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)

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.)

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 …

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.)

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.)

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)

Openation of a 6-in.-dia foam separation column with Sr/sup 89/ tracer and dodecylbenzenesulfonate as a surfactant and foaming agent was continued. The catalytic oxidation of H/sub 2/, CO, and CH/sub 4/ was studied using a nickel- chromepalladium ribbon catalyst. A Mark I prototype fuel assembly was sheared to within 1.5-in. of the end by modifying the gas hydraulic system of the shear. The force required to shear a highly carburized Mark I fuel assembly ductile tubing. Demonstration of the mechanical dejacketing of the SRE Core I fuel burned to approximates 675 Mwd/ton and cooled about 2 years is complete, and decontamination and equipment removal from the segmenting cell is approximately 90% complete. Ten SRE Core I fuel jackets were dissolved in aqua regia and analysis showed negligible U and Pu. A semicontinuous leach run, in which -2 mesh graphite fuel containing 2.6% U was leached in 90% HNO/sub 3/ at 60 deg C, gave only 0.37% U loss. Graphically estimated spectral factors for radiation between tubes within fuel bundles and improved wall radiation factors were rised to calculate the temperature distribution expected within spent fuel elements. Further studies of the dissolution of zirconium oxide by HF in fused salt resulted …

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 …

The time-dependent behavior was investigated of the neutron population in an unreflected unmoderated cylindrical assembly of 90 wt.% U (93.2 wt.% U/sup 235/), 10 wt.% Mo alloy following rapid establishment of a super-prompt critical c ondition with negligible initial neutron population. Reactivity increases up to 11 cents above prompt critical resulted in bursts yielding as many as 1.8 x 10/ sup 17/ fissions with reactor periods as short as 16 mu sec and temperature increases as large as 400 deg C. Pressure waves generated in a portion of the core held in position by an electromagnet for bursts greater than ~6 x 10/sup 16/ fissions initiated the removal of this section about 225 mu sec after the peak burst. (auth)

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)

Data and information are presented concerning analyses of PL-2 transient performance, normal startup and shutdown procedures, and shield design. (J.R.D.)

A number of conceivable reactivity accidents were analyzed, using conservatively pessimistic assumptions and approximations, to permit evaluation of reactor safety. Most of the calculations, which are described in detail, were performed by a digital kinetics program, MURGATROYD. Some analog analyses were also made. None of the accidents which were analyzed lead to catastrophic failure of the reactor, which is the primary consideration. Some internal damage to the reactor from undesirably high temperatures could result from extreme cold- slug accidents, premature criticality during filling, or uncontrolled rod withdrawal. Each of these accidents could happen only by compounded failure of protective devices, and in each case there exists means of effective corrective action independent of the primary protection, so that damage is unlikeIy. The calculated response to arbitrary ramp and step additions of reactivity show that damaging pressures could occur only if the addition is the equivalent of a step of about 1% delta k/k or greater. (auth)

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.)

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.)