High heat flux components for a DEMO fusion reactor: material and technology development

Transcription

1 High heat flux components for a DEMO fusion reactor: material and technology development Matti Coleman Power Plant Physics and Technology Department EUROfusion G. Federici, J-H. You, T. Barrett, C. Bachmann, L. Boccaccini, et al. WPDIV WPBB WPPMI

2 Outline Nuclear fusion and fusion reactors Power exhaust Divertors and blankets Material restrictions Current HHF technologies (ITER) HHF technologies for future reactors (DEMO) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 2

3 Nuclear fusion Stellar nuclear fusion: high temperature and high density plasma required to overcome Coulomb repulsion (H-H reaction chain) Gravitational confinement Nuclear fusion on earth: higher temperature, lower density plasma (D-T reaction) Magnetic confinement 2 1 D + 3 1T 4 2He (3.5 MeV) + 1 0n (14.1 MeV) ~ 10 9 y Tokamak 0.42 MeV 0.42 MeV ~ 4 s 5.49 MeV 5.49 MeV ~ 400 y MeV M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 3

4 Timeline and devices for exploitation of fusion power Fusion is plausible Romanelli et al., (2012) Fusion is feasible ITER Fusion is practical, attractive DEMO Power Plant Fusion is commercially exploited JT-60SA Fusion facilities around the world ~2020 Operation ~2025 ~2050 M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 4

5 DEMO power plant DEMO has to be a representative fusion power station: DEMO has to produce electricity (500 MWe) predictably and safely, with minimal environmental impact Fuel cycle self-sufficiency; breed its own tritium 6 3 Li + 1 0n 4 2He (2.05 MeV) + 3 1T (2.75 MeV) Pave the way for commercial fusion power M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 5

6 Tokamak steady-state heat fluxes Thermal power must be exhausted (and preferably used for electricity generation) 2 1 D + 3 1T 4 2He (3.5 MeV) + 1 0n (14.1 MeV) 80% P α + P aux 20% + X M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 6

7 Tokamak transient heat fluxes Plasma instabilities/events can release significant amounts of energy over very short timeframes: Edge-localised modes (ELMs) ~5-10% W plasma in ~ μs Many other plasma events High thermal energy content (100s MJ vs. few MJ in current devices) More intense damage effects (e.g., material ablation and melting) Steady-state Edge-localised mode Klimov et al., (2009) Courtesy CCFE Whyte, (2008) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 7

8 Divertors and blankets Divertor minimizes the influence of plasma wall interactions on the main plasma, but concentrates the heat and particle flux onto a relatively small area. Blanket deals with reduced heat load but must: breed tritium, shield the magnets, and extract power (preferably at high T) MW/m 2 heat flux PWI and erosion First Wall q w =500kW/m²...? Hirai et al., (2013) Carloni, Arbeiter, Courtesy: et al., (2013) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 8

9 Heat sink materials for fusion High heat flux structural materials Fe Eurofer has poor thermal conductivity ~30 W/mK Cu CuCrZr susceptible to irradiation damage and has a poor operating temperature window and service life Elements by thermal conductivity k = 80 W/mK k = 400 W/mK Radioactivity/tritium Melting point < 500 C Thermal k < 50 W/mK Strength at 300 C Ductility at 200 C Availability/cost M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 9

10 Plasma-facing materials for DEMO Plasma-facing surfaces must be able to withstand contact with the plasma (HHF and erosion), including transients Must be compatible with D-T plasma operation Be Good oxygen getter Low Z lowest plasma radiation May melt even as a result of disruption mitigation High erosion CFC k = 150 W/mK α = 4 με/k Better able to handle transient heat loads (ELMs) and disruptions (very high Eurofer T) CuCrZr PFC manufacturing and qualification difficulties leading to significant rejection rate High erosion k = 30 W/mK k = 350 W/mK Demonstrated tritium α = 10 retention με/k problem α = 17 με/k W Very high melting temperature Low erosion in a reactor Still need to improve thermal fatigue performance Melting/resolidification induced irregularities with consequent melting/evaporation increase due to particle glancing incidence Eckstein et al. (1997) Roth et al. (2009) Physical sputtering M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 10

