Pu output from separations plant was less than scheduled, but year-to- date production exceeded commitment by 4%. The Palm recovery run in Purex was the most successful to date. UO{sub 3} production and shipments met schedule. Purex had two pump failures. When Purex 1WW was centrifuged and treated to recover Ce, most of it remained in the centrifugate; only 14% was recovered. The prototype Pu ozonator in Redox performed well. Test runs on an acid precycle flowsheet and a proposed internal recycle scheme for Palm recovery were initiated in Redox. Recuplex had a change in solvent extraction feed preparation, and an installation of a safe-geometry bottom section on the stripping column. Storage of Purex 1WW wastes was discussed in a meeting. Conversion of Rexuplex to a manufacturing facility was completed. Cost estimates were developed for several alternative Palmolive processing schemes. Process flow diagrams were completed for Sulfex decladding of Yankee elements and Zirflex decladding of Dresden elements.

This document details activities of the irradiation processing department during the month of October, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor operations; Facilities Engineering operation; Employee Relations Operation; and Financial operation.

This compilation presents the phase diagrams for possible materials for nuclear reactors developed at the Oak Ridge National Laboratory over the period 1950 to 1959. The systems covered are: metal, metal-fused-salt, fusedsalt, oxide, hydroxide, and aquecus. (W. L. H.)

Laboratory data of flow rate versus header pressure were obtained for various conditions of nozzle cap removal from C, K and BDF single tube mockups of central zone tube assemblies using I&E slug charges. The data are presented. Suggestions are made for applying the data to DR and H reactors. In general, the effect of a pre-inserted support charge on the flow rate is small, especially with the front nozzle cap on. It should be noted that pre-insertion of an entire (117 inch long) support charge and subsequent front cap replacement is impossible in either a BDF tube with 34 fuel elements or a C tube with 32 fuel elements simply from a length standpoint.

Surface active materials in aluminum nitrate-nitric acid fuel reprocessing solutions were characterized. Polymerized silica, zirconium- modified silica and soluble dibutyl phosphate species were found to contribute to stable emulsion formation. These surfactants were reduced in effectiveness by added acid. (auth)

Uranium oxides in a molten eutectic mixture of NaClKCl were chlorinated by bubbling chlorine gas through the mixture. The reaction product, uranyl chloride. was soluble in the molten salt. Although UO/sub 2/ was the most common oxide used, the reaction was similar in the other oxides. Phosgene and aluminum chloride were also used as chlorinating agents. A dense, crystalline precipitate of pure UO/sub 2/ was prepared by the reduction of the uranyl chloride contained in the molten salt solution. The reduction was accomplished by contacting the salt solution with any of several metals, by reaction with hydrogen or dry ammonia gas, or by electrolysis. Several kilograms of UO/sub 2/ were prepared by electrolysis using graphite electrodes. The physical properties of the material made it potentially useful as a ceramic fuel material. The initial high particle density of the "as-produced" UO/sub 2/ was considered of great potential advantage for adapting this process to the refabrication of irradiated UO/sub 2/ into recycle fuel elements. (M.C.G.)

The solid solubility of yttrium in iron-20 to 40 wt.% chromium alloys was determined by metallographic techniques and found to be extremely small. At 1320 deg C, the maximum solubility is about 0.12 wt. a yttrium. Study of iron- rich alloys of the iron-yttrium system shows that the compound YFe/sub 5/ exists. A eutectic occurs between iron and YFe/sub 5/ at 1257 deg C. The constitution of iron - 20 to 40 wt.% chromium-yttrium alloys ccntaining less than 6 wt.% yttrium was studied between 1250 and 600 deg C. It was found that, upon exceeding the solubility limit, YFe/sub 5/ is formed and cccurs in conjunction with alpha iron - chromium. At about 900 deg C and above 615 deg C, if more than 6 wt.% yttrium is present in the iron-35 wt.% chromium alloy and more than 3 wt. % in iron-40 wt.% chromium alloy, YFe/sub 4/ forms to give a three-phase field of alpha plus YFe/sub 5/ plus YFe/sub 4/. The upper yttrium limit of this phase field was not determined. At 815 deg C, sigma phase forms in the iron - chromium system and comes into equilibrium with YFe/sub 5/. (auth)

This document details activities of the Fuels Preparation Department during the month of July 1959. (FI)

Progress is reported on the separation of airborne particles into size fractions for radioactive analysis. Laboratory studies of scavenging systems were conducted using a latex suspension diluted 1 to 500 parts with polystyrene then atomized with a Lauterbach generator. Tests were conducted for the collection of polystyrene particles by an evaporating water droplet. The results from these tests are included. The size distributions of particles obtained from atomizing latex and 1% gelatin suspensions are tabulated. The latex suspensions were diluted 1:500, 1:100, and 1:10. Future laboratory studies are to be directed toward elimination of charged aerosol particles and the use of radiochemical techniques for determining the amount of material collected by a water droplet. (For preceding period see ARF-3127-7.) (B.O.G.)

