Technological readiness comparison for Helical and Tokamak

Technological readiness comparison for Helical and Tokamak DEMO

A. Sagara1, R. Wolf2 and H. Neilson3

1) National Institute for Fusion Science, Japan 2) Max-Planck-Institut fur Plasmaphysik, Germany

3) Princeton Plasma Physics Laboratory, USA

3rd IAEA DEMO Programme Workshop

11-14, May 2015 , Hefei

@University of Science and Technology of China

Special Session 2

Topics 1 : Contribution of integrated devices to closing the gaps (1)

3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 1 / 34

(1) This opportunity is a very good chance to know an example

of young researcher’s opinions in Japan on Technical Readiness Level (TRL) for Helical/Stellarater systems

in comparison with Tokamak systems.

(2) TRL is one of practical measures for discussing and having common

senses in the research communities. However, it should be noted that

TRL assessments must be made against established criteria*,

taking care not to inflate assessments and being as objective as possible.

There are no clear timelines pointed out in the TRL tables.

(3) In this presentation, therefore, some comments or new information

are added as far as possible for enhancing workshop discussions.

*USDOE example: DOE G 413.3-4A, “Technology Readiness Assessment Guide,” https://www.directives.doe.gov/directives/0413.3-EGuide-04a. (But

currently not applied to fusion by DOE)


https://www.directives.doe.gov/directives/0413.3-EGuide-04a





Ryuta Kasada*1, Takuya Goto*2, Shinsuke Fujioka*3, Ryoji Hiwatari*4, Naoyuki Oyama*5, Hiroyasu Tanigawa*5, Junichi Miyazawa*6

Young Scientists Special Interest Group on Fusion Reactor Realization

1Institute of Advanced Energy, Kyoto University 2NIFS

3Instituteo of Laser Engineering, Osaka University 4CRIEPI

5JAEA

US-Japan Workshop on

Fusion Power Plants and Related Advanced Technologies

8-9 March 2012

UC San Diego, Center for Magnetic Recording Research


The Young Scientists Special Interest Group on Fusion Reactor Realization is an active volunteer group which consists of the spirited young researchers beyond the frame of organizations or specialties who has a burning ambition to realize fusion reactors in their lifetime.

This work has been carried out by this voluntary group including many young researchers.


Visualization of Current States of the Key

Technologies

About TRL

Prerequisites for the TRL definition

Results of TRL evaluation

Tokamak

Helical

General issues

Perspective of R&D on the Key Technologies


After the FUKUSHIMA accident (3.11), R&Ds on energy technologies in Japan is being judged distinctly by the stakeholders whether they can contribute to the energy demand in near-future.

R&D on fusion reactors will be also judged from a viewpoint of not plasma science but energy development.

So some of young researchers (DEMO generation) decided to prepare the explanation to state the current and future of the key technologies for realization of fusion reactors.

A technology readiness levels (TRL) assessment is a powerful tool to visualize the maturity of the technologies under development.


A Technology Readiness Level (TRL) describes the maturity of a given technology

relative to its development cycle. TRL assessment were originally developed by NASA is to provide a common language among the

technology developers, engineers who will adopt/use the technology, and other stakeholders.

Dr. Tillack and the ARIES Team have already reported the TRL evaluation

for the fusion reactor.(2008)

While many types of reactor concepts have been studied in Japan, there were no activity to evaluate the technology maturity based on TRL.

TRL 1 2 3 4 5 6 7 8 9

Tech. A

Tech. B

Tech. C

Tech. D

completed In progress 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 7 / 34

Ref.

UCSD-CER-08-01


To identify Japanese customer (or young researchers) needs: Our target is to realize fusion reactor in 2050 when most of us will

alive.

For Magnetic confinement fusion: We set TRL5-6 for ITER and TRL7-9 for DEMO reactor.

Our TRL evaluation assumes two different scales of DEMO reactors allowing the development of components; such as 1GW fusion power (0.3GWe class) and a 3GW fusion power (1GWe class).

