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BAKHAREV et al.

1

TOKAMAK RESEARCH IN IOFFE INSTITUTE

N.N. BAKHAREV, G.I. ABDULINA, V.I. AFANASIEV, A.B. ALTUKHOV, L.G.

ASKINAZI, N.A. BABINOV, A.N. BAZHENOV, A.A. BELOKUROV, I.M. BUKREEV,

F.V. CHERNYSHEV, I.N. CHUGUNOV, D.N. DONNIKOV, L.A. ESIPOV, D.B. GIN,

A.D. GURCHENKO, E.Z. GUSAKOV, V.K. GUSEV, M.V. ILIASOVA, M.A. IRZAK,

E.M. KHILKEVITCH, N.A. KHROMOV, E.O. KISELEV, V.A. KORNEV, A.N. KOVAL,

D.V. KOUPRIENKO, S.V. KRIKUNOV, O.L. KRUTKIN, G.S. KURSKIEV, S.I.

LASHKUL, S.V. LEBEDEV, A.D. MELNIK, V.B. MINAEV, I.V. MIROSHNIKOV, E.E.

MUKHIN, V.O. NAIDENOVA, A.N. NOVOKHATSKII, A.S. NAVOLOTSKY,

K.YU. OSHUEV, M.I. PATROV, M.P. PETROV, S.Ya PETROV, YU.V. PETROV, I.A.

POLYNOVSKY, A.YU. POPOV, A.G. RAZDOBARIN, D.V. RAZUMENKO, V.V.

ROZHDESTVENSKY, N.V. SAKHAROV, D.S.SAMSONOV, A.N. SAVELIEV, P.B.

SHCHEGOLEV, A.E. SHEVELEV, A.D. SLADKOMEDOVA, A.I. SMIRNOV, V.V.

SOLOKHA, V.A. SOLOVEI, A.YU. TELNOVA, V.A. TOKAREV, S.YU. TOLSTYAKOV,

P.V. TRETINNIKOV, A.S. TUKACHINSKY, V.I. VARFOLOMEEV, N.A. ZHUBR

Ioffe Institute

St.-Petersburg, Russia

Email: [email protected]

M.D. BLEHSHTEIN, V.V. BULANIN, A.V. PETROV, A.YU. YASHIN Peter the Great St. Petersburg Polytechnic University


E.N. BONDARCHUK, S.N. KAMENSHIKOV, A.A. KAVIN, K.M. LOBANOV,

A.B. MINEEV JSC «NIIEFA»,


AL.P. CHERNAKOV JSC «Spectral-Tech»,


T.P. KIVINIEMI, S. LEERINK, P. NISKALA Aalto University

Espoo, Finland

A.V. GORBUNOV NRC Kurchatov Institute

Moscow, Russia

C. LECHTE

Institute of Interfacial Process Eng. and Plasma Technology

70569 Stuttgart, Germany

E.G. ZHILIN Ioffe Fusion Technology Ltd.


S. HEURAUX

Institute Jean Lamour UMR 7198 CNRS, Université de Lorraine

54000 Nancy, France

S.P. PANDYA, S.N. PANDYA

Institute for Plasma Research

Gandhinagar, 382-428, India

OV/5-4

Abstract

The recent tokamak researches in Ioffe Institute are described: energy confinement, Alfven eigenmode, discharge

disruption and SOL width studies in the Globus-M spherical tokamak; ICE, Alfven waves in OH plasma and L-H transition

studies in TUMAN-3M; full-f global gyrokinetic modelling benchmark using synthetic diagnostics in FT-2. Anomalous

absorption and emission in ECRH experiments due to parametric excitation of localized UH waves is described. Progress in

the development of the neutral particle analysis, Gamma-ray spectrometry and divertor Thomson scattering combined with

laser-induced fluorescence diagnostics for ITER is discussed. Globus-M2 status is reported.

1. INTRODUCTION

In 2018, Ioffe Institute celebrated its 100th

anniversary. Plasma physics and thermonuclear fusion researches in

Ioffe Institute has been going on for more than 60 years. At present, research of various aspects of tokamak

physics is conducted on the small tokamaks at Ioffe Institute in a wide range of experimental conditions:

R/a=1.6, BT=0.5(1.0) T, Ip=250(500) kA – Globus-M(M2), R/a=2.4, BT=1.0 T, Ip=180 kA – TUMAN-3M,

R/a=7.0, BT=3.0 T, Ip=40 kA – FT-2. Besides that there is cooperation with other tokamaks all over the world.

Also three ITER diagnostics – neutral particle analysis, Gamma-ray spectrometry and divertor Thomson

scattering combined with laser-induced fluorescence – are being developed. In this overview recent results of

the tokamak researches in Ioffe Institute are described.

2. GLOBUS-M RESULTS

Globus-M [1], which has recently been replaced by a new Globus-M2, was a compact spherical tokamak (ST)

(R ≈ 36 cm, aspect ratio A ≈ 1.5, elongation k<2) equipped with NBI (Neutral Beam Injection) and ICR (Ion

Cyclotron Resonance) heating and LH (Lower Hybrid) current drive systems. It’s distinguishing features were

high normalized Larmor radius and collisionality, high heating power density and small plasma-wall distance.

The main results, dedicated to the study of energy confinement [2], Alfven eigenmodes, discharge disruption

and SOL study are described.

