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in subjects: "Boiling"

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Acoustic measurements of the boiling stability tests on THORS sodium loop. [LMFBR]

Acoustic data of boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility were obtained using three sodium-immersible high temperature microphones. The data was analyzed in both the time and frequency domains and provides the following information: (1) the acoustic signal due to sodium boiling was clearly observed; (2) the signal level and the repetition rate of boiling pulses are directly proportional to the applied heat flux; (3) a typical boiling pulse consists… more

Date: January 1, 1977

Creator: Sheen, S.H.; Bobis, J.P. & Carey, W.M.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

A theoretical prediction of critical heat flux in subcooled pool boiling during power transients

Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions are of fundamenatal interest in the design, operation, safety of boiling and two-phase flow devices. This paper discusses the results of a comprehensive theoretical study made specifically to model transient CHF behavior in subcooled pool boiling. This study is based upon a simplified steady-state CHF model in terms of the vapor mass growth period. The results obtained from this theory in… more

Date: January 1, 1988

Creator: Pasamehmetoglu, K. O.; Nelson, R. A. & Gunnerson, F. S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Further considerations of critical heat flux in saturated pool boiling during power transients

In this paper, we further evaluate our previously postulated transient CHF model. First, we verify the steady-state CHF model on which the transient model is based by using recent macrolayer-thickness data. We also include the effect of thermal storage in the heater that we previously neglected. The use of a simplified approach in the prediction of the instantaneous surface heat flux for given pwoer generation rates considerably improves the predictive capability of the transient critical heat-… more

Date: January 1, 1988

Creator: Pasamehmetoglu, K.O. & Nelson, R.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Physics of reactor safety. Volume II. Quarterly report, April-June 1980

The work in the Applied Physics Division includes reports on reactor safety modeling and assessment by members of the Reactor Safety Appraisals Section. Work on reactor core thermal-hydraulics is performed in ANL's Components Technology Division, emphasizing 3-dimensional code development for LMFBR accidents under natural convection conditions.

Date: August 1, 1980

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Flow instability in a vertical annulus under steady state and transient conditions

Experimental data have been obtained for the occurrence of flow instability in a vertical annulus under steady state and transient down-flow conditions. The results show that the primary factor affecting the onset of flow instability is the Q{sub ratio} parameter (a measure of the void production in a channel). In addition, it is shown that for steady state conditions good agreement exists between the single annulus data and single tube data based upon the same L/D ratio. The results also indic… more

Date: January 1, 1992

Creator: Fighetti, C.; Kokolis, S.; Maciuca, C.; Reddy, D.G.; Yang, B.; Carrano, V. (Columbia Univ., New York, NY (United States)) et al.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Test PCM-5 rod bowing and bow direction reversal. [PWR]

Test PCM-5 was the first bundle test in the PCM Test Series being conducted by the Thermal Fuels Behavior Program of EG and G Idaho, Inc. as part of the Nuclear Regulatory Commission's Fuel Behavior Program. The experiment was performed in the Power Burst Facility (PBF) reactor at the Idaho National Engineering Laboratory. The bundle consisted of nine previously unirradiated PWR-type fuel rods, arranged in a 3 x 3 array within a square cross section flow shroud, with rod-to-rod spacing typical … more

Date: January 1, 1980

Creator: Kerwin, D. K.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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A theoretical prediction of critical heat flux in forced convection boiling during power transients

In this paper, a theoretical prediction of critical heat flux (CHF) during power transients in forced convective boiling is presented. The analysis is restricted to the departure from nucleate boiling (DNB) type of CHF at low qualities. The developed theory is compared with the experimental data available in the literature. The agreement is favorable. The results are discussed in terms of the various parameters affecting the transient CHF. This new model also is compared with the semi-empirical… more

Date: January 1, 1988

Creator: Pasamehmetoglu, K.O.; Nelson, R.A. & Gunnerson, F.S.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Design of a molten-lava, single-tube boiler experiment

No Description Available.

