A. Lumsdaine

Fusion Nuclear Science Facility (FNSF) Before Upgrade to Component Test Facility (CTF)

Abstract The compact (R0~1.2-1.3m) Fusion Nuclear Science Facility (FNSF) is aimed at providing a fully integrated, continuously driven fusion nuclear environment of copious fusion neutrons. This facility would be used to test, discover, and understand the complex challenges of fusion plasma material interactions, nuclear material interactions, tritium fuel management, and power extraction. Such a facility properly designed would provide, initially at the JET-level plasma pressure (~30%T2) and conditions (e.g., Hot-Ion H-Mode, Q<1)), an outboard fusion neutron flux of 0.25 MW/m2 while requiring a fusion power of ~19 MW. If and when this research is successful, its performance can be extended to 1 MW/m2 and ~76 MW by reaching for twice the JET plasma pressure and Q. High-safety factor q and moderate-plasmas are used to minimize or eliminate plasma-induced disruptions, to deliver reliably a neutron fluence of 1 MW-yr/m2 and a duty factor of 10% presently anticipated for the FNS research. Success of this research will depend on achieving time-efficient installation and replacement of all internal components using remote handling (RH). This in turn requires modular designs for the internal components, including the single-turn toroidal field coil center-post. These device goals would further dictate placement of support structures and vacuum weld seals behind the internal and shielding components. If these goals could be achieved, the FNSF would further provide a ready upgrade path to the Component Test Facility (CTF), which would aim to test, for ≤6 MW-yr/m2 and 30% duty cycle, the demanding fusion nuclear engineering and technologies for DEMO. This FNSF-CTF would thereby complement the ITER Program, and support and help mitigate the risks of an aggressive world fusion DEMO R&D Program. The key physics and technology research needed in the next decade to manage the potential risks of this FNSF are identified.

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.

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.

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.

The Material Plasma Exposure eXperiment MPEX: Pre-design, development and testing of source concept

The availability of future fusion devices such as a Fusion Nuclear Science Facility (FNSF) or DEMO 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 high-intensity 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 re-deposition 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 transfer cask 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 ECH system the source concept is being tested in the Proto-MPEX device (100 kW helicon, 200 kW EBW, 30 kW ICRH). Proto-MPEX has achieved electron densities of more than 4×1019m-3 with a large diameter (13cm) helicon antenna at 100 kW power. First heating with microwaves resulted in a higher ionization represented by higher electron densities on axis, when compared to the helicon plasma only without microwave heating.

Overview of Design and Analysis Activities for the W7-X Scraper Element

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A high heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). This paper presents an overview of all of these activities and their current status.

Thermal-fluid modeling of single- and two-phase flows in the W7-X high heat flux divertor scraper element

The Wendelstein 7-X (W7-X) stellarator experiment is scheduled to begin operation in 2015 with steady-state operation planned for 2019. During the steady-state operation, certain simulated plasma scenarios have been shown to produce heat fluxes that surpass the technological limits on the edges of the divertor target elements. To reduce the heat load on the target elements, the addition of a “scraper element” (SE) in ten divertor locations is under investigation. Two aspects of thermal-fluid modeling for the W7-X SE components are analyzed in this paper using the ANSYS CFX 15.0 software: (1) sensitivity to turbulence models in single-phase and (2) two-phase modeling with the standard k-epsilon model. The results from the single-phase survey of seven turbulence models reveal a significant sensitivity to the selected model in the thermal-hydraulic analysis. Two-phase modeling is completed with the k-epsilon turbulence model, but no vapor generation is recorded. A significant temperature gradient between the liquid-solid interface indicates the need for a turbulence model that can achieve higher mesh resolution in the near-wall region.

Analysis of the Wendelstein 7-X test divertor unit scraper element with radiation shields

Early implementation of divertor components for the Wendelstein 7-X stellarator will include an inertially cooled system of divertor elements called the Test Divertor Unit (TDU). One part of this system is a scraper element that is intended to explore methods of mitigating heat flux on the ends of the TDU elements. This system will be in place in 2017, after a run period that will involve no divertor, and will precede steady state operation with actively cooled divertors scheduled for 2019. The TDU scraper element is an experimental device with uncertain requirements and with loading conditions which will developed as a part of the experiment. The pattern of heat flux may vary from currently predicted distributions and intensities. The design of the scraper element must accommodate this uncertainty. Originally the mechanical design was to be based on extensive studies for the monoblock-based design of an actively cooled system. An obvious simplification is the elimination of the manifolding needed for the water cooling. The wall panels on which the panels are mounted are to be maintained at 200C or less. Thermal ratcheting of the tiles, supporting structures, and backing structures is managed with adequate cooldown times, thermal anchors, where allowed, and radiative shields. Water cooling of the shields was proposed and rejected. Better radiation modeling is showing less need for multiple shields, but during initial run periods, the scraper element will have to be restricted to an acceptable operating envelope. Thermal instrumentation is recommended.