the usual instrumental uncertainties. Historically computation effort in atomic data was directed to improving the scope of the atomic models but there is a growing activity to produce assessed uncertainties on ab initio atomic data, e.g. [446] for astrophysics. A number of technical meetings and workshops organized by IAEA have fostered this uncertainty quantification (UQ) effort in the fusion community [447].

Theoretical and computational improvements are still required for atomic data models. For example, it should be noted that there is no R-matrix data available for the equivalent inner shell transitions for W$^{45+}$. Another example is that differences of up to 40% in the S/XB coefficients for Ar$^+$ transitions were found between simple R-matrix and R-matrix with pseudo states (RMPS) models [448]. The later model includes coupling to the continuum levels which are important for neutral and near-neutral ions. In this case using the theoretical spread of values would give a severe over-estimation of the atomic error.

Spectral features, such as the quasi-continuum of figure 92, are composed of many thousands of spectral lines from a number of ionization stages (W$^{27+}$–W$^{45+}$ for the 4–5 nm tungsten feature). The large number of transitions are due to the many possible energy levels when coupling partially open d and f shells. Combining the radiated power from lines, resolved by J total quantum number, of the lowest lying configurations with configuration-average rates for the rest is reasonable for total radiated power, or data for energy-resolved SXR emission [415]. However this approach may not be sufficient for a spectral prediction to be confronted with experiment. Furthermore, configuration interaction effects alter the shape of the QC features without affecting the total power radiated. An example from laser plasmas, a little higher in electron density than tokamak plasmas, shows that transitions between excited configurations also affect the shape of these QC features [449]. Including these, in the necessary J-resolved resolution, adds orders of magnitude more lines [450]. The shortcomings in the atomic spectral model can be seen in the, deliberately simple two-temperature zone model, with the ADAS fractional abundance data of figure 103 and distorted wave generated feature photon emissivities (f-PEC). The qualitative nature of a dominant low temperature part at ~1.6 keV, with some emission from a higher ~3.5 keV component, is compatible with the temperature profile and the cooling effect of a large tungsten influx. The identification of the contributing stages to the spectrum in [325] confirms that the missing piece of the modeled QC feature is due to emission from W$^{26+}$–W$^{29+}$. A more adaptive partitioning scheme in the atomic model will be needed to enable the modeled spectral feature to be fully used in a predictive capacity.

However, the QC feature can still be used in a pragmatic manner because of the nature of the emissivity coefficients. The photon emissivities of the many lines of similar wavelength from the different ionization stages vary little in temperature, a few percent between 1 and 15 keV, across the group so any changes in the measured spectrum can be attributed to changes in the relative abundances. Furthermore, the ratio of the power in the 5 nm QC band to the overall power also has a modest dependence on temperature. In a plasma with

a flat density profile and little transport, the concentration of tungsten ($c_W$) is simply,

$$P_{\text{tot}} = c_W N_e^2 L(T_e, N_e),$$

where $P_{tot}$ is the power contributed by tungsten and $L_z$ is the cooling curve (figure 103). The factor linking the intensity of the QC feature to the concentration of tungsten in a specific calibration measurement can then be extended to a wider range of conditions by adjusting the fractional abundance along the LOS of the observation. This method has been used at ASDEX Upgrade [358, 451] and JET and gives a robust measure of the tungsten concentration without the necessity of a high fidelity fit to the spectrum.

The energy binning of a wide coverage, multi-energy SXR (ME-SXR) detector may not be sufficient to resolve the 5 nm QC feature but the desirable characteristics of 2D imaging and fast response to MHD events remain. Tomographic techniques are used with current SXR diagnostics [69] and the consideration of the detector response is via filtered cooling rates. The fraction of radiated power filtered by the beryllium window and Si-diode detector is shown in the dashed (SXR) curve in figure 103. The SXR emission is moderately sensitive to the filter function, as seen in the mis-matched beryllium window widths of the vertical (250 µm) and horizontal (350 µm JET SXR system [424]). The difference in response can be accounted for in the tomography with a, computationally expensive, correction factor in the contribution matrix. Other effects such as the different energy signatures of each impurity can also be used to improve the SXR analysis [452]. However true ME-SXR requires that the detector response is properly included in the analysis. Incorporating the detector energy behavior into the tomographic analysis has been demonstrated at WEST. The GEM detector response over 1.5–10 keV was used to include a transport model which enabled the tungsten emission at different positions to be disentangled from the SXR signal (utilizing the variation similar to that in [453]). Figure 110 shows that a 5-position LVIC ionization chamber can reconstruct the energy dependent radiated power signature, including lines and the shape of the continuum background, in the 1–100 keV region for a high performance ITER plasma [442]. Figures of merit (FOM) for each pixel in the tomographic analysis are good in the core ($\rho < 0.8$) for strong line emission and for the background above 15 keV but are poor in the edge region. The 200 µm beryllium window cuts off much of the edge emission but the good core FOMs show the potential of ME-SXR for measuring general impurity content and as a way to measure the electron temperature profile at very high core temperatures where TS becomes difficult.

Extracting useful information from some measurements does require a sophisticated model of the emission. A measurement common to all tokamak plasmas is the visible Balmer-α line from hydrogen or its deuterium and tritium isotopes. Even before the current metallic wall choice for ITER, the difficulties caused by contamination of the signal due to reflections from the wall in DIII-D (CFC tiles) and Alcator C-Mod (molybdenum) was quantified as being small (<10%) for divertor views but could be significant (up to 50%) for the
<figure><img src="image_122.png" /><figcaption>**Figure 110.** The reconstruction of the upper 20% (1–100 keV) radiation of the high performance ITER plasma of figure 106. The modeled LVIC detector chamber is filled with xenon at atmospheric pressure with anode sizes of 5, 30, 60, 100, and 150 µm behind a 200 µm thick beryllium window. Reprinted from [442], with the permission of AIP Publishing.</figcaption></figure>

main chamber [454]. A model of the reflections of Balmer-α in an ITER scenario based on the measured BRDF of Inconel™ tiles from JET showed that diffuse reflection, in addition to specular, was significant [455]. A practical mitigation strategy is to terminate the LOS on a viewing dump but this is not always possible so a good synthetic model of the feature is necessary.

An additional complication for main chamber Balmer-α is the strongly non-Maxwellian VDF caused by recycling in the SOL. The effect of both recycling and stray light contamination is seen as an asymmetric line profile that requires a HES and a reflection propagation model to disentangle the contributions and provide a clean measurement. A model of the Balmer-α emission is constructed and used to solve the *inverse problem* [456]. Strong constraints from Zeeman splitting separate the low and high field emission regions and a semi-analytical description of the shape of the stray light from the divertor, assuming multiple specular reflections of a homogeneous divertor source [457], makes this a robust model for fitting. This approach has been successfully validated with JET observations and figure 111 shows the variation in shape of the line from a limiter view of JET [458]. The contributions of the SOL emission from the high field and low field are very different at different phases of the discharge and there is significant emission due to stray light reflections from the divertor.

An additional complication for main chamber Balmer-α is the strongly non-Maxwellian VDF caused by recycling in the SOL. The effect of both recycling and stray light contamination is seen as an asymmetric line profile that requires a HES and a reflection propagation model to disentangle the contributions and provide a clean measurement. A model of the Balmer-α emission is constructed and used to solve the *inverse problem* [456]. Strong constraints from Zeeman splitting separate the low and high field emission regions and a

<figure><img src="image_123.png" /><figcaption>**Figure 111.** A fit to the Balmer-α spectrum from the inboard limiter in the octant 8 of JET (pulse 87 638 with KSRB spectrometer) during plasma ramp-up ($t = 4.1$) and at steady state with ICRH heating ($t = 13.9$ s). The fit parameters for the SOL light from the high field side (HFS) and low field side (LFS) are shown, along with the contribution of stray light from the divertor (DSL). Reproduced from [458]. © 2016 EURATOM. All rights reserved.</figcaption></figure>

semi-analytical description of the shape of the stray light from the divertor, assuming multiple specular reflections of a homogeneous divertor source [457], makes this a robust model for fitting. This approach has been successfully validated with JET observations and figure 111 shows the variation in shape of the line from a limiter view of JET [458]. The contributions of the SOL emission from the high field and low field are very different at different phases of the discharge and there is significant emission due to stray light reflections from the divertor.
Reflections are typically modeled by optical ray-tracing to
produce a transfer matrix relating the reflected light originat-
ing from different sources to each point in the plasma volume
[459]. Tomographic techniques can then be used to incorpor-
ate this contribution in a diagnostic analysis. The accuracy
depends on knowing the reflecting properties of the tokamak
vessel's interior materials (BRDF) and the source and axisym-
metry of the originating light—from the divertor, SOL or core.
The reflectivity of in-vessel components will change during
operations, via depositions and roughening (see subchapter 9
on mirrors) so ideal, or even measured pristine surface BRDFs
may not be useful. However, ideal, rather than real, reflectiv-
ities are sufficient to determine favorable locations for dia-
gnostics, as done with the Visible Spectroscopy Reference
System (VSRS) diagnostic for ITER [307]. Calibrating the
reflectivity response with a small volume plasma has been
demonstrated on COMPASS [460] but this method may not
be feasible to mimic the intense radiation from the divertor.

Combining a comprehensive emission model with a
verisimilar reflection model produces a Balmer-α synthetic
diagnostic [456]. The use of synthetic diagnostics, which is a
methodology to generate numerically what the diagnostic sys-
tem will see, can be used to test the success of the algorithms
used to recover the desired MP. Many spectral emission mod-
els are being incorporated into the CHERAB [136] spectro-
scopy modeling framework, which uses ADAS as a primary
source of atomic data. This implements an, optimized-for-
tokamaks, ray tracing code Raysect [461] to model the reflec-
tions in a realistic geometry. Figure 112 taken from [462]
shows how well a good synthetic diagnostic could be used
to tune an analysis applied to assessing the best placement of
the fields of view (FoV) for the DIM. The results are depend-
ent on the environment and in particular on the unmeasured
and uncharacterized BRDF. With a synthetic diagnostic the
effect can be quantified by adjusting the roughness in the
reflection model and a more refined error estimate, calculated
for each FoV rather than a single overall percentage, may be
determined [463].

A synthetic diagnostic for each spectroscopy system
enables enhanced utilization. Two aspects are explored in
detail in later chapters. Combining diagnostics provides a
much fuller picture of the plasma performance, but its suc-
cess relies on a good description of each input. A synthetic
diagnostic may be parameterized, either by a reduced model
or via machine learning, which enables complex analysis to
be performed in real time.

5.6. Summary

Much of the worldwide effort in passive spectroscopy since the
last ITPA review has revolved around preparing to make quant-
itative, high quality, measurements of the emission expected
from ITER. The ITER environment with significant neutron,
gamma and x-ray fluxes, and long duration discharges, will
pose a challenge to deploying reliable and stable diagnostics.
The spectroscopy of tungsten will be a defining aspect of the
passive emission from ITER because it will radiate from all
parts of the plasma and will be a mixture of quasi-continuum

<figure><img src="image_124.png" /><figcaption>**Figure 112.** A The contribution of reflections to the D-α signal for an equatorial (a) and upper (b) port view of the divertor impurity monitor (DIM) diagnostic for ITER. Simulated signals and the recovery of the real signal by removing the contribution from reflection are shown. Reproduced from [462]. © IOP Publishing Ltd All rights reserved.</figcaption></figure>

spectral features and individual lines. The 74 electrons of tung-
sten, and the need to consider in detail at least 70 of its stages—
even ITER will not be hot enough to ionize the He- and H-like
stages—has been a grand challenge to calculate atomic data
which underpin the models and the analyses that flow from
them [419].

Plasma experiments, dedicated spectroscopy programmes,
new detector technologies, diagnostic system prototypes, fun-
damental atomic data calculations, atomic cross section meas-
urements and the development of better algorithms and soph-
isticated analysis codes have been undertaken in recognizable
fusion institutes but also in many universities and other labor-
atories so that useful measurements will be possible from the
first ITER plasma.

6. Laser aided diagnostics

R. Scannell²⁸, H. Liu²⁶, E. Mukhin²⁹, E. Yatsuka¹⁴,
A. Gorbunov²⁷, L. Giudicotti³⁰, G. Kurskiev²⁹, J. Chen¹⁰,
M. A. Van Zeeland⁴, D. Finkenthal³¹, R. Imazawa³²,
D. Brower³³, A. Sirinelli², T. Akiyama⁴, T. Carlstrom⁴,
D.W. Johnson²¹, M. Lesher⁴, C. Watts² and M. Bassan²

6.1. Introduction

The required plasma measurements to make ITER a success,
of which laser diagnostics are a crucial component, flow from
the ITER overall PRs document [464]. The PRs document
details what must be achieved by the ITER plant based on
the underlying physics. From this document, the particular
requirements for individual diagnostic systems are derived and detailed in the ITER diagnostics requirements document [197]. The means of meeting the measurement requirements for ITER was initially addressed in [3] for the full diagnostic set and in particular for laser diagnostics in section 3.3 of this document. Further progress in diagnostics systems with specific examples was last set out in 2007 and published in [4].

This current paper details the status of the various laser diagnostics that will be used to meet the measurement requirements. Each section contains links to relevant background to properly introduce each diagnostic system.

As well as describing the various diagnostic systems themselves relevant techniques are also discussed. Polarimetric and dual laser TS techniques which are particularly useful for TS systems on ITER are introduced. Neither are routinely employed on existing tokamaks but they are being tested with a view to application on ITER. The polarimetric technique in particular has developed considerably since the last update to the ITER physics basis.

The Laser Induced Flourescence (LIF) diagnostic which will be integrated into the DTS diagnostic is a new advance also relative to the ITER physics basis published in 2007. The Toroidal Interferometer Polarimeter (TIP) system, which is planned to be used for real time density control at ITER, has recently been prototyped and deployed in a test capacity at DIII-D. While the PoPola (Poloidal Polarimeter diagnostic has been in the ITER diagnostic set for some time, it is proposed use for magnetic fluctuation measurements and plasma vertical control are a recent development. The final diagnostic discussed is the Dispersion interferometer which is another recent innovation. Due to its configuration, it is highly resistant to vibrations making it a very robust measurement tool and motivating its application for ITER.

## 6.2. Core TS

The ITER core plasma TS (CTPS) diagnostic was originally envisaged as a LIDAR design [465]. However, the design has been modified to follow a conventional TS layout, shown in figure 113, which operates at high scattering angle. A detailed design trade-off between the LIDAR and conventional design solutions was considered in making this decision. Some of the main factors motivating a conventional design were easier availability of detectors and laser, higher spatial resolution and longer beam dump lifetime.

The CPTS diagnostic must meet requirements on its performance which derive from the ITER PRs [464]. The CPTS diagnostic has a number of common elements with the edge TS diagnostic, since both systems are conventional systems based on Nd:YAG lasers operating at high scattering angle. Particularly important common elements include the beam dump [467, 468] and solutions for laser beam combination [469] as well as laser and collection optics alignment. Particularly important differences with the edge system are the more challenging optical labyrinth (neutron attenuation) trade-off for CPTS, larger first aperture and hence shutter requirement for the CPTS and requirement for the CPTS

<figure><img src="image_125.png" /><figcaption>**Figure 113.** Optical design of CPTS diagnostic. Mirrors are number in order of light propagation from the plasma. The first and second and sixth mirrors from the plasma are free form. The approximate parameters for the furthest in and out most measurement points are indicated. Reproduced from [466]. © 2017 ITER Organization. Published by IOP Publishing Ltd on behalf of Sissa Medialab. All rights reserved.</figcaption></figure>

to measure to higher electron temperature and hence lower wavelength.

The main components of the proposed system are a 5 J, 100 Hz, 4 ns laser system, envisaged to consist of at least 2 lasers for redundancy, an optical system collecting at F/14.7 at $r/a \sim 0.02$ at $\sim 167$ degrees varying to F/7.6 at $r/a \sim 0.85$ at $\sim 155$ degrees. The CPTS system design has capability for measurement of 'p' as well as 's' polarized scattered light, originally proposed by [470] and detailed further in [471]. The proposed CPTS system also includes a 1319 nm 10 Hz 2 J laser operating in conjunction with the main 1064 nm laser. This multi-laser approach has been suggested for each of the ITER TS diagnostics motivated by requirements for calibration and to reach high electron temperature; a recent consideration has been made in [472].

In order to achieve good optical performance and neutron shielding, freeform optical elements have been included in the new optical design. These are challenging to implement, in particular in conjunction with the requirement for a mirror cleaning system for the first mirror but promise benefits in performance. Considering simulated ITER profiles, density and temperature gradients imply that higher spatial resolution is required in the mid-radius and outboard region than in the core. This higher resolution is naturally achieved for a fixed fiber bundle size due to the change in collection F/#.

The main challenges facing the diagnostic are in terms of survivability of PFCs in the harsh operating environment, the long duration of 4700 h of operations and the inaccessibility during this operating period. Some of these key engineering challenges are listed below:

* Survivability of the first mirror in a harsh and unknown plasma environment where deposition is expected [473].

* Demonstration of a mirror cleaning system compatible with the first mirror [474], particularly if free form.

* Engineering implementation of a mirror protection shutter at high reliability. This shutter must contain cooling channels.

* Active alignment of the scattering lasers to the beam dump to maintain sub-cm alignment during plasma pulses.

* Survivability of the laser beam dump.
<figure><img src="image_126.png" /><figcaption>**Figure 114.** Optical transmission of the CPTS diagnostic including full information of spectral transmission of the various collection optic elements, the detector EQE and optical fiber transmission. The optical fiber transmission includes expected degradation for the ‘intermediate’ OH fiber after 10 kGy of radiation. The ‘at risk’ region, which may suffer further transmission losses is highlighted. Reproduced from [466]. © 2017 ITER Organization. Published by IOP Publishing Ltd on behalf of Sissa Medialab. All rights reserved.</figcaption></figure>

As well as diagnosing electron temperature profiles more generally, one of the particular key goals of the CPTS diagnostic is to measure the peak core $T_e$ value. This peak core $T_e$ will be an important indicator of ITERs success. This implies that the diagnostic must measure to the lowest possible wavelength ~400 nm see figure 114. There are a number of factors that make measurement at this very low wavelength difficult, in particular first mirror reflectivity losses are expected to be high in this region, as will transmission losses for optical fibers and Cerium doped lens elements in the interspace. Also, there is significant line emission in the lower wavelength region. However, there are new diagnostic advances that should aid the CPTS in these crucial measurements and help distinguish from any spectral transmission uncertainty at the edge of the wavelength region. These advances are ‘Polarimetric’ TS, measuring the polarization of the scattered light and use of a 1318 nm Nd:YAG laser. The theoretical concepts behind these techniques have been further developed in the last few years. Polarimetric and two-wavelength TS are now being prototyped and tested on existing machines.

### 6.3. Edge TS (ETS)

ETS has an important specification to measure the region of $r/a > 0.85$ with a spatial resolution of 5 mm in order to measure the steep electron density and temperature gradient in H-mode discharges. The electron temperature range covered by ETS is 0.05–10 keV. This temperature range is measured from 590–1060 nm with six-channel polychromator spectrometers.

