The Hybrid Illinois Device for Research and Applications (HIDRA) is a toroidal plasma device at UIUC, formerly known as WEGA when operated in Greifswald. The HIDRA vacuum vessel has a circular cross section and a major radius of R = 0.72 m and a minor radius a = 0.19 m, with a steady state toroidal magnetic field BT < 0.5 T. A limiter can be used with reduced plasma minor radius between 0.10 – 0.15 m. Since HIDRA has the ability for long pulse steady state operation via the classical stellarator configuration, HIDRA has an actual toroidal magnetic field, just like a tokamak. HIDRA also has the capability to operate as a tokamak and thus a pulsing capability during steady operation allows simulation of transient events. Initial plasmas will use 2.45 GHz magnetron heating up to 26 kW and should achieve Te ~ 20 eV and ne ~ 1×1018 m-3. Even though these plasma parameters are much lower than that of larger devices like EAST, the plasma and magnetic fields at the first wall are very close to those produced in HIDRA. The steady state and pulsed capabilities of HIDRA make it an ideal test bed for liquid Li science and technology, where flow, ejection and recycling can be assessed.
Fusion is the process of taking two light atoms and providing them with enough energy that they are able to fuse to form a heavier atom. This is the opposite to fission where atoms are split. This is the energy source that drives the stars! One example of a fusion reaction is the formation of helium from the fusing of deuterium and tritium and releasing 17.6 MeV of energy.
D + T → He + n + 17.6 MeV
Energetically this is the easiest of the fusion reactions and there are many other fusion reactions, that in the future may be utilized. Many universities and laboratories around the world have pursued the dream of fusion energy since the 1950’s. It promises to be a clean source of energy producing no CO2 and no radioactive waste. However, technologically fusion has proven to be quite complicated to achieve over the years, just imagine how do you confine a plasma that is over 100 million oC. There are two main magnetic confinement devices that are studies, tokamaks and stellarators.
Tokamaks use a toroidal magnetic field and a poloidal magnetic field induced from a current that is driven within the plasma itself to form the confining magnetic bottle. The stellarator uses all external magnetic fields that are generated through various magnets, and some of these can get very complicated indeed, as seen with W7-X!
It turns out that the interactions between the plasma and surfaces in fusion reactors play a major role in the ability for these machines to operate successfully with good confinement, minimal transient events and being able to withstand the particle and heat flux to the surfaces. Power flux at steady state operation can be as high as 10 MWm-2 and during a transient event such as a disruption or ELM can get as high as 100 MWm-2. Most materials cannot withstand this and so new ways to protect surfaces or new materials that can withstand the fluxes need to be developed.
HIDRA will be at the forefront of studying plasma material interactions and developing the technology needed for innovative plasma facing components that will make fusion a viable energy source of the future.
 D. Andruczyk, D. N. Ruzic, D. Curreli, J. P. Allain and the HIDRA Team, Fusion Sci. Technol. 68 (2015), 497.
 M. Otte, D. Andruczyk, E. Holzhauer, et al., AIP Conf. Proc. 3 (2008) 993.
 H. P. Laqua, D. Andruczyk, S. Marsen, et al., Proc. 22nd IAEA Fusion Energy Conf. Geneva, 13 – 18 Oct, (2008).
 M. Otte, H. P. Laqua, E. Chlechowitz, et al., Nucleonika, 57 (2012) 171.
 J. Chung, R. König, J. Howard, et al., Plasma Phys. Control. Fusion, 47 (2005) 919.
As part of the Plasma Surface Interactions: Bridging from the Surface to the Micron Frontier through Leadership Class Computing SciDAC project, we are working to better understand how plasma facing components (PFCs) will interact with burning fusion plasmas. Researchers at CPMI are investigating processes like sputtering, re-deposition, and evolving surface roughness with Monte Carlo, binary collision approximation (BCA) codes. We have designed and developed several surface roughness models for TRIDYN (resulting in the Fractal TRIDYN code, or FTRIDYN) and are using these to investigate the role surface roughness plays in ballistic ion-surface interactions.
Click to Enlarge: D on Be Sputtering Yield
W. Möller, W. Eckstein, J.P. Biersack, Tridyn-binary collision simulation of atomic collisions and dynamic composition changes in solids, Computer Physics Communications, Volume 51, Issue 3, November 1988, Pages 355-368, ISSN 0010-4655, 10.1016/0010-4655(88)90148-8.
