Magnetic System for the Upgraded Spherical Tokamak Globus-M2

Transcription

1 1 ICC/P1-01 Magnetic System for the Upgraded Spherical Tokamak Globus-M2 V.B. Minaev 1, V.K. Gusev 1, N.V. Sakharov 1, Yu.V. Petrov 1, E.N. Bondarchuk 2, A.F. Arneman 2, A.A. Kavin 2, N.M. Kozhukhovskaya 2, S.V. Krasnov 2, G.S. Kurskiev 1, A.A. Malkov 2, A.B. Mineev 2, A.N. Novokhatsky 1, I.Yu. Senichenkov 3, V.N. Tanchuk 2, V.I. Varfolomeev 1, E.G. Zhilin 4 1 Ioffe Physical-Technical Institute of the Russian Academy of Sciences, St.Petersburg, Russia 2 D.V. Efremov Scientific Research Institute of Electrophysical Apparatus, St. Petersburg, Russia 3 Saint Petersburg State Polytechnical University, St. Petersburg, Russia 4 Ioffe Fusion Technology Ltd., St. Petersburg, Russia contact of main author: Vladimir.Minaev@mail.ioffe.ru Abstract. The necessity of toroidal magnetic field increase for further gain in plasma parameters is evident from experiments conducted on spherical tokamaks. Modernization of the machines is planned for NSTX (US), MAST (UK) and Globus-M (Russia) and aimed at toroidal magnetic field magnification. For the upgraded spherical tokamak Globus-M2 it means toroidal magnetic field increase from the present value of 0.4 T up to 1 T as well as the plasma current rise up to 0.5 MA. Also the poloidal magnetic field consumption extends from its current value of 0.3 Wb up to 0.34 Wb. The vacuum vessel stays unchanged in order to reduce project costs. Present parts of the magnetic system also will be used as far as possible. The key point of the design is the novel central stack with the inductor wound above. In the current report, conception of tokamak upgrade is discussed and mechanical and thermal stress analysis results for the magnets under increased field and plasma current are presented. 1. Conception of the Tokamak Upgrade As is seen from experiments, the weak point of the design for the spherical tokamak magnetic system is the joint unit between central rod and outer limb of the toroidal magnetic coil. Principal cause of that are strong mechanical and thermal dynamic loads on the high-current electrical joints of the toroidal coil. Appling different technical issues for this unit none of three principal spherical tokamaks (NSTX, MAST and Globus-M) could avoid the operating problems. For example, the toroidal field coil of Globus-M [1,2] consists of solid central column and 16 separate outer limbs joined together. Present coil was repaired twice after electric breakdown and that is the cause of current restriction for this toroidal field coil. Previous experiments [3-5] has shown, that the increasing of the toroidal magnetic field together with the plasma current should promote plasma performance and provide improved conditions for auxiliary heating and current drive. Preliminary plasma simulations show, that significant (a few times) rise of the toroidal magnetic field and the plasma current in Globus may help in achieving conditions necessary for the neutron source prototype modeling [6]. We plan to increase toroidal magnetic field up to 1.0 T and plasma current up to 0.5 MA [7]. Evidently, that field value rise by a factor of 2.5 will entail serious increase in loads in the magnetic system elements. For example, only mechanical stress on the toroidal coil caused by electromagnetic forces becomes 6 times more. The upgraded tokamak also provides increased duration of the plasma shot in order to utilize noninductive CD scenarios, which are necessary for regime development of a compact thermonuclear neutron source prototype. Thereupon new design of magnetic system is aimed at achieving both these goals in one machine, which is the upgraded Globus-M2 tokamak. Physical design details are provided in [8].