2.1. Energy confinement

The thermal energy confinement study was performed in both OH and NBI heated H-mode deuterium plasma

using 26-28 keV 0.35-0.75 MW deuterium beams. The range of the major plasma parameters was as follows:

Ip=0.12-0.25 MA, BT=0.25-0.5 T, absorbed heating power Pabs=0.2-0.8 MW, averaged electron density ne=1.8-

5.5 1019

m-3

. The H-factor was in the range 0.5-1.3. Since IPB98(y,2) functional dependence is poorly suited to

describe the experiment, the regression fit was performed as follows WGLB

=C∙Ip-αI

BTαB

PabsαP

neαn

(Fig. 1). The

following fitting parameters were obtained C= 9.07±0.95, αI=0.77±0.08, αB=1.30±0.06, αP=0.2±0.04,

αn=0.67±0.04 while RMSE=25%. The most important physical quantities that defines perpendicular energy

transport are engineering safety factor qeng~BT/Ip, collisionality ν*~ ne/T2, normalized Larmor radius ρ*~T

0.5/BT

and plasma beta βT~W/BT2. It was found that for conventional tokamaks thermal energy confinement exhibits

strong degradation with βT and depends weak on collisionality. ITER scaling predicts BTτE~ βT-0.9

ν*-0.01

qeng-3

[3],

however different trend is observed for STs. The main feature of the thermal energy confinement in ST is strong

τE dependence on collisionality [4-5]. For MAST tokamak BTτE~ q-0.85±0.2

ν*-0.82±0.1

, for NSTX BTτE~ ν*-0.79±0.1

.

Future fusion devices should operate in the range of sufficiently lower collisionality, while q and βT values will

be close to the parameters of the existing devices. Therefore extrapolation of the experimental data to the lower

ν* values is of interest. Formal regression fit in the form BTτE~ ρ*xρ

βTxβ

νxν

qengxq

yields xρ=-2.7±0.12,

xβ=1.45±0.3 xν=0.45±0.01, xq=0.85±0.05 with RMSE=10%. Suggesting gyro-Bohm dependency (xρ= -3) it

was found that xβ=1.49±0.02 , xν=-0.47±0.01, xq=0.77±0.04 (RMSE=10%), For Bohm case (xρ= -2)

xβ=1.37±0.03, xν=-0.41±0.01, xq=1.04±0.04, (RMSE=11%). The confinement time dependence on

collisionality can be bounded like BTτE~ ν-0.44±3

. Dedicated analysis was performed with a set of plasma

discharges with different collisionality values while the other dimensionless parameters were kept constant. For

this purpose we have selected discharges with BT=0.32, 0.4, 0.5 T and Ip=0.15, 0.2, 0.25 MA. The exponent

indexes calculation was performed by log-linear regression yielding (Fig. 2). The scaling as a

function of ν* are even stronger when the variation of ρ* is taken into account through the Bohm and gyro

Bohm assumptions, with the normalized confinement going as

and

,

respectively. This result differs from the data from NSTX and MAST and is stronger than in ITER scaling [3].

The main reason for confinement improvement is the decrease of the electron heat conductivity. The indexes

changed along minor radius from B/e~ ν*-0.2

to B/e~ ν*-0.6

. The result is consistent with the obtained energy

confinement time dependence on collisionality. Ion heat transport is close to neoclassical level. However, the

presence of the anomalous contribution is observed at low collisionality.

BAKHAREV et al.

3

2.2. Alfven eigenmodes

Existence of the fast particles with velocities exceeding Alfven one in plasma will lead to excitation of Alfven

instabilities that can provoke fast particle radial transport and losses. Investigation of the toroidal Alfvén

eigenmodes (TAE) identified earlier in the experiments with NBI heating on Globus-M [6-7] were continued at

the increased values of the magnetic field and plasma current of 0.5 T and 240 kA correspondingly. We

observed that the TAE induced particle losses per burst of equal amplitude decrease with the increase of BT and

Ip. The decisive role in reduction of the losses plays the plasma current increase, probably due to a decrease of

the fast particle Larmor radius in the poloidal magnetic field. The tor oidal magnetic field increase gives a

weaker effect, but we have to increase the current and field simultaneously to conserve the safety factor value.

The obtained dependence is promising for CFNS on the base of a spherical tokamak, but should be checked in a

wider parameter range, which is one of the tasks to a new tokamak Globus-M2.

A progress was reached in the investigation of the TAE structure and localization. The multichannel Doppler

backscattering reflectometry (DBS) was successfully applied for this purpose during the last experimental

campaign on Globus-M. The description of the method can be found in [8] and its first application on Globus-

M - in [9]. Multichannel probing at frequencies of 20, 29, 39 and 48 GHz was applied. It allowed us to observe

the TAE fluctuations at four locations on major radius simultaneously. Fig. 3 demonstrates radial profiles of the

magnetic fluctuation amplitudes obtained by the described method for the TAE modes n=1 (black curve) and

n=2 (red curve) at 141.2 ms of shot #37001. The

dashed line shows the q profile obtained from the

EFIT code. We can conclude that TAE are localized

at the periphery of the plasma column, in the region

of normalized minor radii ρ from 0.6 to the

separatrix, the modes with different n have different

localization. Modeling of the Alfven continuum and

mode structure for the TAE burst with profiles shown

in Fig.4 were performed with the modified KINX and

CAXE codes [10]. Several global n=1 TAE were

found in the Alfvén continuum gap (the specific heat

ratio Γ=0) assuming fixed boundary condition. One

of them, shown in Fig. 4 with the frequency 155

kHz, seems to agree with the experimental data

also with respect to the mode localization.