11 Irradiation, activation, and operating T windows Radiation damage mechanisms: embrittlement, thermal creep, swelling, etc. to be carefully considered in the design phase Lowest swelling occurs in BCC alloys (Ferritic steels) Activation of materials must be minimised (no geological disposal) Reduced CuCrZr T Activation window FM ~ (200) Steels (Ta 300 repl. C Nb, V repl. Irradiation Ti, Cr repl. Mn) embrittlement at lower Eurofer (EU) 8.9% Cr,1% W, 0.2% V, 0.14% Ta, temperatures 0.12%C Loss of strength and creep at higher temperatures Non-ductile design criteria: extend T window! Eurofer T window ~ C Irradiation embrittlement at lower temperatures Loss of strength at higher temperatures 16 dpa Eurofer CuCrZr 350 C Irradiated 300 C 550 C Unirradiated Gaganidze et al. (2008) Fenici et al. (1994) Tavassoli You et et al. (2009) (2014) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 11

12 Present technological solutions ITER divertor target: Water-cooled (100 C, 10 m/s, 3.3 MPa) Hot radial pressing of pipe through soft Cu interlayer into solid W block Swirl tube increases margin to CHF by ~30% and HTC by ~15% (Raffray et al., 1999) May be feasible in DEMO, but Demonstrated in ITER-like conditions (Gavila performance et al., 2014): limited by 10 MW/m 2 (5,000 cycles) 20 MW/m 2 (300 cycles) degradation of CuCrZr due to irradiation ITER hypervapotron (first wall): Water-cooled (120 C, ~5 m/s, 3.3 MPa) Be tiles brazed to CuCrZr laser welded to SS 316 Used in various locations as Enhanced Heat Flux (EHF) first wall and divertor dome Not suitable for DEMO; Demonstrated in ITER-like conditions (Mazul coolant et al., T too 2011): low, CuCrZr 5 MW/m 2 (1,000 cycles) 10 MW/m 2 (800 cycles) activation and service life, Be erosion, n attenuation Hirai et al. (2013) Raffray et al. (2014) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 12

13 ITER-like DEMO divertor target ITER Structural Design Criteria (SDC-IC) Numerous elastic structural design criteria for irradiated CuCrZr in use at low temperatures. In elastic analyses, the design is driven by two rules in particular: Developing inelastic analysis procedures DEMO does not have the luxury to design by experiment Compliance with structural design criteria (SDC-IC) q = 10 MW/m 2 (steady-state) q = 20 MW/m 2 (excursion) + erosion T < 1800 C to limit W recrystallization Required for erosion/lifetime Pintsuk et al. (2013) Interlayer 14-16mm Raffray 150 et Cal. (1999) 5 MPa 16 m/s T < 300 C to prevent CuCrZr softening/creep Peak wall heat flux < 1.5 x CHF You, Greuner, et al. (2015) Barrett et al. (2015) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 13

14 Interlayer engineering: Thermal break log (E/E Cu ) Design rationale: Engineer the properties of the interlayer between the W block and the CuCrZr pipe to alleviate stresses in both materials Thermo-mechanical decoupling of the W block and the CuCrZr pipe Address the mismatch in thermal expansion of W and CuCrZr reduce thermal stresses Redistribute the heat to the coolant around the pipe reduce the peak heat flux Minimum reserve factor Barrett et al. (2015) log (k/k Cu ) Cu felt Cu foam M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 14

15 W/Cu composites Design rationale: Best of both; Enhance high temperature strength and toughness Strengthen Cu and ductilise W for use as a structural material W f -Cu composite W-Cu laminate 1m Charpy test at 300 C W rod Charpy test at 300 C W/Cu laminate pipe v. Müller et al. (2014) You et al. (2015) Reiser, Rieth et al. (2012) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 15

16 DEMO blanket first wall (FW) Blankets used to breed tritium and extract power Low temperature coolant bad for power extraction High pressure drop bad for power extraction Thick wall bad for breeding Need to avoid direct contact with the plasma as far as possible Two coolants considered for DEMO blankets: H C, 15.5 MPa (PWR) C, 8 MPa Water-cooled FW up to 1.5 MW/m 2 He-cooled FW up to 1.0 MW/m 2 Aubert et al. (2015) Arbeiter et al. (2015) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 16

