Explore Results

open access

Subcooled burnout at high flows

Description: Task II-B of the High Performance UO2 Program calls for testing a series of UO2 fuel rods in the PW loop of the GETR. The purpose of the tests is to obtain a relationship between the extent of central melting and the length of time before the fuel rod fails as a result of interaction between the fuel and the cladding.
Date: unknown
Creator: Janssen, E. (Engineer) & Kervinen, J. A.
Partner: UNT Libraries Government Documents Department
open access

Design and Operating Experience of the ESADA Vallecitos Experimental Superheat Reactor (EVESR)

Description: Summary: "The various design features significant to superheat are described for the 12-1/2 MW., 960 psig, 1050° F, steam-cooled, low-enriched, annular-fueled, experimental superheat reactor built by the General Electric Company at the Vallecitos Atomic Power Laboratory. Results obtained during the first six months of full-power operation, on the emergency cooling system, core thermal performance, and pressure vessel temperatures are presented and compared with predictions. Operating experience… more
Date: February 1965
Creator: Ianni, P. W.; Fritz, J. R. & Law, D. D.
Partner: UNT Libraries Government Documents Department
open access

UO2 Pellet Thermal Conductivity From Irradiations With Central Melting

Description: Abstract: Continued irradiation experience under the AEC - Euratom, UO2 High Performance Program provided five separate and distinct sets of data on UO2 thermal conductivity. Four of these results are expressed in terms of the value of the thermal conductivity. The first two of these measurements were applicable -- strictly -- to poly crystalline UO2. Recently, three additional sets of measurements have been obtained -- all pertinent to UO2 after the formation of large columnar grains. The … more
Date: July 1964
Creator: Lyons, M. F.; Coplin, D. H.; Pashos, T. J. & Weidenbaum, B.
Partner: UNT Libraries Government Documents Department
open access

The Effects of Non-Uniform Flow and Concentration Distributions and the Effect of the Local Relative Velocity on the Average Volumetric Concentration in Two-Phase Flow

Description: Abstract: A general expression which can be used either for predicting the average volumetric concentration or for analyzing and interpreting experimental data is derived. The analysis takes into account both the effect of non-uniform flow and concentration profiles as well as the effect of the local relative velocity between phases. The first effect is taken into account by a distribution parameter, whereas the latter is accounted for by the weighted average drift velocity.
Date: April 1964
Creator: Zuber, N. & Findlay, J. A.
Partner: UNT Libraries Government Documents Department
open access

Fabrication of fuel Cladding From Incoloy Alloy 800 : an Evaluation of Methods

Description: Summary: On the basis of its high temperature, physical and corrosion properties, Incoloy Alloy 800 was selected as a candidate for fuel cladding nuclear superheat applications. At the time of its selection, there was little information or experience with Incoloy 800 in the production of thin-walled, small diameter tubing suitable for nuclear fuel cladding. As a result, special purchasing efforts were required for the procurement of initial tubing used in fuel fabrication. As-received welde… more
Date: April 1964
Creator: Kirby, R. F.; MacMillan, D. F. & Punches, J. R.
Partner: UNT Libraries Government Documents Department
open access

In-Core Instrumentation Development Program Quarterly Progress Report January - March 1964

Description: The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. This technical report discusses the following topics: low versus high cable termination impedance, amplifier considerations, noise considerations, gas and pressure selection, cable selection, effect of gamma, effect of temperature, and remaining problems.
Date: April 1964
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department
open access

Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Tenth Quarterly Report, January-March 1964

Description: Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, transient testing of fuel, fuel performance evaluation, fast-flux irradiation of fuel, and reactor physics and core analysis.
Date: April 1964
Creator: Breizy, C. E.
Partner: UNT Libraries Government Documents Department
open access

AEC Fuel Cycle Program Design and Fabrication of Special Assembly 9-L : Irradiation Performance Test of UO2-Cermet Fuel