2.3. Discharge disruptions

The characterization of plasma current quench during major disruption was studied in the Globus-M spherical

tokamak. A favorable, almost linear dependence of the normalized current quench time tCQ/S on the plasma

current density Ip/S before the disruption was observed [11]. The disruption characteristics depended weakly on

the ion mass. It was shown, that current quench is accompanied by the fast vertical displacement of the plasma

column results in the induction of considerable toroidal current in the vessel wall [12]. Interaction of this current

with the poloidal magnetic field produces electromagnetic loads. To calculate the distribution of induced

toroidal currents and electromagnetic pressure, a mathematical model of the vessel wall was developed. In the

model the wall is approximated by 64 axisymmetric rings with certain electric resistance closed along the torus.

The voltages, induced during the plasma current quench, were measured with a set of 21 loops located on the

vessel wall. Each loop was associated with a group of the neighboring wall elements. The distribution of normal

36 38 40 42 44 46 48 50 52 54 56 58 60

0

2

4

6

8

10

12

14

16

18

20

t = 141.2 ms n=1

n=2

~B

flu

cts

[ga

uss]

R [cm]

#37001 n=1

n=2

0

2

4

6

LCFS

q

q

FIG. 3. Localization of the TAE modes,

measured with DBS and q profile in

shot #37001, 141.2 ms.

FIG. 4. The structure of the TAE mode with fixed

boundary condition in Alfvén continuum gap, frequency

155 kHz. Harmonics in straight field line coordinates and

level lines of normal plasma displacement are shown.

0.2 0.4 0.6-0.4

-0.3

-0.2

-0.1

0

0.1

0.2

0.3

0.4

2/

A

2=0.1307

toroidal mode number n=1

0.0 0.1 0.2 0.3 0.40.0

0.3

0.6

0.9

1.2

1.5

1.8

2.1

~*

-0.4

B E

, T

*m

s

e*

FIG. 2. The dependence of

normalized confinement time on

collisionality.

FIG. 1. The comparison of the W

with WGLB=C∙IpαI

BTαBPabs

αPneαn

regression fit.

OV/5-4

to the vessel wall pressure Pn in the shot with downward plasma displacement during the disruption is shown in

Fig 5. In the example the plasma current before the disruption is Ip ≈ 0.2 MA and the current quench time ~ 0.6

ms. The maximum local pressure on the lower dome of the vessel is about 17 kPa. The pressure is directed

oppositely over a short vessel section. Such complicated distribution of the electromagnetic loads is connected

with a strong inhomogeneity of the poloidal magnetic field in this region. The shape of the vessel domes is near

momentless, but during the current quench a bending moment arises in the momentless dome.

2.4. SOL width

A new Langmuir probe, which can be moved along major radius between tokamak shots, was installed at the

outer midplane. The probe structure allowed measurements of plasma parameters up to about 2 cm inside the

separatrix. One of the crucial parameters of the edge plasma, determining the heat loads to the divertor plates, is

a power decay length λq. In spite of the significance of λq, a full theoretical model, allowing evaluation of this

parameter, doesn’t exist yet and empirical scalings are usually used for this purpose. According to a simple SOL

approach [13] the power flux is proportional to neTe3/2

. Therefore power

decay length at the midplane could be calculated using the following

expression: λq = (1/λne+(3/(2λTe))-1

. Electron temperature and density

profiles in SOL were measured for the two series (Bt = 0.4 T, Ip = 180

kA and Bt = 0.5 T, Ip = 225 kA) of lower single-null discharges with

close q95 values. The results were compared with Eich’s scalings of

2011 [14] and 2013 [15] (see Table 1). Like in the previous experiments

[16] derived values of λq are more close to Eich-2013 scaling.

TABLE 1. EXPERIMENTAL λq AND SCALING ESTIMATIONS

BT, T Ip, kA λq, mm λEich-2011q,

mm

λEich-2013q,

mm

0.4 180 4.9±0.5 7.0 4.0

0.5 225 4.1±0.5 5.4 3.4

3. TUMAN-3M RESULTS

TUMAN-3M is a compact tokamak with the following parameters: R/a=2.4, BT≤1.0 T, Ip≤180 kA, equipped

with the neutral beam injector. The main results of the investigation of the high frequency MHD oscillations in

IC [17] and Alfven frequency range and study of the L-H transition, initiated by GAM activity and pellet–

injection, in Tuman-3M are described.

3.1. ICE

Emission in ion-cyclotron frequency range (ICE) in magnetically confined plasma is usually observed in the

experiments where ion velocity distribution function is locally non-monotonic. This is typically caused by NBI

or RFH (Radio Frequency Heating). In Ohmic regime ICE is observed in plasmas with an essential fraction of

suprathermal ions originated due to nuclear fusion reaction [18].

