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This report discusses Hanford`s contribution to river water temperatures and describes the Pasco and Richland Water Treatment Plants.

An IBM 1620 program is described that calculates the Bragg angles for cubic, hexagonal, and tetragonal crystal systems for use with powder x-ray diffraction techniques. Allowance is made for absent reflections in the case of face- and body-centered cubic and close-packed hexagonal materials.

An I.B.M. 1620 programme is described which calculates refined lattice parameters from X-ray diffraction data according to Hess' Method. The programme may be used for cubic, hexagonal, and tetragonal systems and three systematic error functions are included. The output includes the refined parameters their stand errors, and the misfit of the data at each experimental point.

Annual report submitted by the Girl Scouts of the United States of America to Congress describing highlights from 1962, activities, organizational leadership, and other information about scouting programs.

Report that describes the development, design, and operation of a prototype instrument meant to monitor the As76 concentration in reactor cooling water streams.

The cadmium ratio was measured in an exponential mockup of Borax V as a function of the void fraction. The extent of voids, simulated by lengths of closed polyethylene tubes, ranged from 0 to 40%. The corresponding cadmium ratios ranged from 6.1 to 4.6. The exponential was also used to determine the radial flux pattern across a Borax-type fuel assembly and the fine flux detail in and around fuel rods. For a normal loading the maximum-to-average power generation across an assembly was 1.24. (auth)

A small, lightweight, uranium carbide-fueled reactor with a beryllium reflector surrounding the core was mocked up as Assembly 40. It was determined that the presence of beryllium in the axial and radial reflectors did not endanger the safe loading and operation of the' critical assembly. The actual experiment consisted of determination of the critical mass, measurement of the reactivity coefficients for a large number of fissile and nonfissile materials, the performance of radial and axial fission traverses, and measurement of central fission ratios. The effectiveness of the radial beryllium reflectcr as a control mechanism was determined and the Rossialpha was measured. (auth)

Monthly newsletter of the High Plains Underground Water Conservation District No. 1, discussing the field of underground water. Topics include profiles of water conservation research, annual pre-plant soil moisture survey data, annual Winter Water Level measurement data, and information about the latest water conservation tips.

This note describes a simple cryostatic device, which, by utilizing the latent heat of vaporization of liquid nitrogen, enables accurate control of temperature between ambient and - 150 degrees C. The advantages of this cryostat over other models known are simplicity of design and operation, ability to operate accurately from ambient to - 150 degrees C, economy, and large working space.

This bibliography represents the first cumulative edit of the literature examined by the Radiation Shielding Information Center recently established at the Oak Ridge National Laboratory. Its two major sections consist of a list of documents by subject category and of an author index. The document list covers nine major subjects with a total of 118 categories; however, literature specimens have not yet been entered for all categories. If a document applies to more than one category, it is listed under each. The author index includes cross references to the categories under which each author's publications appear. Preceding the bibliography is a list of the subject categories, their corresponding category numbers and the pages on which they appear in the bibliography,

This report addresses the decay heat cooling analysis of a nuclear rocket engine.

A process was developed for the destruction of HNO/sub 3/ in Purex-type waste solutions using sugar. Polyhydroxy compounds such as sucrose, fructose, glucose, and crude syrups such as black strap molasses rapidly and controllably decompose HNO/sub 3/ in aqueous solutions at temperatures above approximately 85 deg C. The number of moles of HNO/sub 3/ destroyed per mole of sugar fed is a function of the temperature of reaction and the moles of dissolved multivalent metal cations (such as iron and chromium). The efficiency of HNO/sub 3/ destruction does not vary with acid concentration over the range of 0.5 to 8 M. However, destruction of acid does vary with iron concentration. Approximately 12 to 20 moles of HNO/sub 3/ are destroyed per mole of sucrose fed as the iron concentration increases from 0 to 1 M. Products of the denitration reaction are all gaseous, consisting of oxides of carbon and nitrogen. The process was successfully tested in the high level hot cells, the cold pilot plant denitration unit, and in the Purex plant. (P.C.H.)

The Fuel Cycle Facility for the Second Experimental Breeder Reactor (EBR- II), the process equipment, and the operations to be conducted in the facility are described. The Fuel Cycle Facility is a plant for reprocessing, by pyrometallurgical methods, the core and blanket material discharged from EBR-II. The reactor core alloy is uranium-5% fissium and contains about 46 wt.% U/sup 235/ . The blanket material consists of U in which plutonium is bred. Core and blanket subassemblies contained in transfer coffins are transferred between EBR- II and the Fuel Cycle Facility, which is in an adjacent building. The Fuel Cycle Facility consists primarily of an argon-atmosphere cell where fuel processing is done, an adjacent air-atmosphere cell where reactor subassemblies are assembled and disassembled, and an operating area (for personnel) which surrounds the two cells. Because of the high levels of activity expected, the fuel-handling-and- processing equipment is designed for remote operation. Remote processing is accomplished with the aid of bridge cranes, electromagnetic bridge manipulators, and master-slave manipulators. Transfer ports and air locks are used in the transfer of materials and equipment into the air-atmosphere cell and between the two cells. The walls between the argon-atmosphere and air-atmosphere cells and the operating …

