UNT Libraries Government Documents Department - 168 Matching Results
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- 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).
- 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.
- Irradiation behavior of uranium monocarbide fuel experiments NAA 81-3 and AI 3-12
- A report regarding irradiation behavior of uranium monocarbide fuel experiments NAA 81-3 and AI 3-12
- U-10 Wt % Mo Fuel Element: Irradiation in SRE
- 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.
- Graphical Aids in the Calculation of the Shielding Requirements for Spent U²³⁵ Fuel
- Abstract: The data presented herein, in the form of graphs, can be used to obtain the value of this energy.
- Performance of HNPF Prototype Free-Surface Sodium Pump
- Abstract: A free-surface centrifugal pump, incorporating a hydraulic bearing running in sodium, was operated at the conditions required for service in the HNPF.
- Piqua Nuclear Power Facility Operations Analysis Program Progress Report Number 2: 1963
- 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.
- Piqua Nuclear Power Facility Operations Analysis Program Progress Report Number 3: July-December 1963
- 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.
- 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.
- 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.
- Metallurgical Aspects of SRE Fuel Element Damage Episode
- Abstract: An investigation of the metallurgical aspects of the SRE fuel element episode, that occurred July 26, 1959, has been completed.
- 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.
- A Further Evaluation of the Calder Hall Type of Nuclear Power Plant
- Abstract: This report presents the results of plant optimization studies and cost estimates of the reference design for a natural uranium, graphite moderated, gas-cooled reactor and power plant which was described in NAA-SR-1833.
- 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 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.
- Separations Chemistry, Quarterly Progress Report, April-June 1954
- "Scale-up experiments on high temperature fuel recovery processes have included the dummy run phase on the handling of 1-kologram samples of molten, non-irradiated uranium in the hot cell. The next step involves the use of spent X-10 fuel slugs. Small scale experiments with X-10 uranium on the extaction of Pu with Mg show that as much as 80 per cent of the Pu can be removed in pone pass. Treatment of uranium with fused fluorides can remove at least 90 per cent of the Pu in one pass. Oxide scavenging with ZrO2 is very effective in removing rare earths.:
- Thermal Cycling and Leakage Tests of 12-inch Valves for Sodium Service
- Abstract: Tests were performed to determine the effect of thermal cycling on the across-the-seat leakage characteristics of commercially available valves considered for use in the sodium coolant system of the Hallam Nuclear Power Facility.
- A Pebble-Bed Reactor for Stationary Power Plants
- A preliminary study has been made of a solid homogeneous reactor for stationary power plant application. The core consists of graphite spheres impregnated with uranium and thorium, and the coolant is bismuth. This concept possible offers advantages over other solid fuel reactor systems with respect to simplification of core structure, fuel fabrication and fuel handling, and reduction of fuel inventory external to the reactor. From the results of this preliminary study, it appears that the potential cost of electric power from this reactor is competitive with that from other reactor systems which have been proposed for the same application. The Po210 produced in the coolant presents a decontamination problem, but is also possibly a valuable by-producgt.
- 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.
- Second-cycle airox reprocessing and pellet refabricating of highly irradiated uranium dioxide
- "This report describes second-cycle postirradiation examination and AIROX reprocessing-refabricating of uranium dioxide irradiated to an additional 10,000 Mwd/MTU burnup."
- 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.
- 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.
- 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.
- 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.
- Multi-Channel Boiling Stability for Sodium Graphite Reactors
- Abstract: This report presents an analysis of coolant boiling in sodium graphite reactors.
- A Reversing Logarithmic DC Amplifier
- Purpose: Automatic recording equipment was designed for use with a high temperature Sykes experiment in which calorimetric measurements were to be made to temperatures approaching 2000* C. At such high temperatures, radiation becomes the dominant mechanism for heat transfer. The temperature differences which are used to determine the magnitude of this transfer no longer are directly proportional to it, but must be related by the Stefan-Boltzman law of radiation.
- Final Design of Sodium-Heated, Modular, Steam Generators for the SCTI
- Abstract: The following report covers the final design of the modular steam generators.
- 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.
- 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.
- 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.
- General Chemistry, Quarterly Progress Report, April-June 1954
- "General Chemistry investigations reported herein includes: (1) the Organic Coolant-Moderator Program, (2) investigations on zirconium hydride, and (3) analytical chemistry."
- Plutonium: Enriched OMR Cores
- Abstract: The influence of plutonium on the nuclear characteristics of organic moderated cores is studied.
- 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.
- OMR Degasifier Loop Experiment
- Abstract: A loop has been constructed to study the removal of water and highly volatile materials from Organic Moderated Reactor coolant by vacuum degasification. An analysis of the process was made to determine the most important parameters for study during the experimental program.
- Sodium Reactor Experiment Pump Development
- Design and operational techniques are described for a freeze seal type centifugal pump for use in the Sodium Reactor Experiment.
- 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.
- 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.
- 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.
- Large SGR Control Rod Development
- 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.
- SRE Mark II Fuel Handling Machine
- 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.
- 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.
- 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.
- 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.
- 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.
- 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."
- 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.
- Radiation Effects, Quarterly Progress Report, January-March 1954
- No Description Available.
- Radiation Effects, Quarterly Progress Report, July - September, 1953
- No Description Available.
- Radiation Effects, Quarterly Progress Report, October-December 1953
- No Description Available.
- Neutron Leakage from the 30 Megawatt SGR-P4 Reactor
- Abstract: The fast and thermal neutron leakage from the 30 megawatt SGR-P4 reactor has been studied by three independent methods.