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

Abstract

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 (nmore » e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

Authors:
 [1];  [1];  [1];  [2];  [3];  [1];  [4];  [1];  [5];  [4];  [6];  [5];  [4];  [5];  [7];  [1];  [5];  [8];  [9];  [3] more »;  [10];  [4];  [11];  [1];  [2];  [12];  [10];  [2];  [4];  [2];  [2];  [2];  [1];  [6];  [4];  [1];  [2];  [2];  [5];  [1];  [2];  [13];  [12];  [4];  [2];  [6];  [1];  [6] « less
  1. General Atomics, San Diego, CA (United States)
  2. Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
  3. Univ. of Toronto, ON (Canada)
  4. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  5. Univ. of California at San Diego, La Jolla, CA (United States)
  6. Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
  7. Univ. of Texas, Austin, TX (United States)
  8. Univ. of Tennessee, Knoxville, TN (United States)
  9. Dalian Univ. of Technology, Liaoning (People's Republic of China)
  10. Princeton Univ., Princeton, NJ (United States)
  11. Aalto Univ., Espoo (Finland)
  12. Univ. of Wisconsin, Madison, WI (United States)
  13. Institute of Plasma Physics, Anhui (People's Republic of China)
Publication Date:
Research Org.:
Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); General Atomics, San Diego, CA (United States)
Sponsoring Org.:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
OSTI Identifier:
1325161
Report Number(s):
SAND-2016-4859J
Journal ID: ISSN 0029-5515; 640662
Grant/Contract Number:
AC04-94AL85000; AC02-09CH11466; FG02-07ER54917; FC02-04ER54698; AC52-07NA273441; AC05-00OR22725
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Fusion
Additional Journal Information:
Journal Volume: 56; Journal Issue: 12; Journal ID: ISSN 0029-5515
Publisher:
IOP Science
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; divertor concept; plasma-material interactions; DIII-D; advanced tokamak; fusion reactor

Citation Formats

Guo, H. Y., Hill, D. N., Leonard, A. W., Allen, S. L., Stangeby, P. C., Thomas, D., Unterberg, E. A., Abrams, T., Boedo, J., Briesemeister, A. R., Buchenauer, D., Bykov, I., Canik, J. M., Chrobak, C., Covele, B., Ding, R., Doerner, R., Donovan, D., Du, H., Elder, D., Eldon, D., Lasa, A., Groth, M., Guterl, J., Jarvinen, A., Hinson, E., Kolemen, E., Lasnier, C. J., Lore, J., Makowski, M. A., McLean, A., Meyer, B., Moser, A. L., Nygren, R., Owen, L., Petrie, T. W., Porter, G. D., Rognlien, T. D., Rudakov, D., Sang, C. F., Samuell, C., Si, H., Schmitz, O., Sontag, A., Soukhanovskii, V., Wampler, W., Wang, H., and Watkins, J. G.. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices. United States: N. p., 2016. Web. doi:10.1088/0029-5515/56/12/126010.
Guo, H. Y., Hill, D. N., Leonard, A. W., Allen, S. L., Stangeby, P. C., Thomas, D., Unterberg, E. A., Abrams, T., Boedo, J., Briesemeister, A. R., Buchenauer, D., Bykov, I., Canik, J. M., Chrobak, C., Covele, B., Ding, R., Doerner, R., Donovan, D., Du, H., Elder, D., Eldon, D., Lasa, A., Groth, M., Guterl, J., Jarvinen, A., Hinson, E., Kolemen, E., Lasnier, C. J., Lore, J., Makowski, M. A., McLean, A., Meyer, B., Moser, A. L., Nygren, R., Owen, L., Petrie, T. W., Porter, G. D., Rognlien, T. D., Rudakov, D., Sang, C. F., Samuell, C., Si, H., Schmitz, O., Sontag, A., Soukhanovskii, V., Wampler, W., Wang, H., & Watkins, J. G.. Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices. United States. doi:10.1088/0029-5515/56/12/126010.
Guo, H. Y., Hill, D. N., Leonard, A. W., Allen, S. L., Stangeby, P. C., Thomas, D., Unterberg, E. A., Abrams, T., Boedo, J., Briesemeister, A. R., Buchenauer, D., Bykov, I., Canik, J. M., Chrobak, C., Covele, B., Ding, R., Doerner, R., Donovan, D., Du, H., Elder, D., Eldon, D., Lasa, A., Groth, M., Guterl, J., Jarvinen, A., Hinson, E., Kolemen, E., Lasnier, C. J., Lore, J., Makowski, M. A., McLean, A., Meyer, B., Moser, A. L., Nygren, R., Owen, L., Petrie, T. W., Porter, G. D., Rognlien, T. D., Rudakov, D., Sang, C. F., Samuell, C., Si, H., Schmitz, O., Sontag, A., Soukhanovskii, V., Wampler, W., Wang, H., and Watkins, J. G.. 2016. "Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices". United States. doi:10.1088/0029-5515/56/12/126010. https://www.osti.gov/servlets/purl/1325161.
@article{osti_1325161,
title = {Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices},
author = {Guo, H. Y. and Hill, D. N. and Leonard, A. W. and Allen, S. L. and Stangeby, P. C. and Thomas, D. and Unterberg, E. A. and Abrams, T. and Boedo, J. and Briesemeister, A. R. and Buchenauer, D. and Bykov, I. and Canik, J. M. and Chrobak, C. and Covele, B. and Ding, R. and Doerner, R. and Donovan, D. and Du, H. and Elder, D. and Eldon, D. and Lasa, A. and Groth, M. and Guterl, J. and Jarvinen, A. and Hinson, E. and Kolemen, E. and Lasnier, C. J. and Lore, J. and Makowski, M. A. and McLean, A. and Meyer, B. and Moser, A. L. and Nygren, R. and Owen, L. and Petrie, T. W. and Porter, G. D. and Rognlien, T. D. and Rudakov, D. and Sang, C. F. and Samuell, C. and Si, H. and Schmitz, O. and Sontag, A. and Soukhanovskii, V. and Wampler, W. and Wang, H. and Watkins, J. G.},
abstractNote = {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). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.},
doi = {10.1088/0029-5515/56/12/126010},
journal = {Nuclear Fusion},
number = 12,
volume = 56,
place = {United States},
year = 2016,
month = 9
}

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  • 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, whichmore » 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). As a result, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less
  • 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, whichmore » 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 (ne/nGW), 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.« less
  • Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less
  • Unique diagnostic and access features of the DIII-D tokamak, including a sample exposure system, have been used to carry out controlled and well-diagnosed plasma-surface interactions (PSI) experiments. An important contribution of the experiments has been the ability to link a given plasma exposure condition to a measured response of the plasma-facing surface and to thus understand the interaction. This has allowed for benchmarking certain aspects of erosion models, particularly near-surface particle transport. DIII-D has empirically quantified some of the PSI effects that will limit the operation availability and lifetime of future fusion devices, namely, net erosion limiting divertor plate lifetimemore » and hydrogenic fuel retention in deposit layers. Cold divertor plasmas obtained with detachment can suppress net carbon divertor erosion, but many low-temperature divertor PSI phenomena remain poorly understood: nondivertor erosion sources, long-range particle transport, global erosion/deposition patterns, the enhancement of carbon erosion with neon impurity seeding, the sputtered carbon velocity distribution, and the apparent suppression of carbon chemical erosion in detachment. Long-term particle and energy fluences have reduced the chemical erosion yield of lower-divertor tiles. Plasma-caused modification of a material's erosion properties, including material mixing, will occur quickly and be important in long-pulse fusion devices, making prediction of PSI difficult in future devices.« less