Recent Progress in Fusion Technologies under the BA DEMO-R&D in Phase1 in Japan

(FTR/3-2Ra)T. Yamanishi 1), T. Nishitani 1), H. Tanigawa 1), T. Nozawa 1), M.

Nakamichi 1), T. Hoshino 1), K. Hayashi 1), M. Araki 1), T. Hino 2), and S. Clement Lorenzo 3)

1): Japan Atomic Energy Agency, Tokai, Ibaraki, JAPAN, 2): Hokkaido University, Sapporo, Hokkaido, JAPAN,

3): Fusion for Energy, Barcelona, Spain

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Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons

(FTR/3-2Rb)K. Ochiai 1), T. Hoshino 1), Y. Kawamura 1), Y. Iwai 1), K. Kobayashi 1), K. Kondo 1), M. Nakamichi 1), C. Konno 1), T. Nishitani 1), M. Akiba 1)

1): Japan Atomic Energy Agency, Tokai, Ibaraki, JAPAN, 23rd IAEA Fusion Energy Conference, 11-16 October 2010, Daejon, Korea Rep. of

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Recent Progress in Fusion Technologies under the BA DEMO-R&D

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Satellite Tokamak Project IFMIF-

EVEDA Project

DEMO Design and R&D CenterDEMO Design and R&D CenterComputational Simulation CenterComputational Simulation Center

IFERC Project

1. DEMO R&D Subjects and schedule of JA2007

2008 2009 2010 2011 2012 2013 2014 2015 2016 2017

Phase Phase 1Phase 2Phase 2Phase 2Phase 2----1111

Structure materials(Reduced activation ferritic/martensitic )

SiCf/SiC compositesTritium technologyNeutron multiplier(Beryllium compounds)Tritium breeder (Lithium compounds)

Preparation of material and test equipments

Large scale melting. Development of joining andinspection methods

irradiation effects on mechanical properties

Mechanical properties

Preparation of fabrication and test equipment

Test melting and characterization

Irradiation effect on physical properties

Preparation of fabrication and test equipment

Test fabrication and characterization

Preparation of Tritium and RI handling facility

Tritium Accountancy, Basic Tritium Safety data,Tritium Durability

Test fabrication and characterization

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Tritium handlingroom*

Liq. waste tanks

Be handlingFacility(independent HVAC)

Other RI handling rooms**

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*: In this area, tritium and some RI species can be handled (Group A, B, and D). **: In this area, a hot cell and a temperature controlled section are installed. The RI species of Group B, C, and D, and a small amount of tritium are handled.

2. RI handling equipment at Rokkasho siteGroup A B C DRI Tritium Cer-

amics Steel Other metals

Typ-ical RI H-3 P-32

Fe-59, Cr-51Ta-182

W-188,Re-188

Usage GBq/day

3700

370GBq/hood

MBq/day

100

MBq/day

61 95046

MBq/day

4444

Sto-rage

7400 GBq

500 MBq

915 MBq14 GBq690 MBq

220 MBq220 MBqGlove box and

detiritation systems

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2. RI handling equipment at Rokkasho site

3．．．． Recent Progress in Phase1 in Japan3.1 SiCf/SiC CompositesFailure behavior of SiC/SiC composites by various failure modes such as tensile, compressive, and shear modes was studied.Basic data for Radiation-induced conductivity (RIC) and Radiation-induced electrical degradation (RIED) irradiation at room-temperature in air was obtained(See Figure) . 10101010

-8-8-8-8

10101010

-7-7-7-7

10101010

-6-6-6-6

0.00.00.00.0 0.20.20.20.2 0.40.40.40.4 0.60.60.60.6 0.80.80.80.8 1.01.01.01.0 1.21.21.21.2 1.41.41.41.4

during irr.during irr.during irr.during irr.

non-irr.non-irr.non-irr.non-irr.

Dose (MGy)Dose (MGy)Dose (MGy)Dose (MGy)

Conductivity (S/m)

Conductivity (S/m)

Conductivity (S/m)

Conductivity (S/m)

CVD-SiC, gamma-ray irradiation at R.T.CVD-SiC, gamma-ray irradiation at R.T.CVD-SiC, gamma-ray irradiation at R.T.CVD-SiC, gamma-ray irradiation at R.T.

GB = continuously detritiation, negligible tritium permeation.Permation and leakage of the devices in hoods and detritiation systemsDistance between workers and RI = 50 cmConcentration of tritium in room = ~5x10-4 Bq/cm3(~1/1000 of regulation). Concentration of tritium at a stack = ~4x10-4 Bq/cm3(~1/10 of regulation). 0.783 mSv/week for workers (~0.8 of regulation); 1.2 mSv/3 months at a boundary of the radiation controlled cold area (~0.9 of regulation), and 43 μSv/3 months at a site boundary (~0.2 of regulation).

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3．．．． Recent Progress in Phase1 in Japan3.2 Structure MaterialsVarious properties of the plates of the F82H steel melted and forged was studied to optimize the fabrication technology of F82H. Assessment on the specific welding technologies for DEMO was carried out to obtain basic data on the joining of F82H.

