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Configuration of A=2 Based Tokamak for Testing of Reactor Technologies

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Configuration of A=2 Based Tokamak for Testing of Reactor Technologies Russian Federation Agency for Atomic Energy (ROSATOM) ConfigurationConfiguration ofof A=2A=2 basedbased tokamaktokamak forfor testingtesting ofof reactorreactor technologiestechnologies (TRT)(TRT) E.Azizov1, Yu.Arefiev1, O.Buzhinskij1, V.Dokuka1, O.Filatov2, V.Krylov2, P.Khayrutdinov1, V.Korotkov2, A.Krasilnikov1, A.Lopatkin3, A.Mineev2, N.Obysov4, V.Cherkovets1, E.Velikhov5 1State Research Center of Russian Federation, Troitsk Institute for Innovation and Fusion Research (TRINITI), Troitsk, Moscow Region, 142190, Russia 2D.V. Efremov Scientific Research Institute of Electrophysical Apparatus (NIIEFA), Metallostroy, St. Petersburg, 196641, Russia 3Research and Development Institute of Power Engineering, Moscow, 101000, Russia 4Russian Federation Agency for Atomic Energy, Moscow, 101000, Russia 5Russian Research Center “Kurchatov Institute”, Moscow, Russia St. Petersburg, 2005 ITER - principal step on the way to fusion power. ITER is to demonstrate the feasibility of: - ignition and long fusion plasma burning in an efficient unit; - integration of the basic systems of the future DEMO reactor into the acting facility; - real safety of fusion reactors. ITER is cannot help in solving the following problems: - radiation resistive materials; - test and development of systems and components with necessary resources required for steady state operation; - control of reactor processes in steady state burning; - effective stationary systems of auxiliary plasma heating; - closed fuel cycle; - effective blanket. Of great importance is the creation of a test facility for the development of basic systems of DEMO and next generations of fusion reactors. Contribution of leading countries to Modern Fusion Programs Europe – JET, ASDEX-U, TORE-SUPRA, MAST, STELLARATORS; USA – DIIID, ALCATOR, NSTX; Japan – JT-60U; China – HL-2M, HT-7; Russia – T-10, T-11M, Tuman-3, Globus-M, FT-2 Next steps of the leading countries: Europe – ITER, IFMIF, W-7X; Japan - JT-60 SU, IFMIF; USA – NCSX; China – EAST; Korea – KSTAR; India – SST-1; CTF-TRT – may be the best option for the national Russian fusion program. The important step, as most members of the fusion community believe, would be development of TRT for testing of components of future fusion reactors with neutron fluencies typical for reactors. We propose to create TRT on the basis of tokamaks with small aspect ratio. It is necessary to note that the proposed compact spherical tokamaks with reactor parameters, have been made earlier in the USA, the Great Britain, Russia and in China. Their performances, Wilson proposal, etc. are given in Tables 1 and 2. Table 1 Parameters of modern VNS and CTF projects based on tokamaks China China USA, USA, UK, USA, 1998 [3] 1998 [3] M.Peng, M.Peng Robins ARIES 1998 [4] , 1998 on 1998 2000[6J [4] [5] Neutron loads рn, MW/m2 1.02 0.5 1 4 1.5 0.23 Aspect ration А 1.4 1.4 1.4 1.4 1.6 1.6 Major radius R, m 1.4 1.4 1.1 1.1 0.7 2.37 Elongation, k 1.85 1.85 3 3 2.3 3.1 Safely factor q9V 5.5 7.5 9 9 3 3.1 βN 6 5.5 4 7.6 2.6 5.04 Plasma current Ip, MA 9.2 7 10 10 10.3 12 Toroidal field Bt,T 2.5 2.5 2.1 2.1 2.94 1.54 Bootstrep-current faction fBS 0.72 0.81 0.5 0.9 0.3 (*) 0.95 Additional heating power РАих, MW 28 19 40 70 46 47 Fusion gain factor Q 3.6 2.6 1.65 3.7 0.8 1.65 Enhancement factor. Нτ 2.2 2.7 1.5 1.7 1.1 1.7 Table 2 Today proposals of fusion reactor based on compact torus R.Stambaugh H.R.Wilson [8] [9] 2 Neutron loads рn, MW/m 4.1 3.5 Aspect ration А 1.6 1.4 Major radius R, m 3.2 3.42 Elongation, k 2 2.44 Safely factor q9V 3.4 3.2 βN 8.3 8.2 Plasma current Ip, MA 31 31 Toroidal field Bt,T 2.14 1.77 Bootstrep-current faction fBS 0.99 0.92 Additional heating power РАих, MW ~50 50 Fusion gain factor Q 57 64 Enhancement factor. Нτ 1.8 1.6 Proposals differ not only by performances and the technical concept, but also by the purpose. In the USA, the Great Britain and Russia spherical tokamaks with reactor parameters were considered as the basis for future compact reactors. In China they were condidered as the basis for hybrid reactors, and then reactors for transmutation of long- lived SNF. The concept of a compact tokamak as a volumetric neutron source for transmutation was developed in different countries, including in Russia (JUST-T). It is necessary to note that the parameters presented in Table 2 are too high. So, in Wilson‘ proposal it is supposed, that ( li ≈ 0.2 ) and ( βN ≈ 8.2 ). According to the existing physical concept, βN could not be more than 6 ( βN ≤ (4 –6 )li ). The same should refer Stambaugh proposal ( βN = 8.3). Two main directions were considered for development of inexpensive compact spherical tokamaks. The first is TRT, and the second VNS for transmutation. We considered the concept of the compact spherical tokamak reactor as TRT. The concept of compact tokamak as TRT is based on the following assumptions: 1. Aspect ratio А=2 (on boundary between spherical and classical tokamaks); 2. Moderate sizes Ro = 2m, а =1m, k95 =1.7and SN - configuration; 3. βN ≤ 5 ⋅ li; 4. Pf ≈ 50-100 МW; 5. τЕ =H⋅τE,IPB(y,2); H < 2; 6. The neutral beams with deuteron energy of 140 and 400-500 KeV; power РDN <50 МW; 7. Use of ICH and ECH for heating, current drive and plasma parameters control; 8. The possibility to use combined inductive and non - inductive method for formation and ramp-up of current up to the design value; 9. Blanket with the shielding located between the vacuum chamber and toroidal coils on outside contour. Peculiarities of Plasma Confinement in Spherical Tokamaks In the NSTX device a weaker degradation of the energy confinement time τE depending on additional heating power P was observed. At the same ST, a stronger dependence on the toroidal magnetic field Bt was found. According to the MAST database, a stronger dependence τE on the index of the inverse aspect ratio ε seems to exist, 0.81 τE ∝ε The database for the enhanced energy confinement regimes MAST NSTX NSTX The confinement enhanced factor does not depend on q95 and depends weakly on Te and Ti Data base on improved confine modes: ASDEX Upgrade, DIII-D, FT-U, JET, JT-60U, TORE SUPRA DIII-D JT-60U: Squares - NB heating, circles NB + EC heating The improved confine factor does not depend from q95 and weekly depends from Te/Ti While developing the TRT concept the experimental data of tokamaks NSTX and MAST were used. The calculations of scenarios for TRT were carried out using the DINA code. The main points of the calculated model: 1) The initial stage of ramp-up of the plasma current continues up to 2.5 MA with CS; 2) The further plasma current ramp-up and maintenance through the steady-state operation regime is provided by tangential injection of deuterium beams; 3) In calculations two neutral beams with different energies were supposed to be used; 4) The plasma density profile was supposed to be parabolic with a pedestal being equal to 0.9 of the central density; 5) The current drive due to NB injection was calculated according to J.D. Gaffey et al; 6) The bootstrap current was calculated according to O.Sauter et al. Injection of beams with energies from 140 keV to 400 keV and a total power of about 50 MW makes it possible to achieve several goals - plasma heating up to Ta= 7-8 keV; - required conditions for generation of considerable fraction of the bootstrap current; - control over the profiles of plasma current and the safety factor; - to generation up to 50% of the necessary neutron flux. TRT steady state mode (conditions for sustaining) ¾q95 ~4-5 ¾No monotonic profile of q ¾Completely no inductive current ¾HH ~ 1.5-2, βN ~ 4-6 ¾ Ibs/Ip ~50% Main parameters of TRT with maximum neutron loading 20 -3 R=2m; a=1m; Bt=4T; ne=1.5⋅10 m ; k=1.7; Paux=50 MW (400) keV) Plasma current, Ip, MA 8.7 Poloidal beta, βp 1.2 Electron temperatures, <Te>/Teo,keV 14/50 Ion temperatures,<Ti>Tio,keV 15/41 ne/ngw 0.53 Internal inductance, Ii 0.68 2 Average neutron loading, Гп, MW/m 1.3 Fusion gain factor, Q 4.8 Normalized beta, βN 5.8 Bootstrep-current faction, fbs 0.54 Enhanced factor of energy continental 2 time, H(y,2) TRT composition Internal Contact External The vacuum part TMS zone part ТМS chamber Toroidal magnetic system The central solenoid NumberNumber of of toroidal toroidal coils coils 20 20 RippleRipple << 2 2 % % Field on R = 2 m (Т) 4 Field on R00= 2 m (Т) 4 NumberNumber of of turns turns in in coil coil 1414 CurrentCurrent in in the the tur tur (K (KАА) ) 140140 MaterialMaterial CuCu Insulation Al O , ZrO ,SiC Insulation Al2O2 3,3 ZrO2,SiC2 The choice of the configuration with A=2 is defined by: - the ability to use the so-called combined method for start, formation and ramp-up of plasma current; - minimization of the TRT size with warm EMS and high parameters of fusion plasma and, hence with minimum power consumption from the grid; - the possibility to use the physical database of classical tokamaks combined with attractive features of low aspect tokamaks; - more effective use of neutron fluxes from the plasma core for testing.
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