**Response to ST questions from Alternative Concepts Review Panel**

**S. Kaye, M. Ono, R. Goldston (PPPL, Princeton Univ.),**

**M. Peng (ORNL), 27 March 1996 (corrections included)**

**First Category:**

**What is the status of (Spherical Tokamak) research, achievements, and
the justification for proceeding to further:**

**1) What is the present experimental achievements?**

CDX-U: 1 kG, 100 kA, R/a ≈ 1.5, κ ≈ 1.6.

> \* Non-inductive ST start-up (1990-1993):
>
> \- Helicity injection ST start-up demonstrated.
>
> \- ECH produced fully pressure driven ST configuration demonstrated.
>
> This CDX-U result together with HIT (CHI) and DIII-D (ECH) results
> gives us confidence that we can form a tokamak discharge
> non-inductively in future ST.
>
> \* ST inductive start-up with toroidally continuous vacuum vessel wall
> (1994-1995):
>
> \- ECH preionization reduced the required start-up voltage to only a
> few volt.
>
> \- Efficient ohmic start-up with Ejima coefficient ≈ 0.3 obtained for
> up to 100 kA.
>
> This CDX-U result together with the START result give us a high degree
> of confidence in the effectiveness of ECH startup assist.
>
> \* Role of resistive MHD on ST performance (1995-):
>
> \- Lower operational limit of q(a) ≈ 3.5 established up to 100 kA.
>
> \- Increase in the n=1/m=3 activity causes the observed q(a) limit.
>
> \- Mode coupling between n=1/m=1 and n=1/m=3 resistive modes leading
> to an IRE (internal reconnection event).
>
> \- Ohmic ST plasmas are indeed resilient - no hard disruption (as
> found in START).
>
> With improved understanding of resistive MHD phenomena such as IRE, we
> will be in a better position to control them for the ST performance
> optimization.
>
> \* Attractive ST Regime (1995-)
>
> \- MHD quiescent regime (QS) with 2-3 times the central confinement
> compared to the usual regime with resistive mode activity has been
> obtained by a discharge programming. This suggests the need for
> current profile control for future ST experiments.
>
> \- The QS regime with low MHD activity and low collisionality (ν\* \<
> 0.1 for the most of the volume) is well suited for the
> micro-turbulence/transport study.
>
> \- The QS regime with high Te is an ideal target for high harmonic
> fast wave heating.

START: 5 kG, ≤ 250 kA, R/a ≥ 1.25, κ ≤ 3−4.

> ∗ High Elongation
>
> − Stable plasmas with κ up to 4 have been obtained for short time
> scales (a few ms) immediately following current ramp up using inboard
> solenoid alone. The electron temperature profiles were measured to
> have off-axis peak values in the range of 100−200 eV (with about 100
> kA in plasma current) during this phase.
>
> ∗ Stable Operation at High Densities
>
> − Stable operation at high plasma densities (≈1014 cm−3 line average)
> has been obtained consistently for currents in the range of 100−200 kA
> for R/a ≈ 1.4 in START. Greenwald density nGW for such plasmas is in
> the range of 0.6−1.2⋅1014 cm−3. Success in high density operation is
> apparently related to the absence of major disruptions so far in START
> (see \"Disruptivity\" below).
>
> ∗ Confinement Data Base
>
> − High peak electron temperatures (500−1000 eV) have been obtained in
> START with 100-200 kA in plasma current and 5 x 1019 m−3 in density.
> Local outboard electron temperature gradients can approach 100 eV/cm
> in such plasmas.
>
> **−** While the limited range of confinement data seem to fit trends
> of the Rebut-Lallia scaling and exceed those for several other scaling
> expressions, the data points to a clear need for studies using
> auxiliary heating and for plasma currents at the MA level
>
> ∗ Divertor Power Exhaust and Configuration
>
> − Double-null divertor plasmas can be produced since 1995 in START, in
> addition to the inboard-limited plasmas with natural elongation and
> diverted outboard SOL. Significant measurements using probes have been
> carried out for the edge and SOL plasmas.
>
> − For the natural divertor configuration, large fluctuations in
> electron temperature (order 50% relative to background temperatures,
> which are in the range of 30−50 eV) have been measured, while density
> fluctuations remained modest (order 10%).
>
> − For the natural divertor configuration, only order 10% level of
> total heating power (ohmic so far) can be measured to reach the
> inboard limiter. The outboard heat flux estimates based on probe
> information so far can only account for a small fraction of the total
> heating power.
>
> − By comparing the measured outboard SOL thickness and the field-line
> connection lengths for these two configurations, it is found that the
> SOL thickness scales more consistently with the model assuming
> marginal MHD pressure driven instability than assuming typical
> transport coefficients for diffusion and parallel convection.
>
> ∗ Plasma Current Limitations
>
> − START plasmas have operated successfully for edge q values as low as
> 4, although edge q values around 5−6 (b/a ≈ 1.5) are more common for
> flat top operations.
>
> − For the highly elongated plasmas, edge q values as high as 30-40
> have been inferred from equilibrium modeling.
>
> ∗ Disruptivity
>
> − The START ohmic plasmas is yet to suffer a major disruption for as
> long as R/a remains below 1.8 for over 29,000 discharges, for the
> operational range of edge q values mentioned above.
>
> − Relatively large scale Internal Reconnection Events (IRE\'s) have
> been observed, from which the plasma shows a strong ability to
> recover.

