This report details Technical Division activities for the week ending May 22, 1943.

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A comprehensive survey of the literature was made in an effort to identify nonmetallic materials of possible usefulness as liquid coolants. Materials having maximum melting points of 1000 deg F and boiling points of l200 deg F were considered, but boiling points above 2200 deg F were preferred. Melting points, boiling points, densities, heat capacities, and thermal conductivities were tabulated. Approximately 190 materials appeared to have melting and boiling temperatures in a suitable range. A paucity of thermal- conductivty and heat-capacity data prevented further estimates of suitability of all but nine of these materials. Of these nine nonmetallics, only sodium hydroxide appeared to offer possibilities when considered according to the NEPA formula. (M.C.G.)

During the preliminary operaton of a volatilization apparatus it became apparent that a large stopcock in the high vacuum side of the oil diffusion pump was quite unsatisfactory. Since it was necessary to mount the stopcock in a relatively inaccessible location, the difficulty of turning the plug and the consequent necessity of regreasing the valve approximately every week were real problems. Also, it became apparent that an increased pumping speed through the valve would be required. Since a larger stopcock was out of the question, a type of valve reported by Stern was modified in design and was bulit into the apparatus.

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This report provides Part III through VI of the Material and Equipment Section`s activities at New York Shipbuilding Corporation. Fabrication, inspection, and testing of reactor components are detailed.

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This report consists of Chapter Six of the Nuclear Metallurgy Lectures. Chapter six discusses the reduction of uranium, chemical composition of uranium ores, extraction process for uranium ores, bomb reduction, green salt, and dingots.

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The study of interactions and decay of K mesons continued, using the 10- inch liquid hydrogen bubble chamber and emulsions. Four emulsion stacks were exposed for two internal groups and twenty-three stacks were exposed for seventeen groups from outside the Laboratory. Interactions of pi /sup -/ mesons were observed with a 30-inch propane bubble chamber and with emulsions. Ten emulsion stacks were exposed to neutral-particle beams and one stack to the internal 6.2-Bev proton beam. The 30-inch propane bubble chamber and emulsions were used to study the interactions of antiprotons. Eighteen target bombardments in the internal proton beam were made for the chemistry group. Successful tests were completed of two static-electromagnetic velocity spectrometers for the separation of highenergy particles. One was of coaxial construction, the other of parallel-plate construction. (For preceding period see UCRL-6114.) (auth)

Somewhere in the future lies the possibility that the existing HAPO reactors will no longer be required for the production of weapons grade plutonium. This could come about either through advances in technology or through changes in the requirements for this particular fissionable isotope as a weapons material. In any event, should these reactors no longer be utilized for plutonium production, it would be appropriate to consider the possibility of their conversion to other uses, notably that of electrical power production. Accordingly, this study has been made with the objective of investigating the economic and technological feasibility of such conversion. However, it must be recognized that the results presented herein are based on present day technology and economic philosophy and that within the next five years we may experience considerable renovation of these criteria which could alter the conclusions of the study appreciably. The primary benefit to be gained from this preliminary study is to narrow the area of future and more detailed investigation as much as possible so that if and when such a conversion appears likely, a significant fraction of the almost infinite number of combinations of reactor and power plant parameters may be rejected at the outset.

A new method of introducing predetermined quantities of hydrogen into zirconium was developed and evaluated. The procedure is simple, effective, and inherently good in preverting oxygen and nitrogen contamination. Zirconium samples containing four levels of dissolved hydrogenated to 100, 300, 3500, and 7700 ppm hydrogen. Dissolved oxygen was found to decrease hydrogen absorption in both the alpha and beta modifications of zirconium. The effect of hydrogen on density, hardness, and microstructure was determined. X-ray-diffraction measurements were also obtained. (auth)

Results of four experimentul runs in the Fluorox fluidized bed reactor system are reported. The engineering feasibility of UF/sub 6/ production from UF/ sub 4/ by use of dry air of O/sub 2/, 2UF/sub 4/ + O/sub 2/ = UF/sub 6/+ UO/sub 2/ F/sub 2/, in an Inconel fluidized bed reactor at 800 to 850 deg C was demonstrated in two experimental tests in which greater than 90% of the theoretical amount of UF/sub 6/ was collected or measured. Two runs made with crude UF/sub 4/ (produced from unpurified mill concentrate) as the feed material, showed thnt UF/sub 6/ could be produced at 700 to 725 deg C but corrosion on Inconel was prohibitive. (auth)

