J. Lore

Improved understanding of physics processes in pedestal structure, leading to improved predictive capability for ITER

Joint experiment/theory/modelling research has led to increased confidence in predictions of the pedestal height in ITER. This work was performed as part of a US Department of Energy Joint Research Target in FY11 to identify physics processes that control the H-mode pedestal structure. The study included experiments on C-Mod, DIII-D and NSTX as well as interpretation of experimental data with theory-based modelling codes. This work provides increased confidence in the ability of models for peeling–ballooning stability, bootstrap current, pedestal width and pedestal height scaling to make correct predictions, with some areas needing further work also being identified. A model for pedestal pressure height has made good predictions in existing machines for a range in pressure of a factor of 20. This provides a solid basis for predicting the maximum pedestal pressure height in ITER, which is found to be an extrapolation of a factor of 3 beyond the existing data set. Models were studied for a number of processes that are proposed to play a role in the pedestal ne and Te profiles. These processes include neoclassical transport, paleoclassical transport, electron temperature gradient turbulence and neutral fuelling. All of these processes may be important, with the importance being dependent on the plasma regime. Studies with several electromagnetic gyrokinetic codes show that the gradients in and on top of the pedestal can drive a number of instabilities.

The Development of the Material Plasma Exposure Experiment

The availability of future fusion devices, such as a fusion nuclear science facility or demonstration fusion power station, greatly depends on long operating lifetimes of plasma facing components in their divertors. ORNL is designing the Material Plasma Exposure eXperiment (MPEX), a superconducting magnet, steady-state device to address the plasma material interactions of fusion reactors. MPEX will utilize a new highintensity plasma source concept based on RF technology. This source concept will allow the experiment to cover the entire expected plasma conditions in the divertor of a future fusion reactor. It will be able to study erosion and redeposition for relevant geometries with relevant electric and magnetic fields in-front of the target. MPEX is being designed to allow for the exposure of a priori neutron-irradiated samples. The target exchange chamber has been designed to undock from the linear plasma generator such that it can be transferred to diagnostics stations for more detailed surface analysis. MPEX is being developed in a staged approach with successively increased capabilities. After the initial development step of the helicon source and electron cyclotron heating system, the source concept is being tested in the Proto-MPEX device. Proto-MPEX has achieved electron densities of more than 4×1019 m-3 with a large diameter (13 cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared with the helicon plasma only without microwave heating.

Design and Analysis of Divertor Scraper Elements for the W7-X Stellarator

A set of new water-cooled divertor components is being designed for the Wendelstein 7-X stellarator to protect the edges of the primary plasma facing components during the bootstrap current evolution (~ 40 s). These new components, referred to as scraper elements (SEs), will intercept field lines and associated heat flux that would otherwise overload the main target edges in certain operational scenarios. The SEs are calculated to experience peak heat fluxes ~15-16 MW/m2 and will be constructed from carbon fiber reinforced composite monoblocks of a type that has been qualified for ITER. The heat flux distribution and magnitude is calculated from field line following in a 3-D magnetic field that includes the contribution from plasma currents. The heat flux calculations are coupled with an engineering design in an iterative process to generate SEs that meet the design criteria while reducing the geometric complexity of the elements.

The Development of Plasma-Material Interaction Facilities for the Future of Fusion Technology

Abstract A new era of fusion research has started with ITER being constructed and DEMO for power demonstration on the horizon. However, the fusion nuclear science needs to be developed before DEMO can be designed. One of the most crucial and most complex outstanding science issues to be solved is the plasma surface interaction (PSI) in the hostile environment of a nuclear fusion reactor. Not only are materials exposed to unprecedented steady-state and transient power fluxes, but they are also exposed to unprecedented neutron fluxes. Both the ion fluxes and the neutron fluxes will change the micro-structure of the plasma facing materials significantly even to the extent that their structural integrity is compromised. New devices have to be developed to address the challenges ahead. Linear plasma-material interaction facilities can play a crucial role in advancing the plasma-material interaction science and the development of plasma facing components for future fusion reactors.

Effect of nonaxisymmetric magnetic perturbations on divertor heat and particle flux profiles in National Spherical Torus Experiment a)

Small, nonaxisymmetric magnetic perturbations generated by external coils have been found to break the axisymmetry of heat and particle flux deposition pattern in the divertor area in the National Spherical Torus Experiment (NSTX). This breaking by the applied 3-D field causes strike point splitting that is represented as local peaks and valleys in the divertor profiles. In case of n = 3 fields application, the broken toroidal symmetry of the divertor profile shows 120° of spatial periodicity while data for n = 1 fields provide a fully nonaxisymmetric heat and particle deposition. Field line tracing showed good agreement with the measured heat and particle flux profiles. Higher toroidal mode number (n = 3) of the applied perturbation produced more and finer striations in the divertor profiles than in the lower mode number (n = 1) case. Following the previous result of the intrinsic strike point splitting by the n = 3 error fields [Nucl. Fusion 50, 045010 (2010); J. Nucl. Mater. (2011), doi:10.1016/j.jnucm...