HIT: 5 kG, ≈ 200 kA, R/a ≈ 1.5, κ ≈ 2.

> ∗ Noninductive Start Up and Current Drive
>
> **−** HIT has demonstrated successful startup and current sustainment
> (for up to 10 ms) by Coaxial Helicity Injection (HIT) of a low aspect
> ratio tokamak without a transformer.
>
> **−** It has achieved electron temperatures of 100 eV, burning through
> the oxygen impurity, with plasma currents around 200 kA, The plasma is
> characterized by an n = 1 perturbation, which is believed to be
> responsible for the inward flux diffusion associated with this current
> drive.

**2) What is present theoretical understanding?**

> A great deal of theoretical work has pointed to the benefits of the ST
> configuration in producing scientifically interesting, high
> performance plasmas. Analysis has been carried out showing the
> expected behavior of microinstabilities as aspect ratio is reduced
> from 2.5 to 1.25 at fixed q and fixed RBT.[^1] This investigation,
> based on a comprehensive microinstability code which takes into
> account high-n effects, all resonances, and finite-beta, finite Larmor
> radius, and banana orbit effects, showed a dramatic decrease in the
> growth rate of both electrostatic (trapped electron-ηi) and
> electromagnetic (kinetic ballooning) modes for low ν∗ plasmas, as the
> aspect ratio decreases below 1.3 to 1.5. The reason for this
> stabilization is a reduction in the orbit-averaged bad curvature (the
> instability drive) in low aspect ratio equilibria. Another aspect of
> these equilibria, especially at high-beta, is a reduction in the
> trapped particle fraction in the core because of omnigeneity and the
> presence of a magnetic well.
>
> MHD stability studies have shown the attractiveness of the ST
> configuration in producing high-beta/high bootstrap fraction
> configurations. Results of parameter scans in the low- and high-n
> stability studies have shown the potential for increasing βt by a
> factor of three to four as aspect ratio decreases from 2.5 to 1.2,
> achieving values of close to βt=33%, βn=5.5 for a typical Ip=1 MA,
> BT=0.3 T, R/a=0.8/0.64 m, κ=2.0, qa=8, q0=1.1, p0/\<p\>=1.7 plasma.
> The results indicate that stabilization by a conducting shell at
> rwall/a\~1.25 is needed in order to achieve these βt values.
>
> Stability surveys of a range of parameters show that high-beta can be
> attained in several regimes: a \"low-q\" regime (as given above) in
> which the amount of bootstrap current is small (\~20 to 40%), and thus
> substantial external current drive would be required for fully
> non-inductive operation. (The required bootstrap currents, as
> calculated in the MHD equilibrium and stability studies, are
> consistent with the values computed using the Shaing-Houlberg model,
> which is valid in the low aspect ratio regime.[^2] As expected, a
> result of the calculation based on this theory is that the bootstrap
> fractions obtained scale linearly with plasma beta at fixed qa). A
> second regime exists, however, with high bootstrap fractions, in
> excess of 80%. In this regime, qa is higher (\~16), and q0\~2.5. It is
> necessary to optimize both the pressure and the current profiles to
> obtain this high-beta, high and well-aligned bootstrap equilibrium. In
> the highest-beta equilibrium found so far (βt=44%, βn=8.5), 80% of the
> current is carried by pressure gradient driven currents. The remaining
> 20% would need to be provided by current drive near the edge of the
> plasma. Work is continuing to determine if it is possible to increase
> the fraction for pressure gradient driven currents to above 90%
> without reducing the stability beta limit.
