Measured signatures of low energy, physical sputtering in the line shape of neutral carbon emission

The most important mechanisms for introducing carbon into the DIII-D divertors [J.L. Luxon, Nucl. Fusion 42 (2002) 614] are physical and chemical sputtering. Previous investigations have indicated that operating conditions where one or the other of these is dominant can be distinguished by using CD and C2 emissions to infer C I influxes from dissociation of hydrocarbons and comparing to measured C I influxes. The present work extends these results through detailed analysis of the C I spectral line shapes. In general, it is found that the profiles are actually asymmetric and have shifted peaks. These features are interpreted as originating from a combination of an anisotropic velocity distribution from physical sputtering (the Thompson model) and an isotropic distribution from molecular dissociation. The present study utilitzes pure helium plasmas to benchmark C I spectral profiles arising from physical sputtering alone. © 2004 Elsevier B.V. All rights reserved

Measured signatures of low energy, physical sputtering in the line shape of neutral carbon emission

The most important mechanisms for introducing carbon into the DIII-D divertors [J.L. Luxon, Nucl. Fusion 42 (2002) 614] are physical and chemical sputtering. Previous investigations have indicated that operating conditions where one or the other of these is dominant can be distinguished by using CD and C2 emissions to infer C I influxes from dissociation of hydrocarbons and comparing to measured C I influxes. The present work extends these results through detailed analysis of the C I spectral line shapes. In general, it is found that the profiles are actually asymmetric and have shifted peaks. These features are interpreted as originating from a combination of an anisotropic velocity distribution from physical sputtering (the Thompson model) and an isotropic distribution from molecular dissociation. The present study utilitzes pure helium plasmas to benchmark C I spectral profiles arising from physical sputtering alone. © 2004 Elsevier B.V. All rights reserved

Controlling marginally detached divertor plasmas

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e = 5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER's divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B × ∇ B drift into the divertor makes this a significant challenge. If the divertor plasma were to reattach between ELMs, there would be a long (depending on delays in the gas puff system) window of high heat flux before detachment could be re-established. Thus, good understanding of detachment behavior near the threshold for re-attachment is required to properly tune an active control system to maintain ideal divertor performance without reattaching. The top-of-pedestal electron densities during dithering across the bifurcation and during stable marginally detached operation are the same within uncertainty, showing the need for local real-time measurements of the divertor conditions

ERO code benchmarking of ITER first wall beryllium erosion/re-deposition against LIM predictions

Previous studies (Carpentier et al 2011 J. Nucl. Mater. 415 S165-S169) carried out with the LIM code of the ITER first wall (FW) on beryllium (Be) erosion, re-deposition and tritium retention by co-deposition under steady-state burning plasma conditions have shown that, depending on input plasma parameter assumptions and sputtering yields, the erosion lifetime and fuel retention on some parts of the FW can be a serious concern. The importance of the issue is such that a benchmark of this previous work is sought and has been provided by the ERO code (Pitts et al 2011 J. Nucl. Mater. 415 S957-S964) simulations described in this paper. Provided that inputs to the codes are carefully matched, excellent agreement is found between the erosion/deposition profiles from both codes for a given ITER-shaped FW panel. Issues regarding the difficult problem of the correct treatment of Be sputtering are discussed in relation to the simulations. The possible influence of intrinsic Be impurity is investigated

Spectroscopy of Divertor Plasmas

The requirements for divertor spectroscopy are treated with respect to instrumentation and observations on present machines. Emphasis is placed on quantitative measurements.of impurity concentrations from the interpretation of spectral line intensities. The possible influence of non-Maxwellian electron distributions on spectral line excitation in the divertor is discussed. Finally the use of spectroscopy for determining plasma temperature, density, and flows is examined

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D