k-demo status and progress - nucleus · 6. assembly tooling total value( ) : 18.45 ko allocation :...

Keeman Kima, Kihak Ima, Hong-Tack Kima, Sungjin Kwona, Hyun Wook Kima, Hyeon Jeong

Leea, G.H. Neilsonb, C. E. Kesselb, T. Brownb, P. Titusb, and D. Mikkelsenb

aNational Fusion Research Institute, Daejeon, 34133, the Republic of KoreabPrinceton Plasma Physics Laboratory, Princeton, NJ 08543, U.S.A

5th IAEA DEMO WorkshopMay 7th, 2018, Daejeon, Korea

K-DEMO Status and Progress

5th IAEA DEMO Workshop May 7-10, 2018, Daejeon, Korea

Fusion Energy Development

Roadmap in Korea

2


Alternative choice of clean energy for Korea

➢ Renewable energy in Korea ? • Not enough space for solar panel and wind mill• Not enough sunlight and wind

➢ Fission prospect ?• Safety and efficiency has been improved• Remained issues related with long term waste

➢ Fusion prospect ..• Close to realize (ITER and DEMO)• Need further research to be economic


Korea Fusion Energy Development Plan (revised in 2017)

ision

Phase

Policy Goal

Basic Directions

Basic Promotion

Plan

PrimaryStrategy for

Plan 3

Policy Goal for Plan 3

Establishment of foundation for fusion engineering technology development for demonstrating electricity production

❖ Acceleration for development of DEMO core technology❖ Strengthening of basic research in fusion and manpower fostering system❖ Broadening the base of support for fusion energy development

Phase 1 (’07~’11) Phase 2 (’12~’26) Phase 3 (’27~’41)

Establishment of a foundationfor fusion energy development

Development of Core Technology for DEMO

Construction of DEMO by acquiring construction capability

of fusion power plants

▪ Acquisition of operating technology for the KSTAR

▪ Participation in the international joint construction of ITER

▪ Establishment of a system for the development of fusion reactor engineering technology

▪High-performance plasma operation in KSTAR for preparations for the ITER Operation

▪ Completion of ITER and acquisition of core technology

▪Development of core technology for the design of DEMO

▪ DEMO design, construction, and demonstration of electricity production

▪ Undertaking of a key role in ITER operations

▪ Completion of reactor core and system design of the fusion power reactor

▪ Commercialization of fusion technology

Basic Promotion Plan 1 (’07~‘11)

Basic promotion

plan 5(‘27~‘31)

Basic promotion

plan 6 (‘32~‘36)

Basic promotion

plan 7(‘37~‘41)

Basic promotion

plan 3 (‘17~‘21)

Basic promotion

plan 4(‘22~‘26)

Basic promotion

plan 2(‘12~‘16)

Secure sustainable new energy source by technological development and the commercialization of fusion energy


Strategy of the Korea fusion R&D programs :

Paradigm shift from short pulse to steady-state operation

PLT

ALCATOR C

PDX DIII

TFTR

DIII-D

JET/TFTR

JET

TFTR

JT-60U

ALCATOR A

1970 1975 1980 1985 1990 1995 2000 2005 2010

1KW

1MW

1GW

1W

SNUT-79

2015 2020 2040

JET

T-3

(1968)

1965

ATC

KAIST-T

KT-1

KSTAR

ITERConventional Device (Cu)

Superconducting Device

Superconducting Device

Fusi

on P

ow

er

Year

DEMO

FAST FOLLOWER FIRST MOVER

Fusion Energy Devel.

Promotion Law

3rd 5-yr National

Plan started


Role of KSTAR and ITER

KSTAR : Physics Machine ITER : Fusion Engineering

· High Performance (βN > 4) Plasma

Research for Fusion Power Plant

· Fusion Physics Validation

· Tokamak Simulator Development

· Design Requirement for Fusion Plant

· Stable (βN ~ 2) Burning Plasma &

Fusion Nuclear Science Research

(14 MeV Neutron Effect)

· Confirmation of Engineering

Feasibility for Fusion Power Plant

Construction & Operation of

Artificial SUN(=Fusion Power Plant)

[βN > 4 required]


KSTAR :Exploring steady-state high performance

operation

7


KSTAR program

Construction (1995 ~ 2007) First plasma (2008) First H-mode (2010) First ELM suppression (2011) Long-pulse H-mode (> 70s) (2016) Long ELM suppression (>34s) (2017) on going NBI upgrade


