MEETING REPORT
SUMMARY OF WORKSHOP ON 3. potential for stronger vertical stability and fewer or ESTABLISHING THE PHYSICS BASIS no disruptions, easing design of first wall, blanket, NEEDED TO ACCESS THE POTENTIAL and shield structure OF COMPACT TOROIDAL REACTORS, 4. possible cost advantages due to smaller core size and OAK RIDGE, TENNESSEE, lower magnetic fields JULY 19-22, 1994 5. physics that permits, if desirable, higher wall load- ing in smaller unit size INTRODUCTION 6. possibly the only configuration that permits viable advanced fuel reactors. A workshop on establishing the physics basis needed to access the potential of compact toroidal reactors was held Major issues for this configuration are July 19-21, 1994, at Oak Ridge National Laboratory, Oak 1. unshielded single-turn normal-conducting center leg Ridge, Tennessee. The presentations and discussions at the for the toroidal field coil (TFC), which requires reg- workshop covered the low-aspect-ratio (A) tokamak and ular replacement (at fluences <10 MW-yr/m2) and the spheromak. The findings of the workshop, summarized generates added radioactive waste here, covered 2. divertor design and power handling, which are yet to 1. reactor and pathway be developed for any A and may be more challeng- 2. plasma confinement ing for low A 3. magnetohydrodynamic (MHD) equilibrium and 3. recirculating power associated with normal-conduct- global stability ing center leg and current drive, which must be minimized. 4. MHD beta-limit studies Systems code calculations for deuterium-tritium and 5. plasma heating and current drive advanced-fuel low-^4 reactors using the presented physics as- 6. divertor power and particle handling sumptions were recommended by the workshop. The \ow-A tokamak offers the possibility of a steady- 7. present device experimental plans state high-volume plasma neutron source (HVPNS) that 2 8. action items. produces neutron wall loading of 1 to 2 MW/m in small, driven, high-current plasmas (R0 « 0.8 to 1.0 m, Ip « 6 to The workshop benefited from advice from R. A. Blanken 9 MA). Such a device provides a reduced-cost approach for and S. A. Eckstrand of the Office of Fusion Energy of the testing full-function power blankets if high neutron fluence U.S. Department of Energy (DOE). («6 MW-yr/m2) can be obtained. It would complement the capability of the International Thermonuclear Experi- REACTOR AND PATHWAY mental Reactor (ITER) in testing full-size blankets to a lim- ited fluence (~1 MW-yr/m2). Reliable power blankets are Tokamak needed to achieve high duty factors in a fusion demonstra- tion power plant. The low-^4 tokamak provides the following reactor ad- vantages relative to the standard and high-^4 tokamaks: Spheromak 1. better access for remote maintenance of blanket, shield, and divertor The potential advantages for the spheromak are 2. simpler system requiring no inboard blanket, no 1. no need for TFCs, leading to a simplified torus ohmic, and possibly no poloidal divertor coils geometry
210 FUSION TECHNOLOGY VOL. 29 MAR. 1996 210 2. helicity injection current drive, which is projected to Spheromak be efficient in reactor level devices, assuming devel- For the spheromak, the most important need is to de- opment of steady-state helicity injection with small termine core energy confinement in a steadily sustained, energy losses due to the induced magnetic turbulence stable discharge. Spheromak experimental physics made sig- 3. the possibility of combining the divertor, helicity in- nificant progress during the decade in which the United jector for current drive, and fueling system and po- States conducted the research; an electron temperature of sitioning them near the machine axis (small R) for 400 eV was reached in the final tests on the Compact Torus vertical access. Experiment (CTX) (Los Alamos National Laboratory). A strong relationship was recently observed in the The issues facing the spheromak are reversed-field pinch (RFP) experiment of the Madison Sym- 1. lack of data on core energy confinement and high metric Torus (University of Wisconsin) between magnetic beta limit during sustainment fluctuations and plasma heat and particle fluxes Qr and Tr near the plasma edge. 2. unresolved mechanisms and scaling for confinement Magnetic