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The Closest Packing of Spheres (A Unifying Basis for Crystal Structures)
Abstract: An intuitive approach to the understanding of crystal structures is presented in terms of the concept of the closest packing of spheres.
Hallam Nuclear Power Facility, Reactor Operations Analysis Program: Semiannual Progress Report Number 2, March-August 1963
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.
Stress and Elevated Temperature Fatigue Characteristics of Large Bellows
From abstract: Charts in this report show axial stress distribution over exterior bellows surfaces induced by bellows axial deflection, and by internal pressurization. The influence of root rings on stress distribution is presented graphically.
An Evaluation of the Calder Hall Type of Nuclear Power Plant
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.
Steam Cycle Optimization Study for Large Sodium Graphite Nuclear Power Generating Stations
Abstract: This report presents steam cycle optimization studies for large sodium graphite nuclear power generating stations.
200-Mwe Prototype Large SGR: Reactor Structure Design and Evaluation
From abstract: This document presents the reactor structure design and evaluation for a 200-Mwe prototype large SGR.
Piqua Nuclear Power Facility Operations Analysis Program Progress Report Number 4: January-June 1964
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.
Hazards Analysis of the Organic Moderated Reactor Experiment
Introduction: The description of the Organic Moderated Reactor Experiment, (OMRE), its location, its safety system, and operative procedures have been previously detailed. The present report, although dealing with the subject of OMRE safety, has the more detailed intent of (1) determining the behavior of the OMRE under extremely unlikely sets of conditions; and (2) providing additional design information in the areas of reactivity coefficients, burnout heat flux, and reactor control.
OMR (Piqua) Unitized Control-Safety Rod Prototype Tests
Abstract: A unitized magnetic jack driven control-safety rod has been developed for the 45.5 thermal megawatt organic moderated reactor (Piqua).
A Device for Continuous Detection of Hydrogen in Sodium
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.
Steady-State Tests of a Reactor Safety Device
Introduction: This is the fourth report in a series of reports dealing in particular with a reactor safety device of a double diaphragm type.
Remote Maintenance Techniques for the Processing Refabrication Experiment
Abstract: This report outlines the general techniques developed for replacement of PRE in-cell equipment or equipment components.
OMR Control-Safety Rod Component Development Tests
Abstract: A magnetic-jack control-safety rod is under development for the 45.5 thermal megawatt Organic Moderated Reactor. The rod is "unitized," i.e., the poison element, drive, position indicator, and shock absorber are contained in a compact assembly which is inserted in a regular fuel channel opening in the core. Tests to develop components capable of operating under these conditions are described and results are reported.
Reactivity Absorbed by Xenon-135 in the SRE
Abstract: The measurement and calculation of the reactivity absorbed by Xe135 as functions of time after shutdown for the SRE are described.
Organic Moderated Reactor Experiment : Safeguards Summary
Abstract: This report presents a description of the Organic Moderated Reactor Experiment (OMRE), of the hazards associated with this experiment, and all of the safeguards taken to ensure the safety of the operating personnel and the population of the surrounding area.
Hot-Pressure Bonding of OMR Fuel Plates
Abstract: An alluminum-clad low-enrichment, uranium-alloy fuel element of flat plate configuration has been proposed for the Organic Moderated Reactor (OMR).
Plutonium: Enriched OMR Cores
Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.
Maritime Organic Moderated and Cooled Reactor
Introduction: This document describes the results of a six-week conceptual design study of an organic moderated and cooled reactor (OMCR) power plant adapted to a Class T-7 tanker.
Increasing Thermocouple Reliability for in-Pile Experiments
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.
QUICKIE: A Computer Program for Spatially Independent Multigroup Slowing-Down and Thermalization Calculations
Introduction: QUICKIE is a computer program designed to solve the multigroup neutron slowing down and thermalization equations without consideration of spatial dimensions.
Hallam Nuclear Power Facility, Reactor Operations Analysis Program: Semiannual Progress Report Number 3, September 1963-February 1964
From introduction: This report provides industry, plant operators, and the scientific community with information covering the results of the performance analysis.
Inpile Experiments on Retention of Fission Products in 500°F Sodium
Abstract: The results of three separate inpile capsule experiments are presented.
