# Presentation: 821838

## Some Highlights During 50 Years of Fusion Research

- Dale Meade
- Fusion Innovation Research and Energy®
- Princeton, NJ
- United States of America

- 22nd IAEA Fusion Energy Conference
- October 13-18, 2008
- Geneva, Switzerland

## Slide 2

## Fusion Prior to Geneva 1958

- A period of rapid progress in science and technology
- N-weapons, N-submarine, Fission energy, Sputnik, ....
- Controlled Thermonuclear Fusion had great potential
- Much optimism in the early 1950s with expectation for a quick solution
- Political support and pressure for quick results
- Many very  “innovative” approaches were put forward
- Early fusion reactors - Tamm/Sakharov, Spitzer
- Reality began to set in by the mid 1950s
- Collective effects - MHD instability (1954)
- Bohm diffusion was ubiquitous
- Meager plasma physics understanding led to trial and error approaches
- A multitude of experiments tried and ended up far from fusion conditions
- Magnetic Fusion research in the U.S. declassified in 1958

## Fusion Plasma Physics, a New Scientific Discipline, was born in the 1960s

- Theory of Fusion Plasmas
- Energy Principle developed in mid-50s became a powerful tool for assessing macro-stability of various configurations
- Resistive macro-instabilities
- Linear stability analyses for idealized geometries revealed a plethora of microinstabilities with the potential to cause anomalous diffusion Trieste School
- Neoclassical diffusion developed by Sagdeev and Galeev
- Wave propagation became basis for RF heating
- Experimental Progress (some examples)
- Most confinement results were were dominated by instabilities and ~ Bohm diffusion
- Stabilization of interchange instability by Min|B| in mirror - Ioffe
- Stabilization of interchange in a torus by Min<B> in multipoles - Ohkawa/Kerst
- Quiescent period in Zeta due to strong magnetic shear in self organized state
- Confinement gradually increased from 1 B to 5-10 B for low temp plasmas
- Landau Damping demonstrated

## Stabilization of MHD Interchange by Geometry (minimum |B|) in a Mirror Machine

- IOFFE IAEA Salzburg 1961,  J Nuc Energy Pt C  _7_, p 501  1965

- Well Formed

- Increasing Bmultipole

## A Gathering in the Model-C Control Room

- T. Stix , H. Furth, E. Teller, L. Strauss, M. Rosenbluth, M. Gottlieb

- Late 1960s

## 1968-69 T-3 Breaks Bohm, Tokamaks Proliferate

-  Hints of a major advance at IAEA Novosibirsk 1968, but skeptics abound
-  Thomson Scattering (Peacock/Robinson) Dubna 1969 confirms Te ≈ 1 keV
-  Energy confinement ≈ 30 B - Bohm barrier broken for a hot plasma
-  Skeptics converted to  advocates overnight, Model C Stellarator converted to Symmetric Tokamak (ST) in 6 months, T-3 results are quickly reproduced.
- During the 1970’s ~  many medium size (Ip < 1 MA) tokamaks (TFR, JFT-2a, Alcator A, Alcator C, ORMAK, ATC, PLT, DITE,  DIII, PDX, ASDEX, ... were built with the objectives of :
-  Confinement scaling with size, Ip, n, T,.......
-  Auxiliary heating (compression, ICRF, NBI, ECRH, LH )
-  Current Drive (LH, NBI, ... )
-  Impurity control (limiters, divertors)

- 40 years ago

## Fusion was Prepared for a Major Next step when Opportunity Knocked (1973 Oil Embargo)

-  Amid calls for increased energy R&D, Fusion budgets rise sharply
-    -  US Fusion budget increased a factor of 15 in 10 yrs.
-  Four Large Tokamaks approved for construction less than a decade after T-3
-  TFTR conservative physics/strong aux heating  const began 1976
-  JET shaped plasma - const began 1977
-  JT-60 poloidal divertor- const began 1978
-  T-15 Superconducting TF (NbSn) const began 1979

- These were very large steps, taken before all the R&D was completed.
- 	Plasma Current 	0.3 MA   =>    3MA to 7MA
- 	Plasma Volume	    1 m3   =>  35 m3 to 100 m3
- 	Auxiliary Heating	0.1 MW  =>  20 MW to 40 MW

## Optimism about Confinement Increased in the late 1970s

- Trapped Ion instabilities were predicted in the early 1970s to be a threat to the achievement high Ti in tokamak geometries.
- In 1978, Ti ~ 5.8 keV was achieved in a collisionless plasma reducing concerns about Trapped Ion instabilities.  Ti was increased to 7 keV in 1980.
- In ~ 1979 Alcator A with only ohmic heating achieved nE ≈ 1.5 x 1019 m-3 s, consistent with  optimistic scaling E~ na2.

