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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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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).
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).
Reference Design for an OMR-Powered 38,000-DWT Tanker
Abstract: This report presents a reference design of an organic moderated and cooled reactor for the propulsion of a 38,000-dwt tanker.
Study of SCTI Control System
Introduction: This report has been prepared to document the extensive analytical work required to design the control system for the Sodium Components Test Installation (SCTI), constructed for the Atomic Energy Commission at the Atomics International field laboratory.
Refabrication and Encapsulation of Highly Irradiated Uranium Dioxide
Abstract: One hundred gram quantities of uranium dioxide, irradiated to burnup as great as 21,000 Mwd/MTU and previously reprocessed by AIROX (Atomics International Reduction Oxidation), were refabricated into high density pellets and encapsulated for re-irradiation.
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.