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.

Abstract: Uranium monocarbide shows considerable potential for use as a fuel in high temperature, high power density, nuclear power reactors. As Atomics International is proposing its use in sodium graphite type reactors, it was necessary to develop a process for fabricaing sodium bonded uranium carbide fuel elements.

Abstract: The fast and thermal neutron leakage from the 30 megawatt SGR-P4 reactor has been studied by three independent methods.

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.

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

Abstract: Cast and Wrought specimens and restrained wrought specimens of unalloyed uranium were irradiated in the Materials Testing Reactor, as the first in a series of experiments to develop fuel materials for sodium cooled reactors.

"Heat production resulting from the absorption of gamma ray photons in thermal shields and the leakage of neutrons and photons from ferritic thermal shields are investigated. The gamma rays considered arise from three types of reactor radiation -- thermal neutrons, fast neutrons, and core and reflector gammas. The energy spectra of the fast neutron leakage and absorption have been investigated in some detail because of the significant contribution of fast neutrons to the heating of the concrete biological shield."

"Engineering pertinent to the development of the sodium-cooled, graphite-moderated type of reactor was continued. This included work on problems related to the zirconium canned moderator, low enrichment uranium fuel, sodium piping, secondary coolant system, shielding, and the control and safety elements. A large fraction of the work was devoted specifically to problems of the proposed Sodium Reactor Experiment (SRE) configuration. In this connection, an integrated effort was initiated to prepare a complete preliminary design of the SRE by an early date. In addition, two alternate sodium-graphite reactor configurations were studied. One was an intermediate size, 145 thermal megawatt, unit optimized for the production of low cost plutonium. The second was a low power 10 thermal megawatt intended for power production, but in which sodium circulation through the core was entirely dependent upon thermal convection."

Introduction: The purpose of this experiment was to demonstrate that a liquid metal heat transfer system could be built and operated at very high temperatures.

"Scale-up work on high temperature fuel recovery processes has progressed to the point where the (high temperature) vacuum furnace for several operations to the hot cells has been completed and tested under operating conditions. Small scale experiments on high temperature methods for processing molten irradiated uranium fuel have been made with spent X-10 fuel slug pieces. The results of direct Pu evaporation, treatment with fused fluorides and oxide scavenging were every similar to those found with tracer experiments."

From introduction: This report details the extensive modifications to the SRE heat transfer systems required to accomplish increased capacity objectives.

Annual report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the 1965-66 fiscal year.

Abstract: An experimental and analytical research program is described which is designed to meet certain specific needs for data and methods required to make improved predictions of transient voids, burnout, flow, and fuel temperature during extreme accidents in sodium-cooled reactors.

Abstract: This report describes a 300,000-kwe, sodium-cooled, graphite-moderated nuclear power plant based on existing technical information.

Abstract: This report presents steam cycle optimization studies for large sodium graphite nuclear power generating stations.

Abstract: A new concept for continuously monitoring neutron flux has been successfully tested.

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: 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.

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.

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.

Summary: This report contains a description of the final design of the Piqua Nuclear Power Facility (PNPF); an outline of the test and operating procedures, and the organization and responsibilities; and a summary of the hazards and safeguards analyses that have been conducted to evaluate the safety of the facility operations.