fluctuations and radial transport losses are ob- and beta limits served to decrease as the profile of X = j)\(r)/B(r) is flat- 3. feedback or other stabilization required for global in- tened in spheromaks and in RFPs. As the helicity lifetime stabilities for discharges lasting beyond the flux con- increases with electron temperature, this observation leads server timescales. to the prediction that energy losses will decrease as the spher- omak core becomes hotter and, in some simple modeling, The normal-conductor TFC center leg introduces an that the confinement would be adequate when extrapolated added complexity to the low-^1 tokamak relative to the to a reactor. A model of spheromak core confinement based spheromak. The external toroidal field, however, produces on this mechanism has been developed and indicates favor- the high-<7 tokamak magnetic field geometry with the at- able extrapolation to a reactor. tending promises of tokamak-like confinement, high beta The timescale for plasma buildup in this experiment can stability, current drive, and impurity control. The sphero- be shorter than the soak-through time of the flux conserver, mak progress in these areas has been relatively modest in so that no externally applied vertical field is required to comparison. maintain plasma edge position. No auxiliary heating power Paux would be required for this purpose. PLASMA CONFINEMENT Wall conditioning will be important as in tokamaks; progress in the CTX experiments will be obtained only when Tokamak discharge cleaning and wall conditioning are done. The Small Tight Aspect Ratio Tokamak (START) (Cul- ham Laboratory, United Kingdom) experiment and initial MHD EQUILIBRIUM AND GLOBAL STABILITY kinetic instability analysis using modeled equilibria and pro- files have provided enticing results on plasma confinement Tokamak in the low-y4 tokamak. The following properties characterize the low-,4 Confinement results from a mega-ampere, low-/! toka- tokamak: mak are needed to develop a more broadly based empirical confinement scaling. Only very limited data are available 1. Given a simple vertical field (decay index >0), the from START for comparison with scalings; additional in- limiter plasma elongation (K = b/a) increases to >2 and tri- formation is hoped for from the Current Drive Experiment- angularity to >0.6 naturally without using plasma-shaping Upgrade (CDX-U) [Princeton Plasma Physics Laboratory coils as A approaches the ultralow A range of 1.1 to 1.2. (PPPL)], the Helicity Injected Tokamak (HIT) (University 2. This plasma is vertically stable without external of Washington), and START with neutral beam injection control. (NBI) heating. Auxiliary heating with Paux » Poh will be needed to develop scalings with confidence. 3. Strong elongation, triangularity, and toroidicity lead The experiment will investigate the suppression of to strong plasma shaping, producing (Ipq^/aBt0), which trapped particle modes and related turbulence and transport. approaches 200 MA/m • T at ultralow A. START has pro- The initial analysis indicated the following: duced (Ipq^/aBt0) « 20 MA/m-T for A « 1.4 and K « 1.6. 1. Increased good curvature at low A reduces the drive 4. The START plasma has not experienced major dis- for this instability. ruptions for A < 1.8, whereas CDX-U plasmas did recently. The causes for the difference need to be investigated. 2. Strong dependence on A is expected in the range of 5. The scrape-off layer (SOL) flux tube connected to the 1.2 to 1.6 and on v*e at low collisionality. plasma inboard diminishes as A decreases, leaving the SOL 3. Fluctuation diagnostics in such plasmas will be nec- mostly diverted without using divertor coils. essary to evaluate these effects directly. 6. The PPPL calculations and Tokyo Spheromak-3 The experiment will permit study of H-mode physics at (TS-3) (University of Tokyo, Japan) experimental results low A. The present ohmic plasma data provide no informa- suggest that plasma tilt can be prevented for the low-,4 tion on the L or H mode. Effects of divertor configuration, tokamak when Ip/Itf ^ 2 and/or q^ > 1. The dependence wall conditioning, and threshold powers on L- and H-mode of this limit on plasma shaping and profiles needs to be transitions need to be measured. clarified.
