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Title: Fusion nuclear science facilities and pilot plants based on the spherical tokamak

Abstract

Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is $${{R}_{0}}\geqslant 1.7$$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

Authors:
 [1];  [1];  [2];  [1];  [3];  [4];  [5];  [1];  [1];  [6];  [7];  [2];  [2];  [1];  [6];  [2];  [4];  [1];  [6];  [7] more »;  [7];  [1];  [2];  [8];  [2];  [1];  [1];  [1];  [9];  [10];  [7];  [1] « less
  1. Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
  2. Univ. of Wisconsin, Madison, WI (United States)
  3. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  4. Culham Science Centre, Oxfordshire (United Kingdom)
  5. Univ. of Washington, Seattle, WA (United States)
  6. Tokamak Energy Ltd., Oxfordshire (United Kingdom)
  7. Univ. of Texas at Austin, Austin, TX (United States)
  8. College of William and Mary, Williamsburg, VA (United States); Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
  9. Columbia Univ., New York, NY (United States)
  10. Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)
Publication Date:
Research Org.:
Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
Sponsoring Org.:
USDOE Office of Science (SC), Fusion Energy Sciences (FES) (SC-24)
OSTI Identifier:
1335165
Report Number(s):
5280
Journal ID: ISSN 0029-5515
Grant/Contract Number:
EP/I501045; AC02-09CH11466
Resource Type:
Journal Article: Accepted Manuscript
Journal Name:
Nuclear Fusion
Additional Journal Information:
Journal Volume: 56; Journal Issue: 10; Journal ID: ISSN 0029-5515
Publisher:
IOP Science
Country of Publication:
United States
Language:
English
Subject:
70 PLASMA PHYSICS AND FUSION TECHNOLOGY; fusion nuclear science facility; pilot plant; spherical tokamak; tritium breeding; negative neutral beams; super-X divertor; high-temperature superconductors

Citation Formats

Menard, J. E., Brown, T., El-Guebaly, L., Boyer, M., Canik, J., Colling, B., Raman, R., Wang, Z., Zhai, Y., Buxton, P., Covele, B., D’Angelo, C., Davis, A., Gerhardt, S., Gryaznevich, M., Harb, M., Hender, T. C., Kaye, S., Kingham, D., Kotschenreuther, M., Mahajan, S., Maingi, R., Marriott, E., Meier, E. T., Mynsberge, L., Neumeyer, C., Ono, M., Park, J. -K., Sabbagh, S. A., Soukhanovskii, V., Valanju, P., and Woolley, R.. Fusion nuclear science facilities and pilot plants based on the spherical tokamak. United States: N. p., 2016. Web. doi:10.1088/0029-5515/56/10/106023.
Menard, J. E., Brown, T., El-Guebaly, L., Boyer, M., Canik, J., Colling, B., Raman, R., Wang, Z., Zhai, Y., Buxton, P., Covele, B., D’Angelo, C., Davis, A., Gerhardt, S., Gryaznevich, M., Harb, M., Hender, T. C., Kaye, S., Kingham, D., Kotschenreuther, M., Mahajan, S., Maingi, R., Marriott, E., Meier, E. T., Mynsberge, L., Neumeyer, C., Ono, M., Park, J. -K., Sabbagh, S. A., Soukhanovskii, V., Valanju, P., & Woolley, R.. Fusion nuclear science facilities and pilot plants based on the spherical tokamak. United States. doi:10.1088/0029-5515/56/10/106023.
Menard, J. E., Brown, T., El-Guebaly, L., Boyer, M., Canik, J., Colling, B., Raman, R., Wang, Z., Zhai, Y., Buxton, P., Covele, B., D’Angelo, C., Davis, A., Gerhardt, S., Gryaznevich, M., Harb, M., Hender, T. C., Kaye, S., Kingham, D., Kotschenreuther, M., Mahajan, S., Maingi, R., Marriott, E., Meier, E. T., Mynsberge, L., Neumeyer, C., Ono, M., Park, J. -K., Sabbagh, S. A., Soukhanovskii, V., Valanju, P., and Woolley, R.. 2016. "Fusion nuclear science facilities and pilot plants based on the spherical tokamak". United States. doi:10.1088/0029-5515/56/10/106023. https://www.osti.gov/servlets/purl/1335165.
@article{osti_1335165,
title = {Fusion nuclear science facilities and pilot plants based on the spherical tokamak},
author = {Menard, J. E. and Brown, T. and El-Guebaly, L. and Boyer, M. and Canik, J. and Colling, B. and Raman, R. and Wang, Z. and Zhai, Y. and Buxton, P. and Covele, B. and D’Angelo, C. and Davis, A. and Gerhardt, S. and Gryaznevich, M. and Harb, M. and Hender, T. C. and Kaye, S. and Kingham, D. and Kotschenreuther, M. and Mahajan, S. and Maingi, R. and Marriott, E. and Meier, E. T. and Mynsberge, L. and Neumeyer, C. and Ono, M. and Park, J. -K. and Sabbagh, S. A. and Soukhanovskii, V. and Valanju, P. and Woolley, R.},
abstractNote = {Here, a fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR ≈ 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R 0 for a range 1 m–2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is ${{R}_{0}}\geqslant 1.7$ m, and a smaller R 0 = 1 m ST device has TBR ≈ 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.},
doi = {10.1088/0029-5515/56/10/106023},
journal = {Nuclear Fusion},
number = 10,
volume = 56,
place = {United States},
year = 2016,
month = 8
}