For Inertial fusion

Electrical power supply for solo reactor can be 0.2GWe because 5 sets of the reactor can generate 1GWe.

We set TRL6-7 for DEMO and TRL8-9 for prototype power reactor.


TRL Definition

State for magnetic

confinement fusion

(incl. R&D in DEMO)

State for inertial fusion

TRL1 Basic principles observed and reported

Proof of concept Proof of concept TRL2

Technology concepts and/or applications formulated

TRL3 Analytical and experimental demonstration of critical function and/or proof of concept

Component validation Component validation

(including validation of

reactor-core physics and fusion

burning)

TRL4 Component and/or bench-scale validation in a laboratory environment

TRL5 Component and/or breadboard validation in a relevant environment

ITER (Exp. Reactor) TRL6

System/subsystem model or prototype demonstration in relevant environment

Demonstration reactor TRL7

System prototype demonstration in prototypic environment

DEMO reactor TRL8 Actual system completed and qualified through test and demonstration

Prototype reactor TRL9

Actual system proven through successful operations

Remarks After GNEP’s definition Main machines: ITER &

DEMO

Main machines: (NIF, LMJ, FIREX etc),

DEMO, Prototype

GNEP: Global Nuclear Energy Partnership 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 10 / 34

Mt. Takao

599m

Mt. Fuji

3,776m

Everest 8,848m

Attention: Higher mountain is more different to climb; TRL is not same quantity.

Top

Top

Top

Pictures from Wikipedia and Google

half

half

half

Because the expected outcome and the degree of difficulty for commercialization are

different. TRL is the evaluation of individual technical achievement. It is important to

clarify that clear criteria and specific measures rather than the numerical value of TRL.

Yama-girl

Pro. Climber


Summarized by Ryoji HIWATARI & Naoyuki OYAMA


Currently it has become possible to build a technology roadmap by the accumulation of knowledge based on the past reactor designs.

Dr. Okano team build a technology map based on the WBS (work breakdown structure) which consists of 18 area including >1000 issues.

We can evaluate the TRL on the WBS.


TRLs are evaluated for the different electrical power plants of 0.3GWe and 1GWe.

Even supposing the electrical output, fusion power output and the size is not determined.

We show typical examples of the combination of size, plasma performance and engineering conditions.

And we evaluate the TRL on these examples.


Major radius: 6.5m Major radius: 8.5m

0.2GWe for

thermal

efficiency of

30％ (0.3GWe for thermal

efficiency of

40％)

• bN>3.0⇒ITER high-performance plasma

• Thermal efficiency > 30%⇒to be tested in ITER TBM

• Higher strength magnetic field is

needed for TF coil.

• Neutron wall load: ~1.5MW/m2

• Full-sector removal maintenance

• bN=2.0以上⇒ITER normal plasma • Thermal efficiency > 30%⇒to be

tested in ITER TBM


needed for bigger TF coil.


• Module maintenance

1GWe for

thermal

efficiency of


efficiency of

30％)

• bN>4.0以上⇒Advanced plasma ( over ITER) with conducting walls in blanket

• Thermal efficiency > 40%⇒advanced blanket


needed for TF coil



• NBI system efficiency: >30％

• bN>３.0⇒ITER high-perfomance plasma




• Neutron wall load: ~2.０MW/m2



Scale Lab Large Device ITER DEMO

TRL 1 2 3 4 5 6 7 8 9

System design

Core plasma βN~4.0

βN~2.0

TF coil 16T

13T

Blanket T.E. 40%

T.E. 30%

Under discussion on DEMO mission.

ITER normal operation

JT-60SA

ITER scale

Higher magnetic field coil needs R&D on materials.

WCSB-WLK：ITER-TBM

Analysis for the selection and specification required to determine the primary design is underway.

・When the DEMO design is based on the findings of ITER, there is no Tokamak-specific bottleneck technology.

・TF coil technology can be a bottleneck, which depends on DEMO mission.

