Ronald D. Boyd

Subcooled Flow Boiling Critical Heat Flux (CHF) and Its Application to Fusion Energy Components. Part I. A Review of Fundamentals of CHF and Related Data Base

The present understanding of critical heat flux (CHF) in subcooled flow boiling with water is reviewed and fusion reactor component high-heat flux (HHF) requirements are outlined. This survey (Part...

Subcooled water flow boiling at 1. 66 MPa under uniform high heat flux conditions

Steady-state subcooled water flow boiling experiments have been carried out in a uniformly heated horizontal circular channel with an exit pressure of 1.66 MPa and with the mass velocity G varying from 4.4 to 32.0 Mg/m{sup 2} {center dot} s. These experiments are related to high heat flux removal in fusion reactor beam dumps and first walls in compact fusion reactors. For the chosen values of L/D and exit pressure, the measured critical heat flux (CHF) values are higher than any previous values for smooth tubes in the literature. An increase in the exit pressure resulted in an increase in the slope of this relationship. However, the local heat transfer coefficient actually decreased as the pressure increased, for the same power level and mass velocity.

Subcooled Flow Boiling Critical Heat Flux (CHF) and Its Application to Fusion Energy Components. Part II. A Review of Microconvective, Experimental, and Correlational Aspects

Microconvective, instability, experimental, and correlational aspects of subcooled flow boiling critical heat flux (CHF) are summarized. The present understanding of CHF in subcooled flow boiling is reviewed and research directions that will permit the accommodation of higher heat fluxes are outlined. This survey (Parts I and II), which contains a representative coverage of the literature over the last 30 years, is concerned only with CHF in the subcooled flow boiling regime, and unless otherwise noted, all references to CHF are confined to that regime.

Subcooled Water Flow Boiling Transition and the L/D Effect on CHF for a Horizontal Uniformly Heated Tube

Steady-State subcooled water flow boiling experiments were carried out in a uniformly heated horizontal circular channel with a 0.45-MPa exit pressure and with the mass velocity varying from 1.56 t...

Similarities and Differences Between Single-Side and Uniform Heating for Fusion Applications—II: Sine Heat Flux

Quantitative correlation factors and functions are derived that relate the inside peak heat fluxes for cases involving either a single-side or a uniform external heat flux, which is applied to a circular coolant channel. The results are also presented in terms of a map that gives insight on how to improve high-heat-flux accommodation. The results show that similarities and differences depend on the channel geometry, material, and internal convection. 6 refs., 5 figs.

Single-side conduction modeling for high heat flux coolant channels

In the development of plasma-facing components (PFCs), most investigators have erroneously postulated negligible water critical heat flux dependence on the coolant channel length-to-diameter (L/D) ratio above a constant value of L/D. Although encouraging results have been obtained in characterizing peaking factors for local two-dimensional boiling curves and critical heat flux, additional experimental data and theoretical model development are needed to validate the applicability to PFCs. Both these and related issues will affect the flow boiling correlation and data reduction associated with the development of PFCs for fusion reactors and other physical problems that are dependent on conduction modeling in the heat flux spectrum of applications. Both exact solutions and numerical conjugate analyses are presented for a one-side heated (OSH) geometry. The results show (a) the coexistence of three flow regimes inside an OSH circular geometry, (b) the correlational dependence of the inside wall heat flux and temperature, and (c) inaccuracies that could arise in some data reduction procedures.

Local subcooled flow boiling model development

Several existing heat transfer models for uniformly heated channels were examined to accurately represent the boiling curve and to characterize the local heat transfer coefficient under high-heat-flux (HHF) conditions. Comparisons with HHF data showed that major correlation modifications were needed in the subcooled partial nucleate boiling (SPNB) region. Because the slope of the boiling curve in this region is important to ensure continuity of the HHF trends into the fully developed boiling region and up to the critical heat flux, accurate characterization in the SPNB region is essential. Approximations for the asymptotic limits for the SPNB region have been obtained and have been used to develop an improved composite correlation. The developed correlation has been compared with 363 water data points. For the local heat transfer coefficient and wall temperature, the overall percent standard deviations with respect to the data were 19 and 3%, respectively, for the high-velocity water data. 14 refs., 3 figs., 3 tabs.

Heat transfer for fusion component applications

A quasi-automated high heat flux flow boiling facility has been developed for the systematic study of critical heat flux (CHF), heat transfer, and two-phase pressure drop. High heat flux research is important in state-of-the-art electronics and fusion component design. For fusion applications, there are practically no low-pressure data for large values of coolant channel length-to-diameter (L/D) ratio (i.e., 100), channel diameters near 1.0 cm, and medium to high heat flux levels (i.e., 100 to 2000 W/cm/sup 2/). A second step is provided to fill this void. Forced flow boiling (with water) quasi-steady experiments have been conducted on uniformly (resistively) heated horizontal copper tubes. The tubes were 1.02 cm in inside diameter and 117.87 cm long. The inlet water temperature was 20/sup 0/C. For a 1.6-MPa exit pressure, measurements of the CHF varied from the annular flow regime (150 W/cm/sup 2/) to the subcooled flow boiling water regime (425 W/cm/sup 2/). The mass velocity was varied form 0.63 to 3.5 Mg/m/sup 2/ . s. At 1.6 MPa, the transition between the annular and subcooled CHF regimes were measured to occur between 1.03 and 1.26 Mg/m/sup 2/ . s. Large axial variations in the Nusselt number were also measured.