Progress of CFETR Design
Ge Zhuang1, Guoqiang Li2, J. Li2, Y.X. Wan1,2, Y. Liu3, X.L.
Wang4, Y.T. Song2, V. Chan1, Q.W. Yang3, B.N. Wan2, X.R.
Duan3, and CFETR design team
1University of Science and Technology of China, Hefei, China
2Institute of Plasma Physics, CAS, Hefei, China
3Southwest Institute of Physics, Chengdu, China
4Chinese Academy of Physics Engineering, Mianyang, China

27th IAEA Fusion Energy Conference, Ahmedabad, India

22-27 October 2018

China MCF Roadmap

1 GW, Power
Plant Validation
IIDEMO validation, Q>10, CW, 1 GW, >50 dpa
IQ=1-5, steady state, TBR>1, >200 MW, <10 dpa
Phase I IQ=5, 3000 s, 350 MW, steady-state burning plasma
Phase I Q=10, 400 s, 500 MW, Hybrid burning plasma
Advance PFC, steady-state advanced operation
Advanced divertor, high power H&CD, diagnostics
Disruption mitigation, basic plasma
2015
Oct. 23 2018

2020

2025

2030

2035

2040

27th IAEA FEC, Ahmedabad

2045

2050

2055

2060
1

CFETR Missionl Fusion power production of Pf = 200~1500 MW
l Generates steady-state burning plasmas (duty time ~ 50%)l Tests the self-sustainable burning (Q ≧ 25~30, Hα ~ 83-86%)
l Realizes Tritium self-breading (TBR ≧ 1)l R&D for structural and functional materials which have highneutron flux resistive
Buildup the science and technology base for PFPP

Oct. 23 2018

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Key Issues of CFETR Mission

Obtained burning Plasma
for fusion power

Steady-state operation
for fusion energy

1. P = 200-1500 MW
2. Q = 1-10, SSO, hours
3. Q = 20-30 hours-SSO
4. High energetic α heating

5. SSO (Ext H&CD + Higher fb)
6. Hybrid (OH+BS+CD)
7. PSI on the first wall
8. Heat & particle exhaust on Div.

T-breeding by blanket
10. T-plant: extract & reprocessing
11. Materials & components
12. Reliable and quick RH
13. Licensing & safety
9.

Breeding tritium
for T self-sustained

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Course of CFETR Design Eventsl Concept design (2011-2017)- First period (2011-2015)
R = 5.7 m, a = 1.6 m, BT = 4–5 T, Pf = 200 MW- Second period (2015-2017)
R = 6.6 m, a = 1.8 m, BT = 5–7 T, Pf = 1 GW

l Integrated engineering design (2017-, 30 M$)- New version
R = 7.2 m, a = 2.2 m, BT = 6.5 T, Pf = 200 MW - 1 GW

Oct. 23 2018l

Small scale R&D continues (70 M$)lLarge scale R&D will start soon (500 M$)
27th IAEA FEC, Ahmedabad4Outlinel Introduction- New version of CFETR designl CFETR Physics Design- Development of operation scenarios- Consideration of divertor conf. & impurity effects- Investigation of MHD stabilityl CFETR Engineering Design- Magnet system- Vacuum system- Remote handling and maintenance system- Others…

l SummaryOct. 23 2018

27th IAEA FEC, Ahmedabad5Outlinel Introduction- New version of CFETR designl CFETR Physics Design- Development of operation scenarios- Consideration of divertor conf. & impurity effects- Investigation of MHD stabilityl CFETR Engineering Design- Magnet system- Vacuum system- Remote handling and maintenance system- Others…

l SummaryOct. 23 2018

27th IAEA FEC, Ahmedabad

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New version of CFETR design
Key parameters
Major Radius R0

7.2 m

Minor Radius a

2.2 m

Elongation

2

Toroidal B Field BT

6.5 T

Plasma Current Ip

14 MA

Divertor Conf.

