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2023年4月19日发(作者：二年级看图写话)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 d /ale de Lausanne (EPFL), Centre de Recherches
e
en Physique des Plasmas, Association Euratom-Conf d rationSuisse,
e
CH-1015 Lausanne, Switzerland
b Max-Planck-lnstitut f r Plasmaphysik, Euratom-IPP Association, Boltzmannstra 2, D-85748 Garching,
遝
Germany
A r t i c l e i n f o
Article history: Available online 17 March 2011 Keywords: TCV tokamak Neutral beam 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 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 beyond 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 plasmasce narios. The
v
feasibility studies for 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 current

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 current (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
areas of tokamak research
[1].
High power X2-ECH, for relatively low density TCV plasmas ，does not allow operation
at reactor 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 and fill the gap between present predominantly electron
heated experiments and fusion reactor. The思绪万千的意思 ion to electron temperature ratio is of
[2]
particular interest in the projection of the transport mechanisms from existing experiments
to burning plasma. The ratio plays a key role in the transition between 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 higher\"values, close to the ideal 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 fast
ion en ergy is below critical en ergy (Ecrit -20 for hydroge n beam and deuterium
plasma, ) The proposed NBI system would thus also provide TCV
[4].
[5]
with a tool to investigate fast ion and related MHD physics as well as plasma rotation
control for which TCV is already well diagnosed. The behaviour of
[6]

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 inin酒逢知己千杯少下一句是什么 g 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. :5l/Aaiingr # CQig for H-mods
廊日
NB sej^ftELMy
*
%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. 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 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 heat
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
predicte莫道桑榆晚为霞尚满天类似的诗句 d by power law scali nggTPB98(y,2) for
1]
ELMy H-mode and sta ndard power law regressi on for L-mode. The EC power

depositi 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 discharge. About 95% of injected
12
\"
deuterium NB power can be absorbed 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 the 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_ IB heating vs f hin in Hi-mcHdc for tangential co- and countcr-l and normal
典
r

co-Jp irijcctioni-
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 (3) resulting from the fast particle
H
contribution. Fast ion slowing down times in such regimes are of 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 performed 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 considered 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 ar唐诗中秋月李峤其一 e extremely important and become
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 scenario (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].
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 mo贫贱夫妻百事哀上一句是什么意思 dification is not feasible, and the existing
support structures are major problems for NBI plasma access, in particular for the
tangential injection direction. Access for NB injectors through 15cm diameter ports
TH t
JC
Paramelrrs oFlicdled scenarios wilh drurtcrium (D-N Eand hydrogen (H-NB).[nguntingurwy radiits- distanceIwtwrcn beam axis and tokanuk major axis. Run■ 74cni) and nanrul (= 23 cm}
〕也』{柑
NS injection. Half width for Gamsu INB power distribution in dw taka male is 10 cm. NB power frartionE with Full, halFand (]?31 erwrgj aic 64. Z4 d aa. - central ion and sleetnxi
口彳口
temperamn^: PhfP^ - km jmd -cfctCTCfflj hcjtuig power from East FW - ncutijl beam, power without shine-trnugh arid
fast ian wbt losses;power of ohmic hearing, P. - or>-electron ciassical Coulomb c^li si an ihermal cquillibirjtbDai powci*; r - bulk pljsmj energy c!fnent time.
科利-
H
No. Scimairiq\'paramcCrr kcV) Pfc/Rw(kW)
l.D TCV#29475 1 fl- l J neiKE. IJ02/2.73 0^21-0 -193
3.3 0.77 MW.keV. D-NQ. CO-J 63^1.00 741i\'24S 137
13 2.00 MW, ZSIkeV, O-NB, CO^ibal 353/1.73 15001/119 18811503 767