- 30 years ago

## Auxiliary Heating Reveals New Trends 1981 

- Auxiliary heating allowed controlled experiments to reveal the scaling of the global global confinement time.
- Confinement degradation observed as heating power was increased - Low mode scaling would threaten objectives of the large tokamaks, and tokamak based reactors.

- ISX-B

## H-Mode Discovered on ASDEX- 1982 

**H-Mode Discovered on ASDEX- 1982**

-  Facilitated new insights and understanding of transport, and
-  Provided the baseline operating mode for ITER

- F. Wagner, IPP

## Tokamak Optimization

-  By the mid 80s (~ 1984)
-  It was clear tokamak performance would need to be improved, if the tokamak were to lead to an attractive fusion power source.
- The benefits of cross-section shaping for increased confinement and beta were demonstrated and understood in Doublet IIA and Doublet III.
-  The  limit formulation by Troyon and Sykes  provided a design guide for .
-  Empirical scaling formulations (e.g., Goldston scaling) provided guidance for E
-  An understanding of divertors emerged from JFT-2a, PDX, ASDEX, DIII, DITE.
-  A second generation of flexible optimized tokamaks:                                                 	DIII-D, AUG, JT-60U, PBX, Alcator C-Mod were built in the late 1980s to extend  	and develop the scientific basis for tokamaks.

## Large Tokamaks Extend Plasma Parameters

- After about 6 years of construction TFTR, JET and JT-60  began operation 1982-84.
- By the mid 80s, after  4 years of operation the plasma parameter range had been significantly extended
- Ti~ 20 keV and ne(0)E ~ 1.5x1019 m-3 s with neutral beam injection
- ne(0)E ~ 1.5x1020 m-3 s  and Ti~ 1.5 keV with pellet injection
- H-Mode extended to large tokamaks, new improved performance regimes discovered.
- Bootstrap current  and current drive extended to MA levels
- Divertor extended to large scale
- Complex Technology demonstrated at large scale
- Enabling Technology - Neutral beams, pellet injection, PFCs

## Fusion  Temperatures Attained,

**	****	****	****Fusion Confinement One Step Away**

- ni(0)ETi
- increased by ~107
- since  1958

- JAEA

## Significant Fusion Power (>10MW) Produced 1990s

- 1991 JET  90/10-DT, 2 MJ/pulse, Q ~ 0.15, 2 pulses
- 1993-97 TFTR 50/50-DT, 7.5MJ/pulse, 11 MW, Q ~ 0.3, 1000 D-T pulses,
- Alpha heating observed, Alpha driven TAEs  - alpha diagnostics
- ICRF heating scenarios for D-T
- 1 MCi (100 g) of T throughput, tritium retention
- 3 years of operation with DT, and then decommissioned.
- Advanced Tokamak Mode Employed for High Performance
- Improved ion confinement TFTR, DIII-D,  QDTequiv ~ 0.3 in DIII-D 1995
- nET record => QDTequiv in JT-60U DD using AT mode 1996
- Bootstrap and current drive extended
- 1997 JET 50/50-DT  22MJ/pulse, 16 MW, Q ~ 0.65,  ~100 D-T pulses
- Alpha heating extended, ICRF DT Scenarios extended,
- DT pulse length extended
- Near ITER scale D-T processing plant
- Remote handling