Operation goal of KSTAR exploring steady-state

and high performance plasma operation

Parameters DesignedAchieved

(~2017)

Major radius, R0

Minor radius, a

Elongation,

Triangularity,

Plasma shape

Plasma current, IP

Toroidal field, B0

H-mode duration

N

Superconductor

Heating /CD

PFC

1.8 m

0.5 m

2.0

0.8

DN, SN

2.0 MA

3.5 T

300 s

5.0

Nb3Sn, NbTi

~ 28 MW

C, W

1.8 m

0.5 m

2.15

0.8

DN, SN

1.0 MA

3.5 T

73 s

4.3

Nb3Sn, NbTi

~ 10 MW

C, W

(2016)

➢ Key parameters of KSTAR

▪ Mission : To explore the physics and technologies of high performance and steady-state operation that are essential scientific and technological basis for ITER and fusion reactor


Unique features of KSTAR advancing the high

performance operation R&D

▶ Highly performance SC magnets• Lowest error field (𝛿B/B0~1x10-5)• Lowest toroidal ripple (~0.05 %)

0

5

10

0 1 2 3 4

dB/B0 vs bN in KSTAR (measured/linearly projected)

Typical Intrinsic EF level in Ohmicplasmas in other tokamaks

bN

dB/B0

(x10-5)

bN, no-wall ~ 2.6 (nominal)

▶ ITER relevant In-vessel coils• Uniquely top/middle/bottom coils• Reliable ELM-crash suppression (>30s)

▶ Advanced diagnostic systems • Profile and 2D imaging diagnostics• Physics validation of MHD & confinement

▶ Long-pulse heating & CD systems • Long pulse high beta op. using NBI (>70s)• 2nd NBI system is under construction

2500 ms (w/ ITB)4500 ms (w/o ITB H-mode)6500 ms (w/o ITB L-mode)

#16498

q

psi_n

ECEI Image bolometer MSE


Research directions of KSTAR toward K-DEMO

➢ Explore the long pulse high beta operation with ELM-free H-mode

• ELM-crash control with IVCC coils in case H-mode is the only choice

• Establish physics based ELM-crash and routine recipe for suppression

➢ Explore new operating regime with less instabilities

• Expand operation window to higher beta operation with intelligent control of MHDs (sawtooth, NTM, disruption, RWM)

• Hybrid, ITB combined with low edge q, etc.

➢ Validation of MHD and turbulence models for predictive capabilities

➢ Test of innovative engineering and technologies

• Innovative heat removal

• Innovative current drive – explore new scheme with improved efficiency


Reproducible long pulse, NBI-heated H-mode

operation (>70sec)

▶ long pulse H-mode discharge • H-mode discharge : > 70 s (0.45 MA)

• Nearly non-inductive discharge

▶ Alternative operation mode

development• Internal transport barrier (ITB) > 13s

• High beta operation (N > 3.0, P > 3.0) : > 3s

0.45 MA, 70s

0

1

2

3

4

5

6

7

1.8 1.9 2.0 2.1 2.2 2.3

0

100

200

300

400

Ti (ke

V)

ITB @2.5 s

H-mode @4.5 s

Vt (km

/s)

R (m)

H-modepedestal

#16498

ITB foot(ρ=0.46)

0.0 0.2 0.4 0.6 0.8 1.0

0

1

2

3

4

5

6

7

8

1 2 3 4 5 6 7 8 9

0.5

1.0

1.5

2.0

2.5

3.0

Saf

ety

fact

or, q

r

ITB @2.5 s

H-mode @4.5 s

Time (s)

q 0

ITB foot(q=2.1)

ITB mode


KSTAR could assess a very robust way for ELM-

crash suppression (~ 34s) using IVCC

34 sec wall

Record breaking long ELM-crash suppression (~34s)

➢ KSTAR has unique 3 row RMP coils like as ITER (top/middle/bottom)

➢ Record long ELM-crash suppression (~34s) with n=1 RMP achieved in 2017

• ITER comparable operation (q95 ~ 3.4, βN ~ 1.8)

• Validation of predictive model of ELM-suppression with experiments


Long-pulse H-mode research•Long pulse H-mode (>70s)•ELM research & control (>30s)•Alternative operation modes (ITB, low q, ..)

Advanced scenario & MHD research• Stable high beta operation

(N >3.0, Tion ~ 10 keV)• Advanced mode develop.