>
> Non-inductive current drive is an essential element of an ST research
> program because of the high current requirement and the preclusion of
> an OH solenoid to make the center stack as skinny as possible. The
> bootstrap current was described above as a potential means of driving
> the majority of the current in optimized configurations. The bootstrap
> fraction, as discussed, scales with βt at fixed qa, indicating the
> importance of achieving the high-beta equilibrium (≥30%). Collisional
> effects can reduce the bootstrap current significantly especially near
> the edge for the ultra low aspect ratio of 1.25. Theoretical and
> experimental studies of high temperature plasma are needed for the
> ultra low aspect ratios.
>
> For cases in which the ∇p driven current accounts for only a part of
> the total current, other means to drive current non-inductively must
> be found. Such current drive may also be needed for j(r) control.
> Because of the high dielectric constant of ST plasmas, typically
> ωpe2/ωce2\>\>1, the conventional RF current drive techniques, such as
> Lower Hybrid and Electron Cyclotron, are impractical. High Harmonic
> Fast Waves (HHFW) are a promising alternative for heating and current
> drive at low aspect ratio.[^3] HHFW shows strong, single pass
> absorption for high-βt operation, consistent with the high density,
> low TF regime of STs. The damping decrement scales as ne/BT3. Fast
> Waves have good accessibility, and will damp on the electrons up to βi
> values of \~50%, at which point the waves mode convert to the Ion
> Bernstein branch. The advantage of the HHFW is that the resulting
> electron heating can enhance bootstrap fractions, and the power
> deposition can be localized for current profile control. Calculations
> based on the FISIC[^4] code and the Ehst-Karney[^5] parameterization
> for the above nominal discharge parameters and for nφ=2, ne(0)=5 x
> 1019 m-3, Te(0)= 5 keV and Zeff=2, indicate that close to 1.5 MA of
> current can be driven with 6 MW of HHFW. At higher nφ (\~4), the
> driven current can be localized off-axis, with an associated decrease
> in the current drive efficiency such that the total driven current is
> reduced, but still reaches approximately 0.5 MA.
>
> In order to assess credibly these non-inductive current drive
> scenarios, and the MHD stability of the equilibria, it is important to
> allow enough time for the current profiles to relax fully. In the high
> Te plasmas, Te(0)=5 keV, fully time-evolving discharge simulations
> using the Tokamak Simulation Code (TSC) indicate that it requires
> approximately 3 to 5 sec. to allow this relaxation to take place.
>
> The final issue to be discussed in this section is associated with the
> scrape-off layer (SOL) physics. The motivation is to understand the
> divertor heat fluxes that could be expected in the high power density
> ST regime (e.g., P/R\~7.2 MW/m for 6 MW loss power for the plasma
> condition mentioned above), and how to reduce them. As will be
> mentioned in the next section, experimental data in START indicate
> outboard SOL thicknesses that are greater than those expected for Bohm
> transport across the SOL layer.[^6] The SOL thicknesses, however, are
> consistent with those expected for maintaining marginally stable,
> pressure gradient driven MHD modes with line tying at the divertor
> plates.[^7] For the nominal ST parameters indicated earlier, outboard
> SOL thicknesses of order 6 to 10 cm would be expected for the MHD
> model, as compared to 2 to 3 cm for the Bohm model, and the thicker
> SOL would lead to reduced loads on the divertor plates.

