Fourth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).

Abstract: This technical report covers the investigation of 17-4 PH steel components used in the SM-1, SM-1A and PM-2A nuclear power plants, which consist of control rod drive racks, pinions, seal shafts and associated parts. Evidence supporting the continued use of 17-4 PH steel in these applications is presented. Alco's current procurement and processing specifications for 17-4 PH steel are included. The evidence presented is in the form of stress calculations, test and environmental data and the results of non-destructive and destructive metallographic analyses.

Abstract: This report presents metallurgical .analyses and corrosion tests conducted to determine :causes of failure of two SM- 1 reactor vessel studs due to fracture in the threaded areas, after 23 months' operation (March 1959). Conclusions of corrosion tests are: (1) failure of the studs was caused by stress corrosion cracking, (2) contributing factors to failure of the studs were improper control of heat treatment of the studs and absence of treatment for stress relief after thread grinding operation.

From summary: "Experiments to determine effectiveness of additives to reduce radiolytic damage to a mixture of isometric terphenyls in which 93 separate compounds were evaluated."

Comparison of the economics of mechanical shear-leach and Sulfex decladding-core dissolution head end treatments for processing typical tubular bundles of stainless steel-jacketed UO2 nuclear fuels.

Report that describes the Oak Ridge National Laboratory Experimental Gas-Cooled Reactor, problems with the procurement and assembly of its components, and its economic feasibility.

From abstract: "Fuel materials proposed for use in the Advanced Test Reactor were evaluated."

Study conducted for the U. S. Army Reactors Branch of reactors for the irradiation of food.

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.

"The studies of this report were made to indicate the thermal stability at elevated temperatures of fluorocarbons similar to the hexadecafluoroheptane" (p. 4).

"This report covers fire tests conducted on an experimental design of a wooden-jacket shield for radioisotope shipping containers" (p. i).

This report consists of the description, drawings, connections, and schematics of the various control elements that make up the control system of the Engineering Test Reactor Critical Facility (ETRC).