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)

The effect of the redesign of the NERVA reactors as described herein on the transient startup and decay heat operating characteristics was not determined.

Technique used in deriving values of the linac ball tuner settings to produce various fields is discussed briefly, and tables showing ball tuner change numbers and values used in making up the ball tuner change numbers are given. (D.C.W.)

This report is about Binary counters.

The design and construction of the Boiling Nuclear Superheater (BONUS) Power Station at Punta Higuera on the seacoast at the westernmost tip of Puerto Rico are described. The reactor has an output of 17.5 Mw(e). This report will serve as a source of information for personnel engaged in management, evaluation, and training. (N.W.R.)

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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)

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)

The original objective of the work on coated fuel particles at the Nuclear Materials and Propulsion Operation of the General Company was to prepare coated fuel particles less than twenty microns in diameter and to incorporate them n BeO to yield a fuel element in which fuel-matrix reactions, grain growth, and fission product release would be inhibited up to 1650 degrees Celsius.

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)

Mathematical treatment is given the design for a reflectometer to be used with the Brookhaven alternatinggradient synchrotron Diagrams of various reflectometers are displayed. (D.C.W.)

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.)

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.)

Control rod materials, designs, and fabrication processes were compared for their relative economics. Control rod lifetime data were calculated with a simple approximation of nuclear worth depreciation. These data were used in conjunction with the estimated fabrication costs to determine the cost of using several absorber materials in a typical power reactor on a cost-per-year basis. The effect the control system has on core power density and fuel lifetime was included. (auth)

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)

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)

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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 study of irradiation effects on normal Hanford AlSi-canned I&E fuels was started in March, 1959 under the name of the Quality Certification Program. This report describes the irradiation-induced effects for Warp, Tube Filling Capacity, changes in Outside Diameter, Profile Types and Break Strength, in terms of mathematical models as functions of reactor power, time, and temperature. Graphical representation of the response surfaces generated by these models are included. Future Quality Certification programs are presented.

>The Yiftah, Okrert, and Moldauer cross-section set (ANL Set 135) was modified to include new measurements of nu and alpha for U/sup 235/, and to include, in an approximate manner, Hummel, Rago, and Meneghetti's corrections for flux depression at resonance energies in Al and stainless steel. This modified set, ANL Set 635, was used to compute values of k for 22 ZPR-III assemblies of widely varying composition. The DSN neutron transport code was used in spherical geometry and the S4 approximation; shape factors were used to convert from cylindrical to spherical geometry. Seventeen of the calculated values of k lie within plus or minus 1% of a mean value of 1.003, and the remaining 5 lie within plus or minus 2%. In terms of prediction of critical mass, it appears that the procedure used here can achieve an accuracy of 5% to 10% for a wide range of U/ sup 235/-fueled assemblies. (auth)

An analysis is presented concerning the reactor kinetics of TREAT used as a pulsed, engineering test reactor for fast reactor fuel element studies. A description of the reactor performance is given for a wide range of conditions associated with its use as a test reactor. Supplemental information on meltdown experimentation is included. (J.R.D.)

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Based on the classical theory of diffusion and reaction kinetics, the kinetics of radiation of chemical reactions in fluid systems is represented by a system of parabolic partial differential equations. A code for the solution of a mathematical model which allows up to two radicals and two solutes to be present simultaneously is devised; it gives the concentrations as functions of time and space, and the recombination integrals as functions of time. When the model is applied to water, the radicals may be thought of as H and OH, and the solutes as forming together a redox couple. the reactions are thought to be initiated either by a narrow beam of gamma rays or alpha particles. (auth)