**3) Does theory, modeling, simulations, and empirical scalings fit**

**the experimental observations?**

> Though initial results from START, CDX-U, HIT, TS-3, etc. are often
> better than expectations based on the standard aspect ratio tokamak
> data base, the ST plasma has thus far behaved qualitatively like a
> tokamak. Much of the tokamak operational experience and existing
> tokamak theory and analytic models are applicable to ST following
> appropriate adjustments. Much needs to be done in this area. Initial
> comparisons have led to encouraging confirmation as well surprising
> indications for the ST that are yet to be understood. ST physics
> research is therefore likely to provide new opportunities for progress
> in our understanding of toroidal magnetized plasmas. Examples include:
>
> ∗ High Elongation for Hollow Current Profiles
>
> − Free-boundary MHD equilibrium modeling for the START plasmas
> verified that the observed elongations (that can vary from as high as
> 4 at the beginning to 1.4 near the end of the discharge) are
> consistent with the effects of hollow to peaked current profiles.
>
> − The observed vertical stability of these naturally elongated plasmas
> are consistent with the well understood effects of vertical field
> decay index on vertical stability.
>
> ∗ Confinement Interpretations
>
> − The ST confinement data are so far limited to ohmic plasmas. The
> results have nevertheless shown the strong likelihood that most of the
> common scaling expressions will not provide reliable predictive
> capabilities for the ST plasmas.
>
> ∗ Noninductive Startup and Current Drive
>
> − Modeling for startup by helicity injection have benefited from the
> HIT experimental results. The externally applied toroidal field
> increases the circuit impedance of the electrodes when compared to the
> previous HI tests in spheromaks (e.g., CT), leading to an improvement
> in the startup efficiency. This improves our ability to design an
> effective CHI systems for the MA level experiment.
>
> − The ECH-produced tokamak currents in CDX-U and DIII-D have been
> interpreted as primarily pressure-driven bootstrap currents in very
> high q plasmas. This is an encouraging result for application to
> larger ST tokamaks.
>
> ∗ Divertor Power Exhaust and Configuration
>
> − There is at present no reliable understanding of the partition of
> the tokamak plasma heat flux between the inboard and the outboard
> plasma edges. The absence of high heat fluxes to the inboard limiter
> in START needs to be investigated though this result is not
> inconsistent with the observations in DIII-D.
>
> − By comparing the measured outboard SOL thickness and the field-line
> connection lengths for the double-null divertor and the natural
> divertor configurations, it is found that the SOL thickness scales
> more consistently with the model assuming marginal MHD pressure driven
> instability than assuming transport by diffusion and convection. This
> is a high leverage issue owing to the thick SOL predicted by the
> former model for ST.
>
> ∗ Disruptivity
>
> − The mechanisms by which the START plasmas have so far avoided major
> disruptions for edge q values as low as 4 are not understood. Tests
> for high beta collisionless plasmas will shed more light on this
> important subject.
>
> − The IRE\'s observed so far in small ST plasmas (START, CDX-U) show
> significant differences from those observed in standard tokamak
> plasmas. The strong ability for the ST plasmas to recover from IRE\'s
> is not understood.

**4) What constitutes \"proof-of-principle\'\' for this concept?**

> The \"proof-of-principle\" experiment for this concept should address
> the physics issues which are critically needed for attractive ST-based
> VNS, Pilot Plant, and Power Plant applications. These critical issues
> include:
>
> ∗ Noninductive startup, current sustainment and j(r) control,
>
> ∗ Confinement physics especially with additional heating,
>
> ∗ β limit resulting from MHD instability physics, and
>
> ∗ Scrape-off-layer physics.
>
> The regimes of interest to these applications are characterized by the
> advanced physics features of:
>
> ∗ High β \~ 30 − 45%,
>
> ∗ High pressure gradient driven current fraction \~ 50 − 90%,
>
> ∗ Fully relaxed, noninductively sustained current profile,
>
> ∗ Extreme low R/a \~ 1.25 with high natural κ \~ 2, and
>
> ∗ Low collisionality for extrapolating results to reactor
> applications.

**5) What is the potential impact of research on this concept on**

**a) increasing our knowledge of basic plasma physics?**

> ST research will have a strong impact on both basic and fusion related
> plasma physics in that the experimental and theoretical investigations
> will be carried out in a previously unexplored parameter regimes.
> Important results will be generated in truly high-beta/high bootstrap
> collisionless plasmas that exhibit
>
> ∗ Increased orbit-averaged good curvature for suppression of
> electrostatic and electromagnetic microturbulence,
>
> ∗ Near-omnigenous regions of reduced trapped particle fraction in the
> outboard region of the core,
>
> ∗ High trapped particle fractions and mirror ratios near the plasma
> edge, and
>
> ∗ Strong single pass electron absorption of high harmonic fast waves.
>
> The naturally diverted outboard SOL is characterized by
>
> ∗ Strongly unstable magnetic curvature,
>
> ∗ Long connection length of the field lines, and
>
> ∗ Large mirror ratios.
>
> These enhance the MHD instabilities, limit the pressure gradients
> there, and possibly leading to increased SOL thickness and reduced
> heat fluxes on the divertor plates. Investigation in these regimes
> push beyond the limits of present studies at low as well as higher R/a
> configurations.

