Employing a modified arc-melting technique, homogeneous cylinders of stoichiometric UC were consistently produced. The cylinders ranged in size from 0.25-in. dia. by 1 in. long to 0.375-in. dia. by 3 in. long. The top 10% of nearly all cylinders did, however, contain shrinkage voids that could not be eliminated. UC buttons previously prepared by arc melting the elements were placed over a graphite-mold cavity, but in contact with Cu in an Aratmosphere arc furnace. When the UC charge became fully molten, it dropped into the cavity and solidified instantaneously. The measured densities of the arc-cast cylinders were 99.8% of theoretical with compositional limits of plus or minus 30 ppm C. The study was initiated because of a need for pure stoichiometric UC radiation specimens and a need for nonstoichiometric specimens containing controlled C concentrations. Therefore, emphasis was placed on purity control of the starting material and furnace atmosphere, rate of power consumption, accuracy of weighing procedures, and arc-melting procedures. Variations in either of these preparational steps would result in an undesirable pr oduct. (auth)

Operational experience at the SRE and evaluation of alternate methods resulted in the development of an optimized fuel-handling system for the Hallam Nuclear Power Facility Beactor. Increased reliability, easier maintenance, improved safeguards, and broader flexibility are some of the characteristics of this system, which features a manually-operated, shielded cask with reactor atmosphere seals that is usable in a conventional unshielded air-atmosphere building. (auth)

Progress in the development and testing of plutonium fuels is described. Topics covered include: project and facility administration, preparation and characterization of fuel materials, fabrication and evaluation of fuel shapes, fuel element fabrication and evaluation, radiation testing aad evaluation, and reactor physics and engineering parametric studies. (M.C.G.)

Twenty-two papers are included. Separate abstracts were prepared for fifteen of these. The presentations not abstracted concern liquid metal research, evaluation of materials in gaseous and liquid Hg, liquid metal corrosion, of loop materials, effects of K vapor on refrsctory alloys, research program on Na components, Na loop studies, and determination of O in Na. (J.R.D.)

Safety analyses of the launch of SNAP-type units as to launch, ascent, and intermediate or final-stage failures are reported. A preliminary design study to define the configuration and performance of the Surveyor generator is completed. Dispersion patterns are found for radioactive materials released by the high-velocity vertical impact of Surveyor-type generators on the moon. High- velocity testing methods are discussed. Flare and telemetry capsules, designed to simulate fuel cores of nuclear auxiliary power systems in Atlas-type ballistic re-entry trajectories, are in final production stages. The design of a SNAP-3 test generator is completed. Analytical studies are reported on the diffusivity of various particle sizes and on the ablation phenomenon as it pertains to re- entry of radioisotope fuels. A preliminary flare and telemetry capsule test is discussed. Routine coding, assembly, and checkout operations are carried out on radioisotope shielding codes. Arrangements are made to obtain code input information and actual radiation levels from the Sr/sup 90/-fueled weather station generator, as a check problem to verify code calculation results. (T.F.H.)

Investigations are being made to establish the feasibility of using cold extrusion to produce Zircaloy-2 tubular products. Tests on extruding 1 1/4-in. diameter billets were completed. Samples in 46 bar-extrusion tests were extruded with 80% reduction at 400 deg C. Severe lubricant breakdown occurred. Piercing tests were conducted at, room temperature and 200 deg C, using 50% reduction and a 140-degree conical punch profile. Good extrusions were obtained in forraing 1.07-in. OD x 0.052-in. wall tubes with 65% reduction. Low efficiencies in tube- extrusion tests compared with those for bar extrusion were due to the large surface-to-volume ratio of tubular specimens. Tooling was designed and fabricated to produce, in three steps, 12-in. lengths of 0.560-in. OD x 0.030- in. wall tubing from

Isotope power sources are being developed for use in thermionic and thermoelectric power generators for space applications. Suitable radioisotope fuels and heat source containment capsules are being designed for SNAP-type generators with land, sea and space applications. Curium242 and Plutonium-238 are currently being investigated, and methods for purification, compound conversion and fuel form are being established. A Heavy Elements Processing Facility capable of handling and preparing gram quantities of Am/sup 241/ in an aluminum matrix for irradiation to Cm is being activated. In the facility, high integrity aluminum capsules were produced through special encapsulation and welding techniques. Curium purification by ion exchange techniques is under study to determine the operating conditions and design parameters for the Hot Cell Facility. Following a screening process, these container materials were selected for further investigation: Hastelloy C and Inconel X for thermoelectric applications and a Mo-W combination for the curium thermionic generator. Safety, thermal and shielding analyses were conducted on the processes being developed as well as during handling and shipping. (auth)

Work in the processing of Sr/sup 90/ into heat sources for 4 radioisotope-fueled thermoelectric power generation systems is described. The design and engineering analysis of these thermoelectric generators are discussed. Fuel process flow and associated equipment requirements for remote conversion of Sr/sup 90/ feed material to strontium titanate pellets are covered. Previously evolved technical standards concerning raw fuel material specification were coordinated. (M.C.G.)