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  • A potentially attractive next-step towards fusion commercialization is a pilot plant, i.e. a device ultimately capable of small net electricity production in as compact a facility as possible and in a configuration scalable to a full-size power plant. A key capability for a pilot-plant programme is the production of high neutron fluence enabling fusion nuclear science and technology (FNST) research. It is found that for physics and technology assumptions between those assumed for ITER and nth-of-a-kind fusion power plant, it is possible to provide FNST-relevant neutron wall loading in pilot devices. Thus, it may be possible to utilize a singlemore » facility to perform FNST research utilizing reactor-relevant plasma, blanket, coil and auxiliary systems and maintenance schemes while also targeting net electricity production. In this paper three configurations for a pilot plant are considered: the advanced tokamak, spherical tokamak and compact stellarator. A range of configuration issues is considered including: radial build and blanket design, magnet systems, maintenance schemes, tritium consumption and self-sufficiency, physics scenarios and a brief assessment of research needs for the configurations.« less
  • Recent progress(1) in plasma science of the Spherical Tokamak (or Spherical Torus, ST)(2) has indicated relatively robust plasma conditions in a broad number of topical area including strong shaping, stability limits, energy confinement, self-driven current, and sustainment. This progress has enabled an extensive update of the plasma science and fusion engineering conditions of a Component Test Facility (CTF)(3), which is potentially a necessary step in the development of practical fusion energy. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over sizescale > 105 cm2 and depth-scale > 50more » cm, delivering > 3 accumulated displacement per atom (dpa) per year(4). Such chamber conditions are calculated to be achievable in a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~ 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. The ST CTF will test the life time of single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of relatively low cost CTF devices to suit a range of fusion engineering science test missions.« less
  • This article is a response to the Office of Energy Research of the US DOE from the Fusion Energy Advisory Committee on a review of major US magnetic confinement facilities. This response was solicited in response to one of the suggestions made as part of the advisory report `A Restructured Fusion Energy Sciences Program` submitted to the US DOE in early 1996.
  • A broadly based study of the fusion engineering and plasma science conditions of a Component Test Facility (CTF),1 using the Spherical Torus or Spherical Tokamak (ST) configuration,2 have been carried out. The chamber systems testing conditions in a CTF are characterized by high fusion neutron fluxes n > 4.4 1013 n/s/cm2, over size scales > 105 cm2 and depth scales > 50 cm, delivering > 3 accumulated displacement per atom (dpa) per year.3 The desired chamber conditions can be provided by a CTF with R0 = 1.2 m, A = 1.5, elongation ~ 3, Ip ~ 9 MA, BT ~more » 2.5 T, producing a driven fusion burn using 36 MW of combined neutral beam and RF power. Relatively robust ST plasma conditions are adequate, which have been shown achievable4 without active feedback manipulation of the MHD modes. The ST CTF will test the single-turn, copper alloy center leg for the toroidal field coil without an induction solenoid and neutron shielding, and require physics data on solenoid-free plasma current initiation, ramp-up, and sustainment to multiple MA level. A new systems code that combines the key required plasma and engineering science conditions of CTF has been prepared and utilized as part of this study. The results show high potential for a family of lower-cost CTF devices to suit a variety of fusion engineering science test missions.« less