**b) increasing our knowledge of fusion plasma physics (especially
generic fusion plasma physics as well as physics of other confinement
concepts)**

> The investigations mentioned in (a) will also increase the knowledge
> base of fusion plasma physics. Important studies that will enhance the
> fusion physics base include investigations to:
>
> ∗ determine the q-limits for stable, disruption-free operation,
>
> ∗ develop techniques for controlling current profile in both the core
> and outer plasma regions, and for starting up and maintaining the
> plasma current non-inductively,
>
> ∗ reduce the divertor heat loading in a variety of configurations,
>
> ∗ assess the confinement properties of high power, low ν∗ plasmas,
> including H-mode access and L- and H-mode confinement trends, and
>
> ∗ determine the plasma operation scenarios needed for fusion-grade
> plasmas with these attractive features.
>
> Results obtained at low R/a will expand the present higher R/a
> database, and will thus lead to a more comprehensive understanding of
> the physics mechanisms underlying high performance tokamak plasmas.
>
> The ST plasmas of the ultra and extreme low aspect ratios (R/a as low
> as 1.1 or less) are expected to be characterized by a diminishing
> externally applied toroidal field (Bt0) compared to the fields
> generated by the currents in the plasma. Such ST plasmas approach in
> magnetic configuration the low aspect ratio spheromak (with Bt0 = 0
> and βp \< 1), the field reversed configuration (FRC, with Bt0 = 0 and
> βp = 1), and a low aspect ratio rendition of the reversed field pinch
> (RFP, with Bt0 \< 0 relative to the plasma generated toroidal field
> and βp \< 1). Comparison of the plasma behavior among these magnetic
> configurations should shed new light on the differing underlying
> physics mechanisms.

**c) helping develop fusion as an energy source (help develop the**

**data base for fusion development steps such as burning plasmas,**

**volumetric neutron source, etc.)?**

> Verification of the physics principles for the ST configuration (see
> Item 4 above) produces definitive data base for the advanced physics
> regime collisionless plasmas of high-temperature. Such data should be
> adequate for determining the design and performance parameters of an
> attractive D-T-fueled ST fusion core for use in the next step fusion
> development.
>
> In the case of moderate success in verifying the advanced physics
> regime (e.g., β \~ 20−30%, bootstrap fraction \~ 40−50%, and typical
> tokamak-like confinement, etc.), it is estimated that a small fusion
> core with R ≤ 0.9 m and Ip ≤ 10 MA would permit a fusion performance
> at the level of Q ≈ 1. Such ST plasmas, driven with neutral beam
> heating and current drive, would produce a fusion power of 30−40 MW
> and neutron wall loading of ∼1 MW/m2. A remote maintainable ST device
> with such fusion performances for steady state operation would be a
> cost-effective vehicle (e.g., the VNS) for testing nuclear
> technologies needed for future power production.
>
> In the case of full success in verifying the advanced physics regime,
> an ST plasma with R ≤ 1 m and Ip ≤ 15 MA is estimated to permit a
> fusion performance at the level greater than Q = 10. Upon reaching
> fusion burn conditions, the heating and current drive power needed for
> sustainment drops to about 10 MW to produce a fusion power of over 100
> MW and neutron wall loading of ∼3 MW/m2. A remote maintainable ST
> device with such fusion performances for long pulse operation (∼102 s)
> would be a cost-effective vehicle for testing fusion burn near
> ignition. Successful outcome of this test would lead to improved
> economy of the VNS because of the reduced drive power and increased
> wall loading, and highly attractive Pilot Plants producing fusion
> powers in the range of a few hundred MW.
>
> Proof of principle experiment for the ST concept has a very attractive
> potential impact for developing fusion energy among all the
> confinement concepts examined so far. This potential is based on the
> projected possibility for building highly cost-effective fusion cores
> for testing the physics, engineering and technology of fusion energy
> production.