Description: Technical report describing a UO2-Mo cermet fuel assembly fabricated for long-term irradiation performance testing in the Vallecitos Boiling water Reactor. The design and fabrication histories of this assembly are described and pre-irradiation data on each individual rod are presented. Molybdenum was added to improve the bulk thermal conductivity of the fuel, so that fuel temperatures would remain comparatively low during high-power level operation of the fuel element. The molybdenum was inc… more
Date: March 1964
Creator: Ogawa, S. Y.
Partner: UNT Libraries Government Documents Department
open access

General Corrosion of Incoloy-800 in Simulated Superheat Reactor Environment

Description: The 300 series stainless steels had been selected originally as the reference fuel cladding material for utilization in several superheat reactor (SHR) systems being built as part of the United States Atomic Energy Commission (USAEC) program. The adequacy of the general corrosion resistance of Type-304 stainless steel for superheat fuel cladding was confirmed in the Phase I portion of the study reported previously. Fuel jacket failures that occurred in Type-304 stainless clad fuel elements exp… more
Date: March 1964
Creator: Pearl, W. L.; Brush, E. G.; Gaul, G. G. & Wozadlo, G. P.
Partner: UNT Libraries Government Documents Department
open access

Reactor Pressure Vessel Material Surveillance Program at the Garigliano Nuclear Power Plant

Description: Abstract: A materials exposure program has been established in the Garigliano Nuclear Power Plant to measure the effect of neutron irradiation and time at temperate on the mechanical properties of the reactor pressure vessel steel. Base metal specimens were made from portions of the pressure vessel steel, and weld heat-affected zone and weld metal samples were taken from a weldment made from the pressure vessel steel and simulating a pressure vessel circumferential weld since there are no longi… more
Date: March 1964
Creator: Brandt, F. A. & Kobsa, I. R.
Partner: UNT Libraries Government Documents Department
open access

Transition and Film Boiling Data at 600, 1000, and 1400 PSIA in Forced Convection Heat Transfer to Water

Description: Summary: Data were obtained in a two-road test section which consisted of two 7/16-inch diameter heater rods inside a roughly rectangular flow area. The heated length of the rods was 30 inches, with a 15-inch unheated calming length preceding it. Heater wall temperatures were recorded while the heater tubes were trans-versing the critical heat flux and transition boiling; these temperatures were used to calculate heat transfer coefficients. The following general results were obtaining: (a)… more
Date: March 1964
Creator: Hench, J. E.
Partner: UNT Libraries Government Documents Department
open access

Design and Fabrication of Fuel Rods Containing Sintered UO2 Extrusions - Assembly 11L

Description: The extrusion forming of ceramic powders may be economically interesting in the field of nuclear fuel fabrication. When applied to the forming of rod-type uranium dioxide fuel, extrusion processes have been able to produce cylindrical bodies with length-to-diameter ratios much greater than those of the conventional die-pressed pellets. Furthermore, after being sintered, the extrusions have exhibited densities at least as high as those of sintered pellets. Thus, extrusion forming may offer re… more
Date: February 1964
Creator: Megerth, F. H.
Partner: UNT Libraries Government Documents Department
open access

Design and Fabrication of Pellet Fuel Rods Clad With Thin Wall Stainless Steel

Description: Summary: Stainless steel clad nuclear fuel cycle costs can be reduced to those associated with Zircaloy clad fuel or potentially lower by reducing the thickness of the clad tube wall until performance penalties offset the savings associated with the reduction in parasitic neutron absorption. To demonstrate the feasibility and investigate performance capabilities of thin clad fuel rods for power reactor application an assembly was fabricated with 0.0127 cm (5 mil) thick stainless steel cladding … more
Date: February 1964
Creator: Hoffmann, J. P.
Partner: UNT Libraries Government Documents Department
open access

Localized Corrosion of Stainless Steels and High-Nickel Alloys in Simulated Superheat Reactor Environment

Description: Abstract. A program was instituted to study and reproduce the in-reactor intergranular failures of Type-304 stainless steel fuel cladding found in superheated steam. The program was directed toward finding ways to eliminate the cause of failure or to use improved alloys that would be less susceptible to failure. A materials screening test was developed in the out-of-pile superheat facilities with 1.5 ppm chloride added as sodium chloride to the recirculating water in the presence of typical … more
Date: February 1964
Creator: Pearl, W. L.; Gaul, G. G. & Wozadlo, G. P.
Partner: UNT Libraries Government Documents Department
open access

