Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan M. Enoeda, Hisashi Tanigawa, T. Hirose, S. Suzuki, K. Ezato, Y. Seki, A. Yoshikawa,

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan

M. Enoeda, Hisashi Tanigawa, T. Hirose, S. Suzuki,

K. Ezato, Y. Seki, A. Yoshikawa, D. Tsuru,

K. Ochiai, C. Konno,

Y. Kawamura, T. Yamanishi,

T. Hoshino, M. Nakamichi,

Hiroyasu Tanigawa, M. Akiba

JAPAN ATOMIC ENERGY AGENCY

16th International Workshop onCERAMIC BREEDER BLANKET INTERACTIONS

Portland, USA, 11-16, September, 2011

1

Contents1. Importance of Water Cooled Ceramic Breeder (WCCB) Test

Blanket and assumed schedule of WCCB TBM R&D2. Domestic cooperation and R&D flow chart3. Module fabrication technology development - First Wall + Side Wall Assembly, Back Wall Partial Mockup4. Advanced breeder and multiplier pebble development for

DEMO - Increased chemical stability and soundness in high

temperature5. Advanced tritium recovery technology for blanket system6. Neutronics engineering - Verification of tritium recovery rate of Li2TiO3 pebble bed with

DT neutron in FNS

2

ITER Test Blanket Module Program- ITER Test Blanket Module (TBM) Program is to test essential functions of DEMO Blanket in

the real fusion environment with scalable module.- ITER TBM Program is one of the most important development steps.

VV

Plasma

Water Loop

Generat

or -Production of fusion fuel tritium

- Extraction of energy

Fusion Council of Japan stated that ITER TBM Test Program is one of the most important development step. (Aug. 2000)Japan has a position to - act as a Port Master and a TBM Leader to test the WCCB TBM. - participate as a Partner in HCCB/HCSB, LiPb-based TBMs and Li-based TBM.

3

1700mm

600mm500mm

Water Cooled Ceramic Breeder(WCCB) TBM proposed by Japan

Test Blanket Module （ TBM ）

Structure of RAFM (F82H)

Neutron MultiplierPebble Bed (Be)

Tritium BreederPebble Bed (Li2TiO3)

ITER Cross SectionProvisional Port Allocation

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3118 19 20 21 22 23 24 25 26 27 28 29 30

1906 07 08 09 10 11 12 13 14 15 16 17 18ITER Construction ITER Operation

PrS

R

FD

RTentative Milestones,

under discussion

JA PY

AD

1/1 BreederBox MUTesting

1/1 FWHHF Test

TBMFabric. Tech. Eng R&D1/1 TBM

Box HHF Test

1/1 FWMockuoFabricat.

1/1 BreederBox Mock-

up Fabr.

1/1 SWMockuo

Fabrication

Advanced Breeder, Multiplier Development

Tritium Recovery Process Development

Nuclear Performance Validation Tests

Structural Material Validation/ Database

PD

R

CD

R

TB

M #1 S

afy R

eport

Installation in ITER

BW/Box Assembling

1/1 BWFab.

1/1 BWPlate

Fabrication

4

Material Proc., Parts Manufact.

Ancillary System Manufact.

Module Manufact., Assembling

TBM #1 Fabrication

TB

M

Delivery

1/1 BWFabr.Tests

Assemblingof FW/SW

Provisional WCCB TBM Delivery Schedule

20 21

TB

M P

A1.66m

.484m

0.6m

WCCB TBM

32 33

TBMPPTS

Testing

- Large Size Mockup Fabr.- LSMU Internal Pressure Test- Detailed Fabr. Validation- RH Demonstration

- Water Ingress Safety Validation Tests

Prototype #2 TBM Fabrication- Structure Endurance Tests (Errosion-corrosion, Be pebble bed)- Ancil. System Testing- HHF Tests/Heating Tests, function tests of prototype

Large Size Mockupo Testing

Cooperation of Solid Breeder Blanket DevelopmentSystem IntegrationOut-pile R&D, Module fabrication technology, Thermal hydraulic research, - Blanket Engineering Lab. (JAEA)

Material Development, Fabrication Tech.- Fusion Mater. Development Gr., Radiation Effects and Analyses Gr. (JAEA)- Profs Kohyama, Kimura (Kyoto Univ.)- Prof. Serizawa (Osaka Univ)

