design of helical reactor ffhr-d1 and c1 for steady state...

14, March 2014 @ UCSD

Design of Helical Reactor FFHR-d1 and c1for Steady State DEMOfor Steady State DEMO

Akio Sagara,H. Tamura, T. Tanaka, N. Yanagi, J. Miyazawa, T. Goto, R. Sakamoto,

J. Yagi, T. Watanabe, S. Takayama, and the FFHR design groupIn Fusion Engineering Research Project (FERP)

National Institute for Fusion Science,Japan

US-Japan Workshop on Fusion Power Plants and Related Advanced Technologies13-14 March 2014

University of California San Diego, USA

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14, March 2014 @ UCSDYoung Researchers Young Researchers are nominated as are nominated as the TG leaders!the TG leaders!

FERP consists of 13 task groups FERP consists of 13 task groups and 44 subtask groupsand 44 subtask groups the TG leaders! the TG leaders! g pg p

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Presentation Outline1. Introduction2 3D Designs of FFHR-d1 and New Proposals2. 3D Designs of FFHR-d1 and New Proposals

1) SC Magnet and Support Structure2) Neutronics Evaluation

Tamura

3) Flinabe Blanket mixed with Metal Powder4) Fabrication of Helical Coils with Segmented HTS Conductors5) D i f DC li f d ti t5) Design of DC power supplies for superconducting magnet

3. Improvement in Ignition Core Plasma and Design FlexibilityFlexibility1) Physics Analyses of FFHR-d1 Core Plasma2) Start-up Scenario of Ignition Core Plasma3) Improvement of the Pellet Fueling Scenario4) Sub-Ignition Design of FFHR-c1 before Demo

4. Summary and Future Plans

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1. Introduction

◆ From 2010, NIFS launched the Fusion Engineering Research Project (FERP) and started the re designResearch Project (FERP) and started the re-designof the LHD-type helical reactor FFHR-d1 towards DEMO in parallel with R&D researchesDEMO in parallel with R&D researches.

◆ In the first round of FERP the main parameters of◆ In the first round of FERP, the main parameters of FFHR-d1 were selected, establishing the system design code HELIOSCOPE and Direct Profiledesign code HELIOSCOPE and Direct Profile Extrapolation (DPE) method based on LHD plasma datadata.

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14, March 2014 @ UCSDA. Sagara et al., Fusion Eng. Des.87(2012)594.Design Parameters

GW

(on average)(on average)

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2nd RoundDesign of in-vessel components 4.642E+13

1.000E+14

2.154E+14

4.642E+14

1.000E+15

Fast neutron flux (>0.1 MeV)Fast neutron flux (>0.1 MeV)

1E+141E+14

1E+151E+15Blanket andshield B

Blanket andshield B

DivertorDivertor 4.642E+13

1.000E+14

2.154E+14

4.642E+14

1.000E+15

Fast neutron flux (>0.1 MeV)

1E+14

1E+15Blanket andshield B

Divertor

OV coilWC shield

FS+B4C

Flibe+Be/FSbreedingblanket

Design of in-vessel components

Posi

tion Y

1.000E+10

2.154E+10

4.642E+10

1.000E+11

2.154E+11

4.642E+11

1.000E+12

2.154E+12

4.642E+12

1.000E+13

2.154E+13

1E+131E+13

1E+121E+12

1E+111E+11

Helicalcoil

Helicalcoil

Blanket andshield A

Pumpingport

Pumpingport

Posi

tion Y

1.000E+10

2.154E+10

4.642E+10

1.000E+11

2.154E+11

4.642E+11

1.000E+12

2.154E+12

4.642E+12

1.000E+13

2.154E+13

1E+13

1E+12

1E+11

Helicalcoil

Blanket andshield A

Blanket andshield A

Pumpingport IV

coil

FS B4Cshield

Blanket TG3D blanket design

Position X

P

1.000E+08

2.154E+08

4.642E+08

1.000E+09

2.154E+09

4.642E+09

1.000E 10

1E+101E+10

1E+91E+9

1E+81E+8(n/cm2/s)(n/cm2/s)

