在tcv托卡马克中用中性束加热的一般性研究外文翻译

1、英文原文 A scoping study of the application of neutral beam heating on the TCV tokamak Alexa nder N. Karpushov a,?, Basil P. Duval a, Ren Chava na, Emilia no Fable b, Jea n-Michel Mayora, Olivier Sauter a, Henri Weisena a Ecole Polytechnique F ed /ale de Lausanne (EPFL), Centre de Recherches en Physique

2、 des Plasmas, Association Euratom-Conf d erationSuisse, CH-1015 Lausanne, Switzerland b Max-Planck-lnstitut f r Plasmaphysik, Euratom-IPP Association, Boltzmannstra 遝 2, D-85748 Garching, Germany A r t i c l e i n f o Article history: Available online 17 March 2011 Keywords: TCV tokamak Neutral beam

3、 heating A b s t r a c t The TCV tokamak contributes to the physics understanding of fusion plasmas,broadening the parameter range of reactor relevant regimes, by investigations based on an extensive use of the existing main experimental tools: flexible shaping and high power real time-controllable

4、electron cyclotron heating (ECH) and current drive (ECCD) systems. A proposed implementation of direct ion heating on the TCV by the installation of a neutral beam injection (NBI) with a total power of would permit an extension of the accessible range of ion to electron temperatures () to well beyon

5、d unity, depending on the NBI/ECH mix and the plasma density. A NBI system would provide TCV with a tool for plasma study at reactor releva nt ratios T and in inv estigat ing fast ion and MHD physics together with the effects of plasma rotati on and high plasmavsce narios. The feasibility studies fo

6、r a NBI heating on TCV presented in this paper were undertaken to construct a specification for the neutral beam injectors together with an experimental geometry for possible operational scenarios. 1. Introduction TCV is a compact (major radius , minor radius, toroidal magnetic field , plasma curren

7、t of ), high elongated (vessel elongation 3) toroidal fusion experimental machine. High power, real-time controllable, injection of waves at the second (X2, 3MW) and third (X3, 1.5MW) harmonics of electron cyclotron frequency constitute the primary method of heating (ECH) and driving non-inductive c

8、urrent (ECCD) in the plasma with electron densities , electron temperatures, ion temperatures . The flexible plasma shaping and powerful ECH system are used to contribute in many 1. areas of tokamak research High power X2-ECH, for relatively low density TCV plasmas ，does not allow operation at react

9、or relevant ratios of ion to electron temperatures , as the electron-ion classical Coulomb collision thermal equilibration time is significantly longer than the characteristic confinement times. Implementation of direct ion heating at the MW power level would allow the extension of to beyond unity a

10、nd fill the gap between present predominantly electron heated experiments and fusion reacto2r. The ion to electron temperature ratio is of particular interest in the projection of the transport mechanisms from existing experiments to burning plasma. The ratio plays a key role in the transition betwe

11、en ion temperature gradient (ITG) and trapped electron (TEM) mode dominated turbulent energy transport mechanisms. Increasingreduces the ion 3 and electron energy transport as observed in DIII-D H-mode experiments . NBI heat ing may therefore allow TCV plasmas to reach highervalues, close to the ide

12、al limit or beyond at high elongation. Injection of fast atom beams (NBI) into tokamak is a possible and well used method of auxiliary heating. Following ionization and charge-exchange, fast atoms of the beam are trapped as plasma ions and transport energy and momentum mainly to bulk ions if the fas

13、t ion en ergy is below critical en ergy (Ecrit -20 for hydroge n beam and deuterium plasma, )4. The proposed NBI system would thus also provide TCV 5 with a tool to investigate fast ion and related MHD physics as well as plasma rotation control 6 for which TCV is already well diagnosed. The behaviou

14、r of toroidal rotation in the vicinity of an ITB is of particular interest because of its in flue nee on triggeri ng an d/or susta ining the barrier. Target plasmas could in clude ITER-like H-mode shapes together with adva need shapes, rece ntly accessible only in ohmic regimes 【刀 * jD. :5廊日 l/Aaiin

15、gr # CQig NBsejft for ELMy H-mods %财w Fig. IL Electron and ion temperatures vs NB power for ELMy H-mode without and with 1.3 MW X3-EC heating 2. Scenarios of NBI heating experiments Experimental scenarios for the NBI experiments on the TCV are strongly linked to limitations imposed by ECH and ECCD.

