Mission and design requirements on national centralized tokamak nct
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IEA/LT Workshop (W59) combined with DOE/JAERI Technical Planning of Tokamak Experiments (FP1-2) 'Shape and Aspect Ratio Optimization for High Beta Steady-State Tokamak’ 14-15 Feb. 2005 at San Diego, GA. JT-60. Mission and Design Requirements on National Centralized Tokamak (NCT).

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Mission and Design Requirements on National Centralized Tokamak (NCT)

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Mission and design requirements on national centralized tokamak nct

IEA/LT Workshop (W59) combined with DOE/JAERI Technical Planning of Tokamak Experiments (FP1-2) 'Shape and Aspect Ratio Optimization for High Beta Steady-State Tokamak’

14-15 Feb. 2005 at San Diego, GA

JT-60

Mission and Design Requirements on National Centralized Tokamak (NCT)

Y.Miura and the National Centralized Tokamak Facility Design Team

1)Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Mukoyama, Naka, Ibaraki, 311-0193 Japan

  • * the National Centralized Tokamak Facility Design Team

  • M. Akiba1), H. Azechi2), T. Fujita1), K. Hamamatsu1), H. Hashizume3), N. Hayashi1), H. Horiike2),

  • N. Hosogane1), M. Ichimura4), K. Ida5), T. Imai4),S. Ishida1), Y. Kamada1), H. Kawashima1), M. Kikuchi1),

    A. Kimura6), K. Kizu1), H. Kubo1), Y. Kudo1), K. Kurihara1), G. Kurita1), M. Kuriyama1), K. Masaki1),

    M. Matsukawa1), M. Matsuoka7), Y. M. Miura2), N. Miya1), A. Morioka1), K. Nakamura8),

    H. Ninomiya1), A. Nishimura5), K. Okano9), K. Okuno1), A. Sagara5), M. Sakamoto8), S. Sakurai1), K. Sato8),

    R. Shimada10), A. Shimizu8), T. Suzuki1), H. Tamai1), H. Takahashi1), Y. Takase11), M. Takechi1), S. Tanaka11),

    K. Tsuchiya1), H. Tsutsui10), Y. Uesugi12), and N. Yoshida8)

  • 2)Osaka Univ., 3)Tohoku Univ., 4)Univ. of Tsukuba, 5)National Institute for Fusion Science, 6)Kyoto Univ.,

  • 7)Mie Univ., 8)Kyushu Univ., 9)Central Research Institute of Electric Power Industry,

  • 10)Tokyo Institute of Technology, 11)Univ of Tokyo, 12)Kanazawa Univ.


New minimum step to fusion power

New Minimum Step to Fusion Power

NCT

Previous Strategy

Proto

DEMO

Large Tokamaks

⑤Economical feasibility

Commercialization

③SS plasma

④Power production

JT-60(JA)

JET(EU)

TFTR(US)

①Self-sust. burn

②Long burn

ITER

To establish vision for commercialization

for a period of 2030-2050, it

  • should operate in steady state

    (for example, continuously for 1year)

  • should achieve high beta

    (N=3.5(SSTR)-5.5(CREST))

  • should operate reliably (less than 1

    off-normal event in 2 years) at least towards the end of operating period.

Recent strategy

ITER

Commercialization

DEMO

Large Tokamaks

Demonstration of SS operation

High b SS

JT-60

TFTR

→NCT

JET


Mission and design requirements on national centralized tokamak nct

NCT is a domestic research program for advanced tokamak research to succeed JT-60U incorporating Japanese universities accomplishments

NCT

Collaborating Universities or Institutes in FY2004.

Stratified Structure of Fusion Research

Realization of Fusion Energy

Developmental

ITER

Tokamak

IFMIF

Helical Laser

(Developmental)

(Academic)

Fusion Science

Reactor engineering

Plasma Science

Academic Research Basis

Academic


Mission of national centralized tokamak

Mission of National Centralized Tokamak

NCT

ferritic steel

  • Establish high  steady state operation for DEMO and Contribute to ITER

    • Demonstrate high  (N=3.5-5.5) non-inductive operation for more than 100 s in collision-less regime

    • Test compatibility of reduced activation ferritic steel

    • Demonstration of ultra-long (~8 hours) steady state operation


Requirements of nct machine capability

Requirements of NCT machine capability

NCT

Cryostat

TFC

PFC

R F

Present equipment

in JT-60 (reuse)