Coreplasma

Blanketand

shield B

Blanketand

shield BHelical

coilHelical

coil

Position X

P

1.000E+08

2.154E+08

4.642E+08

1.000E+09

2.154E+09

4.642E+09

1.000E 10

1E+10

1E+9

1E+8(n/cm2/s)

CoreplasmaCore

plasmaBlanket

andshield B

Helicalcoil

Neutronsource

Pumping port

Helicalcoil3D neutronics

Superconducting coil Design Integration TG

Estimation of the neutron flux distribution by the 3D simulation code MCNP 【Blanket TG】

In‐vessel components TGDivertor design

TGCooling capability

Design Integration TGConsistency checkNumerical model

~8m~8m~8mDesign of the pumping ducts to realize a high conductance

Pumping

~2m~2m~2m

to realize a high-conductance 【In-vessel components TG】

~2m~2m ~0.5m~0.5m

~4m~4m

~8m~8m

~2m ~0.5m

~4m

~8m

Bottom portBottom port

~4m~4m~12m~12m

Bottom port

~4m~12m

Numerical modeling of the blanket poloidal shape being consistent with plasma and divertor legs at arbitrary toroidal angle 【Design integration TG】Sagara 6 / 17

14, March 2014 @ UCSD2. 3D Designs of FFHR-d1 and New Proposals

1) SC Magnet and Support Structure) g ppIn ISFNT-11: H. Tamura

Large apertures can be secured in the coil support made of 250mm-thick stainless steel 316 with a maximum stress of 600MPa level f t t l ti f 160GJ for a total magnetic energy of 160GJ.

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2) Neutronics evaluations)In ISFNT-11: T. Tanaka

Fast neutron flux (>0.1 MeV)BlanketBlanket

(n/cm2/s) Fast neutron flux (>0.1 MeV)(n/cm2/s)

1E+14

1E+15Divertorpumpingport

Blanket andshield

Blanket andshield

1E+14

1E+15

(n/cm /s)Poloidal

coilDivertorspace

Poloidalcoil

1E+13

1E+12

Helicalcoil

Blanket andshield

Coreplasma

1E+13

1E+12

Helicalcoil Helical

coil

Divertorpumpingport

Coreplasma

Inboardblanketspace:70 cm

1E+11

1E+10

plasma

Helicalcoil

1E+11

1E+10

port

Divertorspace

0 c

More than 10 years’ operation of magnets is feasible using the

1E+9Divertor space1E+9

Flibe blanket+ WC shield

Flibe blanket+ (Ferritic steel+B4C) shield

y p g ginboard side shield made with WC, where the total TBR is 1.08 with 90% Li-6 enrichment.

The divertor targets can be efficiently shielded from fast neutrons, The divertor targets can be efficiently shielded from fast neutrons, expanding the range of material choice (e.g., Cu alloys).

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3) Flinabe Blanket mixed with Metal Powder)

HastelloyHastelloy Flinabe is comparable to Fliberegarding TBR and nuclear shielding

In FED(2014) A.Sagara

Single‐mode Cavity microwavepowerH‐field

Single‐mode Cavity microwavepowerH‐field

regarding TBR and nuclear shielding.

Newly proposed based on preliminaryNewly proposed based on preliminary experiments, to effectively increase hydrogen solubility to increase thermal efficiency up to 46% to increase thermal efficiency up to 46%

with V alloys or 38% with Ferrite.Sagara 9 / 17


4) Fabrication of Helical Coils with Segmented HTS Conductors) gNIFS R&D achieved 45 kA at 20 k, 6 T, and low resistance (< 5n)A 100 KA class conductor sample

proposed as a promising method by connecting half-helical-pitch segments of 100kA class YBCO

Mechanical Bridge-Type Lap JointDeveloped by S. Ito, Tohoku Univ.Collaboration with

Tohoku Univ.

