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Engineering Design Evolution of the JT-60SA Project

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Engineering Design Evolution of the JT-60SA Project

P. Barabaschi, Y. Kamada, S. Ishida for the JT-60SA Integrated Project Team
EU: F4E-CEA-ENEA-CNR/RFX-KIT-CRPP-CIEMAT-SCKCEN
JA: JAEA

1

JT-60SA Objectives

•A combined project of the ITER Satellite Tokamak Program of JA-EU

(Broader Approach) and National Centralized Tokamak Program in Japan.

•Contribute to the early realization of fusion energy by its exploitation to support the exploitation of ITER and research towards DEMO.

ITER

DEMO

Complement ITER towards DEMO
Support ITER

JT-60SA

2

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The New Load Assembly

••

JT60-U: Copper Coils (1600 T), Ip=4MA, Vp=80m3 JT60-SA: SC Coils (400 T), Ip=5.5MA, Vp=135m3

JT-60U JT-60SA

JT-60SA(A≥2.5,Ip=5.5 MA)

ITER (A=3.1,15 MA)
3.0m

6.2m

~4m
~2.5m

1.8m

KSTAR (A=3.6, 2 MA)

1.7m

EAST (A=4.25,1 MA)

1.1m

3

SST-1 (A=5.5, 0.22 MA)

3

High Beta and Long Pulse

• JT-60SA is a fully superconducting tokamak capable of confining break-

max

even equivalent class high-temperature deuterium plasmas (Ip =5.5 MA) lasting for a duration (typically 100s) longer than the timescales characterizing the key plasma processes, such as current diffusion and particle recycling.
• JT-60SA should pursue full non-inductive steady-state operations with high βN (> no-wall ideal MHD stability limits).

4

4

Plasma Shaping

• JT-60SA will explore the plasma configuration optimization for ITER and DEMO with a wide range of the plasma shape including the shape of ITER, with the capability to produce both single and double null configurations.

5

5

  • Ip=5.5MA, Lower Single Null
  • Ip=5.5MA, Double Null
  • Ip=4.6 MA ITER like

Phased Research Plan

• Exploitation within the BA period will aim at the initial research phase:
- HH operation for plasma full commissioning - DD operation for identification of the issues in preparation for full DD operation

• Principles of “Joint Exploitation” later phases agreed between EU and JA • Detailed “Research Plan” jointly prepared

Annual Neutron
Limit
Expected Duration
Remote Handling
Max Power

  • Phase
  • Divertor
  • P-NB N-NB ECRF

1.5MW
Power x Time
1-2 y 2-3y 2-3y >2y

  • -
  • 10MW

x100s
23MW 33MW

phase I phase II phase I phase II

HDDDD

Initial
Research Phase
LSN partial monoblock
+
1.5MW x5s
NB: 20MW x 100s
30MW x 60s duty = 1/30
4E19 4E20 1E21
R&D
Perp.
13MW
10MW
Tang. 7MW
Integrated Research Phase
LSN full-
ECRF: 100s
37MW

41MW monoblock
7MW

Use
Extended

Research
Phase

  • >5y
  • 1.5E21
  • DN
  • 24MW
  • 41MW x 100s

6

Baseline Schedule

-Start of Tokamak Assembly:

Early 2012

-Completion of Tokamak Assembly: October 2015

-First Plasma: Mid 2016.

7

Device re-baselining

• In late 2007, cost concerns prompted the Parties to request a re-design of the device.
• Several ways to reduce costs were found in “old” design:

• Issues found:

ꢀ

Maximum field in TF conductor rather large for cost effective [email protected] design (Operation point : 6.5T at 5K, with 1K temperature margin). Conductor hot spot temp was unnecessarily low Structures rather heavy and complex to fabricate – multiple coil types Nuclear heating too conservatively estimated

ꢀꢀꢀꢀ

…

• Solution:

ꢀꢀꢀ

Reduce TF field, decrease SC strands volume (segregation), decrease Cu section Optimise structural design Recover plasma performance lost by field reduction by plasma volume increase made possible by reduction of magnet size Rationalise requirements and ports

ꢀ

• In parallel some other problems were addressed:

ꢀꢀꢀ

Seismic Loads to be managed by supports Peak TF Ripple in Plasma to be reduced to 0.5% Inductive flux for plasma operation to be increased to reach 100s flattop

8

Design at >6T costly for NbTi (when Tcs~6K)

T=6K

T=5K

9

Aspect Ratio Optimised

•

R/a is the real only free parameter in the design of a Tokamak for a given

nTτ

••

Trade-off between volume and power density R/a determines important engineering factors

ꢀꢀꢀꢀꢀ

Access to In-Vessel Components Wall power density Field and technology of magnets Mechanical loads Controllability of plasma shape

