Paradigm Change in Fusion Science and University Research Centers in Korea

Hyeon K. Park

Pohang University of

Science and Technology

(POSTECH)

at

2008 FISFES workshop

Plasma and Space Science Center,

NCKU, Tinan, Taiwan

11/6-11/8, 2008

Talk

Progress in fusion plasma researchBrief background of fusion plasma research and

International Thermonuclear Experimental Reactor (ITER) project

Paradigm change of the fusion research: steady state capable fusion plasma devices in Asia

Quest for experimentally validated and verified theoretical modeling “Microwave Video Camera”

University based research centers for KSTAR Manpower and accelerated research preparation for

KSTAR.

Magnetic fusion devices Hot plasma is confined by an intricate magnetic field

Tokamak – external magnetic field and magnetic field by a driven plasma current

Stellarator – magnetic field by complex external coils

TEXTOR (Torus-Experim ent for Technology O riented R esearch)

B v

B t

B p

m ajor radius: 1.75 mm inor radius: 0.50 mplasm a current:

0.5 (0.8) M Atoro idal fie ld : 2 .8 Tpulse length: 10 sec

Tokamak configuration Stellarator configuration

Large Helical Device (LHD), NIFS, Japan

External diameter 13.5 mPlasma major radius 3.9 mPlasma minor radius 0.6 mPlasma volume 30 m3

Magnetic field 3 T

Julich, Germany

Plasma pressure and magnetic field profiles Plasma pressure – ions and electrons

Temperature ~ 20 keV (optimum D-T cross-section)

Density ~ 1 x1020/cm3

Energy confinement time ~ 1 second

Poloidal cross-section of the Tokamak plasma

Helical magnetic structure

Rotational transforme.g. toroidal turn (n=5)/poloidal turn (m=1)

Stability of hot plasma in Tokamak

BJp

Stability – balance between magnetic fields, current and plasma pressure

Unstable MHD behavior that can lead to the disruption and/or loss of confinement can not be predicted (e.g. thermal and current quenching)

Harmful MHD activities are rising as the plasma (~<p>/B2) is increased

Need a fully understood physical model of these harmful MHD modes in order to find any remedy

Disruption (sudden termination) (JET)

Cross-field energy transport in Tokamak Transport – free energy due to the

plasma pressure gradient through micro-turbulence

Physical mechanisms (ITG, TEM, ETG and MHD turbulence, etc.) responsible for the energy and particle transport – critical for Advanced Tokamak (AT) modes

Verification and validation of many theories by a decisive experiment are essential

• Fusion is not virtual science Turbulence based transport simulation results from Gyro code (GA)

t

trTtrT ei

eiei

),(

)),(( ,,,

Gyro

Demonstration of scientific breakeven Three large tokamak era: non-steady state device based on Cu

coils (pulse length is limited by the cooling system < ~ 20 sec.) Tokamak Fusion Test Reactor (USA) 1982-1997, Princeton Plasma Physics

Laboratory, USA Fusion power yield: Q ~ 0.3 from D-T experiment

Joint European Tokamak (EU):1983 – present, Culham, Oxfordshore, UK Fusion power yield: Q ~ 0.7 from D-T experiment

JT-60U (Japan):1985 - present, Japan Atomic Energy Agency (JAEA), Japan Q~1.25 extrapolated from D-D experiment

Internal view of JET/plasma dischargeInternal view of JT60-UInternal view of TFTR

Advancement of the fusion research Domination by advanced nations (through three large tokamak era)

Significant progress of theoretical understanding based on advancement of “computational power” in last two decades – Needs “verification and validation”

Established “empirical physics basis” and “engineering requirement” for the ITER – must be supported by “first principle based physics basis”

Scaling laws (engineering and physics) – Needs comprehensive physics of the external machine parameters

3-D MHD stability modeling (M3D)

3-D transport modeling based on micro-turbulence

(GS-2)

Why ITER depends on scaling law? The goal is "to demonstrate the scientific and technological

feasibility of fusion power for peaceful purposes". Demonstration of fusion power yield; Q (output power/input power) ~10 International consortium (Europe, Japan, USA, Russia, Korea, China, and

India) Design is based on “empirical scaling law”

Can we design the next generation without scaling law?

