Practical procedures were demonstrated for partitioning plutonium from uranium in the Purex process with hydrazine-stabilized uranium(IV) nitrate as the reducing reagent for plutonium. Typical partitioning results were 0.04--0.2% loss of plutonium to the uranium product and less than 1 wt% uranium in the plutonium product. With uranium(IV) nitrate as a replacement for ferrous sulfamate, most solids in wastes from the Parex process can be eliminated. No undue hazard is expected from the use of hydrazine as the stabilizing agent. (auth)

Nuclear cross section data for gadolinium, Gd/sup 155/, and Gd/sup 157/ were compiled. Average absorption cross sections were computed assuming Maxwellian thermal spectra for moderator temperatures from 70 to 600 deg F. Poison demand curves were computed for a boiling water reactor with an H/sub 2/O- to-UO/sub 2/ ratio of 2.6, U/sup 235/ enrichments from 1.8 to 3.8 wt%, zirconium or stainless steel as clad, and fuel depletions of 10,000 to 25,000 Mwd/t. Based on the above poison demand curves, poison lumps in the shape of cylinders are being investigated analytically. A calculational method is being developed for detailed depletion studies of cylinders. Gadolinia, alumina, and Al/sub 2/O/sub 3/- Gd/sub 2/O/sub 3/ rods wer e formed and sintered at temperatures to 2700 deg F, in an air atmosphere for two to six hours. Investigations of reaction between Gd/sub 2/O/sub 3/ and Al/sub 2/O/sub 3/ were begun. (auth)

An analytical approach and an IBM machine code were prepared for the design of gas-cooled reactor once-through steam generators for both axial-flow and cross-flow tube matrices. The codes were applied to investigate the effects of steam generator configuration, tube diameter, extended surface, type of cooling gas, steam and gas temperature and pressure conditions, and the pumping power-to-heat removal ratio on the size, weight, and cost of steam generators. The results indicate that the least expensive and most promising unit for high- temperature high-pressure gascooled reactor plants employs axial-gas flow over 0.5-in.dia bare U-tubes arranged with their axes parallel to that of the shell. The proposed design is readily adaptable to the installation of a reheater and is suited to conventional fabrication techniques. Charts are presented to facilitate tlie design of both axial-flow and cross-flow steam generators for gas- cooled reactor applications. (auth)

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Three separate irradiation experiments were completed with Th and Th--U alloys. In the first experiment, threerolled plates of Th and Th--5 wt% U alloy irradiated to total atom burnups up to 1.5% at 200 deg C showed no anisotropic growth and decreased in density at a rate of 1% per at.% burnup. In the second experiment, 15 swaged specimens of Th and of the alloys Th--0.1 wt% U, Th--1.4 wt% U, and Th--5.5 wt% U were irradiated to burnups ranging from 0.3 to 3.6% of all atoms at temperatures in the range of 45 to 200 deg C. Again, no anisotropic growth was observed and densities decreased at rates near 1% per at.% burnup. A Th--1.4 wt% U alloy specimen with 2.0 at.% burnup was found to have retained significant room-temperature ductility. In the final experiment, a group of 44 chill-cast specimens of Th alloys containing 10, 15, 20, 25, and 31 wt% U were irradiated to burnups ranging from 0.16 to 10.1% of all atoms. Maximum irradiation temperatures ranged from 260 to over 1000 deg C. Surface roughening occurred in the alloys cortaining 25 and 31 wt% U. Volume increases at any given temperature for all alloys were linear …

A study was made on the effects of beta treatment variables on the texture, grain size, and distortion in unalloyed dingot and ingot uranium. The effects of delta heat treatment, various cooling regimes, recrystallization, etc., were studied. (D.L.C.)