Lower Single Null

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Proposed CFETR Research Planl Some new features of CFETR design- Higher BT, Lower Ip, Advanced CS (≥ 480 VS), 16 TF coils for easy RH à
More reliable plasma targets and higher confidencel CFETR operation plan (Staged approach)- H/He: 1-2 years- DD:
1-2 years- DT: < 100 MW1 years
200 MW, SSO, T fuel cycle, 5 years
500 MW, SSO, TBR > 1, 3 years- DT: DEMO validation, 1 GW, 5 years
Advance Scenario, > 1.5 GW , Q~30, 2-3 years- Total: ~ 20 years

Oct. 23 2018

27th IAEA FEC, Ahmedabad8Outlinel Introduction- New version of CFETR designl CFETR Physics Design- Development of operation scenarios- Consideration of divertor conf. & impurity effects- Investigation of MHD stabilityl CFETR Engineering Design- Magnet system- Vacuum system- Remote handling and maintenance system- Others…

l SummaryOct. 23 2018

27th IAEA FEC, Ahmedabad

9Primary Tasks of CFETR Physics Designl CFETR physics design mainly focuses on development andoptimization of the operating scenarios with respect to physics
and engineering constraintsl Operating scenarios will
- Predict the fusion performance- Explore and determine a robust operation space possessing good
confinement, MHD stability and acceptable transport level- Evaluate and limit the fraction of helium and other impurity particles
while approaching the desirable fusion performance- Size up the power and particle exhaust compatibility with the chosendivertor configuration- Assess and manipulate the transit and steady heat load to the firstwall and divertor to keep the machine safety-…
Oct. 23 2018

27th IAEA FEC, Ahmedabad

10Integrated Modelling Strategy
l 0D system code used to scope out parameter space- Provides 0th order engineering parameters e.g. R, a, IP, BT- Consistent set of H98, !N, fBS, etc. for target Qfus, Pfus
- Ballpark estimates of Paux
- Does not identify actual operating scenariol Integrated Modeling (IM) used for scenario development- Physics-based models, beyond experimental scaling laws
- Reproducing experimentally demonstrated scenarios- Ensuring consistency of core, pedestal and boundaryl IM informs key engineering design requirements- H&CD, Divertor heat and particle fluxes, fueling- Plasma control and disruption mitigationl IM critical to CFETR diagnostics design and operation- Provides best-guess, hard to measure profile information

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Code Suites for Multi-physics Modelingl Core-pedestal coupling for
scenario design- A workflow was developed
under the framework OMFITl SOL and divertor SOLPS,OEDGE/DIVIMP, …
l MHD stability NIMROD, MARS-F,
AEGIS, GATO …
l Energetic particle NOVA-K, M3D,
ORBIT …
l Pedestal Ana. & Opt. ELITE,
BOUT++ …
l Plasma shape design TEQ, EFITl Discharge simulation TSC,TOKSYS
Oct. 23 2018l Evolving particle densities, Te and Ti,
and momentum- ne/He/impurity profiles evolved and D&T
obey quasi-neutrality

l SOL solutions match core parameters
at pivot point ~ top of pedestal- Heat and particle fluxes, iterate boundary
densities and temperatures

27th IAEA FEC, Ahmedabad

12Fully Non-inductive Operation Scenario Designed with
System Code (0D)
fusion power
power to run plant
Pfusion/Paux
net electric power
Neutron Power at Blanket
normalized beta
bootstrap fraction
H factor over ELMY H_net
current drive powerplasma current
field on axis
Ion/electron Temperature
Electron Density
Ratio to Greenwald Limit
Zeff
Power per unit R
q95_IterOct. 23 2018