Centrifugal-casting techniques were investigated as a method of producing hollow cylindrical extrusion billets of aluminum-35 wt.% uranium. Among the variables evaluated were melt temperature, mold and pouring-spout configurations, mold speed, and method of pouring. With the equipment employed it was found that the best castings were produced utilizing a pouring temperature of 2400 ction prod- , a heavy-walled steel cylinder rotating between 700 and 900 rpm for the mold and a bottom-pouring technique employing a retractable pouring spout. Sound, nonporous billets 26 in. long and 5 in. in diameter were produced with a yield after machining of over 75% of the original charge. The major losses occurred in the pouring spout-and-cup asserably. This loss is relatively unaffected by the casting length; and, therefore, castings of greater length than 26 in. should result in even greater recoveries. (auth)

Results are presented of an evaluation of nuclear safety and raaiation control related to the shipment, mechanical processing, and storage of SRE-1 fuel elements. (auth)

Production of Pu, UO{sub 3}, and Pu metal exceeded forecasts. The 2nd attempt at Purex to recover Zr-Nb resulted in about 1/3 recovery, contaminated with about 1% of the Ce. Palm losses to Purex U product were eliminated, and the Pu content was reduced 5 to 10{times}. Routing the dissolver rinses into 3WB concentrator resulted into improved rinsing efficiency. Unclarified feed was processed through Purex HA column. In a test for using B in Redox, the B was routed completely to the waste; it was not detectable in product streams beyond the first cycle. Almost 1000 g Palm was purified and converted to oxide. Ferrous ion catalyzed the reduction of Palm VI by hydrazine or semicarbazide. Coordination of E-metal and NPR reprocessing at Redox in multipurpose dissolver was studied. An interim fission product recovery program at Purex will be directed toward low-efficiency collection of Pm {sup 147}. Locations for critical incident alarms were selected. (DLC)

Radiolytic gas concentration in the HRT as a function of operating power and temperature have been calculated for the conditions of Run 18. The partial pressure of the radiolytic gas was estimated for the various conditions. The maximum total vapor pressure (the maximum summation of the radiolytic gas partial pressures excess oxygen partial pressures and D2O vapor pressure) was found to exist at the core and blanket outlets for all avcrage temperatures greater than 250 C. Below 240 C at some recombination rate constants and power levels, the maximum pressure will be highest at the heat exchanger exit. In the calculations the transfer of radiolytic gas between the core and blanket systems was neglected. The simplification is equivalent to saying that the radiolytic gas transfer between the systems is equal. The results of the calculations show that this is nearly true when 60% of the total power is generated in the core. The effect of pump-up in each system was included. (auth)

A pyroelectric carrier distillation method, using sodium fluoride as a spectroscopic carrier, was developed for the impurity analysis of plutonium metal. (auth)

BS>The thermal characteristics of various reactors fueled with uranium dioxide fuel pellets are listed. (W.L.H.)

This report for February 1959, 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 comparison of possible types of research reactors for the production of transplutonium elements and other isotopes indicates that a flux-trap reactor consisting of a beryllium-reflecteds light-water-cooled annular fuel region surrounding a light-water island provides the required thermal neutron fluxes at minimum cost. The preliminary desigu of such a reactor was carried out on the basis of a parametric study of the effect of dimensions of the island and fuel regions heat removal rates, and fuel loading on the achievable thermal neutmn fluxes in the island and reflector. The results indicate that a 12- to 14-cm- diam. island provides the maximum flux for a given power density. This is in good agreement with the US8R critical experiments. Heat removal calculations indicate that average power densities up to 3.9 Mw/liter are achievable with H/ sub 2/O-cooled, platetype fuel elements if the system is pressurized to 650 psi to prevent surface boiling. On this basis, 100 Mw of heat can be removed from a 14-cm-ID x 36-cm-OD x 30.5-cm-long fuel regions resulting in a thermal neutron flux of 3 x 10/sup 15/ in the island after insertion of 100 g of Cm/sup 244/ or equivalent. The resulting production of Cf/sup 252/ amounts …

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This document details activities of the irradiation processing department during the month of February 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

For presentation at the 5th Nuclear Congress, Clevelands Apr. 7, 1959. Increasing emphasis has been placed recently on the application of solvent extraction to the recovery of uranium and plutonium from spent power reactor fuels. Zircaloy-2 jackets were removed from PWR blankettype fuels by dissolution with the Zirflex Process, and the UO/sub 2/ cores were dissolved in 10 M HNO/sub 3/. Zirflex treatment of prototype samples irradiated to 2500 Mwd/ ton resulted in satisfactory dissolution rates and losses to the dejacketing solution generally less than 0.2% for U and Pu. U-Zr alloy fuels were dissolved in 6 M NH/sub 4/F and adjusted for solvent extraction by the addition of ride was recycled by metathesis and precipitation. Stainless steel jackets were removed from Consolidated Edisontype fuels by dissolution in 6 M H/sub 2/SO/sub 4/ (Sulfex Process), and the ThO/sub 2/-UO/sub 2/ core was dissolved in 13 M HNO/sub 3/0.04 M F-0.04 M Al. Dejacketing losses in unirradiated samples were about 0.02%. Use of the ORNL Reference Darex flowsheet for APPR processing resulted in solvent extraction feed containing 30 ppm chloride. Mechanical equipment was designed to declad SRE fuels and chop and leach techniques are being developed to treat stainless and Zr …

This report for January 1959, 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.

This document details activities of the irradiation processing department during the month of January 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.

Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)

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