3.1.1. ICE study In OH plasma

Oscillations in ion-cyclotron frequency range (5-100 MHz) were recently detected by fast magnetic probes in

Ohmically heated plasma in the TUMAN-3M tokamak. This phenomenon was called Ohmic ICE (OICE). The

distinguishing feature of Ohmic ICE is that it was not caused by suprathermal fusion products as in that

experiments the ion temperature (Ti~150-300 eV) was too low. The following characteristic features of Ohmic

ICE were found:

— The OICE was detected in the hydrogen and deuterium plasmas by magnetic probes, located inside the

vacuum vessel, both in the inner (high magnetic field side) and in outer (low field side) parts of the tokamak.

— Spectrum of the OICE typically consisted of many harmonics with frequencies evolving in time as toroidal

magnetic field BT(t) (see spectrogram in Fig. 6, where 8 harmonics of OICE are shown together with BT(t))

— OICE intensity was distributed asymmetrically relative to the equatorial plane: it was not possible to register

OICE below the equator.

FIG. 5 Diagram of the normal

pressure on the Globus-M vessel.

Circles - magnetic loops. Arrows

are proportional to the pressure

magnitude.

BAKHAREV et al.

5

— Fundamental frequency of OICE, registered by each of the

probes, depends on the probe position; it corresponds to the ion

cyclotron resonance frequency of the main plasma ions in the

close vicinity of the corresponding magnetic probe, regardless of

hydrogen (H+) or deuterium (D

+) was the main ion (Fig. 7).

— NBI does not influence the fundamental frequency of OICE

regardless of what hydrogen isotope atom is injected [19]; though

the injection can influence the intensity and harmonic content of

OICE.

In [20], the possibility of excitation of drift instabilities in an

inhomogeneous plasma at harmonics of the ion cyclotron

frequency is shown. The model proposed in these works seems to

be the most suitable for describing the phenomenon of Ohmic ICE

observed on the TUMAN-3M tokamak.

3.1.2. ICE study In NBI-heated plasma

ICE observed on the TUMAN-3M tokamak in NBI heated scenario is

produced mainly by minority fast ions (e.g. fast deuterons in hydrogen

plasma, and by fast protons in deuterium plasma) [19, 21] In contrast to many

other experiments, where ICE was found to originate from LFS periphery,

ICE in the TUMAN-3M comes from core plasma region. ICE frequency was

found to follow toroidal magnetic field evolution and to be insensitive to

changes in plasma density. Experimentally observed ICE spectral line consists

of 2 to 4 unevenly spaced very narrow lines, see Fig. 8.

Most probably, ICE in the TUMAN-3M is a result of CAE instability

(with dispersion relation = kvA) excited by stagnation fast ions with narrow

drift orbits localized in the vicinity of R~R0 and having average vertical drift

velocity close to zero [22]. For these ions, having velocity vb and IC

resonance frequency ci, Doppler-shifted resonance condition is =lci+k||v

[23]. Here and k|| are wave frequency and parallel wave vector. For

fundamental frequency hydrogen ICE in deuterium plasma observed

frequency f 13 MHz This is slightly lower than ICR for fast ions,

(ci)/2 = k||vb /2 − 1.5MHz, meaning the wave is counter-

propagating and has toroidal mode number n = − 1. Neutral beam being used

on the TUMAN-3M has non-monoenergetic spectrum, comprising E0, E0/2,

E0/3, 2E0/3 and some other components. Fast ions, produced by ionization of

these components, have different parallel velocities vb and locate at slightly different stagnation orbits, resulting

in different ICR and ICE frequencies. In this framework, the next experimentally observed characteristic

features of TUMAN-3M NBI ICE can be understood: exact harmonics for l=1, 2, 3…; linear dependence of ICE

frequency on toroidal field, and weak dependence on plasma density; fine structure of ICE frequency line.

FIG. 8. (a) Spectrogram of ICE in the TUMAN-3M deuterium plasma: NBI hydrogen ICE − Thin bright line at

~12.5MHz, with second and third harmonics barely seen; Ohmic deuterium ICE – broad line at ~8.5 MHz and its

harmonics (up to 7th), (b) Fine structure of NBI ICE spectrum, (c) Evolution of spectral components of ICE caused by

sawtooth activity.

FIG. 7. Spectra of OICE on the two

magnetic probes: LFS1 (Low Field

Side) and HFS1 (High Field Side) in

D and H plasmas. Dotted lines

denote calculated ion-cyclotron

frequency fci and its harmonics.

FIG. 6. Typical spectrogram of

Ohmic ICE in deuterium plasma. Up to 8

harmonics of fundamental OICE frequency

are observed. Toroidal magnetic field

evolution BT(t) is shown in an arbitrary scale.

OV/5-4

3.2. Alfven waves in OH plasma

Lower-frequency AW-type oscillations with

similar characteristics are observed at low

densities en < 3 10

19m

-3) in both Ohmic and NBI

regimes [24-25]. They exhibit clear Alfvenic

dependence on toroidal field and plasma density.

Comparison of the experimentally measured

frequency evolution with one calculated using

local plasma density in a discharge with strong

perturbation of density profile caused by LH-

transition indicates that AW are localized in the

central region r/a<0.5 (Fig. 9 a) [25] and thus may

be identified as GAE (Global Alfven Eigenmode)

rather than TAE (Toroidal Alfven Eigenmode).