Date: July 1, 1975

Creator: Fewell, M.E.; Hardee, H.C. & Montoya, P.C.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Survey of Nonmetallic Liquid Coolants for Nuclear-Power Piles

A comprehensive survey of the literature was made in an effort to identify nonmetallic materials of possible usefulness as liquid coolants. Materials having maximum melting points of 1000 deg F and boiling points of l200 deg F were considered, but boiling points above 2200 deg F were preferred. Melting points, boiling points, densities, heat capacities, and thermal conductivities were tabulated. Approximately 190 materials appeared to have melting and boiling temperatures in a suitable range. A… more

Date: May 26, 1950

Creator: Shaw, H.L. & Boulger, F.W.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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BOILING NUCLEAR SUPERHEATER (BONUS) POWER STATION. Supplementary Study. Extrapolation to Large Central Station Integral Nuclear Superheat Plant

An evaluation was made of the maximum size plant for which the BONUS reactor plant could serve as a realistic prototype and the design changes required to increase the size and characteristics for the present BONUS design such that it could serve as a realistic prototype for the largest feasible integral-superheat reactor power plant. (M.C.G.)

Date: October 31, 1962

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Effects of water in film boiling over liquid metal melts

Liquid-liquid boiling experiments have been performed with H/sub 2/O and liquid metal melts in the 100-series test matrix (Runs 121, 126, 127) and the VE test matrix. Some of the pre-explosion unstable film boiling data as well as observations from the explosive series have been previously reported.

Date: January 1, 1986

Creator: Greene, G.A.; Finfrock, C. & Burson, S.B.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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MODIFICATION OF THE EXPERIMENTAL BOILING WATER REACTOR (EBWR) FOR HIGHER- POWER OPERATION

Supplement to ANL-5607. Alterations and additions made to the Experimental Boiling Water Reactor (EBWR) plant to permit operation at power levels up to I00 Mw(t) are described. Topics covered include over-all system modifications and additions, nuclear component modifications and additions, and reboiler plant component description. (M.C.G.)

Date: April 1, 1962

Creator: Matousek, J.F. comp.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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The effect of surface gas injection on film boiling heat transfer

A database was developed for heat transfer from a horizontal plate with both film boiling and gas jetting occurring simultaneously, in a pool of water maintained at its saturation temperature. The effect of passing nitrogen through established film boiling is to increase the heat transfer from that surface. At the highest superficial gas velocity measured, approximately 8.5 cm/s, and for a constant surface temperature, the heat transfer increases by a factor of two over the heat transfer with n… more

Date: January 1, 1991

Creator: Duignan, M.R. (Westinghouse Savannah River Co., Aiken, SC (United States)); Greene, G.A. (Brookhaven National Lab., Upton, NY (United States)) & Irvine, T.F. Jr. (State Univ. of New York, Stony Brook, NY (United States))

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Steady-State Film-Boiling Data in Rod-Bundle Geometry and Non-Equilibrium Correlation Assessment

A series of 22 steady-state, rod bundle, dispersed flow film boiling experiments has been performed in the Thermal-Hydraulic Test Facility (THTF), a pressurized-water loop containing 64 full-length electrically heated rods. Test parameters in the upflow experiments cover a wide range of conditions typical of those which might be encountered during a nuclear reactor loss-of-coolant accident. Local equilibrium fluid conditions were calculated using mass and energy conservation considerations. Exp… more

Date: January 1, 1982

Creator: Yoder, G.L.; Morris, D.G.; Mullins, C.B.; Ott, L.J. & Reed, D.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

On the hot-spot-controlled critical heat flux mechanism in pool boiling of saturated fluids

In this paper, we further investigate the hypothesis that the critical heat flux (CHF) occurs when some point on the heated surface reaches a high enough temperature that liquid can no longer contact that point, resulting in a gradual but continuous increase in the overall surface temperature. This hypothesis unifies the occurrence of the CHF and the quenching of hot surfaces by relating both to the same concept, i.e., the ability of a liquid to contact a hot surface. We use a two-dimensional t… more

Date: January 1, 1992

Creator: Unal, C.; Sadasivan, P. & Nelson, R.A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Compressible analysis of inlet plenum pressure rise due to sodium boiling in fuel subassemblies during pump coastdown of an LMFBR

The effect of sodium compressibility and steel elasticity on the rise in inlet plenum pressure occurring during boiling in a loss-of-flow accident in an LMFBR has been investigated using the require consideration in accident analysis. The pressure rise is less for pool than for loop designs. 3 refs., 1 fig., 9 tabs.

