NIFS-FERC

Japanese Perspectives on Liquid Blanket Research and Relating Collaboration

T. Muroga

Fusion Engineering Research Center,

National Institute for Fusion Science, Japan

APEX/TBM Project MeetingNovember 3-5, 2003, UCLA

NIFS-FERC

Japanese Fusion Research Organizations

Two organizations carry out fusion research in Japan

JAERI

Mission given by the government

Project-oriented

ITER official contractor (at present)

Universities (group of independent Professors)Mission defined by their own (more interest-oriented)

Project by mutual agreement

Scientific approach

Playing complementary roles but sometimes causing problems (especially in making national decisions)

NIFS-FERC

Introduction of NIFS-FERS

NIFS (National Institute for Fusion Science) is the inter-university research institute

Coordination and enhancement of University research

LHD as the core project

Fusion Engineering Research Center in NIFS Established in 1999

Coordination and enhancement of University activity on

Structural Materials

Blanket (started 2001, activity still limited)

SC system (nuclear technology related, 2003~)

NIFS-FERC

Introduction of NIFS-FERS (cont.)

Present activities of NIFS-FERSDevelopment of vanadium alloys (JAERI is the core for RAFS)

Fabrication of reference ingots and characterization by universities

MHD coating (started in 2000)

Fabrication and corrosion tests

Netronics (started in 2002)

Liquid blanket, IFMIF

IFMIF-Key Element Technology Verification

Collaboration with universities (Li free surface in Osaka -)

ITER-TBM needs coordination of Universities and thus potential major activity of FERC in the future

NIFS-FERC

Outline of Presentation

What is agreed in Japan as strategy for liquid blanket research?

Roadmap to powerplant

Responsibility sharing between JAERI and Universities/NIFS

Research emphasis in Universities/NIFS 　ITER participation/contribution (Including TBM)

International collaboration

NIFS-FERC

General Classification into Two Lines for Fusion Development

Fast realization of power demonstrationConsidered to be important for public support

Based on modest progress in science/technology

Relatively large budget allocation for near-term

Project-oriented approach

Currently JAERI is the core for this line

Exploration of advanced system Long-term research including fundamentals

Increasing attractiveness (cost, safety, environment) is thought to be crucial for fusion development

Science-oriented approach

Currently University/NIFS is the core for this approach

NIFS-FERC

EU Strategy Has Also Two Lines

• power into grid:

• DEMO & 1st generationpower plants (economicallycompetitive)

• DEMO & 1GPP will be basedon similar/identical physics &technology

• further improve economic& environmental properties

• 2nd generation („advanced“)power plants

• R&D now, due to long leadtimes (materials)

• ongoing physics research

Reference system

Advanced system

Lackner, ICFRM-10

NIFS-FERC

Two Lines for Materials/Blanket in Japan

Fast realization lineRAFM/water as a reference system

RAFM/Supercritical water (optional 1)

ODS/Supercritical water (optional 2)

Relatively large budget allocation for the development

RAFM test planed to be dominant in early stage of IFMIF

Advanced line Presently liquid blanket systems (V/Li and Flibe) and SiC/He

Focusing on fundamentals and key feasibility issues

Science-oriented approach

University/NIFS is the core for this approach

Major subjects of JUPITER-II

NIFS-FERC

Two Lines for ITER-TBM(preliminary discussion)

Fast realization lineRAFM/water TBM

Efforts focused on Day One TBM

Contribute to licensing First Power Generation Plant together with early IFMIF data

Advanced line Liquid blanket systems (V/Li and Flibe) and SiC/He

Plan to start TBM either from Day One or in the later phase of ITER

Agreed to keep these activities irrespective of the selection on Day One TBM

NIFS-FERC

Schematic Roadmap for Materials and Blanket Development in Japan

Materials and Blanket System Development

Reference Material (RAFM) and System

Design Construction Operation

ITER

Power Generation Plant

Irradiation Test, Materials Qualification and System Performance TestIFMIF

Advanced Powerplant Design

(Staged construction and operation)