Access to edge scenarios for testing a scraper element in early operation phases of Wendelstein 7-X

The edge topology of magnetic fusion devices is decisive for the control of the plasma exhaust. In Wendelstein 7-X, the island divertor concept will be used, for which the edge topology can change significantly as the internal currents in a plasma discharge evolve towards steady-state. Consequently, the device has been optimized to minimize such internal currents, in particular the bootstrap current [1]. Nonetheless, there are predicted pulse scenarios where effects of the remaining internal currents could potentially lead to overload of plasma-facing components. These internal currents are predicted to evolve on long time scales (tens of seconds) so their effects on the edge topology and the divertor heat loads may not be experimentally accessible in the first years of W7-X operation, where only relatively short pulses are possible. However, we show here that for at least one important long-pulse divertor operation issue, relevant physics experiments can be performed already in short-pulse operation, through judicious adjustment of the edge topology by the use of the existing coil sets. The specific issue studied here is a potential overload of the divertor element edges. This overload might be mitigated by the installation of an extra set of plasma-facing components, so-called scraper elements, as suggested in earlier publications. It is shown here that by a targeted control of edge topology, the effectiveness of such scraper elements can be tested already with uncooled test-scraper elements in short-pulse operation. This will allow an early and well-informed decision on whether long-pulse-capable (actively cooled) scraper elements should be built and installed.

Three-dimensional equilibria and transport in RFX-mod: A description using stellarator tools

RFX-mod self-organized single helical axis (SHAx) states provide a unique opportunity to advance 3D fusion physics and establish a common knowledge basis in a parameter region not covered by stellarators and tokamaks. The VMEC code has been adapted to the reversed-field pinch (RFP) to model SHAx equilibria in fixed boundary mode with experimental measurements as constraint. The averaged particle diffusivity over the helical volume, estimated with the Monte Carlo code ORBIT, has a neoclassical-like dependence on collisionality and does not show the 1/ν trend of un-optimized stellarators. In particular, the helical region boundary, corresponding to an electron transport barrier with zero magnetic shear and improved confinement, has been investigated using numerical codes common to the stellarator community. In fact, the DKES/PENTA codes have been applied to RFP for local neoclassical transport computations, including radial electric field, to estimate thermal diffusion coefficients in the barrier region for...

Modeling and Analysis of the W7-X High Heat-Flux Divertor Scraper Element

The Wendelstein 7-X stellarator experiment is scheduled for the completion of device commissioning and the start of first plasma in 2015. At the completion of the first two operational phases, the inertially cooled test divertor unit will be replaced with an actively cooled high heat-flux divertor, which will enable the device to increase its pulse length to steady-state plasma performance. Plasma simulations show that the evolution of bootstrap current in certain plasma scenarios produce excessive heat fluxes on the edge of the divertor targets. It is proposed to place an additional scraper element in the 10 divertor locations to intercept some of the plasma flux and reduce the heat load on these divertor edge elements. Each scraper element may experience a 500-kW steady-state power load, with localized heat fluxes as high as 20 MW/m2. Computational analysis has been performed to examine the thermal integrity of the scraper element. The peak temperature in the carbon-carbon fiber composite, the total pressure drop in the cooling water, and the increase in water temperature must all be examined to stay within specific design limits. Computational fluid dynamics modeling is performed to examine the flow paths through the multiple monoblock fingers as well as the thermal transfer through the monoblock swirl tube channels.

Multiphysics Analysis of the Wendelstein 7-X Actively Cooled Scraper Element

Abstract The Wendelstein 7-X stellarator experiment is scheduled to start operation in mid-2015, and to move to steady-state operation in 2019. During this steady-state operation, certain plasma scenarios have been shown to produce heat fluxes that exceed the technological limits on the edges of the divertor target elements. The addition of a so-called scraper element (SE) in the ten divertor locations is being investigated in order to reduce the heat load on these divertor target edges. The ANSYS commercial multiphysics package is used to model the performance of the SE under predicted operational conditions. Computational fluid dynamics (CFD) modeling is performed to analyze the hydraulic and thermal characteristics of the water-cooled SE under thermal loading using the ANSYS CFX software. This multiphysics modeling is performed for the entire SE to ensure that the total pressure drop in the cooling water circuits, the increase in water temperature, and the peak temperature in the CFC all satisfy the design requirements. Because the contour of the SE surface must be machined to a sub-millimeter precision, it is important to determine the amount of thermal expansion experienced by the entire SE. The thermal-hydraulic results are imported into ANSYS Mechanical to perform the thermal-structural analysis. The thermal deformation of the SE is examined to confirm that the component’s position will remain within its operational limits.

Intrinsic plasma rotation and Reynolds stress at the plasma edge in the HSX stellarator

Using multi-tipped Langmuir probes in the edge of the HSX stellarator, the radial electric field and parallel flows are found to deviate from the values calculated by the neoclassical transport code PENTA for the optimized quasi-helically symmetric (QHS) configuration. To understand whether Reynolds stress might explain the discrepancy, fluctuating floating potential measurements are made at two locations in the torus corresponding to the low field and high field sides of the device. The measurements at the two locations show clear evidence of a gradient in the Reynolds stress. However, the resulting flow due to the gradient in the stress is found to be large and in opposite directions for the two locations. This makes an estimation of the flux surface average using a small number of measurement locations impractical from an experimental perspective. These results neither confirm nor rule out whether Reynolds stress plays an important role for the QHS configuration. Measurements made in configurations with the quasi-symmetry degraded show even larger flows and greater deviations from the neoclassically calculated velocity profiles than the QHS configuration while the fluctuation magnitudes are reduced. Therefore, for these configurations in particular, the Reynolds stress is most likely not responsible for the additional momentum.