(hybrid, ITB, low q)• MHD & disruption control

Steady-state & reactor mode research• Tungsten divertor & active cooling• Advanced current drive under test(HFS LHCD & Helicon CD)• Steady-state operation (~300s)

2008

First plasma

(ECH 84 GHz)

Long-pulse H-mode

(NBI~5.5 MW)

(ECH~1 MW)

2017

Heating upgrade(NBI~12 MW)

(ECH~6 MW)

2021

2021

Divertor upgrade(Tungsten divertor)

(Detached divertor)

(Diagnostics)

Advanced current

drive(LHCD~4 MW)

(Helicon CD~4 MW)

2025 ~

2017

Future research and upgrade plan for higher beta and steady-state operation toward K-DEMO


Heating & CD upgrade for higher beta and

Divertor upgrade for steady-state operation

NBI-2 (upgrade)

NBI-1

NBI-2

NBI-1

• Heating & CD upgrade (2018~)• 2nd NBI (6MW, off-axis) / ECCD (6 MW, total) • Innovative HFS LHCD, Helicon CD

• Divertor upgrade (2021~)• Tungsten-based metal• Fully active cooling• Diagnostics for Divertor physics

250 m

m

A4 size

RW SWA


ITER :Construction and operation of reactor-scale

tokamak

16


In-kind contribution of Korea

1. TF Conductor (Completed)

Total Value (kIUA) : 215.01

KO Allocation : 20.2%

KO Contribution (kIUA) : 43.39

2. Vacuum Vessel Main Body

Total Value(kIUA) : 118.61



3. Vacuum Vessel Port




4. Thermal Shield


KO Allocation : 100%

KO Contribution(kIUA) : 26.88

6. Assembly Tooling



KO Contribution(kIUA): 18.45

7. Tritium SDS




8. AC/DC Converters




11. Test Blanket Module*

KO Contribution :

HCCR TBS (TBM System)

kIUA Value : N/A

9. IVC Bus bars




5. Blanket Shield Block




10. Diagnostics




Total Value : 263.58 kIUA* TBMA (TBM Agreement) was signed in 2014.


Fabrication of reactor-scale vacuum vessel and

structures satisfying reactor-grade qualification

• Vacuum vessel : the first Sector 6 is about 76% (end of November).

• Assembly of inner

shells for Lower Port

Stub Extension (LPSE)

has been completed

• Thermal Shield

(VVTS) is about 55%

as scheduled

• Manufacture of Vacuum Vessel Sectors by KO Hyundai Heavy Industries Corporation is making good progress and will ensure to achieve the First Plasma in December 2025;

Fabrication of vacuum vessel and port


Experience of accurate assembling of reactor-

scale heavy structures

• ITER is challenging program requires high accuracy in assembly and alignment of

massive structures

• Elements for the massive 800-ton Sector Sub-Assembly Tools are being delivered to ITER.

• Installation in ITER assembly hall is on-going


K-DEMO Concept Study :R&D for advanced fusion reactor

20


Strategy for designing the K-DEMO

➢ Strategy for K-DEMO :

• Natural Path: KSTAR ITER DEMO

• Two phase operation to mitigate risks in the course of DEMO development

➢ The operation Stage I (PFusion = 2,200 MW)

• At least one port will be designated for the CTF including blanket test facility.

• Demonstrate the self-sufficient Tritium cycle (TBR > 1.05)

➢ The operation Stage II (PFusion = 3,000 MW)

• Major upgrade of In-Vessel-Components

• Demonstrate the net electricity generation (> 400 MWe)

• Demonstrate the competitiveness in COE.

■ Main Parameters

• R = 6.8 m / a = 2.1 m

• B0 = 7.0~7.4 T / B-peak = 16 T

• elongation = 1.8

• triangularity = 0.625

• Plasma current > 12 MA

• Te > 20 keV

■ Other Feature

• Double-null & Single-null configuration

• Vertical Maintenance

• Total H&CD Power = 80~120 MW

• P-fusion = 2200~3000 MW

• P-net > 400 MWe at Stage II

• Number of Coils : 16 TF, 8 CS, 12 PF


➢ Example of the developed steady profiles (Pfusion ~2.1 GW scenario)

0 0.2 0.4 0.6 0.8 10

10

20

30

40

50

0 0.2 0.4 0.6 0.8 10

5

10

15

Density (1019m-3)

Te

ne

ni

Ti

Temperature (keV)

Radius, ρ Radius, ρ0 0.2 0.4 0.6 0.8 1

0

0.5

1

1.5Safety factor, q

0 0.2 0.4 0.6 0.8 10

2

4

6

8

Bootstrap

NB

Total

Jφ (MA/m2)