AW activity in the TUMAN-3M plasma is usually

represented by two types of oscillations: short burst correlated with sawtooth crashes, and longer ones in

between the crashes, see Fig. 9. These two types have close frequency and similar poloidal spectrum, but

different poloidal localization: short burst oscillations are localized predominantly at the top of the torus,

whereas long burst have maximum amplitude at the bottom. In the TUMAN-3M AW are driven not by NBI-

produced fast ions (which velocity is well below Alfven velocity) but rather by magnetic field perturbation

created by sawtooth crashes or by loss of runaway electrons, associated with these crashes. Driving mechanisms

for the two types of AW may be different and require further investigation.

3.3. L-H transition

Experiments in TUMAN-3M show that in low density discharges (<ne> < 0.5.10

19 m

-3) L-H transition is not

present even if strong external sheared Er is applied. Two scenarios are discussed: discharge with electric field

perturbations due to the strong GAM activity; in this scenario the L-H transition is observed after GAM burst in

TUMAN-3M, while in low density discharges in FT-2 (n(0) < 5·1019

m-3

) with stronger GAM activity it is

absent; cryogenic pellet injection in TUMAN-3M, which creates strong density and temperature gradients and,

therefore, radial electric field gradient, and could lead to the L-H transition, but does not guarantee it. Both

scenarios were considered using the model of particle density profile evolution accounting for diffusion

coefficient dependence on Er shear [26,27]. Transport quantities and turbulence parameters were provided by

ELMFIRE gyrokinetic code. The modeling results have shown satisfactory agreement with the experiments: 1)

in TUMAN-3M the GAM parameters threshold for L-H transition is determined (Fig. 10), however for FT-2 the

transition was not observed for experimental GAM parameter sets [28]; 2) in the case of the pellet injection L-H

transition occurred if a part of the pellet was disintegrated in the pellet-guide, resulting in a gas/“snow” puff so

that the pellet ablation is peripheral; on the contrary, if the pellet is evaporated deeper in plasma, there is no L-

H transition [29].

Results of the modeling proved the idea about the importance of the particle source for determining LH-

transition possibility. This role is quite similar to the role of the ion heat source (heating power threshold) – if

particle source is too low for certain scenario, self-sustaining H-mode is not possible. Particle source thresholds

could be determined based on the diffusion coefficient dependence on Er shear (and thus n). The analysis of

non-linear particle flux dependence on n helps to reveal possible causes of LH-transition presence or absence.

Diffusion equation in stationary case has either two stable solutions (for L- and H-mode), or one – for one

confinement mode only, depending on integral particle source value compared to particle flux. The case when

the second stable solution (H-mode) is impossible corresponds to FT-2 low density discharge parameters or

TUMAN-3M deep pellet injection scenarios.

0 2 4 6 8 100

5

10

H-mode

L-modedn

/dr,

10

19 m

-4

t, ms

t=3 ms

t=5 ms

r = 21.5 cm

GAM

L-mode

a)

0 2 4 6 8 100

5

10

H-mode

L-modeL-mode

r = 21.5 cm

t, ms

EOSC

= 4.5 kV/m

EOSC

= 5.5 kV/m

dn

/dr,

10

19 m

-4

b)

GAM

FIG.10. GAM in TUMAN-3M (a, b) and FT-2 tokamaks (с). In TUMAN-M GAM burst of sup-threshold duration

(а) or amplitude (b) leads to LH-transition initiation.

52 54 56 58 60 62 64 66 68 70 720.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

r = 18 cm

17013120

r = 10 cm

r = 0 cm

f, M

Hz

t, ms

kII=1.74 m

-1L -

H tra

nsi

tion

r = 22 cm

a)

73.2 73.6 74.0 74.4 74.81.68

1.76

1.84

-200

-150

-100

-50

0

50

100

150

20073.2 73.6 74.0 74.4 74.8

t, ms

ne, 1019 m-3

filtred signal from 326 probe

b)

FIG. 9. (a) Evolution of measured (dots) AW frequency

and one calculated from local density evolution at different minor

radii (lines) in a discharge with LH-transition at t=57.2 ms, (b)

Correlation between sawtooth oscillations and AW burst.

BAKHAREV et al.

7

4. FT-2 RESULTS

FT-2 is a conventional machine (R/a=7.0, BT=3.0 T, Ip=40 kA), equipped with the LH heating system. FT-2 is

characterized by good discharge reproducibility and therefore is used for researches in different areas. In this

overview the main results of the full-f global gyrokinetic modelling benchmark using synthetic diagnostics are

described [30] followed by the synchrotron radiation measurements report [31].

4.1. Benchmarking of full-f global gyrokinetic modelling results against the FT-2 tokamak Doppler

reflectometry data using synthetic diagnostics

The massively parallelized particle-in-cell simulations of the gyrokinetic (GK) distribution function and the

electric field provide an efficient theoretical tool for studying the nonlinear turbulent plasma dynamics, which,

however, needs a validation and comprehensive benchmarking against the experimental data. In the present

paragraph the results of the global GK particle-in-cell simulations performed by ELMFIRE code are

quantitatively compared with the X-mode Doppler reflectometry (DR) experimental data obtained at the high

magnetic field side of the FT-2 tokamak. Two versions of the DR synthetic diagnostics, a fast linear version

based on the reciprocity theorem [32] and a full-wave version utilizing the IPF-FD3D code [33] are used for this

purpose. The experimental data obtained with standard Doppler reflectometry utilizing variable antenna position

and with the radial correlation Doppler reflectometry (RCDR), which was used in the comparison, characterize

the tokamak plasma turbulent dynamics at different scales: micro- (density fluctuations of trapped electron

mode (TEM)), macro- (the velocity of global plasma flows) and meso-scale (geodesic acoustic mode (GAM)).