2 2 ICC/P1-01 It is necessary to emphasize, the case in point is just the device upgrade. The principal element, which is vacuum vessel with its ports for auxiliary heating and diagnostic systems and technological equipment, stays unchanged. Probably the revision of the in-vessel component design will be necessary in future due to increased radiation load. At present the modernization applies to magnetic system of the tokamak only. It consists of forcing up the operating mode for the toroidal field coil and central solenoid. Increase in toroidal magnetic field and plasma current together with plasma shot prolongation leads to significant heating of the coils in addition to applied mechanical extra loads. At the same time we have to place more strained elements of the magnetic system in the space bounded by the existing vacuum vessel. That is the main cause of drastic revision of the design. The diameter of the central stack in its top joint area is enlarged up to the maximal diameter, allowed by the size of inner hole of the vacuum vessel (see FIG.1.). The inductor is wound directly over the central column and forms non-separable unit together with it. The geometry of outer limbs stays basically unchanged. Nevertheless, improved grade for the conductor (see Table 1) as well as a new design for the magnet support structure was required due to enhanced mechanical loads. Simultaneously significant changes in the design were made for the top and bottom joints of the toroidal field coil (see FIG.2.). Each ganging between straight inner part and outer limb of the turn of the toroidal coil is provided with the help of three screw-bolts M12 and two pins of 14 mm in diameter. In addition all outer limbs are joined with each other at the top and at the bottom by means of FIG. 1. Spherical tokamak Globus-M2 magnetic system design. FIG. 2. Top (left) and bottom (right) joints of the toroidal field coil.

3 3 ICC/P1-01 screw-bolts M12, thus forming toroidal reinforcement belts. The influence of mechanical loads on the electric contact areas is diminished as far as possible. Flexible multilayer busbars are used as bonds. Belleville washers are applied in order to provide reliable electric connection. Poloidal field coils will be subjected to some alterations due to increased plasma current. Nevertheless they take its former places and allow to keep the full set of plasma magnetic configurations available in Globus-M. TABLE 1: MECHANICAL PROPERTIES OF MATERIALS, WHICH USED IN GLOBUS-M2 MAGNETS. Magnetic System Elements Toroidal Field Coil Supporting Structure Material Copper Alloy CuAg0.10P Stainless Steel 12Х18Н10Т (Russian grade close to AISI 304) Tensile modulus Е, GPa Yield strength σ 0.2, MPa Ultimate strength σ в, MPa Bolts, Pins Inconel Two basic operating modes are proposed for the upgraded tokamak Globus-M2: "B-max" mode maximum toroidal magnetic field B t0 = 1 T in the centre of the vacuum vessel (R 0 = 0.36 m), plasma current I p = 0.5 MA, plasma pulse duration Δt p = 0.3 s; "t-max" mode maximum plasma pulse duration Δt p = 0.6 s, toroidal magnetic field B t0 = 0.7 T in the centre of the vacuum vessel (R 0 = 0.36 m). 2. Basic Plasma Shot Scenario Simulation for the Globus-M2 Tokamak Plasma shot scenario simulations for the upgraded tokamak Globus-M2 were aimed at determination currents in the poloidal magnetic field coils in order to provide for appointed plasma current and plasma shape evolution. These data were used for the calculations of mechanical loads and heat generation in tokamak magnets. In the framework of the abovementioned concept two basic scenarios were developed. The first one (so called "Bmax") assumes tokamak operation with the maximal toroidal magnetic field of 1.0 T and plasma current of 0.5 MA. In this case the electric current through each turn of the toroidal coil reaches value of 110 ka. The ramp-up and ramp-down time is equal to 60 ms and determined by power supplies and inductance of the coil. It follows from calculations that plasma shot duration is restricted by toroidal field coil overheating for this case. Second scenario (so called "t-max") is aimed at the maximal duration of the shots with the same value of plasma current 0.5 MA. Toroidal field is reduced down to 0.7 T, which is equal to 77 ka current through the turn. In this case plasma shot duration is limited by volt-second capacity of the poloidal field coils. Also this regime might be used for noninductive current drive scenario application. Plasmas start-up takes place after the toroidal magnetic field reaches flat top. The plasma current ramp-up rate is determined as 10 MA/s for both scenarios. This value was chosen due to Globus-M experimental data analysis.