Using a theory of electron transfers which takes cognizance of reorganization of the medium outside the inner coordination shell and of changes of bond lengths inside it, relations between electrochemical and related chemical rate constants are deduced and compared with the experimental data. A correlation is found, without the use of arbitrary parameters. Effects of weak complexes with added electrolytes are included under specified conditions. The deductions offer a way of coordinating a variety of data in the two fields, internally as well as with each those in another. For example, the rate of oxidation or reduction of a series of related reactants by one reagent is correlated with that of another and with that of the corresponding electrochemical oxidation-reduction reaction, under certain specified conditions. These correlations may also provide a test for distinguishing an electron from an atom transfer mechanism. (auth)

Safety problems that have arisen in the conquest of space and are of a nuclear origin are discussed. Space radiation such as cosmic radiation, geomagnetically trapped corpuscular radiation, and solar flare radiation and radiations inherent in nuclear auxilliary power systems are considered. Problems created by the use of nuclear propulsion systems are emphasized. Launch pad problems and core disposal are discussed (M.C.G.)

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A series of transient tests was performed in Spert I with the Bulk Shielding Reactor II (BSR-II) core to determine the self-limiting properties of this core and to test the response of the reactor control system. The experimental program included measurements of the static parameters, self- limiting transient tests, control-system-shutdown transient tests, low-level start-up tests, and determinations of the reduced prompt neutron lifetime by dynamic and static techniques. The kinetic behavior of the BSR-II self-limiting transient tests showed a general similarity to behavior of self-limiting transient tests performed on other plate-type cores tested at Spert. The control system was found to be more effective in terminating power bursts than was the inherent self-shutdown of the reactor for excursions initiated by reactivity additions of up to an excess reactivity of about 1.8 dollars, corresponding to a reactor period of about 3 msec. For reactivity insertions larger than 1.8 dollars, self-shutdown was more effective than shutdown by external controls in limiting the excursion. In the short-period region where limited or total core damage can occur, no increased protection to the reactor is provided by the control system. (auth)

One capsule containing BeO-30 vol% UO/sub 2/ fuel pellets was irradiated to a uranium burnup of approximately 30,000 Mwd/t. The results of a microscopic examination and physical measurements made during the hot-cell examination of the capsule were described. (auth)

In studies of the effect of alloying elements on the rate of dissolution of Al in mercury-catalyzed nitric acid, it was observed that Co, Ni, and Si present at concentrations of 1 to 2% have significant passivating effects. Fe was slightly catalytic. Passivation due to Si was partly overcome by contacting the passive alloy with active, high-purity Al. Increased catalyst concentration improved the rates when Ni and Si were present. Neither approach was effective when Cu was the passivating element. For application to nuclear fuel reprocessing, it is suggested that for minimum reprocessing costs the potentially passive Al alloys not be used in fuel elements or that, failing this, electrochemical activation techniques be applied at the processing plant. (auth)

The cadmium ratio was measured in an exponential mockup of Borax V as a function of the void fraction. The extent of voids, simulated by lengths of closed polyethylene tubes, ranged from 0 to 40%. The corresponding cadmium ratios ranged from 6.1 to 4.6. The exponential was also used to determine the radial flux pattern across a Borax-type fuel assembly and the fine flux detail in and around fuel rods. For a normal loading the maximum-to-average power generation across an assembly was 1.24. (auth)

ABS>The purpose of the Fast Reactor Test (FARET) program is reviewed. Studies of engineering performance and reactor dynamics and safety are summarized. Design considerations are discussed, including basic core plan, core parameter operating data, instrumentation, coolant, and blanket. Experiments are being planned to study reactivity effects of fuel temperatures, physics irradiation, engineering irradiation, and coolant chemistry. Development of cladding materials, metallic fuel, ceramic fuel, and fuel for reactivity test is described. PERT scheduling is discussed. (M.C.G.)