・Blanket technologies are not specific for Tokamak.


Construction of a new 600 W He refrigerator/liquefier A 600 W He refrigerator/liquefier with an additional function

is replacing.

・ Its completion will be on March 31, 2015.

・ He coolant of variable temperature can be supplied to

test facilities for superconductors and magnets.

Specifications

・Refrigeration / liquefaction capacities

600W at 4.5K / 250L/h

・Supply capacities of He coolant

350 W at 4.55 K (50 g/s SHe)

1.0 kW at 20 K – 30 K (18 g/s GHe)

1.5 kW at 40 K – 50 K (20 g/s GHe)

A unique test facility of superconductors in the world by

operating with a 13 T magnet and a 75 kA power supply

13 T & 70 cm boar 75 kA power supply

Corroboration researches of wide scope are acceptable.

・HTC and new superconductor tests with the 4.5 K – 50 K

temperature range

・CIC conductors and coils cooled by SHe

・Conventional superconductors such as NbTi and Nb3Sn

Unloading of the cold box

Unloading of

a He compressor

S. Hamaguchi, et al., P6-11, 24th ITC

New

function


ブランケット交換で性能向上可能、主・副半径をどうする主要課題

冷却材条件(冷媒、圧力、温度) 変更、発電システム総交換必要性

原型炉概念Demo-CREST(主半径7.5m)で提案中

Major radius: 6.5m Major radius: 8.5m

0.2GWe for

thermal

efficiency of


efficiency of

40％)

• bN>3.0⇒ITER high-performance plasma

• Thermal efficiency > 30%⇒to be tested in ITER TBM


needed for TF coil.



• bN=2.0以上⇒ITER normal plasma • Thermal efficiency > 30%⇒to be

tested in ITER TBM





1GWe for

thermal

efficiency of


efficiency of

30％)

• bN>4.0以上⇒Advanced plasma ( over ITER) with conducting walls in blanket



needed for TF coil



• NBI system efficiency: >30％

• bN>３.0⇒ITER high-perfomance plasma




• Neutron wall load: ~2.０MW/m2


Upgrade by replacement of blanket 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 18 / 34

U.S. ARIES team has compared “advanced” vs.

“conservative” assumptions for 1 GWe tokamak power plant designs

Conservative ARIES-ACT2

Advanced ARIES-ACT1

Ph

ys

ics

A

ss

um

pti

on

s

βNtotal 2.6 (no wall) 5.8 (with wall)

H98 1.25 1.65

n/nGr minimize minimize

Te

ch

no

log

y

As

su

mp

tio

ns

Blanket: DCLL (RAFM)

(ηth~0.44)

SCLL (SiC-comp)

(ηth~0.58)

qdivpeak <10 MW/m2 <15 MW/m2

Re

su

lts

R 9.75 m 6.25 m

BTaxis 8.75 6.0

Pfus 2.6 GW 1.8 GW

QDT 25.0 42.5

QENG 3.1 6.5

Special Issue (12 papers): Fusion Science and Technology 67, Vol. 1 (Jan. 2015)


Summarized by Takuya GOTO & Junichi MIYAZAWA


The NIFS fusion engineering research project (FERP):

has initiated the conceptual design activity of LHD-type helical reactor “FFHR-d1” by 13 task groups.

Taking the knowledge of past commercial type design concept FFHR-2m , the project aims to demonstrate the next generation reactors as soon as possible.

TRL evaluation:

Identification of major R & D items by the Task activities.

Because of no ITER-scale device for helical system, DD experiment phase, large scale engineering component tests, and numerical simulation is defined as TRL=5, 6.

Specific-issues for helical type reactor

Original fabrication and maintenance method should be developed for 3D

structure containing winding helical coil and blanket.

Due to the big size, building and fabrication process is probably special.

A big advantage is that constant heating input is not needed.


