Tests were run to determine the burnout conditions for a nonuniformly heated rod in an annular geometry. Two special electrically heated rods were used; the cosine rod, which gave an approximate cosine axial heat flux distribution; and the truncated cosine rod, identical to the cosine, except for one end cut short. The rod to be tested was placed in a circular tube test section to form the annular flow path for the water coolant. Only the rod was heated; the outer surface (tube) was essentially adiabatic. Orientation was vertical, with flow upward. The tests were run at the following conditions: rod OD, C.540 inch; tube ID, 0.875 inch; hydraulic diameter, 0.335 inch; heated length of cosine rod, 108 inches; heated length of truncated cosine, 91 inches; pressure, 1000 psia; flow rate, 0.84 x 10/sup 6/ to 1.40 x 10/sup 6/ lb/hr-ft/sup 2/; and steam quality at exit, 12 to 35%. For each flow and iniet subcooling, the electrical power was increased until burnout was reached, thus establishing a burnout condition. Each nonuniform rod was instrumented with thermocouples in the region of anticipated burnout. The local heat flux and local quality at each position at which a thermocouple indicated a temperature …

Thesis submitted to University of Tennessee, Knoxville. Activities in a program to determine the wetting behavior of pure metals such as Au, Ag, Ge, Al, and Cu on solid Be are reported. Results of similar investigations of binary alloys such as Be--Ti, Be-Zr, and Be--Pd are also included. The contact angles of the molten metals on Be as a function of temperature, exposure time, and atmosphere were measured. The solid-liquid interfacial reactions occurring as a function of test temperature and atmosphere were investigated, and the liquid- vapor and internal surface tensions for those systems in which interfacial reactions did not appear to occur were calculated. (J.R.D.)

Experimental parameters of ZPR-HI fast reactor assemblies were compared with those calculated through use of a cross-section set derived from that of Yiftah, Okrent, and Moldauer, and initially used for a study of the critical sizes of ZPR-III assemblies. This cross-section set waa also used to study central fission ratios in ZPR-III assemblies. The analysis was extended to prompt neutron lifetimes and the relative reactivity effects of the substitution of materials at the reactor centers. The results of the work were examined in conjunction with those of the critical size and fission ratio studies. (auth)

High-efficiency filters are used in air-handling systems as the primary means of minimizing the amount of radioactive particulate matter released to the atmosphere. The filtration efficiency is successfully measured in situ using an aerosol of dioctyl-phthalate and a forward light scattering photometer. Tests are made to determine the linearity of the Naval Research Laboratory type photometer, the precision of filter efficiency determinations, the effect of prolonged exposure to aerosol on filtration efficiency, and the effect of airflow rate on the efficiency of both efficient and defective filters. (auth)

A velocity selector of the Dash-Sommers type with straight slits was constructed for use with the MTR crystal spectrometer. The rotor, designed to remove undesirable higher order contaninants from the Bragg beam, is built of magnesium and fiberglas-epoxy laminate to minimize weight and is driven directly by a 1/8-hp synchronous motor. Vanes of fiberglas laminate are set into 235 slots milled around the periphery of a 4'' diameter magnesium core. High strength is attained by winding the assembled rotor with glass filaments impregnated with epoxy resin. Effective helical pitch is obtained by setting the rotor axis at an angle to the Bragg beam. Delow 0.25 ev, the rotor gave good suppression, but the transmission was low. (auth)

Declassified 26 Nov 1973. Information on Tory-2C is presented concerning reactor performance, neutronics, design parameters, test program, and physical properties of reactor materials. (DCC)

The branching ratio for the beta decay of Tm/sup 170/ was determined to be 0.22 with an uncertainty of 2%, and the internal conversion coefficient of Yb/ sup 170/ was determined by two different methods to be 7.27 and 8.8 plus or minus 0.5, respectively. Preliminary work on resonance integral and reactivity measured is described. (D.C.W.)

The development of separation processes for the transuranium elements, process-equipment development, HFIR target-fabrication development, design of the TRU facility, design of development facilities, corrosion studies, and analytical research and development are reported. (auth)

A remotely operated pilot piant for the engineering development of the sol-gel---vibratory-compaction procedure is being used to prepare about a thousand Zircaloy-2-clad fuel elements for a zero-power criticality experiment. The feasibility, safety, and economy of remote recovery, preparation, fabrication, and inspection of U/sup 233/O/sub 2/-ThO/sub 2/ fuels are being evaluated. The procedure includes solvent extraction recovery of U/sup 233/ by means of di-sec-butyl phenylphosphonate in diethylbenzene; the sol-gel process, a new procedure for preparing dense oxide fuels; oxide particle sizing; vibratory compaction loading of fuel tubes, followed by end-closure welding, cleaning, and inspecting. Gamma-ray scanning will be used to determine the uniformity of the fuel density. Instruments are provided for remote control of the process and for monitoring the radioactivity and evaluating the hazards to workers at each step of the operation. Twelve pilot-plant runs involving uranium decontamination, sol- gel preparation of 50 kg of oxide, and fuel sizing and ciassification were completed, with only minor difficulties. The results of examination after irradiation of 17 test capsules at burnups to 17,000 Mwd/metric ton of fuel showed the sol-gel-prepared oxide to perform satisfactorily with respect to stability and fission gas release. In laboratory experiments, uranium-thorium dicarbide fragments and microspheres were prepared …

