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30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System
Final design for the 30 megawatt intermediate heat exchanger and steam generator.
30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System: Volume 2, Chemical and Stress Analysis
Chemical engineering analysis and stress analysis for design of the 3 megawatt intermediate heat exchanger and steam generator for service with a liquid sodium heat transfer fluid.
30 Megawatt Heat Exchanger and Steam Generator for Sodium Cooled Reactor System: Volume 4, Operation and Maintenance Procedures
Operation and maintenance procedures for 30 megawatt heat exchanger and steam generator for sodium cooled reactor system.
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.
300,000-KWE SGR Nuclear Power Plant of Current Technology
Abstract: This report describes a 300,000-kwe, sodium-cooled, graphite-moderated nuclear power plant based on existing technical information.
568 Report
Final report for a grant contract documenting information about the scope of the project and results.
1000 MegaWatt Boiling Water Reactor Plant Feasibility Study: Volume 1
From introduction: "Summary report of the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company."
1000 MegaWatt Boiling Water Reactor Plant Feasibility Study: Volume 2
From introduction: "Presents the detailed description of a 1000 MW Electric Power Plant employing one or two Boiling Water Reactors as the steam source."
1000 MegaWatt Boiling Water Reactor Plant Feasibility Study: Volume 3
From introduction: "Contains the appendices and a complete set of drawings related to the 1000 MWe Boiling Water Reactor Plant Feasibility Study performed by the General Electric Company for the United States Atomic Energy Commission."
6144-Channel Time-of-Flight Analyzer
Report describing a 6144-channel analyzer designed and built for the purpose of analyzing time-of-flight during studies of slow neutron scattering at Hanford Laboratories. This includes descriptions of the analyzer, its logic and circuits, and its test mode.
ABCC-NIH Adult Health Study Hiroshima 1958-60. Cardiovascular Project Report Number 6, Heart Size Norm
Data on 13,000 person 15 yr of age or older obtained during detailed clinical examinations, including radiological recorded heart size, were correlated with sex, age, height, and weight of subjects to arrive at a standard heart size for Hiroshima residents This information will be used in investigations cardiovascular disease in the population.
The Absolute Abundance of the Chromium Isotopes in Some Secondary Minerals
From abstract: "Isotopic assays have been made on the chromium in samples from fourteen different chrominiferous minerals from different geographic and meteoritic sources. The results of the assays indicate that it is not possible to unequivocally state that variations in isotopic compositions have been observed."
ABWR: PL-2 Design Report
From preface: This report satisfies the quarterly progress report requirements for PL-1 and PL-2 plant design work for the period ending September 30, 1960 At present time a SL-1 Core 2 is under construction. This is a replacement core for SL-1 (ALPR) and will be identical to a PL-2 core; a PL condenser is under test at the SL-1 facility; final construction plans for PL components and modules which are not site sensitive will be completed in March 1961.
ABWR Design and Development Quarterly Progress Report: July 1 - September 30, 1961
Quarterly design and development progress report on the activities of the Army Boiling Water Reactor (ABWR) Program.
ABWR Design and Development Quarterly Progress Report, January 1 Through March 31, 1962
Quarterly design and development progress report on the activities of the Army Boiling Water Reactor (ABWR) Program.
Accurate Nuclear Fuel Burnup Analyses; Eighth Quarterly Progress Report, (September - November 1963)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses; First Quarterly Report, (December 1961 - February 1962)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses; Ninth Quarterly Progress Report, (December 1963 - February 1964)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analyses; Second Quarterly Progress Report, (March - May 1962)
The objective of the Accurate Nuclear Fuel Burnup Analyses program is to develop more accurate methods for burnup analysis for general use than the current method of analysis of Ca-137 or Sr-90. The program will require from three to five years of effort.
Accurate Nuclear Fuel Burnup Analysis Quarterly Progress Report: Seventh Quarter, June 1963 - August 1963
Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.
Accurate Nuclear Fuel Burnup Analysis Quarterly Progress Report: Sixth Quarter, March 1963 - May 1963
Quarterly progress report on Accurate Nuclear Fuel Burnup Analysis project.
Acid-Base Equilibria in Tertiary Butyl Alcohol
From abstract: "The dissociation of acids in tertiary butyl alcohol has been studied by potentiometric, spectrophotometric, and conductimetric methods. Values for the over-all dissociation of perchloric and picric acids and several tetrabutylammonium salts were estimated by the Fuoss-Kraus treatment of conductance data. Potentiometric studies were carried out at constant ionic strength in order to minimize activity coefficient variations. An acidity scale was established from potentiometric measurements at a glass electrode, and conductance values of dissociation constants. A method was developed for the evaluation of the over-all dissociation constant of weak acids using potentiometric data for hydrogen ion activities and conductance data for the corresponding anion activities. Over-all dissociation constants are reported for perchloric acid, picric acid, 2,4-dinitrophenol, and benzoic acid. Apparent dissociation constants from potentiometric measurements at a constant ionic strength were determined for hydrobromic, nitric, hydrochloric, picric, and p-toluenesulfonic acids."