A 100 KA-class conductor sample is being fabricated to be tested.

segments of 100kA-class YBCO High cryogenic stability, High efficiency for cooling, Facilitation of the winding

GdBCO Tapes

Tohoku Univ. Facilitation of the winding process.

t Sec

tion

Joint

Join

Helical Coils

Th l t i l f i thThe electrical power for removing the Joule heating from all 7800 joints in two helical coils is <4MW (0.1% of Pfus) N. Yanagi, S. Ito, H. Hashizume et al.,

MT 23 (2013) in Bostonfor 20K operation, which is acceptable. MT-23 (2013) in Boston

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5) Design of DC power supplies for superconducting magnet) g p pp p g g

Connected in series and excited by In ISFNT-11: H. Chikaraishi

one power supply.Electro-magnetic forces on the helical

il l b l dcoil are always balanced.The total capacity of the power

li i <15 MW (0 5% f Pf )supplies is <15 MW (0.5% of Pfus).

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14, March 2014 @ UCSD3. Improvement in Ignition Core Plasma and Design Flexibility

1) Physics Analyses of FFHR-d1 Core Plasma1) Physics Analyses of FFHR-d1 Core PlasmaRecently it has been found, When the helical coil pith c = 1.20 is selected, instead of the former 1.25, When the helical coil pith c 1.20 is selected, instead of the former 1.25,

Shafranov shift of the magnetic axis in the high- can be mitigated by controlling the vertical magnetic fields, Bv.

I thi th l i l h t l i hl ½ f th h ti In this case, the neoclassical heat loss is roughly ½ of the heating power and the direct loss of particles calculated by GNET and MORH codes can be reduced to a tolerable level of 10 ~ 20 %.

FFHR-d1c = 1.25

012 w/o B

v

c

-2-10

w/ Bv

improved

Magnetic surface reconstructed by HINT2

10 12 14 16 18 20R (m)

FFHR-d1c = 1.20

code using the profile extrapolated from LHD data by DPE method (by J. Miyazawa)

(higher Aspect ratio)

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2) Start-up Scenario of Ignition Core Plasma2) Start-up Scenario of Ignition Core Plasma

In the quasi-1D particle balance model the feedback control of the

In ISFNT-11: T. Goto

model, the feedback control of the line-averaged electron density is adopted, instead of the fusion power, b th i ti d lbecause there is time delay corresponding to the time constant of the diffusion on the density profile.

By controlling the external heating power, smooth ignition access by a p g ysimple on-off control of the fixed-size pellet injection with 10Hz repetition can be achieved in 300 sec withcan be achieved in 300 sec with maximum heating power of 60MW.

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3) Improvement of the Pellet Fueling Scenario3) Improvement of the Pellet Fueling Scenario

Re-evaluation of the particle diffusioncoefficient which degrades with

R. Sakamoto, submitted to Nucl. Fusion3 15coefficient, which degrades with

increasing heating power density.However, Hi h d i j ti >15k /

3

2

020 /m

3 ]

15

10 Te [k

(a) (b)

High-speed injection >15km/s cannot be attained in present-day technology.

1

0

n e [1

1 00 50 1 00 50

5

0

keV]

/a= 0.9/a= 0.3

Large pellet size may cause unacceptable fusion output variation

1.00.50 1.00.50

20 N =2 0E22

r/a r/a

(c)

Therefore,Relatively shallow pellet penetration to /a = 0.3 with pellet injection velocity

20

10city

[k

m/s

] Np=2.0E22 (~5 % of plasma)

Np=4.0E22 (~10 % of plasma)

Np=8.0E22(same rate as LHD)/a 0.3 with pellet injection velocity

at 1.5 km/s is an acceptable compromise between pellet injection technology and burning plasma

10

0

Velo

c

1 00 80 60 40 20

(same rate as LHD)

technology and burning plasma performance.

1.00.80.60.40.20/a

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4) Sub-Ignition Design of FFHR-c1 before Demo4) Sub-Ignition Design of FFHR-c1 before Demo

P d “b f d t d t t t”

In FED (2014) A. Sagara

Proposed as “before-demo, compact and component-test” to keep design flexibility on FFHR.

