Japanese Universities’ Perspective on Li/V TBM

NIFS-FERC

Japanese Universities’ Perspective on Li/V TBM

T. Muroga

Fusion Engineering Research Center

National Institute for Fusion Science

NIFS-FERC

Outline of Presentation

Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities

Purpose of Li/V ITER-TBMNeutronics examinationConsideration to the Russian designPresent strategyProgress in Key Technology development for Li/V

With emphasis on MHD coating development in Japan

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Roadmap for Materials/Blanket Development

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 (Mostly JAERI responsibility)

Advanced option (Mostly NIFS/University responsibility)

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Position of Li/V Blanket and Li/V ITER-TBM for Japanese Universities

Li/V blanket is categorized into “advanced system” in contrast to RAFM/water blanket as “reference system”

General plan for Li/V ITER-TBM is to start the test in the midst of the ITER operation phase

However, first-day Li/V TBM will be explored in the following cases

If large technological progress is made, we will reconsider the schedule

If other parties propose first-day TBM, we will support it and make effort for Japanese idea to be incorporated into the proposal

Now Russia is proposing ｆ irst-day Li/V-TBM, we will evaluate the proposal and seek for possibility of cooperation

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Purpose of Li/V ITER-TBM(current consensus in Japanese Universities)

Feasibility of no-Be and natural Li blanket Use of 7Li reaction for enhancing TBR in contrast to R

ussian Be+6Li enriched TBM

Validation of neutronics prediction Technology integration for V-alloy, Li and T

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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

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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

■ Tritium self-sufficiency is feasible

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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-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

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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)

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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

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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

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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

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Is Russian Li-Be-V an Attractive Option?

“No-beryllium” is an attractive potentiality of Li/V system

Does Li-Be-V have alternative merits compensating the demerit of using Be? High TBR? – Excess TBR probably not necessary More space for shielding? – Requirement is system

dependent

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■ Top View

■ Side View

Plasma

100 cm0

Self-cooled T breeder

Radiation shield Vacuum vessel

SC magnet

LCFS

First wall

550°C coolant out

SOL

Thermal shield 450°C coolant in

20°C

7 53

T boundary&

Flibe

217 5

JLF-1(30vol. %)

1 1

&

Carbon 10 ~ 20

JLF-1 + B4C (30 vol.%)

Be (60 vol.%)

■ FFHR-II Original Design ■ Flibe Blanket → 　 V/Be/Li Blanket

10 m

2 m■ First Calculation Using FFHR Torus [MCNP4C+JENDL3.2]

Consideration on FFHR

■ Shielding for SCM is one of the critical issues for FFHR

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Carbon[20cm]

JLF1(70%)

+B4C

(30%)[33cm]

JLF1[5cm]

Filler: 58cm

FWV-4Cr-4Ti

[2cm]

Li-1[10cm]

Li-2[15cm]

Be[5cm]

V-4Cr-4Ti[2cm]

V/Li blanket: 36cm

V alloy-wall[1cm x 2]

Total: 92cm

6Li: 30%=> Local TBR (Full coverage): 1.41

Li/Be/V Blanket into FFHR

１ MW/ ｍ２

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0 20 40 60 80 1001.0

1.1

1.2

1.3

1.4

1.5

Li 10, Be 5, Li 15, Carbon 20, J LF1+B4C 33

Loca

l T

BR

Ratio of Li- 6 (%)

0 2 4 6 8 101.0

1.1

1.2

1.3

1.4

1.5

1.6

Loca

l T

BR

Be thickness (cm)

Li 10, Be <x>, Li 15, Carbon 20, J LF1+B4C 33

30% Li- 6

*

*

10 15 20 25 30 351.0

1.1

1.2

1.3

1.4

1.5

1.6

Loca

l T

BR

Thickness of 2nd Li layer (cm)