Report regarding "[t]he Fuel Cycle Facility for the Second Experimental Breeder Reactor (EBR-II), the process equipment, and the operations to be conducted in the facility are described. The Fuel Cycle Facility is a plant for reprocessing, by pyro-metallurgical methods, the core and blanket material discharged from EBR-II. The reactor core alloy is uranium-5 percent fissium and contains about 46 wt.% Uranium-235 . The blanket material consists of uranium in which plutonium is bred. Core and blanket subassemblies contained in transfer coffins are transferred between EBR-II and the Fuel Cycle Facility, which is in an adjacent building." (p. 13)

A series of in-pile experiments was designed for evaluation of a prototype fuel element to operate under approximate ASCR conditions. The fuel surface temperature will be automatically controlled by means of a BF/sub 3/ thermal neutron flux control system. Complete design calculations and results are presented, which should be adequate for presentation of the experiment to the ETR safety committee. (auth)

The United States and the European Atomic Energy Community (Euratom), on May 29, and June 18, 1958, signed an agreement which provides a basis for co-operation in programs for the advancement of the peaceful applications of atomic energy. The work described in this report represents the Joint U.S.-Euratom effort. The over-all development program is designed to obtain the test data and operating experience necessary to eventually realize a 50 percent increase in the output of the Garigliano Nuclear Power Station located at Sessa Aurunca (Campania, Italy). Two tasks are in progress: Task III-F involves the preparation of test specimens of reactor vessel material for irradiation; Task IV consists of the formulation of specification for a complete data logging and computer system.

Thermodynamic considerations indicate that sodium should be favorable for the one-step reduction of UF/sub 6/ to uranium metal. A reaction vessel was developed for the continuous reduction of UF/sub 6/ to metal with batch collection of the products, and several experimental tests established the chemical feasibility of this direct and continuous reduction. Up to 93.5% of the uranium content of UF/sub 6/ continuously reduced by sodium in a reaction vessel was recovered as massive uranium metal in the form of a metal button of acceptable purity. A remaining problem is development of compatible materials of construction. (auth)

The effect of irradiation on the hydrolytic behavior of uranium monocarbide as related to the aqueous chemical processing of these reactor fuels is being investigated. The fabrication, evaluation, and encapsulation of the UC in niobium prior to irradiation is described. Uranium monocarbide buttons were prepared by arc-melting 1.51%enriched uranium metal with the stoichiometric quantity of spectroscopic-grade carbon. The buttons were then remelted and drop- cast into graphite thimbles 3/8 in. dia, and 3 in. long. The cylindrical castings were surface ground and cut into 0.17 in.-thick pellets, which were then polished. Preirradiation evaluation included metallography, chemical analysis, and hydrolysis studies. Microstructures of the final pellets indicated that the monocarbide was nearly stoichiometric; however, some alpha -uranium was present along with carbon contamination from the graphite thimbles. The use of tungsten- tipped electrodes in the arc-melting operation introduced about 0.3% of tungsten impurity. The composition of the pellets, as estimated from hydrolysis studies and chemical analyses was 92% UC, 3.5% UC/sub 2/, 4.0% uranium metal, and 0.3% WC. The UC pellets were encapsulated in niobium and subsequently shrunk-fit into a stainless steel cladding. The capsules were sealed by heliarc welding in a helium atmosphere chamber. (auth)

Three separate irradiation experiments were completed with Th and Th--U alloys. In the first experiment, threerolled plates of Th and Th--5 wt% U alloy irradiated to total atom burnups up to 1.5% at 200 deg C showed no anisotropic growth and decreased in density at a rate of 1% per at.% burnup. In the second experiment, 15 swaged specimens of Th and of the alloys Th--0.1 wt% U, Th--1.4 wt% U, and Th--5.5 wt% U were irradiated to burnups ranging from 0.3 to 3.6% of all atoms at temperatures in the range of 45 to 200 deg C. Again, no anisotropic growth was observed and densities decreased at rates near 1% per at.% burnup. A Th--1.4 wt% U alloy specimen with 2.0 at.% burnup was found to have retained significant room-temperature ductility. In the final experiment, a group of 44 chill-cast specimens of Th alloys containing 10, 15, 20, 25, and 31 wt% U were irradiated to burnups ranging from 0.16 to 10.1% of all atoms. Maximum irradiation temperatures ranged from 260 to over 1000 deg C. Surface roughening occurred in the alloys cortaining 25 and 31 wt% U. Volume increases at any given temperature for all alloys were linear …

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The density and purity of fused UO/sub 2/ from several suppliers was evaluated. Densities of large pieces varied widely, but variations in pycnometer and bulk densities of crushed UO/sub 2/ powder were small. Average oxygen- touranium ratios ranged from 1.94 to 2.14. Impurities visible as microscopic inclusions were U/sub 4/O/sub 9/, metallic uranium, UC, and UN/sub 2/. The chief trace metallic contaminants were aluminum, iron, and silicon. Hydrogen, nitrogen, carbon monoxide, and carbon dioxide were evolved during vacuum extraction. (auth)

The corrections required to deduce unperturbed neutron fluxes from measurements with foil absorbers were determined experimentally for the case of gold foils exposed in a light-water medium. The results are in good agreement with the theoretical predictions of Dalton and Osborn and of Ritchie and Eldridge. (auth)

Quarterly report discussing progress on the Fast Ceramic Reactor Development Program, "an integrated analytical and experimental program directed toward the development of fast reactors employing ceramic fuels, with particular attention to mixed plutonium-uranium oxide" (p. 1).