FUSION TECHNOLOGY VOL. 29 MAR. 1996 211 Spheromak PLASMA HEATING AND CURRENT DRIVE Major spheromak plasma stability properties are (a) Tokamak tilt, shift, and \ow-n external kinks that can be stabilized by a close-fitting flux-conserver over a resistive wall time and In low-,4 tokamaks, NBI, ion cyclotron resonance fre- (b) feedback stabilization of these modes on longer time- quency wave, ion Bernstein wave, Alfven wave, electron cy- scales. Rotation may provide stability against some modes. clotron wave, and lower-hybrid wave are being considered for heating, and NBI and helicity injection are being con- sidered for current drive. MHD BETA LIMIT STUDIES 1. Many radio-frequency schemes are expected to Tokamak present special challenges for high beta studies at low A be- cause of the low field (<1 T) and high density expected. (Ini- Tremendous progress was reported at this workshop in tial calculations of electron Landau damping, performed quantitative theoretical analysis of tokamak beta limits at after the workshop, look very favorable for fast-wave elec- low A. tron heating at the frequency range of the high ion cyclo- 1. For current profiles with q0 ~ 1 and A down to 1.3, tron harmonics.) the Troyon scaling remains roughly unchanged (@Nt rises slightly in some analyses and falls in others), giving up 2. For the next-step mega-ampere experiments, NBI is to -20% without a conducting wall. A standard version of a straightforward method for heating and current drive. Steady-state cases have been calculated for HVPNS (Ip ~ 6 Troyon scaling is used here: (ft) = 2/z
4. Spheromak plasmas are calculated to be unstable to Tokamak resistive interchange modes at all betas. However, the effects Divertor effects and questions relating to the low-^4 to- of high magnetic Reynolds number S and high drift fre- kamak were discussed during the workshop. quency (co*) may contribute to stabilizing these modes or minimizing their consequences at high temperatures and re- 1. High heating power P and low R at the divertor strike actor-relevant parameters. point in future low-^4 tokamaks lead to high P/R. Here R
212 FUSION TECHNOLOGY VOL. 29 MAR. 1996 212 could be significantly smaller than R0. Near-term devices often end in hard disruptions, whereas the START discharges have P/R in the same range as present large tokamaks. exhibit reconnection events without hard disruptions. 2. Divertor considerations for low A are similar to those 2. In TS-3, the global MHD stability of the ultralow-,4 for high A. They include space for divertor plates, baffles (>1.05) configuration formed from a spheromak with to confine neutrals in the divertor chamber, strong pump- q(a) ~ 1 is another area that has been explored. ing in the divertor chamber, and impurity entrainment by 3. Strong stabilization of the microturbulence behavior plasma flow in the SOL. has been calculated for plasmas with A < 1.5. Transport 3. The same divertor solutions will probably apply, but measurements of microturbulence and confinement for the divertor must provide strong pumping to hold the A < 1.5 can be pursued in the collisionless regimes in CDX-U plasma density below the Greenwald limit to make nonin- and START. ductive sustainment of the plasma current attractive for 90% 4. With auxiliary-heated CDX-U (ECH, 8 GHz) and pressure-driven currents. START (NBI) plasmas, MHD stability behavior at increased 4. Large SOL flux expansion for low-,4 plasmas, seen beta can be explored. so far in START, may be useful for controlling heat fluxes; this is a worthwhile test objective for near-term devices. 5. To move toward the ultralow-,4 regime, one must ini- tiate and form tokamak plasmas without induction. Several 5. The double-null configuration may be favored for promising options such as helicity injection, ECH pressure- minimizing power exhaust on the inboard side; this needs driven currents, and compression can be explored in CDX-U, to be proven. START, HIT, TS-3, SPHEX, and MRX. 6. A ballooning limit to the SOL thickness (pressure 6. It is important to develop efficient current mainte- gradient) is interesting. This needs to be investigated more nance options. Similar to the startup problem, helicity in- widely. jection and pressure-drive currents are promising options 7. Can the electrodes for coaxial helicity injection (ei- that can be explored on CDX-U, HIT, and SPHEX. ther for tokamak or for spheromak) do double duty as 7. The spheromak low-A tokamak continuum can be divertor plates? In this case, can a "detached" plasma be ob- investigated in the ultralow-,4 and low-g regimes in the TS-3, tained? These potentially attractive techniques need to be FACT, SPHEX, and MRX. In TS-3, the addition of a mod- investigated. est toroidal field has resulted in stabilization of the tilt-like Spheromak MHD mode. Various spheromak issues can also be pur- sued, if needed, with possible new experiments on HIT and Spheromak configurations appear compatible with a di- CDX-U. vertor, subject to the foregoing discussion and the absence of a significant toroidal field at the plasma edge. Spheromak