The benchmarking against X-mode DR experimental data, has demonstrated a good agreement between the DR

spectra measured and computed using both synthetic diagnostics (see Fig. 11). For all antennae positions used

for comparison both the spectra frequency shift and width, and in many cases the spectra shape, were similar,

thus demonstrating a correct reproduction of the electric field behavior in the FT-2 tokamak by the ELMFIRE

GK code. The mean fluctuation velocities determined by the DR measurements and the synthetic diagnostics

were close. The GAM frequency and amplitude provided by the measurements and by the synthetic DR appear

to be close within a 12% accuracy. However a drastic (factor of 4) excess was found in the decay with growing

frequency shift in the probing wave channels of the RCDR cross-correlation function (CCF) provided by the fast

synthetic diagnostics compared to the experimental one. The quick decrease of the radial correlation observed in

the experiment is attributed to the phase modulation of the probing wave due to the long-scale density

fluctuations, which is shown to be close to π/4 by both the specially performed measurements and the WKB

estimation based on the GK modeling results. In spite of the fact that this value indicates only the beginning of

the transition to the fluctuation reflectometry nonlinear regime [34, 35] (which therefore is not influencing the

DR spectra) it already has a strong impact on the RCDR performance. This conclusion is confirmed by the full-

wave DR synthetic diagnostics correctly accounting for the perturbation to the probing wave propagation

produced by the turbulence, which demonstrates a much better agreement with the experimental CCF, as it is

shown in Fig.12. The only pronounced difference between the measurements and synthetic diagnostics is found

in dependence of the DR signal power on the antenna vertical displacement, which related to the density

fluctuation poloidal wavenumber spectrum (see Fig.13). Dependences of the measured and synthetic DR signal

power on the antenna vertical displacement are well approximated by Gaussian curves. However, the estimated

«poloidal correlation lengths» corresponding to these dependences are different by a factor of 1.5. This

difference could be attributed to the fact that GK computations underestimate the small-scale turbulence level in

its decay region at the high field side of the torus.

FIG. 11. Comparison of DR spectra

for up-shifted antenna vertical

displacement. Circles –experiment;

triangles – fast synthetic DR, squares

– full-wave synthetic DR

FIG. 12. The RCDR CCF against the

channel frequency separation. Stars –

fast synthetic DR; squares – full-wave

synthetic DR; triangles – fast synthetic

DR with phase modulation; circles –

experiment.

FIG. 13. Dependence of the

backscattering power on the

fluctuation poloidal wavenumber

-800 -400 0 400 8000.0

0.2

0.4

0.6

0.8

1.0

f (kHz)

P (

no

rm.

un

.)

experiment

FS DR

FWS DR

-4 -3 -2 -1 0 1 2 3 40.0

0.2

0.4

0.6

0.8

1.0

synth.with phase

measur.

Co

her

ence

f (GHz)

synth.

full wave

-20 -10 0 10 200.00.10.20.30.40.50.60.70.80.91.01.1 -10 -5 0 5 10

(cm

-1)

ya (mm)

Ps (

no

rm.u

n.)

experiment

FS DR

FWS DR

OV/5-4

4.2. Bremsstrahlung and non-imaging synchrotron radiation emission measurements

The studies of Runaway Electron (RE) generation and energy dynamics in tokamaks have gained great

importance from the theoretical and experimental perspective because the generation of high power RE beam

during the plasma disruption may damage in-vessel components of the large size tokamaks, such as ITER.

Bremsstrahlung (BR) measurements of the high energy REs (~2-7 MeV) and observation of synchrotron

radiation emission (SRE) performed in the FT-2 tokamak in the range of 106-156 GHz were used as a basis for

the design and development of a new diagnostic for ADITYA-U tokamak that can detect SRE in the sub-

millimeter band (THz-band) [31)]. An imaging diagnostic approach has been proposed for the first time to

capture spatiotemporally resolved SRE pattern of the low energy confined REs in the energy range 1-20 MeV.

5. ANOMALOUS ABSORPTION AND EMISSION IN ECRH EXPERIMENTS DUE TO PARAMETRIC

EXCITATION OF LOCALIZED UH WAVES

In contradiction to the standard theory [36] predicting high thresholds of anomalous phenomena at ECRH in

tokamaks the anomalous backscattering was observed a decade ago at TEXTOR in modest power level X2

mode neo-classical magnetic island control experiments [37]. This effect was explained in [38] by low-threshold

parametric excitation of two upper hybrid (UH) waves trapped in the vicinity of a local maximum of plasma

density situated in the magnetic island. However, it should be mentioned that the anomalous backscattering

effect was observed at TEXTOR in a wide range of the plasma densities in the local maximum of the density

profile substantially exceeding the UH value for half the pump wave frequency. Under these more general

conditions the trapping of both decay UH plasmons is no longer possible. Nevertheless, one of the primary UH

daughter waves can still be localized, that can lead to the excitation of low-threshold absolute parametric two-

plasmon decay instability (TPDI) [39]. In the present paper we analyze the saturation of the TPDI under the

general conditions when only one of the parametrically driven plasmons is trapped, whereas the second one can

leave the decay region (see Fig. 14). We consider the secondary decay instability of the localized UH wave

producing another UH wave and ion Bernstein wave (see Fig. 15), as a moderator of the primary TPDI. The

general analytical consideration is accompanied by the numerical analysis performed under the experimental

conditions typical of the off-axis X2-mode ECRH experiments at TEXTOR [37] demonstrating the instability

saturation (see Fig. 2). We also estimate the pump power fraction gained anomalously throughout THDI in this

case. As it is shown in Fig. 16, well above the TPDI threshold equal for the TEXTOR parameters to 128 kW the

anomalous absorption rate is equal to 11%.