16、For the elTBs and fully non-inductive scenarios on TCV, the accessible plasma density is limited by the X2 cut-off in curre nt drive and electro n heati ng experime nts. Conv ersely, efficie nt X3 depositi on is obtained for electron density in the range of and 81 The ASTRA code was used to simulate

17、 the plasma response to neutral beam heating in the geometry of the TCV tokamak. The code solves equations for electron and ion temperature and plasma current density with the prescribed electron density profile and total plasma curre nt take n from TCV experime nt. The use of the neoclassical ion h

18、eat conductivity 9 gives that is matched to the CXRS 10 measureme nt. The experime ntal electro n heat con ductivity was no rmalised to obtai n the en ergy confin eme nt time predicted by power law scali ngg1TPB98(y,2) for ELMy H-mode and sta ndard power law regressi on for L-mode. The EC powerdepos

19、iti on profile was calculated by the TORAY ray-traci ng code. 2.1. High density ELMy H-mode regime The target parameters for modelli ng were take n from Ohmic and X3 heated (Table 1, No. 1.0) statio nary ELMy H-mode phases of TCV discharge12. About 95% of injected deuterium NB power can be absorbed

20、by the plasma for tangen tially injected beam. The simulati ons show that can be achieved with7.8MWof NBI and 1.3MWX3-ECH (Figs. 1 and 2).Access to should be attai nable at in creased ()NB or reduced X2-ECH power. The fast ion charge-excha nge (CX) losses on backgro und neutrals strongly depend on t

21、he first wall recycling conditions, the density of background atoms ,obtained from EIRENE modelling, reduces the NB heating efficie ncy by 15% (No. 1.4), CX losses on beam n eutrals are n eglectable. Fig_ 典IB heating vs f hin in Hi-mcHdc for tangential co- and countcr-lr and normal co-Jp irijcctioni

22、- At high plasma den sity and curre nt, n eutral beam inject ion could result in an in crease of the thermal from 2.0 (pure 1.5 MW X3-ECH) to 2.6 (2MW NBI), and could even reach the ideal MHD limit (H3) resulting from the fast particle contribution. Fast ion slowing down times in such regimes are of

23、 the order of ,i.e. shorter or comparable with the bulk plasma en ergy confin eme nt time, so, perturbati on of the ion en ergy Maxwellia n distributio n by fast ions is expected to be small (as in a fusi on reactor). 2.2. X2-EC and NBI heat ing Modelling of NB heating in low density regimes was per

24、formed for 2MW X2-EC heated L-mode reference discharge (#31761, No.2.0). Increase of the NB deposited power per plasma ion at low den sity results in-2 times lower () tha n in high den sity regime NBI power required to access of(scenarios 2.1 and 2.2 and Fig. 3). Near-normal NB injection cannot be c

25、onsidered here due to higher shine-through losses, resulting in first wall overheat of the TCV central column. ASTRA simulations confirm earlier experimental and numerical studies of fast ion orbit losses on the TCV 13 . At low plasma current, fast ion orbit losses are extremely important and become

26、 substantial for counter-Ip NB injection (Fig. 4); losses increase at high ion energy (32% for D-NB and 59% for , scenarios 2.4 and 2.7) and for higher NB atomic mass. NB injection at low plasma density and current provides the possibility to study the fast ion and MHD physics. In the unfavourable s

27、cenario (like 2.4), the delivered by the NB power leads to the creation of a strong fast ion population with a stored energy of few tens kJ that, at low current, significantly contributes to the ideal MHD limit. Fast particle instabilities would dominate the plasma behaviour under these conditions 5

28、. 3. Neutral beams injection layout TCV was not originally designed for neutral beam heating although several relatively wide machine midplane lateral ports were implemented for general diagnostic flexibility. The location of magnetic field coils, for which modification is not feasible, and the exis

29、ting support structures are major problems for NBI plasma access, in particular for the tangential injection direction. Access for NB injectors through 15cm diameter ports TjHc t Paramelrrs oFlicdled scenarios wilh drurtcrium (D-N Eand hydrogen (H-NB).也ngunti柑 ngurwy radiits- distanceIwtwrcn beam ax

30、is and tokanuk major axis. Run 74cni) and nanrul (= 23 cm NS injection. Half width for Gamsu 口 INB power distribution in dw taka male is 10 cm. NB power frartionE with Full, halFand (?31 erwrgj aic 64. Z4 彳口d aa.- central ion and sleetnxi temperamn: PhfP - km jmd -cfctCTCfflj hcjtuig power from East

31、FW - ncutijl beam, power without shine-trnugh arid fast ian wbt losses;科利-power of ohmic hearing, P. - or-electron ciassical Coulomb cli si an ihermal cquillibirjtbDai powci*; rH - bulk pljsmj energy cifi.rH!fnent time. No. ScimairiqparamcCrr r.JTctOKkcV) Pfc/Rw(kW) 躲朋jh tkW ftc(kW) te tms) with FJ.

