Abstract: The Sodium Reactor Experiment Mark II Fuel Handling Machine has been modified to ensure fuel and gas containment during core III operation. A new fuel control system has been designed for the fuel handling machine.

Abstract: A rotary kiln was evaluated for the controlled oxidations of UC.

From abstract: This report summarizes the results of SNAP-10A/Agena developmental testing and final vehicle systems tests.

Abstract: A fuel development program has been conducted, to investigate the behavior of U-10 wt% Mo alloy at the fission rates and temperatures anticipated in Core I of the Hallam Nuclear Power Facility.

Abstract: Atomics International conducted a fuel irradiation program to investigate and define the behavior of U-10 Mo fuel at the temperatures and fission rates anticipated in Hallam Nuclear Power Facility (HNPF) Core I.

Abstract: A limited study of the compound U₃Si₂ has been made in reference to its use as a prospective nuclear fuel in sodium cooled reactors.

Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.

From abstract: The Madelung constant and the inverse twelfth power repulsion factor have been calculated for the wurtzite structure for wide ranges of the crystal parameters and u.

Abstract: The following report covers the final design of the modular steam generators.

Abstract: Cladding and fuel material processing prospects are reviewed, and fuel system possibilities for near-term (~1 mill/kwh) and long-range (<0.5 mil/kmh) fuel cycles are described.

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.

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.

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: 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.

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.

Quarterly report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the first quarter of the 1966 fiscal year.

Abstract: The results indicated that increased reliability can be obtained by using thermocouples made with insulation of increased density and/or a low thermal expansion sheath.

Abstract: This report reviews various means which have been used to control experimental conditions in irradiation experiments, and presents the concept of local flux control.

Abstract: This report describes in detail two designs of a nominal 6-kwe Nuclear Power Plant (NPP), one using thermoelectrics for power conversion and the other using the Mercury-Rankine cycle NPP.

Abstract: This report presents the pertinent results of a series of fatigue tests relating to the evaporator and superheater 37-tube module heads for the sodium heated steam generator described in NAA-SR-9826.

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