The cation exchange characteristics of several commercially available zeolites were evaluated, Equilibrium loading characteristics and exchange capacities were determined experimentally in the binary systems cesium-lithium, cesium-- sodium, cesium--potassium, cesium-Tubidium, cesium -- hydrogen, cesium-- ammonium, strontium-- magnesium, strontium-- calcium, strontium-- barium, and strontium-- hydrogen for several zeolites. Exchangeable cation fractions in the equilibrium solutions were varied from 0.2 to 0.00001, while the total normality of the system constituents was held constant. Results were plotted as equivalent fraction of the traced cation on the zeolite at equilibrium vs. the fraction of traced cation remaining in the equilibrium solution. If an influent composition is known, and falls in the system listed above, zeolite column loading may be read directly in univalent-univalent or divalent-divalent systems. However, in the strontium-sodium or strontium-- hydrogen systems, a k value must be used because squared terms appear in the equilibrium expression as follows: 2 Na- zeolite plus or minus Sr-solution in equilibrium 2 Nasolution + Sr-zeolite, or k = (Sr-zeolite)(Na-solution)/ (Na-zeolite)2(Sr-solution). The k for a given cation is a function of zeolite loading with that cation. Several column 50% loading capacities were determined to randomly test the validity of the above curves to predict column capacities, Several columns …

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Two IBM-7090 codes were written to aid the user of the neutron moments method code RENUPAK. One code computes and punches response function input cards for RENUPAK or NIOBE (another neutron transport code). The second code reads RENUPAK flux tapes and prints out a compact edit including dose rate as a function of distance. (auth)

A Monte Carlo program OGRE-P2 was written for the IBM-7090 to solve the problem of gamma radiation from a slab of laminated regions composed of various materials. The dose rate is calculated on one side of the slab. (J.R,D.)

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The theory is presented in detail of an activation method for measuring the ratio of U/sup 238/ absorptions to U/sup 235/ fissions in a nuclear reactor. The origin of the activities is discussed, the equations are given for interpretation of the measurements, and the expected counting rates are computed for a typical experiment in the EGCR. (auth)

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To permit more accurate evaluation of the incentives for isotopes producing control rods a determination was performed of the tritium production which would result from conversion to lithium HCR`s. A cursory examination of the gains resulting from the use of lithium shows that a significant amount of tritium could be produced in this manner totaling 1.6 percent of the present Pu-239 production on an atom equivalent basis.

Submitted herewith are source and special nuclear material composition of ending inventory reports for the month of July 1966. Normal uranium, depleted uranium, plutonium, thorium, U-233, tritium, deuterium, enriched lithium, neptunium 237 and plutonium 238 are all inventoried.

Investigation of several alternative methods of modifying the C- Reactor overbore test facility in order to terminate the production tests has been completed. The results are presented in this report.

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All of the UC-PuC powder required for the natural U irradiation specimens, and some of the powder required for the enriched U irradiation specimens was synthesized. All of the natural uranium UC-PuC pellets required were fabrlcated by pressing and sintering, and are being ground to size. The average density of the (U/sub 0.8/Pu/sub 0.2/)C/sub 0.95/ pellets was 12.54 g/cm/ sup 3/ (92.4% of theoretical) and of the (U/sub 0.08/Pu/sub 0.2/)C/sub 0.95/ plus 0.1 wt% Ni pellets was 13.15 g/cm/sup 3 (96.8% of theoretical). Cored pellets were fabricated by pressing and sintering, and the holes are being reamed to size to provide a place for an accurately located central thermocouple. The initial clad irradiation specimens were assembled, and the assembly of the first irradiation capsule was started. The test reactor choice was narrowed to ETR and GETR. The GETR is recommended for at least four of the six capsules. (auth)

This report details activities of the N-Reactor Department during the month of July 1964.

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Additional work on the synthesis and fabrication of uranium nitride produced an improved product free of oxide contamination as indicated by x-ray analysis. Further work to increase the density of the sintered pellets is needed. A stock of several pounds of stoichiometric uranium monocarbide was prepared by carbon reduction of uranium dioxide. Pellets having bulk densities ranging from 93 to 96% theoretical were obtained by cold pressing and sfntering. Initial experiments on the fabrication of bars, 3 by 1/2 by 1/4 in., by cold pressing and sintering, resulted in sound but somewhat low-density bodies. A few experiments were conducted on the production of uranium monocarbide from ammonium diuranate. The results indicate that considerable addftional work may be necessary to consistently produce a stoichiometric product. The simultaneous synthesis ard hot pressing of uranium monocarbide was funther studied and pellets with balk densities as high as 96.6% theoretical (based on 100% UC) were produced. However, metallographic examination disclosed the presence of some free uranium metal in all pellets. The synthesis of 1-lb batches of U/sub 3/Si/sub 2/ of improved quality was successfully carried out by a nonquench method. Using the U/ sub 3/Si/sub 2/ so produced, sound pellets with bulk densities up …

Flux excursions in the HFIR and the response of its safety system have been examined for the situations arising from voids swept into the flux trap with the cooling water. It was found that by limiting the addition of void reactivity to 00 or less, temperature coefficients alone can handle the excursion without damage to the core. Curves are given which show the safety system response necessary for larger additions of void reactivity. (auth)

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Approximately ten years ago single-tube tests demonstrated the feasibility of using unfiltered river water as a reactor coolant from the standpoint of aluminum corrosion and film formation. However, some effluent activity penalty was indicated. Inasmuch as both current water plant operation and the characteristics of Columbia River water have changed, it was deemed appropriate to reinvestigate the use of partially treated water as a reactor coolant. This report summarizes the results of a half-reactor test carried out at F Reactor.

This report is about the Advantages of Glass Micro-balloons as Filler Material