<figure><img src="image_127.png" /><figcaption>**Figure 115.** Configuration of the ETS collection optics. Inside the port plug, it consists of only mirrors, and outside the port plug, it consists of mirrors and lenses.</figcaption></figure>

To detect enough photons, the $F$-number of the ETS collection optics varies from $F/15.2$ to 21.7. While using such a high collection optical system, it is a challenge to minimize chromatic aberration over the above-mentioned wavelength band and to be compatible with radiation shielding. The key engineering challenges are the same as the five listed in the CPTS section.

The optical path of the collection optics forms a labyrinth in the port plug to pass the collected light and to shield neutrons from the plasma. No lens is used in the port plug to avoid fracturing of glass due to vibrations during plasma disruptions. The collection optics are capable of collecting light from a normalized minor radius range between 0.80 and 1.08. The outermost LOS corresponds to the face of the DFW [475]. In order to suppress aberrations, the angle of incidence on the curved mirror was made as small as possible. Figure 115 shows the configuration of the ETS collection optics. Considering the radiation environment near the port plug, it is desirable that the field of view of the collection optics can be adjusted without human intervention. Light propagating in the opposite direction from the optical fiber hits the bracket near the upper port. By observing the vignetting of spots projected on the blanket, the relative relationship between DFW and the observation position can be maintained.

The laser beam dump is mechanically installed on the inner blanket. The space provided for the beam dump is 60 mm wide, 115 mm high and 77 mm deep, including a flange for mounting on the blanket. If the laser beam dump fails, the entire FW structure of a blanket module must be replaced by remote handling maintenance equipment. In order to reduce the risk of laser beam dump failure, the beam diameter should be as large as possible. For this purpose, the beam has to always be incident on the center of the laser beam dump. ETS achieves a temporal resolution of 100 Hz by alternately firing two 50 Hz laser beams. The use of two lasers enables us to perform the TS measurements at a temporal resolution of 50 Hz in case that one of the laser systems stops functioning. In order to maximize the light receiving surface of the laser beam dump, allow use of two laser beams, and suppress the streaming of neutrons along the beam optical path, it is necessary to align the two beams coaxially. Since the scattering angle is about 140 degrees and the measured electron temperature is up to 10 keV,
<figure><img src="image_128.png" /><figcaption>**Figure 116.** (a) Bird's eye view and (b) concept of the laser beam dump.</figcaption></figure>

the depolarization of scattered light is at most a few percent. In order to improve the ratio of Thomson scattered light to the background light of randomly polarized light, a polarizer is installed in the collection optics. The functional requirements for laser incidence and laser beam dump are summarized below. The two horizontally polarized beams must be coaxial and always incident at the center of the laser beam dump. The laser beam dump needs to survive for a long time under the severe heat load of ITER.

A rotating half-wave plate provides a lossless method for combining beams together with stable beam pointing. The polarization extinction ratio of the combined beams can exceed 1000 [476].

In order to inject the probing beam into the center of the laser beam dump, a method for detecting the beam position is required. The beam position at the laser beam dump is monitored by four alignment laser beams which propagate parallel to the probing Nd:YAG laser beam and imaging systems installed outside the port plug [469].

ITER is scheduled to operate for about 4700 h of plasma, and when the beam is incident at a repetition rate of 100 Hz, the number of pulses on the beam dump will be on the order of $10^9$. The laser beam dump is made of molybdenum alloy and is designed to withstand high heat loads. To reduce the risk of laser induced damage, the internal structure of the laser beam dump has been proposed as shown in figure 116 [467]. It has a structure in which a large number of 0.5 mm molybdenum alloy sheets bent at several points are arranged in parallel with a clearance of 0.6 mm. In this structure, S-polarized beams are incident at a large angle of incidence (wide area). Therefore, the beam energy absorption density can be reduced to about 1/20 as compared with the case where the angle of incidence is 0. It has been experimentally shown that this structure is advantageous for the survival of the beam dump [468], and it has been confirmed that it can be manufactured [475].

<figure><img src="image_129.png" /><figcaption>**Figure 117.** Effective increase in the number of spectral channels for the different YAG probing wavelengths with various $T_e$ spectra overlaid. Reproduced from [472]. © 2015 IAEA, Vienna. All rights reserved.</figcaption></figure>

In order for the laser beam to be injected into the plasma, it passes through a vacuum window. The vacuum window is also the confinement boundary for tritium and beryllium. Vacuum windows are expected to be exposed to a total gamma dose of several MGy. A 10 MGy gamma ray irradiation was performed on fused silica windows with an antireflection coating, and then a laser injection test was performed. It was confirmed that the laser-induced damage threshold did not deteriorate [477].

### 6.4. Multi-color TS

The main calibration laser candidates for the core TS are: second harmonic of Nd:YAG 532 nm and the alternative Nd:YAG lasing wavelengths 946 nm and 1320 nm. Using different probing wavelengths also leads to an effective increase in the number of spectral channels (see figure 117). This can enhance the diagnostics capabilities if the electron VDF deviates from a Maxwellian for example under a strong ECRH/ECCD [472]. For the temperature range 0.5–40 keV the calibration coefficient for each channel can be determined with an accuracy of about 10%. This is not enough since the calibration coefficients should be determined with an accuracy better than 1%–2%. Therefore, averaging over a hundred pulses is required, that poses a problem for usage of a three level Ruby laser with a suitable wavelength of 694 nm which has a low repetition rate. Due to the extremely wide $T_e$ range there is no optimal choice of a supplementary laser, each choice has its own pros and cons. The use of both additional 1320 nm and 946 nm lasers seems to be preferable because this (a) allows successful operation when the spectral sensitivity changes quickly; (b) allows calibration of the system in the entire temperature range; (c) improves the accuracy in the case of non-Maxwellian electron VDF. Usage of a 1320 nm laser is sufficient for system calibration and accuracy improvement both for high $T_e$ measurements and non-Maxwellian VDF. The Nd:YAG 532 nm laser is relevant for calibration only in special ITER regimes within the limited $T_e$
range. The benefits of additional lasers can be summarized as follows [472]:

* Nd:YAG 1320 nm—advantages: significantly increases accuracy for high temperature measurements $T_e > 10$ keV, spectral calibration is possible for $T_e > 1.5$ keV, increased accuracy in the case of non-Maxwellian electron VDF.

* Nd:YAG 1320 nm—disadvantages: the laser is not yet commercially available, poor $T_e$ accuracy in double laser mode for $T_e < 1.5$ keV, different spectral channels will be calibrated at different temperatures.

* Nd:YAG 936 nm—advantages: calibration is possible over the full temperature range, could be used as the main laser if required.

* Nd:YAG 936 nm—disadvantages: poor $T_e$ accuracy in double laser mode for $T_e > 10$ keV, not useful for non-Maxwellian electron VDF determination, different spectral channels require calibration at different temperatures.

* Nd:YAG 532 nm—advantages: availability, all channels can be calibrated at one temperature, but this range is very short.

* Nd:YAG 532 nm—disadvantages: good $T_e$ accuracy in two laser mode only for $6 \text{ keV} < T_e < 15 \text{ keV}$, spectral calibration is possible only for $8 \text{ keV} < T_e < 20 \text{ keV}$, not useful for non-Maxwellian electron VDF determination.

Measurements of $T_e$ as high as 40 keV in the plasma core of a large fusion device such as ITER encounter a number of issues. The most reliable techniques that can be used are TS and ECE diagnostics. In case of ECE a large relativistic and Doppler broadening will have a significant impact on the spatial resolution of the measurements. For example for regimes with core $T_e(0) = 30$ keV obtaining the required spatial resolution of a/30 will only be possible for the outer region of the plasma column, $r/a > 0.5$ [273, 478]. In case of TS diagnostics, the main issue is a significant blue shift for backscattering geometry, see figure 118, while the operational wavelength range is bounded on the blue side (400–500 nm) due to the collection optics transmission and the strong background radiation [472, 479]. The problem will be made worse due to the darkening of the collection optics and fibers under strong neutron and gamma irradiation. In the worst case scenario the lower wavelength limit could be increased to 700 nm thus limiting the TS system dynamic range to a few keV [466, 472]. At high $T_e$ the $\sigma_{Te}/T_e$ values expected from the 1064 nm laser are limited by the ability to determine the shape of the spectrum, in particular it becomes more difficult to distinguish between increasing temperature and density. Possible solutions to this problem include using an alternative geometry with lower scattering angle, depolarization measurements (see section 2 for details) or using an additional probing wavelength shifted deeper to the near IR range. Nd:YAG lasers have emission lines at 1319 nm and 1338 nm. Using large scale Nd:YAG crystals together with high-power laser diode pumping allows one to realize such lasers operating with pulse energy as high as 1–2 J and a repetition rate of 50–100 Hz. At present a TS system is currently being prepared for experiments at Wendelstein 7-X [480] using a 1319 nm laser. The main advantage of such an approach is the effective shift of spectral channels, see

<figure><img src="image_130.png" /><figcaption>**Figure 118.** (a) Thomson scattering spectra for a scattering angle $\theta = 170^\circ$ and $T_e = 20–100$ keV. The spectra are normalized by the integral in the wavelength range ‘A’. (b) The ratio of integrals in wavelength ranges ‘A’ and ‘B’ versus $T_e$. Tor high $T_e$ the uncertainty drastically increases due strong interdependence of $T_e$ and $n_e$.</figcaption></figure>

figure 119. This allows to significantly reduce the temperature and density error correlation for 20–40 keV for a 1319 nm laser in comparison with a 1064 nm laser thus increasing the measurement accuracy of $n_e$ and $T_e$. Moreover, the other advantage of this technique is the inherent *in-situ* spectral calibration.

## 6.5. Polarimetric TS

Polarimetric TS is an advanced technique for the analysis of TS spectra that exploits the depolarization of the scattering radiation [470]. This is an effect barely noticeable in present-day fusion plasmas, but becomes important in plasmas with $T_e$ of the order of tens of keV, as it is expected in the high $T_e$ regimes of ITER [479]. Polarimetric TS has been suggested as a method to mitigate the loss of accuracy of the ITER CPTS system due to the fact that in the high $T_e$ regimes of ITER the spread of the TS spectrum exceeds considerably the spectral region considered to be useful for the detection of the TS signal. This is the wavelength region $400 \text{ nm} < \lambda < 1060 \text{ nm}$ and its limits are determined by the drop of the silicon APDs quantum efficiency on the long wavelength side and by the region of high background plasma light on the low wavelength side. The spectral region $400 \text{ nm} < \lambda < 650 \text{ nm}$ while also used is considered ‘at risk’ of transmission loss due to darkening of windows and fiber optics from neutron irradiation. When the plasma $T_e$ is high and the TS spectrum spreads outside the useful detection region, the reconstruction of the shape of the TS spectrum becomes difficult and the expected measurement errors increase. Polarimetric TS provides a method to mitigate this, contributing to keeping the $T_e$ and $n_e$ measurement errors low in the high $T_e$ regime [471, 481]. Recently
<figure><img src="image_131.png" /><figcaption>**Figure 119.** Intensity of the integrated scattering signal (into the full solid angle) as a function of $T_e$, for an input laser beam with linear and circular polarization. The intensity $I$ is normalized to its value $I_0$ for a non-relativistic plasma ($T_e$ approaching 0.) The completely polarized and the unpolarized components are also shown. At any $T_e$ the total intensity $I$ is identical for the two input polarizations and increases up to $I = 1.38 I_0$ for $T_e = 40$ keV. The intensity of the unpolarized component is higher for a circularly polarized input beam, reaching 0.25 $I_0$ at $T_e = 40$ keV compared to 0.20 $I_0$ for a linearly polarized input beam. Reproduced from [481]. © 2017 IOP Publishing Ltd and Sissa Medialab. All rights reserved.</figcaption></figure>

the depolarization of TS radiation has been experimentally observed in JET [482].

In order to describe the polarization properties of the scattering radiation it is convenient to use the Mueller calculus [483]. Theory shows [481] that the wavelength integrated radiation scattered by a high $T_e$, relativistic plasma is partially polarized and its Stokes vector can be written as $S = C_{PL}S_{PL} + C_{NP}S_{NP}$ in which $S_{PL} = (1, S_1, S_2, S_3)$ and $S_{NP} = (1, 0, 0, 0)$ are unit Stokes vectors representing the completely polarized and the unpolarized components of the radiation and $C_{PL}$ and $C_{NP}$ are their intensity coefficients. The quantity $P = C_{PL}/(C_{PL} + C_{NP})$ is the degree of polarization and $D = 1-P$ is conventionally called ‘depolarization’ [484]. Both are functions of the electron temperature only. The Stokes vector of the wavelength integrated scattering radiation is computed as $S(T_e, \theta) = M(T_e, \theta)S_{IN}$ where $S_{IN}$ is the Stokes vector of the incident laser beam and $M(T_e, \theta)$ is the Mueller matrix of TS [485]. The scattering signal detected in a given spectral interval can be calculated from the spectral Stokes vector $S_S(\lambda, T_e, \theta) = M_S(\lambda, T_e, \theta)S_{IN}$ in which $M_S(\lambda, T_e, \theta)$ is the spectral Mueller matrix of TS that can be calculated either numerically [484] or by a semianalytic method [486, 487]. Note that these expressions are valid for any polarization state of the incident laser beam. Let us now consider these quantities for TS from an ITER plasma. Figure 119 shows the dependence from $T_e$ of the intensities $C_{PL}$ and $C_{NP}$ and of the degree of polarization $P$ of the scattering radiation from a position near the center of the scattering region in the ITER core plasma.

The unpolarized component increases steadily with $T_e$ and at $T_e = 40$ keV accounts for up to 20% and to 25% of the intensity of an input beam with linear (as is typical) or circular polarization, respectively. For each input polarization the depolarization $D$ has a very weak dependence on the wavelength but in any spectral interval it is only a function of $T_e$. Polarimetric TS allows the determination of $T_e$ by measuring $D$. The technique can be implemented in any spectral interval and therefore does not suffer from limitations of the useful detected spectral range.

One way to exploit this depolarization effect in ITER is to include an additional polarimetric channel into a conventional filter polychromator as in figure 120 [481].

<figure><img src="image_132.png" /><figcaption>**Figure 120.** Layout of a TS detection scheme including n spectral and one polarimetric channel. If the input laser beam has linear polarization as usual, the quarter wave in input is not necessary and the scattering light is separated into the two linear ‘s’ and ‘p’ polarizations. If the input laser beam has right circular polarization, then the quarter-wave plate at 45° must be added to separate the scattering light into left ‘l’ and right ‘r’ circular polarizations. Reproduced from [481]. © 2017 IOP Publishing Ltd and Sissa Medialab. All rights reserved.</figcaption></figure>

The purpose of this polarimetric channel would be to detect the component of the scattering radiation with the polarization perpendicular to that of the incident laser beam, covering a spectral region as wide as possible within the useful detection range. This will provide an additional signal that, contrary to those of the conventional spectral channels, will be increasing with $T_e$, thus contributing to reducing the measurement error in the high $T_e$ range. In figure 121 we show the expected effect of such a polarimetric channel on the accuracy of the measured $T_e$ in ITER assuming that it has the same spectral extent of the full set of spectral channels [481].

These calculations show that the inclusion of the polarimetric channel becomes significant at about 20 keV and at $T_e = 40$ keV can decrease the $T_e$ and $n_e$ measurement errors by about 30% and 50% for an incident laser beam with linear and circular polarization, respectively. In these calculations the possibility of performing TS with a circularly polarized laser beam has been considered. In fact, in TS from a plasma, an incident beam with polarization different from linear perpendicular to the scattering plane has never been done.
<figure><img src="image_133.png" /><figcaption>**Figure 121.** Relative error on $T_e$ and $n_e$ expected using the spectral channels only and with the addition of a polarimetric channel, for linear and circular input polarization in the ITER CPTS. We assumed $n_e = 5 \times 10^{19} \text{ m}^{-3}$, laser energy in the plasma volume $E = 3.2 \text{ J}$, collection optics solid angle $\text{DW} = 3.14 \times 10^{-3} \text{ sr}$ (F/9.8), scattering length $L = 48 \text{ mm}$, scattering angle $\theta = 159^{\circ}$, APD excess noise factor $k = 2.5$ and no plasma light. The solid angle, scattering length and scattering angle are those of the measurement position in ITER at $R = 6.2 \text{ m}$. Reproduced from [481]. © 2017 IOP Publishing Ltd and Sissa Medialab. All rights reserved.</figcaption></figure>

The reason for this is that linear input polarization maximizes the scattering signal for any scattering angle. However in the almost backward scattering condition of ITER the difference between the scattering signal detected with linear and circular polarizations is hardly appreciable. In fact figure 121 shows that the accuracy of the spectral measurement is virtually the same at low $T_e$ whereas the inclusion of a polarimetric channel is more effective at high $T_e$ if the scattering is performed with a circularly polarized laser beam. In ITER however a polarizer, either linear or circular, depending on the polarization of the laser beam, must be inserted in the collection optics, to separate the two polarization components before the fiber optic bundles, as these will destroy the polarization of the scattering light [466, 488]. Such a polarizer must be sufficiently achromatic to operate on the entire detected spectral range and compliant with the requirement of neutron irradiation. A problem that may arise with an input circular polarization is due to the damage threshold of the laser beam dump, whose design was specific for a linearly polarized laser beam [489].

A second possibility to implement polarimetric TS in ITER is to use a double pulse technique, in which two laser pulses with different polarizations either linear horizontal/vertical or circular right/left are sent on the same path into the plasma with a small time delay. In this way only a single fiber bundle is necessary and also the polarimetric channel can be omitted because the unpolarized component will be measured by the same set of spectral channels on the second laser pulse. In this case however it is necessary to measure the power of the two pulses with good accuracy, and also introducing a beam dump of a different design. In conclusion, including a measurement of the unpolarized component of the TS signal may effectively improve the accuracy of the TS measurements at high $T_e$ in

<table><caption>**Table 30.** DTS measurement requirements.</caption><thead><tr><th></th><th>Coverage</th><th>Frequency</th><th>Accuracy</th></tr></thead><tbody><tr><td>$n_e$</td><td>$10^{19}-10^{22} \text{ m}^{-3}$</td><td>50 Hz</td><td>20%</td></tr><tr><td>$T_e$</td><td>1–200 eV<br>0.3–1 eV</td><td>50 Hz</td><td>20%<br>0.2 eV</td></tr></tbody></table>

the ITER CPTS system, counteracting the adverse effect of the limited detection spectral range.

Unfortunately the current constraints on the scattering geometry limit the advantages of polarimetric TS. An alternative access providing a more favorable scattering angle might become available for the ITER later experimental campaigns when the highest temperatures are expected.