3.3 Advanced Neutron MultiplierPreliminary tests using mixed beryllium and titanium (Be, Ti) powder by the plasma sintering method.The formation of Be-Ti intermetallics (Be12Ti, Be17Ti2 and Be2Ti ) was observed for the sintering temperature of 1073~1273K.. Elemental Be and Ti decreased with Increase in sintering temperature. 3.4 Tritium Technology

Tritium analysis technology = imaging plate method; Basic tritium safety data= behaviors in tungsten , stainless steel etc.Tritium durability test = organic compounds durability tests by gamma and beta ray irradiation.

Iconductivity of the Nafion membrane.(□):0kGy, (●): 150kGy, (△):250kGy, (■): 500kGy, (○):1000kGy, (▲):1500kGy

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3.5 Advanced Tritium BreedersA rotating granulation method for making Li2TiO3 pebbles from powder of 5μm sintered at 1200°C.Succeful production of pebbles with a sphericity of 1.04 (See figure).Reprocessing tests: Solvent = Mixture of peroxide hydrogen (H2O2) and nitric acid (HNO3) ; Dissolution rates of Li from lithium ceramic powder=90%.Summary (1)The RI handling equipment has been constructed at Rokkasho site as the first and quite unique facility in Japan, where tritium(3.7 TBq/day), beta and gamma RI species, and beryllium(Be) can simultaneously be used. The tritium, Be, and other RI handling equipment, such as hoods have been designed, made, and installed. (2) As the BA activites, the studies on strucutr materials (F82H) ; SiCf/SiC: advanced tritium breeders(production of pebbles of Li2TiO3) , neutron multiplier(production of beryllides from Be and Ti); and tritium (durability) are carried out.

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R&D on blanket technology at JAEADEMO blanket

R&D on structural materials (SiC, structure materials)

R&D on neutron multiplier (pebble fabrication, characteristics)

R&D on tritium breeder (Li compound fabrication, characteristics)R&D on tritium technology (analysis, material interaction, durability)

BA activities

Integrated test for tritium production (multiplier, breeder, tritium, neutron irradiation)

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1. Background and Objectives• It is an essential issue to verify the nuclear performance concerning tritium

breeding in fusion blanket including ITER-TBM. • Japan and EU have progressed to verify the tritium production ratio in

simulated blanket assemblies with DT neutron irradiation. FusionNeutronics Source (FNS) facility in JAEA has verified tritium production rate of Li2TiO3/Beryllium type.

• However, no DT neutron irradiation experiment for the tritium recovery properties of the solid blanket has been conducted and the tritium recovery ratio is one of urgent technical issues for the development of the fusion breeding blanket system.

• Based on the above DT neutron irradiation technology, JAEA-FNS has conducted the first tritium recovery experiment for the solid breeding blanket as a next step.

Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons

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The Tritium recovery ratio of the ceramic breeding blanket was measured with the 14 MeV DT neutrons for the first time in the world.

DT neutron irradiation arrangement (JAEA-FNS)

JAEA/FNSDT neutronsource

Beryllium cylindrical

bulkBreeding material Container

Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons

765 （48 x 48 tag）

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Inlet (coolant air)

Inlet (sweep gas)

Thermocouple

Breeding material outlet(sweep gas)

Breeding materialφ1-mmLi2TiO3 pebble67.0g (Li-6: 7.5%)

Insulator (SiO2 and Al2O3)

Heater wire (2 t)Cr 23~26%, Al 4~6%, Fe balance

198.8

Overview of container for irradiation

Set up of breeding material pebble, heater and insulator in container

2. Experimental setup

203 mm depth point from surface

Φ630 x 457.2 mm3

Number of DT neutrons = 10 15

measured by 3.5 Mev alpha with 2% error

Gas cylinderHe gas (H2 1.04%)

MFC

100sccmTC

CoolantAir in

Purge Air InLi2TiO3 pebble67g (Li-6: 7.5%)

HeaterUp to 873 K

Heater773K

CuO Bed100.0g

CoolantAir out

Compressor

Bubbler 1 Bubbler 2 Silica Gel(124.3g)(water : 100cc/bottle)

MFC

100sccm

Pump(for purge)

Purge Airout

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Tritium recovery measurement

The tritium recovery ratio of 1.05 ±±±± 0.08 was obtained at 873 K, which indicates that the design of Japanese solid breeder blanket promises a good prospect of tritium recovery at 873 K.

Tritium production measurement

Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons

DT neutron radiation

Eject pebble

Dissolution of pebblein liquid HCl (0.2N)

Li2TiO3 (T)+ HCl

Skimming supernatant (1cc)mixed with scintilator

2LiCl + H2O(T) + TiO2For 3 days

1 day

LSCMeasurement of tritium activity

3. Measurement of amount of tritium produced and recovered

LSC

423 K

1.5 m from inlet to the first bubbler

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4. Conclusion4. Conclusion4. Conclusion4. Conclusion

We have performed the first tritium recover experiment with DT neutrons in the world at FNS in JAEA. From the measured tritium recovery ratio, a good prospective is obtained for the solid breeder blanket at 873 K. In order to progress the investigation of tritium recovery property for the ITER test blanket module and DEMO blanket designs, we are going to examine the dependency of the temperature and sweep gas with DT neutron source as the next step.

Tritium Recovery Experiment from Li Ceramic Breeding Material Irradiated with DT Neutrons