(Licencing) (Blanket test)

Blanket Module Test

Approximate calendar year 2015 2020 2030 2040

Advanced Materials (V-alloy, SiC/SiC --)and System

Fast realization line (Currently JAERI leadership)

Advanced line (Currently NIFS/University leadership)

NIFS-FERC

Recent Activity of ITER-TBM in Universities/NIFS

From Japan, only solid breeders were proposed to ITER via JAERIParticipation to scientific aspects of ITER research by NIFS/ Universities is being enhancedA NIFS-collaboration activity started in 2002, in which liquid blanket test module is explored

Party Proposed TBM-type

JAPAN Solid - Water

Solid - Helium

EU Solid - Helium

Li-Pb - Water

Russia Solid - Helium

Lithium

US Solid - He

Lithium

Party Proposed TBM-type

JAPAN Solid - Water

Solid - Helium

EU Solid - Helium

Li-Pb - Water

Russia Solid - Helium

Lithium

1995 2001

NIFS-FERC

NIFS Collaboration Activity for ITER-TBM

Subject University NIFS Initial Activity

Thermal-structural Analysis

HashizumeHoriike

Takahashi

ImagawaNagasaka

Thermal-structural analysisMHD-reduction

Neutronics Iguchi T. Tanaka T-productionNuclear Heat, After Heat, Activation

T-recovery S. TanakaFukada

Suzuki Hot trapCold trap

Materials MatsuiAbe

Nagasaka Design data

Flibe-module concept

Terai Sagara Concept exploration

Design Integration

MatuiJ AERI

MurogaSagara

Reporting

Subject University NIFS Initial Activity

Thermal-structural Analysis

HashizumeHoriike

Takahashi

ImagawaNagasaka

Thermal-structural analysisMHD-reduction

Neutronics Iguchi T. Tanaka T-productionNuclear Heat, After Heat, Activation

T-recovery S. TanakaFukada

Suzuki Hot trapCold trap

Materials MatsuiAbe

Nagasaka Design data

Flibe-module concept

Terai Sagara Concept exploration

Design Integration

MatuiJ AERI

MurogaSagara

Reporting

SubjectSubject UniversityUniversity NIFSNIFS Initial ActivityInitial Activity

Thermal-structural Analysis

Thermal-structural Analysis

HashizumeHoriike

Takahashi

HashizumeHoriike

Takahashi

ImagawaNagasakaImagawaNagasaka

Thermal-structural analysisMHD-reduction

Thermal-structural analysisMHD-reduction

NeutronicsNeutronics IguchiIguchi T. TanakaT. Tanaka T-productionNuclear Heat, After Heat, Activation

T-productionNuclear Heat, After Heat, Activation

T-recoveryT-recovery S. TanakaFukada

S. TanakaFukada

SuzukiSuzuki Hot trapCold trapHot trapCold trap

MaterialsMaterials MatsuiAbe

MatsuiAbe

NagasakaNagasaka Design dataDesign data

Flibe-module concept

Flibe-module concept

TeraiTerai SagaraSagara Concept explorationConcept exploration

Design Integration

Design Integration

MatuiJ AERIMatuiJ AERI

MurogaSagaraMurogaSagara

ReportingReporting

Examination of Li/V first and then followed by FlibeSupport from JAERIFirst output expected in 2004

NIFS-FERC

Purpose of Li/V ITER-TBM(current discussion)

Feasibility of no-Be and natural Li blanketUse of 7Li reaction for enhancing TBR in contrast to Russian Be+6L

i enriched TBM

Validation of neutronics prediction

Technology integration for V-alloy, Li and T

NIFS-FERC

ITER with Li/V self-cooled blanket - MCNP calculation by T. Tanaka (NIFS) -

Centersolenoid

Vacuumvessel

+Filler

Blanket

Coilstructure

Plasma

[ Inboard ]