FIG. 14. The dispersion curves of

primary and secondary daughter waves

generated in the cascade of decays; 1 –

primary non-trapped UH wave, 2 -

primary trapped UH wave, 1’ –

secondary IB wave, 2’ – secondary

trapped UH waves. The dimensionless

UH frequency profile ( 2 2

UH 0/ ( / 2)f f )-

thick solid curve.

FIG. 15. Temporal evolution of the

averaged energy density of primary

(solid curve) and secondary (dashed

curve) UH plasmons. The dash-dotted

line - the analytical predictions.

FIG. 16. The anomalous absorption

rate versus the pump power.

It should be also underlined that in the case 0 UH min2 x considered in this paper the non-trapped UH wave

propagating perpendicular to the magnetic field outward is converted into the X-mode (see Fig. 14). It crosses

the ECR surface and leaves the plasma at the high-field side. Neglecting ECR absorption of the X-mode, which

is small at the perpendicular propagation in modest temperature plasma of middle scale devices, and taking into

account the power balance in the decay we can estimate the power of this X-mode radiation at the half-pump

frequency as 0 2 0 / 2XP P . In the case considered in this paper it results in 0.05 P0. The larger part of this power

is reflected from the device wall in the form of the X-mode and finally absorbed after conversion in the UHR.

However a smaller part of the power, characterized by cross-polarization factor βXO, is reflected in the form of

O-mode, which is partly absorbed in the ECR and then leaves the plasma. Taking into account the O - mode

BAKHAREV et al.

9

absorption rate O1 we obtain the estimation for the O-mode emission power at the low field side at the half

pump frequency XO O1 00.5 expOQ P . Taking into account that for the conditions of the TEXTOR off-axis

ECRH experiment O1 2 , and assuming βXO=0.01 we get for the case under consideration 5

08 10OQ P . Thus

at the maximal microwave power of 600 kW one can expect to see at the low field side of the tokamak the O-

mode emission of 50 W at the half pump frequency. This emission is observable only in the narrow density

range when the UH density for half the pump frequency is slightly lower than plasma density in the profile local

minimum, however the decay with the upper branch of the non-localized UH wave (electron Bernstein wave) is

still possible. Out of this density range the intensive emission of sub-harmonic 0 2 is not possible;

nevertheless due to the nonlinear coupling of UH plasmons and the pump wave emission of the 03 2 harmonic

can occur. It happens when the following conditions hold

max 0 e

ce m

e T max

5

2e

e

n x c nx

n l n x

, where

e max mine en n x n x , maxen x and minen x stand for the density value in the local maximum and minimum

of the profile accordingly. We get the 03 2 pump harmonic emission power received by antenna as 3 2 0sp T P ,

where the conversion coefficient at the TEXTOR parameters is equal 53 2 5 10T . For the maximal power

utilized in the TEXTOR magnetic island control experiment it results in 30 W of received 03 2 pump harmonic

emission. More details on the researches described in this paragraph can be found in [40].

6. ITER DIAGNOSTICS DEVELOPMENT

Three diagnostics for ITER are being developed in the Ioffe Institute: neutral particle analysis (NPA), Gamma-

ray spectrometry (GRS) and divertor Thomson scattering combined with laser-induced fluorescence (DTS/LIF).

6.1. NPA and GRS

Over the past few years considerable progress has been made in several areas of NPA diagnostic system

development. Current 3D design of NPA system, which consists of low-energy (LENPA) and high-energy

(HENPA) neutral particle analyzers and gamma-ray spectrometer (GRS), is presented in Fig. 17. A detailed

technological analysis of NPAs design was done to conform with French Nuclear Regulations and from the

point of view of manufacturing. A lot of efforts have been made into integration of LENPA and HENPA

analyzers with their supporting subsystems into ITER environment. At present the tests of the most critical NPA

components are underway [41]. Also the problems related to NPA neutron and magnetic shield design are being

resolved [42]. One of the main goals for the analyzers at ITER is the measurement of fuel isotope (D/T) ratio

[43]. Thus, the influence of various plasma phenomena on the NPA ratio measurement has been analyzed. It

was shown that Neutral Beam Injection (NBI) [44], Pellet Injection [45] and sawtooth oscillations [46] may

have some effect on NPA signals, which should be taken into account by appropriate modelling to get sufficient

accuracy of D/T measurements.

Gamma-ray spectrometer is placed in the neutron dump of the NPA system behind LENPA (Fig.11). The

spectrometer is intended to diagnose accelerated ions and electrons by the line-integrated gamma-ray

measurements [47,24] They will provide complementary signals for NPA data on fuel ratio and energy spectrum

of confined α- and other energetic particles (p, D, T, 3He). The gamma-ray spectrometer includes two detectors:

a semiconductor HPGe detector and a scintillation detector with a LaBr3(Ce) crystal. To reduce incoming

neutron flux, a neutron attenuator is installed.