## The Next Challenge -Sustainment of Fusion Plasma Conditions

- Steady-state operation is a highly desirable characteristic for a magnetic fusion power plant. This requires:
- Sustained magnetic configuration
- The stellarator (helical) configuration is inherently steady-state, or
- Advanced tokamak with high bootstrap current fraction and moderate external current drive is also  a possible steady-state solution.
- Effective removal of  plasma exhaust and nuclear heat
- Power density and distribution of removed power
- Effect of self conditioned PFC on plasma behavior
- Helical/Stellarator Resurgence
- Confinement, beta approaching tokamak
- Opportunities for configuration optimization
- Long Pulse Superconducting tokamaks - T-7, T-15, Tore Supra, TRIAM, EAST, KSTAR, SST-1, JT-60SA

## Realizing The Advanced Tokamak 

- Plasma cross-section shaping to enhance plasma current, power production
- 1968 Ohkawa (Plasma Current Multipole), 1973 T-9 Finger Ring,
- 1990s Spherical Tokamaks
- Bootstrap Current (self generated current)
- Predicted 1971 - Bickerton
- First observation 1983 in a mulitpole exp’t - Zarnstorff
- Observed in 1986 in tokamak -TFTR - Zarnstorff
- Beta limit physics “understood” for tokamak
-   = N (Ip/aB) where  = <p>/<B2>, 1983,  Troyon, Sykes
- NTM Stabilization by ECRH ASDEX Upgrade, DIII-D or RS
- Resistive Wall Stabilization DIII-D ~2005
- Confinement enhancement by stabilizing ITG using RS
- Reversed shear with a hollow current  profile provides the above:
- PEP modes on JET 1988
- ERS modes on TFTR 1994
- NCS modes on DIII-D 1994
- RS modes on JT-60U 1995 - record nT
- But all were transient

## Slide 18

## Slide 19

## Slide 20

## Four New Superconducting Tokamaks will Address Steady-State Advanced Tokamak Issues in Non-Burning Plasmas

- EAST:  R = 1.7m, 2MA, 2006

- KSTAR:  R = 1.8m, 2MA, 2008

- JT-60SA:  R = 3m, 5.5 MA, 2014

## Optimizing the 2-D Geometry of a Tokamak

**Optimizing the 2-D Geometry of a Tokamak**

-  Higher -limits at lower aspect ratio recognized in mid 1960s
- START achieved t ≈ 40% in 1991-96, NSTX 2004
- Very Low aspect ratio may allow a Cu TF coil engineering solution in a D-T environment
- What is the optimum aspect ratio for overall system performance?

- MAST

- NSTX

## The Stellarator/Helical  (3-D) Systems

-  The stellarator as first proposed by Spitzer May 1951 was a thermonuclear power generator based on a linear cylinder with uniform magnetic field.  A toroidal stellarator based on a Figure 8 was described later.
-  PPPL Model C - converted to tokamak in 1969, and the main stellarator effort was carried forward by IPP and Japan Univ’s/NIFS through the 70s and 80s.

- Figure 8 Stellarator (Model A ~ 1954)  and Spitzer (1993)

## Slide 24

## Sustained Hi  in Partially Optimized Stellarator W7-AS

- W7-AS

**Sustained Hi ******** in Partially Optimized Stellarator W7-AS**

- • W7-AS was the first stellarator device based on modular non-planar magnetic field coils
- • demonstrated commonality with tokamak physics like access to H-mode confinement regime

## Wendelstein 7-X

**An Optimized Stellarator is Under Construction**

**Major radius: ****	****5.5 m**
**Minor radius: ****	****0.53 m**
**Plasma volume****	****30 m****3**
**Induction on axis:  3T**
**Stored energy:****	****600 MJ**
**Machine mass:****	****725 t**
**Pulse length:****	****  30 min**
**Aux Heating ****	****20-40 MW****	**

- First Plasma 2014

- W-7X is based on W-7AS, and is optimized to reduce bootstrap plasma currents, fast particle loss,
- neoclassical transport, with good flux surfaces , MHD stability and feasible coils.