SS,H2O

Blanket FWVacuumvessel

V-4Cr-4Ti walls,Natural Li

SS (60%),Li coolant (40%)

40 cm

[ Outboard ]

SS (60%),Li coolan (40%)

V-4Cr-4Ti walls,Natural Li

FW Blanket

SS,H2O

40 cm

(*Dimensions from ITER Nuclear Analysis Report)

Vacuumvessel

Input geometry for MCNP calculation *

SS,H2O

1 m

A

A

B

B

A : Standard ITEF-FEAT blanket

B : ITER with V/Li full blanket

NIFS-FERC

330 340 350 360 370 380 390 400 4100.0

1.0x10- 7

2.0x10- 7

3.0x10- 7

4.0x10- 7

5.0x10- 7

Total 6Li 7Li

Tri

tium

pro

ducti

on r

ate (

g/F

PD

/cm

3 )

P osition (cm)840 860 880 900 920 940

0.0

1.0x10- 7

2.0x10- 7

3.0x10- 7

4.0x10- 7

5.0x10- 7

Total 6Li 7Li

Tri

tium

pro

ducti

on r

ate (

g/F

PD

/cm

3 )

Position (cm)

ITER with Li/V self-cooled blanket - Local TBR -

Inboard

Outboard Total

Contribution

of 7Li (%)Li/V

blanket 0.30 0.92 1.22 33Coolant in filler 0.029 0.15 0.18 2.6

Total 0.33 1.1 1.4 ---

Local TBR (Full Coverage)*

(* JENDL 3.2)

Distribution of tritium production rate

FW

Blanket

FillerFiller

Blanket

FW

(a) Inboard (b) Outboard

■ Significant contribution of 7Li to TBR

NIFS-FERC

Neutron spectrum at first wall of Standard and V/Li Blanket

10- 6 10- 5 10- 4 10- 3 10- 2 10- 1 100 101109

1010

1011

1012

1013

1014

1015

1016

Neu

tron

flux

(n/

cm2 /s

/let

harg

y)

Neutron energy (MeV)

ITER- FEAT ITER- Li/ V

Comparison of Neutron Fluxat Outboard First Wall

10- 6 10- 5 10- 4 10- 3 10- 2 10- 1 100 10110- 3

10- 2

10- 1

100

101

102

103

104

Cro

ss s

ecti

on f

or t

riti

um p

rodu

ctio

n (b

arns

)

Neutron energy (MeV)

6Li (n,a) T 7Li (n,na) T

Cross Section for Tritium Production(JENDL 3.2)

■ Significant difference between thermal neutron component in ITER-FEAT and ITER-Li/V

■ Thermal neutron should be shielded in the TBM area of ITER-FEAT for the purpose of simulating V/Li blanket condition

NIFS-FERC

Russian Li/V self-cooled test blanket module 　 - Structure -

■ 6Li enriched coolant 　 (7.5 % ==> 90%)

Plasma

■ Li layer x 2, Be multiplier ==> 6Li (n, ) T

V-5Cr-5Ti

Li layer(6Li : 90%)

Bemultiplier

WCShield

(Reflector)

SS(60%)+

H2O(40%)

Structure of Russian Li/V TBM

505

1720

(Unit : mm)

■ Maximize the 6Li reaction to demonstrate DEMO reactor breeding tritium by 6Li

NIFS-FERC

Plasma

Li layer (1) Li layer (2) [6Li : 90%]

Be WC

SS+H2O

Tritium production rate in Li layersand contribution of 6Li and 7Li

TBM surface

Li layer (1) Li layer (2)

Total : 0.09 (g/FPD)

Russian Li/V self-cooled test blanket module 　 - Tritium production -

0 10 20 30 40 50 600.0

2.0x10- 6

4.0x10- 6

6.0x10- 6

8.0x10- 6

1.0x10- 5

Total TP R TP R from 6Li TP R from 7Li

Tri

tium

pro

duct

ion

rate

(g/

FP

D/c

m3)

