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Work has been progressing on the development of an extrusion process for I&E tubing. The ultimate object of this development project is to compare the extrusion process from a cost standpoint with, the presently used rolling-drilling process. Various experiments have been performed to determine the optimum billet I.D., the proper follower block technique, the proper tooling parameters for the process, and the proper metallurgical structure of the material for optimum results. This extrusion combines many of the past results as a further determination and refinement of the extrusion process to be used in the final semi-production optimization extrusion. The object of this extrusion was to evaluate the resulted of previous extrusion experiments to determine final process conditions prior to an optimization extrusion.

The purpose of this study is to determine if a hazardous condition is created by accidental closing of the coolant loop isolation valves�while the reactor is operating.

The preferred orientation and anisotropy of strain behavior of Zircaloy- 2 were studied as functions of fabrication variables. An inverse-pole-figure technique was used for the preferred orientation determinations. Evaluation of the effects of the fabrication variables on the anisotropy of strain behavior was accomplished by a contractile strainaxial strain analysis. An analysis of strain behavior in the normal direction was developed on the basis of theory of plastic flow of anisotropic metals. A simple intuitively derivable relation was found to exist between the strainstrain analysis and the preferred orientation data. Correlations of the strain-strain data with true-stress-truestrain diagrams and mechanical properties were attempted. The preferred orientation of Zircaloy-2 produced by the Oak Ridge National Laboratory-Homogeneous Reactor Project (ORNL- HRP) metallurgy fabrication schedule (ingot breakdown at 1500 to 1900 deg F, major reduction at 1800 to 1900 deg F or 1350 to 1450 deg F, a heat treatment of 30 min at 1800 at 1550 deg F followed by a water quench or rapid air cool to below 1200 deg F, a final reduction of 25 to 40% at 1000 deg F. and a 3O-min anneal at 1400 to 1425 deg F) was weak compared to that of most of the …

The amount of Cf/sup 252/ and intermediate isotopes which could be made from Cm/sup 244/ in a reactor with high neutron flux was calculated. The variations of different isotopes formed in the bombardment as a function of time are described by a series of linear differential equations. The cross sections used were either weighed measured values or values estimated from the amounts of curium and californium isotopes formed in irradiations of heavy elements. The number of atoms of heavy curium isotopes per initial Cm/sup 244/ atom is given as a function of total flux intervals. The buildup of Bk/sup 249/ and the californium isotopes is also shown. (M.C.G.)

A condenser, suitable for use in the distillation of metals was designed. The temperature of the condensing surface was established by controlling the pressure over boiling NaK-78 contained within the condenser. Performance was evaluated in test units in which pure bismuth was distilled as the test metal. (auth)

The oxidation of uranium(IV) by oxygen and nitrous acid was studied in both 30% TBP-""Ultrasene"" and in aqueous solutions. Emphasis was placed on reactions between uranium(IV) and nitrous acid. Distribution coefficients were measured for uranium(IV) and uranium(VI) in the nitric acid-30% TBP system. The use of uranium(IV) as a reductant for plutonium in the Purex process is discussed. (auth)

A compilation was made of extraction data obtained for tri-n- octylphosphine oxide (TOPO). Diluents, oxidation states of the elements, type and concentration of acid, concentration of the metal and extractant, phase volumes, temperature, phase equilibrium, stripping, decomposition of metal adducts, and determination of extracted elements are discussed. lts extraction characteristics for the elements in each pericdic group are described. (M.C.G.)

The program was written for the Philco-2000 computer to provide a nuclear design computing tool equivalent to the IBM-704 computer program MUFT-4 and to provide a set of routines for a future spatial multigroup program, P/sub 3/ MG-1. In addition to the features of its 704 predecessor, MUFT-5 provides a more complete isotopic edit, an optional blackness coefficient edit, and the use of a complete P/sub i/->i>s library. The resultant program is found to be easier to use because of the simplicity of running several problems, more valuable because of the many additional results that are edited, and more economical because of the changes made in programming. (auth)

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Modifications of EBWR are described, and data from analysis of control- rod operational crud deposit are tabulated. Development and construction progress on Borax V are summarized. Research for ZPR III is reported on a series of uranium oxide critical cores. Comparative data on composition, critical masses, and central fission ratios are tabulated for these assemblies. Developmental work on ZPR-VI, ZPR-IX, and Juggernaut is summarized, followed by a summary of EBR-I operation and development of EBR-II. Studies of reactor steam superheating concepts are reported from which conclusions are listed. In reactor safety studies, data were obtained on the extent of water reaction with various metal combinations for use in cores as function of energy input. In other work in-pile melt down studies were conducted. Operation and development of TREAT are described. Nuclear technological developments are summarized in sections on physics, fuel development, reactor components and materials, separations development, and advanced reactor development. (For preceding period see ANL- 6295.) (J.R.D.)

