japanese universities’ perspective on li/v tbm
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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
NIFS-FERC
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)
NIFS-FERC
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
NIFS-FERC
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
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
■ Tritium self-sufficiency is feasible
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-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
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-FERC
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
NIFS-FERC
■ 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
NIFS-FERC
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
NIFS-FERC
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
NIFS-FERC
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
NIFS-FERC
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
NIFS-FERC
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)
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 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)
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 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
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 (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
NIFS-FERC
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)
NIFS-FERC
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
NIFS-FERC
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
NIFS-FERC
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
NIFS-FERC
(at RTNS-II. LLNL. 1988)(at RTNS-II. LLNL. 1988)
Neutron Irradiation Effects on SCM
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