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 three alterations discussed in this report have an effect on the parameters used in Total Control Calculations for C Reactor: Conversion to smaller diameter and shorter vertical safety rods in six outer control channels; installation of forty-four overbored tubes with short fuel charges in the flat zone and an associated increase in spike enrichment; and centralization of the enrichment ring into a three-lattice unit width in place of three and one-half lattice units previously in use. New parameters are derived in this document for future total control calculations.

A number of metals, carbides, nitrides, oxides, borides, and uranium gulfides are briefly evaluated concerning their possible applications as advanced nuclear rocket materials. (R.J.S.)

We present the results of measurements of the angular distribution of fission fragments produced by irradiation of Au{sup 197} and Bi{sup 209} with various heavy ions. The projectiles, B{sup 11}, C{sup 12}, N{sup 14}, and O{sup 16}, had energies from a few MeV above the Coulomb barrier to 10.4 MeV per nucleon. The gross features of these results can be explained by use of a model and parameters that have been used by others to account for angular distributions of fission fragments from helium-ion bombardments. In detail, however, these results appear to indicate that the models used to predict the average value of the angular momentum of the compound nucleus give values too low near the Coulomb barrier. Furthermore, at high bombarding energies it is necessary to consider the fact that appreciable direct interaction is taking place.

Information of interest to analytical chemists is presented in a report containing both positive and negative results obtained in a total of 58,467 determinations. The data and information are presented in sections concernlng the work of the shift laboratory, special analysis laboratory, spectral analytical group, analytical development group, quality control and standards laboratory, and analytical service for EOCR. Details of methods added to the ICPP analytical manual, and to the ICPP analytical radiochemlcal manual are included. (J.R.D.)

Preparation of enriched fuel elements and target (N) elements for a full H-Reactor E-N demonstration loading was authorized in April, 1960. The objective of this loading was to demonstrate the feasibility of operating a Hanford reactor economically as a plutonium-tritium producer to increase over-all conversion ratio and diversify product output. Irradiation of the E-N load began in May, 1961 and was completed in January, 1962.

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 results of Spert I in-pile transient tests of a rodtype, low- enrichment UO/sub 2/ fuel element are presented. The tests were performed to investigate the possibility of damage to such long thermal-time-constant fuel rods when subjected to short-period power excursions, and to test the effectiveness of an instrumentation technique for measurement of UO/sub 2/ fuel temperatures within the rods. In an initial series of power excursion tests, in which the range of reactor periods was from approximately 1 sec to 7.5 msec, simultaneous measurements were made of the transient temperature at the center of the fuel rod and at the outer cladding surface. Fuel rod rupture occurred during the exponential rise of the 7.5-msec excursion. Similar short-period tests performed on a second fuel rod contain ing no internal thermocouples did not result in cladding failure, supporting the postulation that rupture of the first rod was caused by waterlogging of the UO/sub 2/ as a result of the cladding penetrations made for installation of the internal thermocouples. Calculations of the transient temperature distribution in the fuel rod were made, and the results are found to be in good agreement with the experimental data obtained on the central-UO/sub 2/ and cladding-surface …

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In this report, approximations which have been shown to simplify the calculation of the normal modes of paraffin hydrocarbons are tested for olefinic hydrocarbons by application to C2H4, CH2CD, C2D4. The results indicate that the method should suffice for the low-frequency models below about 0.1 ev energy, and may be adequate for all vibrations as far as he needs of neutron scattering are concerned,

Reproducibly high density UC-PuC pellets containing minor amounts of (U,Pu)/sub 2/C/sub 3/ were made with nickel sintering aid. Average densities of 95% and maximum densities of 96.6% of theoretical were obtained. High density, singlephase UC-PuC pellets were made by sintering at high temperatures. Densities of 94% of theoretical appear to be attainable. Improved identification of microstructural constituents was made by means of metallography, x-ray diffraction, and microprobe analysis. Development of specialized chemical analysis techniques continued. The post-irradiation examination of the UC specimens was completed. Mass spectrometer analysis showed that 0.4% of the fission gases was released. Metallographic examination of the UC did not reveal any channges in structure. Metallographic examination of the metal specimen parts that were in contact with UC was completed. No evidence for interaction with niobium and slight evidence for interaction with stainless steel was found. The latter was not deleterious. Perturbation factors were calculated for the UC- PuC experiments. Negotiations continued, in order to select the most suitable test reactor. (auth)

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This report, from the Chemical Processing Department at HAPO, for April 1962 discusses the following: Production operation; Purex and Redox operation; finished products operation; maintenance; financial operations; facilities engineering; research; employee relations; special separation processing; and auxiliaries operation.

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)

Appendices C and D may further be identified as SL1925 and CF-61-12- 20(Rev.), respectively. A comparative report is presented in which the economics and feasibility of plant protection from nuclear attack by plant hardening, remote siting, and utilization of optional fueling concepts for the coal-fired plant are evaluated. (J.R.D.)

The losses of charged particles from a detector aperture, due to multiple scattering of the particles in thin layers of material preceding the detector, are calculated with the aid of a high-speed digitai computer. The rootmean-square multiple-scattering angle is computed after the method of Moliere, and graphs are presented for protons and alpha particles scattering in nitrogen, aluminum, argon, copper, silver, and gold, and for deuterons and tritons scattering in aluminum and copper. The range of particle energy covered is from 5 to 100 Mev. A second program is used to construct a map from which losses, as a function of the root-mean-square multiple-scattering angle and the scattering configuration, may be directly read. The FORTRAN listings for both programs are given. (auth)

The Pebble-Bed Reactor Experiment (PBRE) was designed to advance the pebble-bed concept by providing a test of characteristic features and make contriliutions to the general development of all-ceramic gas-cooled reactors. The following objectives were established for the reactor experiment: to investigate key features of the pebble-bed concept, including on-stream fuel handling, movement of fuel through bed, and performance of core; to obtain operation and maintenance experience with a system contaminated with fission- product activity; and to investigate the behavior of graphite fuel elements. A fourth objective, study of the behavior of core materials at conditions occurring with exit gas temperatures in the range 2000 to 2500 deg F, was tentatively included. The preliminary design oE a 5-Mw(t) reactor for achieving these objectives was prepared. The core of the PBRE is a 2 1/2-ft-diam, 4-ft-tall cylinder containing approximately 12,000 spherical graphite fuel elements 1 1/2 in. in diameter. Fuel spheres are added to and removed from the core by gravity flow, and these operations are performed while the reactor is at power by using pairs of valves for passage of elements into and out of the high-pressure system. Exposed fuel can be recycled to the top of the core. Helium coolant …