A description is given of the process and equipment for the continuous conversion of plutonium nitrate to tetrafluoride powder for reduction to Pu metal. The conversion involves oxalate precipitation, filtration, calcination, and hydrofluorination. Typical results are described; the product is a free-flowing PuF/sub 4/ with a high degree of separation from impurities. (D.L.C.)

The information presented is a collection of most of the physics information available for the New Production Reactor. The details of some of the physics information, particularly those dealing with exposure and temperature effects, are by no means to be considered the final word since there has been no experimental verification of these effects. However, the gross physics characteristics described are felt to be reasonable representations of the expected physics behavior of N Reactor and should serve as useful guides throughout the startup planning and initial operation.

A type of precision measuring instrument, a flying spot digitizer, for the analysis of bubble chamber data is described. When the flying spot encounters a bubble image, it is attenuated typically 25 to 75% depending on the directness of hit and the bubble chamber photographic and operating parameters. The width of the bell-shaped attenuation curve is approximates the sum of the spot and bubble image diameters, or 40 to 50 mu . The high precision of the FSD derives from the fact that it is possible to find the center of area of the attenuation curve to a standard deviation of 3 to 4 mu out of the much larger full-width at the base. Each flying spot coordinate is of a precision approximates equal to that given by a measuring projector, but typically 16 flying spot points are obtained to one measuring-projector point, so that the effective error is only 1/4 as large. The digital electronics problem associated with a flying spot digitizer is basically the trapping of the grating count at the time a bubble image center is found, the storage of the number temporarily until the computer is free, and then the transmission of the coordinate to the …

A road test, performed on a SNAP 10A mass mockup system, indicated that shock and vibration inputs under actual conditions were not excessive. During the course of travel from the Santa Susana field laboratory to Edwards Air Force Base, and during field trials at Santa Susana, the test system was subjected to only two accelerations exceeding 2 g, and no inputs exceeded 2.4 g. It was concluded that shock and vibration to SNAP systems from highway transporation can be adequately controlled by: selecting the appropriate vehicie, providing proper packaging; specifying route and speed limits for various driving conditions; using ordinary care in loading and unloading; and including suitable shock monitoring instruments with the shipment, (P.C.H.)

The N-Reactor will be charged with tubular fuel elements consisting of two components. Both components, an outer tube with a concentric inner tube, have a uranium core, coextruded with a Zircaloy-2 cladding. The testing in support of N-Reactor fuels has been conducted in the experimental loops in the K-East Reactor. Although the process tubes in the KER loops are smaller than the proposed N-Reactor process tubes, the test loops operate at the water temperatures and pressures that will exist in the N-Reactor. To obtain the desired heat fluxes and core temperatures in KER loop test elements, enriched uranium fuel was used. The cladding surface temperatures and fuel temperatures were near anticipated N-Reactor values. Test conditions, results, and conclusions are presented.

The basic heat transfer and flow equations applicable to studies of flow oscillations in NERVA engines are first linearized, then expressed as difference-differential approximations, and finally written in a form suitable for analog computer solution. Block diagrams are presented which illustrate on analog procedure for solving the linearized, difference-differential equations.

A system of six codes is described that permits the solution of a broad class of problems encountered in the study of neutron thermalization, average cross sections, and spectral models. These programs are all compatible with use on MONITOR for the IBM 7090. The Fortran listings are given. (auth)

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A transfer function for d-c operational amplifiers with generalized input and feed-back networks is derived by application of network theory. This transfer function, more general than others in common use, applies to three and four-terminal networks. It contains, as a special case, the well-known transfer function for d-c operational amplifiers with two-terminal networks. Examples are given. (auth)

During the 18-year history of Hanford operations, a high degree of safety has been provided at reasonable cost by the storage of high-activity liquid wastes in underground tanks, by the percolation of low-activity liquid wastes (principally water) through the soil to the ground water, and by the adsorption or filtration of radioactive materials from gaseous effluents prior to their discharge to the atmosphere. Studies of the Hanford area indicate that current practices could be continued almost indefinitely without jeopardizing environmental safety, provided the stored liquid wastes are periodically transferred to new tanks before the existing ones fail. In 1960, a Fission Product Recovery Program was prepared outlining a plan to meet the Atomic Energy Commission`s needs for isolated fission products. Emphasis was placed on developing processes for isolating selected fission products and on providing interim production capability for recovering the fission products of immediate interest. In 1961, development efforts were channeled almost exclusively toward meeting the immediate needs of the Commission for separated strontium-90 and cesium-137. During this time, it became increasingly evident that substantial economies could be realized, both in immediate development and subsequent operating efforts, if the separate fission product recovery and waste management programs were merged. The …