D.N. Ruzic and H.K. Chiu, J. Nucl. Mater. 162-164 (1989) 904.
R.P. Doerner, D. Nishijima, T. Schwarz-Selinger, Nucl. Fusion 52 (2012) 103003
Study of liquid lithium has increased as of late for its applicability in fusion energy. Liquid lithium implemented in the divertor of a fusion device may prove to be the path to a fusion device that produces significantly more energy than it consumes. Liquid lithium benefits a fusion device by gettering cold hydrogenic species at the wall of a fusion device, raising edge temperatures, and increasing energy confinement times. However, this effect subsides as the lithium passivates, or reacts to form different compounds. Maintenance of a clean lithium surface therefore is important, and so is the study of clean lithium surfaces.
The lithium injector developed at UIUC imparts the ability to place controlled amounts of visibly impurity-free liquid lithium into a vacuum chamber. Loaded with lithium that may be partially oxidized on the surface, ejection through a nozzle ensures via the high surface tension of lithium that the oxide, hydroxide, and hydride impurities that form on lithium surfaces are confined to the injector tube, and that only lithium exits. This injector is used in the Materials Characterization Test Stand (MCATS) chamber to study the contact angle of liquid lithium, as well as used to fill the Liquid-Metal Infused Trenches (LiMIT) that are implemented on the Solid-Liquid Lithium Divertor Experiment (SLiDE) chamber with clean lithium.
Liquid metals have received increased attention within the fusion community as of late. Liquid lithium, especially, has been the target of much interest for its ability to getter impurities and cold hydrogenic species at the walls of fusion devices. Lithium has been shown in several fusion devices to increase energy confinement times and to reduce the frequency of edge localized modes, plasma instabilities which cause large deposits of energy to the first wall of fusion devices. Inclusion of liquid metals in a fusion device requires, however, that the liquid metal wet the substrate on which it is placed. Beading of the liquid is undesirable.
The Materials Characterization Test Stand (MCATS) chamber at CPMI was specifically designed to investigate the wetting phenomena of liquid metals on various fusion relevant substrates. A moveable stage mounted either on a plate heater, or with strip heaters attached to the back side allows for placement of several droplets of liquid metal on various surfaces, in order to study the contact angle of a liquid metal on the surface. Contact angles of liquid lithium on various fusion relevant surfaces have been studied, as well as methods for the reduction of the critical wetting temperature of lithium on these surfaces.
MCATS also allows for the study of the compatibility of liquid metals with various solid surfaces. A pool of liquid metal in a stainless steel cup is mounted on the stage. A rotary motion feedthrough driven by an external motor allows for study of erosion of different solid materials in liquid metal. Most recently, this device showed the strong attack of copper by liquid gallium.
Exposure to very large heat fluxes as well as large radiation loads inducing strong wear threatens to limit the lifetime of solid plasma facing component materials in fusion energy devices. Liquid metal devices, however, do not suffer the same ill effects. Liquid lithium, in particular, has shown promise as a potential candidate for a plasma facing component material. Several methods exist to introduce lithium into fusion devices, however, one of the most unique methods of doing so is the Liquid-Metal Infused Trenches concept of CPMI. LiMIT relies on thermoelectric magnetohydrodynamics (TEMHD) to propel a liquid metal down a series of trenches. TEMHD flow is based on the interaction of a thermoelectric current with a magnetic field.
The source of the thermoelectric current in the LiMIT device arises from the junction between the flowing liquid metal and the trench wall material. A thermal gradient across this interface gives rise to the thermoelectric current. An analogy may be constructed by considering the interface to be composed of two junctions, the top portion of the trench would constitute the hot junction of a thermocouple, while the bottom portion of the trench would constitute the cold junction. Since the interfacial voltage between the two materials is a function of temperature, the temperature difference gives rise to a voltage difference, which drives a current.
The magnitude of the current driven is proportional to the relative thermopower, or the difference in Seebeck coefficient, between the two materials. To generate data on the Seebeck coefficient of a wide variety of materials, an apparatus was constructed at CPMI. Shown in the photo above, measurement of the Seebeck coefficient of many fusion relevant materials is possible. A library of Seebeck coefficients is being compiled, and measurements are ongoing.
In order to get undergraduate and even high school students interested in plasma science, the CPMI often gives tours to highlight our research projects. As part of these tours, plasma demonstrations are shown so that students get a better understanding of the power of plasma!