Mechanical, Fluid Flow, and Heat Transfer Out-Of-Pile Tests on EVESR MKI Prototype Fuel Bundle

Description: Summary: An EVESR MKI prototype fuel bundle was fully instrumented and operated intermittently for a 5-month period at the Pacific Gas and Electric Company’s Moss Landing Power Station. The vessel was operated up to 1000 psi with steam flows from 3000 to 26,600 lb/h, and steam inlet temperatures up to 825 degrees F. Data was recorded for blowout, vibration, flow distribution, heat transfer and pressure drop. The mechanical integrity of the fuel bundle, riser, and jumper system was satisfact… more
Date: February 1964
Creator: Polomik, E. E.; Fritz, J. R. & Ianni, P. W.
Partner: UNT Libraries Government Documents Department
open access

Preoperational Power Stability Analysis of the Consumers Big Rock Point Plant

Description: Summary: An analytical study of the stability of the Big Rock Nuclear Reactor has been performed for the plant as built, and supplements a previous design stability study. The plant has been determined by this analysis to be very stable under every mode of operation anticipated during Phase I of the development program testing. Even under conservative assumptions of system parameters the minimum calculated gain and phase margins do not go below 13.0 db and 46 degrees, respectively. (Nor are … more
Date: February 1964
Creator: Case, J. M.
Partner: UNT Libraries Government Documents Department
open access

Sodium Mass Transfer. [Part] 8. Corrosion of Stainless Steel in Isothermal Regions of a Flowing Sodium System

Description: Technical report describing an analytical investigation made on the mechanism of the "downstream" effect in the corrosion of stainless steel in sodium. A mechanism of iron alloy corrosion is assumed in which the controlling rate is diffusion of iron-oxygen species, probably a FeO-Na2O complex. A mathematical model is developed for the corrosion, and the predicted results agree with the experimental data. The corroding species is probably present in sodium at concentrations of ~10(-8) g Fe/g Na. more
Date: February 1964
Creator: Mottley, J. D. & Epstein, Leo F.
Partner: UNT Libraries Government Documents Department
open access

Sodium Mass Transfer. [Part] XI. 1963 Test Run Reports (January - June)

Description: Technical report describing how corrosion data and exposure effects were obtained by subjecting metallic samples, during programmed test runs to flowing sodium in 6 test loops fabricated with various combinations of three selected materials, Type 316 stainless steel, 2 1/4 Cr-1 Mo alloy steel, and 5 Cr-1/2 Mo-1/2 Ti alloy steel. Information produced by each test run, including operational and metallurgical data and analyses, is presented. Data are shown in tables, graphs, and drawings.
Date: February 1964
Creator: Lockhart, R. W.
Partner: UNT Libraries Government Documents Department
open access

In-Core Instrumentation Development Program Quarterly Progress Report September - December 1963

Description: Introduction: The objective of Project Agreement 22 is to determine the feasibility of using in-core ion chambers to cover the complete reactor neutron flux startup range from 10(4) -5 - 10(13) nv using in-core ion chambers. The counting mode of operation will be used at low neutron flux levels and the RMS voltage fluctuation mode (Campbell Theorem) will be used at high neutron flux levels. The June-September Progress Report (GEAP-4386) shows how the RMS voltage mode can be used, discusses c… more
Date: January 1964
Creator: DuBridge, R. A.
Partner: UNT Libraries Government Documents Department
open access

Sodium-Cooled Reactors Program, Fast Ceramic Reactor Development Program: Ninth Quarterly Report, October-December 1963

Description: Quarterly report discussing progress on the Fast Ceramic Reactor Development Program. Information is reported on vented fuel production, fuel testing in TREAT, fuel performance evaluation, fast-flux irradiation of fuel, and reactor dynamics and design.
Date: January 1964
Creator: Leitz, F. J.
Partner: UNT Libraries Government Documents Department
Back to Top of Screen