The fabricability of dispersion fuels using UO/sub 2/ or UC as the dispersoid and uranium combined with 10 to 15 wt% Mo as the matrix was investigated. Cores containing l7.8 wt% UO/sub 2/ dispersed in U-- 15 wt.% Mo were successfully fabricated to about 80% of theoretical density by cold pressing at 50 tsi, sintering at 1100 deg C, and cold coining at 50 tsi. Comparable results were obtained with UC as the dispersoid. Core fabrication results varied greatly with the type of matrix powder used. Occluded gases, pour density, and surface cleanliness bore important relations to the fabrication behavior of powders. Suitable pressing and sintering results were obtained with prealloyed, calcium-reduced U--Mo powder and with molybdenum and calcium-reduced uranium as elemental powders. Shotted prealloyed powders were difficult to press and sinter, as were elemental and prealloyed powders prepared by hydriding. The cores containing UO/sub 2/ were picture-frame, hot-roll-clad as miniature plates. Molybdenum, Fansteel 82, and Zr--3 wt% Al were investigated as cladding materials. While each bonded well to itself, only the molybdenum-clad core, rolled at 1150 deg C to 10/1 reduction, resulted in dispersions free of ruptures and UO/sub 2/ fragmentation and in strong bonding to the core, evaluated …

Progress is reported on the Molten-Salt Reactor Program. Topics covered include: reactor analysis, component development, design status, metallurgy, radiation effects, chemistry, and fuel processing. (M.C.G.)

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Calculations were made of the powder pattern and of the angles between planes and directions for alpha plutonium in order to provide useful data for interpretation of x-ray patterns. The calculated powder patterns for Cu, Ni, Co, Fe, Cr, and Ti K alpha radiations are tabulated. The angles between planes and the angles between directions are listed, and the standard projections are presented. (auth)

A description is presented of the conceptual design of a high- performance, pulsed reactor called the Power Burst Facility (PBF). This reactor is designed to generate power bursts with initial asymptotic periods as short as 1 msec, producing energy releases large enough to destroy entire fuel subassemblies placed in a capsule or flow loop mounted in the reactor, all without damage to the reactor itself. It will be used primarily to evaluate the consequences and hazards of very rapid destructive accidents in reactors representing the entire range of current nuclear technology as applied to power generation, propulsion, and testing. It will also be used to carry out detailed studies of nondestructive reactivity feedback mechanisms in the shortperiod domain. The facility was designed to be sufficiently flexible to accommodate future cores of even more advanced design. The design for the first reactor core is based upon proven technology; hence, completion of the final design of this core will involve no significant development delays. Construction of the PBF is proposed to begin in September 1984, and is expected to take approximately 20 months to complete. (auth)

From introduction: It is the purpose of this study to present field measurements of dispersion in a river and a canal, whose characteristics are unlike streams in which similar work has been done. The Colorado River and the Pierce Canal were selected for testing.

The objectives of the SNAP 7B program were to design, manufacture, test and deliver a thirty-watt electric generating system for a U. S. Coast Guard Automatic Light Station. This report describes the sixty-watt, strontium-90 thermoelectric generator, the converter, the batteries, and the installation of the generator into the container. The electrical tests of the generator and of the system, the shock and vibration tests, and the tests at the environmental temperature extremes are discussed.

The status of aqueous processing methods for tharium fuels is reviewed. A specially designed 250-ton shear was successfully tested for shearing full-size simulated Consolidated Edison type unirradiated assemblies into 0.25-1.5 in. lengths.

This report addresses the semiannual progress report for the period ending on February 15, 1963.

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An evaluation was undertaken to compare the results of various thicknesses of flat tensile specimens with results of round specimens from the same uranium plate, and to analyze the effect of some of the fabrication and testing variables on the test results. Flat specimens gave good reproducibility, particularly at thicknesses of 0.150 and 0.200 inch, but showed biases when compared to round specimens which varied as a function of thickness. Fabrication variables such as omitting hand polishing produced small increases in the bias and a substantial increase in the scatter of the results. Specimen alignment was found to have a substantial influence on the shape of the curve obtained.

The transfer of uranyl nitrate hexahydrate solution from the, Purex Plant in 200-East Area to the U0{sub 3} Plant in 200-West Area is presently being accomplished with the use of two specially built trailer trucks. This system has been used since the startup of the Purex Plant. Because the production rate of Purex has increased significantly since then, it seems appropriate to ask, ``Is trucking the optimum method under the present and predicted rates of production?`` At the request of Purex Operation, a study has been made of the UNH hauling system, and of some possible alternatives. The purpose of this study was to compare the various advantages and disadvantages of the present and alternate systems, to make recommendations, and to provide a basis for possible management action.

This report addresses the first general analysis of the circular loop as a transmitting antenna, which appears to be that of Hallen.

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This document describes the advantages and limitations of Hanford production facilities. In addition to summarizing the technical parameters of the reactors and separations plants and their mechanical features, the unique aspects of these facilities to the production of special materials in which the Commission may be interested have been discussed. As the primary difference between the B-C-D-DR-F-H reactors and the K reactors and the K reactors is in the number and length of process channels. This report is addressed primarily to the 2000-tube reactors. K reactor characteristics are within the range of lattice and flexibility parameters described.