32、WWXJECW.吟他卜 7.8 R PdP in X3-ECHjMMW l.D TCV#29475 曲 1 fl- l J 材矗neiKE. IJ02/2.73 afo 021-0 -193 3.3 0.77 MW.為 keV. D-NQ. CO-JpB 0/2馆 631.00 741i24S 137 1.00 MW. 25 keV.Cn-JpLTjnp-nt i.U 2.4R/2.06 目旳斡站 知门sa 398 13 2.00 MW, ZSIkeV, O-NB, COp.tangcfibal 353/1.73 15001/119 18811503 767 Hol IJ2 mltiCXlns

33、sn. rtiD(ljCS)-5 k IS15 m ? 232.17 7(3113 967/24 NB usage at low densities is, however, severely limited by excessive shine-through and high inner wall power loads. The maximal acceptable power load of 7.6MW/ for a 1 s duration leads to temperature rise of graphite inner wall tiles!14 of 1000K corre

34、sp onding to shine-through of the 1MW beam with the 15cm foot-print size. A model of a n eutral beam with geometric focuss ing and an gular diverge nc!e5 was performed to calculate the beam tran smissi on and power load on the critical scrapers in the NBI duct. The acceptable。 beam power transmitted

35、 into the tokamak for 1MW, , 1 s beam with 200mA/cm2 extracti on curre nt from the ion optical system located at about 250cm from the TCV port is feasible only with low beam diverge nee: 0.7/0.8for ?10/15cm duct apertures respectively. The tran smissi on of the high power NB through narrow ports dem

36、a nds high curre nt den sity, low diverge nee n eutral beam in jector only reachable, at prese nt, by lower curre nt diag no stic n eutral beams. To allay these requireme nts on beam diverge nee and curre nt den sity a modificatio n TCV vacuum vessel to create new port(s), specifically designed for

37、NBH and fitted between magnetic field coils, is considered. The available gaps between toroidal and poloidal magnetic field coils at the TCV midpla ne are 22cm in vertical and 38cm in toroidal directi on. The desig n of duct with inner =3*帆川口 p-c-wwThtfS-ECM Fg- OKMeniLinn* -and hdroce-ni NB h&M inc

38、 Rw L mode wrh 2: MW X2- EC tie-iEinc Hk. 4. cd- nd igrw亡卩戸 nhvt in L-rvudv T S T Fig- 斤-ill iij-i I i njcct iciri i rii531-meriit f i_ iiivkII n fic-at imi riF VjJc- 1OOQ &oa &00 400 200 1QQ Dudmm Fig 6. NE3 tangency radius vs rnaxirna.1 beam duct diameter. EE -5&C蚤逮 1 /ow fi&id scte wail 十 1 L H 丄

39、 H W W ， minimal aperture of 20 cm, wall thickness 1 cm and 3 cm gaps to toroidal field coils, beam axis tangency radius of 74cm (Fig. 5) was found to be feasible and permits to transmit 90% of the NB power to the plasma for 1MW, deuterium beam with diverge nee (reachable for heat ing beams). The re

40、lati on betwee nand beam duct aperture horiz on tal size for chose n duct wall thick ness and gaps to toroidal coils is show n in Fig. 6. To reduce beam block ing by desorbed gas in the n arrowest part of the beam duct (close to the tokamak entran ce), differe ntial duct pump ing is required. This g

41、eometry could permit two NB injectors (aimi ng in co- and coun ter-curre nt directi ons) on the same port. With proper power adjustme nt, one could obta in scenarios with balanced momentum transfer to the plasma. 4. Conclusion Installation of 1MW, , deuterium, tangential (basic reference) neutral be

42、am injector would significantly increase the experimental capability of the TCV tokamak by extending the operational domain at higher ratio and plasma pressureand widening H-mode operational domain (especially at high density). 1MW of injected power is sufficient to access , taking into account 20%

43、CX fast ion losses on background neutrals. Two balanced co- and counter-Ip orientated injectors with total power of 2MWwould permit the investigation of the effects of NB induced plasma rotation, to reach vratio 2 and study fast ion behavior MHD physics in seenarios such as stationary ELM free H-mod

44、es and fully non-inductive electron internal transport barriers. Lowering the beam energy results in a decrease of the on-axis ion heating power density by broadening the NB deposition profile. At higher beam energies, fast ion orbit losses strongly reduce the heating efficiency, especially for coun