The attenuator is a steel cylindrical housing filled

with lithium hydride (LiH) tablets. Lithium

hydride provides a high degree of absorption and

scattering of neutron radiation, while remaining

relatively transparent for gamma quanta. For

spectrometric measurements under conditions of

a thermonuclear experiment, it was necessary to

develop algorithms for digital signal processing,

providing stable measurements at the counting

rate of LaBr3(Ce) up to 5∙106 s

-1 and up to 5 ∙10

5

s-1

for HPGe detector. The developed data

acquisition system based on the NI PXIe

architecture allows recording the detector signal

FIG. 17. 3D design of ITER NPA diagnostic system.

LENPA and HENPA neutral particle analyzers and gamma-ray

spectrometer (GRS) are shown.

OV/5-4

for at least 400 s with a sampling rate of 250 MHz. At present, the prototyping and testing of the GRS main

components are carried out under conditions close to the ITER ones. The LaBr3(Ce) detector mockup was

successfully tested for a radiation resistance under integral neutron flux of 1.1⋅1013

cm-2

. More details on the

GRS development are presented in the report [48].

6.2. Divertor Thomson scattering combined with laser-induced florescence

The ITER divertor Thomson

scattering (DTS) and Laser-induced

Fluorescence (LIF) (Fig.18) are

active laser-aided diagnostics

providing local measurements of

plasma parameters in the outer

divertor leg i.e. in the area with

extremely steep gradients both

parallel and perpendicular to the

magnetic surfaces. The combined

DTS/LIF diagnostic provides

simultaneous measurement of the

local Te, ne, Ti, ni, na along the

probing laser paths with resolution

of 25 mm along the probing chords.

The set of plasma parameters allows

calculating rates of the processes important for better understanding physics of divertor operational modes.

Challenges of Thomson Scattering implementation in the ITER divertor and the capability to satisfy project

requirements, related to the range of the measured electron temperature and density, were discussed yearly in

[49] and under consideration at this conference [50]. The main challenge for DTS is extremely high-electron-

density and low-electron-temperature plasmas in the regions with predominant recombination. The developed

approach, based on synthetic experiments, shows that the expected Te and ne measurement accuracy is better

than the specified technical requirements, in spite of the pronounced collective effects. Temporal shapes of the

HeI fluorescence can be also utilized for the measurement of ne in the range of 1018

- 1020

m-3

using HeI

collisional-radiative model CRM and known Te expanding measurable range of ne. The new kind of LIF

spectroscopic scheme based on laser induced quenching of the most intensive hydrogen line Hα = 656.1 nm is

proposed for the measurement of H/D/T atomic density in ITER divertor. Dα line quenching was tested

experimentally in a glow discharge plasma with laser- exciting atomic deuterium from 3rd

to one of the upper

states with n = 4÷12. Current status of LIF means measurements of CRM describing relation between the

fluorescence and plasma parameters was presented in [51]. Recent achievements in the hardware development

are multiple. The progress in the development filter polychromators and piezomotors is presented in [52]. Last

achievements in the development of DTS laser system was presented in [53]. Investigation of laser resistance of

plasma-facing diagnostic components was discussed in [54].

7. GLOBUS-M2

Globus-M2 is an upgraded version of Globus-M with the same vacuum chamber and enhanced EM, heating, CD

systems, power sources and diagnostics. An increase in the magnetic field (from 0.4 T to 1.0 T) together with an

increase in the plasma current (from 200 kA to 0.5 MA) in Globus-M2 tokamak should improve plasma

performance and provide enhanced conditions for auxiliary heating and current drive [16, 55-58]. ICR heating in

the case of an increased magnetic field and, as a result, the growth of the resonant frequency up to 15 MHz

becomes more effective due to the enhancement of a single-pass absorption and improvement of the antenna-

plasma coupling. An increase in the efficiency of neutral beam injection is expected due to a significant

reduction in the fast particle losses in the discharges with higher plasma current and magnetic field. A half of the

plasma current can be sustained non-inductively with simultaneous input of two neutral beams of 1 MW each.

The use of the low-hybrid waves (2.45 GHz) at an input power level of 0.5 MW should ensure completely non-

inductive maintenance of the current in the discharge. As a result of joint plasma current and magnetic field rise,

on the one hand, and an increase in the efficiency of the auxiliary heating methods, on the other, one could

expect an improvement in the plasma parameters, including the increase in the electron temperature and the

significant drop in the collision frequency. First plasma in Globus-M2 was produced on April 23, 2018. During

the first campaign 250 kA plasma current and 0.6 T toroidal magnetic field were achieved. A full-scale

experiment with an increased toroidal magnetic field is planned for the end of 2018.

FIG. 18. DTS/LIF equipment situated in ITER Lower Port #8. (a) 3D

view of the equipment. 1 — Strike point on the outer divertor target; 2 — Laser

beams; 3 — Front and back racks used for arrangement of DTS/LIF in vessel

equipment and neutron shield components; (b) Side view on the divertor port

relative positions of the laser probing beams and magnetic field surfaces.

BAKHAREV et al.

11

ACKNOWLEDGEMENTS

Tokamak studies in Ioffe Institute were made within the framework of the programs of fundamental scientific

research of state academies of sciences.

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