Li 10, Be 5, Li <X>, Carbon 20, J LF1+B4C 33

30% Li- 6

*

*: Standard parameter shown in the first page

6Li enrichment

Thickness of Be Thickness of 2nd Li layer

(Changing carbon reflector to JLF-1, TBR: 1.41 =>1.41 (same). Usage of JLF-1 is more attractive for shielding ability)

Parametric Survey

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Results

V Layer (cm)Li Layer (cm)

625

625

620

615

412

Be Layer (cm)

5 5 10 10 10

RAFM shield layer (cm)

58 63 63 68 73

TBR (full coverage)

1.41 1.41 1.54 1.50 1.50

Neutron flux at SCM (1010n/m2s)

6.3 3.1 2.2 1.3 0.78 4.3(FFHR-II)

0.71(Target)

■ Neutron flux at SCM could be reduced to the target level

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Present Position to Russian TBM Design

Li-Be-V system may have a merit of enhancing shielding for SCM

Could be considered as attractive if this degree of enhancement is crucial for protecting SCM

There may be other methods to protect SCMProbably, this potential merit will not deserve

abandoning the merit of no-BePresent philosophy is to cooperate with Russia

and seek for opportunity of testing no-Be TBM

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Progress in Key Technology development for Li/V system in Japan

V-alloy development Production of purified large ingots

– feasibility of recycling Manufacturing technology (tube, welding--) Radiation effects of weld joints

Technology development in relation to IFMIF-KEP Li loop technology Basic study for tritium recovery with Y

MHD coating development (JUPITER-II and domestic activity)

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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 was explored in the US

Ca++

M2Ox

O2-

Ov

Mx+

MLi

V-alloy Li(M)

Ca++

M2Ox

O2-

Ov

Mx+

MLi

V-alloy Li(M)

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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 unstable in pure Li at high temperature, continuous supply of oxygen is necessary to maintain the coating

Once the oxygen is exhausted in V-alloy, the coating start to dissolve

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

Solubility of Er (<<1%) is much lower than CaO

10m/y10m/y

CaO

Er2O3

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

Er2O3 layer was formed on V-4Cr-4Ti by oxidation, anneal and exposure to Li (0.15 wt% Er) at 600C

The coating was stable to 750 hrs, also 700C 100hrs

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

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Oxygen Supply Mechanism

In the oxidized and annealed condition, oxygen is stored as Ti-O precipitates oriented to <100> directions

The stability of the precipitates depends on the oxygen level During Li exposure, the precipitates dissolve, supplying oxygen

into matrix

As-received (annealing 1272K, 2h)

Oxidation (973K, 1h) Oxidation (973K, 1h) + Annealing (973K, 16h)In-situ coating condition

NIFS-HEAT-2 (V-4Cr-4Ti)

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Coating ResistivityResistivity increases by ~12

order of magnitude by formation of Er2O3 layer

Analysis of crack allowance limit (by Sze) suggested 10(-4)~10(-6) lower crack limit than the practical coating defect density.

Goal of the in-situ healing may be set to increase the resistivity of cracked area from complete conduction by 4~6 orders of magnitude – Seems to be feasible –Need further research

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MHD Coating in JUPITER-II

New Candidates Found by Bulk Exposure Tests (Eu2O3, Y2O3, AlN)

Coating Development and Characterization(Eu2O3, Y2O3, AlN, in progress)

In-situ Coating with Er2O3(feasibility demonstrated)

Two layer coating development

Crack allowance estimate

Resistivity measurements in LiCompatibility of the first layer material

Proposal of the coating system TBM design

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Summary

Japanese Universities have an interest in participating in Li/V ITER-TBM

General philosophy is to plan to start the test in the later stage of ITER operation

However, collaboration with Russia (and other potential countries) for first-day Li/V TBM will also be explored

Key technology development is being enhanced by domestic program, IFMIF program and JUPITER-II

The MHD coating is making significant progress. The achievements in JUPITER-II will be applied to designing TBM

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(at RTNS-II. LLNL. 1988)(at RTNS-II. LLNL. 1988)

Neutron Irradiation Effects on SCM