45、ter-beam alignment (Figs. 2 and 4). For a given injection energy and target plasma parameters, the fraction of NB power delivered to bulk ions is higher and shine-trough losses are lower for deuterium beam than for hydrogen (Fig. 4). Due to unacceptable shine-through power load on the central column

46、, only double-path tangential NB injectio n is acceptable for in termedate and low plasma den sities (4x 1019m-3, for deuterium beam). The capability of the NBI operation to use hydrogen ions is essential (1) for on-axis ion heating at high () plasma density and (2) to reduce orbit losses of coun te

47、r-injected fast ions at low ( plasma curre nt. Adjustable beam en ergy of 1530 and likely a wider range should satisfy the concept of TCV a very flexible tokamak and permits to adjust beam power by simultaneous change of beam energy and ion current (maintaining optimal perveance, relationship betwee

48、n beam energy and current). Acknowledgments This work was supported in part by the Swiss National Science Foundation. The authors are grateful to Prof. A.A. Ivanov, Prof. V.I. Davydenko and Dr. T.D. Akhmetov for useful discussions and developing of the neutral-beam propagation code 15. 翻译 在TCV托卡马克中用

49、中性束加热的一般性研究 摘要： TCV托卡马克以现有的实验工具（可形变高功率实时可控的电子回旋共振加 热装置（ECH和电流驱动系统（ECCD为基础进行大量研究，对聚变等离子体 的物理解释和扩展聚变堆相对温度的范围有贡献。 用总功率为的的中性束注入装 置（NBI）,预期可实现在TCV中对离子进行直接加热，使得依赖于NBI/ECH和等 离子体密度的离子与电子温度之比的范围扩大到（）超过单位值。 NBI 系统将为 TCV寸的聚变堆内等离子体的研究和快子 MH吻理结合等离子体旋转效应高B值 的研究提供有力工具。本文对TCV中NBI加热的可行性研究，是通过建立专门的 中性束注入系统和适用于可操纵情况的实

50、验几何尺寸进行的。 关键词：TCV托卡马克,中性束加热（NBH 1. 序总 TCV是一种结构紧凑（主半径，次半径，环状磁场，等离子体电流），高度拉伸 （管拉伸率为 3）的环状聚变实验装置。高功率实时可控，注入波在第二（X2， 3MW和第三（X3,1.5MW电子回旋谐振频率构成主要的加热方式（ECH。在等 离子体中进行非感应电流驱动（ECCD等离子体的电子密度为，电子温度为，离 子温度，可形变等离子体和高功率 ECH系统被用于托卡马克许多方面的研究。 高功率X2-ECH对于相对密度低的TCV等离子体，将不容许在离子电子温度 比的情况下操作， 因为电子离子的经典库伦碰撞热平衡时间比特征约束时间长得

51、 多。在功率为MWt级的情况下对离子进行直接加热，可使增大超过单位值，从 而填补了目前占主导地位的电子加热实验和聚变堆之间的差距。 离子与电子温度 之比在从现有实验到聚变等离子的输运机制的研究计划中起特别重要的作用。 在 离子温度梯度（ITG）和束缚电子模式（TEM之间的转变起关键作用，决定了紊 流能量的输运机制。在DIIF-D H-模式实验中当增加可观察到离子和电子能量输 运的减小。因此NBI加热容许TCV等离子体的B值更高，接近理想极限，甚至在 高拉伸率装置中可超过理想极限 将高速原子束注入托卡马克装置是一种可能的好的辅助加热方法。 但是如果 快离子能量低于临界值（对于氢原子束和氘等离子体

52、 Ecrit , ）能量和动量输 运主要由大量离子完成，将出现电离、电荷交换、快原子束被束缚等现象。预期 的NBI系统只要旋转等离子体控制TCV已被很好的诊断，也将为TCV提供研究快 离子和研究相关的MH勣理提供工具。其中ITB附近的环状旋转行为尤为重要， 因为它将影响触发或持续势垒。目标等离子体能够包括类ITER H模式形状和目 前只在欧姆范围内接近的先进形状。 2. NBI 加热实验的相关情况 TCV托卡马克装置中的实验参数与 ECHffi ECC啲受迫极限有很大关系。对于TCV 中的elTBs和完全非感应情况，可得到的等离子体密度被X2的电流驱动系统和 电子加热装置所限制，与此相反，X3

