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.

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 …

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)

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 calculation of shielding the thickness for plutonium is complicated by the many different energies represented in the gamma radiation emitted during decay of the plutonium isotopes. Dose rate predictions are also frequently confused by gamma from varying content of fission product impurities in the plutonium, as well as other gamma radiation induced through alpha and neutron particle absorption within the source material or its environment. After assumptions are made for these many factors the radiation data for shielding determination is still frequently inadequate because of wide variations in dose rates resulting from self-absorption. The degree if self-shielding is in turn dependent on nature of the plutonium compound, degree of compactness, weight, and over-all geometrical distribution of the source material. By preparing a variety of plutonium samples representing combinations of these varying factors, actual dose rates and gamma spectra, as obtained from them, can then be extrapolated for application to specific situations.

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.)

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.

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.)

Continued process improvements in the separation plants provide an incentive for economics in waste storage costs by utilizing existing facilities to their maximum capability consistent with the radiological hazards involved. The major improvements have reduced waste volumes resulting in increased fission product concentration and energy potential. Analyses have been made to determine the effect of this change on the integrity of the existing structures.

The NPR design calls for carbon steel to be the major constituent in the reactor coolant piping system. The piping and its associated fittings will, in all likelihood, be exposed to atmospheric weather conditions during the period of reactor construction. This type of exposure causes rusting. From experience gained during the startup of KER Loop 1 it is expected that there will be initially high NPR coolant activity levels. The high activity during the startup of KER Loop 1 was partially caused by the activation of rust that was eroded from pipe walls. Prevention of rusting on the carbon steel prior to its introduction into the coolant system would reduce the initial activity levels.

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Soil chemistry studies have been carried on at Hanford for a number of years in support of the disposal of low and intermediate level liquid wastes to the ground. Equilibrium type experiments were carried out to investigate the mechanisms of the reactions of Sr, Cs, and rare earth ions with soils (1, 3, 4, 5). Experiments were also conducted with simulated and actual wastes in soil columns and an empirical method was developed for prediction of radionuclide breakthrough from ground disposal facilities into ground water on the basis of short, laboratory soil columns and equilibrium experiments (6,7). In this method it is assumed that the shape and position of the breakthrough curve is the same under laboratory conditions as in the larger scale field case. It is also assumed that the "column volume" or amount of soil effectively used for fission product sorption is that directly under a disposal facility without considering any spreading of waste solutions. These assumptions are known to be on the conservative side, but the degree of conservativeness is not known.

From abstract: "Steady temperatures and thermal stresses are determined for long hexagonal tubes having round holes for the case of uniform internal heat generation and a fluid-cooled inside surface. Thermal stresses are also determined for thick plates pierced by holes on an equilateral triangular pattern and for hexagonal tubes having a hexagonal hole with rounded inside corners. Stresses in the end regions of tubes are also investigated."

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)

A study of metallurgical problems resulting from variation in grain boundaries, bonding layers, and inclusions in metals and alloys has become increasingly important over the past several years. To keep pace with and aid in these studies, several new techniques have been developed in emission and X-ray spectroscopy. In X-ray, areas as smalls as one square micron can be studied by making them targets for a focused electron beam and observing the X-rays emitted (5,2). Such an instrument would be quite helpful at Hanford, but the high original cost is prohibitive.

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Document issued by the Office of the Attorney General of Texas in Austin, Texas, providing an interpretation of Texas law. It provides the opinion of the Texas Attorney General, Will Wilson, regarding a legal question submitted for clarification: Authority of the Livestock Sanitary Commission to quarantine hides from diseased animals to prevent transmission of disease, under Article 1525b, Vernon's Penal Code.

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.

Samples containing 5 to 10 mg of uranium were extracted with hexone (methyl isobutyl ketone) and titrated coulometrically in sulphate media. Relative standard deviations of 0.45% for samples containing 5 mg and 0.56% for 10 mg were determined by precision studies.

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)

The initial and long-term nuclear characteristics of two internally cooled heterogeneous, graphite-moderated, two-region, molten-salt reactors have been studied. These reactors have doubling times of 22.5 years and 27.5 years. Methods of decreasing the doubling times by removing the Pa233 from the core and be increasing the specific power of the reactor are described.

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)

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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