DC Glow Chamber
Arc mode in DC Glow Chamber
Originally completed in May of 2009, the Solid/liquid lithium divertor experiment (SLiDE) was designed with intention of studying surface tension driven flows in fusion relevant magnetic fields and conditions which would be applicable to divertors and limiters. The work was spurred on by the findings from the Current Drive Experiment-Upgrade (CDX-U) which is housed at Princeton Plasma Physics Laboratory (PPPL). The CDX-U experiment reported the ability of flowing liquid lithium to handle an electron beam generated heat spot of 60 MW*m-2 without notable evaporation. Later a series of experiments on SLiDE project conducted by Dr. Michael Jaworski proved that the strong heat flux mitigation ability of liquid lithium surface is due to thermoelectric magnetohydrodynamics (TEMHD) driven flow. Strong liquid lithium swirling flow was discovered in experiments, which is the first time that TEMHD driven flow is directly proved since it was firstly raised by Shercliff in 1979 . [Video of Lithium Swirling]
Thermoelectric effect, or called Seebeck effect which is commonly known as the key phenomenon for thermocouples, means the temperature gradient along the interface of two different types of material with different thermoelectric coefficient (Seebeck coefficient) can generate a current through the material. When a transverse magnetic field exists and one of the material is in liquid phase the Lorentz force can obviously drive the liquid to flow. Both Lorentz driven force and MHD damping affecting the flow is so called TEMHD.
This phenomenon can be applied to drive the liquid metal in fusion reactors. It has been widely accepted that liquid lithium be used since its ability to flow is extremely beneficial to the plasma facing surface of divertor or limiter in fusion reactors. The flowing liquid lithium surface in fusion reactors has many advantages such as the abilities to lower the impurity, suppress the recycling and transfer the heat. However the magnetic field in fusion reactors strongly damps the liquid metal flow. The Li-Metal Infused Trench (LiMIT) concept was raised to utilize TEMHD to drive the liquid lithium flow in fusion reactors .
In the LiMIT design, many narrow trenches are built in parallel with millimeter thick gaps between each other and the trenches are placed radially as the divertor target plate. The toroidal field is perpendicular to the side walls of trenches and the center surface receives narrow heat flux. Once the temperature gradient establishes between the top and bottom of the trench structure a thermoelectric current is generated in the opposite direction of the temperature gradient. Since this current is perpendicular to the toroidal field the lithium flow is driven along the trenches in the radial direction across the narrow heat flux strike point.
TEMHD driven liquid lithium flow in trenches was proved on SLiDE project. With the help of the fast speed camera the movement of the surface can be directly recorded. The lithium flow seems fast and smooth (0.22 ± 0.03 m/s) when the field is low (0.0589 T) while it becomes slower (0.13 ± 0.02 m/s) and having more surface variation when the field is increased (0.19 T). [Video of High Flow] [Video of Low Field Flow] Other than the function of keeping the liquid lithium surface fresh to face the plasma the heat transfer ability of this TEMHD driven trench flow is also observed by IR camera and embedded thermocouple measurements. IR images show that the uneven temperature distribution always follows the flow direction.
The results from SLiDE project initiate worldwide interests from many fusion research institutes which enable the international collaborations to examine the LiMIT concept on important fusion related devices such as HT-7, EAST, Magnum-PSI and LTX. The test of LiMIT concept as a movable limiter on HT-7 proves that liquid lithium can be driven by TEMHD effect in fusion relevant magnetic field. Future experiments of LiMIT concept on EAST, Magnum-PSI and LTX will further investigate the flow and related heat transfer in strong magnetic fields and how the trench flow will affect the fusion plasma.
 M.A. Jaworski, T.K. Gray, M. Antonelli, J.J. Kim, C.Y. Lau, M.B. Lee, M.J. Neumann, W. Xu and D.N. Ruzic, 104 (2010) 094503
 D.N. Ruzic, W. Xu, D. Andruczyk and M.A. Jaworski, Nuclear Fusion, 51 (2011) 102002
(Updated June 2013)
The basic principle of Laser Assisted Plasma Coating at Atmospheric Pressure (LAPCAP) is to utilize a pulsed ND:YAG laser (t= 5~10 ns, with the intensity of 10^10~10^11 W/cm^2) to knock the atomic particles from the target and feed into the inductively coupled low temperature RF or microwave atmopheric plasma to further increase the efficiency of the deposition process through charged plasma particles such as ions and electrons. Further ,the low-temperature plasma has an advantage of acting as a heating source for the substrate to be coated to have improved coating. This system will allow a high quality, non-porous and uniform coatings on variety of substrates.