**d) developing this concept as a candidate for a fusion power plant?**

> The physics, engineering and technology data for fusion power
> production together with the utility experience of Pilot Plant
> operation would secure the basis for an attractive future DEMO power
> plant based on the ST concept, producing a net electricity of a couple
> hundred MW at a high duty cycle. Successful operation of such DEMO
> power plants will produce the data base needed to optimize the
> economic viability of fusion power plants.
>
> It should be noted that a number of engineering, technology, and
> material issues unique to the ST fusion core will have to be addressed
> successfully before the potential benefits can be accrued. These
> include
>
> ∗ nuclear blankets that handle high neutron wall loading (\~6−10
> MW/m2) for long lifetimes (\~20-30 MW-yr/m2); and
>
> ∗ the single-turn center leg of the TF coil that is designed to carry
> only modest loads, and use highly conducting copper alloys resilient
> to intense neutron damage and modest in activation due to
> transmutation.

**Second Category:**

**What is the appropriate level of research for this concept:**

**1) What are the major experimental and theoretical issues that should
be addressed?**

> The major issues to be addressed for the ST concept fall into four
> broad categories: confinement and transport; MHD stability;
> non-inductive startup, current sustainment and profile control; and
> scrape-off layer (SOL) and divertor physics. It is important to carry
> out these studies in scientifically interesting regimes of high-beta
> and low collisionality.
>
> Under confinement and transport, a key investigation is to study the
> global confinement properties of high power density, low ν∗ ST
> plasmas, including access into the H-mode and the L- vs H-mode
> confinement trends themselves. A related area of interest is the
> possible suppression of electrostatic and electromagnetic turbulence
> at low ν∗ due to an increase in orbit-averaged good curvature. Aspect
> ratio scans are key to assessing this suppression at low R/a. It is
> important also to investigate the effect that fully relaxed current
> profiles would have on transport and MHD stability, and vice-versa.
>
> MHD stability studies are an important part of developing high-beta,
> high bootstrap scenarios. Key elements of this study include
> investigating the low- and high-n stability with respect to varying
> aspect ratio, shaping (elongation and triangularity), conducting wall
> stabilization (including the effect of resistive wall modes), and
> rotation control.
>
> Non-inductive startup, current sustainment and profile control are key
> elements of the discharge scenarios for fully non-inductive high-beta,
> steady-state ST plasmas. Such plasmas are relevant to the basic plasma
> science as well as fusion science relevant to eventual energy
> production. Helicity injection appears to be a credible start-up
> technique, and has the potential for driving current in the outer
> region of the plasma. High harmonic fast waves could be used both for
> heating and for controlling the current profile in the central region
> of the plasma. The bootstrap current can, in certain configurations,
> support over 80% of the total current, thus reducing the recirculating
> power requirements. Important for producing high bootstrap scenarios
> is the production of a high-q, high-beta, low ν∗ equilibrium, with
> careful bootstrap alignment to insure MHD stability. It is also
> important to study high bootstrap current scenarios that require
> little or no edge current drive. Only current profiles that are
> fully-relaxed will allow a credible assessment of the efficacy of the
> current drive techniques and the MHD stability of these non-inductive
> plasmas.
>
> In the final category, SOL and divertor physics, a key investigation
> is to assess the role of pressure-driven transport across the SOL and
> the resulting SOL thickness. Low aspect ratio configurations have high
> mirror ratios (\~3 to 5) and high trapped particle fractions (\~0.75)
> near the edge, and this may influence the properties of the SOL.
> Finally, in ST configurations, it is the outer divertor plate that
> will take up the bulk of the heat and particle flux, and it will be
> important to investigate the effect of varying configurations
> (single-null, double-null, and \"natural\" divertor) on the SOL flux
> tube expansion and on divertor heat loads.

**2) Do the above issues requires**

**(a) launching new experimental facilities and/or theoretical
activities?**

> Yes, the present experimental activities (mainly START, CDX-U and HIT)
> are quite important but not sufficient in size and scope in answering
> the key ST physics questions such as the plasma confinement under
> strong auxiliary heating, beta limit, current profile control, access
> to advanced ST regimes in fully relaxed current profiles, etc. We will
> need a PLT-class ST experiment with adequate capabilities in heating
> and current drive for significant pulse lengths to answer these
> questions.