4 4 ICC/P1-01 Poloidal magnetic field coils and double-null plasma column symmetrical relative to the tokamak mid plane are sketched out in FIG.3. Poloidal field coil position and dimensions, which were used in calculations, are presented in the Table 2. TABLE 2: GLOBUS-M2 POLOIDAL MAGNETIC FIELD COIL DIMENSIONS AND POSITION. R(m) Z(m) ΔR(m) ΔZ(m) N PF PF PF PF PF PF CC CC CC CC CC CC CS CS The central solenoid (CS) of 1.3 m high is designed as a double layer coil with 59 turns per layer. The conductor for new central solenoid is worthy of being noted. It has reduced cross-section of mm 2 against mm 2 for the existing coil. As it will be shown bellow, no overheating occurred during plasma shot. Quite the contrary this circumstance let us enlarge cross-section for the central stack and reduce its heating. Moreover, the poloidal field coil flux consumption of the central solenoid increases from 0.3 Wb up to 0.34 Wb for double-swing regime and maximal current of 70 ka due to enlarged radius of the coil. Three pairs of coils (PF1, PF2, PF3) are responsible for the plasma shaping. Another three pairs of corrective coils (CC1, CC2, CC3), which are fed with the help of a single power supply, are mainly used to compensate stray magnetic field inside vacuum vessel during plasma breakdown. In a case of high plasma current these coils also provide plasma column equilibrium. All above FIG. 3. Double-null plasmas and poloidal field coil allocation.

5 5 ICC/P1-01 mentioned coils are fed with the help of six-pulse thyristor rectifiers. Simulations for the Ohmic and auxiliary (NBI) heated discharges were carried out. PET code [9] was applied. It deduces currents through the poloidal field coils in order to provide plasma column equilibrium with fixed total plasma current, plasma shape, surface flux, plasma pressure and poloidal flux profiles (alternatively, plasma pressure and safety factor profiles so called inverse problem). At the beginning of calculations the currents in the poloidal magnetic field coils were found out to suppress stray magnetic field from central solenoid as much as possible during plasma breakdown. Results of the poloidal magnetic field compensation are shown in FIG.4. One can see that the significant area inside vacuum vessel has the value of the poloidal magnetic field less than 1 Gs. With all this going on the central solenoid provides the poloidal flux consumption of 0.17 Wb for the 70 ka current through the coil in single-swing mode. Plasma loop voltage U L was derived with the help of ASTRA code [10]. The Globus-M experimental database was used. In the Globus-M2 Ohmic shots the loop voltage amounts to V during plasma current ramp-up and comes down to 0.8 V at the plasma current plateau. In the NB heated plasmas the loop voltage during plasma current plateau is slightly lower and is equal to 0.65 V. The poloidal field flux consumption in the double-swing mode of central solenoid operation is shown in FIG.5; corresponding currents through poloidal field coils are presented in FIG.6. It's necessary to note, that X-point position inside vacuum vessel influences significantly on the value of the current through PF1 and PF2 coils. We choose double-null configuration with moderate plasma elongation k 95 = As is seen from FIG.5, noticeable part of the poloidal field flux consumption is provided by "outside" coils. Plasma shot duration comes to ms FIG. 4. Stray poloidal magnetic field compensation. Fixed level of the poloidal magnetic field (in Gs) is shown by dashed lines. Y, Wb 0,15 0,10 0,05 0,00 Ics, ka Ics -> 0 I, ka PF1 PF2-0,05-0,10 <- Ycs CC -0,15-0,20 <- Ytot , t, ms -250 FIG. 5. Poloidal field flux consumption and current throw central solenoid. Doubleswing mode of central solenoid operation. -10 PF t, ms FIG. 6. Currents through poloidal field coils.

6 6 ICC/P1-01 for the maximal flux consumption. 3. Analysis of Thermal Conditions and Mechanical Stresses for the Globus-M2 Magnetic System Thermal analysis of the central stack revealed its significant overheating (ΔT > 100 C) after about 150 milliseconds of operating in the case of increased feeding current of 110 ka and unchanged crosssection of the conductor (Globus-M like) for the inner parts of the toroidal field coil. At the same time the temperature of the central FIG. 7. Comparison of the central stack and the central solenoid arrangement for the Globus-M and Globus-M2 tokamaks. Vessel border is shown by blue circle. solenoid stays within reasonable limits and does not restrict plasma shot duration. Water cooling application does not effect significantly on the temperature conditions and determines the interval between the shots only. Importent rearrangement of the stack and solenoid was made on basis of calculations. Taken into account the restrictions imposed on the central assembly by the dimension of the inner vacuum vessel hole, we enlarged the inner wedge cross-section from initial 425 up to 583 mm 2, reducing the gap between column and vacuum vessel and, as it was mentioned above, diminishing cross-section of the conductor for the central solenoid down to 286 mm 2 (see FIG.7.). At that the outer diameter of the assembly, which is equal to the central solenoid diameter, came to 210 mm, leaving 3 mm gap between solenoid and vessel wall. The design of top and bottom joints of the toroidal field coil was revised too. As it was said, flexible busbars were used as bonds to avoid electric contact deterioration due to "divorcing" of the mechanical joint. As evident from calculations, the renovated design of the toroidal field coil could provide the 1 T flat-top of the magnetic field FIG. 8. The temperature distribution (in C) along the toroidal field coil turn at the end of current pulse flat-top (left) and in 0.3 s after (right).

