rapid. Vapor pressures of the complex as a function of temperature have been measured and the heat of formation of the complex determined. Laboratory experiments show that quantitative removal of the BF3 from the complex can be accomplished by heating. A complete miniature ANCO plant was constructed and operated in the laboratory to test the feasibility of the system. The system was found to operate efficiently with a minimum of attention, and to enrich the isotopes of boron as expected. Based upon the experience obtained with the laboratory ANCO unit, a pilot plant large enough to utilize a 6-inch diameter exchange column was designed. The design calculations of …

Kilogram quantities of delta-stabilized and pure plutonium turnings can be cast directly into ingots of normal quality with high yields. This report describes the equipment and process used.

Past experience has shown that the demand for increase separations plant capacity comes up very regularly. One of the variables which greatly affects plant capacity is cross-sectional area of the individual vessels. Larger areas permit greater flow rates as well as more space for the installation of heat transfer piping (shell and tube concentrators). Design considerations of the separations plants vessels have been based on both circular cylinder and slab geometries. A study has been made to determine other vessel geometries that will result in safe vessels from a nuclear safety standpoint and at the same time offer larger cross-sectional areas than right circular cylinders. Vessels of elliptic as well as annular cross sections have been considered. It is neither the intent of this study to discuss the effects of intersection, vessel piping, etc., nor the pros and cons of fabricating feasibility and structural strength of these different shaped vessels. The main purpose is to make comparisons of cross-sectional areas (capacity parameter) of safe vessels so that vessel shape may be evaluated as one of the parameters in any design study for separation plants.

An alpha scintillation method for the analysis of uranium accumulation in plants as an indicator of uranium deposits in the soil was developed.

Contemplated higher tube power for future reactor operation will probably require a fuel element jacketing material more corrosion resistant than presently available aluminum alloys. Zirconium and its alloys are generally regarded as the most promising jacketing candidates for high temperature operation, particularly for exposures of long duration. In order to obtain assembly, welding, and corrosion data, twenty Al-Si bonded and twenty unbonded Zircaloy fuel elements were prepared for KER loop testing. This report describes the technique developed to weld Zircaloy jacketed fuel elements and presents the results of end closure corrosion testing and metallographic examination.

Abstract. The design, selection, and calculation of the Army Package Power Reactor shielding are described. The APPR-1, a prototype of a package reactor for remote locations, has a primary shield of iron and water. this shield has been adopted to permit fast erection and to provide low transported weight. economically, including transportation cost, the iron water shield is better than a lead water shield and is competitive with a concrete shield for a remote site. Because of the location at Fort Belvoir,Va., the shielding requirements for the APR-1 are considerably more stringent than those for a reactor at a remote base. Since the secondary shielding which surrounds the entire primary system must provide protection for personnel at any location outside the vapor container, concrete is provided for this need.

The work described in this report was done to determine the path of decomposition of plutonium (IV) oxalate and to determine the factors affecting the reactivity of the oxide with the hydrogen fluoride.

The enthalpies and heat capacities of seventeen fluoride mixtures in the liquid state have been determined using Bunsen Ice Calorimeters and copper block calorimeters. The fluoride mixtures were composed of the fluorides of two or more of the following metals: lithium, sodium, potassium, beryllium, zirconium, and uranium. The enthalpies and heat capacities of most of these mixtures were studied in the solid state also. Estimates of the heat of fusion have been made. General empirical equations have been developed which represent the enthalpies and heat capacities of the fluoride mixtures in the liquid and in the solid state.

Summary: The justification of using polyethylene whose hydrogen density of 0.132 gm/cm³ with a distributed void of 15.9 percent as a mockup of water at 70°F and having a hydrogen density of 0.111 gm/cm³ was tested in the FCE. A mockup close to the BORAX core was built and its critical mass determined. Corrections were calculated for differences in the hydrogen desnity and self shielding of the fuel. The effective FCE critical mass agreed with that of the BORAX core to within one percent.

Studies are described of the percutaneous absorption and gastrointestinal absorption in the rat of ruthenium administered in the form of "insoluble particulates." Results are also reported on the gastrointestinal holdup of these particulates.

For some time, it has been the opinion of the personnel of this laboratory and other investigators that an appreciable amount of the corrosion observed on stainless steel in nitric acid solutions might be due to the presence of some of the lower oxides of nitrogen; NO-, NO-2, etc. If this assumption is correct, the elimination of these compounds from process solutions should result in a significant increase in the service life of equipment handling nitric acid, such as acid fractionators. Since the corrosion problems experienced in acid fractionators and concentrators are very severe, an investigation of this theory becomes highly desirable. The purpose of this report is to summarize the work performed, to date, by this laboratory in the investigation of the inhibition of nitric acid corrosion of stainless steel.