Abstract: "An exploding copper wire, one millimeter in diameter, 30 centimeters long, has been developed as a fuse component for a Los Alamos capacitor energy source to be employed in controlled thermonuclear research studies. The fuse allows the passage of the high normal duty "action" (13,700 ampere-coulombs per capacitor) at a 20-second repetition rate. However, it interrupts the circuit to a shorted capacitor in 12 microseconds, thereby protecting the faulted capacitor from explosive energy consumption. The initial phase of the development included observations of various metals (copper, silver, iron, and nickel) as well as various configurations (straight wires, helixes, foils, and tubes). Direct scaling of previous small exploding wire studies at Sandia Corporation was demonstrated with scaling factors up to 700,000."

From introduction: Approximately fifty radioactive deposits and nearly fifty properties not abnormally radioactive were examined during a geologic reconnaissance for radioactive minerals in Idaho, Washington, and western Montana during the period July 1952 -- June 1955. The most important uranium deposits are in or near granitic to quartz monzonitic intrusions of probable Cretaceous age in central and northern Idaho, westernmost Montana, and northeastern Washington. The purpose of these reports is to describe the geology of the areas so that the possibilities for the disposal of high-level radioactive fluid waste in deep wells can be ascertained.

An IBM-704 shielding program is described which solves, by a finite difference method, the multigroup diffusion equation system in slab, cylindrical, or spherical geometry, with a prescribed source distribution given for the highest energy group. A wide choice of boundary conditions has been made available at the outer boundaries and either continuity of flux and current or the black boundary'' condition may be specified at each regional interface. The machine time required to process an average problem is approximately 1 to 2 minutes. This includes reading in the program deck. (auth)

From introduction: This report details the results of instrument recordings of a nuclear explosion where ground zero was near the top of Rainier Mesa.

An analytical prediction of the burnout distributions in particular SM-1 fuel elements and absorber sections after 10.5 MWYR of core energy release is given. The distributions are based on the results of a one-shot, non-uniform burnout calculation, and are pre sented for both fuel and boron -10 depletion. Particular emphasis is placed on those elements removed from the SM-1 core in March 1959, since their subsequent burnup analysis by ORNL should provide a valuable check on the analytical models employed. (auth)

Fuel elements clad with Zr or containing Zr as a diluent can be recovered by a fluoride-volatility process. The first step consists of hydrofluorination of the elements in a bath of molten fluoride salts using an HF sparge. In this case the two salt systems considered were NaF-ZrF/sub 4/ and NaF- LiF. Materials evaluated at Battelle for possible use in the construction of this hydrofluorinator include Inconel, A'' Nickel, copper, silver, Monel, Hastelloy B, Hastelloy W, INOR-1, and INOR-8. The metals were exposed to molten fluoride salts through which HF was bubbled continuously. The data indicate that the NaF-LiF systems are much more corrosive than the NaF-ZrF/sub 4/ system. The systems are most corrosive when the alkali fluoride component is high. An elevation in temperature increases the corrosion significantly as does an increase in the HF flow rate. Hydrogen in the HF flow stream retards the corrosion of the sodiumzirconium salts significantly, but appears to have less effect on the sodium -lithium systems. The areas at the interface of the liquid and vapor phases were most seriously damaged under the exposure conditions usually used. However, appreciable reduction in attack was experienced when zirconium was actually hydrofluorinated. INOR-8 was the most …

FOLDER is a routine whose purpose is to solve a system of linear differential equations on the IBM 650 computer which is equipped with index accumulators and floating-point hardware. A system of K n/sup th/-order differential equations which is represented by a system of K x n first-order differential equations and coded in a prescribed manner is solved numerically by the methods of Milne and/or Runge-Kutta. Flexible operating modes are made possible by the user's designtion of parameters. Output data and format are under complete control of the user. (auth)

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Several experiments have been performed at ORNL with light water solutions of uranyl nitrate (highly enriched in either U/sup 233/ or U/sup 235/) in an essentially bare sphere 27 inches in diameter. Results are presented of several calculations with elementary bare reactor theory and a discussion of the observed discrepancies between the calculated and experimental results. If the observed critical concentration is used in the calculations, the calculated effective multiplication constant is less than unity; thus a higher critical concentration would be predicted than is actually observed. ( auth)

4 -Metallurgical Processing. A direct-cycle fuel reprocessing plant is being designed for pyrometaliurgical processing of discharged power reactor fuel elements. Irradiation tests on cerium-bearing glass samples for shielding windows revealed that the optimum cerium content is less than the nominal amount originally specified. The light output of gammairadiated mercury vapor lamps was determined to be about 55% of original after an exposure of 1.1 x 10/sup 9/ rads. Analyses of the composition of 30 ten-kilogram ingots of natural U-5% fissium alloy prepared in the melt-refining furnace indicate that not all of the added Zr and Mo went into solution. Irradiation tests have shown that natural rubber formulated with an antioxidant (Antiox 4010) is satisfactory for cable insulation at radiation levels to 2 x 10/sup 8/ rads. Four 2-kilogram scale runs were made to study the meltrefining characteristics of high Pu (20%) -- U --flssium alloys. A further investigation was made of the possible Zr contamination of molten U and its Ce alloy resulting from prolonged holding at 1400 deg C in stabilized ZrO/sub 2/ crucibles. Experiments at 1700 deg C showed considerable evolution of CO as a result of reaction of ZrO/sub 2/ with graphite. Reduction of oxide coatings on …

