Edge Plasma and Divertor Issues in DEMO-FNS Project V.Yu. Sergeev 1, B.V. Kuteev2, A.S. Bykov1, V.G. Skokov1, A.A. Gervash3, D.A. Glazunov3, P.R. Goncharov1, A.Yu. Dnestrovskij2, R.R. Khayrutdinov2, A.V. Klishchenko2, V.E. Lukash2, I.V. Mazul3, P. A. Molchanov1,V. S. Petrov2, V.A. Rozhansky1, Yu.S. Shpansky2, A.B. Sivak2, A.V. Spitsyn2 1Peter the Great St. Petersburg Polytechnic University 2National Research Center “Kurchatov Institute”, Moscow 3D.V. Efremov scientific research institute of electrophysical apparatus, St. Petersburg V.Sergeev@spbstu.ru 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 1/20
Challenge: As simple as possible divertor design As compact as withstanding neutron irradiation which provides stable & steady-state divertor operation consistent with reduced heat loads and erosion of solid PFM & high-quality ELM free H mode of plasma core 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 2/20
Outline 1. DEMO-FNS basic parameters 2. Concept of power and particle exhaust 3. PFCs design and choice of materials 4. FW mock-up testing 5. Modelling of DEMO-FNS operation 6. Summary 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 3/20
DEMO-FNS basic parameters [Kuteev B.V. et al 2015 Nucl. Fusion 55 073035] 2.5 ÷ 2.75 Plasma major radius R (m) 2.5 ÷ 2.75 Aspect ratio R / a Elongation κ elong 2.1 Triangularity δ 0.5 Plasma current I p (MA) 5 Magnetic field at geometric axis B t (T) 5 Volume averaged electron density <n> (10 20 m − 3 ) 1 Neutron load to FW (MWm − 2 ) ≤ 0 . 2 Beam energy E b (keV) 500 External power P EC + P NBI , М W 6+30 ≤ 40 Fusion power P Fusion (MW) H = τ E / τ IPB98(y,2) factor 1.0-1.4 Beta normalized β N < 3 Non-inductive current fraction 1 First wall area S w (m 2 ) 162 ÷ 188 Plasma volume V pl (m 3 ) 103 ÷ 113 Lifetime (year ) 30 Duty factor 1/3 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 4/20
Concept of power and particle exhaust Ø A Double Null (DN) divertor configuration is chosen since: Decrease of the divertor heat flux up to1.5 times [ Pitcher C.S. and 1) Stangeby P.C. 1997 PPCF 39 779 ] “Good coupling of the radiating divertor with high performance H- 2) mode plasma” [ Petrie T.W. et al 2008 Nucl. Fusion 48 045010 ] Seems to be more favorable due to more symmetric loads at 3) transients (ELMs, VDE, start-up, ramp-down) [ Petrie T.W. et al 2003 Nucl. Fusion 43 910 ] Ø) V-shaped long outer divertor leg is used: Easier for detachment [ Kukushkin A.S. et al 2011 Fusion Eng. and Des. 86 1) 2865; Asakura N. et al 2010 J. Plasma Fusion Res. SERIES 9 136 ] Better for coupling with high performance H-mode plasma: “The 2) ∼ R variation of the parallel heat flux into the detachment zone 1/ helps to stabilize the location of the thermal front—keeping it in the divertor volume” [ LaBombard B. et al 2015 Nucl. Fusion 55 053020 ] 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 5/20
Concept of power and particle exhaust(1) [Greenwald M. et al 2007 Fusion Sci. Technol. 51 266] Ø H factor plotted against outer gap, showing that confinement does not deteriorate if the plasma wall spacing is larger than 2 to 3 mm. This distance roughly corresponds to the SOL width. [ASDEX Team 1989 Nucl. Fusion 29 1959] “Non-ideal” divertor configuration in which (3050%) of plasma power exhausting Ø through separatrix is distributed onto the first wall is considered for DEMO-FNS [ Sergeev V.Yu. et al 2015 NF 55 (accepted) ]. Plasma-separatrix gap of DEMO-FNS should be 23 cm since � q 5 mm [ Eich T. et al 2011 Ø � Phys. Rev. Lett. 107 215001] 5/2� � q � 1.3 cm of the SOL temperature and density e-folding lengths and deviation SP (� 1 cm) due to ELMs at JET. Control system should provide accuracy of plasma position < 1 cm during steady-state operation. Experiments with reduced plasma-wall gaps would be interesting on contemporary Ø tokamaks (JET, Asdex Upgrade, D-IIID, TCV) where this is technically possible. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 6/20
