To set forth limiting values of internal vapor pressure and effective liquid specific gravity which will permit the maximum utilization of the existing underground waste storage capacity. It is also intended to describe briefly the elements which could contribute to ultimate failure.

It was suggested to the writer by K.A. Eschbach that liquid magnesium might offer certain advantages as a reactor coolant. As a result of this suggestion, a preliminary investigation of the possibilities of this material was made. Definite advantages for a restricted class of applications were found, but a detailed evaluation would seem to require further basic experimental research on the heat transfer, corrosion and flux-mechanical properties of the substance.

This examination was conducted primarily to determine the extent of corrosion existing between the silver bearing alloy and the type 416 stainless steel nut of K reactor units subjected to service conditions and also to establish irrevocably that metallurgical defects exist in the brazes.

Graphite suffers radiation damage when exposed to high energy particles. Many of its physical and chemical properties are changed. One of the effects which is easily measured is an increase in the interlayer or C₀ distance of the crystal lattice. This has been attributed to the displacement of atoms or groups of atoms into interstitial positions of the lattice forcing the planes apart. The C₀ expansion may be annealed out by heating the damaged crystal to high temperatures. This paper deals with the rate of annealing of C₀ damage at several constant temperatures.

The primary objective of this work is to add to the understanding of irradiated graphite. This report describes an experimental technique whereby changes in crystal lattice parameter may be studied as a function of time at constant temperature. The results given here are limited to changes in the C₀ spacing for a typical sample of irradiated graphite in the temperature range 100 to 750 C.

A corrosion testing program has been initiated in Purex and Redox storage tanks to obtain corrosion data on SAE 1020 carbon steel, two alloy-high strength steels, and an alloy steel in neutralized process wastes.

The purpose of this report is to develop an economic selection of conventional reinforced concrete structures of various shapes to establish a design criteria for intermediate level radioactive liquid wastes.