Acid-Base Reactions and Kinetics of the Halates in Fused Nitrates
From Abstract : "The mechanism of the reactions of the halates, bromate, chlorate, and iodate, with dichromate in fused alkali nitrates has been shown to involve a fast equilibrium followed by a slow rate determining strip to give oxygen and halogen gases as final products." Experiments outlined serve as insight into the structure of fused salts or fused electrolytes.
Acid-Base Reactions in Fused Salts. Dichromate-Bromate Reaction
Technical report. From Abstract : "The reaction of Lewis acid and base, Cr2O7= and BrO3_, in fused KNO3 - NaNO3 mixtures has been shown to involve an equilibrium followed by a slow decomposition to gaseous products."
Acute Intravenous and Intraperitoneal Toxicity Studies on Sodium Pentaborate Decahydrate and Sodium Tetraborate Decahydrate
This technical report describes the toxicity observations on mice of varying borate-glucose molar ratios and relative potencies (p) (4) for the pentaborate and tetraborate drug. This report outlines the methods and results of this experiment and provides a discussion of the results.
Addendum to the Spert IV Hazards Summary Report: Capsule Driver Core
From abstract: "explain all important features pertaining to a new pulsed irradiation reactor, the Capsule Driver, Core, and to analyze the potential problems and hazards of operating this reactor in the existing Spert IV facility."
The Adsorption and Surface Reactions of Hydrocarbons on Clean Iridium
From abstract: "The adsorption of ethane, ethylene and acetylene on clean iridium in a field emission microscope has been found to cause characteristic changes in the work function of the iridium surface. Further changes, which are time and temperature dependent, result when such surfaces are heated. Flash filament experiments have shown that the changes in work function upon heating are due to desorption reactions and that the desorbed product consists principally of hydrogen. By assuming a linear relationship between surface coverage and work function, it has been possible to determine the desorption kinetics from the observed rates of work function change at various temperatures. The results are consistent with a mechanism involving stepwise surface dehydrogenation in which a pair of hydrogen atoms is removed from the hydrocarbon molecule in each step, followed by desoption of the adsorbed hydrogen. At very high temperatures the remaining carbon atoms are removed, presumably by evaporation."
Adsorption of Radioactive Gases on Activated Carbon
The purpose of this experiment is to study the quantitative adsorption characteristics of a carbon adsorber bed receiving a radioactive inert gas in a helium stream. An objective of the experiment is to measure the equilibrium transmission of the radio-active gas through a carbon adsorber in order to determine if radio-active decay of the adsorbed gas permits additional adsorption.
Adult Health Study : Review of Substudies, June 1962, Hiroshima and Nagasaki
Data are summarized from a series of studies to determine the late effects of radiation in adult populations of Nagasaki and Hiroshima. Results ae reported from studies on skin aging, hair greying, cardiovascular findings, neuromuscular response, antibody levels, ocular aging, auditory aging, and miscellaneous aging characteristics in persons exposed to radiation from the atomic bombs as adults; growth and development studies on exposed persons born between 1935 and 1945; the incidence of neoplasms in the exposed populations; possible genetic effects of radiation in selected groups; the incidence of tuberculosis and other infectious diseases in exposed populations; hematological studies; metabolic studies, and other related studies in exposed persons and their offspring. Possible future programs are discussed.
Advanced Designs and Special Applications for Fast Breeders
The purpose of this paper is to describe a few of the suggested advanced concepts for fast breeder reactors and to compare these with the standard approach as to their potential advantage. I have attempted to estimate the economic effect of full technical success with each of the proposed concepts. The proposed concepts include: (1) single sodium system, (2) steam-cooled core concept, (3) direct cycle reactor using potassium as reactor coolant and working fluid, (4) molten plutonium-fuel alloy circulated and cooled by a jet of sodium, (5) settled-bed core, (6) molten salt concept, and (7) paste-fuel system.
Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 1. Cost Analysis and Future Development
First part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. I-ii).
Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 2. Plant Conceptual Studies
Second part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. II-i).
Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 3. Analog Simulation of Reactor Plant Transients
Third part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).
Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 4. Steam Driven Coolant Pumps
Fourth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).
Advanced Indirect Cycle Water Reactor Studies for Maritime Applications: Part 5. Spiked Core Concept
Fifth part of the "final report of a study directed toward the evolution, design, and demonstration of the principle design features of interim indirect cycle water cooled and moderated nuclear power plants which will be useful in early cooperative programs between the Atomic Energy Commission and the United States maritime industry" (p. i).
Advanced Pressurized Water Reactor Study [Part 1, Supplement]: Phase 2--A Report
From introduction: This report details the progress of the Advanced Pressurized Water Reactor Study.
An Advanced Sodium-Graphite Reactor Nuclear Power Plant
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.
Advanced Test Reactor: Final Shielding Design Report
From abstract: "This report has been prepared as a design reference document describing the calculation methods and engineering results for shielding analysis work done during the design of the Advanced Test Reactor."
