Organic reactor technology is critically evaluated and areas of research and development work now lacking or inadequate for the successful development of this reactor concept are indicated. The development programs for present organic and heavy water moderated concepts appear generally adequate to reach specific goals. However, the narrow scope of the organic reactor program should be broadened to assure coverage of areas where the application of novel principles might result in marked economic benefits. Further work, principally of a basic nature, is recommended in the fields of chemistry, processing, management, and thermodynamic properties of coolants, in fuel development, and in concept evaluation. (N.W.R.)

Electrical resistivity properties of polycrystalline 39 wt % thorium-- magnesium eutectic are reported for the solid from room temperature to its melting point at 589 deg C and as a liquid from its melting point to 900 deg C. The electrical resistivity of the eutectic at the melting point was 69.5 microhm- centimeters; it decreased to a value of 64.8 microhm-centimeters at 900 C. Tantalum tubing was used to contain the alloy in the molten state. (auth)

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Given multigroup fluxes and adjoint fluxes of any cylindrical R-Z configuration, 2D PERT may compute: the prompt-neutron lifetime; the relative worth of various delayed neutrons; the integrals of capture, fission, etc., of given materials over any given region; local perturbations, i.e., danger coefflcients; and integrated perturbations, i.e., reactivity effect of uniform variation in the cross sections affecting a whole region. 2D PERT is programmed for a 32K IBM-704 using 3 tape units. The code is written in FORTRAN with the exception of two SAP subroutines. Input fluxes and adjoint fluxes are on tapes which may be obtained either directly from CUREM output or manufactured by a special tape-writing routine. Homogeneous cross sections and variations of these cross sections are either read in as input information or are computed by the code from a microscopic-cross-section library and atomic densities given as input. A combination of these methods may be used. (auth)

Measurements of material bucklings for graphite uranium systems are summarized. A comprehensive listing and guide to the original data sources is provided. Complete information on physical and nuclear properties of the lattice and the geometry of the exponential assembly is included, along with some of the auxiliary data taken. The fuel sizes vary from 0.925 to 2.5 in. in diameter for five different fuel geometries. The lattice spacings vary from 4 3/16 to 15 in. Over 300 measurements of material buckling are included. (auth)

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Separate abstracts were prepared for 43 sections of this report. (C.H.)

A second order form of the cylindrical P-3 equations is obtained for the case of an isotropic source. The boundary conditions and numerical method are discussed. Input preparation and operating instructions are included. (auth)

By use of the multisphere neutron spectrometer, the neutrons emitted from a sample of PuF4 were analyzed for average neutron energy. The results indicated an average neutron energy of 1.25 plus or minus 0.25 Mev. The room background had a fast neutron energy of 1.00 plus or minus 0.25 Mev, with a reasonably large contribution from scattered neutrons. A 110 gram sample of PuF/ sub 4/ gave a dose rate reading of about 9 mrem/hr at a distance of 30 cm, which corresponds to a neutron yield of 8.0 x 10/sup 5/ n/sec. The room background was about 0.5 mrem/hr. Neutrons originating in the PuF/sub 4/ processing equipment were measured with the 10 inch sphere neutron survey instrument. Comparable readings were obtained with the converted PeeWee neutron survey instrument by using a correction factor of 30, or a fast neutron energy of 0.5 or 0.4 Mev. Either of these factors enabled a monitor to obtain a more accurate dose rate with the PeeWee than was previously possible. (auth)

References (165) on diffusion in ceramic systems such as oxides, silicates and glasses, borides, and carbides and graphite, are given to books, reports, and U.S. and foreign journals published from 1904 to 1961. Data on the frequency factor, Da, and activation energy, Q, are given for various elements and systems. A separate author index is also included. (P.C.H.)

Strontium recovery from Purex 1WW was investigated with simulated feeds and tracer activities. Initial experiments demonstrated recovery of over 70% of the strontium by sulfate precipitation from partially neutralized 1WW by either increasing the sulfate concentration to about 3 M or by adding carriers such as lead. Precipitation of iron was avoided by addition of one or more moles of tartrate per two moles of iron. Precipitation at elevated temperatures and addition of lead after pH adjustment were shown to be beneficial. Strontium recoveries of over 95% were achieved by precipitation at about 80 deg C at pH values of 0.4 to 4.0 with sulfate concentrations of 0.67 to 3 M and over 0.02 M lead carrier added. High sulfate concentrations were required at low pH, but the sulfate concentration is not critical above pH 1. Some separation of strontium from cerium was observed at pH 2 to 4, with the degree of separation being dependent on both tartrate concentration and pH. Recovery of strontium from the sulfate precipitate by leaching with nitric acid, sodium acetate, or oxalic acid was unsatisfactory because of the large volumes required. Metathesis with sodium hydroxide removed most of the lead as plumbite and converted …

Liquid radioactive wastes are converted into solids, with volume reduction factors of 3 to 8, by flash drying on finely screened, porous, inert solid particles (e.g. alumina) in a fluidized bed at 320 to 550 deg C. The wastes may be either aluminum nitrate-containing wastes from the processing of MTR-type fuel elements, or Purex Process wastes. Ruthenium is found to be the only volatile fission product in this temperature range. Methods are described for its removal from the fluidizing gas. (T.F.H.)

