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)

None

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)

Precipitation, evaporation, and extraction feed preparation conditions are established for the removal of zirconium and fluoride from fuel dissolver product solutions by the addition of sodium formate. A sparingly soluble complex fluozirconate is formed. Ninety-five to 99% of the zirconium and fluoride is separated from the uranium losses of 0.1% or less. Chemical material balances, based on experimental data, were developed for two flowsheets. In one flowsheet, sufficient nitric acid is added to the combined wash solution and filtrate produced during the precipitation step to destroy the formate ion (which inhibits uranium extraction) and to prevent post-precipitation during the evaporation of these solutions. The other flowsheet calls for addition of sufficient nitric acid to destroy the formate ion, but not enough to prevent post- precipitation during the concentration step. Post-precipitation removes additional zirconium and fluoride, but necessitates an additional solids- separation step. (auth)

Completion of zero-power critical experiments with the ORNL Health Physics Research Reactor is reported. A description is given concerning these experiments which were used to determine the critical size, fission-rate distributions, reactivity calibrations of its movable parts, the temperature coefficient of reactivity, and the reactivity effects of the presence of neutron- reflecting materials adjacent to the reactor. (J.R.D.)

None

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)

Calcium-magnesium precipitation apparatus was used to reduce the concentrations of these elements in ORNL tap water, used as a substitute for waste water of low level of radioactivity, prior to strontium removal by foam separation. With and without alkali and flocculator chambers and with a stirred sludge of ratio height to diameter equal to 1/1 to ~4/1, use of 5 x 10/sup -3/ M each of NaOH and Na/sub 2/CO/sub 3/ and 2 ppm Fe/sup 3+/ reduced the dissolved Ca + Mg concentrations to 1 to 2 ppm as calcium. Simultaneously, a strontium DF of 20 to 200 was achieved, and, by adding Grundite clay in the proportion ~0.5 1b/ 1000 gal, a cesium DF of 10 to 40 was achieved. (auth)

None

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 …

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)

A system for generating detailed tables of gamma ray cross section data has been devised for use on the IBM7090 computer. This sy;tem obviates the preparation of large amounts of cross section data. It also provides a scheme for rapid access to these tabulated values. (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)

ABS>Hot cell demonstration of the Zirflex decladding process coupled with a modified Purex solvent extraction process was completed using specimens of Zircaloy-clad UO/sub 2/ irradiated to levels of 6150-14,600 Mwd/TU. Soluble losses of uranium and plutonium to the decladding solutions were about 0.05%. Centrifugation of the decladding solution is probably necessary to remove up to 1% of the UO/sub 2/ present as fines resulting from the fracture of low (93 to 95%) density pellets; high (96%) density pellets produced few fines. Approximately 5 hours were required to dissolve the UO/sub 2/ core material (14,000 Mwd/TU) in 4M HNO/sub 3/ versus 6 to 7 hours for unirradiated pellets to produce a solvent extraction feed of 100 g U/l and 3M HNO/sub 3/. Gamma decontamination factors for uranium in the Purex CU stream and plutonium in the BP stream were increased by factors of 2 to 10 from the normal 1.3 x 10/sup 3/ and 2.1 x 10/sup 3/, respectively, by pretreatment of the solvent extraction feed with dincetyl monoxime or its degradation product, oxalic acid. Preliminary data indicate radiation damage degrades the solvent, 30% TBP diluted with Amsco 125- 82, upon one pass through the mixer-settler banks with feed solutions irradiated …

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 …

Critical studies were made with a simulated, large, dilute power reactor having uranium carbide as fuel. The uranium in the core was 30.7% enriched, and the atomic ratio of uranium to carbon was 0.946. The critical mass was 503.01 kg U/sup 235/ and the critical volume 574.47 liters. Central reactivity coefficients, effective fission crosssection ratios, heterogeneity effects, reactivity worth of distributed materials, foil irradiations, and the average prompt neutron lifetime were measured. Multigroup calculations using the Yiftah, Okrent, and Moldauer crosssection set overestimated k for the critical configuration by 4.7%. (auth)

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 development of apparatus and laboratory techniques for the study of materials compatibility with 1500 to 2200 deg F potassium was attempted. Techniques for corrosion tab preparation, dry box capsule filling and sampling, and vacuum filling and sampling are described. Apparatus for rotating capsule testing to 2000 F, rotating capsule testing to 2400 deg F, and anisothermal see-saw'' capsule testing is also described. (auth)

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)

An addition is presented for the IBM-7090 program MURGATROYD to produce a rough graph of reactor power versus time. A sample of output is included for the case given as an example. (J.R.D.)

Tests were performed to obtain station performance data at various steady-state generator loads. The station 0 was operated for four-hour periods at steady state conditions and levels of 5, 21, 42, and 61 Mw gross generator output. The various readings are presented in tabular form. A list is given of equipment in service during the test. All plant components operated satisfactorily during the test. (M.C.G.)