## 6.6. DTS

The ITER DTS measures $T_e$ and $n_e$ in the outer divertor leg providing the link between upstream and target parameter measurements (e.g. from Langmuir probes). The diagnostic measurement requirements are presented in table 30. At the very low temperatures, corresponding to the region in which strong recombination and thus divertor plasma detachment will occur the requirement on measurement accuracy is reduced. This is reasonable in the sense that one of the most important roles of the DTS will be to provide experimental evidence that strong recombination is occurring in the target vicinity and to validate plasma boundary code simulations of the detachment behavior. Since the recombination rate increases more rapidly in the region with $T_e$ below ~0.5 eV in comparison with the interval ~0.5–1.0 eV, it suffices for code validation that experiments provide a measurement, which can distinguish roughly between these regions [490].

The required temporal resolution is based on particle confinement times in the divertor plasma. In ITER, these timescales equate to ~10 ms defining the minimum timescale upon which ‘global’ changes might be expected in the divertor plasma (not including, of course, very fast turbulent processes occurring on small spatial scales). The baseline time resolution of the DTS system of 20 ms (50 Hz) seems reasonable, especially given the possibility of achieving, if necessary, a factor of two higher resolution (10 ms) through simultaneous application of the spare laser system. The DTS optical layout uses crossed probing and viewing beams with front-end laser launcher located beneath the divertor cassettes. The laser chords (see figure 122) enter along the bottom of the divertor cassette and, then, travel upwards into the outer divertor SOL, passing through an ~20 mm gap between adjacent divertor cassettes. Several interchangeable probing chords with common collection optics improve the diagnostic reliability.

Laser chords #1 are very close to the separatrix, coinciding with the main stream to the outer divertor leg, and are the most informative for detachment monitoring and control. Laser chords #2 and #3 are less informative but are launched
<figure><img src="image_134.png" /><figcaption>**Figure 122.** DTS probing chords and spatial channel arrangement on two fiber bundles. Numbering starts from the bottom of the divertor leg.</figcaption></figure>

from mirrors located further from the outer divertor target but better protected from plasma interactions.

The standard processing of the TS spectrum assumes that the spectral width depends on $T_e$, while the total scattered signal intensity is linearly proportional to $n_e$. When studying a sufficiently dense and cold divertor plasma the Debye length approaches the differential scattering vector and the deviation of the TS spectral shape from Gaussian becomes very pronounced [491]. The intended processing of TS spectra is fitting the measured spectra with theoretical expectations, estimating the $n_e$ and $T_e$ measurement errors from simple analytical equations [472]. The DTS measurement capabilities were assessed by estimation of $n_e$ and $T_e$ relative errors for TS spectra with pronounced collective effects using the following algorithm:

(a) estimation of TS signals in spectral channels based on the known engineering parameters;

(b) multiple inverse problem solution (103 runs) recovering $T_e$ and $n_e$ from the TS signals simulated for given $n_e$ and $T_e$ with allowance for random deviations, where $k \sim 2.5$ APD excess noise, $N_i$ and $N_{bg}$ i-number of signal and background photoelectrons in ith spectral channel, $N_{amp} \sim 50$ contribution of the detector and amplifier noise reduced to the input of detector (minimal error of the scattered signal was assumed as 2.5%, taking into account probable systematic errors due to calibration);

(c) estimation of the $T_e$ and $n_e$ standard deviations from the obtained distributions of $T_e$ and $n_e$.

The assessed relative errors of the measured $T_e$ and $n_e$ for the probing chord #1 are summarized in figure 123. Figures 123(a) and (b) show that the measurement errors of both $T_e$ under $n_e = 2 \times 10^{20} \text{ m}^{-3}$ and of $n_e$ under $T_e = 0.3$ eV satisfy the requirements (plasma parameters near strike point). The same analysis for plasma parameters typical for upper part of the divertor leg is presented in figures 123(c) and (d).

The most important divertor plasma parameter is the degree of plasma detachment/attachment to divertor targets. Detachment requires ion-electron recombination in the plasma volume and this in turn requires the achievement of

<figure><img src="image_135.png" /><figcaption>**Figure 123.** The expected $\delta T_e/T_e$ and $\delta n_e/n_e$ for run #1514 of the SOPLS 4.3 software package corresponding to a peak power flux density of 8 MW m$^{-2}$. The background radiation intensity scaled from matching $n_e^2$ dependence with five-fold overestimating. The dashed lines mark acceptable accuracy. (a) $\delta T_e/T_e$ at $n_e = 2 \times 10^{20} \text{ m}^{-3}$. Solid curves are analytical approximation, crosses-errors estimated in a numerical experiment; (b) $\delta n_e/n_e$ at $T_e = 0.3$ eV; (c) $\delta T_e/T_e$ at $n_e = 10^{19} \text{ m}^{-3}$—minimum density for the top of the divertor leg; (d) $\delta T_e/T_e$ at $n_e = 5 \times 10^{19} \text{ m}^{-3}$—the minimum density for the bottom of the divertor leg. The curve dashed line takes into account the blackbody radiation of the divertor tiles at a temperature of 2000 °C.</figcaption></figure>

very low plasma temperatures (<1 eV) at which there is a transition from ionizing to recombining conditions. The ionization front location does not directly reflect the level of plasma detachment if there is at the same time a vertical displacement of the magnetic configuration, which can occur, for example, as a result of control errors or plasma instabilities. If these two effects can be sufficiently deconvoluted, it is possible to envisage the DTS as a potential sensor in a control loop designed to affect the degree of detachment. The profiles of $T_e$ along the DTS probing chord #1 are shown in figure 124. Arbitrary vertical shifts of -50 mm, 0 and +50 mm have been applied to the profiles, simulating strike point movements. The figure 124(b) shows that the ratio of areas under curve taken for the upper and lower regions of the probing DTS chord gives a matrix of points representing the extent of movement and ionization front position. An operational diagram of this type could be envisaged as a way in which the peak heat flux could be controlled by DTS.

In the presence of inaccessible components (e.g. vacuum windows, mirrors inside the vacuum system), the spectral characteristics of the in-vessel optical components, which may change with time, have to be calibrated. For DTS, the procedure must include relative calibration for the temperature, to validate the relative spectral response curves and avoid incorrect measurement of the TS spectrum shape, and absolute calibration for the density, which requires absolute sensitivity. The relative calibration of slowly varying spectral
<figure><img src="image_136.png" /></figure>

**Figure 124.** Illustration of deconvolution and decomposition of the influence on the plasma parameters in the outer divertor leg of two phenomena—different plasma detachment and different vertical displacement of the magnetic configuration. (a) $T_e$ as a function of distance $L$ along the probing chord, modeled for three runs of SOLPS #1511, #1514, #1540 with different degrees of detachment and, therefore, different peak stationary target power flux density on the target (10, 8, 6.5 MW m⁻²) and for three vertical displacements of the magnetic configuration −50, 0 and +50 mm; (b) relationship between (Integral A) and (Integral B) of the $T_e$ curves. Reproduced from [490]. © 2014 IAEA, Vienna. All rights reserved.

transmission is done comparing in different spectral channels the TS signals from two laser wavelengths [490]. The spectral channels used for measurements in the 5 eV–500 eV range can be calibrated using ratio of the TS signals from the 1064/1047 nm. If the $T_e$ range is increased up to several keV's, e.g. during ELM activity in which toroidally asymmetric structures carrying plasma with temperature characteristic of the H-mode pedestal, the corresponding spectral channels can be calibrated using signal ratio of 1064/946 nm.

Recent experiments carried out using two diode pumped Nd:YAG and Nd:YLF lasers have demonstrated a possibility to measure $T_e$ within the range $T_e = 0.02–1$ keV [492], see figure 125. The electro-optical shutters of the lasers were synchronized thus making it possible to detect TS signals with time delay of a few tens’ nanoseconds. The characteristics of the probing and measuring system that was used allow multi-laser TS measurements in a steady-state regime, even for the case when the spectral characteristic of the collection optics was artificially distorted.

The absolute calibration of the diagnostic as a whole, including the entire optical path will use a light source with known intensity under the same geometry and alignment conditions. The main approach is to use Rayleigh and/or Raman scattering on a gas target. Both have advantages and

<figure><img src="image_137.png" /></figure>

**Figure 125.** A comparison of the electron temperature measured at the Globus-M2 tokamak measured using multi-laser approach (spectral calibration of the system was not used, Nd:YAG 1064 nm and Nd:YLF 1047 nm lasers with 50 ns time delay) and conventional technique.

disadvantages. The main limitation in calibrating by Rayleigh scattering for the DTS is stray light. The expected stray light delay is longer than the DTS laser pulse duration (3–4 ns) and potentially can be temporally separated. The TS/Rayleigh scattering cross-section ratio for $N_2$ is 2210 at 1064 nm, which determines the signal value from the gas target with a density of $2.2 \times 10^{16}$ cm⁻³ (0.8 mbar) equivalent to the TS signal corresponding to $n_e = 10^{13}$ cm⁻³. The practical advantage of the less intensive Raman compared with Rayleigh scattering is in its broad spectrum, which makes it suitable for both absolute and relative calibration of the spectral channels (see detail in Mukhin *et al* [490]). Temporal shapes of LIF-derived HeI fluorescence can also be used as an extra method for absolute calibration of the collection system transmission. The fluorescence duration varies from 10 to 100 ns over the $n_e$ range $10^{17}–10^{21}$ m⁻³, its dependence on $T_e$ from 0.3 to 200 eV is much weaker [491].

A well-known practical problem facing optical diagnostics on large fusion devices, especially in ITER divertor, is the eroded first-wall materials deposited on the front-end optics during normal operation, when shutters are unable to protect. This leads to decreasing diagnostic signals and can result in substantial errors in the interpretation of the data. There are several mitigating techniques for this. The plasma sputtered materials can reach the divertor diagnostic optics during pressure jumps of ~1 ms pulse duration, when plasma blobs with core plasma parameters strike the divertor targets. The eroded material transport by diffusion is estimated using the MC code KITe (Kinetic Impurity Tracing) [493]. The impurity transport caused by convection flows will be minimized by a dust protecting box, situated between plasma and laser launcher following: (1) deviation of the hydrodynamic flow away from the optical axis and (2) absorption/pumping on the box walls [494–496]. The front-end laser optics will be thin quartz windows, which have to be self-cleaned by the laser ablation [497, 498]. The front-end collection optic will be cleaned by RF plasma [499–507]. The innovative diode-pumped Nd:YAG
<table><caption>**Table 31.** LIF measurement requirements.</caption><thead><tr><th></th><th>Range</th><th>Frequency</th><th>Accuracy</th></tr></thead><tbody><tr><td>n<sub>He</sub></td><td>10<sup>17</sup>–10<sup>21</sup> m<sup>-3</sup></td><td>50 Hz</td><td>20%</td></tr><tr><td>T<sub>i</sub></td><td>0.3–200 eV</td><td>50 Hz</td><td>20%</td></tr></tbody></table>

laser 2 J/50 Hz/1064 nm is the main laser and the auxiliary 1 J/50 Hz/946 nm Nd:YAG and 2 J/50 Hz/3 ns Nd:YLF lasers were developed for probing and calibration by TS at the shifted wavelength. The short laser pulse duration helps to reduce both the accumulated plasma background and to separate in time the TS and stray light signals [508]. A digital filter polychromator with ultra-low noise pHEMT based preamplifiers $F_{3db} = 250$ MHz and onboard digitized signal processing system was developed for DTS measurements with probing laser pulses of 3–5 ns [473, 509–513].

## 6.7. LIF in the divertor

LIF is an active laser spectroscopy diagnostic of atoms/ions in plasma. The technique is based on laser excitation of atom/ion allowed transitions and subsequent measurement of fluorescent signals. LIF allows measuring of densities ($n_a, n_i$) using corresponding collision-radiative models (CRM) and provides information about electron temperature ($T_e$) and density ($n_e$). Other types of LIF measurements are related to pumping spectral line scanning with a narrowband tunable laser: temperature of the atoms/ions from Doppler broadening, projection of the particle velocities from Doppler shift, the magnetic field from Zeeman splitting etc.

The LIF diagnostic shall measure helium density ($n_{He}$) and ion temperature ($T_i$) in ITER divertor according to the requirements specified in table 31. The main role of LIF is to measure helium density distribution, characterizing the efficiency of the helium ash removal. In combination with a vacuum pumping system data and gas analyzer, LIF allows monitoring the helium ash level in the main plasma. Besides this, LIF is a backup diagnostic used for measurements of $T_i$. In combination with a DIM, which is the primary $T_i$ diagnostics, LIF allows knowing the ion temperature distribution along the probing chords. The $T_i$ distribution is essential for evaluating the pressure balance in the divertor plasma [514].

The diagnostic is combined with the DTS by using common laser-injection and signal collecting optics. The beams will be launched into the outer leg of the divertor close to the separatrix with laser mirrors arranged beneath the divertor cassettes (figure 122). Both the DTS and LIF signals will be collected through the gap between neighboring divertor cassettes providing more than 20 independent spatial points. Separation of LIF and DTS signals will be carried out in the diagnostic room with selective mirrors transmitting LIF and reflecting DTS spectral range. The similar measurement chords of LIF and DTS allow using $n_e$ and $T_e$ from DTS for evaluation of $n_{He}$ with CRMs from LIF signals.

The LIF allows measuring both helium atoms ($n_{HeI}$) and hydrogen-like ions ($n_{HeII}$) densities. The measurements of $n_{HeI}$

<figure><img src="image_138.png" /><figcaption>**Figure 126.** Relative errors of $n_{HeI}$ (top) and $n_{HeII}$ (bottom) measurements (left ordinates) and $n_{HeI}$, $n_{HeII}$ densities respectively (right ordinates) for #2505 (DT) and #2327 (He) SOLPS runs. Reprinted from [516], Copyright (2019), with permission from Elsevier.</figcaption></figure>

can be realized by two spectroscopic schemes: a triplet—laser pumping the $1s2s \ ^3S \rightarrow 1s3p \ ^3P$ transition (388.9 nm) and viewing fluorescence in $1s3d \ ^3D \rightarrow 1s2p \ ^3P$ (587.6 nm), and a singlet—pumping $1s2s \ ^1S \rightarrow 1s3p \ ^1P$ (501.6 nm) and viewing $1s3d \ ^1D \rightarrow 1s2p \ ^1P$ (667.8 nm) [514, 515].

$n_{HeII}$ ion has a hydrogen-like structure of excited states limiting a set of spectroscopic schemes suitable for LIF. A new LIQ approach was developed for diagnostics of hydrogen and hydrogen-like ions [514]. LIQ for $n_{HeII}$ is based on quenching the most intense line in the visible range 468.6 nm ($n = 4 \rightarrow 3$ transition) by laser pumping to the upper $n = 5$ state (transition $n = 4 \rightarrow 5$, line 1012.3 nm).

1 kHz pulsed OPO is suitable both for $n_{HeI}$ (LIF) and $n_{HeII}$ (LIQ) measurements. Figure 126 shows the expected measurement accuracy for DT and Helium scenarios of ITER. OPO gives $n_{HeI}$ and $n_{HeII}$ relative errors <20% in the lower and middle parts of the probing beam, where $n_{HeI} > 10^{17} m^{-3}$.

Measurements of the ion temperature can be carried out on the main DT plasma impurity—helium ions. The measurement requires a tunable laser with a spectral width of about ten times narrower (at least) than the absorption line width. The SNR for $T_i$ should be higher than for density measurements. A short-pulse laser (like OPO) does not provide the required SNR; thus, the time-modulated laser is proposed to achieve the accuracy by the signal accumulation over the long time. The transition $n = 4 \rightarrow 5$ (1012.3 nm) is chosen for the
<figure><img src="image_139.png" /><figcaption>**Figure 127.** Simulation of the 1012.3 nm HeII spectral line shape for estimation of the $T_i$ measurement accuracy. Reprinted from [516], Copyright (2019), with permission from Elsevier.</figcaption></figure>

<figure><img src="image_140.png" /><figcaption>**Figure 128.** Relative accuracy (right ordinates) of $T_i$ LIQ measurements and $T_i$ distributions (left ordinates). Reprinted from [516], Copyright (2019), with permission from Elsevier.</figcaption></figure>

468.6 nm HeII line quenching due to minimal required laser power [516].

CRM calculation of the expected quenching and background signals allows estimating the accuracy of the $T_i$ using numerical simulations (figure 127) for Helium and DT ITER scenarios. The spectral line shapes were simulated considering calculated signals and noise sources. Doppler broadened emission spectra, once separated from Stark and Zeeman broadenings, provide a convenient means of making $T_i$ measurements. The laser beam is linearly polarized to reduce the intensity of π-component and of corresponding influence of Zeeman splitting.

Results of the $\Delta T_i$ estimations are shown in figure 128. The $T_i$ relative errors were calculated for the LIF/DTS ITER divertor geometry using the following equation: $\Delta T_i = (T_{init} - T_{meas})/T_{init}$, where $T_{init}$ is the temperature taken from the SOLPS and $T_{meas}$ is determined from the simulated spectral line shape. The simulation was carried out $10^4$ times for every observation point. The $n_e$ and $T_e$ (from DTS) were used to consider Stark broadening. Zeeman splitting was estimated for the magnetic field $B = 5$ T and for the toroidal laser polarization.

The calculations show that the ability to measure $T_i$ with relative errors <20% in the lower part of the probing chord is limited by the pronounced Stark effect. In the upper part of the divertor leg, the measurements are complicated by the low HeII ions density.

The LIQ technique can also be applied for hydrogen (deuterium, tritium) diagnostics in the ITER divertor [491] The laser pumping of one of the Paschen lines (e.g. $P_α = 1875$ nm) quenches $H_α = 656$ nm. Compared to the conventional LIF for hydrogen with laser pumping and viewing of the fluorescence in $H_α$, or laser photoionization (LII) of $H_β = 486$ nm by TS laser, LIQ combines benefits of both LIF and LII. The LIQ requires a low-power laser and allows spectral line scanning based measurements like LIF. The wavelength shift of detected radiation versus pumped transition reaches hundreds of nanometers and the laser stray light can be easily blocked as well as in LII.

The CRM assessments of the hydrogen density LIQ diagnostic in ITER divertor show possibility to measure $n_{HI} > 10^{17} m^{-3}$ with relative errors better than 20% by 1 kHz pulsed OPO. Using a time-modulated narrowband tunable laser allows scanning of the pumping $P_α = 1875$ nm spectral line shape. Therefore, LIQ gives the possibility to measure local $n_H/n_D$, $n_D/n_T$ ratios.

LIQ diagnostics of hydrogen atoms was already tested at Globus-M tokamak. The $H_α$ signal was suppressed by laser radiation with $\lambda = 1005$ nm exciting transition $n = 3 \rightarrow 7$ (using OPO laser: $\Delta\lambda = 2000$ pm, $\tau = 10$ ns, $E = 2.2$ mJ). As well as OPO, the time-modulated fiber laser with $\lambda = 1875$ nm, $\Delta\lambda = 1200$ pm, $P_l = 5$ W and repetition rate of 1 MHz was specially designed and manufactured for the $n_{HI}$ measurements in SOL and divertor plasma of Globus-M. The experiments show the possibility of measuring $n_{HI} > 10^{16} m^{-3}$ averaging quenching signals over 10 ms in SOL plasmas.

