# Presentation: 850703

## Optimization of a High- Steady-State Tokamak Burning Plasma Experiment Based on a High- Steady-State Tokamak Power Plant

- D. M. Meade, C. Kessel, S. Jardin
- Princeton Plasma Physics Laboratory

- Presented at IEA Workshop on Optimization of High- Steady-State Tokamaks
- General Atomics
- February 14, 2005

## Tokamak Based Power Plant Studies have Identified Attractive High- Steady-State Configurations with A ≈ 4

- • Three decades of systematic studies in the US have surveyed  the range of possibilities available for a tokamak power plant and have identified the ARIES-RS/AT (A = 4) designs as  the most attractive possibilities.
- •  Other studies favoring highish aspect ratio for high- steady-state include:  TPX (A= 4.5), ITER-HARD (A = 4), ASSTR (A = 4)
- • The FIRE design studies initiated in 1999 adopted the ARIERS/AT physics and plasma technology design characteristics including :

- •  Strong shaping –> Double null    x = 0.7, x = 2, Aspect ratio ≈  4
- •  Reactor level  BT=  6 - 10 T  and plasma density =  2 - 4 x 1020m-3
- •  LHCD/ICFWCD - no momentum input    •  All metal PFCs  W divertor
- •  Internal fast control coils   •  RWM coils integrated into FW of port plugs
- - *LN cooled coils* provide sufficient pulse length and small size –>  low cost

## Optimization of Cu Coil BPX (e,.g, FIRE)

- •  Optimization Depends on Goals and Constraints
- 	Realistic engineering constraints must be imposed
- •  Optimization of inductively-driven BPX with Cu Coil (e,.g, FIRE)
- 	1991 CIT Study  (LLNL Super Code-Galumbos) W. Reiersen
- 	2000 FIRE (FIRE Sale)  J. Schultz
- 	2001 FIRE (BPSC) S. Jardin,C. Kessel, D. Meade, C. Neumeyer
- •  Optimization of High- Steady-State Modes in FIRE
- 	2002 SOFE Meeting FIRE AT C. Kessel
- 	2002/2004 IAEA FIRE D. Meade et al

- References
- 1. J. Galumbos et al  Fusion Tech. **13**, 93, 1988
- 2. W. Reiersen
- 3. J. Schultz -
- 4. S. Jardin, C. Kessel et al  Fusion Science and Technology **43**, 161 2003
- 5. A High-Aspect-Ratio Design for ITER,  J. C. Wesley et al Fusion Tech. **21,** 1380 1992.
- 6. Y. Seki et al., (1991). Rep. JAERI-M 91-081, JAERI. Naka.

## Compact Ignition Tokamak Optimized at A = 3.5

**Compact Ignition Tokamak Optimized at A = 3.5**

## The Systems Code was Updated and Calibrated Based on 3-D Finite Element Stress Calculations for FIRE.

- E = 0.144 I0.93 R1.39 a0.58 n200.41 B0.15 Ai0.19 0.78 P heat -0.69 H(y,2)
**Density Limit:**           		n20 < 0.75 nGW = 0.75 IP/a2
**H-Mode Power Threshold:**  	Pth > (2.84/Ai) n200.58 B0.82 R a0.81
**MHD Stability:**     		N =  / (IP/aB) <  3.0
- PAUX,  Q = PFUSION/PAUX, qCYL _or_ qMHD,,  ZEFF  all held fixed
**Engineering Constraints:**	1. Flux swing requirements in OH coil (V-S)
- 			2. Coil temperature not exceed 373o K
- 			3. Coil stresses remain within allowables
**Configuration Concept:**  	1. OH coils _not linking_ TF coils, or
- 			2. OH coils _linking_  TF coils—ST-like

**The Systems Code was Updated and Calibrated Based on 3-D Finite Element Stress Calculations for FIRE.**

- Kessel, Jardin 2002

## Optimization of Cu Coil BPX (e,.g, FIRE) Using BPSC

## Slide 7

## Optimization at Smaller Size and Higher Aspect Ratio as Confinement Improves

**Optimization at Smaller Size and Higher Aspect Ratio as Confinement Improves**

- Major Radius
-  (m)

- 10T

## Optimization is not Sensitive to Variation of Elongation with Aspect Ratio

**Optimization is not Sensitive to Variation of Elongation with Aspect Ratio**

- Major Radius
-  (m)

## Normalized pulse length (CR)in FIRE is the same as ITER 

**Normalized pulse length (********CR****)in FIRE is the same as ITER **

- N = 1.8
- Pf/V = 5.5 MWm-3
- fbs ≈ 25%

## Optimization of High- Steady-State* Modes in a Cu Coil BPX (e,.g, FIRE)

- • Optimization of AT Modes in a specific FIRE (Fixed A, R, B < Bmax, etc)
- •  Use a 0-D Systems Code to calculate a large data base(~ 50, 000) of possible solutions as parameter space is scanned.
- •  Impose engineering constraints on pulse length (TF ohmic and nuclear heating, divertor target and baffle heat loads, vacuum vessel nuclear heating and first wall surface heating) to define operating space.
- •  Use J_solver and PEST to validate stability and required current profiles.
- •  Use TSC to confirm evolution of  integrated discharge.

