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Study on Flow Instabilities in Two-Phase Mixtures

Description: Various mechanisms that can induce flow instabilities in two-phase flow systems are reviewed and their relative importance discussed. In view of their practical importance, the density-wave instabilities have been analyzed in detail based on the one-dimensional two-phase flow formulation. The dynamic response of the system to the inlet flow perturbations has been derived from the model; thus the characteristic equation that predicts the onset of instabilities has been obtained. The effects of v… more
Date: March 1976
Creator: Ishii, M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Studies of Unprotected Loss-of-Flow Accidents for the Clinch River Breeder Reactor

Description: Studies of unprotected loss-of-flow accidents in the CRBR for various rates of flow coast-down and with various options in the SAS 3A code did not lead to conditions for a violent disassembly. Maximum fuel temperatures using the SLUMPY module for disassembly were in the range 4000-4500 deg C. An approximate treatment of the LOF-driven TOP accident, not properly modeled by SAS 3A, indicates the possibility of some increase in accident severity. The effect of fission gas in dispersing fuel was no… more
Date: April 1976
Creator: Hummel, Harry H.; Pizzica, P. A. & Kalimullah
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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A Summary on Experimental Methods for Statistical Transient Analysis of Two-Phase Gas-Liquid Flow

Description: Much work has been done in the study of two-phase gas-liquid flows. Although it has been recognized superficially that such flows are not homogeneous in general, little attention has been paid to the inherent discreteness of the two-phase systems. Only relatively recently have fluctuating characteristics of two-phase flows been studied in detail. As a result, new experimental devices and techniques have been developed for use in measuring quantities previously ignored. This report reviews and s… more
Date: June 1976
Creator: Delhaye, Jean-Marc & Jones, Owen C., Jr.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fuel Dynamics Loss-of-Flow Test L3 : Final Report

Description: The behavior of FTR-type, mixed-oxide, pre-irradiated, ''intermediate-power-structure'' fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data reported here leads to a postulated scenario (sequence and timing) of events in the test. This scenario is presented, together with the calculated timing of events obtained by use of the SAS code.
Date: June 1976
Creator: Fischer, A. K.; Lo, R. K. & Barts, E. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Fluid Forces on Two Circular Cylinders in Crossflow

Description: Fluid excitation forces are measured in a water loop for two circular cylinders arranged in tandem and normal to flow. The Strouhal number and fluctuating drag and lift coefficients for both cylinders are presented for various spacings and incoming flow conditions. Results show the effects of Reynolds number, pitch ratio, and upstream turbulence on the fluid excitation forces.
Date: June 1985
Creator: Jendrzejczyk, J. A. & Chen, Shoei-Sheng
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Leakage Flow-Induced Vibration of an Eccentric Tube-in-Tube Slip Joint

Description: Eccentricity of a specific slip-joint design separating two cantilevered, telescoping tubes did not create any self-excited lateral vibrations that had not been observed previously for a concentric slip joint. In fact, the eccentricity made instabilities less likely to occur, but only marginally. Most important, design rules previously established to avoid instabilities for the concentric slip joint remain valid for the eccentric slip joint.
Date: August 1985
Creator: Mulcahy, T. M.
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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Design-Development and Operation of the Experimental Boiling-Water Reactor (EBWR) Facility, 1955--1967

Description: The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(… more
Date: November 1990
Creator: Boing, L. E.; Wimunc, E. A. & Whittington, G. A.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Analysis of Proposed Gamma-Ray Detection System for the Monitoring of Core Water Inventory in a Pressurized Water Reactor

Description: An initial study has been performed of the feasibility of employing an axial array of gamma detectors located outside the pressure vessel to monitor the coolant in a PWR. A one-dimensional transport analysis model is developed for the LOFT research reactor and for a mock-PWR geometry. The gamma detector response to coolant voiding in the core and down-comer has been determined for both geometries. The effects of various conditions (for example, time after shutdown, materials in the transport pa… more
Date: December 1987
Creator: Markoff, Diane Melanie
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Development of a MK-II Loop to Simulate Reactor Hydraulic Conditions

Description: The Mk-IIC Integral Loop was modified to provide an in-pile experimental apparatus that would simulate the subassembly coolant flow rate and inlet pressure head of the Fast Test Reactor (FTR). There were two main design changes. First, the safety dump tanks were removed from the Mk-IIC loop and replaced by a second annular linear induction pump (ALIP). Second, a flow restricting orifice was sized so that the hydraulic requirements of prototypical test-section coolant velocity and pressure head … more
Date: January 1979
Creator: Page, R. J. & Robinson, L. E.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Final Report on Test L4, a Loss-of-Flow Experiment

Description: The behavior of FTR-type, mixed-oxide, pre-irradiated "high-power-structure" fuel during a simulation of an FTR loss-of-flow accident was studied in the Mark-IIA integral TREAT loop. Analysis of the data leads to a postulated scenario (sequence and timing) of events in this test. This scenario is presented, together with the calculated timing of events obtained by use of SAS code.
Date: December 1976
Creator: Eberhart, James G.; Lo, R. & Barts, E. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Occurrence of Critical Heat Flux During Blowdown with Flow Reversal

