This report studies tritium oxides and tritium in water vapor, a product of P-10.

it is desired to know the accuracy and precision of uranium content determination with the γ-photometer. Several possible sources of error (involving both bias and variability) arise when the problem is fully considered. As analysis of data collected for the evaluation of this technique has progressed, there has been a continued discovery of new sources of error and bias further complicating this evaluation. These possible sources have not yet all been systematically examined and thus a complete answer cannot be given at this time. The following is an attempt to state the problem and make recommendations concerning its over-all solution. These recommendations are not assumed to be the totality of those possible; however, they will provide a starting point.

Various methods and procedures for preparing uranium dioxide bodies for metallographic examination and study have been utilized by various investigators. Unfortunately, these methods are not in general satisfactory for use on irradiated uranium dioxide. Since radiation damage studies on irradiated uranium dioxide require use of electron microscopy, and reveal structure by virtue of surface relief. This report covers an investigation of etching methods compatible with radiometallurgy and electron microscopy requirements.

This report discusses the mathematical and statistical questions concerned with the estimation of a leak rate from data collected during a reactor containment vessel pressure test such as that performed on the PRTR vessel in May, 1959. A mathematical method is suggested in Section 3 for the construction of a total number of gas molecules in the containment vessel time series using vessel absolute pressure and temperature readings at several positions within the vessel. A formula for the precision of the series is given in terms of the individual instrument precisions. The question of accuracy and its relationship to the temperature gradient within the vessel is also considered.

The feasibility of removing and recovering cesium-137 from various HAPO process solutions by precipitation of Cs2ZnFe(CN)6 has been demonstrated previously. Pilot plant studies of calcination and steam hydrolysis of non-radioactive Cs2ZnFe(CN)6 precipitates by members of the Process Equipment Development Operation are currently in progress. In support of these pilot plant studies, experiments were performed to determine the extent, if any, to which cesium volatilizes during calcination and hydrolysis of Cs2ZnFe(CN)6 precipitates containing cesium-137. Experimental procedures and results are presented in this report.