As a part of Project CG-775, 100-K Area Water Plant Expansion, the capacity of the 181-K river pump installations will be increased. Recommendations for the raw water requirements based on a reactor flow of 175,000 gpm were presented in document HW-55877. Since then a new fuel element has been developed for the K reactors, resulting in a lower reactor system curve. It is recommended that the design criteria for the 181-K river pumps be based on a total raw water flow of 213,000 gpm with spare pumping units and a maximum flow of 232,000 gpm for each plant.

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The purpose of this document is to update 190-C pump capacity information previous released in HW-52449{sup 1} and HW-58580{sup 2}. Improvements in motor cooling has resulted in raising the previous 3500 HP limit to 3660 HP{sup 3} thus increasing total pumping capacity.

This annual report (1958) from the Hanford Atomic Products Operation. Various programs are briefly described including irradiation processing, fuels preparation, chemical processing, research and development, and supporting operations.

The recent increase in the incidence of groove pitting on X-8001 clad fuel elements in the old reactors apparently refutes the earlier hypothesis that surface segregation of the secondary phase of this alloy was the primary cause of the unique, preferential attack sustained during irradiation. Components received within the past fifteen months have exhibited essentially none of the segregation. On the other hand, recent evidence suggests that localized penetration of the autoclave film on X-8001 may influence groove attack. The implications of this hypothesis include the necessity of special handling to preserve the autoclave film integrity or possibly elimination of the film altogether. Either certain conditions or properties of the X-8001 alloy or unusual autoclave conditions intermittently produce non-uniform autoclave films. If some of these film conditions are a result of non-uniform alloy structure in the cans, they may contribute to the groove pitting attack. This report presents the design of a test to compare the scratched and non-uniform autoclave films with uniform unscratched controls under special irradiation conditions to compare the incidence of groove pitting.

A core parameter study of the operating costs was performed for a 300- Mwe sodium graphite reactor, a scale-up of the Hallam Power Reactor. The results of the study indicate that the core design is nsar optimum and that core modifications would reduce the power costs by less than 5%. The lattice spacing, fuel rod diameter, and sodium flow can be varied within a rather broad range without significant changes in power generation costs. The effect of the fuel cladning thickness is more significant; fuel cycle costs can be reduced if stainless steel canning is replaced with zirconium canning. Use of UC in place of uraniummolybdenum fuel would also permit cost reductions. (D.L.C.)

A series of critical experiments with blocks of 2% U/sup 235/--enriched UF/sub 4/-C/sub 25/sub 5/H/sub 52/ was initiated at the ORN L Critical Experiments Facility. Thus far assemblies with H:U/sup 235/ atomic ratios of 195 and 294 were built in parallelepipedal and simulated cylindrical geometries, both reflected and unreflected. From the results the minimum critical masses for reflected spheres were determined to be 16.3 and 8.5 kg of U/sup 235/ for fuel mixtures with H:U atomic ratios of 195 and 294, respectively. The minimum critical masses for unreflected spheres of these two fuel mixtures are 24.3 and 12.7 kg of U/sup 235/ respsctively. (auth)

The basic theory of operation of the triggered spark gap is established, and qualitative and quantitative engineering design data are given. From the basic twoelectrode gap, a three-electrode or triggered gap model is established with its static and dynamic triggering characteristics shown. Several geometry conditions such as gap spacings trigger electrode hole sizes and insulator effects are discussed, showing their influence upon the triggering mechanism. A suggested trigger mechanism is given based on that proposed by Sletten and Lewis for the trigatron and modified to fit the present analysis. (auth)

A controlled-temperature irradiation capsule was operated containing small fueled specimens at 160O to 1650 F. The design involved calculating the specimen heat-generation rate and designing an insulating gas gap around the specimens to achieve the desired temperature. Electric heaters were inserted to help control temperature. The thickness and composition of the gas gap were modified prior to operation on the basis of information on probable neutron flux obtained from a nuclear mock-ups and on the basis of information on the thermal resistance of various gas annuli obtained from a thermal mock-up. The desired irradiation temperature of 1625 F was achieved with a variation of sintering time 25 F. (auth)

Thorium oxide is formed by the calcination of thorium oxalate precipitated under carefully controlled conditions. Material is produced with mean particle diameters of 1 to 5 mu . Some of the thorium oxide had uranium added to it by decomposing uranyl carbonate on the thorium oxide followed by calcination. Most of the oxides prepared were calcined to 1000 deg C or more and size classified to remove particles greater than 10 mu . The oxides were prepared in 150-lb batches, with a complete cycle requiring 24 hr. (auth)