Organic Moderated Reactor Experiment Progress Report: August-October 1956
Report describing technical progress on the design and construction of an Organic Moderated Reactor Experiment (OMRE), to be operated at the National Reactor Testing Station in Arco, Idaho. "This is the second report of the series, and covers the period from August 1, 1956 through October 31, 1956" (p. 3).
Final Design of Sodium-Heated, Modular, Steam Generators for the SCTI
Abstract: The following report covers the final design of the modular steam generators.
Sodium Reactor Experiment Power Expansion Program: Heat Transfer Systems Modifications
Abstract: Under the Power Expansion Program (PEP), modifications have been made to the Sodium Reactor Experiment (SRE) facility to improve plant reliability and permit an increase in power to 30 Mwt, with a reactor coolant outlet temperature up to 1200°F.
Carbide Fuels in Fast 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.
Operating Experience with Heat Transfer System Pumps at the Hallum Nuclear Power Facility
Introduction: It is the purpose of this report to describe the operating and maintenance experience obtained at HNPF on the sodium heat transfer pumps.
Organic Moderated Reactor Experiment Progress Report: October 1955-July 1956
Report describing technical progress on the design and construction of an Organic Moderated Reactor Experiment (OMRE), to be operated at the National Reactor Testing Station in Arco, Idaho. "This is the first report of the series and coves the period from the initiation of the project to July 31, 1956. Also included as an appendix to the report is a detailed description of the OMRE facility" (p. 3).
A Rotary Kiln for the Controlled Oxidation of UC
Abstract: A rotary kiln was evaluated for the controlled oxidations of UC.
UC Fuel Element Design and Fabrication
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.
Reactor Safety Quarterly Progress Report: May-July 1956
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.
SNAP 10A Structural Analysis
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.
Structural Test on the Final SNAP 10A Prototype System
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.
The High-Temperature Irradiation Behavior of Hypostoichiometric Uranium Monocarbide
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.
Organic Reactor Waste Gas Analyzer
The design of a waste-gas treatment system for organic moderated reactors requires a knowledge of reactor waste-gas composition, generation rate, and radioactivity. To obtain data on these variables a continuous stream analyzer was constructed to analyze the waste gas from the Organic Moderated Reactor Experiment (OMRE).
Experimental Evaluation of a Sodium-to-Sodium Heliflow Heat Exchanger at Temperatures up to 1200°F
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.
Evaluation of Irradiated OMRE Fuel Elements First Core Loading
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.
Preliminary Test of Natural-Circulation Double-Tube Steam Generator
Abstract: Testing of the Natural Circulation Steam Generator has been conducted with the Sodium Reactor Experiment. A necessary modification of the original feed-water control system was utilized.
Coolant Flow and Outlet Temperature: Computer-Monitors for the Hallam Nuclear Power Facility Plant Protective System
Abstract: The design and application of two computers for the HNPF protective system is discussed.
Power Flattening in Sodium Graphite Reactors by Spatial Variation of Moderator Properties
Abstract: In the present study, the variation of moderator composition was postulated to be effected by the inclusion of varying amounts of beryllium oxide in the graphite of an SGR.
Development of a Fuel Handling System for an Organic Moderated Reactor
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.
A Multichannel Digital Recording System
Abstract: This report is a description of a 200 channel digital recording system used to record high temperature strain gage outputs and associated temperatures.
Piqua Nuclear Power Facility Operations Analysis Program Progress Report Number 5: July-December 1964
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.
HNPF Control System Response During Transient Conditions
Abstract: This report documents the results of power ramps inposed on the HNPF during the month of January 1964.
Capsule Irradiation of Unalloyed Uranium at High Temperatures
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.
Flux Control for Irradiation Experiments
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.
SNAP 2: Structural and Dynamic Analysis
Abstract: The structural design criteria and the system basic loads for the SNAP 2 compact power unit are presented.
Boiling Studies for Sodium Reactor Safety: Part 1, Experimental Apparatus and Results of Initial Tests and Analysis
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.
The FAIM Code: a Multigroup, One-Dimensional Diffusion Equation Code
Abstract: FAIM is a general multigroup, one-dimensional diffusion equation code programmed in FORTRAN language for the IBM 7090 computer.
Critical Experiments on Slightly Enriched Uranium Metal Fuel Elements in Graphite Lattices
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.