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Cycle study performed on a closed loop Brayton cycle incorporating a gas-cooled nuclear reactor, regeneration, and compressor intercooling. The purpose of the study was to determine the effect of various cycle parameters on efficiency and cost.

Report issued by the APDA over an evaluation of the toroid sodium resistivity meter. Studies were conducted on carbon and oxygen contamination. As stated in the introduction, "the meter was also subjected to extensive testing in which variations in temperature, flow rate, and supply voltage were made in order to determine signal stability and effective range" (p. 5). This report includes tables, illustrations, and photographs.

The expanding use of Zircaloy-clad Zr-enriched U alloy fuel suggested a need for higher capacity reprocessing facilities. The most ungent requirement is for continuous dissolution equpment to match the solvent extraction capacity already available at Idaho Chemical Processing Plant. Because of the rapid disaolution rates possible with aqueous HF on clean metal as well as on oxidized fuels, and subsequent flowsheet economics, it is the reagent of choice. Laboratory measurements and a survey of the properties of UF/sub 4/ indicated highest solubility in the dissolver effluent was obtained at higher total fluoride concentrations and lower fluoride-to-Zr mole ratios. Since Zr salts were shown to be susceptible to precipitation at higher fluoride and Zr concentrations, an optimum dissolver effluent compositton was selected for a Dilute flowsheet at about 5M total fluoride and a 1.1M Zr. The operation of two bench-scale dissolvers verified the feasibility of a continuous process, and Monel and Carpenter20(Nb) were confirmed as suitable materials of construction for the dissolver. Using the Dilute Flowsheet, U solublilty equivalent to the dissolution of fuels containing 2% U, overall, was achieved. Correlations of dissolver effluent flow rate with effluent composition and with total Zr dissolved are given for guidance in engineering scaleup. …

The following report summarizes an investigation made to determine the extent to which zirconium and zirconium alloys exhibit delayed failure as caused by a combination of absorbed hydrogen and applied stress during the period September 15, 1961, to September 14, 1962.

From abstract: "This report describes and evaluates the variables affecting contamination of precise devices...In the present report these variables are discussed in more detail and the basis for their inclusion in the listing are given. The relative importance of the several parameters will be discussed in another report."

Quarterly report describing the progress and development of improved zirconium alloys for service in superheated water and steam. This report covers the period between July 1 to September 30, 1962 and was conducted by the United States and the European Atomic Energy Community (EURATOM).

Abstract: As a part of continuous geologic and geophysical investigations at the Nevada Test Site, gamma-radioactivity measurements are being made in tunnels and drill holes to determine semiquantitatively the natural gamma radioactivity of geologic formations and the distribution of nuclear-explosion-produced radioisotopes.

The existing pumping capacity and piping intertie arrangement between 181-D and 183-DR makes it necessary to continuously cross-feed treated water from 183-DR flumes to insure adequate reactor flow to 105-DR to meet present production levels. This arrangement means that pumping equipment in both 181-D and 183-D is kept in use to the extent that repair or productive maintenance work must be accomplished during reactor outages, or production levels must be reduced to release a pumping unit during operation.

Progress reports are presented for the following tasks: physical and mechanical metallurgy; plutonium fuels development; and special fabrication of plutonium and other radioactive materials.

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The purpose of Test K-4-13 detailed in this report was to evaluate the behavior of N Reactor fuel elements at conditions equivalent to or more severe than those expected in N Reactor during Phase I) equilibrium operation. Specific objectives of Test K-4-13 were: Measurement of fuel distortion and swelling; Evaluation of irradiation damage to the jacket uranium bond and to the uranium core; and Evaluation of the performance of the fuel element support system.

The objective of this test is to determine the concentrations of radioisotopes in the effluent from a Zircaloy tube and an aluminum tube when both are charged with Zircaloy clad fuel elements and cooled with deionized water. Deionized water is currently being used as the coolant for a Zircaloy and an aluminum tube at KE Reactor. To date the fuel for these tubes has been aluminum clad. Progress is being made in assessing the relative importance of fuel element film and tube film on the radioisotope content of the effluent as well as the contribution of aluminum corrosion to effluent activity. However, there is no information available at present on a channel in which corrosion is reduced to a minimum. It is anticipated that the data supplied by this test will permit a complete evaluation of the variables delineated above.

This report, for September 1962 from the Chemical Processing Department at HAPO, discusses the following; Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.

A continued decrease in flow measured during the periodic high velocity flushes indicated that corrosion buildup in the high tank discharge lines was deceasing backup flow to the reactor. An accurate determination of the high tank flow rate to the reactor was needed to evaluate present back-up adequacy. Riser pressure and flow rate were also measured with throttled high tank discharge lines to establish the low flow and critical portion of the reactor hydraulic characteristic curve. This report presents the numerical results of the test conducted on October 4, 1962.