Concept of power and particle exhaust(2) Ø Impurity puff is being tested as a tool for obtaining the detachment conditions in simulations [ Kukushkin A.S. et al 2011 Fusion Eng. and Des. 86 2865; Asakura N. et al 2010 J. Plasma Fusion Res. SERIES 9 136 ] and experiments on JET [ Giroud C. et al 2012 NF 52 063022 ], Asdex Upgrade [ A. Kallenbach 2015 NF 55 053026 ], D-IIID [ Petrie T.W. et al 2008 Nucl. Fusion 48 045010 ]. Local neon puff is being considered for DEMO-FNS. Ø ELMs are the concern of divertor and FW designs as well. According to JET report [ Loarte A. 2006 Proc. 7th ITPA Divertor and SOL Physics ] 2000 Type I ELMs with energy content 1 MJ each caused erosion of 3 mm of Be wall which is unacceptable for the steady-state operation. Ø External injection of lithium as pellets, droplets or liquid jets into the edge plasma could control ELMs and/or H-mode regime as it was demonstrated using Li dust and granules injection in NSTX [ Bell M.G. et al 2009 PPhCF 51 124054 ], EAST [ Hu J.S. et al 2014 Fusion Eng. Des. 89 2875 ] and DIIID [ Osborne T.H. et al 2015 NF 55 063018 ]. We rely on such external injection of lithium to mitigate or better to suppress ELMs in DEMO-FNS. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 7/20
PFCs design Divertor (ITER like, pipe-type) First wall (thin, arched plate-type) [Sergeev V.Yu. et al 2012 Plasma Phys. Rep . 38 521] [Sergeev V.Yu. et al 2015 NF 55 (accepted)] 16 mm 10 mm 16 mm 3 mm 2 10 mm 3 2 3 1 6 mm 4 5 4 1 Sketches of PFC elements for DEMO-FNS: (1) heat carrier pipe (left) and plate (right) made of chromium–zirconium bronze, (2) sectioned beryllium coating, (3) brazing layer, (4) stainless-steel base (vacuum vessel), (5) pipe for heat carrier (water) flow. ANSYS code simulations with heat transfer coefficient values of about 80-90 kW/(m2K) Ø was done. It reveals that both designs can withstand heat loads up to 10 MW/m2. Further development of the shaping procedure for FW design instead of the cutting Ø procedure of the target is in progress, which will allow manufacturing curved surfaces of the first wall panels. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 8/20
Choice of materials [Sergeev V.Yu. et al 2015 NF 55 (accepted) and references in] Ø Evaluated temperatures satisfy temperature limitations described above for the 5� 10 MW/m2 heat loads at Be, CuCrZr bulks and Be-CuCrZr interface. For heat loads lower than 1 MW/m2 CuCrZr heat-sink material can be replaced to SS . Ø CuCrZr/Be/SS keep their properties at 5/10/30 dpa correspondingly. Neutron Ø irradiation of PFC 20 dpa during the DEMO-FNS lifetime. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 9/20
FW mock-up: design and manufacturing 2 3 1 4 5 Sketch (a) and photo (b) of mock-up of the PFC element for FW [ Mazul I. et al. 2012 Fusion Eng. Des. 87 437 ]: (1) heat carrier plate made of chromium–zirconium bronze, (2) sectioned beryllium coating, (3) brazing layer, (4) stainless-steel base (vacuum vessel), (5) pipe for heat carrier (water) flow. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 10/20
Mock-up: experiments on Tsefey- M Ø Photo (a) and sketch (b) of test-bed experiments with the mock-up of the PFC element on the electron beam facility Tsefey-M [ M.Rodig et al 2000 FED 51–52 715 ] at Efremov Institute. Ø The mock-up has 2 sets of Beryllium tiles: 3х16х16 mm – 18 pieces and 3х7х16 mm – 6 pieces. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 11/20
Mock-up: results of experiments Ø Parameters of test-bed experiments § Area of heat loads 50 cm2 § Range of heat flux 5-10.5 MW/m2 ± 5% § Cross section of water cooling 204 mm2 § Water velocity 7 m/s § Water rate 1.4 kg/s § Water pressure at the entrance 2 MPa ± 5% Ø 1 stage: Power load was lasting 30 sec and was incremented up to 5 MW/m2 (1.0, 2.0, 3.0, 3.5, 4.0, 4.5, 5.0 MW/m2). Then, the cycle load (15 sec load, 15 sec pause) was done at 5.0 MW/m2. The mock-up had successfully sustained during 1000 cycles. The surface temperature was about 260 C. Ø 2 stage: Power load was increased with a step of 0.5 MW/m2 to reach the surface temperature of 600 C. It is occurred at 10.5 MW/m2. Then, the cycle load (15 sec load, 15 sec pause) was done at 10.5 MW/m2. The mock-up had successfully sustained during 100 cycles. 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 12/20
Temperature fields of the mock-up Ø Temperature (260 600 C) of tiles demonstrate the uniform radiation (IR) Ø No tiles lost a thermal contact at both at 5 MW/m2 (left – shot #120) and at 10.5 MW/m2 (right – shot # 95) 1st IAEA Technical Meeting on Divertor Concepts, 29 Sept – 2 October 2015, Vienna, Austria 13/20
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