TritiumExtraction System

DEMO ReactorSystem

Blanket Module

DEMO Reactor Design- Reactor System Gr. (JAEA)- Prof. Ogawa (the Univ. Tokyo)- Dr. Okano (CREIPI)

In-pile R&D, Breeder/multiplier development- Blanket Irradiation Tec. Gr. (JAEA)

Neutronics / Tritium Production Tests with 14MeV neutrons- Fusion Neutronics Gr. (JAEA) Cooling System

5

Tritium Recovery SystemDevelopment, Tritium Control- Tritium Tech. Gr. (JAEA)- Profs Fukada, Nishikawa (Kyushu Univ.)- Profs Tanaka, Terai 　 (the Univ. Tokyo) - Prof. Hino (Hokkaido Univ.)- Profs Okuno, Oya (Shizuoka Univ.)

Plasma Facing Materials- Prof. Hino (Hokkaido Univ.)- Prof. Tanabe (Kyushu Univ.)- Prof. Ueda (Osaka Univ)

- 6 -

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R&D Items and Development Flow of Blanket

Element Tech.Development

Engineering Scale Evaluation

Pebble Bed Fabrication Technol.

Thermo-Hydraulics of High Heat Flux Cooling

Endurance and PropertiesEvaluation of Pebble Bed

In-pile Irradiation Test Technol.

Thermo-hydraulicsof Coolant Network

Partial Module Irradiation Tests, PIE Facility

Integrated Functional Testby Large Scale Mockup

Thermo-Mechanichs ofPebble and ContainerIncluding Compatibility

Therm. Mech, Characteristicsof Blanket Structure

In-pile IrradiationPerformance Evaluation

Breeder / Multiplier Pebble Fabrication Tech.

First Wall, BoxFabrication Technol.

Neutronics ofBlanket SystemTritium Production

Neutronics Tests

Pebble Bed Thermo-Mechanichs

Chemical Compatibilityof Breed./ Mult. Pebble Beds

TB

M F

abrication & Installa

tion in

ITE

R

Thermo-Hydro Dynamics of Coolant

Fusion Neutronics Performance Evaluation

Soundness, Safety, Chemical Compatibility

Mass Fabrication Technol.for Breed. and Multipl.

Pebble FabricationTechnology

Blanket Safety Basic Tests, Coolant Corrosion and Permeation

SafetyDemonstration

Tritium Permeation and Barrier

Structure Corrosionby Coolant Flow

Advanced Materials FabricationBreeder Reprocessing Technol.

Tritium Recovery ElementaryTechnology Development

Tritium RecoverySystem Demonstration

Tritium Recovery System EvaluationBlanket Tritium

Recovery Technology

Development of Integrated Simulation Code for Blanket Tritium Behavior

Blanket System TPR Evaluation

Blanket TPR Confirmationby Simulated Mocups

Irradiation Mockup Design

20102005 2014

FW/SW Assembling Tech.

Simulation Code Development and Experimental Verification

Corrosion Rate Evaluation

FW/ SW Assembly Mockup Fabrication, BW Fabrication Tech.

2009TBM Large Mockup

Function Tests

Development of High Temperature, Long Life Breeder / Multiplier Pebble Materials

Fusion Neutron Tritium Recovery Experiment

Advanced Tritium Recovery Process Developmentt



Progress of Fabrication Technology Development

2006First Wall

2009FW/SW

Assembly

2007Pebble BedContainer

2008Side Wall

2012Large Scale

TBM Mockup

2010-2011Back Wall

2021Installation

to ITER

2015TBM

Fabrication start

7

Real Scale FW Mockup and Heat Flux Test

176mm

25mm

Cross-section of First Wall

8mm

Coolant Inlet

CoolantOutlet

The mockup was high heat flux tested with a heat load of 0.5 MW/m2, 30sec for 80 cycles.

Neither hot spots nor thermal degradation were observed.

Expected heat removal performance was demonstrated.

HHF tests in DATS facilityPeak Heat Flux: 0.5 MW/m2

Beam Duration: 30 sWater Temperature: 300 oCWater Pressure: 15 MPaFlow Velocity: 2 m/s

8

10mm x 1450mmL drilling

1.55

m（

C. c

hann

el 1

.45m

）

0.4m

1 mm accuracy was achieved at the end of 1.45 m depth drilled holes.