with W.7.8 R PdP in X3-ECH jMMW
吟他卜
曲材『矗
為
pB
1.00 MW. 25 keV. Cn-JpTjnp-nt i.U 2.4R/2.06
L
H IJ2 mltiCXlnssn. rti](ljCS)-5 IS m 23^2.17 7(3^113 967/24&
OLK
D15?
1.00 MW, 25 keV, O-NB, CNlft4. tangcnl ial 152 1.5
p
1.00 MW, 25 keV, Q-NB, C0-rf rwmuL 225 697/110 823/251
pp
1 JOO MW, 25 keV, H-Nfl, COngcflitial 1.7 14^2.20 3%3.l\'242
••饰胡也说
TtV#317fi] at (refen?iKe. K2) 2.D 0.75^4.2] -52 73
0.30 MW. 25icev, IPNB,CMpHtinemt
匚碍
⑹
0.50 MW. 25fceV. D-NQ.1 S2.00 15B 337^91 4S7f-92 6J9
匚口斗中口什昶师也
1 .oa MW. 2 5 kev. e NB. cn-J un^nEiil
pu
MW, 25 . CNTR4p
B
口册细”卽
0.5fl MW, 25lkcV H-N9, COtangemI 2.5 X41/2.21 25|\'1.57 ^57/82
t
•乩竝
0.5fl MW, 50 IkcV, O-NB, CO-J tan^cmUal 243/110 93
pp
030 MW, SO . CNTR-Jp. tangcnlial 2J 1J81/3J5 85/7

躲朋
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ftc(kW) tms)
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i-jnodf wth 2MWX2 EtHJTrfO^-S -2Z5 kA
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11
with near normal injection (tangency radius) and through a single ?10cm aperture n ear-ta
ngen tial inject ion port with the axis pass ing n ear the inner wall at has bee n analysed in
[ 13]
. Shine through for is workable at the high densities; 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 of 1000K corresp onding to shine-through of the 1MW beam with
!
14]
the 15cm foot-print size.
A model of a n eutral beam with geometric focuss ing and an gular diverge nce was
!
5]
performed to calculate the beam tran smissi on and power load on the critical scrapers in
the NBI duct. The acceptable^beam power transmitted into the tokamak for 1MW, , 1 s
。％
beam少年游柳永长安古道马迟迟 with 200mA样的拼音 /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
dema 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 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*p-c-ww
■■帆川口
Th tfS-ECM
F>g- O^KMeniLinn* -and h^droce-ni NB h&M inc Rw L mode w^rh 2: MW X2- EC tie-iEinc~
H.-
KCDNH
4. nd igrw in L-rvudv
亡卩」戸」
vt
€T S T
Fig- ill iij-i I i njcct ic^iri ^i rii531-meriit f i n fic-^at imi riF
斤-
_
IIIVKJC
IIVJ
-
1

1OOQ
&oa
&00
400
200
1QQ

/ow s^cte
fi&idwail
十
1
蚤
逮

EE -5&
&C
H ■ ■ ■ W W ■ ■ ■，
L
H

丄 ___ ____
Dud mm
Fig„ 6. NE3 tangency radius vs rnaxirna.1 beam duct diameter.

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, deute黄台之瓜 rium beam with diverge nee
(reachable for heat ing beams). The relati 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 geometry 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 beam 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% 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-modes 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 counter-
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, only double-path tangential NB injectio n is acceptable for in
termedate and low plasma den sities (<4 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 ter-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 between 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]
x

翻译
在TCV托卡马克中用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值的研究提供有力工具。本文对TCV中NBI加热的可
B
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等离子体的值更高，接近理想极
B
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 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 ECHECC啲受迫极限有很大关系。对于TCV 中的
ffi
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2%,可忽略不计。
V
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，为快离子和理的研究提供
MH
D^
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勺理想极限有突出作用。在这些条件下，快粒子不
B

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23cn）和一个直径为10cm的切向注入窗口附近的单孔，注入束的
v
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 1000K。
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（0.6~1.0MW
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 25keV、偏离度小于
1MW
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（、扩展高密度区的H-模的可操纵范围1MW勺注入
B）
功率可达到〉2keV,考虑20%勺中性背景条件下电荷交换（CX 快离子损失。两种不同方向

的平衡注入装置（Co-和Counter-）,总功率为2MW，可用来研究中性束感应等离子体旋转
效应，以及的快离子行为。 MHD物理在比如
定态ELM自由的H模式和完全非感应中内部输运壁垒的研究。降低中性束能量将导致轴
向离子加热功率密度降低。由于扩宽了中性束（NB的沉积分布。在更高的束能量时，快离
子轨道损失导致强烈减少中性束的加热效率，尤其是 Counter- 伐檀原文带拼音中性束如图（ 2）和图
（4）所示。在给定注入能量和目标等离子体参数的情况下，氘中性束的小部分中性束功率
进入大量离子的比率比氢高， shine-through 损失比氢的低，如图（ 4）所示。由于在中心
柱上的不可接受的 shine-through 功率荷载，对于中低密度中性束，只有通过双重路径中性
束注入才能达到目的。利用氢离子的中性束注入的必备条件：
（ 1）轴向离子加热要求中性束密度高
（2）减少Counter-l 在低等离子体电流，注入快离子轨道损失。中性束能量可调节范围
p
为，似乎需要一个更宽的范围才能满足要求。允许通过中性束与离子电流的同步改变调节
中性束功率。
致谢
这一工作得到了 Swiss 国家科学基金的部分支持，作者对 A.A. lvanov 教授, V.l.
Davydenko教授和T.D. Akhmetov博士的有效讨论和中性束传播程序的开发表示感谢。
目录