- * Steady-state = 100% non-inductive,  q/q < few % for several CR , div, FW

## 0-D Operating Space Analysis for FIRE AT

- Heating/CD Powers
- ICRF/FW, 30 MW
- LHCD, 30 MW
- Using CD efficiencies
- (FW)=0.20 A/W-m2
- (LH)=0.16 A/W-m2
- P(FW) and P(LH) determined at r/a=0 and r/a=0.75
- I(FW)=0.2 MA
- I(LH)=Ip(1-fbs)
- Scanning Bt, q95, n(0)/<n>, T(0)/<T>, n/nG, N, fBe, fAr

- Q=5
- Constraints:
- (flattop)/(CR) determined by VV nuclear heat (4875 MW-s) or TF coil (20s at 10T, 50s at 6.5T)
- P(LH) and P(FW) ≤ max installed powers
- P(LH)+P(FW) ≤ Paux
- Q(first wall) < 1.0 MW/m2 with peaking of 2.0
- P(SOL)-Pdiv(rad) < 28 MW
- Qdiv(rad) < 8 MW/m2

## FIRE’s Q = 5 AT Operating Space

- • A data base of ~ 50,000 operating points is calculated with 0-D code
- • Engineering constraints are imposed to generate the operational boundaries shown below
- • Potential operating points are examined in more detail-PEST, TSC, etc

## FIRE AT Mode Operating Range is Limited by Nuclear Heating of Vac Vessel & First Wall Not by Cu Coils

- Nominal operating point
- 	• Q = 5
- 	• Pf = 150 MW,
- 	• Pf/Vp = 5.5 MWm-3 (ARIES)
- 	• ≈ steady-state 4 to 5 CR
- Physics basis improving (ITPA)
- 	• required confinement H factor
-         and N attained transiently
- 	• C-Mod LHCD experiments
- 	  will be very important
- First Wall is the main limit
- 	• Improve cooling
- 	• revisit FW design

- Opportunity for additional improvement (optimization).

## “Steady-State” High- Advanced Tokamak Discharge on FIRE

**“****Steady-State” High-******** Advanced Tokamak Discharge on FIRE**

- Pf/V = 5.5 MWm-3
- n ≈ 2 MWm-2
- B = 6.5T
- N = 4.1, t = 5%
- fbs = 77%
- 100% non-inductive
- Q ≈ 5
- H98 = 1.7
- n/nGW = 0.85
- Flat top Duration
-    = 48 E
-    = 10 He
-    =  4 cr

- FT/P7-23

## Additional Opportunities to Optimize FIRE for the  Study of ARIES AT Physics and Plasma Technologies 

**ARIES AT (********N**** ≈ 5.4, f****bs**** ≈ 90%)**

**12**

**Additional Opportunities to Optimize FIRE for the  Study of ARIES AT Physics and Plasma Technologies**

## FIRE-AT Approaches the Parameters Envisioned for  ARIES-Power Plant Plasmas

| Physics Items | FIRE-AT | ARIES-RS/AT |
| --- | --- | --- |
| x | 2.0 | 2.0 |
| x | 0.7 | 0.7 |
| Configuration | DN | DN |
| N t | ~ 4,  5% | ~ 5,  9% |
| % Non-inductive | 100 | 100 |
| % bootstrap | 77 | 90 |
| Equilibration % | ~100 | steady |
| Plasma rotation | Very low | Very low |
| RWM Coils
(rel. to First Wall) | Integrated
With FW | Inside TF
Outside VV |
| On axis CD | ICFW | ICFW |
| Off axis CD | LH | LH |

| Technology Items | FIRE-AT | ARIES-RS/AT |
| --- | --- | --- |
| B(T) | 6.5 | 6 - 8 |
| Ip(MA) | 5 | 11 |
| Core Power Density (MWm-3) | 5 | 5 - 6 |
| FW -N (MW/m-2) | 2 | 4 |
|   FW - Prad (MWm-2) | 15 | 100 |
| First Wall | Be | Mo |
| Div Target (MWm-2) | 5 - 20 | 5 - 20 |
| Divertor Target | W | W |
| Pulse Length(s, cr) | 40, 5 | months |
| Cooling       Divertor, First Wall | Steady
inertial | Steady
steady |

## Concluding Remarks

- •  FIRE is very close to the optimum aspect ratio and size for an inductively-driven H-Mode burning plasma experiment using LN-cooled coils.
- •  The present FIRE configuration is also capable of producing AT plasmas with characteristics approaching those of ARIES-RS with pulse lengths sufficient to study High- Steady-state burning plasmas with fusion power densities of ≈ 5 MW m-3.
- • The present FIRE AT regimes are limited by the first wall and vacuum vessel and not the TF coil.
- 		- improve FW and Vac Vessel cooling  ––>   6 CR
- 		- change TF conductor to OFHC (Bt ≤ 7T) ––>  ≈ 12 CR
- •  A bottoms up optimization of a FIRE for AT operation only has not been done, but the present case must be fairly close to the optimum.