53、的有效电子密度要求在的范围内，要求电子 温度。 ASTRA被用来模拟在TCV勺几何尺寸条件下，等离子体对中性束加热行为的反应。 这一程序通过求解， 在实验中已得到的电子密度和总等离子体电流的条件下， 离 子和电子的温度以及等离子体电流密度。 发现利用新古典离子加热电导率 （）得 到的值与CXRS勺测量值相符。实验电子加热电导率（）被规范后可以得到可由 得到能量约束时间（IPB98 （y,2 ） ELMyH模）。ECH的能量沉积由TORay射线跟 踪软件计算得到。 2.1高密度ELMy H模式范围 模型的目标参数是通过欧姆加热和 X3加热释放的定态ELMy H扌莫式、对于切向 入射的能量为的氘中

54、性束约有 95%被等离子体吸收。模拟表明用0.8MW中性束注 入（NBI）和1.3MW的 X3-ECH可得到，通过增大中性束的功率（）或减小X2-ECH 的功率，使也能达到。中性束背景情况下快离子电荷交换损失（CX与第一层循 环条件，背景原子密度。通过 ECREN模型得到，将使中性束加热效率减小将近 15%中性束电荷交换损失率V 2%,可忽略不计。 2.2 X2-EC 和中性束加热 低密度中性束加热模型， 增加中性束能量沉积功率使得低密度等离子体中的离子 比高密度中的低2倍。NBI系统的功率要求接近由于更高的（）的shine-through 损失，将导致TCV中心圆柱第一层被过度加热，故在此不

55、能将中性束认为是近似 正常中性束注入。 ASTRA模拟值与更早的TCV中快离子损失的实验数值相符合。 在低等离子体电流 情况下，快离子的轨道损失是极其重要的，而且使 NB注入变得。能量损失随中性束原子质量的增加及离子能量的增高而增加。低 密度等离子体和低等离子体电流的中性束注入，为快离子和mhD理的研究提供 了可能。当不合适的参数（如 2.4 ）由MD传递的（）功率导致产生了能量为几十 千焦的大量快离子，在低电流时，对于MHD勺B理想极限有突出作用。在这些条 件下，快粒子不稳定性将决定等离子体行为。 3. 中性束注入设计 TCV最初并不是为中性束的加热设计的，虽然为了便于实现对等离子体进行 整

56、体诊断， 开设了几个相对较宽的横向窗口。 不可调节的磁场线圈排布和现存的 支撑结构设计，特别是切向注入方向是NBI设计的主要问题。中性束注入装置通 过一个直径为15cm的正常注入窗口（切向半径v 23cn）和一个直径为10cm的切 向注入窗口附近的单孔，注入束的轴向半径=65cm在文献13中已进行过详细分 析。当中性束的密度过高时，=23cm处的发光效应就会起作用，然而对于低密度 情况中性束的利用被过度的发光损失和内壁高功率荷载所严格限制。 最大可接受 功率荷载为7.6MW/m2其中尺寸为15cm的1MW勺中性束可导致在1s内石墨层内 壁的温度上升，当发光损失率为 10%时，可使温度上升 10

57、00K。 一种几何聚焦和角度偏离的中性束模型， 被用来计算中性束输运和中性束的 临界功率荷载。1MW 25keV勺中性束可将80%的功率输运进入托卡马克中。位于 离TCV窗口约250cm处的离子光学系统中性束的外部电流为 200mA/cm2仅对于 低的束偏离0.7/0.8，直径为10/15cm的注入孔才可实现。目前通过低电离诊断 中性束，高功率输运（0.61.0MW的中性束通过窄的窗口，需要高电流密度， 低角度偏离的中性束注入装置才可达到。 为了降低对中性束角度偏离和电离密度的要求，需要改变 TCV真空管的设 计，设立新的窗口，特别是为了中性束加热和合适的 线圈。在TCV中换装和径 向磁场线圈

58、的间隙是：垂直方向为 22cm环状方向为38cm这种管道设计为内 层最小孔径为20cm壁厚1cm有3cm的环状磁场线圈间隙、中性束轴向切向半 径为74cm发现这种设计方案可行，可实现对功率为 1MW 25keV、偏离度小于 1.2 （可实现的加热中性束）氘束完成 90%的功率输运。在图 6 中画出了中性束 切向半径Rtan、中性束孔径尺寸、壁厚、环状线圈间隙之间的关系。为了减小 靠近中性束注入托卡马克入口处的最窄部分的管内气体影响， 需要有不同的管道 抽气装置。 这种几何尺寸可容许在同一窗口安装两个中性束注入装置。 通过适当 的功率调节，可得到由平衡动量转移到等离子体中的参数。 4. 结论 1MW、25keV、氘中性束，切向中性束注入装置将通过扩大可操纵范围：提高 离子电子温度比Ti/Te、增大等离子体压强（B）、扩展高密度区的H-