Scale Lab LHD/ITER-EDA LHD-DD/ITER

& DEMO-EDA

DEMO

TRL 1 2 3 4 5 6 7 8 9

System integration

Core plasma

Supeconducting magnet coil with low Temp. system

Blanket technologies

Internal components

Plasma heating

Plasma measurements

Tritium & safety

Power supply system

Building

To determine key parameters of the reactor FFHR-d1.

By extrapolation of the LHD experiments and numerical analysis.

Various conductors are examined. It is necessary to develop an

efficient method of winding.

Materials evaluation and neutronics design have

been progressed mainly for molten salt blanket.

Studies on 3D layout design and remote maintenance is needed.

Long pulse, high-efficiency

Not helical-specific issue.

Needs for active control of magnetic axis.

Scale is larger than the tokamak building. The schedule should be

consistent with the main component construction process.

Under validation for LHD-DD experiments

(using LHD and ITER tech.)


Progress of 100-kA-Class HTS Conductor and Coil Design

HTS Conductor

Current 100 kA ×

1 hour

FFHR-d1 Helical Coil

(Helium gas cooling 20 K)

Mechanical

Lap Joint

4.2 K

100 kA-class HTS

conductor design (2014)

100-kA-class High-Temperature Superconducting

(HTS) conductor renovates the world record 100 kA current sustained for 1 hour [1, 2]

Tensile test for a single-tape joint was conducted Contact pressure of 50 MPa is needed to withstand the

shear strength evaluated by 3D-FEM [3]

(1) N. Yanagi et al., FIP/P8-21, 25th IAEA Fusion Energy

Conference (2014), St. Petersburg, Russia.

(2) S. Ito et al., Plasma and Fusion Research 9 (2014)

3405086.

(3) S. Ito et al., IEEE Transactions on Applied

Superconductivity 25 (2015) 4201205.

3D-FEM analysis of EM stress

on the helical coil and conductor

Tensile and shear strength test

for single-tape joint

Collaboration with

Tohoku Univ.

Maximum shear

strength obtained by 3D-FEM

analysis 50 MPa


(comments from Helias) C. Beidler, F. Warmer, R. Wolf et al.

HELIAS as an advancement of the optimized stellarator W7-X

Central points which are also important for TRL

• Plasma performance: W7-X designed is based on an self-consistent optimization

w.r.t. neoclassical transport, plasma equilibrium, exhaust concept and fast ion

confinement

Good confinement properties and high b require high density

Development of scenarios with optimized fuelling, density and impurity control

Issue: Conditions to excite fast ion driven instabilities difficult to achieve in W7-X

• Plasma heating based on steady-state ECRH

Extrapolation to high density needs to be demonstrated

• Comparatively low magnetic field ITER superconductor and coil technology is sufficient

• Blanket integration under investigation ( Karlsruhe Institute of Technology)

• EU Roadmap assumes intermediate step between W7-X and commercial

Stellarator-PP, in particular for demonstrating burning plasma and fast ion confinement (further improvement of configuration)

Step to commercial Stellarator-PP based on this and technology development

from tokamak DEMO

Decision point how to proceed when W7-X high performance steady-state

plasmas have been demonstrated (~2025) 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 24 / 34

Summarized by Ryuta KASADA & Hiroyasu TANIGAWA


Scale Lab Large device ITER DEMO

TRL 1 2 3 4 5 6 7 8 9

Blanket: RAF-Solid-Water (BA etc) (ITER-TBM

etc)

(ITER-TBM

etc)

(DEMO-

TBM etc)

Blanket : advanced

Diverter (for ITER) Should be discussed

Diverter (for DEMO)

S.C. coil (for ITER) Nb3Sn+JJ1+Insulator

S.C. coil (advanced) Nb3Al+?St. Mat. +?Insulator

Tritium fuel cycling system

Vacuum tech. (for ITER) Should be discussed

Vacuum tech. (for DEMO)

Tritium recovery system

Tritium safety system

ITER-TBM like component tests is needed.

Effect of neutron irradiation should be examined.