1
总论错误！未定义书签
1.1
项目摘要 ............................ 错误！未定义书签
1.2
编制依据与研究范围 .......................... •错误！未定义书签
1.3
建设规模 .................................... 错误！未定义书签
1.4
主要建设内容 ................................ 错误！未定义书签
1.5
投资估算及资金筹措 .......................... •错误！未定义书签
1.6
工程效益 .................................... 错误！未定义书签
2
投资环境及建设条件分析 .................. 错误！未定义书签
2.1
投资环境分析 ................................ 错误！未定义书签
2.2
建设条件分析 ................................ 错误！未定义书签
3
项目建设的必要性与可行性分析 .................... 错误！未定义书签
3.1
项目建设的必要性 ............................ 错误！未定义书签
3.2
项目建设的可行性 ............................ 错误！未定义书签
4
开发区规划与交通量预测 .................. 错误！未定义书签
4.1
项目区总体规划 .............................. 错误！未定义书签
4.2
项目影响范围的交通量预测 .................... 错误！未定义书签
5
工程建设方案 ............................ 错误！未定义书签
5.1
设计标准及设计规范 .......................... •昔误！未定义书签
5.2
道路设计方案 ................................ 错误！未定义书签
5.3
给排水工程设计 ..............................
错误！未定义书签

5.4
道路照明错误！未定义书签
5.5
电力综合管沟
错误！未定义书签
5.6
道路绿化工程
错误！未定义书签
5.7
交通安全及管理设施
错误！未定义书签
节能分析错误！未定义书签。
错误！未定义书签。
错误！未定义书签。
错误！未定义书签。
错误！未定义书签。
6.1
设计依据
6.2
项目概况
6.3
项目对所在地能源供应状况的影响
6.4
项目用能方案、用能设备

6.5
项目能源消耗量、能源消费结构、效率水平和能源管理水平错误！未定义书签
6.6
节能措施分析评价

6.7
节能措施建议
6.8
结论
7
环境保护
7.1
大气环境质量
7.2
交通噪声
7.3
振动环境质量
7.4
日照环境质量
误！未定义书签
错误！未定义书签。
错误！未定义书签。
错误！未定义书签。
错误！未定义书签错
误！未定义书签错
误！未定义书签错
误！未定义书签错
8
组织机构与人力资源配置
8.1
施工组织机构
8.2
项目部的职责
.................... 错误！未定义书签
........................... 错误！未定义书签
........................... 错误！未定义书签

8.3
人力资源配置错误！未定义书签。
9
项目实施进度 .................................... 错误！未定义书签。
9.1
建设工期 ................................... 错误！未定义书签。
9.2
工程实施进度安排 ........................... 错误！未定义书签。
10
征地拆迁 ....................................... 错误！未定义书签。
10.1
项目建设用地面积 .......................... 错误！未定义书签。
10.2
建设用地现状 .............................. 错误！未定义书签。
10.3
征地赔偿 .................................. 错误！未定义书签。
10.4
道路两侧需拆除建筑物与安置办法 ............ •错误！未定义书签。
11
投资估算与资金筹措 ............................ 错误！未定义书签。
11.1
投资估算 .................................. 错误！未定义书签。
11.2
投资筹措 .................................. 错误！未定义书签。
11.3
工程利润分
析 .............................................. 错误！未定义书
签。
1250
招标投标管理........................................................
12.150
招标依据 ......................................................
12.250
招标范围 ......................................................
12.350
招标组织形式 ...................................................
12.450
招标方式 ......................................................

13 53
社会影响分析
1 2. 5
招标遵循的原则
51
13.153
优化投资环境，提升开发区服务功能 ..............................
13.254
项目对社会的影响分析 ..........................................
13.354
项目与所在地互适性分析 ........................................
13.4 54
社会评价结论 ..................................................
1455
结论与建议 ..........................................................
14.1 55
结论 ...........................................................
14.2 55
建议 ...........................................................