Technical Report Archive & Image Library (TRAIL) - 168 Matching Results
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- 11,400 KW Nuclear Power Plant Employing an Organic Moderated Reactor: Preliminary Description
- 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.
- 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.
- 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.
- Analog Models for HNPF Control and Protection Studies
- 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.
- Analysis of Stresses in Bellows
- 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 Technical Progress Report, AEC Unclassified Programs: 1965
- Annual report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the 1964-65 fiscal year.
- Annual Technical Progress Report, AEC Unclassified Programs: 1966
- Annual report describing progress on unclassified research programs funded by the U.S. Atomic Energy Commission during the 1965-66 fiscal year.
- Annual Technical Progress Report, AEC Unclassified Programs: Fiscal Year 1968
- 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.
- Application of Fast Neutron Removal Theory to the Calculation of Thermal Neutron Flux Distributions in Reactor Shields
- 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.
- Application of Nuclear Power Plants (SNAP Units) to the Manned Orbiting Research Laboratory (MORL)
- 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.
- Basic Chemistry of High Temperature Inorganic Systems, Semiannual Progress Report: July-December 1956
- 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.
- Boiling Depressurization Transients
- Abstract: An analytical and experimental effort toward a better understanding of the mechanisms involved in nuclear power reactor loss-of-pressure accidents is 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.
- Boiling Studies for Sodium Reactor Safety: Part 2, Pool Boiling and Initial Force Convection Tests and Analyses
- 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.
- Calandria Core Weld Joint Development
- 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.
- Calculations of the Madelung Constant and Inverse Twelfth Power Repulsion Factors for the Wurtzite Crystal Structure
- 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.
- 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.
- 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.
- Chemical Development, Quarterly Progress Report, October-December 1953.
- Introduction - The work of the Chemical Development Group has included studies on the thermal and radiation stability of organic materials suitable for reactor coolants, the thermal and radiation stability of zirconium hydride, reactor safety devices involving chemical systems, and general analytical development.
- The Chemical Effects of 1 Mev Electrons on BrF3 at 25 degrees C
- "An investigation of the chemical effects of 1-Mev electrons on BrF3 at 25 degrees C has been carried out. Pressure measurements taken during the irradiation suggest the presence of Br2 and BrF5 as decomposition products and a fractional distillation of the irradiated liquid confirmed their presence. The extent of decomposition was determined both by fraction distillation and spectrophotometric methods. The radiation effect seemed to reach saturation when approximately 10 per cent of the BrF3 was destroyed. The exposure necessary for the decomposition products to reach a concentration of half the saturated value was calculated to be 2.7 microampere hours/cc BrF3 while the "G" value was found to be 1.5. A qualitative comparison of irradiation dosages from the Statiltron with that expected from spent fuels revealed that little decomposition of BrF3 reagent is to be expected from 1-say cooled Hanford fuel (in pile for 100 days) while in the case of 1-day cooled MTR type fuel (in pile for 12 days) a saturated effect might be realized in 1-3 hours. Since at most only 10 per cent of the BrF3 is destroyed it is concluded that BrF3, from a radiation resistance standpoint, is a suitable standpoint, is a suitable reagent for the processing of short cooled fuels."
- 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.
- The Closest Packing of Spheres (A Unifying Basis for Crystal Structures)
- "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).
- A Conceptual Design of a Thorium-Uranium (233) Power Breeder Reactor
- From abstract: A conceptual design study has been performed for a sodium cooled, graphite moderated, thermal power-breeder reactor utilizing the Thorium-Uranium 233 breeding cycle. Several aspects of the design of the system are considered but no attempt has been made to supply all the details. It appears that the design presented is feasible and will allow the production of economic power as well as full utilization of thorium resources.
- The Continuous Neutron Flux Monitor Project: An Interim Report
- Abstract: A new concept for continuously monitoring neutron flux has been successfully tested.
- 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.
- Corrosion and Activity Transfer in the SRE Primary Sodium System
- 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).
- 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.
- Design Modifications to the SRE during FY 1960
- 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.
- Design of an Experimental Bowable Fuel Element for the SRE
- Abstract: An experimental bowable fuel element was developed to study temperature oscillations in the second core loading of the SRE.
- Development, Acceptance and Qualification Testing of the SNAP 10A Ejectable Heat Shield
- 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.
- 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.
- Development of High-Temperature Electrical Ground Test Heaters for the SNAP 10A Program
- Introduction: The development and qualification of the system acceptance test heaters and the reactor simulator heater are described in this progress report.
- 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.
- The Distribution of Tracer Plutonium and Fission Products Between Molten Uranium and Solid Uranium Oxide, Carbide, and Nitride
- "A study has been made of the distribution of tracer fission products and plutonium between small samples of molten uranium and solid uranium oxide, carbine, and nitride. The distribution showed the same behavior i general for all three materials: 1. The rare earth elements, Cs, Ba, and Sr were extracted primarily into the solid scrub phase. 2. Zirconium and Nb partially concentrated in the scrub phase. 3. Plutonium, Mo, and Ru tended to remain completely in the metal phase. The distribution of activities agreed with trends predicted from the thermodynamic data. Uranium oxide appeared to be the most desirable scrub material for removing large amounts of fission products from the uranium while leaving beind the Pu. In addition the uranium metal was not severley contaminated by dissolved oxide."
- Dynamic Void Fraction Measurement System
- Abstract: Various methods and techniques of measuring void fractions in boiling heat transfer media are discussed.
- Effect of Carburization on the Low Strain Rate Behavior of Type 304 Stainless Steel: Interim Report
- 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.
- Effect of Reactor Irradiation on the Thermal Conductivity of Uranium Impregnated Graphite at Elevated Temperatures
- "An experiment to determine the effect of reactor irradiation on the thermal conductivity of uranium-impregnated graphite at elevated temperatures as described. The results show a decrease in the thermal conductivity saturating at [approximately] 60 percent at a temperature of 700 degrees C; at [approximately] 50 percent at a temperature of 1000 degrees C; and at [approximately] 25 percent at a temperature of 1300 degrees C. It was found that after irradiation at a given temperature, exposure at a higher temperature resulted in an increase in the thermal conductivity. The converse was also observed. Within the precision of measurement there was no difference in effed between temperature changes produced by varying the fission rate in the samples and changes produced by varying the power in an external heater."
- Engineering Evaluation of a Mixed Alloy Fuel Element Irradiated at Elevated Temperatures in the SRE
- 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.
- Environmental Monitoring Semiannual Report: January-June 1962
- From summary: This report summarizes environmental monitoring results for the first six months of 1962.
- Equilibrium Dissociation Pressures of the Delta and Epsilon Phases in the Zirconium-Hydrogen System
- 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.
- Evaluation of Coolant Impurity Removal Equipment at the OMRE
- 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.
- 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.
- 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.
- Evaluation of Zirconium Hydride as Moderator in Integral, Boiling Water-Superheat Reactors
- 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.
- 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.
- Fabrication Modification Development for OMRE Third Core Loading
- Abstract: This report describes the fabrication of elements for the OMRE third core loading.
- 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.
- Fatigue Characteristics of 37-Tube, Modular, Steam Generator Head
- 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.
- Feasibility Study of a 1000-Mwe Sodium-Cooled Fast Reactor: Volume 1 - Technical and Economic Potential
- From abstract: The results of a feasibility study of a 1000-Mwe sodium-cooled fast-reactor are presented.