Introduction: On November 14, 1961, an environmental test program commenced on the first SNAP 10A system structure, S10A-PSM-1.

Abstract: Pressure stress-rupture specimens of thin walled Type 304 stainless steel tubing have been tested at temperatures to 1300°F in the presence of an internal diffusion limited carbon source.

From abstract: This document presents the reactor structure design and evaluation for a 200-Mwe prototype large SGR.

Abstract: In fulfillment of the general objective of developing information on two-phase flow required in the safety evaluation of sodium cooled reactors, pool and forced-convection boiling of sodium were studied both experimentally and analytically.

From abstract: The development, acceptance, and qualification tests performed on the SNAP-10A Ejectable Heat Shield and components, the results of those tests, and the conclusions drawn from the results are presented in this report.

Abstract: An analytical and experimental effort toward a better understanding of the mechanisms involved in nuclear power reactor loss-of-pressure accidents is presented.

From abstract: The fuel element assembly was successfully irradiated in the SRE to a maximum burnup of 5300 Mwd/MTU, at a peak fission rate of approximately 1.5 x 10E13 fissions/cm3-sec and a maximum central temperature near 1200F.

From abstract: The results of a feasibility study of a 1000-Mwe sodium-cooled fast-reactor are presented.

Abstract: This report documents the results of power ramps inposed on the HNPF during the month of January 1964.

From abstract: Impact tests on the Mark III Hanford safety element indicate that the design is mechanically adequate and does not constitute a personnel hazard.

Abstract: this report discusses and summarizes all stress analysis done on the SNAP 10A system; it also mentions many of the structural tests which were accomplished.

Abstract: This report presents the results of the structural tests which were performed on the S10A-PSM-1A system to qualify the basic structural design for the anticipated launch environment.

From abstract: A series of three multi-compartmented experiments was designed to investigate the irradiation behavior uranium carbide fuels, as a function of composition, temperature, and burn-up; and the mechanical properties of Type 304 stainless steel, as a function of temperature and integrated thermal neutron flux.

From abstract: A development program was initiated to design, fabricate and test an absorber column with a 10 to 12 year lifetime, for use in the proposed LSGR.

From summary: This report summarizes environmental monitoring results for the first six months of 1962.

Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.

Abstract: The design and initial test of cutting and welding equipment developed to remotely cut and re-weld the bottom process tube joint are discussed in this report.

Progress report for the Piqua Reactor Operations Analysis Program describing observations and analyses at the Piqua Nuclear Power Facility (PNPF). The program goals are to monitor operations and collect data in order to ensure that the plant's operation is safe, to improve design and performance, to evaluate the performance and lifetime of the plant's components and systems, to evaluate plant safety and safeguards, and to disseminate all information to the scientific community.

Abstract: Perforated BeO plates were thermal shock tested to evaluate the effect of: (1) localized temperature variations adjacent to the perforations, and (2) radial gradients across the entire plate.

Progress report for the Piqua Reactor Operations Analysis Program describing observations and analyses at the Piqua Nuclear Power Facility (PNPF). The program goals are to monitor operations and collect data in order to ensure that the plant's operation is safe, to improve design and performance, to evaluate the performance and lifetime of the plant's components and systems, to evaluate plant safety and safeguards, and to disseminate all information to the scientific community.

Abstract: Two in-pile capsules containing mock-up uranium carbide fuel rods were subjected to a total of five neutron bursts in the TREAT facility with initial reactor periods ranging from 0.26 to 0.10 sec.

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"An experiment to determine the effect of reactor irradiation on the thermal conductivity of uranium-impregnated graphite at elevated temperatures as described. The results show a decrease in the thermal conductivity saturating at [approximately] 60 percent at a temperature of 700 degrees C; at [approximately] 50 percent at a temperature of 1000 degrees C; and at [approximately] 25 percent at a temperature of 1300 degrees C. It was found that after irradiation at a given temperature, exposure at a higher temperature resulted in an increase in the thermal conductivity. The converse was also observed. Within the precision of measurement there was no difference in effed between temperature changes produced by varying the fission rate in the samples and changes produced by varying the power in an external heater."

"Engineering was continued on the development of sodium cooled, graphite moderated type reactors. General studies were carried out as well as studies specifically devoted to the following: a. full scale poser-only plant, b. thirty-mega watt pilot plant, the SGR, c. sodium reactor experiment, the SRE. This work consisted of theoretical analysis of various aspects of nuclear performance; economic investigations of different fuel element, cooling system and plant arrangements; and experimental investigations related to the properties of certain materials and to the development of components. Preliminary consideration was given to alternative reactor arrangements employing liquid hydrocarbon moderators and high temperature coolants other than sodium. In addition to a summary of the general design features of the SRD, a program was prepared outlining the proposed use of this installation.