From abstract: "The study evaluates the utilization of superheat and moisture separation techniques in the SM-2 design to improve the steam plant reliability and thermal efficiency. Ease of packaging, steam plant transient response, and capital costs are also considered. The reference design used for comparison is the steam cycle employing saturated steam at 600 psia described in APAE No. 40, Volume II."

Abstract: This technical report contains a description of the nuclear analysis performed upon the SM-2 core and vessel for the period September 1, 1958 to December 31, 1959. Calculations are given for core reactivity, power distributions, lifetime, reactor control, kinetics, radiation problems, fuel and poison burn-ups, and the nuclear effects of poisons, temperature, and geometry. Wherever possible, experimental data is included in order to test the validity of the analytical models. The SM-2 nuclear analysis was performed by Alco Products, in. under Tasks 1, 8, and 10 of Contract No. AT(30-2)-326 for the Atomic Energy Commission.

From abstract: "This report describes the design of the mobile nuclear power plant which is to be the prototype of a mobile, low-powered nuclear power plant intended to furnish electrical power in remote locations."

Discussing photographs of rocks from Basalt Mesa

Introduction: Development of a process for manufacturing fuel elements for the Yankee reactor and fabrication and testing of a prototype assembly is reported.

From introduction: A thorough investigation, both experimental and theoretical was carried out on plaque type radiators. Five phases of the work are reported: preliminary design analysis, experimental measurements, computational program and analytical studies, followed by a general discussion of source design principles. The results are summarized in graphs and formulae which may be applied to other design requirements.

Extensive experimental program undertaken to provide data on the operation and testing of a bismuth-to-bismuth dual-loop system.

Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.

Abstract: Accurate measurement of OMRE fuel-element surface temperatures is important for evaluation of the performance of organic compound as moderator and coolant materials in nuclear reactors; a mixture of polyphenyls has been used in reactor operation to-date.

Abstract: Equipment has been developed to handle the fuel elements and control rods for the Piqua (OMR) reactor.With the handling machine, which consists of a shielded cask mounted on a gantry, a fuel element can be replaced in the core in about 27 minutes.

Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.

"The study of the toxicity of Ra and MsTh(Ra/sup 228/) in humans was continued. Several Thorotrast cases were also examined. Physical and clinical studies were made on 151 persons. In all cases possible gamma ray measurements were made of Ra and MsTh decay products retained in the body, alpha -ray measurements of Rn and Rn/sup 220/ in breath, a complete medical history was taken, a complete x-ray examination was made of the skeleton, and urinalysis, hematological, and blood chemistry studies were made."

Introduction: In conjunction with the final design and development of a 70 MW sodium intermediate heat exchanger and a sodium steam generator, an analysis is required which can be used as a basis for a determination to scale up or scale down the designs. Included in this analysis are those considerations leading to the recommendation of the best prototype test unit and to some of the limits imposed on scaling up or down when considering future applications of designs other than those actually tested. In addition, these considerations include aspects required to accurately predict the performance, operation, mechanical reliability, and feasibility of fabrication of the 70 MW design.

Abstract: Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of commercially available valves considered for use in the sodium coolant system of the Hallam Nuclear Power Facility.