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.

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.

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

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.

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)

We have previously transmitted interim recommendations of goal exposure plans for enriched I&E material at the C and K reactors and for natural I&E material at the K Reactors. We indicated in the references that corrections for changes in inlet temperature would be forthcoming. The purpose of this letter is to transmit revised goal exposure equations.

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.

Selection of new graphites for nuclear reactor moderator applications can be accomplished either by: (1) an evaluation of the behavior of untested materials processed by proven methods of manufacture when tested under conditions typical of the proposed application, or, (2) a development program exploring new materials, processes and environments which results in graphites meeting or exceeding design requirements. The latter approach is not easily adapted to construction schedules and therefore can at this date contribute little to the 105-N program. The program for selecting 105-N reactor graphite is governed basically by two factors: dimensional stability of graphite under the proposed operating conditions is the major design requirement to be met and the types of petroleum coke and processes used in manufacturing artificial graphites directly influence their behavior under irradiation. Since the evaluation and procurement of graphite for the 100-K reactors, carbon companies have established a number of new coke sources. Some major changes in processing graphite have also been developed; however, these have not reached full production scale and cannot be considered sources for 105-N graphite. Consequently, candidate graphites for the 105-N reactor are those representing the new coke sources plus Texas Lockport coke, the only previously evaluated coke now …

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.

A list of particle-accelerator installations, giving location, type, dimensions, particles accelerated, and energy, is presented. (A.C.)

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)

The need for breeding does not appear to be highly cost for a moderately optimistic expanding nuclear power economy between 1960 and 2000. Since the expansion rate of the US nuclear economy is assumed to be high at least 2/3 of the U-235 recovered from natural uranium is used to supply reactor inventory. It is the remaining 1/3 of the available U-235 that can be saved by breeder breeders or a breeder and converter are the doubling time and a parameter expressing the total fissile inventory per magawatt of electricity. In fact, the need for new raw material in any given year is reduced more by specific power than by changing from a converter to a value of total inventory per magawatt of electricity and the content and value of plutonium or U-233 than on raw material cost. The use of 12% vs. 4% annual lease charge can change the inventory costs more significantly than either the Pu (or U-233)/U-235 value ratio or raw material cost. Net fuel burn costs vary more with the product of net conversion ratio and Pu (or U-233)/U-235 value ratio than with the cost of raw material. (auth)

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)

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)

subscale fiat-plate fuel subassemblies measuring up to 2 by 2 by 20 in., incorporating 16 Zircaloy-2-clad uranium -zirconium fuel plates and 15 coolant channels, were successfully fabricated by the gas-pressurebonding process. Autoclave design, specimen-container configuration, and the design and preparation of specimen components were among the problems studied. Uneven heating of the bonding specimen, caused in part by convective heat transfer, was minimized by developing a special four-element muffie-type resistance heater for the autoclave and by filling void space between the specimen and heater with copper plates and sand. With this arrangement there was only a 10 to 15 deg F variation over a 36-in. zone at 1500 deg F. A type 304 stainless specimen container design, resembling in cross section a swastika, allowed distortion-free bonding of specimens without supplementary jigging. The practicality of building up an entire assembly from simple, rectanrular pieces was demonstrated. This technique makcs it easier to maintain the close dimensional tolerances required for fabrication of flat-plate assemblies and elimirates the expensive machining required to produce one-piece picture frames. Flat Ti-Namel stock inserted between the Zircaloy-2clad fuel plates during assembly of the components was removed after bonding was completed by acid etching to form the coolant …

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.)

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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 …

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.

This report discusses risks associated with the underground storage of high-level radioactive wastes from the Purex Plant at Hanford in 1957. Radiation doses and contamination of the Columbia River resulting from storage tank leaks are discussed.

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.

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.)