Fabrication of Real Scale Side Wall

WCSB TBM

- Real scale Side Wall was fabricated. Cooling channels were machined by drilling.

-10 mm x 1450 mm L cooling channels were formed within 1 mm accuracy at the end of the drilled holes. 1700 mm L is available.

Fabricated Side Wall

17701770

232

232

Welding test of 1/1 scale FW/SW using thick plates

9

Fabrication of FW and SW Assembly Mockup

1/1 FW mockup (with cooling channels) and 1/1 SW mockups (with cooling channels) were assembled by EB welding.

Distortion on FW side is less than 1mm, and distortion in hight is less than 3mm. Welding soundness was inspected by UT.

Welding technique and procedure, welding support were confirmed.

FW

SW1.5 m

0.4 m

1489.2, 1488

1489.5, 1489

417.6417.7

417.7417.6

417.8418.3

417.7417.6

1490

417

.5

FW-SW Assembly Mockup

EB Welding Support

EB Welding

10

FW/SW Assembly Mockup after Assembling process on the welding support

Fabrication of Back Wall Partial Mockup by Normal Steel

232 mm

90 mm

60 mmEB Welding

Cooling Channel

Coolant Header

A partial mockup of the back wall of the Test Blanket Module, which has major structure feature of coolant channels, header and a shear key, was fabricated by using conventional steel, by EB welding of the shear key.

The fabrication technique and procedure for the back wall were confirmed.

11

Experimental Apparatus for Flow Assistred Corrosion of Structural Material by High Temperature and Pressure Water Flow

Flow Assisted Corrosion Experimental Apparatus by High Pressure and Temperature Water

- A disc of a test material is rotated in an autoclave of high pressure and temperature water.

- Test specimen of 100mm diameter disk is rotated up to 2000 rpm. Equivalent superficial velosity at the edge of the disk is 10 m/s

- Water condition is available up to 340 oC 15MPa.

- Flow parameter is estimated by comparison between Flow Visualization Experiments and numerical simulation.

12Flow Visualization Experiments

Numerical Simulation

Pressure Cylinder

Rotating Test Piece

Hydraulic Analysis of the Rotation Disc Test Section

Hydraulic Analysis of the Rotation Disc Test Section

It was found that the shear stress is higher than 740 Pa where corrosion layer was peeled off.

3m/s (300℃,15MPa) Corrosion layer was peeled off.

0.7

8.7

4.3

( m / s )

Hydraulic Simulation

Experiments of Flow Assisted Corrosion by High Pressure and Temperature Water

Experiments of Flow Assisted Corrosion by High Pressure and Temperature Water

Vel

ocity

[m/s

]

Distance from the center [mm]

Observed Flow Velocity Distribution (7mm above the disk)Consistency with the simulation was confirmed.

Visualization Experiment

1 m/s No Flow3 m/s

5.2 m/s

Trace Particle

Evaluation of Corrosion of Structural Material by High Temperature and Pressure Water Flow

Flow directionShear Stress on Wall

Header

Blanch Channel

ww

223

213

212

Re=9.4× 10 5 (5.0 m/s)

(Pa)

(m/s)

平均流速流線(m/s)

Flow Line

Re=3.6×105

(4.4 m/s)

Hydraulic analysis of coolant flow in Side Wall headers and channels.

By Hydraulic analysis, it was clarified that shear stress of more than 740 Pa appeared near the part where coolant split into blanch channel from the header.

13

Development of Advanced Neutron Multiplier PebbleBeryllide synthesis process -Plasma sintering -

Raw material powder

Punch and Die unit

Additions of : 1) Pressure 2) Current (for activation and heating)

The plasma sintering direct sintering from material powder - Enhancing powder particle activeness for sintering - Reducing high temperature exposure

Plasma Sintering Conditions

Raw material : Be & Ti powder

Powder purity : >99wt%

Powder size : <50µm

Sintering time : 20min

Pressure : 50MPa

Temperature : 1273K

XRD profiles and EPMA analysis for clarification of sintering temp.

[at 1273K]

Be Be2Ti

Be17Ti2Be12Ti(Beryllides: Be12Ti, Be17Ti2 and Be2Ti)

Be : 2%Beryllides: 98%

(1) It was shown that spark plasma sintering is applicable for synthesis of Be12Ti.

(2) By the experiments of sintering temperature effect on Be12Ti synthesis, It was clarified that sintering in 1273 K achieved largest fraction of Be12Ti.

