Plutonium Spike Fuel Elements for the Plutonium Recycle Test Reactor: Part 2 - The Mark 1-H

Plutonium Spike Fuel Elements for the Plutonium Recycle Test Reactor: Part 2 - The Mark 1-H

Date: March 1961
Creator: Sharp, R. E.
Description: Report describing the fabrication of of the I-H aluminum-plutonium spike enrichment fuel elements for the Plutonium Recycle Test Reactor (PRTR), including the materials used and design specifications. Appendix begins on page 28.
Contributing Partner: UNT Libraries Government Documents Department
Evaluation of an Engineering Demonstration of the Modified Zirflex and Neuflex Processes for the Preparation of Solvent Extraction Feeds from Unirradiated Zirconium-Base Reactor Fuels

Evaluation of an Engineering Demonstration of the Modified Zirflex and Neuflex Processes for the Preparation of Solvent Extraction Feeds from Unirradiated Zirconium-Base Reactor Fuels

Date: March 1964
Creator: Kitts, F. G.
Description: Report that investigates the concept of using ammonium fluoride-oxidant systems to dissolve spent U-Zr-Sn reactor fuels in both batch and continuous operation.
Contributing Partner: UNT Libraries Government Documents Department
MINIMIZER: A Computer Code for Determining Minimum Fuel Cost

MINIMIZER: A Computer Code for Determining Minimum Fuel Cost

Date: July 1965
Creator: Eschbach, E. A.; Deonigi, D. E. & McConiga, A. F.
Description: Report that describes the Hanford Laboratories MINIMIZER code and demonstrates its accuracy in providing fissile fuel costs in a precise and automatic manner.
Contributing Partner: UNT Libraries Government Documents Department
Progress in Nondestructive Testing: A Summary of Hanford Achievements in These Programs Under General Electric, 1952 - 1964

Progress in Nondestructive Testing: A Summary of Hanford Achievements in These Programs Under General Electric, 1952 - 1964

Date: August 1964
Creator: Walker, R. A. & Russell, J. T.
Description: Report discussing the development of Hanford Laboratories' nondestructive testing programs and a summary of their results. Programs include AlSi fuel testing, N-Reactor fuel testing, and reactor research and development.
Contributing Partner: UNT Libraries Government Documents Department
An Alpha Scintillation Tester for Uranium Surface Contamination of N-Reactor Fuel

An Alpha Scintillation Tester for Uranium Surface Contamination of N-Reactor Fuel

Date: July 1963
Creator: Jackson, C. N., Jr.
Description: 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).
Contributing Partner: UNT Libraries Government Documents Department
Pressure-Pulse Propagation in Two-Phase One- and Two-Component Mixtures

Pressure-Pulse Propagation in Two-Phase One- and Two-Component Mixtures

Date: March 1971
Creator: Henry, R. E.; Grolmes, M. A. & Fauske, H. K.
Description: Report issued by the Argonne National Laboratory discussing the propagation velocity of pressure pulses in two-phase mixtures. As stated in the abstract, "in this study, a comprehensive analytical development of the propagation velocity of small pressure pulses is presented for bubble, annular, stratified, droplet, and slug-flow regimes. Particular attention is given to the influence of flow regime on the momentum transfer between phases" (p. 9). This report includes tables, illustrations, and a photograph.
Contributing Partner: UNT Libraries Government Documents Department
Heat Transfer in an Annulus with Asymmetric Heating

Heat Transfer in an Annulus with Asymmetric Heating

Date: February 1959
Creator: Lisin, Alexander Vladimir; Mackewicz, W. B. & Reynolds, William C., 1933-2004
Description: Report concerning the "determination of the validity of using tube flow correlations for calculating pressure drop and heat transfer characteristics in thin annuli," the "validity of using Stein's heat flux asymmetry correction for parallel planes to calculate the heat transfer in asymmetrically heated thin annuli," and the "investigation of the effects of eccentricity on pressure drop, heat transfer, and temperature distribution in concentric ring fuel elements" (p. 5).
Contributing Partner: UNT Libraries Government Documents Department
Casting and Fabrication of Core Material For Argonne Low Power Reactor Fuel Elements

Casting and Fabrication of Core Material For Argonne Low Power Reactor Fuel Elements

Date: December 1959
Creator: Salley, R. L. & Burt, W. R., Jr.
Description: Report issued by the Argonne National Laboratory over studies conducted on the Argonne Low Power Reactor. As stated in the abstract, "this report describes the manufacture of 1150 fuel core blanks, of which 816 were used in the fabrication of fuel elements. Development, casting, hot and cold rolling, cleaning, and punching of core blanks are discussed, as are nondestructive testing and evaluation of the manufacturing processes" (p. 7). This report includes tables, illustrations, and photographs.
Contributing Partner: UNT Libraries Government Documents Department
Engineering development of fluid-bed fluoride volatility processes