The new sector-focused cyclotrons have more energetic and better focused beams at the extraction radius than do ordinary cyclotrons. While the first characteristic requires a more intense electric field, the second permits this field to extend over a smaller volume. By tailoring the electrode geometry to these characteristics of the beam, the required deflector gap and electrode surface become smaller, and a higher gradient can be held without sparking. A different compromise between radioactivity, power dissipation, and resistance to spark damage must be made in selecting electrode materials. Carefully designed electrostatic deflectors perform very well in sector-focused cyclotrons of intermediate energy. Deflector efficiencies of about 50% and external-beam intensities of 20 {micro}A have been obtained in the Lawrence Laboratory's 88 -Inch Cyclotron.

To evaluate the crudding characteristics of ammonium hydroxide conditioned coolant it is necessary to use a thermocouple element. The present design has accentuated structural integrity, reliability and constant heat flux. A radical departure from past designs was made by selecting a stainless steel cladding and enriched ceramic fuel. Three thermocouples are inserted in the cladding to increase the reliability. By the use of two thermocouple elements it will be possible to obtain important preliminary information on preferential crud deposition while simultaneously evaluating the crudding characteristics of ammonium hydroxide conditioned coolant.

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At the request of Lawrence Radiation Laboratory, four plutonium strips 0.050 inches thick, 0-50 inches in width and 6 inches in length have been examined by metallographic, chemical, X-ray, density and dilatometric methods. The data obtained will help calibrate high energy electrical discharge equipment used to rapidly heat plutonium.

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A description of the digital readout and drive portion of the Advanced Reactivity Measurement Facility II is presented along with the philosophy used in selecting the system block diagram and components, accuracy considerations, and a preliminary evaluation of performance and usability. The presentation discourages duplicating of the equipment; advantage should be taken of the opportunity to improve upon it. To this end guideposts are provided as well as a documentation of those aspects of the design that are considered worthy of duplication. (auth)

The main purpose of this report are to provide 1) a reference source for the interchange of analog engine model information with AGC, and 2) a reference point for future WANL analog model development. Being a model status report, the present state of the models are presents.

Two long-term test cells were fabricated and tested. One was loaded with a 30 UC--70 ZrC sample and the other with a 90 UC--10 ZrC samrple. The rates of vaporization of these samaples in vacuum were determined both in the absence and in the presence of a Ni anode maintained at 983 deg K with an electrode spacirig of 20 mils, hr order to set the baseline values. Measurement of the rates of vaporization of these samples under these conditions in the presence of Cs vapor at a pressure of about 1.8 torr are in preparation and will be continued for 1000 hr. Operation of the first test cell, containing the 90 UC- -10 ZrC emitter, was terminated after 510 hr because of leakage through a faulty metal gasket sealing the electron-gun compartment. The second test cell, containing a vapor-deposited W-clad UC emitter, was assembled and operated for a period of 111 hr. The test was terminated because a leak developed between the cell compartment and the electron-gun compartment. A control cell containing a solid W emitter is being assembled to check the modifications made in the design of the electron-gun channber and the filament in order to overcome such …

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The status of radioactive contamination of the Clinch and Tennessee River systems from Nov. 1961 to April 1962 is reviewed. Data are included from studies on the fate of radioactive materials discharged to the Clinch River by the Oak Ridge National Laboratory, the mechanisms of dispersion of radionuclides released to the river, the direct and indirect hazards of waste disposal practices during the period, an evaluation of the over-all usefulness of this river for radioactive disposal purposes, and an evaluation of long-term monitoring procedures. The management of liquid wastes at ORNL is discussed and results are reported from studies on the accumulation and movement of radionuclides in White Oak Creek basin, the contamination of river biota, hydrologic measurements and analyses, measurements of radioactivity in the river system and in community water systems downstream from the Clinch River, and calculation of estimated radiation dosages from drinking Clinch River and Tennessee River water or immersion in the water at various downstream points. Results are included from a preimpoundment study of ecological conditions of Melton Hill Lake and estimates of the effects of Melton Hill Lake and power releases from Melton Hill Dam on hydrologic conditions in the river system. (C.H.)