Date: May 1, 1980

Creator: Kalimullah & Hummel, H.H.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Non-equilibrium aspects of water reactor safety

A critical review of the thermal non-equilibrium phase change models applicable to the water reactor safety calculations is presented. Four specific areas are discussed: (1) rapid depressurization (flashing), (2) post-dryout heat transfer, (3) direct contact condensation, and (4) sub-cooled boiling. The models have been judged from a mechanistic viewpoint, and the areas where further work is needed have been pointed out. It is suggested that when a phase change occurs due to a difference in the… more

Date: July 1, 1977

Creator: Jones, O.C. Jr. & Saha, P.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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An improved steam generator model for the SASSYS code

A new steam generator model has been developed for the SASSYS computer code, which analyzes accident conditions in a liquid metal cooled fast reactor. It has been incorporated into the new SASSYS balance-of-plant model but it can also function on a stand-alone basis. The steam generator can be used in a once-through mode, or a variant of the model can be used as a separate evaporator and a superheater with recirculation loop. The new model provides for an exact steady-state solution as well as … more

Date: January 1, 1989

Creator: Pizzica, P. A.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Heat transfer near spacer grids in rod bundles

Heat transfer data from several sources have been assembled which show the effect of spacer grids on local heat transfer within a rod bundle. Both single phase (air and steam) data and two phase (steam/water) data show heat transfer augmentation in the grid region. Heat transfer improvement immediately beyond the grid ranges from a few percent to over fifty percent in these experiments, depending on flow conditions. The data are examined using several nondimensional parameters which relate the … more

Date: January 1, 1985

Creator: Yoder, G.L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

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Summary report on four foot septifoil cooling experiment

Cooling parameters for some of the SRS reactor internal components are computed using the Transient Reactor Analysis Code, TRAC.'' In order to benchmark the code, the Safety Analysis Group of SRL requested an experiment to provide measurements of cooling parameters in a well defined physical system utilizing SRS reactor component(s). The experiment selected included a short length of septifoil with both top and bottom fittings containing five simulated control rods in an unseated'' configuratio… more

Date: October 1, 1991

Creator: Randolph, H. W.; Collins, S. L.; Verebelyi, D. T. & Foti, D. J.

Item Type: Report

Partner: UNT Libraries Government Documents Department

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Dynamic boiling tests in a 19-pin simulated LMFBR fuel assembly

A series of quasi-steady state tests were made to study dynamic boiling and to determine a boiling region which would not lead to immediate dryout and pin failure in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility (previously the Fuel Failure Mockup), an engineering-scale sodium loop for testing simulated LMFBR fuel assemblies at normal and off-normal operating conditions. These tests were made on an assembly consisting of 19 electrically heated pins of the same configuration as FF… more

Date: January 1, 1977

Creator: Wantland, J. L.; Clapp, N. E.; Fontana, M. H.; Gnadt, P. A. & Hanus, N.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

History match simulation of Serrazzano geothermal reservoir

The simulator SHAFT79 of Lawrence Berkeley Laboratory has been applied to field-wide distributed parameter simulation of the vapor-dominated geothermal reservoir at Serrazzano, Italy. Using a three-dimensional geologically accurate mesh and detailed flow rate data from 19 producing wells, a period of 15.5 years (from 1959 to 1975) has been simulated. The reservoir model used is based on field measurements of temperatures and pressures, laboratory data for core samples, and available geological … more

Date: August 1, 1980

Creator: Pruess, K.; Weres, O.; Schroeder, R.; Marconcini, R. & Neri, G.

Item Type: Report

Partner: UNT Libraries Government Documents Department

open access

Two-dimensional modeling of sodium boiling transients in simulated LMFBR fuel bundles

A two-dimensional code for analysis of sodium boiling in LMFBR fuel assemblies has been developed at ORNL. This code, THORAX, has been used to analyze tests in 19- and 61-pin electrically-heated, simulated LMFBR fuel assemblies in the THORS facility. THORAX has simulated well the transient growth of the two-dimensional boiling region and the resulting static flow instability leading to dryout. Extrapolation of results to a full size fuel pin bundle shows that two-dimensional effects are reduced… more

Date: January 1, 1982

Creator: Rose, S. D.; Dearing, J. F.; Carbajo, J. J.; Levin, A. E.; Montgomery, B. H. & Wantland, J. L.

Item Type: Article

Partner: UNT Libraries Government Documents Department

open access

Critical heat flux concerns during the flow instability phase of a DEGB LOCA

Arguments are presented that support the proposal that a separate burnout risk analysis, for the Flow Instability (FI) phase of a LOCA, not be required for reactor restart. With expected reactor power limits, flow instability will occur before critical heat flux (CHF). Since FI power limits preclude the occurrence of flow instability in a bounding accident, a DEGB LOCA, the risk of CHF and attendant burnout is negligible. A review of RDAP data revealed that in the past reactor assemblies operat… more

Date: August 1, 1990

Creator: Shadday, M.A. Jr.

Item Type: Report

Partner: UNT Libraries Government Documents Department