Parameters A.1
A.2
A.2
A.3 A.4
100MW 200MW 500MW 1GW DEMO-level
Pf
120
229
482
974 2192
Pinternal
199
196
223
238 265
3.06
5.87
Qplasma
1.56
11.89 28.17
Pnetelec
-107
-58
30
232 738
Pn/Awall
0.12
0.23
0.49
0.99 2.23
βN
1.00
1.20
1.50
2.0
3.0
fbs
0.40
0.40
0.40
0.50 0.75
HITER98Y2
1.12
1.25
1.32
1.41 1.42
Pcd
77
75
82
82
78
10.34
12.92
Ip
8.61
13.78 13.78
BT
6.5
6.5
6.5
6.5
6.5
Ti(0)/Te(0)
18
24
32
36
32
n(0)
0.48
0.52
0.61
0.78 1.31
nbar/nGR
0.57
0.51
0.48
0.57 0.96
Zeff
2.45
2.45
2.45
2.45 2.45
9.42
11.66
P/R
8.52
15.69 30.70
q95_iter
8.87
7.39
5.91
5.54 5.54
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1GW Non-inductive Operation Scenario by CorePedestal Coupling Simulationl Preliminary results- No self-consistent tritium fueling- Deviation (~30% for Paux) VS system codeli ~ 0.64
fGR,vol ~ 0.89
Oct. 23 2018l NBI à 500 keV (68 MW,CD) + 100 keV (10MW, rotation drive)l EC à maintain large radius RS and control
qmin > 2 to avoid low n deleterious MHD
modes
l Large BS current à RS and reduce CD
power requirementl Moderate q95
Simulation

Sys. CodePf (GW)

1.0

0.97

Q

9.1

11.9

PEC/PNB

31/78

82(tot)βN,th(βN,tot)

2.05(2.36)

2.0(~)HITER98Y2

1.11

1.41

fbs(%)

59

50

Ip (MA)

12

14

INB/IEC (MA)

4.0/0.9~27th IAEA FEC, Ahmedabad

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Hybrid Operation Scenario Designed with System Code (0D)
Parameters

fusion power
power to run plant
Pfusion/Paux
Neutron Power at Blanket
normalized beta
bootstrap fraction
H factor over ELMY H_net
Ohmic fraction
current drive powerplasma current
field on axis
Ion/electron Temperature
Electron Density
Ratio to Greenwald Limit
Zeff
Power per unit R
q95 IterOct. 23 2018

Pf
Pinternal
Qplasma
Pn/AwallβN
fbs
HITER98Y2
fohm
Pcd
Ip
BT
Ti(0)/Te(0)n(0)
nbar/nGR
Zeff
P/R
q95_iter

B.1
100MW
114
190
1.54
0.12
1.00
0.40
1.01
0.30
74
8.61
6.5
13
0.67
0.79
2.45
7.58
8.87

B.2
200MW
250
196
3.35
0.25
1.20
0.40
1.09
0.30
74
10.34
6.5
17
0.74
0.72
2.45
9.33
7.39

27th IAEA FEC, Ahmedabad

B.2
500MW
558
202
7.65
0.57
1.50
0.40
1.18
0.30
73
12.92
6.5
24
0.82
0.64
2.45
12.63
5.91

B.3
1GW
1128
222
15.30
1.15
2.00
0.50
1.19
0.30
74
13.78
6.5
24
1.16
0.85
2.45
19.11
5.54

B.4
DEMO-level2192
75795.16
2.23
3.0
0.75
1.54
0.24
3
13.78
6.5
34
1.23
0.90
2.45
22.97
5.54
15

1GW Hybrid Operation Scenario
by Core-Pedestal Coupling Simulationl Preliminary resultl NBI à 1 MeV (32 MW,CD) + 600 keV(11 MW, CD & rotation drive)l EC à maintain flat q profile and control
qmin > 1l Moderate li à plasma stability
l ~300 Volt-sec (8-hours in flattop)- No self-consistent tritium fueling- Deviation (~25% for Paux) VS systemli ~ 0.93
fGR,vol ~ 1.0
Oct. 23 2018

Simulation

Sys. code

Pf (GW)

0.92

1.1

Q

10

15

PEC/PFW/PNB

30/20/43

74(tot)βN,th(βN,tot)