No candidates for structural material and insulator,

Significant advance from ITER is needed for steady state operation of DEMO.


Flinak Loop

SC magnet (3T

perp. to flow)

Li-Pb Loop

Oroshhi-2 in NIFS(2015~) A. Sagara et al., FS&T, in press


LiPb loop

FLiNaK loop

3 T superconducting

magnet

(EM: electro- magnetic)

Phased experiments on MHD effects on LiPb Corrosion and mass transfer under no-equilibrium and DT

Heat transfer on FLiNaK under high B Hydrogen charging and recovery

Metal powder mixed system Operation of S-CO2 system for Flinak

A. Sagara, FED 89 (2014) 2114–2120

Future plan

10MPa S-CO2 compressor

and turbine system

~100kW

A. Sagara et al., FS&T, in press

nano-fluid


Tritium cycle

to reduce the total inventory

R. Sakamoto, NIFS Annual repo. 2010-2011. A. Sagara et al., Fusion Eng. Design 87 (2012) 594.


Scale Lab Large device ITER DEMO

TRL 1 2 3 4 5 6 7 8 9

Recovery of Li

from seawater

Enrichment of

Li-isotope

Back-end tech.

Only basic research in Japan.The Hg amalgam is not suitable for environments.)

Cost can be drastically reduced due to the demand of Li-ion buttery.

• Although limited to the magnetic confinement fusion systems, with respect to the

divertor technology, the existing technology intended for ITER is a Dead-End

Technology. It is necessary to expedite the technology development for DEMO.

• For handling tritium in large facilities, the challenge is particularly tritium recovery

system. The first entire operation of the fuel circulation system is examined in ITER.

Extraction and proposals for improvements of the problem is expected. Towards

the DEMO reactor, development of tritium measurement-control technology,

specifically for the technology to predict and measure the inventory is needed.

• Lithium isotope enrichment technology development is very important.

These common issues mostly need facilities such

as IFMIF for the evaluation of blanket, divertor

evaluation facility, and Tritium pilot plant, in addition

to ITER.


As conclusion of this talk


As results of the TRL evaluation for the realization of three types of fusion reactor concept, it is seen that the specialty of each concept has high TRL but some of engineering issues has extremely low TRL. For example, TRL on measurement and control of burning plasma scenario is

very low.

For blanket technologies including structural materials, smooth progress

has been confirmed by TRL evaluation. Now jump to the R&D on "mass manufacturing technology" is required.

TRL related to the tritium fuel system in all concepts is relatively low. R&D with large-scale handling is needed.

It is essential to establish a method for lithium procurement. As for the magnetic confinement concepts, it is necessary to harmonize

research and development of the divertor component based on their integrated design because the current technology is probably dead-end one for the DEMO divertor

There are many difficulties in engineering issues from ITER to DEMO. It is necessary to set the test phase in DEMO (like a DEMO-TBR) and to make engineering machines to test diverter, blanket with tritium handling.


The TRL assessment summarized the current state of the technical challenges of each element, clarified the criteria and lead to the "visualization" of technological bottleneck for the realization of various fusion reactor concepts.

The young researchers have successfully obtained the point of view of reactor design aiming to realize fusion reactors and mutual understanding between young researchers of different fields in fusion engineering and science has been promoted. 3rd IAEA DEMO Pro. WS, May 12, 2015, Sagara 33 / 34

Summary of comments 1. TRL assessments must be made against

established criteria*, taking care not to inflate

assessments and being as objective as possible.

2. There are no clear timelines pointed out in the TRL

tables. Feedforward to a road map is expected.

3. To fulfil all the required developments in the orange

bars, a huge effort is required. Compatibility with

the current research efforts should be assessed.

4. While ITER will be the first test-bed for a fusion

blanket, other developments towards DEMO are

not so clear.

5. For helical system, without burning devices,

extrapolation to the core plasma is the key issue.


Documents

Technological readiness comparison for Helical and Tokamak