## Reactor Scale Magnet Technology

- •  Reliable large scale (≤1.6 GJ) Cu magnets at B~ 5T have been used in the tokamak operational environment for many years - many issues overcome
- •  A growing experience base in Superconducting Magnet Technology
- 		Magnetic Mirror SC coils in the early 70’s and early 80s
- 		First tokamak SC experiment T-7 1979
- 		Large Coil Project mid 1980s
- 		Large tokamak SC experiment T-15, Tore Supra  1988
- 		EAST, KSTAR, and (SST-1) advanced tokamaks ~2007
- 		ITER CS Coil Demo
- •  ITER will demonstrate reactor-scale SC magnets (≤43 GJ) at B ~5.3T
- 		additional work to be done but this area has made great progress
- • Significant benefits from continued development to higher B and/or higher T

## An International Team is Forged to Develop a New Energy Source

- • Agreed to “cooperation on fusion research” November 21, 1985  Geneva
- • The IAEA provides the framework for International Collaboration
- • By Dec 2005, EU,JA, RF, KO,CN, IN and US had signed ITER agreement

- Gorbachev and Reagan

**An International Team is Forged to Develop a New Energy Source**

## ITER is Now Underway

- ITER Site Under Construction

- First Plasma planned for  2018
- First DT operation planned for ~2022

- Reactor scale

## Slide 30

## Inertial Confinement Fusion, Early Days

- Radiation compression of DT to produce fusion energy demonstrated in the early 50s in Greenhouse George Cylinder test (and others).
- Invention of the laser in early 60s offered the possibility of a programmable repetitive driver for micro targets. Research continued on intense particle beam drivers in USSR and US.
- Idealized calculations in late 60s suggested 1kJ needed to achieve breakeven using micro targets and direct drive.
- 1972- Nature article by Nuckolls et al with computer modeling of laser driven compression   *Nature* Vol. 239, 1972, pp. 129
- Laser driven experiments at LLNL and elsewhere from mid 70s to mid 80s (Nova), revealed importance of plasma instabilities and driver uniformity,  raising required driver energy to MJ range.

## Construction of NIF/LMJ - ICF Burning Plasmas

- Classified Centurion-Halite nuclear tests in ~1986 reported to have validated compression modeling
- Many aspects of US ICF declassified in Nov 1994, allowing target designs to be discussed.
- Omega project reports gain of ~ 1% using direct drive of a DT capsule in 1996.
- Fast Ignitor concept (1995) offers possibility of reduced driver energies
- There has been dramatic progress in driver intensity and pellet fabrication in the past 40 years, and many challenges remain.
- Multiple paths in drivers (Glass, KrF, Z-pinch) are being pursued.

## NIF Enabled by Rapid Advance in Laser Technology

- • Glass laser energy has increased 106
- • Fusion energy will need:
- 	increased efficiency
- 	increased repetition rate

**NIF Enabled by Rapid Advance in Laser Technology**

## Target Designs with Varying Degrees of Risk Provide Adequate Gain for all Driver Concepts

- FI Expt’s  -    Omega, FIREX, HIPER

**Target Designs with Varying Degrees of Risk Provide Adequate Gain for all Driver Concepts**

- Tabak Snowmass

## Ignition Campaign - starting 2010

**Ignition Campaign - starting 2010**

## Some Overall Highlights

- A strong scientific basis has been established for fusion.
- Diagnostics and Plasma Technology  (Aux heating, CD, pellet inj) enabled progress.
- Several promising paths to fusion, each working on optimization and sustainment.
- Temperatures needed for fusion achieved  - in many facilities.
- Confinement needed for fusion is being approached - one step away.
- Complex fusion systems have been operated at large scale.
- Fusion systems using fusion fuel (DT) operated safely.
- Fusion could move much faster if required resources were applied.
- Now on the threshold of energy producing plasmas in both magnetic and inertial fusion.

## Facilities to Produce Fusion Energy are under Construction

**ITER**

**NIF**

**Facilities to Produce Fusion Energy are under Construction**

- First D-T  	~2022
- Fusion Gain, Q 	  10
- Fusion Energy/pulse	200,000 MJ

- First D-T  	~2010
- Fusion Gain, Q 	10 - 20
- Fusion Energy/pulse	40 MJ

## Fusion energy will begin powering the world.

***The Highlight for the next 50 years. ***

## Acknowledgements

- E. Frieman, I. Bernstein, K. Fowler, J. Sheffield, R. Stambaugh, M. Kikuchi, O. Motojima, D.Campbell, M. Watkins, F. Wagner, J. Callen, J. Willis, S. Dean, S. Milora,  R. Goldston,  E. Marmar