Abstract: The preliminary design is described for a small electric-power-generating plant powered by an organic moderated reactor. System and component requirements are discussed and possible design configurations and equipment are described.
Abstract: This report describes an advanced sodium-cooled, graphite-moderated nuclear power plant which utilizes high-pressure, high-temperature steam to generate electricity at a high thermal efficiency.
Abstract: This report, intended as a working document, contains analytic representations and analog models of the Hallam Nuclear Power Facility as used in studies of the Control and Protection Systems.
Abstract: Design charts and systematic design forms are presented for simplified calculations to check the number of convolutions and thickness required to limit the deflection and pressure stress range in three types of bellows.
Annual report with the objectives of evaluating, producing, and maintaining an up-to-date set of basic nuclear data; producing and evaluating multigroup constants; and improving of present day methods of neutronic calculations as related to microscopic and macroscopic nuclear data, for unclassified research sponsored by the U.S. Atomic Energy Commission during FY 1968.
Abstract: A calculational method is presented which may be used to determine fast and thermal neutron flux distributions at deep neutron penetrations in hydrogenous shields.
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.
From abstract: The vapor pressure determination of solid ThF4 has been completed. Experiments to determine the heat of formation of ThF4 by studying the reaction of ThF4 with SiO2 are described.
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: 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: 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.
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.
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: 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.
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.
"An intuitive approach to the understanding of crystal structures is presented in terms of the concept of the closest packing of spheres. The qualitative features of the concept are sorted out and correlated by successively treating single, double, triple, and multiple layered arrays of closest packed spheres" (p. ix).
Abstract: An evaluation extending over a two-year period was made of primary system sodium and of stainless steel, zirconium, and beryllium specimens exposed in the hot and cold legs of a bypass loop in the primary system of the Sodium Reactor Experiment (SRE).
Abstract: A series of clean critical experiments was performed in the SGR critical facility utilizing 2 wt % enriched, uranium metal, hollow cylinder, fuel elements, in AGOT graphite moderator.
Abstract: The means to prevent the recurrence of tetralin leakage into the SRE sodium systems are discussed. Included is a description of the redesign of system components to utilize alternate coolants such as nitrogen, air, and NaK.
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: This report describes the features of several systems which were rejected as inadequate, and the evaluation of a design leading to the construction of a prototype cask and its associated equipment.
Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.
Abstract: A device to detect the presence of hydrogen in sodium has been developed. Such a device, installed in a sodium heated steam generator, would signal the presence of water in the sodium resulting from a leak in the sodium-water barrier.
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: A fuel material evaluation was made by destructively examining a full-scale experimental fuel element, irradiated in the SRE to a maximum of 850 Mwd/MTU.
Abstract: Pressure-temperature isochores were obtained for zirconium-hydrogen alloys, spanning the H/Zr composition range of 1.430 to 1.910. The studies were confined to the temperature limits of 300 to 900ºC, and the pressure limits of 0.01 to 10.0 atm.
Abstract: The experimental application of centrifugal clarification, precoat filtration, conventional filtration, and adsorption to the removal of impurities from a bypass stream of irradiated reactor coolant at the Organic Moderated Reactor Experiment is described and evaluated.
Abstract: Irradiated fuel elements from the Organic Moderated Reactor Experiment (OMRE) first core loading have been examined and evaluated to determine: (1) the stability of the floating plate fuel element design, (2) the stability of the stainless steel clad UO2 - stainless steel cermet core fuel plates under irradiation and exposure to the organic coolant, (3) the extent and nature of deposits on the fuel element services, and (4) the distribution of burnup in the fuel elements.
Abstract: Presented herein is the preliminary design of a natural uranium, graphite moderated, CO2-cooled reactor and power plant similar to, but larger than, the British Calder Hall plant, with a net electrical output of 130 MWE.
This report summarizes the results and conclusions of a study made to evaluate the merits of using zirconium hydride as a solid moderator in an integral boiling water-nuclear superheat reactor of the pressure vessel type.
Abstract: Because of the outstanding heat transfer efficiency of sodium, it is necessary in sodium-cooled reactors to consider and attempt to prevent the occurrence of adverse stresses as a result of thermal transients in the system.
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.
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.
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: Electrical resistance measurements and metallography were employed in a study of the kinetics of the gamma to gamma prime transformation in the uranium-molybdenum system at 16 wt% molybdenum.
From summary: The full 140 element loading of the core was completed on October 10, 1962. At this point, critical operation was begun for operator training and post-critical testing purposes.
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