2.09(2.3)

2.0(~)HITER98Y2

1.16

1.19

fbs(%)

49

50

Ip (MA)13

14

INB/IEC/IFW/IOH

1.4/0.8/1.6/3~/~/~/4

27th IAEA FEC, Ahmedabad16

H & CD Scheme Considerationl EC: necessary tool for current profile control- Optional freq/power: 190 GHz ~ 250 GHz, 20 ~ 40 MW- HFS top launched with high freq. for efficient off-axis ECCD
- LFS above midplane for flexible location of ECCD
- Optional application in NTM controll LH: efficient far off-axis current drive- Optional freq/power: 4.6 GHz or beyond, ~20 MW- HFS launched above midplane ( adapting for toroidal field along counter-clockwise
direction) for CD at r/a >= 0.7

l HHFW: efficient off-axis or near-axis current drive- Optional freq/power: 0.8 ~ 2 GHz, ~20 MW- Optional launched positions: HFS, LFS- High CD efficiency at r/a < 0.6

l NB: broad current drive and possible significant rotation drive- Option for CD: 600 keV ~ 1 MeV NNBI, 16 ~ 32 MW, (1 ~ 2 beam)- Option for rotation drive: 100 keV PNBI, 10 MW, (1 beam)
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Considerations of Plasma Shape and Divertorl ITER-like plasma configurationκsep=2.0, %& = 0.39, %' = 0.45l A divertor coil (DC1) is addedfor possible advanced divertor
configuration (Snowflake+)l Plasma-facing MaterialsW

Conventionall Different Divertor
configurations–Conventional–Small Angle Slot（SAS）–Snowflake+κsep=2.0%& = 0.39%' = 0.45
dRsep= 6.0 cm

l Optimization target–Ppeak ≤ 10 MW/m2

SAS–Te ≤ 5-10 eV–ne-sep ≤ 5×1019 m-3–Zeff-ped ≤ 3

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Ar Injection Can Effectively Reduce the Divertor Heat Loadto below 10 MW/m2

Neutral density (m-3)

D

Oct. 23 2018

Ar radiation (W/m-3)

D2

27th IAEA FEC, Ahmedabadl Simulation performed withSOLPS codel The peak heat fluxes on bothinner and outer divertor are
below 10 MW/m2
l Total radiation is higher than
80%, mainly by Ar impuritiesl Detachment occurs at the
strike points, but still too
high Te in far SOL regionl Fueling dilution and fusion
performance degradation incore region should be
carefully concerned for highradiation scenarios
19Blanket has strong stabilization effect on vertical instability
Real blanket structure

Modeled blanket structure

Back support
structure
ManifoldFirst wall

Vacuum vessell Simulations are performed with TSC
and TOKSYS codesl Blanket modules (BM) are modeledwith three-layer structures. Resistivity
is evaluated and scannedl Calculations show BM could
significantly reduce the growth rate
Passive stru
includedVV

Growth rateof VD (/s)

Out ofcontrol

VV + BM()

VV + BM(7.610-6
ohm*m)

2.2

18.1

l Internal coils are still necessary tocontrol the vertical instability. It is
under assessment
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Operation in Grassy ELM regime?

l Type-I ELM must be avoidedl Mostly likely, RMP coils will not be
installed
l According to experimental dataclassification, bp and n* from EPED1
for the reference scenario put it in thegrassy ELM regimel BOUT++ and other codes are being
used to verify the ELM prediction
Oyama N. 2008 J. Phys.: Conf. Ser.