The use of a scalloped cross-sectionsl coolant channel has been suggested as a possible solution of the fuel-element bowing problem inherent in the septafoil type of geometry. Using simplified assumptions, a method has been developed for calculating the red spacing and scallop size necessary to produce equal average fuel-element surface temperatures in the central and peripheral regions of the coolant flow channel at the mid-section of each fuel-red cluster under a given set of reactor flow conditions. Since the extent of rod-bowing is related to the surface temperature distribution, this requirement should minimize fuel-element deflection. ( auth) l6629 In heavy-water-cooled and -moderated power reactors such as NPD-2 and CANDU the coolant will be insulated from the moderator by a gas layer external to the pressure tubes. Therefore, these tubes must be designed to operate at the maximum coolant temperature. However, by placing the insulation inside the pressure tubes they can be kept cold; a thiner wall or a higher reactor operating temperature can be used. and a wider range of pressure tube materials be considered. Two kinds of internal insulstion are considered: the cooled pressure tube in which stagnant heavy water is the insulation, and the cooled stress tube in …

This report presents a reactor design to meet the needs of the basic research program at Argonne National Laboratory. The program requires some irradiations in thermal neutron flux approaching 5 x 10/sup 15/n/ (cm/sup 2/ )(sec) and some beam experiments having over 10/sup 15/ n/(cm/sup 2/) available. Parametric studies indicate that the highest peak thermal flux per unit of reactor power can be obtained in an internal H/sub 2/ reflector having a radius of about 6 cm. Since the design effonts recognized total reactor cost as a mator parameter, an all H/sub 2/O-cooled system was adopted as most expedient. Further, the design and calculations effont was directed toward operating conditions which require little or no research effort. This latter point is a recognition both of the over-all cost plus a desire to meet the needs of research in the shortest possible time. (auth)

Some of the apparatus and techniques being used in order to simulate environmental conditions to test supersonic vehicle components are presented. Some specific tests along with plans for future tests are discussed. (W.L.H.)

In the reaction of silicon halides with graphite to form silicon carbide, thermodynamic conditions were determined for the formation of SiC, free of elemental silicon. The use of these conditions was designed to limit the rate of coating formation by the rate of diffusion of carbon through the coating, and render the operation independent of the vapor-flow factors which normally limit the uniformity of vapor-deposited coatings. Although a wide range of pressure- temperaturecomposition conditions was explored, it was not possible to duplicate the success previously obtained in applying the method to NbC, TaC, TiC, and ZrC coatings. Fundamental differences in the characteristics of the carbides which may account for the difference in behavior are the lack of a range of homogeneity in beta SiC crystal structure, and the fact that SiC undergoes a modification from the cubic beta to a hexagonal form at l900 to 2000 deg C.There remains the prospect of forming a uniform SiC ''sponge'' by the present process which can be subsequently impregnated with metallic silicon to form an oxidation-resistant cpating. (auth) l6200 Preliminary results were obtained on the value that commercially pure Pu (95% Pu/sub 235/ and 5% Pu/subp 240/) has when used as nuclear fuel. …

The dynamic behavior of a fast reactor, when the neutron flux is considered as a function of time, is considered. The kinetics of a fast reactor can be grouped into three distinct areas of interest; the first being the normal operating conditions where all the changes are brought about in a slow manner. and the resulting flux changes being small in comparison with the steady stuff flux. Since the available reactivity and the power density of most large thermal reactors is so small, and the heat capacity is so large, nothing but small deviations from design conditions would occur before the control rods were inserted. Thus reactor kinetics traditionally has meant linear kinetics, which in the mathematical interpretation leads to linearized kinetic equations. The second area is where there is much stronger coupling between reactivity and geometrical changes in the core. A fast reactor has a much higher power density than a thermal reactor and geometrical changes will therefore be more effective on reactivity. A fast reactor needs a greater total amount of U-235 because the fission cross section of U-235 is several hundred times smaller at neutron energies of the order of 0.1 Mev as compared to thermal energies. …

Classical electromagnetic theory indicates that a conducting metallic shield can reduce the magneticradiation loss from a hot plasma undergoing D-D burn to less than 1% or two orders of magnitude. (auth)

Report discussing the use of chemical concentrations of certain elements in uranium deposits of the Colorado Plateau as a guide for the size of the deposits.

Report discussing a study of the distribution, volume, and lithologies of sedimentary rocks of the Colorado Plateau, and of the distribution of elements contained within these rocks.

The following is a semiannual progress report, preliminary in nature, covering work done in the period of December 1, 1958 to May 31, 1959.

A detailed description of fabrication methods used for TREAT fuel elements is presented. Photographs of equipment and assemblies are included along with tables and flowsheets. (J.R.D.)

Abstract: A log-period amplifier which is combined with power supply on a rack-mounted chassis with a 7-in. panel is described.

The latest HAPO Five Year Program review, indicates that significant increases in Pu production from the eight existing Hanford reactors may be achieved. These production increases would be attained by a combination of several methods including increased process water flow rates, reactor coolant outlet temperature, improved time operated efficiency, conversion ratio and induced transient reactivity looses. In order to provide a realistic basis for budgeting to meet these or other increased production goals, it to necessary that a study program be undertaken to determine in general terms the more nearly optimum plant changes required and to evaluate the economic and technical feasibility of achieving future process conditions. The purpose of this document is to present a plan for the execution of the prepared study. Included in the study outline are the basis study consideration, problem assignments and schedules, required manpower estimates, and tentative cost estimates.

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