For purposes of analysis and experiment the centrifugal type downflow separator was divided into the inlet nozzle, separating zone, and outlet nozzle. The analysis and experiments have resulted in a new outlet design, a method of determining separating length, and a more effective inlet nozzle. The results have caused a reduction in pressure loss from 5 ft of water for the reference design to 1.5 ft of water for the new design at a flow rate of 1400 gpm. A reactor core riser and downcomer region was mocked-up in the large air-water tank. Void fraction in the downcomer region was measured as a function of water velocity, water temperature, inlet gas flow rate, and riser geometry. Results show that the void fraction in the downcomer is essentially zero until a threshold downcomer velocity is reached. The void fraction then rises rapidly with increasing water velocity to approximately l1% and then appears to remain constant. Test data from this experiment are being correlated using a dimensional analysis technique. An initial prediction equation was developed. (auth)

A small, lightweight, uranium carbide-fueled reactor with a beryllium reflector surrounding the core was mocked up as Assembly 40. It was determined that the presence of beryllium in the axial and radial reflectors did not endanger the safe loading and operation of the' critical assembly. The actual experiment consisted of determination of the critical mass, measurement of the reactivity coefficients for a large number of fissile and nonfissile materials, the performance of radial and axial fission traverses, and measurement of central fission ratios. The effectiveness of the radial beryllium reflectcr as a control mechanism was determined and the Rossialpha was measured. (auth)

On the basis of 4800 hr exposure to 680 l F water, a number of ternary compositions were shown to have corrosion resistance superior to Zircaloy-2. In addition, the strength and hydrogen pickup properties of the alloys were generally improved over Zircaloy-2. The promising alloys were based on the binary materials Zr- 1Sb, Zr- 1Cr, Zr0.5Nb, and Zr-0.5Sn to which small percentages of Te, Ge, Cr, or Fe were added. At present, 680 l F water corrosion data for an exposure time of 5000 hr are available for the modified ternary alloys. Development of materials for potential service in 750 and 900 l F steam proceeded in a manner similar to that for 680 l F water application. On the basis of corrosion resistance and strength, the alloys Zr3Cr- 1Fe, Zr-3Cr- 0.25Te, and Zr- 1V- 1Fe were considered highly promising and initially acceptable. However, hydrogen pickup properties, which were about the same as Zircaloy-2, were judged as unacceptable. In an attempt to improve this characteristic as well as to further enhance corrosion resistance and strength, an additional series of ternary alloys was prepared. The compositions studied in 900 l F steam did not exhibit satisfactory corrosion resistance. (P.C.H.)

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This report addresses the chemical characterization of selected core samples from the vicinity of Tatum Salt Dome, Lamar County, Mississippi.

This document reports on: Kiwi-B(1B, 2A-CF, 4A, 4A-CF, 4B-CF, 4D, 5, 6A), NRDS activities, graphite fuel element development, metal fuel element development, materials and fabrication, and nuclear flight safety. (DLC)

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Estimates are made relating radionuclide movement to ground water velocity as part of the safety program for a proposed experiment to detonate nuclear devices within the Tatum salt dome. The estimates are based on distribution coefficients obtained from laboratory studies. Core samples obtained from hydrologic test well HT-3, Tatum salt dome, Lamar County, Mississippi, were equilibrated with radionuclides in solutions simulating aquifer waters found in the area. The combinations of Cenozoic sand and silty clay, and quality of water of the area were studied and summarized. The distribution coefficients obtained for different radionuclides were tested and indicate retardation factors from 1.3 to 857 for the travel time of these radionuclides when compared to the travel time of water in the aquifer system. Laboratory results indicate that migration of any radioisotope inadvertently introduced to the aquifers in the vicinity of the dome as a result of proposed nuclear test explosions would be extremely slow. Revised estimates of the rate of dissolved radioisotope movement will be made on the basis of further laboratory studies utilizing chromatographic adsorption columns of 0.5 to 4.0 feet in length.

This report contains documented citations of articles and books about San Francisco's surroundings.

On September 13 and 14, 1962, 8.3 curies of iodine-131 were emitted from the Hanford Redox Plant at a rate of from 0.35 to 0.65 curies per hour for a period of approximately 18 hours. During the emission, the plume trajectories were plotted from meteorological data and samples were collected across the predicted plume trajectories at several altitudes and at distances up to 50 miles from the Plant. Filter, charcoal trap, and air samples were collected during the emission using aircraft; and caustic scrubber, vegetation, and milk samples were collected during and after the emission at selected ground stations in the usual manner. Appropriate background samples were collected before the test began. The data and conclusions are given in this report.