Another useful approach of LIF in ITER is the electron density measurements basing the analysis of HeI fluorescence pulse shapes [491]. The CRM developed for $n_{HeI}$ describes the temporal behavior of the fluorescent signals depending on the laser pulse shape, exciting transition and local values of $n_e$ and $T_e$. Particularly, LIF can be used to measure $n_e$ in the range of $10^{17}-10^{21} m^{-3}$ under $T_e = 0.3-200$ eV by using pulsed OPO laser.

The CRM simulation shows that the fluorescence duration (FWHM) varies from 10 to 100 ns over the $n_e$ range of $10^{17}-10^{21} m^{-3}$ while dependence on $T_e$ in the range of 0.3-200 eV is much weaker. The lower limit of $n_e > 10^{17} m^{-3}$ is determined by reduced dependence on $n_e$ since relaxation of the excited level and hence the fluorescence signal duration is determined mainly by spontaneous emission. Test measurement of $n_e$ was carried out in Globus-M tokamak using TS collection optics. The LIF signal temporal shape satisfies $n_e = (2.00 \pm 0.65) \cdot 10^{19} m^{-3}$ that correlates well with the TS-data $n_e = (2.4 \pm 0.2) \cdot 10^{19} m^{-3}$.

Comparison of $n_e$ by LIF with $n_e$ by DTS, which depends on the collection system transmission, can also be used as an
extra method for absolute in-situ calibration of the collection optics transmission in ITER.

## 6.8. Toroidal interferometer-polarimeter (TIP)

The primary system planned for real-time density control in ITER is the five channel combined TIP. TIP will also contribute to density profile reconstruction and play a secondary role as a diagnostic of core density fluctuations [4]. Initial design studies of the TIP were carried out in 1998 [517] and, in 2013, a more detailed design was published [518] following the successful CDR at ITER. The TIP presented at the CDR featured a traditional two-color vibration compensated interferometry measurement [519–523] with the 10.59 µm CO₂ and 5.42 µm CO laser lines combined with a 10.59 µm CO₂ polarimetry measurement of Faraday rotation using the R and L-wave technique [524–526]. Following prototyping activities, it is now envisioned that the CO laser will be replaced by a 4.6 µm QCL on ITER [527].

The two-color interferometry approach on TIP is an established technique that relies on using two separate colinear interferometers operating at different wavelengths to deconvolve dispersive plasma-induced phase shifts and motion induced phase shifts. In the cold-plasma approximation, the line-integrated density from TIP will be given by

$$ \int n_e dL = \frac{\lambda_{CO2}}{r_e (\lambda_{CO2}^2 - \lambda_{QCL}^2)} \left[ \phi_{CO2} - \frac{\lambda_{QCL}}{\lambda_{CO2}} \phi_{QCL} \right] $$

where $r_e = 2.82 \times 10^{15}$ m, $\lambda_i$ and $\phi_i$ are the wavelengths and interferometric phase shifts respectively of each laser. Plasma-induced phase shifts of the 10.59 µm CO₂ laser beams in ITER are expected to be $10^3–10^4$ degrees and motion-induced phase shifts will likely be $10^4$ Deg. (1 mm) or larger [518]. In contrast, the polarimeter measures a phase shift given by

$$ \phi_{pol} = 2C_p \lambda_{CO2}^2 \int n_e \vec{B} \cdot d\vec{L} $$

with $C_p = 2.62 \times 10^{-13}$ rad T⁻¹ and B is the local magnetic field, from which the Faraday rotation or field weighted line-integrated density can be obtained. Because polarimeter phase shifts are expected to be less than 180 Deg. [518] and the magnetic field is known, the inclusion of polarimetry allows the two-color system to recover unambiguously from fringe skips at all densities, up to and beyond the Greenwald limit as well as the potential to use the polarimeter itself for feedback density control.

The present TIP layout on ITER is shown in figure 129, where it can be seen that each of the five TIP chords will originate in the diagnostic hall from five independent optical tables. The plasma leg beam lines will be enclosed and transmit each chord to and from the ITER vessel through multiple containment zones. In equatorial port 9 (EP9) an optical labyrinth will transmit each chord through the FW where they will fan out and be sent to a set of corner cube retroreflectors that are placed in either of the other equatorial ports or neutral

<figure><img src="image_141.png" /><figcaption>Figure 129. TIP Layout on ITER. (a) View showing TIP layout spanning various zones. (b) TIP chords in ITER vessel and retroreflector (CCR) locations.</figcaption></figure>

beam ducts. The optical train in EP9 features five independent first mirrors and a common FW hole, to minimize penetration sizes and to reduce risks associated with deposition, erosion and neutron streaming.

While elements of TIP including separate CO₂ laser-based two-color interferometry [520, 521] and R&L-wave polarimetry [528, 529] had been made previously, at the time of the CDR, no previous system had been demonstrated with the combination of characteristics required for TIP, i.e. combined two-color IR interferometry and polarimetry measurements, long ~100 m beam paths and steady-state ~1000 s, low-noise measurements. In 2015–2016, a fully functioning TIP lab prototype with the scale of that expected in ITER was constructed and tested [527]. Those tests included several phases, from bench testing to a full 120 m beam path with FPGA-based DAQ and a feedback alignment system that was subjected to motions typical of an ITER bake cycle. Over the course of the testing, the initial conceptual design was refined significantly as issues were identified and overcome. As mentioned, a major change was the shift from a CO to QCL used as the shorter wavelength vibration compensation laser (a 5.22 µm laser in lab tests but planned as a 4.6 µm laser for ITER). Following lab tests, the TIP prototype was moved and installed on DIII-D with encouraging initial results; DIII-D TIP prototype measurements compare well with expectations and the system design is capable of meeting ITER measurement requirements.

The DIII-D TIP prototype layout is shown in figure 130, where the ~100 m roundtrip beam path for the plasma and reference legs can be seen. A detailed description of the table and components are given in Van Zeeland et al [527] and [530]. A key element not shown is the optical table enclosure which reduces airflow, dust and regulates the temperature to help maintain alignment of the components. Like the ITER TIP, the DIII-D prototype features a tangential view
TIP Prototype Layout on DIII-D

<figure><img src="image_142.png" /><figcaption>**Figure 130.** TIP prototype layout on DIII-D.</figcaption></figure>

and feedback alignment system to keep the system aligned throughout motion associated with pulsed fields during a discharge as well as daily thermal cycles.

TIP prototype measurements of a DIII-D plasma are shown in figure 131, where the interferometer and polarimeter measurements are shown in figures 131(a) and (b) respectively along with the measured phase shifts. During DIII-D discharges, the measured phase resolution for the polarimeter and interferometer is 0.05 Deg. (100 Hz bandwidth) and 1.9 Deg. (1 kHz bandwidth), respectively [530], which are to be compared to the ITER target values of 0.1 and 10 Deg [518]. The corresponding line-integrated density resolution for the vibration-compensated interferometer is $nL = 1.5 \times 10^{18} \text{ m}^{-2}$, and the magnetic field-weighted line-integrated density from the polarimeter is $nBL = 1.5 \times 10^{19} \text{ Tm}^{-2}$. Both interferometer and polarimeter measurements compare well to expectations based on calculations using a synthetic TIP diagnostic which takes as inputs TS measured density profiles and EFIT magnetic equilibrium reconstructions [530]. As described in [531], the TIP prototype active feedback alignment system was essential to obtaining these low-noise data. In addition to line-integrated equilibrium density, $1 \text{ MS s}^{-1}$ TIP prototype data clearly demonstrate the ability to resolve core density fluctuations such as Alfven eigenmodes in the 100 kHz range (see figure 7 of [531]).

While TIP prototype results are encouraging, several open issues and design questions remain. TIP faces many of the same issues discussed in the context of other laser-based diagnostics discussed in this summary including primarily first mirror degradation. At present, no shutter or mirror cleaning is envisioned, passive mitigation through the use of IR wavelengths, a single small aperture and a novel retroreflector design is the planned approach. An important issue that was identified in testing is atmospheric interaction and specifically, dispersion causing uncompensated interferometric phase shifts due to environmental variation. An example of this effect is given in figure 30 of [527], where a 0.5 C change over ~40 m of the air-filled beam line caused an uncompensated phase shift of about 25 Deg. or 250% of the TIP interferometer error budget. R&D work will be needed to identify a path forward to handle significant temperature variations but one option includes positive pressure in the beam lines with clean dry air and monitoring temperature variations then removing the uncompensated phase shifts through some experimental fit

<figure><img src="image_143.png" /><figcaption>**Figure 131.** TIP prototype measurements of a DIII-D plasma. (a) Two-color interferometer. (b) Polarimeter.</figcaption></figure>

to the effect. A similar approach has been employed on W7-X [532]. An alternative but difficult approach is the evacuation of the beam lines. Another open issue is the TIP primary vacuum window; leading candidates are $BaF_2$ and $ZnSe$. $BaF_2$ has been used on several devices including the DIII-D TIP prototype, has acceptable transmission and a low Verdet constant; however, it is sensitive to thermal shock and water exposure. While $ZnSe$ is stronger than $BaF_2$, its higher index of refraction requires an AR coating (which use proprietary coatings with no guarantee of reproducibility) and Faraday rotation in the window must be corrected for to meet polarimeter measurement requirements. Presently, neither window material has been formally approved for use on ITER.

### 6.9. PoPola

PoPola injects thirteen far-IR (FIR) laser beams (wavelength: 119 µm) into the plasma as illustrated in figure 132. PoPola measures the polarization change of the FIR laser beams attributed to interaction between the plasma and the FIR laser beams. An equilibrium reconstruction code estimates the profile of plasma current, or equivalently safety factor $q(r)$, by using the PoPola measurement data together with other diagnostics such as magnetics and CPTS. The $q$ profile is used for physics, real-time suppression of neo-classical tearing modes (NTMs) and real-time reverse shear control. The design activity has been carried out, and progress has been made, for instance a retroreflector [533], a FIR laser, gamma and neutron irradiation tests of a piezo actuator, magnetooptic effect of quartz (window material), remote alignment method of laser beam position and passive stabilization of laser beam position (called as 'flexible light guide') [534–536]. This section focuses on two topics; equilibrium reconstruction taking into account both $T_e$ effect and coupling between the Faraday effect and the CM effect, and a measurement technique.
<figure><img src="image_144.png" /><figcaption>**Figure 132.** PoPola beam paths through *rz* cross section of ITER.</figcaption></figure>

The $T_e$ effect and coupling between the Faraday effect and the CM effect are not mentioned in the previous papers, IPB [3] and Progress in IPB [4], but are of significant importance are for discussing the measurement accuracy of *q* profile identification. Polarimetry involves the measurement of a change in the polarization state, caused by the Faraday effect and the CM effect. The Faraday effect mainly changes an orientation angle, $\theta$, of the polarization state and depends on the electron density of plasma, $n_e$, and magnetic field parallel to probing laser light, $B_\|$. The CM effect mainly changes an ellipticity angle, $\epsilon$, of the polarization state and depends on $n_e$ and magnetic field perpendicular to probing laser light, $B_\perp$. When $T_e$ is high, changes of $\theta$ and $\epsilon$ depend on not only $B$ and $n_e$, but also $T_e$ owing to breakdown of the cold plasma dispersion relation. For burning plasmas such as ITER plasma, the electron temperature is expected to be several tens of keV, and thus the $T_e$ effects on the dispersion relation becomes important [537, 538]. In addition to the $T_e$ effect, the coupling between the Faraday effect and the CM effect needs are also significant in the measurement condition of PoPola. Finally, the change of $\theta$ and $\epsilon$ are not expressed by using simple line-integral form, but a vector differential equation called a Stokes equation. So, it was an open question whether the *q* profile can be reconstructed with required accuracy (10% [197]) when the $T_e$ effect and the coupling between the Faraday effect and CM effect are not negligible. The PoPola design team demonstrated for the first time that the *q* profile can be reconstructed even when the governing equation for the polarization change is given by the complex vector differential equation (i.e. Stokes equation) [539]. The PoPola design team found

that the required accuracy of 10% cannot be met in the case of the steady state scenario of ITER plasma operation when the $T_e$ effect is not taken into account. Thus, the Stokes equation taking into account the $T_e$ effect needs to be solved for reconstructing the *q* profile. Moreover, it was found that the accuracy of *q* profile reconstruction using both $\theta$ and $\epsilon$ are higher than that using just $\theta$. This is because the data of $\epsilon$ includes the information of the poloidal magnetic field because of the coupling and also includes the information of the toroidal magnetic field caused by the plasma current. Although the measurement of $\epsilon$ was originally proposed for measuring line-averaged electron density, the measurement of $\epsilon$ is now implemented by PoPola for improving *q*-profile identification. The study regarding the *q*-profile identification leads to the specification of PoPola, which is that the measurement accuracy of $\theta$ and $\epsilon$ are 1° and 6°, respectively, and target to 0.1° and 0.6°, respectively.

The next topic is the measurement technique. This is one of future works that was listed in the Progress in IPB [4]. The PoPola design team examined the four techniques; the rotating waveplate method, the dual PEM method, the Dodel-Kunz method and JET polari-interferometer method. The rotating waveplate method was selected as the preferred option. The major reasons of the selection is the applicability to the wavelength of 119 µm, the capability of measuring the ellipticity angle ($\epsilon$), the robustness to the deformation of in-vessel mirrors, mechanical vibration and the instability of laser frequency. This section explains only the selected method, and detailed information about the other methods is reported in [540]. The rotating waveplate method consists of a rotating quarter waveplate and a linear polarizer. Detector signal is modulated by the rotation of the quarterwave plate and Fourier analysis of the detector signal provides information identifying the polarization state. It is a well-known technique in the field of polarimetry, but there was no previous experience of measuring at the required time resolution of 10 ms in real time. The theory of the rotating waveplate method is simple and well-known, but practical application with high accuracy is not simple under non-ideal conditions such as concentric distribution of birefringence and non-flatness and wedged shape of the waveplate. The applicability of the rotating waveplate method to PoPola was demonstrated by using both He-Ne laser and quarter waveplate of multi-order retardation [541]. No drift of measurement data was found for 1000 s. The achieved time resolution was 3.3 ms, and the achieved precision of the orientation angle, ellipticity angle, and the Stokes parameter was 0.33°, 0.096°, and 1.2%, respectively. This error will be reduced by using the FIR laser because the error is attributed the run-out of the rotating waveplate, and the run-out normalized by wavelength of FIR laser is much smaller than that of He-Ne laser.

## 6.10. Expanded measurement capability potential for PoPola

Since publication of the Progress in IPB [4], two new measurement capabilities have been identified as possible for PoPola:
<figure><img src="image_145.png" /><figcaption>**Figure 133.** Faraday-effect polarimeter using R- and L-wave technique observed magnetic turbulence during a density ramp in a DIII-D H-mode discharge: (a) line-averaged electron density; (b) spectrogram of Faraday fluctuation measurement of magnetic turbulence.</figcaption></figure>

(1) internal magnetic field fluctuations and (2) plasma vertical position.

### 6.10.1. Internal magnetic fluctuation measurement.
Interest in internal measurements of magnetic fluctuation and non-axisymmetric 3D effects in ITER have significantly increased, due to needs of early detection of disrupting MHD instabilities [542] and predictions of unstable electromagnetic turbulence in ITER relevant plasmas [543]. Internal magnetic fluctuation measurements, a capability not available from any currently planned ITER diagnostic, may be the key to address these issues. Recent advances in polarimetry have shown that magnetic fluctuation measurements can be accomplished by using the R- and L-wave technique [526], which is also employed for the ITER TIP diagnostic described in section 8. The technique can detect the Faraday effect with low noise at sub-microsecond time scale, enabling measurements of Faraday effect fluctuations. The noise can be further reduced by using two detectors at each chord to suppress electronic noise via the application of correlation techniques [544]. The Faraday fluctuation measured via a horizontal radial chord near the plasma magnetic axis is dominated by magnetic fluctuations along the chord [545], providing a direct internal magnetic fluctuation measurement. In addition, electron density fluctuations along the same chord can be simultaneously measured by adding a local-oscillator [526]. By using the R- and L-wave technique for chords at or near the magnetic axis, broadband magnetic turbulence (100–500 kHz) is observed in ELMy H-mode DIII-D plasmas, shown in figure 133 which correlates with a density ramp and pedestal performance degradation [546]. Determined by the probe beam width (~6 cm), the similar measurements on ITER would be sensitive to low-k ($k\rho_s < \sim 0.1$ for a 10 keV plasma in ITER) plasma instabilities, such as electron drift waves, MHD and fast ion driven instabilities.

<figure><img src="image_146.png" /><figcaption>**Figure 134.** Layout for combination of R- and L-wave technique and rotating waveplate method in PoPola.</figcaption></figure>

For PoPola, to achieve the magnetic fluctuation measurement capability, one would need to implement R- and L-wave technique for the chord(s) nearest the magnetic axis. This can be done without compromising PoPola's primary mission to determine the *q*-profile. There are two possible approaches. First approach is to combine the rotating waveplate method described in section 9 with the R- and L-wave technique, as illustrated in figure 134. The beam $\omega_0$ with linear polarization is the probe beam for the rotating waveplate method and generated directly by laser. The probe beams of the R- and L-wave technique, beams $\omega_R$ and $\omega_L$ with counter-rotating circular polarizations, can be generated either by additional lasers with ~10 MHz frequency difference from $\omega_0$, or by splitting a portion of the $\omega_0$ beam and shifting frequency using an acousto-optic modulator. The rotating waveplate and R- and L-wave methods are integrated by combining all the probe beams together before leaving the optical table, and splitting the return beams into two parts while preserving their polarizations. By appropriately separating the probe beam frequencies, the presence of beam $\omega_0$ will not affect the R- and L-wave measurements. The presence of beams $\omega_R$ and $\omega_L$ do affect the rotating wave plate measurement, which is avoidable, e.g. by chopping beam $\omega_R$ and $\omega_L$ periodically: a clean rotating waveplate measurement is accomplished when beams $\omega_R$ and $\omega_L$ are blocked. The chopping frequency is optimized so that good rotating waveplate measurement at the millisecond scale and R- and L-wave measurement at the microsecond scale are realized. Another approach is to only use the R- and L-wave technique for the desired chord(s) near the magnetic axis. In this approach, the R- and L-wave measurement would also provide the *q*-profile constraint. A potential concern is that non-negligible contamination from CM effect in the R- and L-wave measurement [541, 547] may affect the accuracy of *q*-profile. However, this concern is largely alleviated by realizing that the absolute error due to the contamination is small (sub-degree level) for chords near the magnetic axis.

### 6.10.2. Plasma position measurement.
Nine chords of the PoPola system view the plasma horizontally from an equatorial port and can therefore be exploited to provide a measurement of the plasma vertical position. The plasma vertical position
<figure><img src="image_147.png" /><figcaption>**Figure 135.** Vertical plasma position on DIII-D showing agreement between Faraday-effect (using 3 central chords) and EFIT results.</figcaption></figure>

is determined by the magnetic axis where the radial magnetic field goes to zero and horizontal position is determined by the magnetic axis where the vertical magnetic field goes to zero. To directly measure the radial magnetic field, the double-pass, radially-viewing, PoPola system can be used.