P-NBI

P-NBI

15 m

N-NBI

VV

reuse

Sector Coil

Stabiliser Plate

Divertor

13.5 m

  • A super-conducting device with break-even-classplasma performance

  • Capability of steady state high- (N=3.5-5.5) plasma with full non-inductive current drive,

    required for the DEMO for

    more than 100 s

  • Flexibility in terms of

    plasma aspect ratio,

    plasma shaping control,

    and feedback control

best use of existing JT-60 infrastructure


Heating and current drive systems for nct

Heating and Current Drive systems for NCT

NCT

1. P-NB(85keV)/ Co-injection : 4 units Current Profile control

2. P-NB(85keV)/ Counter : ???Rotation control

3. P-NB(85keV)/ Perpendicular: 8 unitsHeating Profile control

4. N-NB(350-400keV)Co-inj. : 2 unitsCurrent Profile control

5. ECW : 110GHz, 4 unitsNTM suppression

P-NB: 85 keV

Tang.: 4 units

Perp.: 8 units

ECW: 110GHz

4 units

possible upgrade

N-NB:

350-400 keV

Tang.: 2 units


Higher plasma shape for a high n

Higher Plasma Shape for a High-N

NCT

JT-60ASDEX-UJETDIII-D

S=2.3-7.4

S=3.1-3.6

S=3.0-5.4

S=2.0-2.2

6

5

4

3

2

Research target of NCT

Normalized bN

DIII-D experiment

JT-60

ITER

Ip

q95

S

234567

aBT

Shape parameter S

~ A-1{1+k2(1+2d2)}

  • Extension of the flexibility in the plasma shape is key issue for a high-bN plasma operation where the research target of NCT is addressed.

    Observed in DIII-D experiment [M.R. Wade et al., PoP 8 (2001) 2208]

  • In order to improve a shape parameter, low aspect ratio as well as high elongation and high triangularity is considered in NCT design.

Presented by T.S. Taylor at DOE/JAERI Technical Planning of Tokamak Experiments and Large Tokamak Workshop in Naka at 7-8 Feb. 2001


Two options of nct nct 1 nct 2

Two Options of NCT (NCT-1 & NCT-2)

NCT

NCT-1

NCT-2

400 turn

4m

432 turn

NbTi

NbTi

NbTi

Nb3Sn

NbTi

294 turn

144 turn

168 turn

NbTi

144 turn

NbTi

Ip=4.0MA

Bt=3.43T

A=2.85

k95=1.57

Ip=5.5MA

Bt=2.76T

A=2.6

k95=1.84

144 turn

6m

6m

NbTi

NbTi

96 turn

144 turn

Nb3Sn

NbTi

NbTi

418 turn

160 turn

Nb3Sn

208 turn

Nb3Sn

-4m

540 turn

400 turn


Typical plasma parameters for 2 options with sn dn

Typical Plasma Parameters for 2 Options with SN & DN

NCT

2

1

3

4

NCT-1

NCT-2

1 m

1 m

1 m

1 m


Divertor pumping is important for long pulse operation design criteria for divertor geometry

Divertor Pumping is important for long pulse operation- Design Criteria for Divertor Geometry-

NCT

Cryopanel

Pump S

Outside divertor

Inside divertor

Leg l

For divertor performance

S≥ 100m3/s, l≥ 0.4m

S. Sakurai et al., Plasma Phys. Cont. Fusion 44 (2002) 749.

Criteria of k is estimated for given A, d to ensure the divertor pumping and leg length.

NCT-2

NCT-1

high-k, d, and low-A makes divertor narrow

dx=0.55


Optimization of shape parameter by a and

Optimization of Shape Parameter by A and, 

5

Stabilizer for

NCT-2 DN

NCT

4

3

2

Z (m)

1

0

-1

-2

-3

-4

1

2

3

4

5

R (m)

-5

3.5

Aspect Ratio A

3.0

Divertor Pump

2.0

4

5

6

7

8

2

3

Shape Parameter S

Matsukawa’s presentation

Criteria : Divertor pumping speed ≥ 100m3/s -> X-point height limit

Leg length limit

  • - With the trade-off of divertor pumping, S-parameter goes up to 8 at A ~ 2.6 in NCT-2 design.

  • With the divertor pumping of ≥100 m3/s, S-parameter ~ 7 is expected.

  • - Restrict of  causes the decrease of S in low A.

Flexibility in plasma shape and aspect ratio is extended in consistent with the sufficient divertor performance.


Critical n for mhd stability on plasma aspect ratio

Critical N for MHD stability on Plasma Aspect Ratio

NCT

9

Double null

k95 = 1.8

Negative shear

qmin= 2.4

Parabolic P(r)

A=2.5(n=1)

A=2.5 (n=2)

A=3.0 (n=1)

8

A=3.0 (n=2)

bN

7

Critical

6

5

4

3

1.2

1.4

1.6

1.8

2.0

2.2

2.4

2.6

r

/a

w

NCT design

Kurita’s presentation

  • Dependence of critical bN on plasma aspect ratio for n=1 and n=2 mode is estimated as a function of the ratio of rW/a (ERATO-J code).