By carefully controlling the agitation, temperature, concentrations, and reagent addition rate, it was possible to produce a thorium oxide product within 0.3 mu of any selected size between 2.0 and 3.0 mu . The rate of addition of oxalic acid to a thorium nitrate solution was the variable used in the current series of tests. (auth)

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The physical limitations of existing storage tanks for radioactive wastes from separations plants are defined as a guide for preparing process and operating criteria for the existing tank forms to assure continued integrity of the tanks. A "safe-load" curve for each of the four groups of tanks based on current technology is presented. Loading conditions, operation procedures, and thermal stresses are discussed. (M.C.G.)

This document details activities of the irradiation processing department during the month of September, 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 from separations plants and output of unfabricated Pu exceeded commitments. Purex plant set a new record high for U processed. Production and shipments of UO{sub 3} met schedules. Purex solvent extraction battery performed below normal, probably because of poor solvent quality. NaOH addition to Redox coating removal waste is being reduced. A 3fold improvement in Recuplex product Al impurity was achieved by means of a specific gravity difference > 0.15 between dilute aqueous feed and extractant. Sintered, high-silica crucibles are being tested in RMA production line in Finished Products Operation. Scope design of a fission product shipping cask was completed; powder temperature should be below 440 F for 1 MCi cerium-144 + impurities. Feasibility of using one outside Purex canyon entrance (stairwell opening) for relief damper opening was tested and found to be insufficient. A drawing of the 6-inch continuous centrifuge being evaluated as a vacuum drum filter on RMA button line was reviewed. Casks were designed for the NPR project. (DLC)

The results are presented of a survey of calculated gamma-ray levels at many points on the surface of the operating floor of the containment building for the Enrico Fermi reactor. That portion of the floor surveyed lies directly above the equipment compartment. The calculations were made with the aid of an IBM-650 electronic computer. The main source of radioactivity which gives rise to gamma radiation above the floor is the radioactive sodium-24 in the primary coolant system. This system was considered to be completely filled with sodium, and activated to an equilibrium activity of 0.05 curies/cc, which corresponds to infinite reactor operation at 500 megawatts power. No fission product contamination was considered for these calculations. The operating floor is 5 feet thick and of concrete and steel. The results of the survey indicate that above the equipment compartment the surface dose on the operating floor will in no case exceed 0.9 mr/hr at the expected full operating power of 430 megawatts. Included as appendices are derivations and methods of corrections from one set of concrete and steel thicknesses to another. (auth)

The initial and long-term nuclear characteristics of two internally cooled heteroingeneous graphite-moderated two-region molten-salt reactors were studied. The reactors have doubling times of 22.5 and 27.5 years. Methods of decreasing the doubling times by removing the Pa/sup 233/ from the core and by increasing the specific power of the reactor are described. (auth)

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Modified Zirflex process flowsheets were developed for recovering uranium from the newer power reactor fuel alloys after discharge from the reactor. The STR (1% U97% Zr-2% Sn) and EBWR Core-1 (93.5% U-5% Zr-1.5% Nb clad in Zircaloy-2) fuels are used as examples of low- and high-uranium fuels, respectively. A dissolvent of 6 M NH/sub 4/F yields a solution of zirconium and a precipitate of ammonium uranous fluoride. In one process, ammonium hydroxide is added to produce insoluble hydrous oxides of uranium, zirconium and niobium. The NH/sub 4/F-NH/sub 4/OH supernatant is removed by filtration, partially evaporated, and recycled as dissolvent. The uranium and zirconium oxides are dissolved in nitric acid to yield a solvent extraction feed solution of low fluoride content. In an alternative process nitric acid and aluminum nitrate are added to the ammonium fluoride fuel solution to oxidize U(IV) to soluble V(VI) and prepare a stable solution suitable for solvent extraction. Chromic acid is also added in the case of the STR fuel. In a variation of this flowsheet for the EBWR fuel, only- enough 6 M NH/sub 4/F is added to dissolve the cladding. Nitric acid and aluminum nitrite are then added io dissolve the core. Insoluble niobic …

Samples containing 5 to 10 mg of uranium were extracted with hexone (methyl isobutyl ketone) and titrated coulometrically in sulfate media. Relative standard deviations of 0.43% for samples containing 5 mg and 0.56% for 10 mg were determined by precision studies. (auth)