Japan considerations on design and qualification of PFC's for near term machines (ITER)

1. Japan considerations on design and qualification of PFC's for near term machines (ITER) Satoshi Suzuki1, Satoshi Konishi2 1 Japan Atomic Energy Agency 2 Kyoto University

2. Contents • Development / provisional procurement activity of the ITER divertor outer vertical target • Small divertor mock-ups • Non-destructive examination by using infrared thermography • Development of the first wall of the ITER test blanket module (TBM) • Provisional high heat flux test of the full-poloidal length TBM first wall mock-up • Development of the DEMO divertor • Small divertor mock-ups made of ferritic steel with tungsten armor

3. Contents • Development / provisional procurement activity of the ITER divertor outer vertical target • Small divertor mock-ups • Non-destructive examination by using infrared thermography • Development of the first wall of the ITER test blanket module (TBM) • Provisional high heat flux test of the full-poloidal length TBM first wall mock-up • Development of the DEMO divertor • Small divertor mock-ups made of ferritic steel with tungsten armor

4. Inner vertical target (EU) Outer vertical target (JA) Dome(RF) Cassette body (EU) ITER Plasma Facing Components • Among the divertor high heat flux components, JAEA is going to procure all of the outer vertical target • 22 plasma facing units / 1 cassette • 54 cassettes + 6 spares --> 1320 plasma facing units ITER divertor Cross sectional view of ITER

5. Outer vertical target • Design heat flux • Tungsten part : 5MW/m2 • CFC (Carbon Fiber Composite) part : 10 - 20MW/m2 • Coolant condition • Water, 4MPa, 100oC • Swirl flow by twisted tape (CFC part) • Bonding of armor materials • Braze • HIP (Hot Isostatic Pressing) • Prior to the procurement of the vertical target components, the manufacturing process and the thermal performance of the components should be validated/demonstrated by using small mock-ups. ("Prequalification" activity) Outer vertical target (1 cassette) with "Monoblock" geometry

6. Optimization of the manufacturing process by using small mock-ups • Armor : CFC (CX-2002U) • Tube : CuCrZr • Bonding : Brazing (Tbraze=980oC) • Ni-Cu-Mn • Ti-Cu-Ni • Metalizing of CFC • Present : Ti-Cu (5 - 10%-Ti) • Accuracy of machining of CFC hole • +/- 5 microns to the target value • High heat flux test has been performed to check the soundness of the braze interface and to demonstrate the thermal performance. A small vertical target mock-up simulating the CFC part of the vertical target

7. CFC at 5 MW/m2 (EB heating) Tungsten ~500mm ITER vertical target Qualification Prototype High heat flux test of the small mock-upsand fabrication of a qualification prototype • Braze filler : Ni-Cu-Mn (980oC x 0.5h, followed by Ar gas quench) • Metalizing : Present (Ti-Cu, 7.5 - 10%-Ti) • Aging : 475oC x 2h • Finally, the mock-up fabricated by using parameters above could withstand the heat flux of 20MW/m2 for 1000 cycles. • Based on the successful result of the small mock-ups, a medium scale mock-up (Qualification Prototype) has been manufactured. • This mock-up includes most of the critical technical issues (bonding of CFC/Cu, curved cooling tube, etc...) on the fabrication of the ITER divertor vertical target. This mock-up will be high heat flux tested in Efremov institute in 2008.

8. CFC at 5 MW/m2 (EB heating) Tungsten ~500mm ITER vertical target Qualification Prototype High heat flux test of the small mock-upsand fabrication of a qualification prototype • Braze filler : Ni-Cu-Mn (980oC x 0.5h, followed by Ar gas quench) • Metalizing : Present (Ti-Cu, 7.5 - 10%-Ti) • Aging : 475oC x 2h • Finally, the mock-up fabricated by using parameters above could withstand the heat flux of 20MW/m2 for 1000 cycles. • The high heat flux test of the first mock-up (VTQP-1) has been completed in November, 2008. This mock-up has successfully withstood the heat flux requirements as follows; • CFC : 10MW/m2 x 1000 + 20MW/m2 x 1000cycles • W : 3MW/m2 x 1000 + 5MW/m2 x 1000 cycles • Based on this, Japan Domestic Agency (JADA) has been qualified to proceed the subsequent divertor procurement.