•

Optimum A is largely a function of technology, machine size, and scaling laws

  • The Toroidal Field, and the density
  • increase with Aspect Ratio

  • decrease with Aspect Ratio
  • The Plasma volume the Plasma Current, and

τ

E

10

Main Results of re-Design on TF Magnet

2007 Design

NEW
IDR

CDR 2007

2008
TF Total TF Energy [GJ] TF coils #

1.51

1.06
18
TF Tension [MN]

Number of turns in 1 coil

Peak Field [T]

6.80

90

4.70

72

6.43 30.0

5.65

  • 24.4
  • Total length of conductor [km]

Total Structure Mass [ton]

Strand type

~650

~280

NbTi
Conductor current [kA] Cable space dimensions [mm] N SC Strands

25.3

25.7

18x22

324

25.1x25.1

720
-

  • N Cu Strands
  • 162

Tmarg (after burn) [K]

~1.1

~1.4

SC Strand mass in windings of 18

New Design

  • 92
  • 34

coils [ton]

Segregated copper mass in 18 coils [ton]

-

14

11

Main Results of re-Design

Additional Design Changes: •Cryoplant: heat load re-assessed based on new operation scenarios with reduced neutron yield and reduction of magnet cold structure. From 16kW to ~10kW and much improved dynamics •Cryostat: Single Wall with simplified geometry •Vacuum Vessel: Better distribute VV ports to provide a simpler out of plane support to the TF magnet structure. Reduction of shells

•PF system:

•Balance the distribution of turns in the EF system, •6 Coils (Vs 7), •Broader betap-li range, •28% less conductor •4 Independent CS coils (Vs 3)
•TF Ripple Lowered: with Ferritic Insert will be <0.5%

•Inductive Flux Increased: to ensure 100s flattop

12

Machine Parameters

• After successful completion of the re-baselining in late 2008, the JT-60SA project was launched with the new design with the following parameters.

Cryostat

Basic machine parameters

P-NBI
P-NBI

Plasma Current Toroidal Field, Bt Major Radius, Rp Minor Radius, a Elongation, κX Triangularity, δX Aspect Ratio, A
Shape Parameter, S Safety Factor, q95 Flattop Duration
Heating & CD Power
N-NBI
5.5 MA 2.25 T
2.96

ECH

1.18 1.95 0.53

N-NBI

2.5 6.7 ~3
100 s
41 MW 10 MW 24 MW
7 MW
15 MW/m２
P-NBI ECRF
Divertor wall load

13

13

Heating and CD Systems

Variety of heating/current-drive/ momentum-input combinations

NB: 34MWx100s

Positive-ion-source NB 85keV 12units x 2MW=24MW

COx2u, 4MW CTRx2u, 4MW

Perpx8u, 16MW

Negative-ion-source NB 500keV, 10MW for off-axis NBCD

ECRF: 110GHz, 7MW x 100s

9 Gyrotrons, 4 Launchers with movable mirror

14

14

Magnet

•

All SC Magnet

ꢀ

Nb3Sn for CS

ꢀ

NbTi for TF, EF

•••

18 TF Coils 6 EF Coils 4 CS independent modules

15

Conductors

Coil

Type of strands

TF

NbTi

CS

Nb3Sn 20 ~9
Operating current (kA) Nominal peak field (T) Operating temperature (K) Number of SC strands/Cu strands Local void fraction (%) Cable dimensions (mm) Central hole (id x od) (mm) Conductor ext. dimensions (mm)
25.7 5.65 <5K 324 / 162 32
<5.1K 216 / 108 34

TFC conductor

18.0 x 22.0 -22.0 x 26.0
Φ21.0 7 x 9 27.9 x 27.9

Conductor type

No. of coils Type of strand

EF-H

EF3, 4

NbTi

20.0 6.2

EF-L

EF1, 2, 5, 6

←

CS conductor

Operating current (kA) Nominal peak field (T) Operating temperature (K) Number of SC/Cu strands Local void fraction (%) Cable dimensions (mm) Central hole (id x od) (mm) Conductor ex. dimensions (mm)

←

4.8

  • 4.8
  • 5.0

450 / 0 34
216 / 108 34
21.8 7 x 9 27.7
19.1

←

EF-L&H conductor

16

25.0

16

TF Magnet features

TF winding pack

Connections area

• Cable in Conduit Conductor , 72 turns, 25.7kA each

OIS

• 6 double pancakes, 6 turns/pancake. Helium inlets in high field side – joints in external low field side

EF1 hanging support arm

• Windings enclosed in Steel Casings • Steel casings supported to ground vertically and toroidally - connected in inboard curved regions by “Inner Intercoil Structure”

• Steel casings guided toroidally by “Outer Intercoil
Structure” to support out of plane loads.