A hydraulic analysis of the Pathfinder boiler core was completed. A description of flow paths including a schematic diagram is included. Pressure drops along primary and leakage fiow paths were calculatsd. After the resistance of each of the leakage paths was identified and the available driving forces in terms of pressure drops associated with the active fiow were established, the leakage rates were easily determined. The dependence of pressure drops on core operating parameters was investigated. The boiler flow rates were also determined. The natural circulation flow rate as a function of power level was estimated at normal condltions of temperature and pressure. The relationships between the boiler parameters of fiow, power, pressure, and inlet subcooling which arise through the laws of conservation of mass and energy are presented. (M.C.G.)

Machining was completed on the s-tube flow element, and a spring, for use at a constant known temperature, was constructed for the flow element. Calculations for geometry factor, detector efficiency, and effective absorption along the useful radiation path were completed for the reed densitometer. The flow rate control was calibrated and performed satisfactorily. Response to step transient changes occurred with a damping constant of about 0.6 and an effective response time of 10 sec. Sources of error in the s-tube flow rate measurements are discussed. (M.C.G.)

A nonnuclear explosion involving an evaporator occurred in a shielded cell in the Radiochemical Processing Pilot Plant at Oak Ridge National Laboratory on Nov. 20, 1959. Plutonium was released from the processing cell, probably as an aerosol of fine particles of plutonium oxide. It is probable that this evaporator system had accumulated -1100 g of nitric acid-insoluble plutonium in the steam stripper packing; the explosion released an estimated 150 g inside Cell 6, with about 135 g in the evaporator subcell, and about 15 g in the larger main cell. No radioactive material was released from the ventilation stacks; no contamination of grounds and facilities occurred outside of a relatively small area of OaK Ridge National Laboratory immediately adjacent to the explosion. No one was injured by the explosion, and no one received more than 2% of a lifetime body burden of plutonium or an overexposure to sources of ionizing radiation either at the time of the incident or daring subsequent cleanup operations. The explosion is considerdd to be the result of rapid reaction of nitrated organic compounds formed by the inadvertent nitration of about 14 liters of a proprietary decontaminating reagent. In cleanup the contamination was bonded to the …

Albino rats were exposed to a In/sub 2/O/sub 3/ dust aerosol of mean concentration 64 mg/m) for 3 months. The retention of In in the tissues and the mobilization of In from the lungs and tracheobronchlal lymph nodes were determined. The results indicate chemical toxicity. (D.L.C.)

A survey and re-evaluation of previous work on polyurethane and polyurethane--epoxy copolymer systems was initiated to determine which phases are more productive than others. The effects of varying excess polyol content were studied. Work was carried out to define the amounts of butamediol that the system Adiprene L-213-ferric acetyl acetonate can tolerate and still be thoroughly cured. An experimental diallyl phthalate formulation was made and tested. A survey is being made of potential sources of kilogram quantities of Th/ sup 2//sup 3//sup 0/ in the United States. Analyses of various samples indicated concentrations almost three times as great as previously reported. The half-life of Rain was recalculated to include new data, and the results are reported as a function of the resolution time of the gas flow proportional counter. A method is being developed for determining the coincidence correction in porportional alpha counters by following the decay of a short-lived isotope. Additional determinations were made on the density of liquid cerium by the vacuum pycnometer method using tantalum pycnometers of about 0.5 cm/sup 3/ capacity. The viscosity of molten plutonium metal was investigated from 648 to 950 deg C with a cup viscometer. The construction of the hightemperature colorimeter was …

The results of an investigation into the effect of irradiation on the quality and integrity of several commercial siliocnized-silicon carbide (Si-- SiC) coatings for graphite are summarized. Si-- SiC-- coated graphites were oxidationtested at 1000 deg C in air both before and after irradiation. Data show that base graphites for coating must be fairly isotropic and must be stable toward high temperature radiation damage if the coating is to survive in-reactcr conditions. (auth)

Six examples are given of the application of analog computers in the fields of reactor engineering, heat transfer, and dynamics: deceleration of a reactor control rod by dashpot, pressure variations through a packed bed, reactor kinetics over many decades with thermal feedback (simulation of a TREAT transient), vibrating system with two degrees of freedom, temperature distribution in a radiating fin, and temperature distribution in an irfinite slab with variable thermal properties. (D.L.C.)