New fusion research facilities in Asia Steady state capable devices are critical for the physics

and engineering basis for the fusion plasma research New superconducting tokamak devices are merging to Asian

countries – Japan (LHD, JT-60SA), China (EAST), Korea (KSTAR) and India (SST)

LHD, NIFS, Japan

SST-1, IndiaEAST, Hefei,China

KSTAR,NFRI, Korea

JT-60SA,JAEA, Japan

Diagnostics of fusion plasmas

Extremely hostile environment- invasive method can not survive due to extreme conditions (~100 MoK) Passive – (a) spectroscopy of plasma emission (ranging from ~1 mm

to hard X-ray) (b) energetic particle analysis – neutrals and neutrons (thermal level to ~MeV range)

Active – sophisticated techniques for local plasma parameter measurement such as Te, Ti, ne, ni, impurities, magnetic field, current density, etc.

Visualization: Tomography based on emissions and active system In general, emission at a fixed wavelength is a function of multiple

plasma parameters Viewing ports and number of chords for the emission and active

system are very limited and the plasma temporal behavior is extremely fast: quality of the imaging in medical field (MRI and CT) in plasma study is a challenging task

Emission that depends on a single plasma parameter (i.e., electron cyclotron emission (Te)) is the most promising

Example: Visualization of MHD Physics

?

Analogous to evolution of diagnostic capabilities from Stethoscope to MRI

Precise predictive capability of MHD physics (Sawtooth, NTM, and RWM)

Characteristic MM wave frequencies in plasma diagnostics

Second harmonic of ECE (brown) is suitable for Te measurement Frequency is ranging from 95 ~130

GHz ECH frequency at 110 GHz can be

blocked by notch filter

Scattering Experiment Frequency range is 60 GHz ~ 1THz

Microwave Imaging Reflectometry Frequency is ranging from 60 – 80

GHz

Taylor et al., Rev. Sci. Instrum. 56, 928 (1985)

TEXTOR plasma

Microwave video camera concept

ECE measurement is an established tool for electron temperature measurement in high temperature plasmas

Sensitive 1-D array detector, imaging optics, and wide-band mm wave antenna, and IF electronics are required for 2-D imaging system

Te fluctuation measurement Real time fluctuations can be studied up to ~1% level Fluctuation studies down to 0.1 % level have been performed using long time integration

Conventional 1-D ECE system 2-D ECE imaging system

Sudden break up of a stable magnetic surface in a time scale much shorter than energy transport time

Sawtooth oscillation is a magnetic self-organization via magnetic reconnection process

Classical MHD instability (sawtooth oscillation)

Reconnection process via visualization

Comparison with the full reconnection model Remarkable resemblance with the images from the simulation result of the full

reconnection model (Sykes et al. single fluid MHD model) Magnetic topology change (reconnection) occurs as the island is formed based on “Y-

point” reconnection” (slow process) In experiment, no clear heat flow until a sharp temperature point is developed Reconnection occurs through a pressure finger and the initial stage forms “X-point”

(fast reconnection) Critical physics is missing in this model Quasi-interchange mode model is likely a wrong model

Comparison with the ballooning mode model Low field side

Similarity: “Pressure finger” of the simulation at low field side (middle figure) is similar to those from 2-D images (“a sharp temperature point”)

Difference: Heat flow is highly collective in experiment while stochastic process of the heat diffusion is clear in simulation.

Simulation results from Nishimura et.al.Plasma condition (p ~0.4 and t ~2 %) is similar to the experimental results

High field side Reconnection at high

field side is forbidden in Ballooning mode model

Reflectometry in high temperature plasmas

Incoming wave is reflected at the cut-off layer (rc) similar to ionospheric sounding

Reflected waves contain information of the shape of the cut-off layer

Fluctuating phase of the reflected signal is

drr

rk

cr

oo0

)(

)(~~

where ko is probe beam wave-number

)(~)()( rrr o is plasma permittivity

(rc)

Microwave imaging system for density fluctuation measurement

MIR system is capable of measuring poloidal wavenumbers simultaneously (e.g., agreement between the wheel spacing and measurement is excellent in the laboratory tests)

Extensive laboratory tests were completed

2 θ

20

k σ1

k 2D

Poloidal rotation of turbulence induced by NBI Starts at electron diamagnetic

direction at speed of +21 km/sec and settles at -12 km/sec during OH phase.