Parameters Value

PF (MW) 2080

Q (Fusion

Gain)18.8

PNB (MW)110

(500keV)

IP (MA) 15.5

fBS 77 %

βN 2.8

H98y2 1.2

Tped (keV) 8.3

nped (/m3) 9.9x1019

Simulation of K-DEMO operation

Operation scenario and H&CD optimization

• NBCD depending deposition(~1 MeV, NUBEAM code)

• FWCD (TORIC code)


Magnet System

■ 2D drawing and 3D modeling of

the K-DEMO magnet systems

■ 16 TF coils, PF coils – 6 upper,

6 lower. 8 CS modules

23


Magnet and Structure

24

Dual winding packs with 2 types of CICC

→ High magnetic field

with huge Cost savings

(High Jc Strands & No Radial Plate)

■ TF Winding


Cross-Section of TF Coil

Selected for Detailed Study (Maintenance Space = 2.5 m)

Considering Vertical Maintenance Scheme▪ R = 6.8 m, a = 2.1 m

▪ Small CICC Coil : 18 x 10 turns Large CICC Coil : 12 x 5 turns (Total : 240 turns)

▪ Magnetic Field at Plasma Center : ~7.4 Tesla (Bpeak ~ 16 Tesla, T-margin > 1 K)

▪ Nominal Current : 65.52 kA

▪ Conductor Length : • LQP = ~900 m (Quadruple Pancake) (Total : ~450 ton)

• SDP = ~930 m (Double Pancake) (Total : ~280 ton)

3220 mm

2050 mm

3135 mm

2525 mm 2925 mm

ClearanceFilled with Glass Fiber(5 mm)

Ground Wrap(5 mm)

180 mm

1018 mm

13510 mm

14528 mm

550

mm

13595 mm13795 mm

13948 mm14348 mm

Space (Filled with 316) for Turn Transition and Feed Through (143 mm) 1

90 m

m

106 m

m


Ground Wrap(5 mm)


Cross-Section of CS Coils

Number of Turns : 12 (Total SC strand weight : ~190 tons)

Number of Layers : CS1, CS2, CS3 & CS4 : 36 layers

Magnetic Field at Center : ~13.3 Tesla (Bpeak ~ 13.5 Tesla @ 43 kA)

Conductor Unit Length : 800 m (CS1, CS2, CS3 & CS4 : UL x 6)

Gap Between Coils : 50 mm

Nominal Current : 43 kA

Flux Swing (half) : ~94 V·sec

CS1, CS2, CS3 & CS4 Coil


Cross-Section of PF Coils

Number of Turns : 8 turns for PF1~4, 12 for PF5, and 2 for PF6

Number of Layers : 20 layers for PF1~4, 36 for PF5 and 4 for PF6

Conductor Unit Length : 620, 755, 890 and 1030 m for PF1~4

980 & 1010 m for PF5 and 770 m for PF6

Coil Center Position : (2980, 8310), (3660, 8310), (4340, 8590) – PF1~3

(5020, 8750), (12762 & 13158, 7500), (14880, 2950) – PF4~6

PF1~4

PF5

PF6

PF1~4

PF5

PF6

Gap : 150 mm


CICC Parameter

Parameter TF HF TF LF CS PF1-4 PF5-6

■ Cable pattern

No. of SC strandsNo. of Copper strandsSpiral Dimension (mm)

(3SC)x4x5x6x5 + 2 Helical Spirals

1800-

ID 5 / OD 8

(((2SC+2Cu)x5)x6+7Cu)x6 + Central Spiral

360432

ID 7 / OD 9

(2SC+1Cu)x3x4x4x6+ Central Spiral

576288

ID 8 / OD 10

(2SC+1Cu)x3x4x4x5+Central Spiral

480240

ID 7 / OD 9

■ Void Fraction (%) 30.9 30.54 38.3 36.7

■ Strand Type

High Jc (> 2600 A/mm2) Nb3Sn Strand 0.82 mm diameter

(~450 ton + ~280 ton)

ITER type (Jc ~ 1000 A/mm2) Nb3Sn Strand0.82 mm diameter

(~190 ton + ~90 ton)

NbTi Strand0.82 mm diameter(~90 ton)

■ Cu/non-Cu of Strand 1.0

■ Insulation 1.6 mm (including Voltage Tap)(0.1 mm Kapton 400% + 0.3 mm S glass 400%)