Engineering development of fluid-bed fluoride volatility processes

Date: 1964
Creator: Vogel, G. J.
Description: None
Contributing Partner: UNT Libraries Government Documents Department
Analysis of Failure of Type 304 Stainless Steel Clad Swaged Powder Fuel Assembly

Analysis of Failure of Type 304 Stainless Steel Clad Swaged Powder Fuel Assembly

Date: October 3, 1963
Creator: Lees, E. A.
Description: From introduction: "The purpose of this report is to describe the observations made during the post-irradiation examination of HPD-2S, and to discuss possible modes of failure.
Contributing Partner: UNT Libraries Government Documents Department
High Power Density Development Project: Interim Report, 300 MWe HPD Conceptual Design Study

High Power Density Development Project: Interim Report, 300 MWe HPD Conceptual Design Study

Date: January 5, 1962
Creator: Grayhek, V. G.
Description: From introduction: "Preliminary design and analysis of the 300 MWe core."
Contributing Partner: UNT Libraries Government Documents Department
PRTR second generation shim assembly

PRTR second generation shim assembly

Date: November 1964
Creator: Rasmussen, D. E.
Description: From introduction: "This document discusses the design, fabrication, assembly, and testing of a second generation shim rod assembly, and testing of a second generation shim rod assembly prototype for use in the Plutonium Recycle Test Reactor."
Contributing Partner: UNT Libraries Government Documents Department
Standardization and Evaluation of Grain Size Test for Uranium Fuel

Standardization and Evaluation of Grain Size Test for Uranium Fuel

Date: July 1964
Creator: Silker, W. B. & Thomas, C. W.
Description: Report that "describes the selection of "standard fuel cores for calibrating an ultrasonic tester designated UT-2C" (p. 3) at Hanford Laboratories.
Contributing Partner: UNT Libraries Government Documents Department
Plutonium Spike Fuel Elements for the Plutonium Recycle Test Reactor: Part 1 - The Mark 1-G

Plutonium Spike Fuel Elements for the Plutonium Recycle Test Reactor: Part 1 - The Mark 1-G

Date: March 1961
Creator: Freshley, M. D.
Description: Report describing "[t]he fabrication of of the first aluminum-plutonium spike enrichment fuel elements for the Plutonium Recycle Test Reactor (PRTR) at Hanford" Laboratory (p. 2).
Contributing Partner: UNT Libraries Government Documents Department
Plutonium Recycle Critical Facility: Final Safeguards Analysis Supplements

Plutonium Recycle Critical Facility: Final Safeguards Analysis Supplements

Date: October 1964
Creator: Swanberg, F.
Description: Report containing supplementary material for "Plutonium Recycle Critical Facility, Final Safeguards Analyses." This material expands upon previously supplied safety information and resolves previously unreviewed safety questions. Appendix begins on page A.1.
Contributing Partner: UNT Libraries Government Documents Department
Fast Reactor Core Design Parameter Study

Fast Reactor Core Design Parameter Study

Date: March 1960
Creator: Atomic Power Development Associates
Description: Report describing parametric studies of eleven fast reactor fuel systems undertaken to determine the design and economic factors for producing electricity. The methods used for making the parametric studies are described, as well as the results of these studies. Appendices begin on page 103.
Contributing Partner: UNT Libraries Government Documents Department
The HGCR-1, a Design Study of a Nuclear Power Station Employing a High-Temperature Gas-Cooled Reactor with Graphite-UO₂ Fuel Elements

The HGCR-1, a Design Study of a Nuclear Power Station Employing a High-Temperature Gas-Cooled Reactor with Graphite-UO₂ Fuel Elements

Date: 1959
Creator: Cottrell, William B.; Copenhaver, C. M.; Culver, H. N.; Fontana, M. H.; Kelleghan, V. J. & Samuels, G.
Description: "The preliminary design of a 3095-Mw(thermal), helium-cooled, graphite- moderated reactor employing sign conditions, 1500 deg F reactor outlet gas would be circulated to eight steam generators to produce 1050 deg F, 1450-psi steam which would be converted to electrical power in eight 157-Mw(electrical) turbine- generators. The over-all efficiency of this nuclear power station is 36.5%. The significant activities released from the unclad graphite-UO₂ fuel appear to be less than 0.2% of those produced and would be equivalent to 0.002 curie/ cm/ sup 3/ in the primary helium circuit. The maintenance problems associated with this contamination level are discussed. A cost analysis indicates that the capital cost of this nuclear station per electrical kilowatt would be around 0, and that the production cost of electrical power would be 7.8 mills/kwhr" (p. iii).
Contributing Partner: UNT Libraries Government Documents Department
High Power Density Development Project: Third Quarterly Progress Report, October - December, 1960

High Power Density Development Project: Third Quarterly Progress Report, October - December, 1960

