Publications

Results 26–50 of 58
Skip to search filters

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Fusion Science and Technology

Youchison, Dennis L.; Ulrickson, M.A.

Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling. This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 °C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

More Details
TYPE Journal Article YEAR 2011
OSTIDOI

Prediction of critical heat flux in water-cooled plasma facing components using computational fluid dynamics

Youchison, Dennis L.; Ulrickson, M.A.

Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling. This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

More Details
TYPE Conference YEAR 2010
OSTI

Silicon carbide tritium permeation barrier for steel structural components

Buchenauer, D.A.; Kolasinski, Robert K.; Youchison, Dennis L.; Garde, J.; Holschuh, Thomas V.

Chemical vapor deposited (CVD) silicon carbide (SiC) has superior resistance to tritium permeation even after irradiation. Prior work has shown Ultrametfoam to be forgiving when bonded to substrates with large CTE differences. The technical objectives are: (1) Evaluate foams of vanadium, niobium and molybdenum metals and SiC for CTE mitigation between a dense SiC barrier and steel structure; (2) Thermostructural modeling of SiC TPB/Ultramet foam/ferritic steel architecture; (3) Evaluate deuterium permeation of chemical vapor deposited (CVD) SiC; (4) D testing involved construction of a new higher temperature (> 1000 C) permeation testing system and development of improved sealing techniques; (5) Fabricate prototype tube similar to that shown with dimensions of 7cm {theta} and 35cm long; and (6) Tritium and hermeticity testing of prototype tube.

More Details
TYPE Conference YEAR 2010
OSTI

High heat flux testing of a helium-cooled tungsten tube with porous foam

Fusion Engineering and Design

Youchison, Dennis L.; Lutz, Thomas J.; Williams, B.; Nygren, Richard E.

Utramet, Inc. fabricated one-piece heat exchanger tubes of chemical vapor deposited (CVD) tungsten (W), each with an internal porous mesh fused along either 51 or 38 mm of the axial length of a tube 15 mm in outer diameter. The open porous mesh has a structure of joined ligaments that combines relatively low resistance to flow and a large area for heat transfer. In tests at the Electron Beam Test Stand (EBTS) at Sandia National Laboratories, the maximum absorbed heat load was 22.4 MW/m2 with helium at 4 MPa, flowing at 27 g/s and with inlet and outlet temperatures of 40 and 91 °C and a pressure drop of ∼0.07 MPa. The preparation and testing of the samples was funded through a Phase I grant by the US Department of Energy's Small Business Innovation Research Program. The paper reports the surface temperature distribution indicated by an infrared camera, test conditions, a post-test examination in a scanning electron microscope and other details. © 2007 Elsevier B.V. All rights reserved.

More Details
TYPE Conference YEAR 2007
ScopusOSTI

Features of plasma sprayed beryllium armor for the ITER first wall

Journal of Nuclear Materials

Nygren, Richard E.; Youchison, Dennis L.; Hollis, K.J.

Two water-cooled mockups with CuCrZr heat sinks and plasma sprayed beryllium (PS Be) armor, 5 and 10 mm thick respectively, were fabricated at Los Alamos National Laboratory and thermally cycled at Sandia at 1 and 2 MW/m2. The castellated surface of the CuCrZr mechanically locked the armor. The resulting PS Be morphology controlled cracking during thermal cycling. Post test examinations showed transverse cracks perpendicular to the surface of the armor that would relieve thermal stresses but not degrade heat transfer. The mockups and two others previously produced for the European Fusion Development Agreement had somewhat porous armor, with a thermal conductivity estimated to be about 1/4 that of fully dense beryllium, due to the low (600-650 °C) substrate temperature during deposition specifically requested by EFDA to avoid subsequent heat treating of CuCrZr. Some melting of the armor was expected and observed in the tests. © 2007 Elsevier B.V. All rights reserved.

More Details
TYPE Conference YEAR 2007
ScopusOSTIDOI

ITER first wall Module 18 - The US effort

Fusion Engineering and Design

Nygren, Richard E.; Ulrickson, M.A.; Tanaka, T.J.; Youchison, Dennis L.; Lutz, Thomas J.; Bullock, J.; Hollis, K.J.

The US will supply outboard Module 18 for the International Thermonuclear Experimental Reactor. This module, radially thinner than other modules with a "nose" that curves radially outward to mate with the divertor, has the potential for high electromagnetic (EM) loads from vertical displacement events and high heat loads. The 316LN-IG shield block and first wall (FW) panels must be slotted to mitigate the EM loads and progress in developing the design is summarized. The FW has beryllium (Be) armor joined to a water-cooled CuCrZr heat sink with embedded 316LN-IG cooling channels. The US Team is considering possible fabrication methods as the design develops. Brief results of high heat flux experiments at Sandia on mockups with plasma-sprayed Be armor prepared at Los Alamos National Laboratory are noted. © 2005 Elsevier B.V. All rights reserved.

More Details
TYPE Conference YEAR 2006
ScopusOSTI
Results 26–50 of 58
Results 26–50 of 58