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Russian Federation Agency for Atomic Energy (ROSATOM)

Russian Federation Agency for Atomic Energy (ROSATOM). Configuration of A=2 based tokamak for testing of reactor technologies (TRT).

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Russian Federation Agency for Atomic Energy (ROSATOM)

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  1. Russian Federation Agency for Atomic Energy (ROSATOM) Configuration of A=2 based tokamak for testing of reactor technologies (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

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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.

  7. Table 1 Parameters of modern VNS and CTF projects based on tokamaks

  8. Table 2 Today proposals of fusion reactor based on compact torus

  9. 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).

  10. 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 ( N  8.2 ). According to the existing physical concept, Ncould not be more than 6 (N (4 – 6 )li ). The same should refer Stambaugh proposal (N= 8.3).

  11. 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.

  12. 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); • Moderate sizes Ro = 2m, а=1 m, k95=1.7 and SN - configuration; • 3. N 5  li; • 4. Pf 50-100 МW; • 5. Е=H E,IPB(y,2); H < 2; • The neutral beams with deuteron energy of 140 and 400-500 KeV; power РDN <50 МW; • 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.

  13. Peculiarities of Plasma Confinement in Spherical Tokamaks In the NSTX device a weaker degradation of the energy confinement timeτEdepending 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 seems to exist,τE0.81

  14. 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

  15. 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

  16. 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: • The initial stage of ramp-up of the plasma current continues up to 2.5 MA with CS; • The further plasma current ramp-up and maintenance through the steady-state operation regime is provided by tangential injection of deuterium beams; • In calculations two neutral beams with different energies were supposed to be used; • The plasma density profile was supposed to be parabolic with a pedestal being equal to 0.9 of the central density; • The current drive due to NB injection was calculated according to J.D. Gaffey et al; • The bootstrap current was calculated according to O.Sauter et al.

  17. 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.

  18. TRT steady state mode (conditions for sustaining) • q95 ~4-5 • No monotonic profile ofq • Completely no inductivecurrent • HH ~ 1.5-2, N ~ 4-6 • Ibs/Ip~ 50%

  19. Main parameters of TRT with maximum neutron loading R=2m; a=1m; Bt=4T; ne=1.51020m-3; k=1.7; Paux=50 MW (400) keV)

  20. TRT composition Internal part TMS External part ТМS The vacuum chamber Contact zone Toroidal magnetic system The central solenoid Number of toroidal coils 20 Ripple < 2 % Field on R0 = 2 m (Т) 4 Number of turns in coil 14 Current in the tur (KА) 140 Material Cu Insulation Al2O3, ZrO2, SiC

  21. 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.

  22. About 30-50% of current is provided by the inductive method without reverseof magnetic field of CS. • After completion of the inductive stage, CS can be removed from the tokamak till the next operating cycle. • neutral injection provides attainment of the current plateau and its sustainment; • RF-fields can be used; • the time for attainmentof the stationary condition is defined by skin-processes and amounts to 30-60 min.

  23. The dependence of calculated TRT parameters on aspect ratio

  24. Start and ramp-up of current, transition to steady state regime: • CS is used for break-down initiation and plasma core formation. 30-50 % of current is due to the inductive way without reverse of the CS magnetic field. After the end of the inductive stage CS can be withdrawn from the central post of tokamak till next cycle (starter); • Neutral bean injection is used for attainment of the current plateau and its maintenance. RF-fields can be used, too. • The time for attainment of the current plateau is determied by sin-processes and amounts to 30-60 min.

  25. The steady state regime of maintaining toroidal magnetic field in a tokamak can be supplied by a water cooling. For this purpose it is necessary in multiturn coil provide a parallel channel for water on through central post and an outer parts. • Magnetic field on an axis of a plasma core 3.9 T • Current density in a conductor 3.1107 А/m2 • Power supply of toroidal coils 122 МW • The total cross section water cooling channels 0.31 m2 • Thermal gradient on an input and an output cooling systems Т ~ 100 • Water flow rate ~ 1.5 m3 / c

  26. Central solenoid (CS) will consist of multiturn sections Magnetic flux 6 Vs Field on axis 12 Т Total current 70 МАV One-direction current in the central solenoid changes from maximum down to null. CS provides the start and ramp-up of plasma current up to 2.5 МА. During transition time to steady-state regime CS without current is removed from the radiation area.

  27. The Vacuum Vessel Some versions of the vacuum vessel are considered:  one-shell all-welded construction with stiffening ribs and 10 horizontal ports; the one-shell demountable construction devided into two equal parts with 10 ports and stiffening rib; the two-shell construction. Overall dimensions of the vacuum vessel: With an outer arrangement of blanket R=3.15 m h=5.5 m With inner arrangement of blanket R=3.8 m h=5.5 m

  28. Thermal loadings in divertor area It is assumed that 80 % of a thermal flow of TRT plasma goes to divertor and is divided by separatrix proportionally between inner and outer plates in the ratio 3:1. TRT loading 34 and 11 МW SOL Width 2-3 cm Flow expansion ratio 3 Width of thermal release > 6 cm Thermal loading at 15-20о decline angle 10-15 W / m2

  29. TRT mission is to solve the following problems: - development of plasmaphysical technologies (additional heating plasma system, density, current and temperature profiles control and maintenance of steady-state mode); - test of materials, components, elements, and some separate systems of fusion reactors ; - development of method to suppress current disruption instability; - testing for workability of the systems for control over processes in the fusion plasma and on plasma-wall edge; - test of material and components under high neutron fluencies; - test of the methods to increase the resources of the first wall and divertor; - development and test of methods for on-line control of fusion reactor parameters in the steady-state regime; - development of fuel cycle, etc.