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Resistive Wall Mode Should be Stable for the Steady-state
Scenariol MARS-K has been used to calculate the stability of RMWs, with uniform rotationl The steady-state scenario is marginally unstablel A small rotation of Ω0/ ΩA<0.01 could make the RMW stable

Growth rate of n = 1 RWM
Oct. 23 2018

27th IAEA FEC, Ahmedabad22α particle drive is weakly destabilizing for TAE and RSAE insteady state scenario
n=3 (AWEAC)

l Linear calculations show the α particle
drive weakly destabilizingl NOVA-K shows the damp effects could
make TAEs and RSAEs marginally stablel Effects of nonlinear AEs and EPMs areunder investigation

n=3 (NIMROD)

n=3 TAE (NOVA-K)

n=4 (AWEAC)

n=4 (NIMROD)

n=3 RSAE (NOVA-K)

See Y. Hou TH/P2-7
Oct. 23 2018

27th IAEA FEC, Ahmedabad23

Outlinel Introduction- New version of CFETR designl CFETR Physics Design- Development of operation scenarios- Consideration of divertor conf. & impurity effects- Investigation of MHD stabilityl CFETR Engineering Design- Magnet system- Vacuum system- Remote handling and maintenance system- Others…

l SummaryOct. 23 2018

27th IAEA FEC, Ahmedabad

24

CFETR Engineering Design
Measurement capability
aiming to do research,
control and safety, etc

INTEGRATEDDESIGN 
NBI, ICRF, ECRH,
LHCD, etc

Establish design
standardization,
manual, quality
management, etc.

H & CD

Magnet

CFETR

DesignStandardization

RemoteHandling

Maintenance and replacement
for internal components, NBI,
RF antennas, diagnostics, etc.
Oct. 23 2018

Strategy and procedure
for assembling, interface
compatibility andstandardization, etc.
TF, PF, CS, and theirsupports, feeders

VV

Internal
Components

Vacuum Vessel,
Neutron shielding,port, etcBlankets, interface between pipe andBlanket, Water cooled breeder blankets,
Helium cooled breeder blanket, divertor, etc

27th IAEA FEC, Ahmedabad

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Magnet System (Toroidal Field Coils)l Design completed
l EM & Stressanalysis done

Max Force @TF coil ~736 Mpa; Max. Deformation ~ 16.5 mm
6.5 T @ R = 7.2 m; 174 Turns; 84.6
kA/TurnOct. 23 2018

ITER TF

EU-DEMO[2015]

CFETR TF

No. of Coil

18

18

16

Current per Turn

68 kA

81.7 kA

84.6 kA

Total inductance

17.34 H

32.68 H

32.5 H

Total Storage Energy

40.1 GJ

109.08 GJ

116.34 GJ

Storage Energy per Coil2.227 GJ

6.06 GJ

7.27 GJ

27th IAEA FEC, AhmedabadMagnet System (Central Solenoid)l High temperature superconductor(Bi2212) + low temperature
superconductor (Nb3Sn) à a
maximum 19.9 T@ 51.25 kA/turn.
l Each module has 720 turns,
powered independently
l Maximum 400 VS flux with amaximum rate of field swing of~1.2 Ts.

Perspective view of the CFETR CS
Oct. 23 2018

27th IAEA FEC, AhmedabadVacuum system
Vacuum Vessel+Thermal Shield+Dewar

H ~15 mmm

W ~ 9.1 m Thickness ~50
8 Sectors, 4/6/8 ports

H ~ 15 m, W ~ 9.4 m, Thermal Shield~40mm

H ~ 29 m, D ~ 38 m

l

.

l

.

l

.
.

l

Background Pressure ~ 1x10-4 Pa

.
.

.
.

.

pressure drop @ cooling pipe
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Vacuum systeml Torus with D-shaped cross-section, 4 upper vertical ports, 8 lower ports and 6
equatorial ports- 4 upper ports à maintenance and disassembly of blanket.
- 6 lower ports à divertor maintenance and the cryo-pumps.
- 8 equatorial ports à NBI, diagnostic and some RH tools.

l Inner, outer shells and stiffening ribs joined by welding.
l Material of the VV is 316L(N)-IG.