**(b) expanding the current experimental and theoretical activities?**

> Yes, the present level of resource allocated to the ST experiments is
> very small (\< 0.5% of the U.S. fusion budget). The level of ST
> theory, modeling, and other studies is at present very modest.

**(c) or they can be addressed at present level of research?**

> The present level of research can be continued to address several of
> the physics issues in support of larger ST experiments in high beta
> collisionless regimes. Some degree of upgrade of existing small ST
> experiments (such as the CDX-U by replacing the center stack and
> enhancing the power supplies, etc.) can be envisioned toward 300 − 400
> kA range device with only a short pulse length ≤ 100 msec. The
> limitations of the small facilities (START, CDX-U, HIT, etc.) can be
> put in the table below in contrast with the experimental capabilities
> needed to prove the ST physics principles.

  ---------------------------- ---------- -------- ------------ ----------
                               **I (MA)** **R      **Paux       **τpulse
                                          (m)**    (MW)**       (s)**

  ---------------------------- ---------- -------- ------------ ----------

  ---------------------------- ---------- -------- ------------ ----------
  **CDX-U**                    0.1        0.3      0.02         0.02

  ---------------------------- ---------- -------- ------------ ----------

  ---------------------------- ---------- -------- ------------ ----------
  **HIT**                      0.2        0.3      0            0.01

  ---------------------------- ---------- -------- ------------ ----------

  ---------------------------- ---------- -------- ------------ ----------
  **START**                    0.25       0.32     1            0.05

  ---------------------------- ---------- -------- ------------ ----------

  ---------------------------- ---------- -------- ------------ ----------
  **GLOBUS-M (approved)**      0.35       0.4      1.5          0.2

  ---------------------------- ---------- -------- ------------ ----------

  ---------------------------- ---------- -------- ------------ ----------
  **Proof-of-Principle**       1          0.8      6            5

  ---------------------------- ---------- -------- ------------ ----------

**3) What is the world-wide research plan (outside U.S.) to address
the**

**above issues?**

> The main thrust of the world-wide research plan (outside U.S.) has
> been to build a few 100 kA to MA range ST devices with ohmic heating
> mainly to extend the tokamak data base for aspect ratios below 2. The
> most significant of these are the Mega Ampere Spherical Tokamak (MAST)
> at Culham just approved by EURATOM, and the smaller GLOBUS-M under
> construction at the Ioffe Institute, Russia.
>
> The MAST experiment represents an appropriate extrapolation of the
> successful START experiment to the MA level of plasma current, and a
> cost-effective utilization of the Culham experimental facility. The
> main features and capabilities of MAST as approved presently include:
>
> ∗ Poloidal field coils internal to the vacuum vessel,
>
> ∗ Plasma startup primarily by combining radial compression and
> solenoid induction,
>
> ∗ Absence of a nearby conducting shell limiting the stability beta,
>
> ∗ Neutral beam injection as the primary auxiliary heating and current
> drive, and
>
> ∗ A pulse length of up to 1 s at 1 MA in plasma current.
>
> These baseline design features do not permit outboard rf launchers and
> limit the achievable bootstrap current fraction to about 50% at 1 MA
> in plasma current.
>
> An ST experiment that emphasizes features of interest to fusion core
> applications to VNS, Pilot Plant, and Power Plants will have features
> and capabilities[^8] complementary to the above. These include:
>
> ∗ Poloidal field coils external to the vacuum vessel,
>
> ∗ Plasma startup by solenoid induction and/or by noninductive means,
>
> ∗ Use of conducting shell for stability at high beta,
>
> ∗ RF heating and current drive launched from the plasma outboard,
>
> ∗ A pulse length of 5 s to permit fully relaxed current profiles that
> are dominated by pressure gradient driven currents but controlled by
> noninductively driven currents.
>
> These features, given adequate heating and current drive power (≥6 MW
> for 5 s pulse length), will permit full scale investigation of the
> advance physics regimes in ST plasmas.