High Temperature Plasma (~1,000°C) Assisted Laser Ablation YSZ is rotating at 2 rpm
Low Temperature Plasma Deposition (~100°C) High Temperature Plasma Deposition (1,000°C)
The plasma facing components in fusion reactors are subject to a number of high flux/high energy events during normal operation of the device. The Divertor Edge and Vapor shielding eXperiment (DEVeX) was developed to produce a pulsed, high density, high temperature plasmas capable of simulating the conditions in events such as type I Edge Localized Modes (ELMs). Several theoretical models predict that evaporated and ejected material from the surface of the PFCs will become ionized during these events forming a vapor cloud, which will absorb some of the incident plasma energy and reduce the erosion and heat loading on the surface of the PFCs. The end goal of DEVeX is to study the dynamics of the vapor cloud that is formed and provide a benchmark for the theoretical models.
Lithium magnetron on DEVeX running in an argon atmosphere to deposit lithium on the target surface.
Previous work conducted on the ESP-gun experiment showed that in order to produce high density plasmas in a theta-pinch, a fast rise time (< 2 microseconds), high current capacitor bank is required. Subsequently, a capacitor bank is being designed and built that can operate up to 60 kV with 64 kJ of stored energy, which should be more than adequate to produce the desired plasmas.
Several diagnostics, such as a triple langmuir probe (TLP) and an optical spectrometer, are present in the target chamber to measure the density and temperature of the simulated ELM plasma. Additional diagnostics, such as an infrared temperature diagnostic, are planned as the experiment progresses. In order to investigate material erosion by exposure to a burst of high density plasma in a laboratory setting, plasma diagnostics and target analysis of a theta pinch device are carried out. A series of quadruple-probe diagnostics show that plasma sustains approximately for 100 μs at each pulse, with 1.0±0.2(10)21 /m3 plasma density and 12.5±2.5 eV electron temperature. However, when preionization takes place in conjunction with the main bank discharge, density and electron temperature show a few spikes at 1021-1022 /m3 and 30±10 eV. The temperature rise of the target is measured by thermocouple and corresponds to the energy deposition estimated by plasma parameters.
Based on the measurements of the incident plasma parameters, it is shown that the device produces 1021-1022 /m3 and 20-40 eV plasma parameters at various pressure ranges. The temperature changes measured by a thermocouple yield 8±0.3 oC and 33±5 oC for non-preionization and preionzation and they correspond to 3.5±0.1 J, 14±0.5 MW/m2 and 14±2.2 J, 58±8.7 MW/m2 respectively. It is in agreement with the calculated energy delivery 2.5±1.7 J and 21±14 J predicted by from measured plasma parameters.
In the near future, DEVeX will be incorporated with the current generation SLiDE project where it will be used or the pushing of divertor/limiter plasma facing components. Such work is greatly important as in the edge and divertor region of the plasma, tokamaks experience off-normal events that cause intolerable damage to components of the reactors, which will become a much more serious issue in ITER. To simulate extreme events in a tokamak and provide a test-stand for liquid lithium plasma-facing components, a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) device will provide fusion-relevant plasmas incident on targets with flowing liquid lithium surfaces.
Theta coil, Pyrex tube, optical diagnostics, guiding magnets
(Updated June 2013)
Crucial to the understanding of plasma material interactions is the interaction of ions with surfaces. Acceleration of ions through the plasma sheath, as well as bombardment from high energy ions heated within the bulk plasma contributes heavily to the modification of surfaces exposed to plasma, particularly high energy plasmas, such as those seen in fusion.
Equipped with two different ion guns, the Ion-Surface Interaction Experiment has the ability to study a wide range of scenarios involving ion bombardment on surfaces. A Colutron ion gun allows for velocity and neutral filtered ion bombardment between 1 eV and 2 keV. A NTI lithium ion gun adds to the ion bombardment capability of the system. Magnetic sector and quadrupole residual gas analyzers as well as a quartz crystal microbalance are included to diagnose the interaction of accelerated ions with various surfaces.
The diagnostic suite allows for monitoring of sputtering yields as well as desorption of gaseous chemical sputtering products. IIAX in tandem with the Material Research Laboratory at the University of Illinois can also investigate changes in surface chemistry and morphology from ion bombardment.