Research Needs for Magnetic Fusion Energy Sciences - US Burning ...

Promising new ideas have been developed for modifications to present-day divertors that may
substantially improve their capability to disperse high-heat flux and perhaps provide control of
elms. These new approaches include the super X divertor that guides the near-separatrix sol
flux tubes to a larger major radius to increase the surface area available for power deposition, and
the snowflake divertor concept that produces a second-order null in the poloidal magnetic field
at the X-point. engineering innovations to achieve superior target (tile) alignment can make possible
the effective use of small incidence angles, lowering peak power deposition. another promising
approach is the use of liquid metal (li, sn, Ga) divertor surfaces that can increase the heatflux
capability by flowing the heated material to a non-divertor cooling, and by eliminating most
erosion concerns. There are also ideas about flow through solids such as gas-injected carbon or
boron. all of these areas are only emerging concepts that require substantially more analysis and
definitive experimental tests. Given the need for a large improvement in this area, we advocate a
substantial program to systematically assess them.
TransienT impaCT
Can the impact of transients, long pulse, and elevated temperature operation be controlled?
The impact of transient events on plasma-material interactions with demo-relevant plasma
facing components is critically important. an unmitigated disruption can significantly erode
or damage plasma facing surfaces. in high confinement mode (h-mode) operation, repetitive
elm instabilities expel plasma thermal energy that heats divertor plasma facing component
surfaces and limits their service life. analysis shows that the maximum elm size that can be
tolerated without reaching the threshold energy density of ~1.0 (mJ/m 2 ) for significant erosion
is small in iteR (~10% core plasma energy content). likewise,vdes and runaway electrons
pose major challenges.
The tritium fuel cycle and vessel inventory is a significant concern for iteR due to potential
for co-deposition of tritium with eroded PFc material. in demo the issues of the tritium fuel
cycle are completely different due to the high-operating temperature. These effects (impact of
transients, long pulse, and elevated temperature) can be cost-effectively studied in dedicated,
non-toroidal experiments designed to understand the basic PWi processes in a controlled environment.
These can, for example, be operated in pulsed mode to simulate disruptions and
elms to better understand the effects of these phenomena on PWi over longer exposure times
than can be currently achieved in today’s fusion devices. similarly, the effect of extended pulse
lengths and elevated wall temperature on the PWi process can be addressed in simplified devices,
at lower cost, and much sooner, without requiring the complexity of the full tokamak
system. of course, the toroidal effects will also need to be investigated, and new advanced toroidal
devices will be required for those purposes. in particular, neither iteR nor the upcoming
asian long pulse tokamaks will have the optimum combination of hot wall, long pulses,
high-power density, and high-duty factor capability required to test solutions to these issues
in a toroidal system. While the specifications in terms of pulse length, duty factor, fractional
operation in deuterium (vs. hydrogen) and both flexibility and accessibility for Pmi and coreedge
integration studies need to be completed, it is clear there exists an important role in the
world program for such devices.
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