P osition (mm)

SUS+

H2O

SS316TBMframe

Plasma

Li/VTBM

NIFS-FERCVerification of (1) Coolant circulation (2) MHD coating

Verification of(1) Neutron transport(2) Tritium production from 7Li

Inlet/outlet pipes

Tentative design of Li/V TBM

505

1720

Plasma SS(60%),H2O(40%)

Li layerV-4Cr-4Ti

Li : ~0.027 m3

210

210

470

Tentative design of Li/V self-cooled TBM by NIFS/Universities

(Unit : mm)

■ Thick Li tanks for verification of neutron transport

■ Verification of TPR for 7Li

SS(60%),H2O(40%)

Plasma

SS316TBMframe

Li/VTBM

NIFS-FERC

Tritium production rate in Li layers

Tentative design of Li/V self-cooled TBM - Tritium production -

Plasma

Covering by B4C

Contribution of 7Li to tritium production

■ For verification of tritium production from 7Li (n, n)T reaction

- Reduction of thermal neutrons by B4C shielding

(3)Li layer (1)

(2)

Li layer (1) Li layer (1)

SS(60%),H2O(40%)

(4)(5)

(2)(3)

(2)(3) (4) (5)(4) (5)

0 100 200 300 400 5000

20

40

60

80

100 Nat. Li Nat. Li+B4C(7.5mm)

Con

trib

utio

n of

7 Li

(%)

P osition (mm)0 100 200 300 400 500

0.0

1.0x10- 6

2.0x10- 6

3.0x10- 6

4.0x10- 6

Nat. Li Nat. Li+B4C(7.5mm)

Tri

tium

pro

duct

ion

rate

(g/

FP

D/c

m3 )

P osition (mm)

NIFS-FERC

Experimental parameter for Li/V TBM - Adjustment by B4C shield -

Changes in contribution of 7Li by B4C covering

■ Contribution of 7Li to TPR can be adjusted by thickness of B4C shield

10 cm in front side

10 cm in rear sideLi/V blanket

Li/V TBM

Russian TBM

0 5 10 150

20

40

60

80

100

Con

trib

utio

n of

7Li

to

TP

R (

%)

Thickness of B4C cover (mm)

10 cm in front side

10 cm in rear side

NIFS-FERC

Future Participation to ITER-TBWG(discussion not started)

Discussion on participation of University/NIFS to ITER-TBWG will start soon

Possible options may beTune the present blanket activity to TBWG schedule

Start engineering design for V/Li TBM

Concept definition and start engineering design for Flibe TBM

Keep the present pace with weaker interaction with TBWG

In this case, we will not strongly propose Day One TBM

Keep the present advance blanket research activities irrespective of the selection on Day One TBM

NIFS-FERC

JUPITER-II

JUPITER-II is a mission-defined collaboration program

Advanced blanket (in contrast to JAERI’s FS/water)

Task plan and Check and Review

Use of core facilities (HFIR, STAR --), which are

unavailable in Japan, is the rationale for the collaboration

(transferring budget from J to the US)

Major change of the framework need re-evaluation by standing committees and will face a risk

Most Japanese JUPITER-II participants have strong scientific interests in the present tasks and have small incentive to make extensive change in the framework

NIFS-FERC

JUPITER-II Possible Fine Tuning(unofficial, Muroga private idea)

“We cannot propose any concept for ITER-TBM at present with the lack of corrosion data” (Sze-Muroga e-mail agreement)

Shift some effort from vanadium irradiation to MHD coating/corrosion

MHD coating is the critical issue for both long term blanket development and entry to Day One TBM

REDOX,Flibe-materials interaction should be enhanced

MHD related design activity should be enhancedLenient requirement to MHD coating for V/Li

However, HFIR and Tritium activity must be maintained because of program need and participants incentives

NIFS-FERC

Comment/questions to US Discussion on TBM Selection

What is the philosophy of selecting TBM?