Description: A small-scale experiment using Freon-11 at 130 degrees F (54.4 degrees C) and 65 psia (0.45 MPa) in a well-instrumented, transparent annular test section was used to study the occurrence of critical heat flux (CHF) during blow-down with flow reversal. The inner stainless steel tube of the annulus was uniformly heated over its 61-cm length. Inlet and exit void fractions were measured by a capacitance technique. Flow-regime transition was observed with high-speed photography. A 1-hr contact time… more
Date: January 1977
Creator: Leung, J. C. M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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COMMIX-1AR/P. a Three-Dimensional Transient Single-Phase Computer Program for Thermal Hydraulic Analysis of Single and Multicomponent Systems

Description: The COMMIX-1AR/P computer code is designed for analyzing the steady-state and transient aspects of single-phase fluid flow and heat transfer in three spatial dimensions. This version is an extension of the modeling in COMMIX-1A to include multiple fluids in physically separate regions of the computational domain, modeling descriptions for pumps, radiation heat transfer between surfaces of the solids which are embedded in or surround the fluid, a k-{var epsilon} model for fluid turbulence, and i… more
Date: July 1991
Creator: Blomquist, R. A.; Garner, P. L. & Gelbard, Ely M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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F1 Phenomenological Test on Fuel Motion : Final Report

Description: TREAT F-series tests are being conducted to provide data on fuel motion in an LMFBR during a hypothetical loss-of-flow accident. Fuel and fuel-boundary conditions in an LMFBR subassembly following sodium voiding and dryout under loss-of-flow conditions are simulated in each F-series test. Simulation is achieved with a single fuel element surrounded by an annular nuclear-heated wall in a dry (no sodium) test capsule. The area inside the heated wall was selected to represent the area inside the p… more
Date: May 1978
Creator: Argonne National Laboratory. Reactor Analysis and Safety Division.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Combined Motion of Fuel and Coolant Due to Fuel-Coolant Interactions under High Ramp Rate Reactivity Insertion

Description: An analysis has been made of the combined motion of fuel and coolant due to fuel-coolant interactions following a massive fuel failure in a high-ramp overpower transient. The motion of fuel and coolant was described using a two-fluid model formulation in which the mixture of sodium liquid and vapor and of fission gas, on the one hand, and the fuel particles, on the other, were treated as two superimposed continua. The method of solution employed a numerical procedure called the ACE method, a m… more
Date: July 1978
Creator: Chang, K. I. & Cho, D. H.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Loss-of-Flow Test L5 on FFTF-Type Irradiated Fuel

Description: Test L5 simulated a hypothetical loss-of-flow accident in an LMFBR using three (plutonium, uranium) dioxide fuel elements of the FTR type. The test elements were irradiated before TREAT Test L5 in the General Electric Test Reactor to 8 at. % burnup at about 40 kW/m. The pre-irradiation in GETR caused a fuel-restructuring range characteristic of moderate-power structure relative to the FTR. The test transient was devised so that a power burst would be initiated at incipient cladding melting afte… more
Date: March 1978
Creator: Simms, R.; Gehl, S. M.; Lo, R. K. & Rothman, Alan B.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Evaluation of Coolant Impurity Removal Equipment at the OMRE

Description: 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.
Date: October 15, 1964
Creator: Barbour, P. & Davis, W. W.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Proceedings of the Third Post-Accident Heat Removal Information Exchange November 2-4, 1977, Argonne National Laboratory, Argonne, Illinois

Description: Papers presented at the third Post-Accident Heat Removal Information Exchange concerning heat distribution and criticality considerations, particulate-bed phenomena, pool heat transfer and melt-front phenomena, behavior of heated concrete and sodium-concrete interactions, design-related studies, gas bubbling and boiling effects, and materials interactions at high temperatures and experimental methods.
Date: 1978?
Creator: Baker, Louis, Jr. & Bingle, James D.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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An Evaluation of Mercury Cooled Breeder Reactors

Description: Abstract: The technical feasibility and economic potential of fast breeder power reactor systems cooled with boiling mercury have been investigated by American-Standard under the United States Atomic Energy Commission's New Reactor Concepts Evaluation Program.
Date: October 13, 1959
Creator: Advanced Technology Laboratories
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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One-Dimensional Drift-Flux Model and Constitutive Equations for Relative Motion Between Phases in Various Two-Phase Flow Regimes

Description: In view of the practical importance of the drift-flux model for two-phase flow analysis in general and in the analysis of nuclear-reactor transients and accidents in particular, the kinematic constitutive equation for the drift velocity has been studied for various two-phase flow regimes. The constitutive equation that specifies the relative motion between phases in the drift-flux model has been derived by taking into account the interfacial geometry, the body-force field, shear stresses, and t… more
Date: October 1977
Creator: Ishii, M.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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Final Report on TREAT Test R3, a Single-Pin Loss-of-Flow Experiment

Description: The R3 was the first of the R-series fuel-element meltdown tests performed to support modeling and code development for analysis of hypothetical, whole-core accidents in LMFBRs. Test R3 served as a proof test for the subsequent R4-R8 sequence of seven-pin tests examining coolant, cladding, and fuel behavior under thermal and hydraulic conditions representative of a hypothetical loss-of-coolant (LOF) with failure to scram in FFTF.
Date: July 1977
Creator: Holtz, R. E.; Grolmes, M. A.; Spencer, B. W.; Miller, C. E.; Kramer, N. A.; Testa, F. J. et al.
Item Type: Refine your search to only Report
Partner: UNT Libraries Government Documents Department
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