The plasma vertical position, $Z_F$, can be related to the Faraday measurements, $\psi_F$, via $Z_F = -\psi_F(0)/\psi'_F(0)$, where $\psi'_F(0) \approx \Delta\psi_F/\Delta z$ [531]. The equation gives the center position of the innermost flux surface, $Z_F$, directly from the Faraday rotation measurement. The $\psi'_F$ is calculated from the slope $\Delta\psi_F/\Delta z$ at $z=0$. This is a simple linear relation between position and Faraday rotation measurements near the magnetic axis, $z \sim 0$, which provides essential information for a PCS [548]. This vertical position is independent of the time derivative, therefore, it is NOT subject to noise from long-time integration. Zero-crossing of Faraday effect measurement provides higher sensitivity to vertical position than other methods such as density or SXR emissivity profile measurements. It should be noted that vertical position is determined directly from the two line-integrated measurements and no inversion is needed. This approximation can enhance the capability of performing real-time control since the line-integrated Faraday-effect measurement can have excellent time resolution (especially when using the R- and L-wave approach) and requires no other equilibrium information.

This approach was recently demonstrated on both the EAST [531] and DIII-D tokamaks where horizontally viewing polarimetry chords are available. An example from DIII-D is shown in figure 135, where the Faraday measurement agrees well with the EFIT results (using external magnetics) through a programmed 25 cm vertical position scan. This approach can be used to independently and non-inductively measure the plasma position thereby providing a cross check for flux loop measurements. Since the Faraday measurements are internal to the plasma, they can also be exploited for plasma control related to vertical displacement events or disruptions.

<figure><img src="image_148.png" /><figcaption>**Figure 136.** Basic concept of homodyne dispersion interferometer.</figcaption></figure>

The plasma position, is the ratio of Faraday rotation to its gradient and is therefore insensitive to the CM effect itself as it largely cancels. This result simplifies use of polarimetry measurements for plasma position sensor in a PCS.

The value of a PoPola measurement of plasma position is threefold for ITER; (1) provides an independent, non-inductive measurement of position, (2) can be used to cross check and validate flux loops or serve as a backup in the event of flux loop failure in harsh ITER radiation environment, and (3) can provide an early warning of plasma vertical position changes related to instability or disruption.

### 6.11. DI for ITER

As described in the previous section, the two-color interferometer approach employed by the TIP can separate the phase shifts caused by a plasma and by mechanical vibrations. DI [549] is a relatively new interferometry configuration and is essentially a variation of the traditional two-color interferometer in that it also uses two different wavelengths. While a standard two-color interferometer measures the phase shift for each wavelength separately, and can solve for vibrations and plasma-induced phase shifts independently, the phase shift measured with DI contains essentially no vibration-induced phase shift, instead measuring the phase shift caused by dispersion of a medium. An additional advantage of DI is a simpler optical configuration. DI uses a single laser source and detector and has no requirement for a path-length matched reference chord.

The concept of DI is shown in figure 136. While a standard two-color interferometer is equipped with two laser sources with different wavelengths, DI utilizes a nonlinear crystal to generate the second harmonic component as the second color. The first nonlinear crystal generates the second harmonic component, after which a combination of co-linear fundamental and second harmonic components are launched through the plasma. After passing through the plasma, another nonlinear crystal is used to generate additional second harmonic power and the remaining fundamental component is blocked out. The dispersion interferometer measures the interference between the two second-harmonic beams only. In this approach, the vibration-induced phase shift is optically subtracted, because the two beams experience the same vibrations. In an ideal configuration, the measured phase shift results solely from dispersion through the plasma [549].

Starting with a homodyne configuration, several diagnostic techniques like phase modulation [550–553] and heterodyne detection [554], have been applied to the DI for improvements of time and density resolutions. As an example, the DI configuration on LHD is shown in figure 137. Since DI
<figure><img src="image_149.png" /><figcaption>**Figure 137.** Dispersion interferometer on LHD.</figcaption></figure>

is relatively insensitive to mechanical vibrations, the optical frame, where most optics were placed, had no vibration isolation and was directly installed to the diagnostic platform in the LHD hall. The laser source was a CO2 laser with a wavelength of 10.6 mm and output power of 7.5 W. The 5.3 mm component was generated using a AgGaSe2 crystal. The LHD DI utilized phase modulation with a PEM to improve the density (the phase) resolution. The probe beam injected into the LHD VV was reflected by a corner cube mirror installed inside the LHD vessel. The reflected probe beam returned to the optical frame, and was detected. The detailed evaluation method of the plasma phase shift and the electron density is described in [552]. The line density resolution, which was defined as the standard deviation of a 3 s long signal acquisition, was $2.5 \times 10^{17} \text{ m}^{-2}$ (0.6 deg.), with a response time of 0.1 ms. The measured electron density showed good agreement with the far IR laser interferometer. The response time was good enough to measure rapid density increases due to pellet injection [552].

The relatively simple configuration of a DI reduces its maintenance and the risk of components failure. The absence of a reference leg makes the system's footprint smaller enabling integration into limited spaces. Advantages of the DI like excellent line density resolution, ease of integration and reliability are reported from several devices, such as TEXTOR [550, 551], LHD [552], W7-X [553], and DIII-D [554]. They motivated the inclusion of DI in the ITER diagnostic set. Figure 137 shows a conceptual view of Density Interferometer Polarimeter (DIP) on ITER. DIP is a CO2 laser-based phase modulated DI combined with a polarimeter using PEMs. As with TIP, the polarimeter is incorporated in order to compensate for fringe skips, should they occur. The fundamental component, which is typically filtered out after the 2nd nonlinear crystal, is used for the polarimetry measurements. A PEM-based polarimeter has been previously demonstrated on JT-6U [555] and does not affect DI measurements, since it measures the polarization angle of the wavelength components which is not used finally in the standard DI. The combination of the DI and the PEM polarimeter was tested on a bench

<figure><img src="image_150.png" /><figcaption>**Figure 138.** Concept of the combined dispersion interferometer and polarimeter as will be implemented on ITER's DIP.</figcaption></figure>

as shown in figure 138. In the prototype DIP, simultaneous measurements of the interferometer phase shift and the polarization angle were successfully demonstrated [556].

## 6.12. Conclusion

Since the ITER physics basis update in 2007, laser aided diagnostics have advanced considerably with many diagnostics going from outline concepts to detailed integrated design solutions ready for implementation.

With this advancement in design many particularly challenging diagnostic components have gone through testing and prototyping stages. For example, following prototyping and testing the TIP diagnostic now plans to use a QCL as its' secondary laser as opposed to a CO laser in addition to the main CO2 laser. The ETS diagnostic has tested laser beam dump designs for longevity and a new laser beam combiner based on a rotating half-wave plate has been developed. Similarly, demonstration of a rotating waveplate for PoPola has been shown to measure Stokes parameters with no drift for over 1000 s.

The challenges of the diagnostic designs have lead to development of new hardware. New lasers have been specifically developed for the DTS diagnostic with diode pumped 2 J 50 Hz 1064 nm main laser and calibrating 1047 nm and 946 nm lasers with high pulse power and repetition rates. Additionally a digital polychromator design has been developed integrating digitizers within the traditional filter polychromator unit. There are still considerable challenges to be overcome, for example the CPTS diagnostic would greatly benefit from a free form front optic design. The ability to integrate such a large free form mirror into ITER and shutter and clean it has yet to be demonstrated.

New diagnostic techniques have also been developed since the last ITER physics basis update. Laser induced fluorescence has been proposed as a technique to measure Helium density and ion temperature. Similarly laser induced quenching, which pumps individual transition lines using an OPO laser, provides
the possibility to measure the $n_H/n_D$ and $n_D/n_T$ ratios. Existing diagnostics are expanding their remits taking on new measurements. The PoPola system is now expecting to measure plasma vertical position and magnetic fluctuations. DI is also proposed which using a simple optical configuration can provide phase shift measurements without any vibration induced phase shift.

The theoretical basis for measurements is also being advanced. Recent simulations have also shown that the PoPola must take into account the $T_e$ effect in order to accurately diagnose the q-profile. While the proposal for polarimetric TS is not new, there has been very significant development in the theoretical basis behind this measurement capability. The development of polarimetric TS has been driven by that requirement being highlighted by ITER due to the need to accurately diagnostic high $T_e$.

Importantly laser aided diagnostic techniques proposed for ITER are being trialed on existing Tokamaks. On Globus-M a 1 MHz OPO laser was deployed in order to measure $n_e$ using laser induced quenching. The TIP combined two-color interferometry and polarimetry measurements were first tested in the lab and then installed on DIII-D. The first observation of polarimetric TS measurements were recently obtained on JET. Vertical position measurements using PoPola were demonstrated on DIII-D. DI was demonstrated on a wide number of machines including LHD and TEXTOR. One of the key outputs of these tests on existing machines is it shows potential issues that need resolution, for example interferometric phase shifts were observed by TIP due to small temperature differences along the beam path. These lab demonstrations of advanced diagnostics and techniques on existing machines are crucial to de-risk the laser diagnostic implementation for ITER.

## Appendix I—Abbreviations

<table><thead><tr><td>Abbreviation</td><td>Meaning</td></tr></thead><tbody><tr><td>AR</td><td>Anti-reflection</td></tr><tr><td>APD</td><td>Avalanche photo diode</td></tr><tr><td>CCR</td><td>Corner cube reflector</td></tr><tr><td>CM</td><td>Cotton-Mouton</td></tr><tr><td>CPTS</td><td>Core plasma Thomson scattering</td></tr><tr><td>CRM</td><td>Collisional radiative models</td></tr><tr><td>DAQ</td><td>Data Acquisition</td></tr><tr><td>DFW</td><td>Diagnostic first wall</td></tr><tr><td>DI</td><td>Dispersion interferometry</td></tr><tr><td>DTS</td><td>Divertor Thomson scattering</td></tr><tr><td>ECCD</td><td>Electron cyclotron current drive</td></tr><tr><td>ECE</td><td>Electron cyclotron emission</td></tr><tr><td>ECRH</td><td>Electron cyclotron resonance heating</td></tr><tr><td>EFIT</td><td>Equilibrium FIT</td></tr><tr><td>EQE</td><td>Effective quantum efficiency</td></tr><tr><td>ETS</td><td>Edge Thomson scattering</td></tr><tr><td>FIR</td><td>Far infra-red</td></tr><tr><td>FPGA</td><td>Field programmable gate array</td></tr><tr><td>FWHM</td><td>Full width at half maximum</td></tr><tr><td>HeNe</td><td>Helium neon</td></tr><tr><td>LHD</td><td>Large helical device</td></tr><tr><td>LIDAR</td><td>Light imaging detection and ranging</td></tr><tr><td>LIF</td><td>Laser induced fluorescence</td></tr></tbody></table>

<table><tr><td>LIQ</td><td>Laser-induced quenching</td></tr><tr><td>Nd:YAG</td><td>Neodymium doped Yttrium aluminum garnet</td></tr><tr><td>Nd:YLF</td><td>Neodymium doped Yttrium lithium fluoride</td></tr><tr><td>OPO</td><td>Optical parametric oscillator</td></tr><tr><td>PEM</td><td>Photo-elastic modulator</td></tr><tr><td>PoPola</td><td>Poloidal polarimeter</td></tr><tr><td>QCL</td><td>Quantum cascade laser</td></tr><tr><td>RCP</td><td>Right circular polarization</td></tr><tr><td>SNR</td><td>Signal to noise ratio</td></tr><tr><td>SOL</td><td>Scrape off layer</td></tr><tr><td>TIP</td><td>Tangential interferometer polarimeter</td></tr><tr><td>TS</td><td>Thomson scattering</td></tr><tr><td>VDF</td><td>Velocity distribution function</td></tr></table>

# 7. Microwave diagnostics

M. Austin$^{34,4}$, S. B. Korsholm$^{35}$, Y. Liu$^{2,26}$, S. Danani$^{18}$, C. Muscatello$^{2}$, W. L. Rowan$^{34}$, V. Vershkov$^{24}$, G. Wang$^{33}$, J. L. Xie$^{36}$, M. Zerbini$^{10}$ and Y. L. Zhu$^{37}$

## 7.1. Introduction

Plasma diagnostics based on microwave radiation (inclusive of millimeter-wave and higher frequencies) have a long history going back over 70 years. Early diagnostic work is summarized by Heald and Wharton [557], with later summaries by Hartfuss [558] and Donné [4]. Looking forward, many aspects of microwave diagnostics will be favorable for next step fusion plasma devices. Both active (e.g. reflectometry, polarimetry) and passive (e.g. ECE) diagnostics are usually able to operate with launchers and receivers that are safely distanced from the edge of the reactor grade plasmas. And the radiation is generally able to be transmitted by low loss methods (mostly waveguide) that are not subject to radiation damage and degradation. Additionally, the associated sources and detectors of these systems are readily available at reasonable cost, and due to the transmission lines, can be stationed well away from the fusion device. Overall this bodes well for robust and reliable measurements from microwave diagnostics.

As described in the sections below, there have been advances in microwave systems which all lead to more accurate and reliable measurements of key plasma parameters. Refinements in theory, techniques, and hardware bring the diagnostics to full readiness for burning plasma conditions. Although most of the sections apply to ITER, the advances described will be relevant for fusion devices beyond that project, like a demonstration or pilot plant machine. Additional breakthroughs and new techniques will almost certainly be required for microwave diagnostics to meet the needs of the energy producing follow-ons to ITER.

Robustness will be an important quality for microwave systems as they will be subjected to the harshest conditions seen to date. If they can meet the prescribed reliability, the systems will be key to primary control systems that the machine will rely on to protect itself. However, certain challenges remain. In most cases (excluding the sources for CTS) the power levels for both active and passive diagnostics lie in the watts to nanowatts ranges. This is in the face of multiple megawatts
<table><caption>**Table 32.** ITER measurement parameters that microwave diagnostics contribute to.</caption><thead><tr><th></th><th>Primary</th><th>Backup</th><th>Supplementary</th></tr></thead><tbody><tr><td>Electron cyclotron emission</td><td>Core T<sub>e</sub><br>NTM δT/T<sub>e</sub></td><td>ELM temperature transient</td><td>β<sub>P</sub><br>P<sub>rad</sub><br>E<sub>max</sub> runaway<br>I Runaway<br>Edge T<sub>e</sub><br>TAE δn/n<sub>e</sub> & δT/T<sub>e</sub></td></tr><tr><td>Low field side (LFS) reflectometer</td><td>ELM density transient<br>Edge n<sub>e</sub><br>TAE δn/n<sub>e</sub> & δT/T<sub>e</sub></td><td>Core n<sub>e</sub></td><td>Profiles of electron density<br>V<sub>pol</sub></td></tr><tr><td>High field side (HFS) reflectometer</td><td>ELM density transient<br>Core n<sub>e</sub><br>TAE δn/n<sub>e</sub> & δT/T<sub>e</sub></td><td></td><td>Line-averaged electron density</td></tr><tr><td>Collective Thomson Scattering (CTS)</td><td>Alpha density profile<br>Alpha energy spectrum<br>p, D, T, <sup>3</sup>He Energy Spectrum</td><td></td><td>n<sub>T</sub>/n<sub>D</sub> in plasma core<br>Core T<sub>i</sub><br>Edge T<sub>i</sub><br>n<sub>H</sub>/n<sub>D</sub> Core</td></tr></tbody></table>

of microwaves expected to be employed in reactor devices for ECH (certainly ITER). The challenges for the instruments are twofold: (1) to not be blinded by the high power direct and scattered power, and (2) to not be damaged by the same. The latter is expected not to be too serious a challenge since the ECH will be in known, narrow frequency values so appropriate stop band, high- or low-pass filtering can be employed to prevent interference of the measurement or destruction of the instrument. In fact all diagnostic systems for ITER have some protection of this sort for the stray gyrotron power.

But there is another potential source of interference for microwave diagnostics that will have to be dealt with, and that is intense, essentially random bursts of strong microwave emission from the plasma itself. A type of this phenomenon, bursts at ECE frequencies, was first reported from one of the early tokamaks to operate at high auxiliary heating power, TFTR [559]. Subsequently bursting at ECE frequencies has been reported from many other tokamak plasmas, as seen in [560, 561] and the articles referenced within those. Although there has not been a comprehensive study so far, from these reports one sees that a common situation in most cases is low collisionality. That is, bursting emission from plasmas in the range of ECE occurs for plasmas heated to high electron temperature and relatively low density. This is what leads to the deduction that in ITER and other future fusion devices, bursting emission will be a common and frequent occurrence.

The upshot is that this emission will be an ever present annoyance, and sometimes threat, for both active and passive millimeter wave diagnostics in these future devices. With the frequency of emission bursts so random and uncertain, it will be difficult to devise avoidance and protection schemes. And with the expected high electron beta in the plasmas, it is conceivable that the intensity of these bursts will be enough to inflict damage; there is anecdotal evidence from both DIII-D and AUG tokomaks that bursts have damaged microwave

diagnostic instruments. A reasonable plan to deal with the bursts might be to have the capability, at least at the instrument end, to insert suitable filters once the nature of that devices bursts are understood and become predictable.

So moving forward, the microwave diagnosticians can expect some challenges in dealing with power coming from the reactor-grade plasma itself. The technology is available to deal with the issues, in terms of frequency-domain and time-domain filtering; it is a solvable problem. Therefore no doubt the advanced hardware and techniques described in the following sections will be able to make their full contributions to the fusion effort.

## 7.2. MPs for ITER

Microwave diagnostics contribute to a variety of ITER plasma MPs. Those MPs provide the plasma information such as Core/Edge $T_e/n_e/T_i$, ELM Temperature/Density Transient, MHD induced $T_e/n_e$ fluctuations, dynamics of alpha particles and other energetic particles, etc. A complete list is summarized in table 32.