Date: January 1, 1961
Creator: Holland, L. K.
Description: From introduction: "Development of nuclear reactor cores having high power density, long fuel life, and low fabrication costs is the objective of this program sponsored by the AEC."
Contributing Partner: UNT Libraries Government Documents Department
Evaluation of Thorium Fuels for the Heavy Water Organic Cooled Reactor

Evaluation of Thorium Fuels for the Heavy Water Organic Cooled Reactor

Date: July 1966
Creator: unknown
Description: From abstract: "A summary of work done by The Babcock & Wilcox Company to evaluate thorium fuels for the Heavy Water Organic Cooled Reactor (HWOCR)."
Contributing Partner: UNT Libraries Government Documents Department
An Analytical Study of the Feasibility of Irradiating U233/Th232 Metal Fuel Experiments in EBR-II

An Analytical Study of the Feasibility of Irradiating U233/Th232 Metal Fuel Experiments in EBR-II

Date: 1979?
Creator: Foltman, A. J. & Meneghetti, D.
Description: Recent concerns about the proliferation and diversion of plutonium have lead to reconsideration of Uranium-233/Thorium-232 fuel cycles. Although thorium fuels have been studied earlier, much of that work is incomplete; consequently, additional irradiation studies will be necessary.
Contributing Partner: UNT Libraries Government Documents Department
Monte Carlo-Based Validation of the ENDF/MC²-II/SDX Cell Homogenization Path

Monte Carlo-Based Validation of the ENDF/MC²-II/SDX Cell Homogenization Path

Date: April 1979
Creator: Wade, D. C.
Description: The results are presented of a program of validation of the unit cell homogenization prescriptions and codes used for the analysis of Zero Power Reactor (ZPR) fast breeder reactor critical experiments. The ZPR drawer loading patterns comprise both plate type and pin-calandria type unit cells. A prescription is used to convert the three dimensional physical geometry of the drawer loadings into one dimensional calculational models. The ETOE-II/MC²-II/SDX code sequence is used to transform ENDF/B basic nuclear data into unit cell average broad group cross sections based on the 1D models. Cell average, broad group anisotropic diffusion coefficients are generated using the methods of Benoist or of Gelbard. The resulting broad (approx. 10 to 30) group parameters are used in multigroup diffusion and S(su n) transport calculations of full core XY or RZ models which employ smeared atom densities to represent the contents of the unit cells.
Contributing Partner: UNT Libraries Government Documents Department
Fuel Cycle Programs, Quarterly Progress Report: July-September 1978

Fuel Cycle Programs, Quarterly Progress Report: July-September 1978

Date: January 1980
Creator: Steindler, M. J.; Ader, M.; Bernstein, G.; Flynn, K.; Gerding, T.; Jardine, L. J. et al.
Description: Quarterly report of the Argonne National Laboratory Chemical Engineering Division regarding activities related to properties and handling of radioactive materials, operation of nuclear reactors, and other relevant research. Fuel cycle work reported for this period includes testing of hydraulic performance and extraction efficiency of eight-stage centrifugal contactors, testing of a flowsheet for the Aralex process, evaluation of ruthenium and zirconium extraction in a miniature centrifugal contactor, study of zirconium aging in the organic phase and its effect on zirconium extraction and hydraulic testing of the 9-cm-ID contactor.
Contributing Partner: UNT Libraries Government Documents Department
The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

The Status and Development Potential of Plate-Type Fuels for Research and Test Reactors

Date: March 1979
Creator: Stahl, D.
Description: A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments.
Contributing Partner: UNT Libraries Government Documents Department
Leaching Action of EJ-13 Water on Unirradiated UO₂ Surfaces under Unsaturated Conditions at 90 C : Interim Report

Leaching Action of EJ-13 Water on Unirradiated UO₂ Surfaces under Unsaturated Conditions at 90 C : Interim Report

Date: July 1991
Creator: Wronkiewicz, D. J.; Bates, John K.; Gerding, Thomas J.; Veleckis, Ewald; Tani, B. & Hoh, J. C.
Description: A set of experiments, based on the application of the Unsaturated Test method to the reaction of uranium dioxide with EJ-13 water, has been conducted over a period of 182.5 weeks. One half of the experiments have been terminated, while one half are still ongoing. Solutions that have dripped from uranium dioxide specimens have been analyzed for all experiments, while the reacted uranium dioxide surfaces have been examined for only the terminated experiments. A pulse of uranium release from the uranium dioxide solid, in conjunction with the formation of dehydrated schoepite on the surface of the uranium dioxide, was observed during the 39- to 96-week period. Thereafter, the uranium release decreased and a second set of secondary phases was observed. The latter phases incorporate cations from the EJ-13 water and include boltwoodite, uranophane, sklodowskite, compreignacite, and schoepite. The experiments are being continued to monitor for additional changes in solution composition and secondary phase formation, and have now reached the 319-week period.
Contributing Partner: UNT Libraries Government Documents Department
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