Technical feasibility and ?

Japan : Roadmap to Powerplant

What is the community selecting TBM?

Liquid breeder for Japan :

University/NIFS including B, M, T, S –

(If materials people are not involved heavily, the impact of the decision on materials program must be small)

Why two options (only because of budget?)

Number of available port no longer the factor

What is the fate of the concepts not selected for the first-day TBM? (Longer-term strategy)

NIFS-FERC

End of presentation

NIFS-FERC

MHD Coating – Necessity–

Insulator coating inside the ducts a possible solution

MHD Pressure Drop

・ Load to pumping system

・ Force to structures

Pressure Drop ： proportional to

Flow length 、 Velocity 、 B2 、 Duct thickness 、 Conductivity of Li and Duct 　　　

Magnetic Field

Li FlowForce

Duct

NIFS-FERC

MHD Coating Candidates (1)–Free Energy

Stable ceramics in a quite reducing condition

Selection from the free energy data

CaO 、 Y2O3 、 Er

2O3 、CaZr(Sc)O3 、

AlN 、 BN100℃

1 x 10-3 2 x 10-3 3 x 10-3

１／Ｔ (K)

Ｖ２Ｏ５

Ｌｉ２Ｏ

NIFS-FERC

MHD Coating Candidates (2)– Bulk Compatibility

Potential candidates

Y2O3

Er2O3

AlN with N control

CaZr(Sc)O3 (~700C)

others

10m/y

Japan-US JUPITER-II Collaboration (Pint, Suzuki et al. 2002)

NIFS-FERC

MHD Coating DevelopmentPresent Efforts

Development of coating technology

RF-sputteringEB-PVDArc Plasma Deposition

Characterization of the coating

ResistivityHigh temperature stabilityCompatibility with LiRadiation induced conductivity

In-situ coating technology 104

106

108

1010

1012

0 200 400 600 800 1000

coatingtempdata2

RF-sputtering AlN coating (ohm*cm)RF-sputtering Y2O3 coating (ohm*cm)RF-sputtering Er2O3 coating (ohm*cm)Arc-P-Depo. Er2O3 coating (ohm*cm) by Dr. FujiwaraEB Depo. Y2O3 coating (ohm*cm) by Dr. Pint

Temp.(C)

Japan-US JUPITER-II Collaboration (Suzuki, Pint et al. 2003 ）

NIFS-FERC

In-situ Coating

The in-situ coating method has advantages as,

possibility of coating on the complex surface after fabrication of component

potentiality to heal the cracks without disassembling the component

CaO coating has been explored

Ca++

M2Ox

O2-

Ov

Mx+

MLi

V-alloy Li(M)

Ca++

M2Ox

O2-

Ov

Mx+

MLi

V-alloy Li(M)

NIFS-FERC

Problems of the CaO Coating and New Effort on Er2O3

It was found that the CaO coating, after formation, dissolved at high temperature (600, 700C)

CaO bulk is inherently not stable in pure Li at high temperature, continuous supply of oxygen is necessary to maintain the coating

Er2O3 is much more stable at high temperature

It is expected Er2O3, once formed, be stable in Li for a long time

Er2O3 is stable in air, combination of dry-coating and in-situ coating is more feasible

10m/y10m/y

CaO

Er2O3

NIFS-FERC

In-situ Er2O3 Coating on V-4Cr-4Ti

Er2O3 layer was formed on V-4Cr-4Ti by oxidation, anneal and exposure to Li (Er) at 600C

The coating was stable to 300 hrs

The resistivity was ~1013 ohm-cm

0 5 10 15 20 25 300

0.5

1

1.5

2x 105 Er2O3-layer-0062_1.PRO

Sputter Time (min)