The table 33 lists the requirements of the MPs for those where microwave diagnostics have a primary role. All of those MPs are either for advanced plasma control (AC) or for physics studies (PHY). Among those MPs, some of them can only be measured by a single system. For example, ECE is the only system that contributes to NTM $δT/T_e$, and CTS is the only system contributing to both alpha energy spectrum and the p, D, T, and $^3$He energy spectrum. Therefore, the requirements of those MPs shall be completely met by the sole system. For other MPs that several systems contribute to, the requirements of those MPs shall be met by the combination of those systems. In consequence, the requirements of the MPs that microwave diagnostics have a primary role are driving the system design.
<table><caption>**Table 33.** Measurement requirements of the measurement parameters for which microwave diagnostics have a primary role.</caption><thead><tr><th>Parameter</th><th>Condition</th><th>Range</th><th>Time Res.</th><th>Spatial Res.</th><th>Accuracy</th><th>Role</th></tr></thead><tbody><tr><td>ELM transient</td><td>r/a > 0.85</td><td>5 × 10<sup>18</sup>–3 × 10<sup>20</sup> m<sup>-3</sup></td><td>10 ms</td><td>5 mm</td><td>5%</td><td>PHY</td></tr><tr><td>Core T<sub>e</sub></td><td>r/a < 0.85</td><td>0.5–40 keV</td><td>10 ms</td><td>a/30</td><td>10%</td><td>AC</td></tr><tr><td>Core n<sub>e</sub></td><td>r/a < 0.85</td><td>3 × 10<sup>19</sup>–3 × 10<sup>20</sup> m<sup>-3</sup></td><td>10 ms</td><td>a/30</td><td>5%</td><td>AC</td></tr><tr><td>Edge n<sub>e</sub></td><td>r/a > 0.85</td><td>5 × 10<sup>18</sup>–3 × 10<sup>20</sup> m<sup>-3</sup></td><td>10 ms</td><td>5 mm</td><td>5%</td><td>AC</td></tr><tr><td>NTM δT/T<sub>e</sub></td><td>T<sub>e</sub> > 1 keV</td><td>1 × 10<sup>-2</sup>–5 × 10<sup>-2</sup></td><td>0.1–10 kHz</td><td>Island wid ≥ 20 mm</td><td>1%</td><td>AC</td></tr><tr><td>TAE δn/n<sub>e</sub> & δT/T<sub>e</sub></td><td>—</td><td>5 × 10<sup>-6</sup>–5 × 10<sup>-4</sup></td><td>30 kHz–0.5 MHz</td><td>n = 2–50</td><td>30% (rel)</td><td>PHY</td></tr><tr><td>Alpha density profile</td><td>—</td><td>1 × 10<sup>17</sup>–2 × 10<sup>18</sup> m<sup>-3</sup></td><td>100 ms</td><td>a/10</td><td>20%</td><td>PHY</td></tr><tr><td>Alpha energy spectrum</td><td>—</td><td>0.1–3.5 MeV</td><td>100 ms</td><td>a/10</td><td>20%</td><td>PHY</td></tr><tr><td>p, D, T, <sup>3</sup>He Energy Spectrum</td><td>—</td><td>0.1–1 MeV</td><td>100 ms</td><td>a/10</td><td>20%</td><td>PHY</td></tr></tbody></table>

### 7.3. ECE diagnostic for ITER

#### 7.3.1. Introduction.
In ITER, ECE in the range from 100 GHz to 1 THz will be used for measurement of electron temperature ($T_e$) profiles with high spatial and temporal resolution, for detection of high frequency instabilities (MHD, neoclassical tearing modes (NTM), Alfven eigenmodes, turbulence) and for measurement of the energy loss due to ECE. Both profile and NTM information are required by the Plant Control System in real time. A complicating factor is that recent results in DT plasmas [562, 563] have reinforced an earlier result [49] in which ECE did not agree with Thomson scattering, an alternate diagnostic for electron temperature profiles. The ECE diagnostic has the facility to detect the non-thermal electron distributions, which may cause this. Development of the diagnostic is a collaboration among the USDA, the INDA, and the IO.

The ECE diagnostic will measure electron temperature profiles for full-field and half-field ITER scenarios. Plasma access for ECE with X-mode polarization is illustrated for one full-field ITER scenario in figure 139. The figure represents an H-mode scenario with an axial toroidal field 5.3 T and with peak $T_e$ of 22.5 keV and peak electron density of $1.3 \times 10^{20}$ m$^{-3}$ (IMAS scenario 130 506, run 403, time 230 s, DT fuel, $I_p = 15$ MA). Figure 139 illustrates that the second harmonic X-mode EC emission can be used for measurements in the pedestal and confinement regions. The core of the plasma is accessible to ECE measurements using 1st harmonic O-mode.

The ECE diagnostic comprises three parts, front end, polarization splitter units (PSU) and transmission lines, and detectors. Each will be described in the next sections.

#### 7.3.2. The front end.
The front end (figure 140) collects the ECE from the plasma along two high-resolution views of the plasma, one view radial and the other oblique. One challenge is to collect sufficient emission with high spatial resolution to allow high temporal resolution. Since the captured plasma emission contains information both in its intensity and its polarization, a second challenge for the front-end design is to preserve both. The front-end system is in the middle DSM of equatorial port 9 (DSM EQ9, figure 140). The large distance from the diagnostic flange to the plasma center (>3 m) and the large plasma diameter (>6 m) require collection with minimum divergence angles. Quasi-optical (QO) [564] collection

<figure><img src="image_151.png" /><figcaption>**Figure 139.** Calculated X-Mode ECE access and limits. Solid lines indicate cold plasma ECE approximation. Yellow and purple dashed lines show location of the relativistic shift of emission peaks for second and third harmonics, respectively. Dot-dash lines indicate relativistic broadening of emission for second (yellow) and third (purple) harmonics.</figcaption></figure>

optics offer one means for achieving all of these requirements [565]. Compactness is another requirement. The design of the QO collection optics for ITER is similar to the compact Ellis-Austin periscope, which was developed for DIII-D [566] and achieves the required high spatial resolution with modest frequency dependence using just one ellipsoidal focusing mirror.

In figure 140, an ellipsoidal mirror for each view (M1 for the radial view) collects emission from the plasma. A flat mirror (M2) redirects the emission through the vacuum window (W) to the system aperture (A) which sets the spatial resolution in the plasma. The calibration source (HS) can be switched into the system for intensity calibration using a flat mirror. More on this later. The sizes of the optics are chosen to be a factor of three larger than the Gaussian beam width derived from the QO analysis. That assures transmission of more than 98% of the emission and much simplifies the secondary ECE requirement to measure the total ECE from the plasma.

The size of the collection volume of the ECE beam is tracked from the high field side (HFS) of the plasma through
<figure><img src="image_152.png" /><figcaption>**Figure 140.** Front end components and the PSU. Component labels are for the radial view and are defined in the text.</figcaption></figure>

<figure><img src="image_153.png" /><figcaption>**Figure 141.** The size of the Gaussian beam from the waveguides to the scenario axis of ITER along the optical path. The size is taken to be 3" (Gaussian width). The reference location is the system aperture.</figcaption></figure>

the entire QO system to the entrance to the waveguides in figure 141. The locations of the QO components are shown and their sizes are fixed by this QO calculation. The design results in component sizes that are consistent with the restricted space in the equatorial port yet large enough to reasonably meet the ITER measurement requirements. The size of the observed volume varies across the plasma (distance >4000 mm in figure 141), and as expected there is a frequency dependence. Still, the spatial resolution requirements for ITER are met. Since the initial design, the front-end optics have been modified repeatedly and successfully to follow the progress of port integration and that has demonstrated its robustness and flexibility.

Discrepancies between electron temperature derived from analysis of ECE and Thomson Scattering occur in some instances and appear to be due to departures from Maxwellian distributions [562, 563, 49, 564–569]. The ECE collected from an oblique view is sensitive to non-Maxwellian distributions [569] and has been used to evaluate the discrepancy. The resolution of this discrepancy was shown [49] to require analysis of higher harmonic ECE. This will require a wide bandwidth detection system such as the Fourier transform spectrometer (FTS). An oblique view is included in this diagnostic design (figure 140) which is quite similar to the radial view described

above. In addition, an FTS of the Martin-Pulpett interferometer type is included in the ITER ECE instrument array. The oblique view can provide a redundant measure of $T_e(R)$. The need for an oblique view then follows from the primary role to measure electron temperature with high reliability.

Mirrors are used to collect ECE. Since they are in-vacuum, indirectly exposed to plasma, and inaccessible for extended periods, degradation is a significant issue that requires in-situ mitigation. A calibration source is integrated into the design of both radial and oblique views. Each can be directly substituted for the plasma emission by the switching of a single mirror in each case and this provides a reference calibration for the ECE. The hot calibration source for the radial view is shown in figure 140 (HS). The primary component of the hot source is a SiC surface that is contoured [570] to produce a large diameter, uniform source of blackbody emission. When heated to its target temperature of 700 °C, its blackbody emission is easily distinguishable from the ambient emission of the other front-end components. The design is complex, as it must produce a uniform emission across a large surface for the many hours required for calibration. Design and successful prototyping are described in detail elsewhere [571–573].

### 7.3.3. PSU and transmission line.
The front-end optics are designed to preserve the polarization to a high degree, but the emission is conducted to the detection instruments, which are located more than 40 m from the tokamak, via purged waveguides. Waveguides will not preserve the polarization, so the polarizations for each view are separated in the PSU (figure 140), a QO unit. The polarized emission for each view is conducted to the detectors via a separate waveguide.

The transmission line is smooth-walled circular aluminum tube, which meets loss requirements of <15 dB (70–400 GHz) and <22 dB (400–1000 GHz). The waveguide will be purged with dry air to limit absorption features in the ECE spectrum.

### 7.3.4. Detection.
After all this, the ECE emission spectrum arrives at the instruments where it is spectrally analyzed, digitized, and the data sent to the archive. There will be two standard radiometers [574, 575]. One covers the range from 220 GHz to 340 GHz. While it pushes the limits for contemporary designs, a similar unit was deployed at C-Mod [576]. The limitations of that design were addressed in a prototype [577, 578]. At full field, it will cover the confinement region and pedestal in X-mode. A second, conventional radiometer will operate in the range of 122–230 GHz. At full field, it will be used for core measurements in O-mode. There will be two FTS spectrometers to cover the range of 70–1000 GHz with 3.75 GHz resolution. These will facilitate detection of non-thermal distributions and measurement of energy loss.

### 7.3.5. Prospects.
The design of all components are nearing completion. Prototyping of components is in progress. Port integration is actively in progress.
## 7.4. A low-loss microwave transmission line for ITER ECE

The ITER ECE diagnostic will be used to determine the plasma electron temperature by measuring the intensity of ECE in the frequency range 70 GHz–1 THz. One of the challenging requirements for the ITER ECE diagnostic TL is achieving low attenuation for the long distance (~43 m) between the front end and the instrument room over the broad frequency range. This is particularly challenging because of the low power (~few nW) thermal radiation emitted from the in-situ calibration sources located in the front end, which plays a vital role in determining the accuracy of measurements performed.

To meet the ITER ECE transmission line measurement requirement, an 8 m long prototype TL was developed at ITER-India. This TL is made up of oversized smooth walled straight waveguide sections (each of length 2 m and ID of 72 mm), miter bends and waveguide joints. The experimental set-up for attenuation measurement of this prototype line consists of a millimeter wave black body source, a FTS and the TL with required supports for proper alignment and coupling. The FTS is based on a Martin-Puplett interferometer (MPI) configuration [579] and covers the spectral range from 70–1000 GHz. The MPI operates under vacuum, has high throughput, excellent time resolution of 10 ms, and a scan displacement of 15 mm that provides a spectral resolution better than 10 GHz [580]. There are two input ports available on the FTS. One or both of them can be connected to the oversized waveguide (up to 99 mm). A 3 mm thick polypropylene window is provided for both of the FTS input ports to separate the FTS from all of the external interfaces and maintain a vacuum seal. A manually actuated shutter is provided on each of the input ports to enable blocking of the radiation from the input ports that are not in use. Eccosorb® is attached to these shutters which acts as a reference radiation source at room temperature. The input ports on the FTS are standard ISO 100F-type vacuum flanges. For the prototype TL measurements, port 1 of the FTS was kept at LN₂ temperature and port 2 was kept at high temperature ~500 °C, in order to increase the SNR. The recorded interferograms (30 000) were coherently added to average out the random noise. The attenuation was estimated by taking the ratio of the spectrum (derived by performing FFT of interferograms) measured for ~8 m TL length to the reference spectrum. The average and standard deviation of attenuation are derived for several data sets of attenuation measurement and are shown in figure 142. It is found that the error in TL attenuation is ~60% for frequencies below 250 GHz and <5% for frequencies above 250 GHz.

Considering the uncertainty associated due to various factors such as microwave source power, TL alignment, detector sensitivity and atmospheric effects (i.e. line and continuum absorption) a safety factor of two has been taken for deriving the ITER ECE transmission line loss for 43 m length as shown in figure 143. Since the TL is not under vacuum, water vapor absorption lines are observed in the spectrum at 380, 448, 552, 752, 916 and 988 GHz. It has been observed that, the overall transmission is improved in case of rough vacuum measurement as both the water line absorption and

<figure><img src="image_154.png" /><figcaption>**Figure 142.** Measured transmission attenuation for ~8 m length smooth waveguide TL.</figcaption></figure>

<figure><img src="image_155.png" /><figcaption>**Figure 143.** Estimated loss for ITER ECE TL for smooth-walled waveguide (~43 m length and 6 Miter Bends), based on measured attenuation of prototype system.</figcaption></figure>

the continuum absorption are lower in an evacuated TL [581]. It is also inferred from comparison between air and vacuum measurements that there is some level of continuum absorption of the radiation in this frequency range by atmospheric gases. The broadband transmission attenuation is decreased by ~2 dB (average value) above 500 GHz.

From the above measurement results, it is determined that a smooth walled TL gives a maximum loss of 13 dB in the 500–800 GHz range, and even better performance in the 100–350 GHz which is the range for primary measurement of the ECE Diagnostic. This meets the ITER ECE transmission line attenuation requirements of <15 dB attenuation in frequency range 70–400 GHz and <22 dB attenuation in 400–1000 GHz.

The results have been assessed to understand the cause of large error in TL attenuation for frequencies below 250 GHz. This can be attributed to the low sensitivity of the FTS detector, 1/f noise and low source power. FTS is a wide band system with coverage up to 1800 GHz (upper limit) but with weaker sensitivity in the lower frequency range. A reduced error could be achieved with longer integration times, e.g. ~8 h. However, if the detector element is replaced by one with enhanced low frequency sensitivity, then this problem can be resolved. It is therefore concluded that in case of high signal level from the plasma and having an FTS with optimized detector, will resolve the concern of large error in attenuation
in the low frequency range of the spectrum even in case of low signal (as in the case of the calibration source). In order to further improve the prototype TL attenuation results, another hot source, having good temperature uniformity and temperature stability, is also being considered, which will improve the S/N ratio. Also, there is scope of improvement in the FTS data analysis techniques which are also being looked into. Further experiments are being planned in near future with improved data analysis and hot-cold techniques.

### 7.5. Correlation ECE measurement of electron temperature turbulence

Multi-field turbulent fluctuations, e.g. turbulence in temperature, density, magnetic field, radial electric field, etc have been considered possible mechanisms to explain the observed anomalous radial transport and therefore define the particle and energy confinement in magnetic confinement fusion devices. The correlation ECE (CECE) radiometer has been developed as an effective electron temperature turbulence diagnostic for core to edge pedestal measurement in the last two to three decades on tokamaks and stellarators (referring to a 2007 review paper by Watts [582] and the references therein, and [583–591] for CECE systems on existing machines). The CECE measurement of electron temperature turbulence has contributed to various physics studies, e.g. in earlier days, core heat transport in TEXT-U [592], sawtooth physics in TEXT-U [593], core transport in W7-AS stellarator [594], hot ion instabilities in Tore Supra [595], and others (see review article [582]). Recently, CECE has been utilized to study core transport in I-mode [596], negative triangularity shape plasma operations [597, 598], edge pedestal transport in RMP ELM-controlled H-mode [599] and quiescent H-mode [600], NTM physics [601], and validation of gyrokinetic simulations [602], etc.

The ECE signals are comprised of blackbody thermal radiation with inherent fluctuations [603] generally larger than or at the same level of electron temperature turbulence. The CECE diagnostic utilizes two closely spaced (generally radially spaced) ECE channels which contain uncorrelated thermal noise as well as correlated electron temperature turbulence; small turbulent electron temperature fluctuations are extracted by performing correlation analysis between the two ECE channels [582]. In ITER, ECE has a range of roles in the delivery of the core electron temperature profile and detection of NTMs, the measurement of the H-mode pedestal, and plasma energy, radiated power, runaway electrons, edge electron temperature profile and high frequency instabilities [604]. The CECE technique for electron temperature turbulence measurement can be applicable in ITER and contribute to transport and MHD physics studies in various operational modes and scenarios in ITER.

Recently, a new, improved analysis of correlation ECE data to accurately determine turbulent electron temperature spectra and magnitudes has been proposed [605], which has been verified both numerically through the simulation of synthetic ECE radiometer data, as well as through analysis of experimental data from the CECE system on DIII-D. The new

formulation places coherency-based analysis on a firm foundation and corrects some currently published methodologies. This new method accurately accounts for bias error in the coherence function and correctly calculates noise levels for a fixed data record length. It provides excellent accuracy in determining temperature fluctuation level (e.g. <10% error) even for a small realization number in the ensemble average. The method also has a smaller uncertainty (i.e. error bar) in the power spectrum when compared to the more standard cross-power method when evaluated at low coherency. Direct calculation of system noise level using correlation between randomized IF signals is recommended.

The new form of the normalized electron temperature fluctuation power spectrum $P_{\tilde{T}}(f)$ is derived [605] as:

$$P_{\tilde{T}}(f) = \frac{1}{B_{IF}} \frac{\gamma_c(f)}{1 - \gamma_c(f)}, \qquad (7.1)$$

Where $B_{IF}$ is the intermediate frequency (IF) bandwidth and $\gamma_c(f) = \sqrt{\gamma^2(f) - \gamma_b^2(f)}$, where $\gamma^2(f)$ is the coherence spectrum between the two ECE signals, $\gamma_b^2(f) = (1 - \gamma^2(f))^2/M$ is the bias error in calculating $\gamma^2(f)$, $M$ is the realization number for calculating $\gamma^2(f)$. This new formulation has several advantages. The first being that an average or DC measurement of the ECE signals are not required for the evaluation of $P_{\tilde{T}}(f)$ since the calculation of the coherency spectrum $\gamma(f)$ is not sensitive to DC components. Secondly, the standard cross-power spectrum for evaluating $P_{\tilde{T}}(f)$ tends to have a bias at low realizations and/or low coherency cases. In addition, the statistical error is smaller than the cross-power spectrum with the same ensemble average calculations at low $\gamma(f)$ [605].

In CECE literature the minimum measurable electron temperature fluctuation level $(\tilde{T}_{\text{RMS}}/\tilde{T})_{\text{min}}$ is often predicted from the following radiometer noise formula [582]:

$$\left(\frac{\tilde{T}_{\text{RMS}}}{\tilde{T}}\right)_{\text{min}} = \sqrt{\frac{1}{N^{1/2}} \frac{2B_{\text{vid}}}{B_{\text{IF}}}}, \qquad (7.2)$$

where $N = 2B_{\text{vid}}dt$ is the number of data points, $B_{\text{vid}}$ is the video bandwidth, and $dt$ is the data record length. In fact, formula (7.2) can be found [605] to be the square root of the standard error of measured $(\tilde{T}_{\text{RMS}}/\tilde{T})^2$ using correlation analysis. The simulations and DIII-D data show that the actual system noise level is underestimated by equation (7.2) (by tens of percent). A direct calculation of the minimum measurable electron temperature fluctuation level $(\tilde{T}_{\text{RMS}}/\tilde{T})_{\text{min}}$ using randomized IF signals and equation (7.1) is recommended.