2.0 mm (including Voltage Tap)(0.1 mm Kapton 400% + 0.4 mm S glass 400%)

■ Jacket Thickness 5.0 mm 4.5 mm 5.0 mm 4.0 mm

■ Twist Pitch (mm)1st stage2nd stage3rd stage4th stage5th stage

80 ± 5140 ± 10190 ± 10245 ± 15415 ± 20

80 ± 5140 ± 10190 ± 10300 ± 15

-

27 ± 545 ± 1085 ± 10150 ± 15420 ± 20

35 ± 575 ± 10135 ± 10285 ± 15410 ± 20

■ Wrapping TapeSub-cable wrap thickness

Sub-cable wrap widthCable wrap thicknessFinal wrap width

0.08 mm, 40% coverage15 mm

0.4 mm, 50% coverage7 mm


CICC Dimensions and Trial Fabrication

TF LF CICC TF HF CICC CS CICCPF CICC


Inter-coil Joint Scheme of Magnet

ITER CS Inter-coil Joint Scheme used• Joint Resistance ~0.2 n-ohm/joint


Structural Analysis of TF Case

TF coil case stresses contoured to 900 MPa maximum !!!

2050 mm

3135 mm

2525 mm 2925 mm


Ground Wrap(5 mm)

~ 1 mm

Elastic deformation occurs from the top-right corner of TF inboard side and the maximum stress at the top-left corner can be reduced.

A detail analysis required to make the stress almost uniform ( averaged stress)


Blanket and Neutronic Analysis

RAFM coolant channel

Mixture of Breeder and Multiplier

(Li4SiO4 + Be12Ti)

Coolant inlet/outlet

(Press. H2O)

Purge gas line

(He)

First Wall

(W)

RAFM back plate

Design description of blanket:

▪ Water cooled solid breeder type

▪ FW heat flux: 0.5 MW/m2 (peak)

▪ Neutron wall load: ~2.5 MW/m2 (avg.) / ~3.9 MW/m2 (peak)

▪ Breeder and multipliers: Mixture of Li4SiO4 and Be12Ti ceramic pebbles

▪ Structural material: RAFM

▪ First wall material: Tungsten

▪ Coolant: PWR water condition - 15 MPa, 290~330 °C

Concept of blanket module (Layered structure)



Nuclear heating (MCNP v6.1) and thermo-hydraulic analysis (ANSYS/CFX v18)

to develop the layered structure

Temperature distribution in the layers of the

blanket module

Radial nuclear heating distribution

of blanket module

~700 °C limit



TBR production rate in IB/OB

midplane blanket modules

Tritium breeding ratio: Global TBR ~1.03 (with the pressurized light water)

Breeding regions

TBRs in poloidal sections



Radiation shielding and displacement damage analyses

Nuclear heating in 16 TF coils (~9

KW)

20 (RAFM)

4 (Tungsten FW)

The maximum displacement damage(dpa/fpy) in

the blanket outboard module

Max. DPA/FPY in blanket structure


Inboard

Target

Outboard

Target

Central

Dome

Outlet

Inlet

Cassette Body

Concept of Divertor

36

■ Overall configuration of the K-DEMO divetor system

➢ Details of High Heat Flux Unit

Tungsten mono-block

Vanadium interlayer

RAFM cooling tube• Coolant – PWR-like

(inlet 290 oC,

T = ~ 40 oC)

➢ The K-DEMO divertor system is divided into 32 toroidal modules with the symmetric upper and lower.

5th IAEA DEMO Workshop May 7-10, 2018, Daejeon, Korea37

Concept of Divertor

■ Thermo-mechanical analyses for the outboard divertor target: on-going

Design Parameters HHF unit with RAFM HHF unit with CuCrZr

Max. Temp. of monoblock 1240 ℃ 1232 ℃

Max. Temp. of tube 550 ℃ 297 ℃

Width of monoblock 21 mm 28 mm

Height of monoblock 26.5 mm 28 mm

Top thickness 4.5 mm 8 mm

Lateral thickness 3.0 mm 5.5 mm

Thickness of tube 0.4 mm 1.5 mm

Thickness of interlayer 0.2 mm 1 mm

Water Condition @ inlet 15 MPa, 290 ℃ 4 MPa, 70 ℃

< Thermo-mechanical Analysis for a Monoblock w/ Support>

▪ Thermal load

- 10 MW/m2

W

RAFM Tube

Monoblock

Support

Backplate

MaterialCalculated

[ MPa]