**4) What should be the US role/contribution to the world-wide research.
In particular:**

**(a) Is it necessary to have more than ONE new international
experimental facility?**

> As in the case for conventional tokamaks, stellarators, and other
> confinement concepts, there is a clear merit for building in the world
> more than one ST experiments at the proof-of-principle level. With
> scarce resources, however, it is increasingly necessary and desirable
> to make sure that the world ST experimental activities and
> capabilities are well focused and complementary as much as possible
> and practical. Well coordinated planning and execution of the world ST
> research program is therefore needed. This has been the case for the
> ST research since its initiation about 7 years ago, through
> collaborations and workshops among small and growing research efforts.
> The success of this process is underscored by a rapid increase of
> small ST experiments in operation or under construction, including
> CDX-U, HIT, Medusa (U.S.), START, SPHEX (U.K.), TS-3, TST (Japan),
> GLOBUS-M (Russia), and ETE (Brazil).
>
> The initial success of the small ST experiments has increased the
> interest in the world fusion community in building new devices to
> prove the physics principles needed for attractive future D-T-fueled
> ST applications. The broad range of the physics issues identified so
> far for this purpose have led to cost-effective designs that are
> highly complementary in the basic experimental capabilities. This will
> substantially increase the confidence of the fusion community in
> adequately addressing the key physics issues in a timely manner. In
> order to cover the full range of scientific issues, it is necessary to
> build more than one new ST experimental facilities at the MA level in
> plasma current.

**(b) Given the world-wide plan, which areas US program should focus**

**on?**

> The US program should focus on advanced ST physics and ST related
> innovation aimed at attractive ST regimes from the VNS and power
> reactor point of view. These areas of advanced tokamak and innovation
> are traditionally the strength of the US fusion program. Such a focus
> will ensure an adequate investigation of the rich physics contained in
> the plasma science of high temperature magnetized plasmas. Though
> these research areas are quite challenging requiring community wide
> (experiment, theory, modeling, and reactor studies) participation,
> they represent the highest leveraged research areas with high
> potential pay-off requiring relatively modest resource allocation.
> This strategy of fusion research is consistent with the recent
> recommendations of FEAC.

**Third Category:**

**It is understood that as a concept is developed further and is matured
and is (its) potential impact is realized and/or confirmed, larger
research program including experimental facilities will be required.
What is an appropriate mix of research activity for this concept among
large national facilities and mix of small supporting experiments,
theory, modeling, and simulations, and concept design studies?**

> To ensure that the full potential of the ST concept is developed and
> available for future energy applications, it will be necessary to
> continue innovative studies that could lead to truly practical fusion
> power sources. Innovative studies will require, in addition to the MA
> proof-of-principle experiments, an appropriate mix of small well
> focused supporting experiments, theoretical investigations, improved
> modeling and simulation, and advanced concept design studies.
> Particular areas of interest include the possibility of incorporating
> advantages from potentially related configurations such as the
> spheromak, the FRC, and the RFP.
>
> Assuming that the ST test at the MA level is successful, an ST
> experiment in the range of 5−10 MA in plasma current will produce
> performance equivalent to TFTR and JET. Such an ST device will require
> only modest size (R ≤ 1 m) and magnetic field (Bt0 ∼ 2 T). The
> facility investment can be kept modest (a small fraction of the
> TFTR/JET cost) by utilizing existing conventional tokamak facilities.
> Such an experiment should verify the physics performance of a
> D-T-fueled fusion core under significant self-heating.

[^1]: Rewoldt, G., W. Tang, S. Kaye, and J. Menard, accepted for
    publication in Physics of Plasmas, 1996

[^2]: Houlberg, W., private communication, 1996

[^3]: Ono, M, Phys. of Plasmas [2]{.underline} (1995) 4075

[^4]: Brambilla, M. and T. Krucken, Nuc. Fusion [28]{.underline} (1988)
    1813

[^5]: Ehst, D. and C.F.F. Karney, Nuc. Fusion [31]{.underline} (1991)
    1933

[^6]: Hugill, J. , U.S./Japan ST Workshop, PPPL, Nov. 1995

[^7]: Peng, M. , IAEA, Seville, 1994

[^8]: Y-K. M. Peng, R. J. Goldston, R. J. Colchin, E. B. Hooper, S. M.
    Kaye, M. Ono, R. E. Siemon, R. Stambaugh, A. J. Wootton, "Summary of
    Workshop on Establishing the Physics Basis Needed to Access the
    Potential of Compact Toroidal Reactors, Oak Ridge, Tennessee, July
    19−22, 1994," to appear in Fusion Technology.