For ITER there would be significant challenges to getting useful data from a high resolution CECE instrument but some form of it could likely be implemented, i.e. a system focusing on the low-field-side edge. This could be part of the ECE or reflectometer systems; for the LFS pedestal region the plasma would be in the near field of both system's optics and hence have the best antenna spatial resolution. Additionally the pedestal $T_e$ values are expected to be such that relativistic broadening would be small, again allowing for sufficiently small ECE volumes for CECE to have decent resolution. These
measurements could contribute to understanding how turbu-
lent and mode fluctuations affect the pedestal, its height and
the region’s transport properties.

7.6. CTS for ITER

The primary role of the ITER fast ion CTS diagnostic is to
provide information on the VDF of the confined fast ions—in
particular the fusion born alpha particles. The CTS diagnostic
was included in the ITER Baseline in 2007 as an enabled dia-
gnostic. This means that primarily the in-vessel parts of the
diagnostic will be developed and installed for first plasma. The
in-vessel front-end part of the diagnostic has passed its FDR
in October 2021 [606].

The fast ion CTS diagnostic is a relatively young diagnostic
first being proposed for JET in the 1990s [607, 608]. Below
we will briefly outline the development maturing the fast ion
CTS diagnostic to becoming an ITER diagnostic. Following
this the present design and status of the ITER CTS diagnostic
is presented.

In brief, the basic principle of fast ion CTS is that a power-
ful source beam—typically a gyrotron microwave beam in
the frequency range near the fundamental or second har-
monic at a power of several 100’s kW—is injected via
quasi-optical mirrors. The probe beam interacts with micro-
scopic fluctuations—primarily electron density fluctuations—
and scattered radiation is emitted [609]. A quasi-optical
receiver transmission line picks up a portion of the scattered
signal that contains information about the dynamics in the
volume constituted by the overlap of the probe beam and the
receiver beam. By choosing the scattering geometry and the
probe frequency appropriately, one can ensure that the spec-
tral information in the received signal primarily stems from
collective motion of the electrons in turn originating from the
dynamics of the ions. Hence, a properly designed fast ion
CTS diagnostic can provide information on the VDF of the
confined fast ions both spatially and temporally resolved. For
more details, see [609, 610].

7.6.1. *Developments of the CTS diagnostic.* The fast ion CTS diagnostic was proposed, implemented, and tested on JET with moderate success [607, 608]. The primary challenge for the JET CTS diagnostic system was the choice of frequency in the ‘window’ between the fundamental and second harmonic EC resonance. With the high temperatures of JET, the background (ECE) noise hampered the SNR.

Following the JET CTS experience, an important step
towards the maturation of the diagnostic was made in the
implementation at TEXTOR [611] and later ASDEX Upgrade
[612]. In parallel to this development feasibility studies for
a CTS diagnostic for ITER were performed at Risø National
Laboratory [613–624]. These early works laid the foundation
for the inclusion of CTS in the ITER Baseline, and the system
proposed in the feasibility study is conceptually alike the cur-
rent ITER CTS design—being described in the next section.

While the ITER CTS diagnostic has been chosen to focus
on fast ion physics, the versatility of the CTS diagnostic has

been demonstrated at TEXTOR and ASDEX Upgrade [610, 625]. For specific scattering geometries, CTS can provide information on the fuel ion ratio in the core of a burning plasma, and geometries and equipment can be optimized to provide bulk ion information such as temperature and rotation velocity [616–618].

Furthermore, it was demonstrated on ASDEX Upgrade that bulk ion features in the CTS signal are so powerful that measurements of e.g. bulk ion temperature, $T_i$, and bulk ion rotation, $v_{rot}$, can be made even while EC resonances are in the plasma [619]. This means that bulk ion CTS can be made in piggy-back by simply detecting the scattered signal due to the heating gyrotrons. This opens up new possibilities for more flexible CTS diagnostic systems on machines with ECRH.

While the CTS diagnostic is still a relatively young dia-
gnostic, it is being implemented on several machines for ion
measurements e.g. on LHD [620, 621] and W7-X [622].

7.6.2. *Status of development of the ITER CTS diagnostic.*
The ITER CTS diagnostic will operate using a dedicated gyro-
tron operating at 60 GHz delivering at minimum 1 MW with
50% modulation duty cycle for ECE background subtraction
[606]. The scattered signal is picked up by an array of seven
receiver beams intersecting the probe beam in seven con-
fined scattering volumes across the plasma diameter (see
figure 144). The spatial resolution of the diagnostic varies
across the diameter with a radial extent of the volume of 20 cm
near the edge and of 50–80 cm in the core of ITER. The tem-
poral resolution is 100 ms fulfilling the ITER measurement

<figure><img src="image_156.png" /><figcaption>**Figure 144.** Geometry of the ITER CTS diagnostic. Probe beam in blue. Shown for the ITER steady-state scenario, $I_p = 10$ MA, $B_t = 5.3$ T.</figcaption></figure>
<figure><img src="image_157.png" /><figcaption>**Figure 145.** Drawings of the ITER CTS diagnostic in EPP12 drawer #3. The (pink) beam to the upper left is the gyrotron probe beam that intersects with the seven (blue) receiver beams in defined locations across the plasma diameter.</figcaption></figure>

requirements for the parameters that CTS has as primary function:

* alpha density profile

* alpha energy spectrum

* p, D, T, ³He energy spectrum.

The probing frequency of 60 GHz has been chosen in order to reduce the ECE background noise, which is very low in ITER at frequencies below 65 GHz [623]. This also means that the diagnostic must be operated in X-mode. However, this leads to another challenge, that the probing beam passes the fundamental EC resonance inside the portplug—an issue we return to shortly. At the same time, the frequency range is very close to an X-mode cut-off, leading to significant refraction of both the probe and receiver beams. Nevertheless, extensive modeling and analysis of the diagnostic performance has been performed and the proposed ITER CTS diagnostic will be able to measure the fusion born alphas within the desired accuracy [624, 626–630].

The in-vessel part of the diagnostic is located in an exclusive drawer (#3) of EPP #12. All in-vessel components have been developed and designed under contracts with Fusion for

Energy (see figure 145). The in-vessel part of the fast ion CTS diagnostic for ITER passed the PDR in March 2020 and the FDR in October 2021.

Besides the more generic challenges of ITER diagnostic systems, such as radiation and neutronic damage, the CTS diagnostic has some specific challenges. The primary challenge is the fundamental EC resonance for 60 GHz being located in the middle of the EPP #12. The risk is that an EC-assisted breakdown of residual neutral gas could happen, since the neutral pressure is known to increase during plasma discharges—as seen in JET. This would lead to plasma formation in the CTS in-vessel transmission line, which in turn would lead to absorption of fractions of the 1 MW gyrotron beam inside the port plug transmission line components. The risk mitigating design feature is the use of a longitudinally split electrically-biased waveguide (SBWG). The functionality of such a SBWG has been demonstrated in prior implementations on e.g. DITE [631, 632]. The principal idea is that any free electrons inside the SBWG are swept to the positively biased half so fast that they cannot start an ionization cascade, which would otherwise lead to plasma breakdown inside the waveguide. The design and extensive modeling results are described in [631, 632].
The present paper cannot describe all aspects of the design and analysis, hence the reader is referred to the following literature describing the development towards a complete design of the ITER CTS diagnostic [610, 631–639].

## 7.7. LFS reflectometer for ITER

The Low-field side reflectometer (LFSR) for ITER will supply simultaneously three important plasma measurements: (1) electron density profile, (2) electron density fluctuations, and (3) poloidal plasma rotation. In addition, an indicator of edge transients (e.g. L-H transition, ELMs) will be derived from the time evolution of the electron density profile. The ITER PCS will use real-time measurements of the density profile and poloidal rotation from LFSR for advanced feedback control. In addition to feedback control input to ITER, LFSR data will be useful for plasma stability and turbulence studies. It is capable of detecting low-to-intermediate wavenumber fluctuations, enabling an extensive range of physics studies. The PPR that was originally foreseen in the 2007 ITER Baseline [4] has been abandoned on ITER because integration became prohibitively too difficult and the so-called captive components would not be delivered on time

The broad measurement capability of LFSR is enabled by an array of six monostatic antennas which inject from an equatorial port on the outboard side of the ITER vessel. Low-loss transmission lines, consisting of corrugated, overmoded waveguide and miter bends, transmit the 30–165 GHz, O- and X-mode signals to and from the ITER plasma. Integrated transmission-line components serve a range of purposes, such as protection from high-power stray RF radiation, accommodation of transmission-line displacement, and simultaneous measurement of reference and plasma phases throughout the discharge. Broadband transmission signals are realized by full-band microwave transceivers combined with quasi-optical multiplexing. Both frequency-modulated continuous wave (FMCW) operation for profile and fluctuation measurements and fixed-frequency continuous wave operation for poloidal rotation (Doppler) measurements are incorporated. Considering the fast-paced evolution of microwave and analog-to-digital conversion technologies, collection of FMCW data with bandwidths well above 1 GHz is planned for LFSR. With this capability, LFSR will perform full-band, ultrafast sweeps (<1 µs) with unprecedented time resolution. Density and plasma rotation profiles will be processed in real-time with <10 ms time resolution by FPGA modules. Detailed descriptions of the end-to-end design of LFSR can be found in previously published works [640, 641].

The design of LFSR has been optimized to probe the edge region ($r/a > 0.85$) for a broad range of ITER conditions. However, penetration into the core is possible under certain scenarios. Figure 146(a) shows full-field ITER baseline profiles for electron density ($n_e$), electron temperature ($T_e$), and magnetic field (B). O-mode and X-mode cutoffs and EC resonances at the midplane are plotted in figure 146(b); relativistic modification to the electron mass is included. Contour plots of cutoff frequency versus B, $T_e$, and $n_e$ are shown in

figures 146(c) and (d) for upper X-mode and O-mode polarizations. The $T_e$ profile is mapped to the B profile at the midplane according to the profiles shown in figure 146(a). They are plotted along the vertical axes of figures 146(c) and (d). The horizontal dashed line indicates the magnetic field near the edge ($R = 8.40$ m). The vertical dashed line indicates the Greenwald density with plasma current of 15 MA and minor radius at the midplane of 2 m. With upper X-mode, the outboard $n_e$ profile can be probed from the far SOL to the top of the pedestal, where $n_e$ can be as high as $1 \times 10^{20} \text{ m}^{-3}$. For lower $n_e$ at the top of the pedestal, access into the core is possible, limited by the upper-bound probe frequency (165 GHz contour in figure 146(c)). The $n_e$ profile in ITER is expected to be flat in the core, so O-mode measurements provide limited coverage. However, the measurement is still useful and provides a partially redundant and B-independent dataset to compare and verify with the upper X-mode measurements.

Various performance metrics of LFSR have been studied at a dedicated test facility consisting of a full-scale transmission line, representative of the current design. Two of the most important metrics to reflectometry are antenna efficiency and power budget. Previous works have investigated these performance metrics in detail [641, 642], and the results are summarized here. The front-end of LFSR includes the antenna assembly, ECH protection mirror, phase calibration mirror, and dual-disc vacuum window. The microwave signal traverses each of these components twice (one on transmit and second on receive). Mode conversion caused by these components can ultimately lead to degraded signal-to-noise. Representative antenna patterns measured for 55, 120, and 160 GHz are shown in figure 147. The profiles are peaked near the axis and exhibit a side-lobe power level which is generally less than –20 dB of the main lobe. The effect of the front-end components on the waveguide mode content is quantified with a phase retrieval algorithm using antenna pattern measurements collected at several distances from the antenna. The $HE_{11}$ mode is the ideal propagating mode and couples to a free-space Gaussian beam with 98% efficiency [643]. The results of the mode analysis give the percentage of the total beam power contained in the $HE_{11}$ mode and higher-order modes as a function of frequency. The $HE_{11}$ mode content for 55, 120, and 160 GHz is 94.4%, 88.1%, and 87.5% respectively, i.e. it is highest at low frequency. In other words, loss of about 1 dB due to mode conversion can be expected at the highest frequencies. Inspection of the radiated profiles and the results of the mode analysis indicate the front-end components of LFSR impose only a small and acceptable fraction of mode conversion.

The metric used to assess the upper bound on the spatial resolution of LFSR density profile measurements is the range precision. The range precision depends explicitly on the SNR and the RF bandwidth [644], providing a means to assess the effect of output power, loss, and noise on the performance. Selecting a reasonable bandwidth of $\delta f = 0.25$ GHz, a SNR of 26 dB or larger is required to obtain a spatial resolution of 5 mm or better. Four main factors determine the ultimate SNR of the system: (1) power output of the transceivers, (2) loss due
<figure><img src="image_158.png" /><figcaption>**Figure 146.** (a) Full-field ITER baseline profiles of $n_e$, $T_e$, and B. (b) Corresponding cut-off and cyclotron frequency profiles with relativistic correction. Contour plots of constant frequency (in GHz) versus $n_e$, B, and $T_e$ for (c) upper X-mode and (d) O-mode polarizations. The horizontal dashed line indicates the field at the plasma edge (R = 8.40 m). The Greenwald density is plotted as the vertical dashed line.</figcaption></figure>

to TL components, (3) plasma-induced losses, and (4) plasma-induced noise. Factor 1 is measured directly at the LFSR test facility. Factor 2 is also measured directly and compared to theoretical estimates of the loss. Factor 3 is estimated with full-wave reflectometer simulations for a range of density fluctuation parameters. Factor 4 is calculated assuming ECE is the main source of noise. A plot of the SNR is shown in figure 148, which does not include loss from turbulence and assumes the plasma midplane is vertically aligned to the antenna.

The horizontal dashed line indicates the minimum SNR of 26 dB to achieve 5 mm spatial resolution. Except for a narrow region around 60 GHz, SNR > 26 dB over the full frequency range of LFSR. The drop at 60 GHz is due to a 60 dB notch filter to protect the LFSR microwave electronics from scattered signal from the CTS diagnostic. In general, turbulence increases loss, and the magnitude of additional loss depends strongly on the spectral and spatial characteristics of the turbulence. Over the range of fluctuation parameters run in the full-wave simulations, turbulence-induced losses up

to 10 dB resulted. Even with this additional loss, LFSR still achieves acceptable SNR except near the edges of the discrete microwave bands (30, 50, 75, 92, 110, 137, 165 GHz) that comprise the full range.

The ITER LFSR will provide important measurements of electron density profile, density fluctuations, and plasma rotation. The design largely represents a traditional reflectometer in terms of basic functionality, but its combination of low-loss waveguide components, broadband-enabling technologies, and ultrafast, real-time process capabilities will permit measurements with high robustness and accuracy. Data from a full-scale test facility and modeling results suggest that high signal-to-noise is achievable with the current design, enabling sub-centimeter spatial resolution of the edge density profile.

## 7.8. HFS reflectometry (HFSR) for ITER

HFSR is based on the analysis of the amplitude and phase of the reflected mm-waves probing plasma from the inner
<figure><img src="image_159.png" /><figcaption>**Figure 147.** Radiated beam profiles for 55 GHz, 120 GHz, and 160 GHz.</figcaption></figure>

<figure><img src="image_160.png" /><figcaption>**Figure 148.** Signal-to-noise ratio considering transmission-line loss, plasma-antenna coupling loss, and ECE noise. The horizontal dashed line indicates the minimum SNR of 26 dB required for 5 mm spatial resolution.</figcaption></figure>

circumference of the plasma torus. As the cutoff position of the reflection surface depends on the probing frequency, it is possible to investigate the radial density profile and its fluctuations by a frequency scan. The probing of the plasma can be made with the ordinary (O) and two extraordinary (X-) waves. The cut-off positions of these waves are shown in figure 149. The cut-off position of the O-wave depends only on the density. So it cannot access to the plasma core in the case of a flat density profile, expected in ITER [645], and can be used mainly at the edge. The cut-off position of the High- and Low-Frequency extraordinary waves (XU and XL) depends both on density and magnetic field, which gives the possibility to access the plasma core. But as seen from figure 149, in the case of HFS launch only the XL mode can access the plasma core. The great advantage of this wave is the utilization of lower frequencies and a small influence of the relativistic high temperature effect. Taking into account the operation density range of ITER discharges, frequency bands for O-mode of 18–140 GHz and for XL-mode of 12–90 GHz were chosen. The required wide frequency range, exceeding one decade, caused significant technical challenges in transmission of mm-waves along

<figure><img src="image_161.png" /><figcaption>**Figure 149.** Radial dependencies of different frequencies in ITER for scenario 2. Dashed lines correspond to the cold plasma case ($T_e = 0$), solid ones to the real plasma temperature. $f_{xl}$—lower cut-off for X-mode; $f_o$—O-mode cut-off; $f_{ce}$—electron cyclotron frequency; $f_{xu}$—upper cut-off for X-mode; $2f_{ce}$—second harmonic of the electron cyclotron frequency.</figcaption></figure>

the curved trajectory from external source to the ITER HFS and locating efficient antennas in the slit between the blanket modules. The first problem was solved by the use of rectangular waveguides (inner cross-section 20 × 12 mm) and special forms of the bends (hyperbolic secant) [646]. In order to decrease the induced eddy currents during plasma disruptions and to withstand the thermal stresses under neutron heating the Inconel Alloy 718 was chosen as the waveguide material inside the VV. The inner surface of the waveguides is coated with a thin Cu layer. The second problem was solved by the use of special antennas with horn for launch and mirror antenna for the receiver.

Recently the HFSR system was enhanced with the addition of a refractometry channel. The concept of refractometry was proposed in [647]. It analyses the phase of the wave which is launched from the low field side (LFS) and received at the HFS. Refractometry uses extraordinary wave in the transparency window between the cyclotron frequency and the lower extraordinary cut off.

The HFSR reflectometry layout in section #8th is shown in figure 150. One can see the HFSR horn-mirror antenna (1) which is connected through a 90 degree bend and the transmission line along the vacuum chamber with the upper port. The vacuum part of the transmission system is terminated by the double vacuum window, located at the vacuum port extension tube. The refractometry antenna is located only in equatorial port #8 in front of the HFSR antenna, as shown in figure 150. The HFSR is comprised of four systems. Two are located in the upper port of section #8 and the other two systems are located in sections 9 and 17. The top view of HFSR locations is shown in figure 151.

Each atmospheric transmission line begins from the first vacuum window and ends in the Diagnostic Area, where the mm-wave equipment and instrument and control systems are located.

The two systems in sector 4 are used for density profile measurements in O and XL mode. The systems in sectors 9
<figure><img src="image_162.png" /><figcaption>**Figure 150.** HFSR in-vessel layout in upper and equatorial ports of section #8. 1—HFS reflectometer antenna system; 2—Transmitting refractometry antenna.</figcaption></figure>

<figure><img src="image_163.png" /><figcaption>**Figure 151.** A top view indicating the general locations of the HFSR diagnostic. (1—Vacuum Vessel, 2—Upper port, 3—Interspace, 4—Port Cell, 5—Gallery, 6—Diagnostic Area in Buildings 74/13).</figcaption></figure>

and 17 are used for the measurement of density fluctuations in XL and O modes, respectively.