FFHR-d1 FFHR-d1A FFHR-d1B FFHR-c1.0 FFHR-c1.1 FFHR-c1.2 Rc (m) 15.6 13.0 10.4 Vp,vac (m3) 1,877 1,421 823 823 419 Bc (T) 4 7 5 6 4 0 5 6 Bc (T) 4.7 5.6 4.0 5.6 Wmag (GJ) 162.5 223.5 67.8 125.1 61.4 c 1.25 1.20 1.0 0.85 f 5 1 3 7 1 9 1f 5.1 3.7 1.9 1 0 9.1 9.1 4.5 2.4 Paux (MW) 0 27 53 40 49 Pfusion (GW) 2.7 3.0 1.5 0.43 0.065 0.25 0.13 Q 16 1.2 6.2 2.6 n (MW/m2) 1.5 1.5 0.73 0.21 0.05 0.18 0.14 ~ 0.5 dpa ~ 2 dpa / year

i iblis possible

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14, March 2014 @ UCSDAs a consequence,

a multi-path strategy for FFHR-d1 has been introduceda multi path strategy for FFHR d1 has been introduced FFHR-d1A is the base for 3D

designs with a modified Aspect ratio b d i th Sh f hift

In FED (2014) A. Sagara

by reducing the Shafranov shift,

FFHR-d1B is a flexible design for the ignition core with the magneticthe ignition core with the magnetic (B) field enhanced by 20%,

FFHR-d1C is another flexible design with Configuration optimization of vertical

f field coils,

FFHR-c1 is a sub-ignition version as “before demo compact andas before-demo, compact and component-test”

(FFHR-2m is a commercial reactor.)( )

(FFHR-z1 is an ideal reactor.)

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Summary1. Fusion Engineering Research Project (FERP) in preparation for DEMO has

been launched in NIFS by starting the re-design of the LHD-type helical reactor FFHR-d1.

2. In the first round, the main parameters were selected.

3. The second round is preparing detailed 3D design of the superconducting magnet support structures, and 3D neutronics analyses, where the divertor targets can be efficiently shielded from fast neutrons.

4. A new Flinabe blanket mixed with metal powder was proposed.4. A new Flinabe blanket mixed with metal powder was proposed.

5. Fabrication of helical coils by connecting half-helical-pitch segments of 100kA-class YBCO high-temperature superconductors is proposed as a

i i th dpromising method.

6. Also in progress is improvement of the first round of the core plasma design, ignition start-up analyses, and fueling scenario. g , g p y , g

7. As a consequence, a multi-path strategy on FFHR-d1 has been introduced with versions of -d1A, -d1B, and -d1C, where design flexibility is expanded to include sub ignition with options FFHR c1 for "before demo compactto include sub-ignition with options FFHR-c1 for before demo, compact, and component-test."

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Roadmap of FERP

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Large Supplementary Budget of2.4 Billion JPY in FY2012

(1) SC Magnet‐ 13 T SC magnet test facility, ...

(2) Blanket‐Twin loop with 3T SC magnet, ...p 3 g ,

(3) Low‐Activation Material‐HIP device TEM creep test device ‐HIP device, TEM, creep test device, ...

(4) DivertorHigh heat load test de ice of 0 MW/m2 ll t t d ‐High heat load test device of 10 MW/m2, pelletron tandem accelerator of 1 MV, ...

(5)Tritium(5)Tritium‐Gas / liquid analyzer, …

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15 T SC Magnet FacilityR&D b SC t TGR&D by SC magnet TG

The superconductor testing facility at “Superconducting Magnet Systems Research Laboratory” will be upgraded (after 25 years operation) to increase the bias magnetic field from 9 T to 15 T so that 100 kA‐