The chlorination of mixtures of uranium and fission product oxides in various molten halide systems by sparging with a chlorine-carbon monoxide mixture was investigated. The chlorination reaction causes the suspended oxide to form species that are soluble in the molten salt and removes some of the fission product elements by volatilization of the chlorides. The rate ot oxide dissolution and the fission product behavior both proved to be dependent upon the composition of the molten halide medium used. (auth)

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A manual is presented for those who use or service the transistorized instruments for nuclear spectroscopy: the transistor amplifier; the snip-snap single-channel analyzer; the fast coincidence unit; and the biased amplifier and linear gate. A general description is given for each instrument along with the specifications, a description of the circuit, and a procedure for initial testing. (auth)

The High Flux Beam Reactor, HFBR, is cooled, moderated, and reflected by heavy water and designed to produce 40 Mw with a total epithermal flux of ~1.6 X 10/sup 15/cm/sup -2/ sec/sup -1/ and a flector thermal maximum flux of 7 X 10/sup 14/ cm/sup -2/ sec/sup -1/, using a core formed by ETR plate-type fuel elements in a close-packed array. The hazards summary is given in terms of site description, reactor design, building design, plant operation, disposal of radioactive wastes and effluents, and safety analysis. (B.O.G.)

The M-1 code is a Monte Carlo code that applies to cylindrical geometry when solving for the flux from a pre specified radiation source. The source is a gamma and beta emitter and the solution is for the flux of each energy group and of each region of interest in regard to the emitter. A region is a volume of the system bounded by two planes perpendicular to the axis of symmetry and two cylinders (one cylinder if the region includes the axis of symmetry). The code can be used to solve for a maximum of 30 energy groups and 280 regions. The M-1 is coded in Fortran for a 32,000-word 7090 and requires that the energy intervals be prespecified as well as a complete description of the geometry of the system. A specification of materials in the system must also be given. The number of particles to be followed must be specified by the user. Since the technique of splitting can be employed here and so that splitting can occur, a description of the manner in which the system is divided (geometrically) must also be given by the user. A detailed description of the input required by the code …

Spherical particles of beryllium oxide containing and enclosing UO/sub 2/ particles (-1O micron) were prepared by dispersing a suspension of UO/sub 2/ in a concentrated viscous solution of a basic beryllium saIt in a liquid organic medium, drying, and firing. The spheres produced were porous and would require densification to make the beryllium oxide protective to the UO/sub 2/. Precipitation of beryllium hydroxide or carbonate on UO/sub 2/ particles suspended in solutions of beryllium salts under various conditions produced no actual coating of the UO/sub 2/ particles. (auth)

Methods and procedures are described for removing a radioactive primary sodium pump or a primary cold trap with handling casks from the Hallam Reactor. (D.L.C.)

The efficiency of a 0.5 in. dia packed column with 1/16 in. nickel helices for the separation of the binary system UF/sub 6/-- BrF/sub 5/ was investigated. Several distlllations were performed wlth the system methylcyclohexane and nheptane for purposes of callbration. For both systems, pressure-drop measurements at various flow rates were determined and the flooding rates were determined from these. Experiments to determlne the equilibrium time were also conducted with both the organic and inorganic systems used for calibration. The separation efficiency was calculated as Htu/sub g/. The flooding rates determined were 570 for methylcyclohexane, 2680 for uranlum hexafluoride, and 2200 lb/(hr) (ft//sup 2/) for bromine pentafluoride. Equilibrium times of 34 and 24 hr were found for the organic and inorganic systems, respectively. The Htu/sub g/ was found to be 1.2 in. for flows of 50 to 450 lb/(hr)(ft//sup 2/) for the organic system. The Htu/sub g/ for the inorganic system was 1.3 in. at flow rates above 285 lb/(hr)(ft/sup 2/) and was found to increase to almost 3 in. at flow rates below this. (P.C.H.)

A safety evaluation of the SNAP 7E thermoelectric generator system is described. Analyses were performed to assess the radiobiological effects in event of a fuel release and the shielding was evaluated to determine the safe working limits for personnel. The entire evaluation is based on a fuel loading of 31,000 curies of radiostrontium. It is concluded that the safety criteria are met and there is reasonable assurance that this generator is safe for its intended mission. (J.R.D.)

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A program is described which is devoted to heat transfer and pressure drop measurements on single tube sections with coolant flow only on the inside. The tests were conducted on simulated PM-1 fuel elements. Data are included and data reduction methods are discussed. (J.R.D.)

The hazards evaluation was modified to reflect certain changes made to the equipment as a result of operating experience. These changes included: the addition of a startup interlock circuit; the modification of a startup interlock circuit; several minor modifications to the control rod actuators; and the addition of the tube-sheet cooling system. (M.C.G.)