Blackened by reduction

No change

White sample(Li2TiO3)

Li2TiO3

with added Li

The color of Li2TiO3 changed from white to black in a hydrogen atmosphere at high temperatures. This color-change corresponds to reduction of Li2TiO3.

Li2TiO3

without added Li

After heating at 1273K for 10h in 1%H2-He

Development of Advanced Tritium Breeder Durable to Reduction in Hydrogen Atomoshpere

Development of Li2TiO3 with excess Li to improve the resistance of reduction at high temperatures

In the case of Li2TiO3 with added Li, the color did not change, indicating that this sample was not reduced in the hydrogen atmosphere. Chemical Stability

Gel

Water

GelationLiOH•H2O and H2TiO3

Gel-spheres

Li2TiO3 with excess Li

Diameter 1.18mm

Sphericity 1.04

Density 89%T.D

Grain size 2 - 10μm

Development of Pebble Fabrication Technology of Li2TiO3 with excess composition of Li

Trial fabrication of pebbles of Li2TiO3 with excess Li composition by sol gel method

Raw material Granulation SinteringSlurryLi2TiO3 with excess Li

Pebbles of Li2TiO3 with excess Li was granulated by sol gel method from slurry.

Li2TiO3 with excess Li was synthesized from mixed LiOH•H2O and H2TiO3

Sol-gel method is applicable in pebble fabrication of Li2TiO3 with excess amount of Li.

Advanced Tritium Recovery Technology Development - Principle -

Study on Application of Electrochemical Hydrogen Pump (Ceramic Proton Conductor) as a continuously operating HT and HTO recovery process

Driving force of tritium extraction•Pressure difference•Electric potential difference

Experimental validation of Tritium transport property to evaluate applicability, using tritium gasExperimental conditions

Sweep Gas In (H2, HT, H2O, HTO/He

Sweep Gas Out (He/O2

Principle of Electrochemical Hydrogen Pump

Voltage

Advanced Tritium Recovery Technology Development - Result of transport property measurement -

Tritium was extracted by applying voltage. DF and recovery ratio were 1.5 and 0.4.

- Principle was demonstrated.- One-through Decontamination Factor and T recovery rate

were 1.5 and 0.4 by a single tube with 0.2 l/min He + HT gas flow.

- Scale up is the further issue for adopting this principle.

PREVIOUS EXPERIMENT (Offline) Tritium Recovery Experiment from Li Ceramic Breeding Material

Irradiated with DT NeutronsWe have conducted a tritium recovery experiment for solid breeding blanket with DT neutrons for the first time in the world.

Tritium recovery measurement

DT neutron irradiation arrangement

The experiment shows the tritium recovery ratio for the mock-up is 1.05 0.08 at 873 K, which indicates that the design of Japanese solid breeder blanket promises a good prospect of tritium recovery up 873 K.

Tritium production measurementPebble dissolution method with a weak acid (HCl)

JAEA/FNSDT neutron source

Beryllium bulk

Breeding material Container

Gas cylinderHe gas (H2 1.04%)

MFC

100sccmTC

CoolantAir in

Purge Air InLi2TiO3 pebble67g 6-Li: 7.5%)

Heater

Up to 873 K

Heater

773K

CuO Bed100.0g

CoolantAir out

Compressor

Bubbler 1 Bubbler 2

Silica Gel

(124.3g)(water : 100cc/bottle)

MFC

100sccm

Pump(for purge)

Purge Airout

Experimental Setup for On-line Measurement of DT Neutron Production Experiment

• The neutron intensity was about 1.5 x 1011 neutron/sec.

• The sweep gas He + H2 1.04% flow rate kept 100 standard cm3/min

• After the irradiation, water vapor fraction in the sweep gas line was measured with a dew-point meter. It was an order of 1000 ppm.

• After the run, 1 cm3 water in each trap bottle was mixed into a liquid scintillator and measured with a liquid scintillation counter (LSC), which was calibrated with a standard HTO (50 Bq/cc) sample within 2 % accuracy.