Reduction of small charges of uranium tetrafluoride with magnesium proved to be successful. By hand-tamping of UF/sub 4/-- Mg blend, tap densities ranglng between3.1 and 3.4 g/cc were obtained. The reduction yields for these densities ranged from 72.79 to 93.71%. In the case of machinecompacted UF4-- Mg blend having tap densities from 3.58 to 3.68 g/cc, reduction yields were higher, ranging between 91.45 and 97.2%. Machine-compacted blends gave much more uniform temperature distribution curves during the preheating period, as a result of higher tap densities. The best yields were obtained by firing a machine- compacted blend containing 5% Mg excess at a furnace temperature of 650 deg C, giving an average crude metal yield of 96.3%. However, the high carbon content of 174 ppm in the crude uranium biscuits obtained by compacts reduction, as a result of hydrocarbon binder presence, appeared to be a disadvantage. Attempts were also made to demonstrate the initiation of the reduction reaction at temperatures lower than 500 deg C by taking x-ray-diffraction patterns of the samples of the UF4-- Mg charges heated up to various temperatures. (auth)

A plugging meter designed for the Hallam Nuclear Power Facility was tested. Plugging performance was satisfactory and excellent reprcducibility was achieved. Although the time limitations precluded complete testing of the automatic control equipment, the tests performed gave satisfactory indication of correct performance. (auth)

The general differential equation of a thermally radiating fin is presented for the situation of a uniform heat source along the base of the fin. An electronic analog solution of a fin of nonuniform thickess with one- dimensional steadystate radiant heat transfer is presented. The specific example is concerned with a constant rate of variation of thickness; however, any general configuration could be used. (auth)

Nuclear reactor space electric power units under development in the U. S. include the SNAP-2, SNAP-8, and SNAP-10. The electric power output of these Systems for Nuclear Auxiliary Power (SNAP) extends over three decades of power, from 300 watts to 6O kilowatts. The major operational, installation, and handling characteristics of these nuclear power units are described. In particular, some limitations and restrictions with regard to payload, shielding, and radiation environment are described with respect to the power plants, their mode of installation, and system weight. The ground handling and safety, as well as the over-all safety, aspects of space reactor utilization are also described. The SNAP-10 power unit is a demonstration system that utilizes thermoelectric power conversion, The SNAP-2 power system utilizes a similar compact nuclear reactor, which is cooled with liquid sodium-potassium alloy, and coupled to a small mercury vapor turbine generator power conversion system. The SNAP-8 system is a direct outgrowth of the SNAP-2 power plant development. It wiil deliver 30,000 watts with one mercury vapor turbine generator system and weighs about 1400 (auth)

The flux of hydrogen gas through Hastelloy B, hot forged to 20% reduction was determined as a function of membrane thickness, pressure differential, and temperature. The experimental data were fitted to the equation J = (1810 P/sup 1/2//X exp (-8422/T) where J is the flux (cc at STP/ (cm/sup 2/- hr)); P is the pressure (atmospheres); X is the membrane thickness (mm); and T is the temperature ( deg K). (auth)

A theoretical treatment is made of the stress-strain and plastic flow properties of spherical shells of stainless steel under dynamic loading. The results are used to calculate the response of a homogeneous reactor core to inential pressure pulses. The importance of the duration of the pressure pulse as compared with a natural time or period of the material in plastic as well as in elastic deformation is noted. For a small homogeneous reactor core vessel, the natural plastic "swing time" is found to be about 0.25 msec. The rise time of the pressure pulse is 2.5 msec while the decay time is approximately 12.5 msec. Thus the rise time of the pressure pulse is about 10 times as large as the ""swing time,'' which means the static pressure approximations hold quite well, and maximum energy absorption is determined for the conditions under which the vessel will fail. (auth)