Advanced Test Reactor Servo Regulator Rod Test Program
From abstract: "Verify rod mechanism characteristics reported by the United Shoe Machinery Corporation, demonstrate that rod mechanism characteristics were compatible with reactor kinetics and that over-all reactor system behavior was stable, and to recommend system modifications needed for satisfactory performance."
Advanced Test Reactor Turbo Report
From abstract: "The time-dependent behavior of the Advanced Test Reactor was calculated by the Babcock & Wilcox Company on the Philco 2000 computer, using the Turbo depletion program."
AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel
Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was incorporated into the compacts either as fibers or as a thin coating on individual UO2 particles. Fuel pellets were produced from these materials by vacuum hot pressing. The distribution of the molybdenum in both types of cermet fuels appeared favorable to good heat transfer. The fibers were oriented predominantly in the radial planes of the pellet as a result of the uni-directional compaction during the hot-pressing operation. In the pellets made from the coated particles, a continuous network of molybdenum occurred as a result of the coating welding together during the hot-pressing operation. The test assembly contains eight fuel rods; three contain UO2-Mo cermet, three contain the cermet produced from the coated particles, and two are for reference and contain the conventional sintered UO2 pellet fuel. The nominal outside diameter of the fuel rods is 1.308 cm (0.515 inch), and the clad wall thickness if 0.051 cm (0.020 inch). the cladding material is Type-304 stainless steel. The fuel pellets were all centerless ground to achieve a uniform outside diameter and thereby control the pellet-to-clad diametral clearance within a range of 0.076 to 0.102 mm (0.003 to 0.004 inch). Operation of the fuel rods will be at high specific power levels with surface heat fluxes of about 157 W/cm(2) (~500,000 Btu/h-ft(2)). The assembly was designed for a lifetime of 4.1 x 10(20) fission/cc (15,000 MWD/T) exposure.
Fuel Cycle Program Progress Report: Thirteenth Quarter, July-September 1963
Quarterly progress report discussing activities related to the Vallecitos Boiling Water Reactor (VBWR) and related facilities
Aerial Radiological Monitoring System Part 4: Equipment and Procedures Through Fiscal Year 1966
From abstract: "This report describes the Aerial Radiological Measuring System (ARMS-II) operated by EG&G, Inc., for the Division of Biology and Medicine, U. S. Atomic Energy Commission."
An Aerodynamic Raindrop Sorter. Technical Progress Report No. 1.
A pilot model of an Aerodynamic Raindrop Sorter was constructed along the lines suggested by mathematical analysis. The function of the analyzer is to sort natural rain according to drop size and to collect the sorted drops for further analysis. The pilot model, a small wind tunnel inclined at 45 deg to the horizontal, demonstrated the feasibility of aerodynamic rain drop sorting over a wide range of drop sizes. (auth)
Aeroradioactivity Survey and Areal Geology of Parts of East-Central New York and West-Central New England (ARMS-I)
Report concerning "[a]n airborne gamma-radiation survey of Connecticut, Rhode Island, and parts of New York, Massachusetts, New Hampshire, and Vermont" (p. 5) made between 1958 and 1960 that indicated that a broad range of radioactivity exists in those areas depending on the type of bedrock. Correlations are drawn between this radioactivity and the geology of the region.
Aircraft Reactor Test Removal and Disassembly
Report documenting the dissection of a reactor called the Aircraft Reactor Test (ART). Includes the removal of the reactor from its test cell, component removal, and plans for a for a disassembly building facility.
An Algorithm for Construction Feasible Schedules and Computing Their Schedule Times
"An algorithm for the generation of feasible schedules and the computation of the completion times of the job operations of feasible schedule is presented. Using this algorithm, the distribution of schedule times over the set of feasible schedule—or a subset of feasible schedules—was determined for technological orderings that could occur in a general machine shop. These distributions are found to be approximately normal. Biasing techniques corresponding to “first come first serve,” random choice of jobs ready at each machine and combinations of these two extremes were used to compute distributions of schedule times."
Alpha Particle Radiolysis of Anion Exchange Resins
Technical report. From Abstract : "Irradiation of 'Dowex' 1, 'Permutit' S-1, and 'Permutit' SK anion exchange resins with alpha particles results in losses in ion exchange capacity and in 'apparent per cent crosslinkage'. The order of decreasing radiolytic stability for these properties in 'Permutit" SX > 'Permutit' S-1 > 'Dowex' 1."
An Alpha Scintillation Tester for Uranium Surface Contamination of N-Reactor Fuel
Report that "describes a nondestructive tester and some of its applications in measuring 10 to 100 µg of uranium surface contamination on unirradiated, low enrichment, uranium fuel elements" (p. ii).
Alternating Direction and Semi-Explicit Difference Methods for Parabolic Partial Differential Equations
"The energy method is applied to study the stability of two types of difference approximations to parabolic partial differential equations, the alternating direction methods Douglas, Peaceman, and Rachford, and a new semi- explicit method. Each difference scheme is proved to be unconditionally stable. These results apply to parabolic equations with variable coefficients, defined in cylindrical domains with an essentially arbitrary bounded base."
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