Schematic view of the DT Neutron Tritium Recovery Online Experiment

Li2TiO3 Pebble Bed (6Li 7.5%)

Heater

Former Trap Bottles for HTO (H2O

100cm3/bottle)

Exhaust Gas

Latter Trap Bottles forHTO (H2O 100cm3/bottle)

Trap bottle change by remote handling

MFC 100cm3/min

DT Neutron Source(1.5 x 1011/n/sec)

Gas Cylinder(He+H2 1.04%)

CuO Bed (100g)For Oxidization of HTO

Concrete Wall (2m thick)

Be Block Assembly JAEA/FNSDT neutron source

Beryllium bulk

Breeding material Container

DT neutron irradiation experimentBreeder capsule arrangement

Experimental Setup for On-line Measurement of DT Neutron Production Experiment

• The neutron intensity was about 1.5 x 1011 neutron/sec. • The sweep gas He + H2 1.04% flow rate kept 100 standard cm3/min

• After the irradiation, water vapor fraction in the sweep gas line was measured with a dew-point meter. It was an order of 1000 ppm.

• After the run, 1 cm3 water in each trap bottle was mixed into a liquid scintillator and measured with a liquid scintillation counter (LSC), which was calibrated with a standard HTO (50 Bq/cc) sample within 2 % accuracy.

Schematic view of the DT Neutron Tritium Recovery Online Experiment

Li2TiO3 Pebble Bed (6Li 7.5%)

Heater Concrete Wall (2m

thick)

Be Block AssemblyFormer Trap Bottles forHTO (H2O

100cm3/bottle)

Heater (773K)

Exhaust Gas

Latter Trap Bottles forHTO (H2O

100cm3/bottle)

Trap bottle change by remote handling

MFC 100cm3/min

DT Neutron Source(1.5 x 1011/n/sec)

Gas Cylinder(He+H2 1.04%)

CuO Bed (100g)For Oxidization of HTO

DT neutron irradiation experimentOn-line tritium measurement setup

Water bottles for tritium recovery

Breeder Capsule

Result

In order to deduce the tritium recovery ratio, we adopted our previously measured TPR data, 7.46 x 10-14 Bq/g/DT neutron with experiment error of 8 %. As a result, the present tritium recovery ratio was 0.96. It is indicated that the tritium recovery of Japanese TBM has a good prospective at 573 K.

It is also shown from the result that the total recovered HTO was significantly larger than the total recovered HT and its ratio is about 0.9. It was considered that larger HTO recovery was due to larger water vapor (1000 ppm) in the sweep gas line. It seems that the tritium produced in the Li2TiO3 pebbles easily reacts on water vapor rather than H2 in the sweep gas at such low temperature. In future, we will conduct an additional experiment with a cold trap system (e.g. dry ice and/or molecular sieve) in the sweep gas line.

From the measurement, HT release showed delay compared with HTO release.

The horizontal axis is elapsed time of tritium recovery and the vertical one is fraction of recovered tritium radioactivity

The total tritium radioactivity was about 8.66 kBq. The number of DT neutron irradiation was 1.74 x 1015. Thus the TRR was 7.11 x 10-14 Bq/g/DT neutron (normalized in Li2TiO3 weight and neutron flux)

-1 0 1 2 3 4 5 6 7 80.0

0.1

0.2

0.3

Total T = 8.66 kBq

Fra

ctio

n (

pe

r to

tal t

ritiu

m B

q)

Irradiation time (hour)

HTO HT

Temperature 573 K

DT neutronirradiation TRR/TPR =0.96

HT

HTO

Conclusions

1. In the fabrication technology development of WCCB TBM, real scale F82H First Wall and Side Walls were successfully assembled with enough small distortion. Also, partial mockup of the Back Wall was fabricated to confirm the fabrication route.

2. In the advanced multiplier and breeder pebble development for DEMO blanket, Be12Ti rod , pebble of Li2TiO3 with excess Li composition, which have increased chemical stability in high temperature, were clarified.

3. In Advanced Tritium Recovery Technology Development, the principle of Electrochemical Hydrogen Pump was demonstrated and basic tritium recovery property was clarified. It was observed that scale up is a further issue.

4. Tritium Production and On-line Recovery Experiment by DT neutron irradiation at JAEA-FNS showed that the tritium recovery ratio was 0.96 0.08 compared to the evaluation by neutronics experiments. It was expected that the tritium recovery data is used for verification of Tritium production performance of TBM.

24

Development of the Water Cooled Ceramic Breeder Test Blanket Module in Japan M. Enoeda, Hisashi Tanigawa, T. Hirose, S. Suzuki, K. Ezato, Y. Seki, A. Yoshikawa,

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