  • Critical bN increases in lower aspect ratio, which suggests the advantage of low aspect ration on N.

  • optimization of pressure profile is being studied.

  • Critical N dependence on shape factor S will be presented by Kurita’s talk.


High n together with large q dt equ

High N together with large QDTequ.

NCT

  • NCT should have a potential to investigate high N at large QDTequ

  • conditions: q95~3.5, PNB=25MW, HHy2=1.5, fGW=0.5-1

  • In the case of N~5.5, QDTequ <0.2

  • In the case of QDTequ~1 with fGW ~0.6-0.8,

    N ~2.6-3(NCT-1), N ~2.9~3.3(NCT-2)


Parameters of and for two options

Parameters of * and * for two options

NCT

  • In the regime of

    N=2.5-5.5, *<0.01

    for fGW=0.5-1.0.

  • Conditions:

    • q95~3.5,

    • PNB=25 MW,

    • HH(y,2)=1.5,

    • fGW=0.5-1.0

ITER

ITER


High n with full non inductive scenario

High N with full non-inductive scenario

NCT

  • 15MW with HHy2=1.5-1.6(25MW with HHy2=1.2-1.3), It is possible to have 1.5-1.7MA plasma with high N and full CD conditions

  • For q(0)>1, it is necessary to increase q95 (because of the central CD).

A=2.6,1.7MA, 1.5T, q95=8.8,

bN=3.6, HHy2=1.64, fGW=0.63

PNB

A=3.3,1.5MA, 1.8T, q95=5.1,

bN=3.7, HHy2=1.51, fGW=0.50

NNB

PNB

PNB

jTOTAL

fBS=0.68

jTOTAL

jBD

jBD

fBS=0.60

jBS

NNB

jBS

PNB


For a high performance off axis n nb is a candidate

For a high performance, off axis N-NB is a candidate

NCT

P-NB

N-NB

  • By a modification of N-NB beam line, it is possible to increase performance of high-N with full CD.

  • The modification increases the capability of the

    current profile controllability.

  • A=2.6, Ip=3.0MA, Bt=2.1T,

    q95=6.1, qmin=2.0, N=4.0,

    HHy2=1.99, fGW=0.50, N-NB

    (3MW, 400keV),

    P-NB (22MW, 85keV)

NCT-2

jTOTAL

jBD

jBS

fBS=0.69


Controllability of eccd for ntm suppression

Controllability of ECCD for NTM suppression

NCT

1

0.5

0

Controllability of ECCD is estimated by ray-tracing and Rutherford equation to deduce the required power.

normal shear with qo=1

fECE= 110 GHz

For m/n=3/2 mode in low-A, slightly high power is required due to the bload resonance width.

Required power is available for 30 s

r

6420

Drive current density (MA/m2)


Heat and particle control

Heat and Particle Control

NCT

・FW + SOL widthlimit

・a and IP is increasing

・a=0.7m : constant

・Plasma moves to inside -> higher IP -> higher BT -> higher density

First wall + SOL width

VV

Stabilizer

TFC

Divertor

Region

・Constant IP

Density window for NCT-1 & 2 ~ same

  • high pumping capability is set on NCT

  • To reduce divertor heat load and to keep plasma clean, the higher density is better

Radiation Power

Zeff

Zeff, Radiation Power

required radiation power

~imput power-10MW

Density


Assessment of two options

Assessment of Two options

NCT

There is a strong probability that NCT-2 option will be National Centralized Tokamak


Day long operation

Day long operation

NCT

• Plasma-wall-interaction with a long time scale (~8 hours)

• Avoidance of disruption against the external perturbation


Example of day long operation

Example of Day-long operation

NCT

  • Demonstration of controllability for ultra-long time scale

Example simulation for day-long operation

available both in NCT-1 and NCT-2

BT=1.3T

IP=1MA

Rp=2.85m

ap=0.85m

q95=5

k95=1.76

d95=0.46

Vp=70.7 m3

bN=3.66

W=2.36 MJ

tE=0.236 sec

HHy2=1.6

Zeff=1.5 (O2)

PNBabs=7.73 MW

neav=3.36e19m-3

fGW=76%

fBS=56%

QDTeq=0.143

Pfusion=1.1 MW

Full CD


Summary

Summary

NCT

  • Design of NCT is in progress to establish high  steady state operation for DEMO and to contribute to ITER

  • The shape and aspect ratio are important parameters for NCT.

  • Recent evaluation for the designs, there is a strong probability that NCT-2 option which has A≥2.6, ≤2, S≤7 and BT≤3T will be selected as National Centralized Tokamak.


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