Image courtesy of Vlad Soukhanovskii
The National Spherical Torus Experiment Device at Princeton Plasma Physics Laboratory (left), and a schematic of magnetic field lines in the snowflake divertor configuration (right).
Heat escaping from the core of a twelve-million degree nuclear fusion plasma device was successfully contained by a snowflake-shaped magnetic field to mitigate its impact on device walls.
One of the grand challenges of the magnetic fusion research is to “tame the plasma-material interface”—to develop an interface between the hot plasma in the nuclear fusion reactor and the low-temperature material wall. The novel snowflake configuration diverts and dissipates heat lost from the high-temperature core plasma, thus alleviating thermal loads on the material wall.
Strong magnetic fields shape the hot plasma in the form of a donut in a magnetic fusion plasma reactor called a tokamak. Confined plasma particles move along infinite magnetic field lines inside the tokamak. Some particles and heat, however, tend to escape because of transport and magnetohydrodynamic plasma instabilities. A separate part of the vacuum vessel called a “divertor chamber” is used to divert away and collect lost heat and particles. If the plasma incident on the divertor surface is too hot, melting of the plasma-facing components and loss of coolant can occur. Under such undesirable conditions, the plasma-facing component lifetime would also be an issue, as they would tend to erode too quickly. The snowflake divertor concept was developed theoretically by Dmitri Ryutov and colleagues within the Fusion Energy Sciences Program at Lawrence Livermore National Laboratory (LLNL). The experiments led by LLNL scientists on the National Spherical Torus Experiment (NSTX) and DIII-D tokamak user facilities at Princeton Plasma Physics Laboratory and General Atomics, respectively, confirmed that all predicted magnetic properties could be realized without any additional hardware. The experiments at NSTX and DIII-D demonstrated a drastic reduction of heat load on divertor plasma-facing components and compatibility with high performance high confinement core plasma regimes. These, as well as other on-going experimental and numerical modeling efforts in USA, Switzerland, Italy and China, provide support to the snowflake divertor configuration as a viable plasma-material interface for future tokamak devices and for fusion development applications.
V. A. Soukhanovskii
Lawrence Livermore National Laboratory
Office of Science Fusion Energy Sciences (FES) program
Ryutov, D., Geometrical properties of a snowflake divertor Phys. Plasmas 14, 064502 (2007).
Ryutov, D. et al., The magnetic field structure of a snowflake divertor, Phys. Plasmas 15 (2008) 092501.
Soukhanovskii, V. et al., Taming the plasma-material interface with the snowflake divertor in NSTX, Nucl. Fusion 51 (2011) 012001.
Soukhanovskii, V. A. et al., Snowflake divertor configuration studies in NSTX, Phys. Plasmas 19 (2012) 082504.
Allen, S. L. et al., Initial Snowflake Divertor Physics Studies on DIII-D, Paper PD/1-2, IAEA FEC 2012.
DIII-D user facility: https://fusion.gat.com/global/DIII-D
NSTX user facility: http://nstx.pppl.gov/
DOE Laboratory, Industry, SC User Facilities, FES User Facilities, DIII-D, NSTX