9. UT sensor CFC (porous) CFC (porous) Development of a new NDE facility"FIND" (Facility of Infrared Non-destructive examination for Divertor) • In parallel to the optimization of the manufacturing process, acceptance test methods have also been provided. • In addition to the conventional methods (UT, RT, He leak testing for the braze/weld joint), infrared thermography examination is essential for the non-destructive examination of the vertical target. Internal defect of CFC or interfacial defect between CFC and soft copper can not be detected by UT. UT can detect the interfacial defects between CuCrZr tube and soft copper interlayer.

10. Development of a new NDE facility"FIND" (Facility of Infrared Non-destructive examination for Divertor)

11. Development of a new NDE facility"FIND" (Facility of Infrared Non-destructive examination for Divertor) • By monitoring of the thermal transient of the mock-ups due to switching of the hot/cold water, internal defect of the CFC or interfacial defect of the CFC/Cu joint can be detected. • Time-to-time differential of the thermal transient between defected and non-defected mock-ups gives quantitative evaluation of the defect size and the location with the help of FEM analysis.

12. Contents • Development / provisional procurement activity of the ITER divertor outer vertical target • Small divertor mock-ups • Non-destructive examination by using infrared thermography • Development of the first wall of the ITER test blanket module (TBM) • Provisional high heat flux test of the full-poloidal length TBM first wall mock-up • Development of the DEMO divertor • Small divertor mock-ups made of ferritic steel with tungsten armor

13. ITER TBM Test Program • ITER TBM Program is to test the necessary functions of DEMO Blanket in the real fusion environment with test module scalable to DEMO blanket. • Demonstration of production of fusion fuel tritium • Demonstration of extraction of energy • Demonstration of shielding performance • ITER TBM Test Program is one of the most important development step. • Based on the results from the basic R&Ds on material testing and fabrication trial, near-full-scale mock-up has successfully been developed. Cross sectional view of ITER

14. Full poloidal length TBM first wall (FW) mock-up WCSB TBM (a sub-module) front plate 250mm 1500mmL x 176mmW - HIP condition：1100ºC, 150MPa, 2h - 15 rectangular coolant channels FW （F82H) rectangular tube rectangular tube 1500mm 10 microns Overall view of the TBM-FW mockup with coolant manifolds front plate ~18cm 8mm Cross Section Fabricated FW mockup

15. High heat flux test H+ Ion Beam TBM FW Mockup Infrared camera Image In ~0.5MW/m2, 30sec, 80 cycles of heat load, hot spot due to bond defect was not observed. Expected heat removal performance was demonstrated. HHF Test Condition Heat Flux: 0.5 MW/m2 Beam Pulse: 30 s Coolant Inlet P.： 15 MPa Coolant Inlet T.: 280 oC Flow velocity: 2 m/s Preliminary high heat flux test of the TBM-FW mockup has been carried out under high temperature pressurized coolant condition. Further thermal cycling test of this mockup is planned in this year.

16. Contents • Development / provisional procurement activity of the ITER divertor outer vertical target • Small divertor mock-ups • Non-destructive examination by using infrared thermography • Development of the first wall of the ITER test blanket module (TBM) • Provisional high heat flux test of the full-poloidal length TBM first wall mock-up • Development of the DEMO divertor • Small divertor mock-ups made of ferritic steel withmonoblock tungsten armor

17. Small divertor mock-up for DEMO application • As a basic R&D to achieve DEMO divertor, a small mock-up made of F82H with monoblock tungsten armor has been fabricated by using HIP bonding technique. at 5 MW/m2 (EB heating) • HIP@970oC x 150MPa x 2h • 750oC x 1.5h High heat flux test The central two tungsten lamellae were overheated due to the delamination of the HIP bonded interface. Further improvements of the HIP bonding condition for the F82H and tungsten joint are essential.

18. Summary • R&D activities on the PFCs for ITER and DEMO reactor have extensively been performed in JAEA. • For the ITER divertor components, the braze bonding technique has been optimized. The ITER divertor qualification prototype has successfully been fabricated and soon be high heat flux tested in Efremov institute. • In addition, the thermal performance of the TBM first wall with full poloidal length have successfully been demonstrated, and also the soundness of the HIP joint of F82H was demonstrated. • On the other hand, the HIP bonding condition for the tungsten and F82H joint should be investigated and improved as a bonding method for the DEMO divertor. • Further development of the bonding technique should be necessary to realize the DEMO divertor. The development of the structural material (F82H) will be continued, in parallel to the development of the TBM components.