EF support pedestals

Gravity support

Reinforcement ribs

17

View of Magnet - assembly

18

TFC Cold Tests

•••

All TFC will be cold tested at full current before shipping to Japan Test Facility will be in CEA-Saclay PA prepared with test facility specifications and cold testing specifications

ꢀ HV Tests ꢀ Leak Tests ꢀ Flow, Pressure Drop ꢀ Dimensional Stability during and after cooldown ꢀ Resistance, Joints ꢀ Tmargin, Quench tests (2 coils)

••

Cryostat and jigs being fabricated. Cryostat to be delivered in Saclay in first half 2011

19

Power Supplies

•

Largely a full new set of Power Supplies compatible with long pulse operation have been designed and will be procured. These will include:

•

TFC, grouped in 3 units of 6 coils

ꢀ 6 pulse unidirectional, 26kA, 80V ꢀ 3 Fast Discharge QPC Units (1.5KV to ground)

•

PFC, each coil with:

ꢀ 12 pulse, 20kA, ~1KV ꢀ Booster (old units) or new Switching Network Unit ꢀ Fast Discharge QPC Unit

•

In-vessel Coils

••

QPC Units: Under final tendering phase SCM-PS:

ꢀ Technical Specs completed, ꢀ Procurement Arrangement in final stage

20

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  • Design and Construction of the KSTAR Tokamak

    Design and Construction of the KSTAR Tokamak

    IAEA-CN-77/OV7/1 Design and Construction of the KSTAR Tokamak G. S. Lee and the KSTAR Team National Fusion R&D Center, Korea Basic Science Institute, Taejon, Republic of Korea e-mail contact of main author: [email protected] Abstract. The extensive design effort has been focused on two major aspects of the KSTAR project mission, steady-state operation capability and “advanced tokamak” physics. The steady-state aspect of mission is reflected in the choice of superconducting magnets, provision of actively cooled in-vessel components, and long- pulse current-drive and heating systems. The “advanced tokamak” aspect of the mission is incorporated in the design features associated with flexible plasma shaping, double-null divertor and passive stabilizers, internal control coils , and a comprehensive set of diagnostics. Substantial progress in engineering has been made on superconducting magnets, vacuum vessel, plasma facing components, and power supplies. The new KSTAR experimental facility with cryogenic system and de-ionized water-cooling and main power systems has been designed, and the construction work has been on-going for completion in year 2004. 1. Introduction The mission of the Korea Superconducting Tokamak Advanced Research (KSTAR) Project is to develop a steady-state-capable advanced superconducting tokamak to establish a scientific and technological basis for an attractive fusion reactor [1,2]. To support this project mission, three major research objectives have been established: (i) to extend present stability and performance boundaries of tokamak operation through active control of profiles and transport, (ii) to explore methods to achieve steady-state operation for tokamak fusion reactors using non-inductive current drive, and (iii) to integrate optimized plasma performance and continuous operation as a step toward an attractive tokamak fusion reactor.
  • Progress of the KSTAR Experiments and Perspective for ITER Scientific Researches

    Progress of the KSTAR Experiments and Perspective for ITER Scientific Researches

    Progress of the KSTAR experiments and perspective for ITER scientific researches Yeong-Kook Oh on be half of KSTAR team and collaborators National Fusion Research Institute, Daejeon, Korea Email: [email protected] KSTAR is a superconducting tokamak aiming to explore the long-pulse high beta confinement. In the 2012 experimental campaign, the duration of the H-mode flattop has been extended up to 16 s at 0.6 MA and plasma current level in H-mode reached at 0.9 MA by adopting the real-time plasma shape control and 3.5 MW neutral beam injection. The equilibrium operating space could be extended surpassing the n=l ideal no wall limit with betaN and betaN/li up to 2.9 and 4.1, respectively. The pedestal formation and characteristics were investigated according to L- and H-mode transition and during the edge localized mode (ELM). As one of the ITER high priority research topics, exploring the ELM mitigation or suppression by applying n=l or n=2 resonance magnetic perturbation (RMP) field or by injecting the supersonic molecular beam. The toroidal rotation changes were inspected for the ohmic and H-mode plasma by applying the ECH or 3D magnetic field. Various experimental researches were conducted according to the proposals including the disruption mitigation by using massive gas injection, fast ion loss detection under the edge perturbation, plasma wall interaction and others. 1 . Introduction The Korea Superconducting Tokamak Advanced Research (KSTAR) device has a mission to explore the scientific and engineering research under the high performance and steady-state plasma confinement, which are essential for ITER and K-DEMO [1].The key designed parameters of the KSTAR are compared recent achievement in the table 1 .The basic designed parameters are toroidal field (TF) of 3.5 T? plasma current of 2 MA, and pulse length up to 300 s.