7 7 ICC/P1-01 for at least 300 ms period ("Bmax" mode), or 0.7 T during 700 ms ("t-max" mode). The temperature of the central rod and the central solenoid does not exceed 95 C for both cases. Some overheating up to 130 C could arise directly in the top and bottom electric contact areas at the end of B-max shot. Nevertheless, it is sharply localized far from inter-turn insulation. The temperature distribution along the toroidal field coil at the end of current pulse and in 0.3 s after is shown in FIG.8. It should be noted that significant local overheating (extra C) can arise in the FIG. 9. Toroidal field coil displacements (in m) under the maximal electromagnetic load at the end of tokamak shot and the maximal thermal load (t=347 ms, Ip=0). top and bottom joints of the toroidal field coil as a result of faulty contact, if the resistance increases from its regular value of 2 μω up to 5 μω (that is possible due to increased electromagnetic loads). Apparently temperature monitoring of the contact areas is required to avoid emergency conditions in operation. Stress analysis of the toroidal field coil and the supporting structure has been performed with the help of the finite element model, which included all the main elements of the construction. FIG.9. shows the toroidal field coil displacements under the maximal electromagnetic load at the end of tokamak shot without plasma current and the maximal thermal load (t=347 ms I p =0). Stress intensity in the outer limb of the toroidal field coil and in the most loaded upper belt zone is shown in FIG.10. Structural analysis shows, that the upgraded machine can provide more than operating cycles and 3000 of them with maximal values of the toroidal magnetic field and plasma current. The last restriction is connected with the fatigue strength of the bolts, which form the toroidal belts around the top and bottom joints of the toroidal field coil. This limitation can be removed by means of periodic replacing the bolts. FIG. 10. Stress intensity (in MPa) in the outer limb of the toroidal field coil (left) and in the most loaded upper belt zone (right) under the maximal electromagnetic load at the end of tokamak shot and the maximal thermal load (t=347 ms, Ip=0).

8 8 ICC/P Current Status of the Tokamak Upgrade The general design of the magnets is ready. The detailed design is close to completion. Raw wedges for the central stack were ordered at the KME Germany. The central stack is under construction now. The cable lines were replaced with copper bus bar lines to reduce Ohmic losses in the feeding circuits of the toroidal field coil and the central solenoid. That allows to increase slightly plasma current and pulse duration in the campaign ahead. The work was supported by RF Ministry of Education and Science (contr. No , No ) and by the RFBR grant References: [1] GUSEV, V.K. et al., Technical Physics, 44 (1999) No. 9, 1054 [2] BELYAKOV, V.A. et al., Plasma Devices and Operation, 9 (2001) No. 1-2, 39 [3] GUSEV, V.K., et al., Nucl. Fusion, 46 (2006) No. 8, S584 [4] GUSEV, V.K., et al., Nucl. Fusion, 49 (2009) No. 10, # [5] GUSEV, V.K., et al., Nucl. Fusion, 51 (2011) No. 10, # [6] KUTEEV, B.V., et al., Plasma Phys. Rep., 36 (2010) No. 4, 281 [7] GUSEV, V.K., et al. (Proc. of 38 th EPS Conf. on Plasma Phys. 2011), ECA Vol.35B (2011) P4.094 [8] GUSEV, V.K., et al., This Conference, EX/8-3 [9] GALKIN, S.A., et al., Nuclear Fusion, 37 (1997) No. 10, 1455 [10] PEREVERZEV, G.V., YUSHMANOV, P.N. // Max-Plank IPP rep (2002)