Publications

Results 1–25 of 44
Skip to search filters

The effect of paramagnetic shift during thermal quench on internal components in fusion devices

Fusion Engineering and Design

Ulrickson, M.A.; Kotulski, J.D.

A plasma current disruption is usually initiated by impurity influx that causes a rapid decrease in plasma thermal stored energy (thermal quench). Thermal quench occurs in 500-2000 μs on a large device like ITER. Depending on the β value, the plasma may be either paramagnetic or diamagnetic. Thermal quench causes a large shift in paramagnetism (or diamagnetism) and a corresponding change in toroidal flux. The flux swing can be 1-2 Weber with the rate of change of the toroidal field between 25 and 150 T/s for a device like ITER. The toroidal field shift induces poloidal current in the vessel and possibly in internal components. We have developed a method for simulating the thermal quench field shift that is compatible for use with the electromagnetic simulation codes. The method is based on a radially thin shell having the shape of the last closed flux surface with poloidal current driven to duplicate the toroidal field shift. The magnitude of the current and its time history are adjusted to duplicate the flux change during a disruption thermal quench. We will present the results of using this method to simulate the induced currents in a vacuum vessel having two shells. © 2012 Elsevier B.V. All rights reserved.

More Details
TYPE Conference YEAR 2012
ScopusOSTI

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

A comparison of two-phase computational fluid dynamics codes applied to the ITER first wall hypervapotron

IEEE Transactions on Plasma Science

Youchison, Dennis L.; Ulrickson, M.A.; Bullock, James H.

Enhanced radial transport in the plasma and the effect of ELMS may increase the ITER first wall heat loads to as much as 4 to 5 MW/m2 over localized areas. One proposed heatsink that can handle these higher loads is a CuCrZr hypervapotron. One concept for a first wall panel consists of 20 hypervapotron channels, each measuring 1400 mm long and 48.5 mm wide. The nominal cooling conditions anticipated for each channel are 400 g/s of water at 3 MPa and 100 °C. This will result in boiling over a portion of the total length. A two-phase thermalhydraulic analysis is required to predict accurately the thermal performance. Existing heat transfer correlations used for nucleate boiling are not appropriate here because the flow does not reach fully developed conditions in the multi-segmented channels. Our design-by-analysis approach used two commercial codes, Fluent and Star-CCM+, to perform computational fluid dynamics analyses with conjugate heat transfer. Both codes use the Rensselear (RPI) model for wall heat flux partitioning to model nucleate boiling as implemented in user-defined functions. We present a comparison between the two codes for this Eulerian multiphase problem that relies on temperature dependent materials properties. The analyses optimized the hypervapotron geometry, including teeth height and pitch, as well as the depth of the back channel to permit highly effective boiling heat transfer in the grooves between the teeth while ensuring that no boiling could occur at the back channel exit. The analysis used a representative heat flux profile with the peak heat flux of 5 MW/m2 limited to a 50 mm length. The maximum surface temperature of the heatsink is 415 °C. The baseline design uses 2 mm for the teeth height, a 3 mm width and 6 mm pitch, and a back channel depth of 8 mm. The teeth are detached from the sidewall by a 2-mm-wide slot on both sides that aids in sweep-out and quenching of the vapor bubbles. © 2006 IEEE.

More Details
TYPE Journal Article YEAR 2010
ScopusOSTI
Results 1–25 of 44
Results 1–25 of 44