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REACTOR PHYSICS QUARTERLY REPORT JANUARY, FEBRUARY, MARCH 1970

Description: The objective of the Reactor Physics Quarterly Report is to inform the scientific community in a timely manner of the technical progress made on the many phases of reactor physics work within the laboratory. The report contains brief technical discussions of accomplishments in all areas where significant progress has been made during the quarter.
Date: May 1, 1970
Creator: Schmid, L. C.; Clayton, E. D. & Heineman, R. E.
Partner: UNT Libraries Government Documents Department

STATISTICAL ANALYSIS AND MODELING OF GAP CONDUCTANCE DATA FOR REACTOR FUEL RODS CONTAINING UO2 PELLETS

Description: The purpose of this study was threefold: To examine a large body of in-reactor data for UO{sub 2} pellet fuel where internal temperatures have been measured, and to estimate fuel surface temperature, cladding inner temperature, and thus gap conductance under a consistent set of assumptions. To estimate the variance (due to experimental uncertainties) in the inferred gap conductance values. To attempt a correlation of the gap conductance values with design parameters and operating conditions (e.g., with cold diametral gap and local linear power).
Date: May 1, 1974
Creator: Lanning, D. D.; Hann, C. R. & Gilbert, E. S.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The schedule for the installation of the PAR slurry loop experiment in the ·South Facility af the ORR has been reviewed and revised. Changed design philosophy to include maintainability of the loop auxiliaries required extension of the main construction period to the end of,June, with a three month period ending in September, for test and run in during which construction will be completed.
Date: February 28, 1958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrication, and installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 56 per cent complete, fabrication 27 per cent complete-and installation 11 per cent complete:
Date: March 31, 0958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The status of the overall design, fabrication, and installation of the component items of the PAR loop experiment in the ORR is reviewed.
Date: January 15, 1958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The schedule for the installation of the PAR slurry loop experiment in the South Facility of the ORR has been reviewed and revised. The design, fabrications and Installation is approximately two weeks behind schedule at this time due to many factors; however, indications are that this time can be made up. Design is estimated to be 75% complete, fabrication 32% complete and installation 12% complete.
Date: April 30, 1958
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

PAR Loop Schedule Review

Description: The status of the overall schedule for design, fabrication, and installation of the component items of the PAR loop experiment in the ORR has been brought up to date and reviewed. This is the first review since the original issue of the schedule
Date: October 16, 1957
Creator: Schaffer, Jr. & W.F.
Partner: UNT Libraries Government Documents Department

PARAMETRIC REACTIVITY TRANSIENT ANALYSES FOR THE FFTF NUCLEAR PROOF TEST REACTOR

Description: Fault tree techniques have been used to identify possible failure paths within the NPTR which could lead to core disassembly. The analysis o f the various faults has led to formulation of design requirements, protective system requirements, and administrative restraints required to prevent accidents from these faults. Transient analyses were performed using the heat transfer-nuclear kinetics codes, Nutiger-II, FORE-II, and MELT-II . To verify results, intercomparison studies were made between the codes. The codes were i n good general agreement. Each code was found to exhibit different advantages and disadvantage. Inherent reactivity feedback effects were assessed in the analysis. With the assumed core parameters, there appears to be sufficient Doppler to prolong a nuclear transient to allow protective action to prevent fuel from melting. The use of average values of the feedback coefficients smeared over the entire core does not appear to be an acceptable method with spacially dependent temperatures. In the thermal analysis, the fuel pin gap coefficient and sodium film coefficient do not appear to be highly sensitive parameters for transient analysis. Power transients resulting from reactivity insertions of from 2$/sec to 20$/sec have been examined in detail. Sodium will be molten before fuel melting occurs for accidents within this range. For the smaller ramp rates (< 4$/sec), sodium nay even reach vaporization temperatures before any fuel melts. Power transients terminated by effective protective action were investigated. It i s believed p o s s i b l e t o design a scram system, with the . present state of the art, to prevent sodium from melting for a reactivity ramp up to at least 6$/sec. This same system would prevent fuel melting for a reactivity ramp up to 15$/sec. Sodium thermal expansion will play a very important role in a core disassembly. When the average sodium ...
Date: January 1, 1970
Creator: Schade, A. R.
Partner: UNT Libraries Government Documents Department

POSTIRRADIATION EXAMINATION OF CENTRIFUGALLY BONDED EBR-11 DRIVER FUEL

Description: BNW was requested by AEC/DRDT to assist ANL with the postirradiation examination of EBR-II driver fuel (U + 5Fs) pins that became shortened during irradiation. To date, three irradiated pins and one nonirradiated pin were examined.
Date: July 31, 1969
Creator: Leggett, R. D.; Hann, D. R.; Last, G. A. & Gruber, W. J.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL COMPONENT DESIGN DISCRIPTION FOR THE REACTOR NUCLEAR CONTROL COMPONENTS No. 33