3. Temperature‐variable

pg ( 5 y p ) g 9 5class conductor samples will be tested at temperature 4 – 50 K

World’s Highest B in a 700 mm Bore current leads

50 m

min a 700 mm Bore

Coil sample with 100 kA‐class superconductor

twisting

bending1. Outer coil2. Inner coil

700 mmSagara 21 / 17


Twin Loops with 3T SC Magnet R&D b Bl k t TGR&D by Blanket TG

CoolerCoolerOrosh2i‐2

World’s Highest B⊥

m

FLiNaK

LiPbEM

pump Mechanical pump

Orosh i 2‐Operational Recovery Of SeperatedHydrogen and Heat Inquiry‐ Forced circulation loops of FLiNaK

~5

SC magnet3 T

SC magnet3 T

Test sections

1.5m/s

‐ Forced circulation loops of FLiNaK(~500 ℃) and LiPb (~300 ℃)‐ Integrated test stand with a SC magnet of 3 T 22

~10 mHeater Heater

magnet of 3 T

Basic configuration of Orosh2i‐2

Specifications of OroshSpecifications of Orosh22ii‐‐2 2 ・Pipe diameter： 1 inch・Normal operation temp.： FLiNaK 500℃, LiPb 300℃・Maximum flow velocity： ~1.5 m/s ・Inventory： ~100 LMaximum flow velocity 1.5 m/s Inventory 100 L・Magnetic field： max. ~3 T（CS magnet）, 50 cm Φｘ 15 cm

Simulation of temperature and flow velocity in fusion blanket Integrated tests of MHD pressure drop, control of laminar and turbulence flow, g p p, ,hydrogen and heat recovery, corrosion behavior etc. under intense magnet field.

Test stand for elemental technologies developed in collaborative studies.Sagara 22 / 17


HIP Device, TEM, Creep TestR&D l ti ti t i l b Bl k t TG

Nano‐particle dispersion

strengthened Surface morphology and composition

Field Emission SEM（existing）

R&D on low-activation materials by Blanket TG

strengthened materials

Ch t i ti f

Field Emission SEM（existing）

Mechanical

Alloying

Mechanical

AlloyingCharacterization of nano‐particle dispersion

Evaluation of high Planetary ball milling TEM

Raw PowderRaw Powder

temperature strength and low temperature ductilityTarget （non‐irradiation）

TEM EDS

Yield stress：300 MPa <

Thermal creep：< 1 % at 100 khrHIP 5 mm 50 mm < 1 % at 100 khr

DBTT：< RT

Establishing fabrication

(Hot Isostatic Press)

High temperature creep test facility

Miniature size

Standard size

High temperature isostatic sintering and Establishing fabrication

technologytest facility

High speed impact test facility

isostatic sintering and joining test system（Fabrication of medium size test pieces）Sagara 23 / 17


High Heat Load Test DeviceR&D b I V l C t TGR&D by In-Vessel Components TG

ACT‐2‐Ultra high heat flux test standUltra high heat flux test stand‐ 10 MW/m2 of heat loading by 300 kW electron gun‐ Large vacuum vessel‐ R&D on material, cooling media, and bonding techniqueR&D on material, cooling media, and bonding technique‐ Realistic scale components

300 kW electron gun

Schematic of ACT‐2

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Accelerator for Material TestR&D b I V l C t d Bl k t TGR&D by In-Vessel Components and Blanket TGs

Non‐destructive analysis ‐ By 1 MV pelletron tandem accelerator‐Quantification of the retained H‐Multiple analysis of RBS, ERD, NRA,

③NRANa(Tl)D th fil f H i t-ray

and PIXE 1 MV tandem accelerator

Sample Material

HERD

4HeRBS =12 deg.=18 deg.

Depth profile of H isotopes

X ray

-ray

HERD2.8MeV- 4He2+

etc. ②ERDSSD①RBSSSD

g

④PIXESDDQuantitative analysis of Quantitative analysis of the plasma particles

X-ray

④ SDDdeposition element

300040005000

unts RBS

HSMo 40

6080

ount ERD

of the plasma particlesHigh resolution (ppm) element analysis

0100020003000

0 100 200 300 400 500 600

Cou

Channel Number

HSi

02040

0 100 200 300 400 500C

oChannel Number

SH

y

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design of helical reactor ffhr-d1 and c1 for steady state...

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