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Experimental Study of the Neutronics of the First Gas Cooled Fast Reactor Benchmark Assembly (GCFR Phase 1 Assembly)

Description: The Gas Cooled Fast Reactor (GCFR) Phase I Assembly is the first in a series of ZPR-9 critical assemblies designed to provide a reference set of reactor physics measurements in support of the 300 MW(e) GCFR Demonstration Plant designed by General Atomic Company. The Phase I Assembly was the first complete mockup of a GCFR core ever built. A set of basic reactor physics measurements were performed in the assembly to characterize the neutronics of the assembly and assess the impact of the neutron… more
Date: December 1976
Creator: Bhattacharyya, S. K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Design and Fabrication of the EBR-II Environmental Instrumented Subassembly : Test XX07

Description: The EBR-II environmental instrumented subassembly (EISA), or Test XX07, was designed, fabricated, and irradiated to provide in-core measurements of irradiation conditions. This subassembly contained 57 elements with uranium metal-5 wt.% fissium fuel and four non-fueled elements. It was instrumented with two permanent-magnet flowmeters, 13 coolant thermocouples at various axial and radial locations, 10fuel-centerline thermocouples, and two self-powered detectors. The subassembly was irradiated i… more
Date: 1977
Creator: Gillette, J. L.; Ploncsik, J.; Smaardyk, A.; Walker, D. E.; Filewicz, E. C.; Longnecker, A. A. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Studies of Axial-Leakage Simulations for Homogeneous and Heterogeneous EBR-II Core Configurations

Description: When calculations of flux are done in less than three dimensions, leakage-absorption cross sections are normally used to model leakages (flows) in the dimensions for which the flux is not calculated. Since the neutron flux is axially dependent, the leakages, and hence the leakage-absorption cross sections, are also axially dependent. Therefore, to obtain axial flux profiles (or reaction rates) for individual subassemblies, an XY-geometry calculation delineating each subassembly has to be done a… more
Date: August 1985
Creator: Grimm, K. N. & Meneghetti, D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Radiative Cooling of a Voided Subassembly

Description: A treatment is formulated for surface-to-surface radiative heat exchange between fuel pins and between pins and duct wall of a nuclear reactor subassembly voided of coolant. Specific attention is given to the case of equal power generation in each pin with uniform duct-wall temperature. Detailed temperature profiles and heat flux values are reported for hexagonal-ring subassemblies ranging in size from one to nine rings. It is found that a duct wall at 1153 degrees K can cool by radiation even … more
Date: 1976
Creator: Chan, S. H.; Condiff, D. W. & Grolmes, M. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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TRANS4 : a Computer Code Calculation of Solid Fuel Penetration of a Concrete Barrier

Description: The computer code, TRANS4, models the melting and penetration of a solid barrier by a solid disc of fuel following a core disruptive accident. This computer code has been used to model fuel debris penetration of basalt, limestone concrete, basaltic concrete, and magnetite concrete. Sensitivity studies were performed to assess the importance of various properties on the rate of penetration. Comparisons were made with results from the GROWS II code.
Date: July 1980
Creator: Ono, C. M.; Kumar, R. & Fink, J. K.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Summary of Treat Experiments on Oxide Core-Disruptive Accidents

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… more
Date: February 1979
Creator: Dickerman, Charles Edward; Rothman, Alan B.; Klickman, A. E.; Spencer, B. W. & DeVolpi, Alexander
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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EBR-II Environmental Instrumented Subassembly XX08 : Engineering and Assembly

Description: Subassembly XX08 is a fueled and instrumented subassembly designed primarily for an ongoing program to investigate the thermal-hydraulic core environment within EBR-II under normal and off-normal plant operating conditions. XX08 contains 58-xenon-tagged, EBR-II Mark-II driver-fuel elements. The Mark-II fuel is expected to provide XX08 with an irradiation lifetime three times as great as that attained with its predecessor, XX07, i.e., a 9 versus 2.9% burnup. A burnup of 9 at.% is equivalent to a… more
Date: May 1978
Creator: Smaardyk, A.; Filewicz, E. C.; Longnecker, A. A.; Poloncsik, J.; Tokar, J. V.; Walker, D. E. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Comparisons of Finite-Element Code Calculations to Hydrostatically Loaded Subassembly-Duct Experiments

Description: The Liquid Metal Fast Breeder Reactor (LMFBR) core structure consists of a matrix of hexagonal subassembly ducts. Evaluation of the safety aspects of the core structure requires that reliable computational procedures be available to predict the deformation response of the subassembly configuration to postulated local energy releases. Finite-element computer codes have been developed to calculate deflections and strains of a hexcan subassembly wrapper subjected to internal and external dynamic p… more
Date: January 1977
Creator: Ash, J. E. & Marciniak, T. J.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Calandria Core Weld Joint Development

Description: 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.
Date: June 30, 1965
Creator: Roberts, J. G.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Thermal Analysis of Type 3 Elements in the SM-1, SM-1A and PM-2A Cores

Description: Report describing a study which analyzed thermal characteristics of Type 3 elements planned for installation in the SM-1, SM-1A and PM-2A plants, for both steady state and loss of flow transient conditions. Results of this analysis for steady state conditions indicate that the SM-1, SM-1A and PM-2A Type 3 cores will operate safely at design and scram conditions.
Date: March 30, 1962
Creator: Davidson, S. L. & Segalman, I.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Scale-Up of Alternate HRT Core