PRESENT DEVICE EXPERIMENTAL PLANS Device experimental plans for the spheromak were not presented at the workshop. Tokamak At present, there are two ohmic low-,4 tokamaks in ACTION ITEMS the world: START and CDX-U. HIT is a helicity-injection- driven low-,4 tokamak. There are three spheromak or near- The workshop participants agreed to take actions to spheromak low-,4 devices: TS-3, SPHEX, and the Flux bring forth nationally based low-,4 toroidal experiments at Amplified Compact Toroid (FACT) (Himeji University, the mega-ampere level in the United States. These critical Japan). action items for the low-,4 tokamak were determined as Regarding near-term improvement plans of present de- follows: vices, START plans to add an NBI capability to test auxil- iary heating, confinement, and beta limit in the 200- to 1. agree on desirable mission elements for a next-step low-,4 tokamak 300-kA range. The CDX-U will increase electron cyclotron heating (ECH) power to 100 kW at 8 GHz in addition to up- 2. determine if there is a role for more than one device grading the ohmic solenoid power supply capability. The (in different scales, presently existing or new) HIT plans to add ohmic current drive capability. An ohmic low-,4 tokamak is being built and should 3. determine device characteristics that follow from become operational in late 1994 at the University of Tokyo. each mission element for the large-scale and perhaps The Magnetic Reconnection Experiment (MRX) facility is smaller scale devices being readied at PPPL, which can also investigate the near- 4. determine the sites that can provide these character- spheromak low-,4 regime. The Instituto de Pesquisas Espaciais istics and the required costs (Brazil) is building a small spherical tokamak experiment ETE to be ready in 3 yr. 5. ask DOE to select site(s) so that a national design ef- There are several important areas in which the present fort can be carried out. and near-term devices can contribute toward future experi- Action Item 1: Desirable Mission Elements for Next-Step ments at the 1-MA level. Low-A Tokamak Experiment 1. One area is in testing of the q limit. As the q(a) is Action item 1 was accomplished at this workshop and is pre- pushed down toward 4 for A ~ 1.4, the CDX-U discharges sented in Table I in accumulative progression.
FUSION TECHNOLOGY VOL. 29 MAR. 1996 213 Action Item 2: Mission Elements for Smaller Low-A Toka- Action Item 3: Provisional Mission-Driven Device Charac- mak Devices teristics for a Next-Step Low-A Tokamak For action item 2, it was agreed that there are very impor- Action item 3 was completed at this workshop, and the tant physics activities for ongoing \ow-A tokamak experi- device characteristics as driven by the mission are provided ments (HIT, CDX-U) and other possible device upgrades. in Table I. However, a consensus was not obtained at the workshop Action Item 4: Determination of the Sites that Can Provide regarding when these activities would be completed and These Characteristics and the Required Costs whether their continuation or other similarly scaled devices would be needed during operation of the next-step low-,4 to- It is proposed that each institution interested in hosting kamak experiment. Some of the possible mission elements the next-step \ow-A tokamak be asked to evaluate the cred- for ongoing or future smaller scale low-^4 tokamak experi- its provided by their existing facility to support these device ments were agreed to as follows: characteristics. In addition, each group should provide rough costs associated with site upgrades needed to support the full 1. testing of q limits set of desired device characteristics. Each institution has the 2. investigating transport physics mechanisms (turbu- freedom to choose a subset of the device characteristics that lences) at low A effectively uses the existing facility. 3. investigating innovative divertor geometries Action Item 5: Request for DOE Selection of Site(s) for Ex- ecution of National Design Effort 4. studying disruption characteristics and beta limits under moderate heating The host-site summaries should be brought together for the purpose of calibration by the working group before the 5. developing innovative startup schemes extrapolata- end of September, 1994. The results will be forwarded to ble to a next-step low-v4 tokamak DOE to facilitate the Department's decision-making process. 6. developing innovative current drive concepts ex- No action items were defined for the spheromak re- trapolatable to a next-step \ow-A tokamak search at the workshop. The following material was included after the workshop as an addendum. 7. studying spheromak configurations and the sphero- A two-step approach is recommended for developing mak \ow-A tokamak continuum. spheromaks, wherein the critical issues are addressed pro- These missions roughly correspond to the first three gressively. In the first step, core energy confinement would phases of Table I with an emphasis on exploratory studies. be investigated in a low-cost short-pulse experiment. The
