Michael A. Ulrickson

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

Abstract 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.

Effects of Hypervapotron Geometry on Thermalhydraulic Performance

Plasma disruptions and edge localized modes can result in transient heat fluxes as high as 5 MW/m2 on portions of a tokamak reactor first wall (FW). To accommodate these heat loads, the FW will likely use water-cooled hypervapotron heatsinks to enhance the heat transfer. In this article, we present the results of a computational fluid dynamics (CFD) study using 70 °C inlet water at 2.7 MPa to investigate the tooth height and backchannel depth of 50-mm-wide hypervapotrons with 6-mm-pitch and 3-mm side slots. We compare a popular design with 4-mm-high teeth and a 5-mm backchannel to a more optimal case with 2-mm-high teeth and a 3-mm backchannel under nominal heat loads (0.5 MW/m2) on a 100-mm-heated length and under single-phase flow conditions. Better heat transfer in the latter case and the smaller backchannel permit a factor of two reduction in the required mass flow while maintaining the same beryllium armor surface temperatures near 130°C. The shallow teeth and smaller backchannel allow the 40 fingers in a typical panel to flow in parallel and simplify the water circuit. A comparison of the two hypervapotron designs during off-normal loading (5.0 MW/m2) and two-phase flow then follows. The design with 2-mm teeth has a 3.5% higher beryllium surface temperature of 648°C and reduces the critical heat flux (CHF) by ~2%. Hypervapotron width also plays a role in heat transfer and CHF. CFD results for 36 and 70 mm wide hypervapotrons compared to the 50-mm case reveal similar thermal performance at low heat flux, but a reduction in CHF with increasing width. This study highlights the necessary compromise between design margin during transient events, effective heat transfer under nominal conditions, limitations on finger width, and the simplicity needed in the water circuit design.

Electromagnetic Analysis of Forces and Torques on the Baseline and Enhanced ITER Shield Modules due to Plasma Disruption

An electromagnetic analysis is performed on the ITER shield modules under different plasma-disruption scenarios using the OPERA-3d software. The models considered include the baseline design as provided by the International Organization and an enhanced design that includes the more realistic geometrical features of a shield module. The modeling procedure is explained, electromagnetic torques are presented, and results of the modeling are discussed.

Divertor design for the tokamak physics experiment

In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4{times} L-mode), high beta ({beta}{sub N} {ge} 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74{degrees} from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m{sup 2} with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities.

The analysis of the electromagnetic loads on selected ITER blanket shield modules due to induced eddy and halo currents

The prediction of electromagnetic loads on the ITER blanket modules during a plasma disruption is considered for two different blanket modules and different disruption events.

Electromagnetic Analysis of Forces and Torques on Selected Components of the ITER Blanket System due to Plasma Disruption

Abstract The ITER device is based on the tokamak concept of magnetic confinement in which the plasma is contained by the use of strong magnetic fields. The nearest structure to the plasma is the blanket system which provides shielding to the vacuum vessel and the superconducting magnets. There are potential abnormal operating environments where the plasma currents inside the tokamak are disrupted and induce eddy currents in the blanket (first wall and shield module). These currents interact with the large magnetic fields to produce forces in the blanket which could potentially cause mechanical failure in the first wall, shield module, or vacuum vessel. For this reason the design and qualification of the ITER blanket system requires appropriate high-fidelity electromagnetic simulations that capture the physics of these disruption scenarios. A number of different geometries will be discussed revealing the effect of different first wall designs and shield modules on the forces and torques experienced by these assemblies during plasma disruption. The key features of the modeling procedure will be presented including the plasma current modeling and geometric modeling of the first wall, shield modules, and vacuum vessel. The eddy current calculation is performed using the Opera-3d software.

First Wall Qualification Testing at SNL

The first wall of ITER will be a replaceable sandwich of beryllium tiles, water-cooled copper alloy, and a water-cooled stainless steel backing. The first wall will be subject to approximately 30,000 pulses of surface heating at levels of 0.2 - 0.5 MW/m2 and 5 - 10 W/cm3 of volumetric nuclear heating. At these low heating levels, the main failure mechanism is predicted to be fatigue, particularly at the interface between the copper alloy and beryllium tile. Six different party teams are proposing to produce the first wall for ITER To qualify the processes and materials for producing the first wall, small mock-ups will be subject to fatigue testing at Sandia National Laboratories (SNL) Plasma Materials Test Facility and at a European Union test facility. We propose that the failure of a joining process is determined by an increase of surface temperature over nominal temperatures for a given surface heat flux. If the joint between a Be tile and Cu alloy degrades, the path from the heated surface to the coolant in the copper alloy increases, which should result in a higher surface temperature. This paper will document the test setup and preliminary analysis of the fatigue testing and failure criteria at SNL.

The tokamak physics experiment: tokamak improvement through advanced steady state control

Abstract The achievement of a long-pulse ignited discharge with over 1000 MW of fusion power in the International Thermonuclear Experimental Reactor will be an important goal for the next phase of the world magnetic fusion program. However, improvements in the physics are needed to design a more economically attractive tokamak power reactor than the present data base would support. Advanced, steady state plasma controls are the key to realizing these improvements. The Tokamak Physics Experiment has a flexible heating and current drive system for profile control; a flexible poloidal field system that supports a strongly shaped double-null poloidal divertor plasma configuration over a wide range of profiles; and a divertor designed for dispersive operation, flexibility, and remote handling. The machine performance in deuterium is sufficient to produce a reactor-like bootstrap current profile and to confine fast electrons for localized current profile control. A conducting structure, plasma rotation, field error compensation coils, and profile control are used to provide stable plasma configurations with β up to twice the Troyon limit and bootstrap current fraction approaching unity. The facility will be designed for 1000 s pulses initially to minimize the influence of initial transients on system behavior, but the pulse length can be extended through upgrades of external systems if necessary.

Transient electromagnetic analysis of selected blanket modules of the ITER blanket system due to plasma disruption

The prediction of electromagnetic loads on the ITER blanket modules during a plasma disruption is considered for two different blanket modules and different disruption events. The key features of the analysis procedure will be presented including the modeling the blanket modules (first wall and shield block), the plasma disruption current, and the integration of these models to produce a high-fidelity electromagnetic simulation that describes a plasma disruption event. The electromagnetic calculations are performed using the Opera-3d software. The loads due to the eddy currents were calculated for two outboard blanket modules and a number of selected disruption scenarios. Once these loads have been calculated they can also be exported for additional postprocessing to assess the mechanical loading effects.

Thermalhydraulic optimization of hypervapotron geometries for first wall applications

Plasma disruptions and Edge Localized Modes (ELMS) may result in transient heat fluxes as high as 5 MW/m2 on portions of the ITER first wall (FW). To accommodate these heat loads, roughly 50% of the first wall will have Enhanced Heat Flux (EHF) panels equipped with water-cooled hypervapotron heat sinks.