Description: "In order to determine the factors involved in the scale-up of cores with concentric inlet and outlet pipes, a 48 inch carbon steel flow model, geometrically similar to a 6 foot diameter core, has been assembled and tested...Visual studies were made of dye and gas behavior in the sphere, and quantitative measurements of point residence times were obtained through the use of conductivity cells actuating a Brush recorder. Static pressure drop across the core was measured."
Date: May 7, 1954
Creator: Lesem, L. B. & Harley, P. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Vapor-Explosion Experiments with Subcooled Freon

Description: Vapor-explosion experiments were conducted in a well-wetted Freon-22 and mineral-oil system in which the initial temperature of both the Freon and the mineral oil were varied over a wide range. These experiments were specifically conducted to investigate the importance of interface temperature in determining the explosive behavior of a given system. The results clearly demonstrate that the interface temperature developed upon intimate liquid-liquid contact is a valid characterization of the exp… more
Date: June 1977
Creator: Henry, Robert E. & McUmber, Louis M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Reactor Physics Studies in the GCFR Phase III Critical Assembly

Description: The third phase of the gas cooled fast reactor (GCFR) program, ZPR-9 Assembly 30, is based on a multi-zoned core of PuO2-UO2 with radial and axial blankets of uranium dioxide. Studies performed in this assembly will be compared to the previous phases of the GCFR program and will help to define parameters in this power-flattened demonstration plant-type core. Measurements in the Phase III program included small sample reactivity worths of various materials, central reaction rates and reaction ra… more
Date: March 1980
Creator: Argonne National Laboratory. Applied Physics Division.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Evaluation of Reactor Core Materials for a Gas-Cooled Reactor Experiment

Description: From introduction: "On February 1, 1956, Batelle was awarded a contract by the Army Reactor Branch (ARB) to select, develop, and test core materials which could be used successfully in conducting a Gas Cooled Reactor Experiment (GCRE). The prime objective of the GCRE would be to evaluate small portable reactor systems for military application...The present report is concerned with the GCRE activities at Batelle during approximately the 7 months' period following the first report of this series,… more
Date: July 11, 1957
Creator: Keller, Donald L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Canning graphite for gas-cooled reactors

Description: From abstract: "A preliminary investigation was made of techniques and materials for canning graphite to protect it for use at high temperatures in a nitrogen-oxygen atmosphere"
Date: 1959
Creator: Paprocki, Stan J.; Carlson, Ronald J. & Bonnell, Paul H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Flux Distributions and Critical k [infinity] Values for Cylindrical Reactors Simulation the MTR

Description: From summary: "In order to estimate the effect of placing bismuth and thorium layers near the MTR core, radial flux distributions and k[infinity] values were calculated for cylindrical reactors having cylindrical layers of these materials of the same height as the core and beryllium reflector and placed either next to the core or embedded in the beryllium reflector."
Date: June 25, 1952
Creator: Bray, D. T. & McMurry, Henry Lewis, {}
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Theoretical Calculations on ZPR Experimentation: Part 2. The Calibration of the Single Central Cross Control Rod

Description: From Abstract: "The report is concerned with a theoretical analysis of the experiment known as the Calibration of the 3" Cross Control Rod. The experiment provided 29 points on a curve of reactivity (in hours per centimeter) versus depth of control rod insertion. An attempt to reproduce this curve theoretically has been quite successful."
Date: January 18, 1951
Creator: Garabedian, Henry L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Results of Experiment 1: FCE Calibration With BORAX Core

Description: Summary: The justification of using polyethylene whose hydrogen density of 0.132 gm/cm³ with a distributed void of 15.9 percent as a mockup of water at 70°F and having a hydrogen density of 0.111 gm/cm³ was tested in the FCE. A mockup close to the BORAX core was built and its critical mass determined. Corrections were calculated for differences in the hydrogen desnity and self shielding of the fuel. The effective FCE critical mass agreed with that of the BORAX core to within one percent.
Date: October 1, 1956
Creator: Starr, E. & Toops, Edward Chassell
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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The Variation in Maximum Power Density and Maximum Heat Flux of the Core and Blanket Thorium Plates in the Converter Reactor with Irradiation Time

Description: This report describes how to achieve the variation in maximum power density and maximum heat flux of the core and blanket thorium plates in the converter reactor with irradiation time.
Date: June 23, 1952
Creator: Sletten, H. L.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Interim Report of Nuclear Analysis Performed on SM-2 Core and Vessel : September 1, 1958 to December 31, 1959.

Description: Abstract: This technical report contains a description of the nuclear analysis performed upon the SM-2 core and vessel for the period September 1, 1958 to December 31, 1959. Calculations are given for core reactivity, power distributions, lifetime, reactor control, kinetics, radiation problems, fuel and poison burn-ups, and the nuclear effects of poisons, temperature, and geometry. Wherever possible, experimental data is included in order to test the validity of the analytical models. The SM-… more
Date: May 27, 1960
Creator: Bobe, P. E.; Birken, S. H.; Byrne, B. J.; Clancy, E. F. & Fried, B. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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