Inte

nsity

Er4dO1s

V2p3

0 5 10 15 20 25 300

0.5

1

1.5

2x 105 Er2O3-layer-0067_1.PRO

Sputter Time (min)

Inte

nsity

O1s

Er4dV2p3

0 5 10 15 20 25 300

0.5

1

1.5

2x 10

5Er2O3-layer-0035_1.PRO

Sputter Time (min)

Inte

nsity

O1s

Er4dV2p3

0 5 10 15 20 25 300

0.5

1

1.5

2x 10

5Er2O3-layer-0030_1.PRO

Sputter Time (min)

Inte

nsity

O1sEr4d

V2p3

Oxidation at 700C

6 hr

1 hr

Oxidation only Oxidation and anneal at 700C for 16 hr

XPS depth profile after exposure to Li (Er) at 600C for 100 hr

~100 nm

Er

V-4Cr-4Ti

Yao. 2003

NIFS-FERC

Need for Collaboration with Design People

Requirement to the coating performance depends strongly on the design

System design is necessary to quantify the requirement to the coating

Clever design would make the requirement lenient

New idea of coating will be obtained by collaboration with design

Laminar coating structure, etc.

NIFS-FERC

Meeting Summary for Crack Fraction Allowance for the MHD Coating (Sze Aug.03)

For a single pipe, with a perfect insulating coating, the allowable crack fraction was <10(-7)    

For a real coating, 10(-2) is achievable, while 10(-4) might be achievable. 

If we start with a poor coating, the allowable fraction can be higher, maybe 10(-4), with a higher MHD pressure drop.

There are other ways to increase the allowable crack fraction, such as change the aspect ratio of the channel, change the boundary conditions of the flow channel.

The boundary condition of the flow channel, such as the contact resistance between the fluid and the wall, may have major impact on the crack fraction.   

The change of the designs may have major impact on the crack allowance.

NIFS-FERC

Impact of Sze Summary on the Coating Development in Japan

Experimental examination of the resistance between the (flowing) Li and the wall covered with cracked coatings at high temperature is of high priority. 

The goal of the in-situ healing may be set to increase the resistivity of cracked area from complete conduction by 4 order of magnitude

Increased collaboration between materials and design people in Japan

NIFS-FERC

Design Effort to Reduce Requirement to the Coating

Optimization of channel structure for reducing the requirement to the coating

Coating may be necessary only on limited flat surfacesInsulator ribs may be inserted instead of coated ribs/walls

Other suggestions on laminar coating structure, enhanced heat transfer, etc

Model (c) Model (d) (coated ribs and back wall) ( + coated end ribs)

Model (a) (insulator ribs) Model (b)(+ coated back wall)

5mm 5mm

0.625mm805mm

205mm

0.625mm 1mm

Model (c) Model (d) (coated ribs and back wall) ( + coated end ribs)

Model (a) (insulator ribs) Model (b)(+ coated back wall)

5mm 5mm

0.625mm805mm

205mm

0.625mm 1mm

(unit : kPa/m)insulator /HT-9 1 10-3 10-6 10-9

Ideal case of (c) 339 49.1 2.12 0.49Model (a) 339 60.5 53.4 58.4Model (b) 339 58.7 6.88 1.01Model (c) 339 127 9.95 1.01Model (d) 339 48.1 6.68 1.02

(Hashizume)

NIFS-FERC

Summary

Collaboration by Materials, Blanket and Design people are increasing on V/Li system in Japan

Progress in developing vanadium alloys toward engineering maturity

Enhanced Li technology by IFMIF-KEP

Increased accessibility for the liquid blanket people to ITER-TBR

Collaboration on MHD-coating development by materials and design people

One of the goals of the collaboration is to propose V/Li ITER-TBM

The collaboration is enhancing research for other advanced blanket systems (Flibe --)

The collaboration covering Material, Blanket and Design people in the US will accelerate the progress, and should be enhanced in the framework of JUPITER-II