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Japan PFC/divertor concepts for power plants

Japan PFC/divertor concepts for power plants . T retention and permeation. Problems of T retention would not be serious…. Wall temperature will exceeds 600 ° C. But, if coolant temperature is low (ex. 300 ° C for water), hydrogen isotope trapping near coolant tubes may not be negligible.

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Japan PFC/divertor concepts for power plants

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  1. Japan PFC/divertor concepts for power plants

  2. T retention and permeation • Problems of T retention would not be serious…. • Wall temperature will exceeds 600 °C. • But, if coolant temperature is low (ex. 300 °C for water), hydrogen isotope trapping near coolant tubes may not be negligible. • T permeation to coolant and dynamic retention effect should be considered. • T recovery system from the coolant will have heavy load, if significant permeation flux exists. • Diffusion barrier of T on inner surfaces of coolant tubes will confine T in wall materials, which could increase T retention (dynamic retention, only existed during plasma operation). This effect on the degradation of materials should be investigated. • Design of high heat flux components and firsts wall of blankets need to take this mobile T effect into consideration.

  3. Helium effect on W • Helium effect on surface roughening of tungsten becomes significant over 800 °C. • Nano-fiber (cotton-like) morphology appears at relatively low temperature. • At higher temperature, bubble structure grows together with recrystallization. • These surface modification will probably lead to enhanced erosion and dust generation, which would not be acceptable. • This effect appears under mixed plasma (D, T & He (5-10%), actual burning plasma) conditions. • This could be the most serious surface effect of tungsten in DEMO. • Surface protection by low-z material coating would be necessary

  4. He ion irradiation effects (~1600 K) NAGDIS-II, Nagoya Univ. Surface He bubble formation and recrystallization with He bubbles at grain boundaries could cause enhanced erosion and dust formation Surface He bubble He at grain boundaries Grain Ejection 1 µm He exposure at 1600 K, then D plasma exposure at 550 K D. Nishijima et al., J. Plasma Fusion Res. 81 (2005) 703.

  5. Effect of He plasma on various grades of W M. Baldwin (UCSD), TITAN Workshop 2008 He plasma effects take place for any tungsten material He ion effects at elevated temperatures would be inevitable.

  6. 49th Annual meeting of DPP (2007) M. Baldwin et al. • Be-W alloy and W-C layers (< µm) inhibit He induced morphology. This could be the key technology to use solid wall materials (tungsten) for years.

  7. Pulsed heat effect • Disruption and ELM’s should be suppressed (sufficiently mitigated) for DEMO……. • This would be too strict to realize fusion reactors. • Slight surface melting causes cracking. • Tungsten surfaces with He bubbles (nano-structure) are vulnerable to pulsed heat. Tungsten surfaces easily melt by the heat pulse less than the melting threshold. • Surface protection by low-z material coating with appropriate thickness would be also effective for this.

  8. Surface cracking of W by pulsed heat load S.Pestchanyi, et al., Fusion Eng. Des.82 (2007) 1657. Surface and cross section of W exposed to 0.9 MJ/m2 (0.5 ms duration, 100shots) by QSPA (Quesistationary Plasma Accelerator) • Surface melting and crack formation took place by ELM like heat load

  9. Protection of wall surface • In order to avoid He ion irradiation effects, surface low Z layer is effective. • Choice of low Z material • Carbon: High erosion, T retention (may not be serious in DEMO), and dust formation in remote area are concerns • Beryllium : Mixed layer formation with W, leading to enhanced erosion of tungsten. • Boron : not easy to form thick coating due to brittleness • Boron would be the candidate, but needs more investigation. • This idea was originally proposed by N. Noda, then C. Wong. • Deposition area control and dust collection • Deposition control and regular dust collection (if any) would be needed. • Usually, first walls are erosion zone. Is it possible to make in-situ coating on the first walls?

  10. Neutron Effect • Neutron irradiation effects • Increase in DBTT (Ductile Brittle Transition Temperature) • Void swelling • The above data were taken with fission reactor neutrons. • Increase in T trapping • Not significant at elevated temperatures (>600 °C) • 14 MeV neutron effects are not known for W • Almost no data for 14 MeV neutron irradiation to W. • Transmutation (W  Re  Os) is not negligible for DEMO reactors. • Helium production becomes significant at this energy. • Definitely, we need to study 14 MeV neutron irradiation effects of tungsten. But how?

  11. Transmutation of W by neutron irradiation • Transmutation of W by fusion neutron (Noda et al. J.N.M. 258-263(1998) 934.) • W:5% Re:0.02% Os (3 MW y/m2) • W:10% Re:0.1% Os (6 MW y/m2) • W:25% Re:1.0% Os (15.5 MW y/m2) • Thermal conductivity decrease with increasing Re concentration • Fujitsuka et al. JNM 283-287 (2000) 1148. pure W W95Re5% W90Re10% ~2 years ~1 year Neutron load (3 MW/m2)

  12. New tungsten development • Preferable property for tungsten • High toughness and high yield strength at elevated temperature • High recrystallization temperature • Negligible increase in DBTT by neutron irradiation • UFG-W (Ultra Fine Grained W) • High recrystallization temperature • Highly resistant for neutron irradiation • Preferable properties under high flux plasma exposure • almost no D blistering, observed for PM (powder-metallurgy) tungsten • no enhancement of D retention • Preliminary results from the high density plasma device (UCSD) • Development of fabrication technique of mono-block size UFG-W is planned.

  13. W–0.5TiC–H2 W–0.5TiC–Ar pure W UFG-W : high resistance to neutron irradiation H. Kurishita, et al., J. Nucl. Mater. to be published (2008) It also showed less neutron induced damage (black dots in photo). UFG-W showed less hardening than pure W by neutron irradiation. After n irradiation Before n irradiation

  14. Durability test of PFCs for years of operation • Exposure time in present exp. is much shorter than DEMO • Ion fluence: 1030~1031 m-2 for divertors in DEMO • At present, 1027 ～1028 m-2 • Neutron fluence for DEMO : >5 MWa/m2 • What kind of test conditions are needed? • Complicated conditions for divertor • Heavy irradiation by 14 MeV fusion neutron • radiation damage, transmutation, He production • High heat flux Thermal stress (irradiation creep?) • High fluence He (&D,T) ion irradiation from edge plasmas • ITER engineering phase can provide opportunity for studying high ion fluence effects. But neutron fluence is not enough. • What is the most realistic method for the test? • CTF-like device, IFMIF with plasma sources, using first phase of DEMO, or something else?

  15. Summary • Surface effects of helium bombardment and pulsed heat load are serious concerns for use of tungsten for DEMO. • Pulsed heat load could be mitigated (or suppressed), but He effects are not negligible. • Surface protection by low Z materials is one of the key technologies. For this, in-situ deposition control (in-situ deposition) would be necessary. • 14 MeV neutron effects must be studied. • Testing of tungsten PFC’s must be made under year-long complicated conditions. • High heat flux, heavy neutron irradiation, and high plasma ion flux. • Tungsten with neutron resistance should be developed. • UFG-W is one of the candidate.

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