3D model Vacuum
Vessel for CFETR

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Divertor Structurel Divertor targets divided
into two halves on each
module, totally 72
divertor modules, eachone ~11 tons. RH fromlower port

Conceptual engineering design of the CFETR divertor structure.

l Cooling water à outertarget à inner target àbafflesl Cassette cooled
separately àtargets/baffles RH
separately from Upper
ports by Multi-Purpose
Deployer
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Progress on the Blanket Designl Helium cooled ceramic breeder
blanket (HCCB) design completed
l Evaluate neutron energy deposition
and wall load @ Fusion power = 1 GW,
2 GW
l Start the water cooled ceramic
breeder (WCCB) design
Oct. 23 2018

27th IAEA FEC, Ahmedabad31

Remote Handling & Maintenancel Blanket RHM: Inboard &outboard blankets à fromupper ports by a corridor
with a crane à hot cell.
l Divertor RHM: circular
movement à lower
horizontal port àcask à
hot cell.
l MPD : equatorial port àmaintenance of small
pieces, inspection,
diagnosis.

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Remote Handling & Maintenance
Mechanical load on Manipulator jointsl Mechanical analysis on RHM of internal components;l Establish overall control architecture, hardware and software control integrationarchitecture @ Heterogeneous control architecture theory.
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Design and R&D for H&CD Systems
LHW

NBI
l Based on ITER NBI design, complete
preliminary design of N-NBI System, R&D of
key technologies of CFETR N-NBI
l Promote research of RF source, high RF power,
long pulse ion sourcel Achieve substantive results on isolationtransformer for RF power transmission

Oct. 23 2018l Complete preliminary design of low power
microwave power source driving circuit,control scheme of power and phase, andInvestigate high-power klystron, andauxiliary power equipmentl Carry out high-field coupling and highfield antenna simulation study, R & D of
key components of the transmission line

27th IAEA FEC, Ahmedabad

4.6GHz 500kW/CW Klystronmodel and structure34

Design and R&D for H&CD Systems
ECW

Helicon Wavel Complete the ECRH system design, R&D
of key technologies for Gyrotronl Start the effectiveness analysis andperformance evaluation of ECCD undervarious conditions (beam injection
position, antennas incident parameter,different gyrotron freqs 170GHz, 230GHz

l Start design and analysis antennaof travelling wave

Toroidal angle

Helicon waves traveling wave antenna module

Poloidal angle

Performance of 230GHz ECCDZa = 0 m)(a) ECCD kA/MW;bPeak location of
JCD;c2nd Harmonic absorp. ratio%)Traveling wave antenna modules arranged in theblankets, satisfying high power requirements

27th IAEA FEC, Ahmedabad35

Outlinel Introduction- New version of CFETR designl CFETR Physics Design- Development of operation scenarios- Consideration of divertor conf. & impurity effects- Investigation of MHD stabilityl CFETR Engineering Design- Magnet system- Vacuum system- Remote handling and maintenance system- Others…

l SummaryOct. 23 2018

27th IAEA FEC, Ahmedabad

36Summaryl New design with R = 7.2 m / a = 2.2 m & high BT.
l Detailed designs of physics and engineering are under the way.
• Progress of physics design- Fully non-inductive and hybrid mode scenarios with performance that meets
the CFETR mission have been developed- Broad operation range in !N and !p, stable with wall at r/a = 1.2- Helium dilution fHe cannot exceeds 0.2 to meet Pfus target- Radiation in the core acceptable up to Zeff ~ 3- Tungsten fraction at the edge can’t exceed 4e-5 to stay in H-mode• Progress of engineering design- Concept design of key systems completed, detailed engineering design of thesystems ongoing.

l CFETR will be fully open to our cooperators, your input in very
valuable for the success of the project.
Oct. 23 2018

27th IAEA FEC, Ahmedabad37

Acknowledgements

We are grateful to General Atomics, PPPL, LLNL,
Wisconsin, U. York, MPG-IPP and U. Toronto for the useof their physics code suites and data, and their helps

Oct. 23 2018

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Thank you for your attention !

Oct. 23 2018

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