3Sm

[ MPa]

W 696 1800

RAFM 497 500

< Thermo-hydraulic Analysis for a high heat flux unit >

▪ Mechanical loads:

- Water Presessure:15 MPa

- EM load:10 kN


DPA Analysis on K-DEMO Divertor

■ Estimation of displacement damage on divertor target (dpa/fpy)

RED : Tungsten, 0.3 ~ 2.6 dpa/fpy

Black : Cu, 0.9 ~ 10.9 dpa/fpy


Vertical Maintenance Concept

■ Vertical maintenance scheme of in-vessel components

➢ For maintenance, in-vessel components are toroidally subdivided into:

▪ Blanket: - 16 inboard sectors

- 32 outboard sectors

▪ Divertor: 32 modules (upper & lower each)

➢ Sectors / modules are removed through the enlarged top vertical ports of

vacuum vessel.

39


DEMO Core Technology

Development Plan

40


K-DEMO Core Technology Development

Design basis

tech.

Material basis

tech.

Machine &

system eng.

tech.

Tokamak core

plasma tech.

Reactor

system

integration

tech. Safety &

licensing tech.

H&CD &

diagnostics

tech.

Heat retrieval

system tech.

Fusion

materials tech.

SC magnet

tech.

Fusion materials

development center

Blanket

test

facility

Extreme

scale

simulation

center

PMI test

facility

Fusion neutron

irradiation test

facility SC conductor

test facility

● 3 Major Research Fields, 7 Core Technologies, 18 Detail Technologies

and 6 Major Research Facilities were identified to develop K-DEMO.


PMI Test Facility (Chonbuk Province)

MAGNUM-PSI (Cf.)

● 2.4 MW High-Temperature Plasma Test Facility

- Plasma-Material Interaction(PMI) Test facility - Upgrade Plasma Facility for PMI Test- Additional, Blanket Test Facility


Superconducting Conductor Test (SUCCEX)

Facility

• Background field : 16 Tesla

• Split-pair Solenoid Magnet System

• Inner-bore Size : ~ 1 m

• Two Test Modes :

✓ Sultan-like sample test mode

✓ Semi-circle type conductor sample test mode

(Cf.) SULTAN

• Background field : 11 Tesla

• 100 kA SC Transformer for

the short sample test

SULTAN-type Semi-circle-type

■ SUCCEX (SUperConducting Conductor EXperiment)


Energy (MeV) 20 100

Peak Current (mA) 0.1 ~ 20 0.1 ~ 20

Max. Duty (%) 24* 8

Max. Ave. Current (mA) 4.8 1.6

Pulse Width (ms) 0.05 ~ 2 0.05 ~ 1.33

Max. Repetition Rate (Hz) 120 60

Max. Beam Power (kW) 96 160

Emittance (mm-mrad) 0.22(x),0.25(y) 0.3 / 0.3

20 & 100 MeV KOMAC Proton

Linac

50 keV Injector

3 MeV RFQ

20 MeV DTL100 MeV DTL

20 MeV Beamlines100 MeV

Beamlines

SRF TB

RI Semiconductor

Life-Medical App.

Materials

Basic Science

RI

Medical App.

Neutron Source

Basic ScienceAerospace App.

Nuclear Materials

Features of 100 MeV linac

50 keV Injector (Ion source + LEBT)

3 MeV RFQ (4-vane type)

20 & 100 MeV DTL

RF Frequency: 350 MHz

Beam Extractions @ 20 or 100 MeV

5 Beamlines for 20 MeV & 100 MeV

Neutron

Irradiation

Test Lab.


Fusion Reactor Materials R&D

Extreme Environment

Material R&D Hub


Summary

▪ Korea fusion R&D programs are supported under “Korea Fusion Energy Promotion Law”.

▪ KSTAR made outstanding progress in achieving steady-state high performance plasma and achieving ELM-free H-mode

▪ KSTAR will explore advanced higher performance operation to contribute K-DEMO by heating & CD

▪ The contribution to ITER construction and operation could convince the burning plasma operation with reactor-scale device.

▪ Design of K-DEMO accompanied by simulation is on-going to get reasonable values of TBR, heat flux, and performance.

▪ Separate R&D program for K-DEMO technology development is TBD.


Thank you for your attention !

Stronger efforts are necessary to ensure the fusion energy for the next generation..

k-demo status and progress - nucleus · 6. assembly tooling total value( ) : 18.45 ko allocation :...

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