The concentration gradients of uranyl ion in aqueous and organic solutions were analyzed by taking a macro photograph of the desired gradient by monochromatic (436 m mu ) light transmitted by the solution normal to the gradient in an appropriate diffusion cell. Two Druhm runs were terminated due to malfunction of the sodium metering system and the third run was terminated when the UF/sub 6/ nozzle ruptured. Calculations of particle temperature versus time relations for the flame denitration-calcination method of preparing metallic oxide from nitrate solutions indicate that the times required for heat transfer are controlled by the rate of radiant heat transfer to particle surfaces instead of by conductive heat transfer within the particles. A completed experimental study indicated that electrolysis in a cell with a mercury cathode and a platinum anode is a practical process for removing nickel from HRT fuel solution. The apparent diffusion coefficient of uranium loading on Dowex 21K was shown to be directly related to the resin size. An explosion of sufficient violence to blow apart the Pyrex pipe dissolver occurred during the fifth Darex dissolution of simulated SRE fuel probably from a rapid gas phase reaction between hydrogen and oxidizing gases such as …

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The objective of this test is to demonstrate the feasibility of projection fuel elements for use In existing process tubing and to determine the reduction in rupture rates or hot-spot incidence so achieved. This test is to authorize, (a) charging 20 columns of bumper type fuel elements and 20 columns of control elements per reactor into B, D, DR, F, and R Reactors for irradiation up to 1200 MWD/T exposure, and (b) irradiation of four columns each of enriched (0.947%) bumper and enriched (0.947%) normal type fuel elements until two ruptures are sustained in each (or until one group shows a significant improvement).

Experimentally determine the overall isothermal temperature coefficient of the SER up to the design operating temperatures.

This document provides details of activities of the Hanford Operations Office during the month of September 1959. (FI)

This report describes the design and testing of a 1.17-Bev/c separated K{sup -} beam designed in the fall of 1958 in connection with a 15-in. hydrogen bubble chamber experiment. At the target the K{sup -}/{pi}{sup -} ratio was 1/140. At the chamber, 4.0 K{sup -}-meson decay lengths from the target, and after two stages of electromagnetic separation, the K{sup -}/{pi}{sup -} ratio was 12.5, corresponding to a total pion suppression by a factor of about 10{sup 5}. The K{sup -} flux at the chamber was 0.87 K{sup -} mesons per 10{sup 10} protons impinging on the target.

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An investigation was conducted into the possibility of alloy additions to Zircaloy-2 to diminish hydrogen absorption during aqueous corrosion. The nickel in Zircaloy-2 is believed to be the major constituent responsible for the relatively high hydrogen absorption. Additions of up to 0.5 wt.% antimony, arsenic, bismuth, or tellurium were selected on the basis of their known ability to poison the catalytic effects of nickel in hydrogenation reactions of other systems. Results of tests conducted for a total of 224 days in 600 and 680 deg F water and 750 deg F steam revealed no decrease in hydrogen absorption in modified Zircaloy-2 containing the aforementioned alloy additions. Hydrogen absorption increased when these alloying elements were present in the range of 0.1 to 0.2 wt.%. Corrosion resistance also decreased with alloy additions in these ranges. A 2-atm. partial pressure of hydrogen in the steam or above the water did not affect hydrogen absorption in the alloys appreciably. The hydrogen partial pressure did not affect time to transition in corrosion rates, but did appear to produce higher weight gains than degassed water. (auth)

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

The Expansion Study Task Force has evaluated several cases of Hanford reactor operation at power levels considerably higher than is presently obtained in the six older reactors. These higher power levels result in more rigorous operating conditions of temperature, heat flux, neutron flux, hydraulics, reactor control, etc. The purpose of this document, the various components of which were prepared by Process and Reactor Development Sub-Section personnel, is to assess the technical feasibility of operation under the proposed conditions, and to delineate those specific areas of development effort which may be necessary to provide adequate support for an expansion program.

A performance evaluation was made of a prototype liquid cooled freeze trap with sodium at 350 and 1000 deg F. The sodium freeze-off function was adequate for all test conditions encountered. The freeze-off occurred satisfactorily with the larger clearance provided by a test modification to provide 0.030 eccentricity to the rotating shaft. Turning the freeze-trap handle was successful in opening the unit for gas venting when 350 deg F sodium was used. For a seal formed with 1000 deg F sodium, 16 turns of the trap handle gave no measurable gas venting at pressures up to 30 psi. Melting out the seal opened the vent satisfactorily. All the major problems encountered during the test were mechanical and associated with the rotating feature of the unit. (M.C.G.)

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Neutron multiplication measurements were performed on two identical finite Pu-metal slab assemblies separated and reflected by plexiglas. (auth)

This document presents a nomograph from which the relationship between reactor operating parameters, tube power, and outlet temperature can be correlated with rupture rate. The index indicates the severity of the reactor climate during irradiation and does not include the metal quality parameters defined in the rupture rate equation. The general form of the Operational Severity Index Equation is OSI=P{sup 3.3}/1000{times}t{sub 0}{sup 8.7}/100, where OSI, is the unitless Operational Severity Index, P is the tube power in kW, and t{sub 0} is the tube outlet temperature, in degrees C.

The most adverse power distribution was revised based on a comparison of PDQ calculations and measurements made during the SM-2 flexible experiments. A review of the basic nuclear data and calculational models employed in the SM-2 nuclear analysis was rnade. A comparison between initial reactivily, hot-to-cold reactivity change, and xenon reactivity with experiment was rnade. Based on a revised power distribution, the core flow requirement was reestimated to be 7800 gpm. Tentative designs of the core support and fuel element structure were prepared and evaluated for pressure drop and flow distribu-tion. The ETR and MTR irradiation programs are suramarized. The TIG process for welding elements is discussed. Specimens of Eu/sub 2/O/sub 3/ dispersions in stainless steel were autoclave tested. Static deflection messurements indicated that a fuel element with cold rolled plates will have a deflection aproximately 18% lower than annealed plates. measurement of plate collapse on two elements indicated possible collapse in the range 140 to 164% of rated flow. Flow distribution and pressure drop tests were made for several core support structure configurations. Mockup experiments on the SM-2 initial cold, clean and SM-2 mid-life cores were completed. Limited power distribution and flux distributions were performed in the clean mockup. …

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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 November 1959 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)

BS>Some of the fundamental relationships in a nuclear rocket engine are discussed. The equations required to calculate the performance of the rocket are presented. The problems associated with these calculations are also pointed out. (auth)

The C Reactor was proposed for producing tritium. To evaluate the performance of enriched U-Al J elements and natural Li-Al alloy target (N) elements, 60 charges containing both J and N pieces were irradiated under a variety of conditions in C Reactor. No ruptures were sustained; however Tube 3276-C was discharged because of a suspect. Corrosion rates of J elements were not worse than for natural U irradiated under same conditions. Differences between corrosion of J elements prepared by three different methods were not significant.

In accordance with the NPR technical criteria which require that provisions be made for useful isotope production in the control rods, both isotope-producing and non-production rods have been designed. Design Analysis has been requested by Reactor Plant Design to specify the number and location of the isotope rods for the initial installation. This choice, however, cannot be made without knowledge of the fuel element characteristics and a prediction of the reactor operating techniques. Some of the factors affecting the use of isotope-producing rods are discussed in this letter, and the following recommendation is made.