• Attempt to achieve long-pulse density control for increased neutral beam current drive fraction using improved fueling and lithium conditioning
• Develop high non-inductive current fraction plasmas with high-beta and high bootstrap fraction under sustained conditions
• Assess the dependence of integrated plasma performance on collisionality (Milestone R11-2)
• Develop and implement improved plasma control techniques to achieve advanced operating scenarios

Research Milestone R(11-2):  Assess the dependence of integrated plasma performance on collisionality.

The high performance scenarios assumed for next-step ST devices such as NHTX and ST-CTF are based on operating at lower Greenwald density fraction and significantly lower pedestal collisionality than NSTX. Building on the research of the FY2010 boundary physics milestone R(10-3), this milestone would extend research on high-performance plasmas toward lower density and collisionality and systematically assess integrated performance (such as non-inductive current fraction, confinement, core and pedestal stability, pulse-duration, impurity content) of long-pulse H-mode plasmas as a function of density and collisionality. Two possible tools for accessing reduced plasma collisionality are the Liquid Lithium Divertor (LLD) and the upgraded HHFW system capable of higher power and with resilience to ELMs. Assuming it has successfully demonstrated particle pumping in FY2010, the LLD would be utilized to vary plasma density and temperature by varying its pumping through control of parameters such as the strike-point position, flux expansion, the temperature, and thickness of the lithium layer. Further, the plasma integrated performance would be assessed as a function of boundary shape - in particular the strike point location and triangularity - to assess the possible trade-off between improved MHD stability (higher triangularity) and increased pumping efficiency (lower triangularity). Assuming the upgraded HHFW system has demonstrated sustained core electron heating during high-power NBI-heated deuterium H-mode in FY2010, HHFW power would be assessed for increasing the bootstrap current and the neutral-beam current-drive in long-pulse H-mode scenarios. The influence of these advanced pumping and heating capabilities on NSTX high-performance plasmas will be compared to time-dependent simulation codes such as TSC and TRANSP to develop a predictive capability for advanced ST operating scenarios.

ITPA Participation

• IOS-5.1  Ability to obtain and predict off-axis NBCD
• IOS-5.2  Maintaining ICRH Coupliing in expected ITER Regime

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