# Presentation: 606581

## Progress in Spherical Torus Research

- Roger Raman
- University of Washington, Seattle
- and the NSTX National Research Team
- ICC Workshop
- 28-30 May 2003, Seattle, WA

## STs bridge a gap between certain ICC concepts and a tokamak

- Spherical tokamaks present both the **strong toroidicity** effects of compact tori
- (notably magnetic shear) and the **good stability** properties provided by the
- external toroidal field of conventional tokamaks.

*  **Ludwig, **INPE*

## Reduction of Aspect Ratio leads to a new physics regime requiring new enabling systems

- High
- Large bootstrap current (> 70%)
- High dielectric constant  ~ 30-100, requires RF waves at high harmonics of i
- New methods to initiate current without reliance on inboard coils
- Increased divertor and wall loading

## Validation of high  and bootstrap current fraction physics and solenoid-free plasma startup and sustainment would lead to attractive reactor designs

- ARIES ST parameters
- R 			3.2 m
- a			2 m
**			3.4
- Ip			29 MA
- FBS		96%
- PCD		28 MW
- BT			2.1 T
- T			50%
- Pfusion 	   	2980 MW

- Najmabadi et al, *Fus. Eng. Desig.* **65** (2003) 143

## Capability to build low cost machines will speed up fusion development path 

- Most of the current driven by the plasma
- Small auxiliary current drive needed
- Low values of BT leads to cheaper machines to enable construction of intermediate designs to demonstrate a level of scientific understanding that allows for extrapolation with confidence

## ST Program has machines that make a connection with tokamaks as well as with spheromaks

- NSTX & MAST: PoP Experiments
- Pegasus: Very Low Aspect ratio experiment, RF current drive to develop sustainment and startup
- HIT-II: CHI for Startup and sustainment
- CDX-U: Li-Wall development, RF for sustainment
- Other STs also contribute to these studies (TST-2, ETE, Globus-M, HIST)

## Low cost has allowed construction of many small machines

*M. Peng, ORNL*

## Results from the large STs

## NSTX has shown high  capability at 1 MA

- T = 35% determined by magnetic analysis
- BT = 0.3T, A = 1.4,  = 2.0,  = 0.8
- N ~ 6 also obtained on MAST
- High beta discharges on both machines had no disruption, limited by operational constraints

*D. Gates, PPPL*

- NSTX

## Substantial bootstrap current fraction achieved in NSTX

- Goal is to control profiles of both pressure & current to maximize stability and bootstrap current contribution

*J.Menard, PPPL & S.Sabbagh, Columbia*

- • Vloop 	     ~ 0.1V
- 	for ~ 0.3s
- • INBI/Ip       = 0.18
- • Ibootstrap/Ip = 0.42
- • Inon-ind/Ip   = 0.6

## 0.1MA of HHFW current drive inferred by circuit analysis

- HHFW are compressional fast Alfvén waves,
-        ~ kVA ~ (6-12) D
- For NSTX  ~ 4% - 38%, choose / k11 ~ Vte

- • 2 discharges with similar ne(r) , Te(r)
- • V not caused by dIi/dt

*H.R. Wilson, PPPL , P. Ryan, ORNL*

## CHI has generated substantial toroidal current in NSTX

- Goal is to control discharge evolution to promote reconnection of toroidal current into closed flux surfaces

## NSTX

**NSTX**

**MAST**

- MAST & NSTX show good confinement (Tau-E > 100ms), in general agreement with ITER scaling (IPB98y2), while extending the range of R/a in database

- Little difference in L and H mode and no power degradation on MAST
-  Similar results for Tau-E from MAST and NSTX at high TF
- MAST has obtained H-modes in a ohmic plasma

*MAST & NSTX Teams*

## 

- In the ‘Natural Divertor’ configuration, inboard limited plasmas in an ST have an **expanded outer SOL** and reduced, evenly spread contact on the centre limiter.
- MAST Natural Divertor shot #4573 exhibited H-mode features: ELM-free periods with spontaneous density rise between ELMs.

**Formation of the Natural Divertor at low R/a:**
*As aspect ratio A = R/a reduces, exhaust plume expands and is diverted from the centre post*

****

*M. Gryaznevich, MAST*

## Large STs have similar cross-section and main plasma parameters as conventional aspect ratio tokamaks of similar size

**MAST**** ****and**** ASDEX-Upgrade ****in Europe**

**NSTX**** ****and**** DIII-D ****in USA**

***Recent results:***
**NSTX:**
- t>30%; ******N ****~ 6*** *>N(no-wall); pol~ 1, Hpby2~1.8, Te~ 3.5keV;
**MAST:**
******N ****~ 6*** *>N(no-wall); pol~ 2, Hpby2~ 2; Ti, Te~ 3keV; G > 2

|  | MAST | NSTX |
| --- | --- | --- |
| R (m) | 0.85 | 0.86 |
| A (m) | 0.65 | 0.68 |
| k | 2.4 (3) | 2.5 |
| Ip (MA) | 1.2 (2) | 1.5 (1) |
| BT (T) | 0.5 | 0.45 |
| PAUX (MW) | 3 NBI (5)
1 EC (1.5) | 5 NBI (5)
4 FW (6) |
| Tpulse (s) | < 0.65 (5) | 0.55 (5) |

- Red shows design values

*MAST and NSTX*

## New fueling method (inboard mid-plane gas injection) developed on MAST

**Inboard fuelling**

- Inboard midplane fuelling improves H-mode access** ***(A Field)*

**Outboard fuelling**

- Initial estimates of fuelling efficiency ~ 50%

- 8-pellet gas gun injector (used on RTP, FOM);
- 3 pellet sizes: 0.5, 1, 2x1020 atoms of deuterium
- Pellet velocity 300 - 1200m/s