The proposed operating conditions for fuel elements to be charged into the NPR require the fuel to be of an extended surface geometry and maintain adequate strength and corrosion resistance in 300 C water. A contract was let to Nuclear Metals Inc. to produce by co-extrusion lengths of fuel rod containing both natural and 1.6% enriched uranium of irradiation quality for fabrication into fuel elements. The fuel rods used in the irradiation test represent the first enriched uranium rods coextruded in 0.030 inches of Zircaloy-2 to be irradiated and examined at Hanford. The rods used for this test were fabricated into four, 4 rod cluster fuel elements thus allowing adequate space between individual rods for expansion in the case of a fuel rod failure. This rod was of particular interest since it contained an irregular uranium-Zircaloy-2 interface. The purpose of the irradiation was to determine the dimensional stability of coextruded fuel rods and to determine whether the irregularity in the bond interface had any effect upon the irradiation performance of the fuel. Fuel elements were irradiated in 200 C water in the KER Loop 2 facility to an exposure of 0.28 a/o burnup (2,200 MWD/T). Post irradiation examination showed that …

At the present time, plants in the 100 to 500 Mw size range are more numerous and carry the greatest portion (over 50%) of the total steam-electric plant load in the US utillties industry. The contribution of plants of over 1,000-Mw capacity is increasing more rapidly than any other size classification and at present represents about 10% of the total capacity. By 1962 the TVA will have six plants with capacities of over 1,000-Mw. The largest steam-electric plant in the US is the TVA plant at Kingston, Tenn., with a nameplate capacity of 1,440Mw. Turbine-generator wits are also following a trend of ever-increasing size. In present construction, the 150 to 200 Mw size range for units ls the most common and represents the greatest contribution to capacity. Two units of 500-Mw nameplate rating each, the largest in the US, are on order by the TVA, and an 80 Mw unit is contemplated. (auth)

By applying the relationships between lattice parameters and composition in the solid solution range, a method was presented to analyze boron in boron carbide. The purpose was to establish a means of determining the boron composition in a boron carbide solid solution in the presence of excess carbon or boron. The different unit-cell dimensions sometimes reported for a given solid mixture are usually attributed to the existence of solid solutions. If the crystal lattice parameters vary linearly with composition, a curve can be constructed to determine composition from lattice parameter measurements. Toward this end four samples of boron carbide of known nominal composition were selected. For each sample precision lattice parameter measurements and chemical analyses of boron and carbon content were obtained to establish relationship between lattice parameters and boron composition. (auth)

This document presents experimental data obtained during a series of tests which were completed at 105-D and DR Reactors in February and March, 1960. No analysis of the data is included in this document. The tests were: PT-301-I, II -- Reactor cold, full flow, BPA power failure; PT-278-A, III B -- 1170 MW, full flow, BPA power failure; and PT-278-A, II -- 1190 MW, full flow, poison push causing bulk surge and scram.

Low attack rates (0.1 to 0.5 mpy) were displayed by coupon specimens of type 347 stainless steel, titanium RC-55, and Zircaloy-2 which were exposed for 2877.5 hr in an oxygenated slurry of Th--8% U oxide, 116.5 hr in water, 6.9 hr in 5% HNO/sub 3/, and 4.3 hr in 3% triscdium phosphate during mns SM-6 through SM-9 in the slurry blanket mockup. The leading coupon of type 347 stainless steel showed a slightly higher rate than the other stainless steel ccupons due to entrance effects. Specimens of SA-212-B carbon steel displayed average attack rates of 2.9 mpy. (auth)

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The objective of this test authorization is to increase reactivity and thus reduce short term enrichment requirements by increasing graphite temperature limits during low exposure operation following full central zone discharge.

This report discusses the delayed neutron characteristics of a reactor which are a function of the distribution of fissions in the various fissionable isotopes. The delayed neutron characteristics of the NPRL delayed fraction and decay constants, are presented as functions of exposure to 2000 MWD/T for both room temperature and operating temperature. It is of importance to note that the delayed neutron fraction decreases from 0.693% to 0.539% with increased exposure. Thus 22% less reactivity change represents a prompt critical condition at 2000 MWD/T, compared to the zero exposure condition.

Tests were performed to obtain station performance data at various steady-state generator loads. The station 0 was operated for four-hour periods at steady state conditions and levels of 5, 21, 42, and 61 Mw gross generator output. The various readings are presented in tabular form. A list is given of equipment in service during the test. All plant components operated satisfactorily during the test. (M.C.G.)

The voltage and frequency response of selected SM-1 plant system parameters to step load changes was analyzed using analog computer measurements. The analog model was that developed for analysis of the SM-2 design. The approach to the analysis, formulation of the model, and analog recordings are presented. The data will be used to prove reliability of the analog model by comparing analog data with test data to be taken at SM-1. (auth)