TABLE I Desirable Progressive Mission Elements and Required Device Characteristics for Midsize Low-v4 Tokamaks
Phase Proposed Mission Elements Required Device Characteristics
Ohmic Determine the range in q (or Ip/aBtQ) for well-confined, Ip ~ 1 MA, Tfiattop ~ 1 S stable tokamak plasmas for A = 1.3 to 1.8 Noninductive Develop techniques for plasma startup to full operating Helicity injection or other noninductive startup current, in the absence of a central transformer startup system (additional systems can be considered as upgrades) Noninductive current drive system (considered as upgrade)
PWX > POH Determine energy confinement scaling at low A Paux ~ 4 MW, rflattop -Is, £>° - D+ NBI (= 1.15 to 1.8) Inertial-power-handling hardware Investigate beta and disruption limits in limited parameter range Investigate scrape-off characteristics of low-,4 plasmas
RQUX ^ POH Determine beta-limit scaling at low A; test effect of Paux ~ 10 MW, Tflattop ~ 1 s (upgrade nearby conducting wall requirement depends on limits and Investigate disruption behavior of high-(/3r)f \ow-A te achieved) plasmas
Long pulse Investigate high pressure-driven current fraction, Paux for Tfiattop ~ 10 s, inertial plasma- high-<0,> operating regimes, and approach to steady facing components (PFCs) 2 state at low A Paux for Tfiattop ~ 10 s, actively cooled Test external current drive schemes PFCs Steady-state operation in advanced regimes
214 FUSION TECHNOLOGY VOL. 29 MAR. 1996 214 TABLE II Mission Elements and Required Device Characteristics for Near-Term Spheromak Experiments
Phase Proposed Mission Element Required Device Characteristics
Core energy Demonstrate core energy confinement by quiescent Plasma equilibrium established by spheromak confinement ohmic electron heating to Te ~ 0.4 keV in a injection into a conducting flux conserver sustained, short-pulsed spheromak Plasma formation and sustainment by helicity Evaluate core energy confinement (effective thermal and power injection from a coaxial gun conductivity) and correlate with level of magnetic Helicity injector matched to the plasma to turbulence minimize magnetic turbulence Determine beta limits and their relationship with Clean vacuum and wall conditions transport Tpulse > 2 ms, lp ~ 1 MA, Paux ~ 0 Optimize operation of a sustained spheromak (e.g., Magnetic fluctuation diagnostics particle control and helicity match to the plasma) Obtain initial data on long-pulse operation Long-pulse Transfer the equilibrium maintenance from the flux Vertical field coils to support the plasma experiment conserver to external poloidal field coils Feedback or other stabilization (e.g., via plasma Stabilize the tilt and shift modes on timescales long rotation) for the tilt and shift modes compared with the soak-through time of the flux Quiescent ohmic heating by the helicity-driven conserver current Achieve Te < 1 keV Divertor to handle particle and power losses Evaluate ohmic heating extrapolation to a reactor Extensive diagnostics set: Is auxiliary heating required? Density (Thomson scattering, interferometer) Study the physics of multi-kilo-electron-volt Temperature (Thomson scattering, impurity spheromak plasmas Doppler broadening) Magnetic (pickup loops, motional Stark effect, O-to-X mode conversion)
Tpu[se > 100 ms, Ip > 1 MA
plasma pulse in this experiment would be long compared E. B. Hooper with the energy confinement time (at a few hundred electron- Lawrence Livermore National Laboratory volts) but short compared with the magnetic field soak- P.O. Box 808 through time of the confining and stabilizing flux conserver. Livermore, California 94550 Good confinement results would encourage the second step, wherein a long-pulse experiment would address the confine- ment physics of high-temperature (>l-keV) spheromaks and 5. M Kaye stability of equilibria maintained by external poloidal coils. M. Ono This could be conducted in a mega-ampere low-,4 tokamak Princeton Plasma Physics Laboratory experiment, saving cost and permitting the common use of Princeton University diagnostics. However, the design of such a facility to address P.O. Box 451 both configurations requires resolution. Princeton, New Jersey 08544 The mission elements and device characteristics for the spheromak experiments are provided in Table II. R. E. Siemon Y. -K. M. Peng Los Alamos National Laboratory Oak Ridge National Laboratory P.O. Box 1663 P.O. Box 2009 Los Alamos, New Mexico 87545 Oak Ridge, Tennessee 37831-8071 R. Stambaugh R. J. Goldston Princeton Plasma Physics Laboratory General Atomics, Fusion Group Princeton University P.O. Box 85608 P.O. Box 451 San Diego, California 92186-9784 Princeton, New Jersey 08544 A. J. Wootton R. J. Colchin University of Texas at Austin Oak Ridge National Laboratory P.O. Box 2009 Austin, Texas 78712 Oak Ridge, Tennessee 37831-8072 June 2, 1995
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