- Pellet injection provides internal fuelling in H-mode:
*K Axon, C Ribeiro, S Shibaev*

- Similar inboard fueling results on NSTX

*MAST*

## Other results from large STs

- MAST uses a merging-compression method using the outer PF coils to generate solenoid-free startup current
- Power handling studies on MAST indicate that most of the heat load goes to the outboard divertor legs
- Unlike in conventional aspect ratio tokamaks, an empirical ***N** *4*l**i* limit has not been justified in STs. This means that reducing *l**i *through profile control would allow further extension of the beta limits

## Results from the medium sized  STs

## Pegasus explores extremely low-aspect ratio physics in high- plasmas, with the goal of minimizing the central column while maintaining good confinement and stability

- t up to 20% and IN up to 6.5 achieved *ohmically*

- ST - Spheromak
- Overlap region

- Pegasus parameters
- A 	1.12 - 1.3
- R	0.2 - 0.45 m
- Ip	0.16 MA
- BT	~0.15 T
- 	1.4 - 3.7

*Pegasus Team*

## Toroidal field utilization exhibits a “soft limit” around unity

- Maximum Ip ~ Itf in almost all cases
- Limit is not disruptive, IP saturates or rolls over
- Large resistive MHD instabilities degrade plasma as TF decreases
- Reduced available volt-seconds as TF is reduced

- IP/Itf is a figure of merit for
- access to low-A physics

*Pegasus Team*

## New method for CHI startup on HIT-II generates useful non-inductive current and improves the performance of inductive discharges

- New method *“Transient CHI startup”* developed on HIT-II
- Sequence of eleven discharges shows how CHI startup is robust and saves volt-seconds
- Sequence is initiated after Ti-wall conditioning. There is no wall conditioning between shots
- Method is applicable to a pre-charged transformer

*HIT-II Team*

- HIT-II Parameters
- R/a 	0.3/0.2
-        	1.5
- Bt	0.5 T
- Ip	265 kA

## Under gas loaded wall conditions, good coupling to induction can be regained if CHI discharges can be initiated at lower neutral gas pressures

- Indicates need for good pre-ionization methods
- Method does not rely on any time changing poloidal field coil currents, therefore attractive for reactors where PF coils will be outside blanket structures
- Method to be tested on NSTX

*HIT-II Team*

## CDX-U studies role of Lithium PFCs on plasma operations and practical implementation issues

- Recycling & Fueling
- Impurity reduction
- Performance enhancement
- Radiation lossses, core Li accumulation
- Safety issues
- Motion of liquid during PF ramps, disruptions

- Heat/Li shield

- Tray temp. monitored

- TiC coated shield on CS

- • Tray has a radius of 34cm, width of 10cm, depth of 64cm, & Temp. ~ 250C
- • Electrical break between two halves, Liquid Li injected into both halves

*CDX-U Team*

- CDX-U parameters
- R/a 	0.34/0.24m
- Ip	70 kA
- Bt	0.2T

## Initial Li experiments with tray temp. @ 300C show improvement in plasma performance

- Plasma experiments started within hours of filling the tray
- Required five fold increase in gas fueling
- Immediate 30% increase in Ip
- 30% increase in discharge duration
- Virtual elimination of OII, CII radiation
- 2-3x reduction in D emission
- No evidence for liquid Li ejection due to disruptions

- Plasma experiments started within hours of filling the tray

*CDX-U Team*

- 0ne half of CDX-U is shown

- UCSD
- Li injector

- Center line

## Ion heating by reconnection (during  an IRE) observed by impurity ion Doppler in TST-2

- Impurity temperature increases up to 400-550 eV from 50-100 eV
- Te and ne decrease

- TST-2 parameters
- R/a   	0.36/0.23
- Bt	0.4 T
- Ip	200 kA(design)

- Achieved parameters
- T 	5.7% (N 2.7)
- E 	3 ms
- Ip	90 kA
- Bt	0.2 T

*TST-2 Team, Japan*

## Non-inductive current initiation by ECH in TST-2

- ECH (2.45 GHz)   1 kA / 1 kW
- Low gas pressures  low collisionality
- Vertical field with positive curvature   trapped electrons

*TST-2 Team, Japan*

## ETE goals are to study Alfvén wave heating and edge plasma conditions

- Ip: 40 - 60kA, pulse up to 12ms
- Evidence for IREs
- Large currents induced in the walls
- Density ≤ 4E19 m-3, Te ≤ 160 eV

- New machine in the process of being fully commissioned
- ETE Parameters
- R/a 	0.3/0.2
- 	1.6 - 1.8
- 	0.3
- Bt	0.4 - 0.6T (<0.8T)
- Ip	0.2 - 0.4MA (design)

* **ETE Team, Brazil*

## Summary 

- Remarkable progress in ST physics and technology in just a few years
- PoP Experiments have been able to validate many of the predicted physics
- T ~ 35% achieved on NSTX
- Good E > 100 ms achieved by NSTX and MAST
- Good progress with divertor power loading studies on MAST
- HIT-II has found a viable solution for plasma startup by CHI
- Liquid Li experiments on CDX-U are showing immediate plasma performance improvement
- Pegasus is finding a MHD limit as Ip approaches IT, results may eventually relate to spheromaks
- Temperature increase during IREs seen on TST-2, current drive methods being developed