Temperatures are given at which two liquid phases form in a synthetic homogeneous reactor fuel solution and its concentrates. The data show a two-liquid-phase boundary temperature of 332°C for the Particular HRT Fuel composition and a flat minimum temperature of 305°C for the initial solution concentrated between 300 and 329°C are presented to indicate solution stability in this temperature region. Some related comments on current HRT operations are given.

The use of mercury as a solvent in the recovery of uranium from spent fuels is of the interest at Oak Ridge National Laboratory. The vapor pressure of mercury is lowered by increased concentration of uranium. By dew-point measurements, the vapor pressure at 175°C was found to very between 2 and 8mm of mercury, and at 375°C, between 300 and 1100 mm of mercury, depending upon composition as described below. Plots of the log of mercury vapor pressure vs. the reciprocal of absolute temperature gave a family of straight lines. Each line corresponded to one of the composition: UHg2, UHg3, UHg4, and a saturated solution of UHg4 in Hg. No Mutual solubility of the intermetallics was indicated.

In the operation of nuclear reactors, nuclear fuel reprocessing plants and in-pile experiments, special provision must be made for disposal of gaseous fission products to prevents contamination of the atmosphere to an unacceptable degree. A disposal process is described in which the noble gas fission products, krypton and xenon, are delayed relative to the sweep gas by physical adsorption as they pass through an adsorbent such as activated charcoal. A theoretical plate analysis, and has been verified experimentally. The retention time for a gas present in trace concentration is proportional to the amount of charcoal in the adsorber bed and to the adsorption coefficient which is evaluated experimentally for a particular combination of materials and conditions. The retention time is inversely proportional to the volume flow rate if the sweep gas.

Several experiments have been performed at ORNL with light water solutions of uranyl nitrate (highly enriched in either U^233 or U^235) in an essentially bare sphere 27 inches in diameter. This report presents the results of several calculations with elementary bare reactor theory and a discussion of the observed discrepancies between the calculated and experimental results. If the observed critical concentration is used in the calculations, the calculated effective multiplication constant is less than unity' thus a higher critical concentration would be predicted than is actually observed.

After two batches (~ 340 kg) of fluoride salt from the ARE were reprocessed, pilot plant operations were terminated because of a leak through which an estimated 780 g of uranium (as UF6) escaped. Of the 21 kg of highly enriched uranium in the feed, 93.12% was collected as UF6 product, 0.13% represented measured losses, and 3.72% was unaccounted for (leak). An additional 3.03% was reclaimed from NaF beds and equipment washes. The product met both chemical purity and activity specifications for product level UF6. Decontamination from fission products was essentially complete. A gross gamma D.F. was apparently limited by the low activity of the feed salt.

The technique of high-frequency titrimetry has been applied to the determination of thorium, uranium, sulfate, and free acid. In Part I of this report, the reproducibility of the method for the titration of standard solutions which contained 50mg of thorium in the absence of interferences is established. The coefficient of variation of the method, under these conditions, was found to be less than one per cent. In Part II, the effect of uranium on the high-frequency titration of thorium, as well as the application of the method to actual samples, is discussed. Uranium in a ratio of 5 to 1 to thorium can be tolerated. When the method is applied to the analysis of representative samples, the coefficient of variation is one per cent.