17 HHF solutions in DEMO Plasma-facing HHF solutions will vary in response to the incident loads <1.5 MW/m MW/m 2? <10 20 MW/m 2 Traditional FW designs: He channels H 2 O pipes Other solutions: W-Eurofer hypervapotron W-Eurofer monoblocks W-CuCrZr monoblocks M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 17

18 Summary and outlook High heat fluxes and high particle fluxes are inherent in the confinement and control of fusion plasmas (10 20 MW/m 2 ). For DEMO the power handling capability of the divertor is a major driver of reactor size. Paucity of choice in terms of materials (Cu, Fe, and W), due to activation and plasma interactions. Present ITER technologies are not directly applicable for DEMO. Multiple HHF technologies under investigation: ITER-like monoblock Thermo-structural break W/Cu composites Functional grading Others Thank you WPDIV WPBB WPPMI M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 18

19 References Romanelli et al., Fusion electricity: A roadmap to the realisation of fusion energy, 2012 Klimov et al., Experimental study of PFCs erosion under ITER-like transient loads at plasma gun facility QPSA, Journal of Nuclear Materials, no , pp , 2009 Whyte, The challenges of plasma-surface interactions for ITER & beyond, 2008 Hirai and Escourbiac, Status of ITER tungsten divertor design and technology R&D, 2013 Eckstein and Lázló, Sputtering of tungsten and molybdenum, Journal of Nuclear Materials, 183, pp19-24, 1991 Roth et al., Tritium inventory in ITER plasma-facing materials and tritium removal procedures, Plasma Physics and Controlled Fusion, 50, 2008 Barrett et al., Enhancing the DEMO divertor target by interlayer engineering, Fusion Engineering and Design, 2015 Gavila et al., High heat flux testing of EU tungsten monoblock mock-ups for the ITER divertor, Fusion Engineering and Design, 2014 Raffray et al., The ITER blanket system design challenge, Nuclear Fusion, 54, 2014 Mazul et al., Preparation to manufacturing of ITER plasma facing components in Russia, Fusion Engineering and Design, 86, pp , 2011 Ward, Fusion and the characteristics of fusion power plants, KIT Fusion Summer School, 2013 Escourbiac, High heat flux and critical heat flux (CHF) tests in support to ITER HHFC, International HHFC Workshop, USCD, CA,,2008 Fenici et al., Effect of fast-neutron irradiation on tensile properties of precipitation-hardened CuCrZr alloy, Journal of Nuclear Materials, , pp , 1994 Pintsuk and Loewenhoff, Impact of microstructure on the plasma performance of industrial and high-end tungsten grades, Journal of Nuclear Materials, 438, 2013 You et al., Conceptual design studies for the European DEMO divertor: rationale & first results, International Symposium on Fusion Nuclear Technology, 2015 Tavassoli et al., Status of RAFM and ODS steels development in EU, Joint IAEA/EC Meeting, 2009 Gaganidze et al., Irradiation Programme HFR Phase IIb SPICE Impact testing on up to 16.3 dpa irradiated RAFM steels, Final Report for Task TW2-TTMS 001b-D05, 2008 Aubert et al., WCLL FW developments, EUROfusion IDM reports, 2015 Reiser et al., Tungsten foil laminate for structural divertor applications Basics and outlook, Journal of Nuclear Materials, 423, pp1-8, 2012 You et al., Divertor PFC concepts considered in the DEMO divertor project, European Fusion Programme Workshop, 2014 Pereslavtsev et al., Report for TA WP13-SYS-02 System Level Analysis T08 Activation and Radiation Dose Map Calculation, 2014 v. Müller and You., W/Cu composites concepts, IDM 2MBPTL, 2014 Coleman et al., Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO, Fusion Engineering and Design, 89, , 2013 Loving et al., Pre-conceptual design assessment of DEMO remote maintenance, Fusion Engineering and Design, 89, , 2013 M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 19

20 Back-up slides Matti Coleman Power Plant Physics and Technology Department EUROfusion G. Federici, J-H. You, T. Barrett, C. Bachmann, L. Boccaccini, et al. WPDIV WPBB WPPMI

21 DEMO fusion reactor Courtesy CCFE Coleman et al. (2013) Loving et al. (2013) M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 21

22 DEMO development fission path Performance attributes DEMO Time / Resources Committed M. Coleman EuCARD 2 meets Industry Workshop CERN 06/11/15 Page 22