Becomes chaotic during beam slowing down time scale

How universal is the zonal flow?

U1U2

Further test of MIR system for robust operation MIR system has been applied to the plasma measurement

Curvature matching condition from plasma cut-off layer is not as sharp as expected from infinite conductivity assumption of modeling

Correlation length based on phase information is not consistent with that based on amplitude of reflected waves (inherent conventional reflectometry problem)

MIR system is back to POSTECH to understand the issues that we learned from plasma application Fundamental difference between plasma cut-off and perfect reflector:

dielectric multi-layer reflector Vs. metal surface. Doppler reflectometry shares the same fundamental problem of the

conventional reflectometry.

Extensive laboratory tests will be conducted with simulation study 1.5D and/or 3D EM simulation (PPPL) will be compared with laboratory

test to clarify the outstanding issues.

Multi-frequency Illumination for 2-D turbulence

A simultaneous “comb” of illumination frequencies can probe multiple cutoff layers, as each distinct frequency reflects from a distinct cutoff layer

Measurement of multi-layer turbulence flow such as “zonal flow” in the core of tokamak plasma

Ultimate Goal – Korean Fusion ReactorITER Korea – License for Demo Korean Fusion Reactor

KSTAR as Premier Fusion Research DeviceKSTAR – Manpower/Critical Physics/Technolgy

(NFRI, National Institutes and University based Programs)

핵융합에너지개발진흥 기본계획 (2007.8) 핵융합에너지개발 진흥법 (2007.3)

Korean Fusion Energy Plan

• KSTAR is the world best SC tokamak device• Excellent engineering team and strong support by Government and Society (ITER program)• Lack of research experience and manpower• History of lack of long-term commitment

Present Status

• New innovative diagnostic tool for physics research on KSTAR• Preparation of High beta SS operation; harmful MHD control, verification of transport physics, effective current drive via profile control

Future Direction

Status of Korean Fusion Science Research

EAST, KSTAR

2010 2015 2020 2025 2030

ITERJT-60SA

DIII-D,ASU,JET

2010 2015 2020 2025 2030

Cu device~10 sec.ITER~1000 sec.

JT60-SA~100 sec.

EAST~1000 sec.

KSTAR~300 sec.

(Time; years)

Perfo

rman

ce (lo

g)

(AU

)

• KSTAR is a Premier SS SC Tokamak Device• Promote KSTAR as the best Research Device prior to ITER and JT-60SA

(Q~10)

(Q~1)

Intermediate planLong term plan

Co-existing plan w/ ITER/SA

Phys./Eng. goalwith Medium

Phys./Eng. goalwith high SS

Direction of KSTAR Research Plan

Plan for International Collaboration

KO/US/JPmanagement

(Manpower and Ancillary systems)

(10~15 yrs)

KSTAR TeamInfrastructureImprovement

University center (core physics, diagnostics)

University center(fusion engineering

center)

University center (edge physics

divertor)

KAERI(Heating/CD)

KSTAR operation (10~15 year initial contract): KO/US/JP management team can play a key role in developing “steady state tokamak physics” of US and JPUniversity center: work with KSTAR management to support manpower and

specific programs indicated in the belowRemote Collaboratory: fully deployed for two phase operation of KSTAR

(Day shift: JP/KO and Evening shift: US/KO )

Remote Collaboratory

University Fusion Research Centers in Korea

(SNU) Fusion Engineering Center

Core physics centerCore diagnostic centerControl research centerHeating research centerCurrent drive research center

Simulation centerEdge physics centerDivertor research centerPlasma facing material center Edge plasma research center

Manpower and Capability through focused research program at University Centers

KSTAR ITERTokamak Plasma Physics Study

ViaVisualization tools

Macro Stability Physics (MHD)

Energy Transport (Turbulence)

Optimum CD for SS operation(LHCD & ECH) via

(profile control & coupling)

Verification ofcritical tokamak

physics via new tools

Contribution toITER and world Fusion program

LHCD heat load

LHCD antenna ECH launcher

ne fluctuationTe fluctuation LHCD & ECH Launcher Design

POSTECH Research Center