  30. The problems to be solved for development of compact tokamak TRT 1. Experimental verifications of TRT plasma-physical model and feasibility of steady-state operation; 2. Radiation-resistant materials and insulation; 3. High power stationary neutral beam injectors; 4. The stationary divertor capable of receiving 10-15 МW/m2 thermal fluxes; 5. First wall protection; 6. Efficient water cooling system of ТМС; 7. Tritium recovery.

  31. Tokamak - VNS for МА transmutation Use of thermonuclear neutrons for transmutation – one of possible practical applications of tokamaks - VNS with Q ~ 1 - 2. • This direction for efficient SNF utilization offers the following advantages: • Possibilities to use transmutation’s blankets • of various type with reproduction of tritium including: • Solid-state with high-temperature or the low-temperature coolants; • Liquid metal blanket; • 2. Elimination of Pu from transmutation cycle; • 3. Nuclear safety.

  32. Plasma currents, Ip,МА 5 Poloidalbeta, p 1.3 Electrons temperature <Te>/Teo, Kev 6.8/21 Ion temperature<Ti>/Tio, Kev 7.1/24 ne/ngw 0.63 Internalinductivity, li 0.72 Mean neutron loading Гп,MW/m2 0.3 Fusion gain factor of, Q 1.2 Normalizedbeta, N 3.5 Fraction of bootstrap-currents, fbs 0.44 Enhanced factorН(у,2) 1.6 VNS of parameters for transmutationR = 2m; a = 1m; Bt = 4T; ne = 1.5 1020m-3;K = 1.7; Paux = 45 MW (140 / 400 KeV)

  33. Blanket The blanket prototype for transmutation of minor actinides

  34. Efficiency estimation of minor actinides transmutation in blanket Composition of calculated zone Zone 1 – 50 % steel, 50 % coolant Zone 2 – 15 % steel, 50 % coolant, 35 % МА Zone 3 – 15 % steel, 40 % coolant, 45 % МА Zone 4 – 75 % steel, 25 % water. Coolants: liquid lithium; liquid lead; water.

  35. Integral parameters of blanket and МА fission rate 82 % of fusion neutron flux falls on the blanket. Neutron flux on the blanket is 0.423 МW/m2

  36. Some additional problems to be solved during the development of the compact tokamak TRT as the VNS for MA transmutation. 8. Type and composition of the transmutation blanket, choice of coolant with strong magnetic fields taken into account; 9. Self-sustaining electric power; 10. Heat utilization from the TMS cooling system; 11. Safety of operating cycle and system.

  37. Cost estimation of main TRT systems: The basic subsystems – electromagnetic system (EMS), system of additional heating, blanket , divertor, vacuum vessel. The basic contribution to EMS cost is made by the toroidal coils (TMS). It is accepted that ZEMS 1.5 ZTMS. The specific cost of normally conducting magnetic system is taken as 0.1 M $/t. The weight of TMS ~ 400 tons, so ZEMS 60M $. As is known, the specific cost of the additional heating system 2M $/MW. For (ICR + ECR) power of 45 and 25 МW ZAH  140 M $. The specific cost of ITER is used for cost estimation of the TRT blanket, i.e. 0.4 M $/m2. As the blanket area is ~ 100 m2, Zbl 40 M $. Since the VNS blanket for transmutation is more complicated and powerful system, the estimation gives Zbl, 100 M $.

  38. Cost estimation of main TRT system: It is supposed that the TRT divertor will have an ITER-like construction, but approximately three times smaller sizes, its dimentions will be smaller by abont a factor of 3, thus Zdiv  30-40 M$. The TRT tokamak have to include power supply system and energy conversion system. Cost of the power supply system by analogy with ITER is ZPS  40 M $. As estimated by experts, the cost of the energy conversion systemfor the blanket is ZEC  200 M $. The total cost of TRT for technology and material test is about ZTRT  250-300 M $, for transmutation of minor actinides - Z  500-600 M $.

  39. Experimental complex at the State Researsh Center TRINITI which can be used for the TRT project • The experimental hall (404030 m3) with concrete shielding 3 m in thickness; • Assembly hall - 404030 m3; • Apparatus rooms - 800 m2; • Rooms for engineering service system - 1000 m2; • Laboratory rooms - 2000 m2; • Personnel rooms - 2000 m2; • Testing facilities rooms - 1000 m2; • Stationary power of 300 MW supplied by a separate substation; • System of short-turn power supply of 1000 МW and pulse power of 1011W; • Tritium complex with the possibility to operate with 10 gram of tritium per day. • The total cost of the available experimental complex is estimated as 500 M$.

  40. Conclusion • The currently existing physical, technological and engineering data base is enough to develop a compact and relatively very inexpensive TRT; • The TRT development could serve as the basis for the National Russian Fusion Program. • TRT with more moderate physical parameters could be used as a VNS-tokamak for Minor Actinides Transmutation.

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