Description: This document describes the nuclear control component of the Fast Test Reactor (FTR) . The Fast Test Reactor is t he central test bed of the Fast Flux Test Facility (FFTF). The Reactor Nuclear Control Components described in this document consist of neutron absorbing elements which are moved axially into and out of the general area of the active portion of the core. The electromechanical drive housings are mounted on the reactor cover and are part of the pressure boundary of the vessel and cover during reactor operation. Drive lines (extension shafts) extend through the cover and are connected to the poison element. The element is guided in the core throughout its travel by a flow duct which resides in the c or e grid plate receptacle. Sodium coolant from the primary loop is used to cool the neutron absorbing rods. This Conceptual Component Design Description (CCDD) suppor t s and expands the requirements of the Overall Conceptual Systems Design Description . Specifically , it establishes the reactor nuclear control design requirements; describes the current reference design of this control; defines safety, operational · and maintenance principles; indicates unresolved problems ; and identifies the principal interfaces affecting the reactor control component functional requirements and design configuration. Included are functions and design requirements, a physical description of the system, safety considerations, principles of operation, and maintenance principles.
Date: October 1, 1969
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL FACILTY DESIGN DESCRIPTION FOR THE INERT GAS CELL EXAMINATION FACILITY NO. 71

Description: The purpose of this Conceptual Facility Design Description (CFDD) is to provide a technical description of the Inert Gas Cell Examination Facility such that agreement with RDT on a Conceptual Design can be reached . The CFDD also serves to establish a common understanding of the facility concept among all responsible FFTF Project parties including the Architect Engineer and Reactor Designer. Included are functions and design requirements, a physical description of the facility, safety considerations, principles of operation, and maintenance principles.
Date: December 12, 1968
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL SYSTEM DESIGN DESCRIPTION FOR THE BUILDING ELECTRICAL POWER SYSTEM No. 12

Description: The Building Electrical Power System, transmits, distributes and controls the electrical power utilized by the equipment of all of the FFTF system. Included are system functions and design requirements, the physical description of the system, safety considerations, principles of operations, and maintenance principles.
Date: July 3, 1968
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL SYSTEM DESIGN DESCRIPTION FOR THE HEATING AND VENTILATION SYSTEM NO. 25

Description: The Heating and Ventilation System provides the atmospheric conditions in FFTF as required to assure that plant objectives can be met in a safe and practical manner. Included are functions and design requirements, a physical description of the system, safety considerations, principles of operation, and maintenance principles.
Date: December 5, 1969
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL SYSTEM DESIGN DESCRIPTION FOR THE PLANT FIRE PROTECTION SYSTEM No. 26

Description: The fire protection system for the FFTF is required for the safety of personnel and the protection of property. Included are functions and design requirements, a detailed description of the system, safety considerations, principles of operation, and maintenance principles.
Date: August 20, 1968
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY CONCEPTUAL SYSTEM DESIGN DESCRIPTION FOR THE PRIMARY ELECTRICAL POWER SYSTEM NO. 11

Description: The Primary Electrical Power System, provides the electrical service from the electric utility system to the FFTF site. Functions and design requirements are given, followed by the Conceptual Systems Design Description. Included are the physical description, safety considerations, principles of operation, and maintenance.
Date: January 1, 1968
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY DEVELOPMENT MONTHLY PROGRESS REPORT

Description: Measurements of the reaction of methane with sodium have been extended to 1400°F, where 100 ppm in helium completely reacted in 3-1/2 hours. It was also determined that the volume of sodium in the system has little effect on the reaction rate, An important milestone was achieved in the development of electrical heaters for simulation of fuel pins for use in future heat transfer studies on FFTF fuel assembly models. In a demonstration test, a tantalum coaxial heater was operated continuously for six hours at a heat flux of 1,000,000 Btu/hr-ft{sup 2} in water at 320°F outlet temperature, The heater then operated at 1,300,000 Btu/hr-ft{sup 2} for an additional 4-1/2 hours before the test was terminated due to electrode failure. The fabrication of the test facility required to demonstrate and establish the feasibility of gas cooling of FFTF fuel elements during inspection and disassembly was completed. A preliminary study was made to determine the feasibility of utilizing facilities such as the Hydraulic Core Mockup, the Seven-Duct Mockup, tile Reactor Cover Mockup, and the Sodium Facilities Building for Fuel Handling Machine and Radioactive Maintenance equipment interface testing. With modification, each or all of the above facilities could be applied in the program. Final PNL recommendations will be made in February. Based on a summary of stainless steel volume expansion data at exposures to 8 x 10{sup 22} nvt, it appears that as mucn as 10% volume increase may occur at 10{sup 23} nvt and irradiation temperatures of 800°F to 1100°F. A preliminary assessment, from computer simulations, of radiation damage processes yields the following ranking in relative damaging power for void production in fuel cladding: EBR-II > DFR > FTR. This ranking is based on the number of primary knock-on atoms (PKA) with energy greater than 3 keV produced per interaction with ...
Date: January 1, 1968
Creator: Woodfield,, F. W.
Partner: UNT Libraries Government Documents Department

FAST FLUX TEST FACILITY MONTHLY INFORMAL TECHNICAL PROGRESS REPORT AUGUST 1969

Description: This report was prepared by Battelle-Northwest under Contract No. AT(45-1)-1830 for the Atomic·Energy Commission, Division of Reactor Development and Technology, to summarize technical progress made in the Fast Flux Test Facility Program during August 1969 .
Date: September 8, 1969
Creator: Astley, E. R. & Cabell, C. P.
Partner: UNT Libraries Government Documents Department