As HFSR uses a very wide frequency range, quasioptical couplers are used to combine/decombine the signals in five frequency bands in single waveguides [648].

The diagnostic back-end utilizes voltage controlled oscillators (VCOs) as the main mm-wave source in the frequency band 8–20 GHz. The VCO is controlled by a highly-integrated control unit that also contains a 125 MHz digital-to-analogue converter to form the required tuning voltage shape, 20 V DC amplifier, several sine generators as IF sources and IQ detectors. A frequency-calibration system that is based on RF coax-cable delay line is used to pre-pulse measurements of

VCO tuning characteristic and tuning voltage shaping. RF coax-cable delay lines reduce the difference of signal path in RF and LO legs. This approach has been successfully tested in the T-10 tokamak [649]. 1 GS s⁻¹ ADCs will be used in the corresponding acquisition system and 10 MS s⁻¹ ADCs in subsystems for fluctuation and live-averaged density measurements.

The planned mm-wave source frequencies lead to an electron density profile measurement range 3 · 10¹⁹ m⁻³–2.65 · 10²⁰ m⁻³ and the line density measurement range 8 · 10¹⁸ m⁻²–8 · 10²⁰ m⁻². The density profile will be measured using a FMCW technique [650]. The diagnostic is planned to provide the cutoff layer position with an absolute error of about 1 cm and a relative error of about 5%. The relative density error at the measurement point will be about 10%. It is planned to have about 20 density points per frequency band that roughly corresponds to 30 density points in the plasma core in the baseline ITER scenario. A 5 µs frequency sweep with 10 kHz repetition rate is planned now as the main measurement scenario. This allows to have spectrogram averaging to reduce the influence of turbulence keeping the time resolution about 1 ms. It should be noted that the repetition rate is determined by the existing I&C hardware solution whereas the rest of back-end equipment allows to operate with 100 kHz repetition rate.

Since HFSR cannot provide the line-averaged density via standard reflectometer measurements, this parameter will be estimated by a see-through channel in Sector 4. Due to uncertainties in the determination of the line-averaged density, the diagnostic will provide the integrated electron density along the viewline in the range 8.0 × 10¹⁸–8.0 × 10²⁰ m⁻², that roughly corresponds to 2.0 × 10¹⁸–2.0 × 10²⁰ m⁻³ averaged density. The measurement accuracy is determined by hardware limitations that lead to an absolute error of 8.0 · 10¹⁸ m⁻² and reconstruction error of about 2%. It should be emphasized that in real-time operation one can expect an increase of the error to 10% because of strict latency limitations.

Small-scale turbulence measurements can provide information on broad-band turbulence and quasi-monochromatic modes (TAE, RSAE, MHD etc) in the frequency band 30 kHz–2 MHz. Since the HFSR utilizes XL mode probing, the reflected signal is sensitive to fluctuations of both electron density and magnetic field at the cutoff layer.

## 7.9. System-on-Chip (SoC) technology for millimeter-wave diagnostics

There are serious issues raised by the harsh environment of ITER which is the first magnetic confinement fusion device to be able to produce net fusion power and will generate radiation levels, i.e. neutrons and gamma rays, that are orders of magnitude higher than present-day experimental machines as well as stray microwave and millimeter wave radiation and microwave bursting. There is thus a strong need to protect diagnostics electronics. This is partially accomplished by the use of chicanes, filters, limiters, and siting in the diagnostic hall beyond a shield wall. However, this does not suffice and
<figure><img src="image_164.png" /><figcaption>**Figure 152.** W-band and F-band receiver chip development for deep plasma temperature measurement on DIII-D. The pre-LNA, balanced mixer, IF-amplifier, and internal LO multiplier are merged on a single chip.</figcaption></figure>

therefore motivates the development of electronics with higher performance and capability, as well as robustness to survive the hostile burning plasma environment.

Monolithic, 'system-on-chip' (SoC) millimeter wave integ-
rated circuit technology, provides dramatic advances for
millimeter-wave diagnostics in fusion reactors, including over
30 dB signal gain and 85% electronics noise suppression
[651, 652]. The transmitter and receiver chips have been
developed with extremely high-level integration size (square
millimeter level for each piece). The prototype approaches
have been fabricated and employed as individual modules
on DIII-D diagnostics, including V-band (55-75 GHz), E-
band (72-80 GHz), W-band (75-110 GHz), and F-band (110-
140 GHz). W-band and F-band receiver chips for plasma
temperature measurement have been developed and success-
fully applied on DIII-D. The pre-LNA, balanced mixer, IF-
amplifier, and internal LO frequency multiplier/driver are
merged onto a single chip (see figure 152), which can
be packaged as a general module for radiometer, reflec-
tometer, interferometer, polarimeter, and scattering system
applications. The W-band image-rejection SoC receivers have
been successfully designed and fabricated using both Taiwan
Semiconductor Manufacturing Company (TSMC) 90- and
40 nm CMOS as well as WIN Semiconductors 0.1 µm GaAs
processes.

For SoC technology, the chip material should have the cap-
ability to tolerate the strong radiation, which dictates the use
of the gallium nitride (GaN) wide bandgap semiconductor
material (3.39 eV versus 1.43 and 1.11 for GaAs and Si,
respectively). GaN has been selected due to its high power,
high breakdown voltages, and low noise beyond 200 GHz
capabilities [653] here with the HRL T3 40 nm process provid-
ing *f*<sub>T</sub>*f*<sub>max</sub> of 200 GHz/400 GHz and the T4A 20 nm process
providing *f*<sub>T</sub>*f*<sub>max</sub> of 329 GHz/558 GHz [654].

A key focus is on the characterization of the millimeter
wave performance under radiation and neutron active envir-
onments of both individual GaN devices and SoCs. The SoC
approach of microwave transmitter and receiver modules have
wide application potential in the fusion plasma diagnostics
field. For example, the full frequency bandwidth in W/F bands
or higher provides wide radial observation windows for ECE

Radiometer and Density Reflectometer diagnostics for fundamental parameter measurements in burning plasma.

7.10. Polarimeter-interferometer system based on terahertz solid state source (SSS)

With the rapid progress of Terahertz technology in recent years, terahertz SSSs with modest power output have been applied in interferometer system on several machines [655–658] to replace the traditional FIR lasers. Compared with FIR lasers, the solid-state sources have many advantages, such as compactness, stable frequency, turn-key operation and easy maintenance. The major shortage, which is the modest power output (around mw), can be amended by the significant improvement of detection sensitivity of planar diode mixers.

The multichannel heterodyne interferometer on KTX util-
izes two beams based on the principle of the Michelson inter-
ferometer. One beam serves as reference signal with fixed fre-
quency about 650 GHz, and the other serves as probing signal
with tunable frequency (630–660 GHz). The SSS with max-
imum output power approximately 2 mW is composed of an
amplifier and multiplier chain with a local oscillator (fixed at
13.45 GHz for the reference beam and a finely tunable syn-
thesizer range from 12.9 to 13.95 GHz for probing beam).
The planar diode mixers are optimized for high sensitivity
~750 V W⁻¹ and then followed by pre-amplifiers with gain
factor 500. To fully use the beam power, the whole system
was set on a light weight and compact platform (~3 m × 3 m)
and installed just beside the VV ring [659]. The optical
design has been optimized for uniform distribution of beam
light and beam width minimization. The interferometer sys-
tem with five-chord signals has recently been successfully
installed on the machine [660]. The beam width across the
plasma is about 20 mm and the minimum phase noise can
reach 0.004π with 1 MHz bandwidth, which corresponds to
about 0.1% density fluctuation amplitude. The density pro-
file has been reconstructed by fitting the line-integrated data
into an equilibrium program. The future work is to develop a
polarimeter-interferometer for simultaneous measurements of
Faraday rotation angle and electron density by upgrading the
existing multi-channel interferometer system based on solid-
state sources.

7.11. The potential of THz-TDS diagnostics for next step
fusion experiments

7.11.1. *THz time domain spectroscopy (TDS).* The Terahertz (THz) band of the electromagnetic spectrum is nowadays defined as the frequency range [0.1–10]THz (1 THz = $10^{12}$ Hz), between microwaves and mid-IR electromagnetic radiation, where electronic and optical technologies overlap. The last decade has seen an enormous development of spectroscopy techniques in this area, with the full development of the so-called TDS. The starting point of this approach lies in the fact that femtosecond laser pulses (usually ~800 nm wavelength) can be used to generate extremely broadband (100 GHz–30 THz) single-cycle THz pulses with less than
<figure><img src="image_165.png" /><figcaption>**Figure 153.** Schematic of a fiber optics THz-TDS setup for plasma diagnostics applications. In addition to the hyper-hemispherical lenses shown here, Gaussian collimation optics will be required to match emitter and receiver across the long beam path in tokamaks.</figcaption></figure>

1 ps duration, through the well known process of photoconductive generation of Terahertz radiation (spark-gap emitter) [661]. Such transients contain a truncated cycle of the electric field and therefore, according to the Fourier Theorem, have a frequency spectrum extending over a broad range. With the appropriate choice of emitter, the THz, the mid-IR and even the FIR radiation regimes can be covered in their entirety.

In a typical THz-TDS spectrometer (figure 153) the receiver detector, gated by a fraction of the laser beam generating the pulse, is used to measure the THz pulses amplitude. By varying the path difference between emitter and receiver the full extension of the pulse waveform is covered, so that the complete knowledge of amplitude and phase of the transmitted (or reflected) THz pulse, i.e. its spectrum, is obtained via Fourier transform. This is the TDS technique. The spark-gap emitter THz power is fairly low (microWatt), but the combination of detector gating and a suitable modulation of the emitter bias to drive a lock-in amplifier, increases the SNR to diagnostic-grade values, even for the short measurement time required for plasma phenomena. With the use of photoconductive polarization-resolving detectors the two polarizations of the probing wave can be resolved simultaneously [661,662]. Another great advantage of those techniques comes from the nature of THz-TDS components, based on solid state devices operating at room temperature, hence cheaper, more stable and compact than those used in conventional FIR and microwave applications. Finally, the pump and gate laser beams can be channeled to the emitter and receiving heads using optical fibers, gaining extreme flexibility in the routing, hence easing up the tokamak access issues. Since optical fibers are dispersive components, a group velocity delay compensation stage will be required. Finally, the use of the wide range of

<figure><img src="image_166.png" /><figcaption>**Figure 154.** Spectral transmission of tokamak diagnostics components measured over the relevant FIR frequency range by the THz-TDS spectrometer in figure 153.</figcaption></figure>

frequencies associated with the probing THz pulse, permits the determination of several sample parameters simultaneously.

All these aspects have raised interest in THz-TDS plasma measurements for density and Faraday rotation in the region of microwave and FIR since the early days [663]. The applicability of THz-TDS to the diagnostic environment has already been tested in laboratory by sampling the spectral response of FIR diagnostic components over a broad frequency range (figure 154) and with preliminary table-top plasma measurements [664].

### 7.11.2. Tokamak diagnostics applications.
One of the most promising diagnostic applications of THz-TDS is interferometry [558], a technique based on the measurement of plasma refractive index through the detection of the phase difference between a microwave beam traversing the plasma column vertically (i.e. parallel to the tokamak axis) and a reference beam travelling the same distance in air. A THz-TDS spectrometer with emitter and receiver located at the opposite ends of a vertical chord, will provide the same information of the interferometric arrangement. In addition by using several spectral components of the THz beam the plasma refractive index (hence the density) can be measured directly and the calibration accuracy of the diagnostics is increased.

Polarimetry measures Faraday rotation of the plane of polarization of a FIR beam which leads to the calculation of the local magnetic field strength [558]. By selecting the relevant frequency of the wide THz-TDS spectrum the same measurement can be performed with a high degree of simplification, also providing a straightforward implementation of the combined measurement of interferometry and polarimetry, without the extra components normally required by the heterodyne-style detection [557].
A more challenging diagnostic application of THz-TDS comes with reflectometry [557, 665], in which the phase variation of a selected frequency component reflected on the critical density (or cutoff) plasma layer where the refractive index vanishes, can be used to diagnose the plasma density profile or the density fluctuations. Traditionally a source at fixed frequency, or swept across the required frequency range is used [666]. The THz-TDS reflectometer will be instead an evolution of the Ultrashort Pulse Radar technique [667], with the broadband spectrum required for the diagnostic obtained through a Terahertz switch. By resolving the phase-shift of selected components the radial density profile can be obtained via Abel inversion.

The THZ diagnostic development is presently progressing along the lines of transition from table-top THz-TDS spectroscopy to plasma diagnostics on a tokamak, which will require a complete change of the layout and work scales. Paths will be in the order of meters, and the ‘sample’ is a large plasma column. The measurement time will necessarily be short, in order to follow the timescale of plasma phenomena (milliseconds). The coupling of emitter and receiver across the tokamak geometry distances will have to be optimized with tailored Gaussian beam optics telescopes. All considered, it can be safely stated that THz-TDS has great potential for evolution of diagnostics capabilities, beyond the routine plasma density and temperature measurements, towards the measurement of magnetic field, charge density and conductivity, with unprecedented resolution and reliability [668].

## 8. Influence of radiation effects on the refinement of the ITER design

I. Ďuran³⁸, A. Gusarov³⁹ and K. Vukolov²⁷

### 8.1. Introduction: requirements for diagnostic systems for ITER

The requirements for diagnostic systems for ITER are specified in the SRD-55 [197] and in the PRs [669]. For each parameter the measurement range and the diagnostic systems performing the measurements are defined. An important difference between ITER and already existing machines is a significant increase of nuclear radiation loads combined with a long-term stable operation. In particular, the in-vessel diagnostics will be subject to high energy neutron and gamma radiation fluxes and may not survive long enough to be considered as robust in the nuclear phases, when it will be most needed to guarantee a safe operation. It is necessary to evaluate the radiation hardness of each diagnostic and if necessary to consider mitigating measures to achieve their reliable operation. This evaluation must be confirmed by radiation testing in representative conditions.

### 8.2. ITER radiation environment

During nuclear operation, the ITER plasma will generate intense fluxes of fast neutrons and MeV-range γ-rays [670]. The principal source of radiation is the D–T reaction, which

generates neutrons with a mean 14.1 MeV energy. Subsequent neutron-matter interaction results in a broad neutron energy spectrum, and it also gives rise to secondary gamma radiation, with both prompt and delayed components. Therefore, the radiation environment at a sensor location is very complex and varies from one location to another.

Numerous diagnostic systems must perform correctly during the ITER life-time to guarantee a successful machine operation. Thanks to carefully selecting locations and to dedicated shielding, the nuclear loads on the sensors are significantly lower as compared to those to the ITER first wall materials. However, the sensors are considerably more vulnerable because they need to maintain over the long time not only the mechanical integrity but also specific properties such as optical transmission, electrical insulation, thermal conductivity, etc. This further highlights the potential problems of material transmutation and helium accumulation which might have significant impact on sensors performance and reliability.

It is necessary to analyze the impact of the nuclear environment on both each individual component and the system as a whole. This analysis requires an integrated approach to the design, when radiation effects and various environmental parameters such as elevated temperatures, vacuum, vibrations, must be considered simultaneously. Moreover, the systems must be compatible with multiple additional requirements, like remote handling operations, personnel safety and nuclear facility licensing requirements.

In the present publication we illustrate the complexity of the analysis using several diagnostics operating on different physical principles.

Simulating the ITER radiation environment presents a significant problem. In practice, there is no possibility to exactly reproduce ITER conditions in radiation tests. NGs can generate neutrons with corresponding energies, but the fluxes are very low and can hardly be considered as representative. To achieve the desirable total neutron fluence, the exposure has to be performed in a fission reactor, where the energy of neutrons is significantly lower than 14 MeV.

Exposure of materials to high energy radiation results in the generation of microscopic defects and changes of macroscopic properties. In the case of fast neutrons formation of lattice defects via kinetic energy transfer is usually considered as the most important effect. This effect is usually expressed by the ‘displacement per atom’ parameter (dpa). It is better to analyze such physical processes by means of numerical analysis, because it is practically impossible to make it experimentally for ITER conditions.

The dpa is defined as a ratio between atomic displacements (lattice defects) occurring within a given volume and a number of atoms within the same volume. To produce a permanent displacement, the energy transferred to an atom should be bigger than its binding energy. Otherwise, the transferred energy will dissipate without any stable defect created. If the energy transferred to an atom is significantly higher than the binding energy, the displaced atom will move through the lattice and can create more defects. According to the modified Kinchin–Pease model, the total number of defects is defined as [671]:
$$N = 0 \quad \text{for } T < E_d$$

$$N = 1 \quad \text{for } E_d < T < 2.5 E_d,$$

$$N = \frac{T}{2.5E_d} \quad \text{for } T > 2.5E_d,$$

where T is the kinetic energy of the incoming neutron and $E_d$ is the energy required to remove an atom from a lattice position. The integral dpa rate $d(t)$ is calculated as:

$$d(t) = \int dE \sigma_d(E) \varphi(E, t),$$

where E is energy, $\sigma_d(E)$ is the dpa cross-section and $\varphi(E, t)$ is the neutron flux energy spectrum and the total dpa is obtained by integration over time t.

Thanks to its apparent simplicity the dpa is often taken as a convenient figure of merit to quantitatively characterize neutron radiation damage. For example, 0.1 dpa means that one atom out of ten was displaced from its lattice position, which provides a simple visualization for the actual radiation effect. Alternatively, the ionizing radiation dose in Grays (1 Gray equals 1 Joul absorbed within 1 kg of material) is used as a measure of radiation damage. However, there is less obvious link between the value of ionizing radiation dose and the actual microscopic changes within irradiated material as it is a more abstract quantity.

Nevertheless, care should be taken when using a dpa value to characterize radiation damage. The amount of dpa cannot be measured and it is not directly proportional to changes in material properties. The threshold displacement energy $E_d$ plays a crucial role in the dpa concept. For some materials, there are big variations in the threshold displacement energies predicted by different models. In case of crystalline materials, each crystal direction has its own $E_d$ due to crystal anisotropy. It is possible to calculate an average over different directions. However, in case of directional radiation (accelerators) of a monocrystalline material the sample orientation will play an important role.

To prepare an irradiation test of an ITER component, there are several important conditions which should be respected. Firstly, it is important to achieve the neutron fluence and gamma dose at levels corresponding to those expected during operation of the component on ITER. The aim is to approximate as close as possible the kinetics of gamma-neutron processes occurring in ITER during the test. Moreover, several times higher neutron fluence can be required when samples are irradiated in a nuclear reactor to compensate the quantitative difference in the fission neutron energy compared to the thermonuclear neutrons. Secondly, if the effect of the neutron irradiation is not decisive in the process under study, then only gamma irradiation, usually at a Co-60 source, can be used for the tests. In some cases tests should be performed at a DT neutron source with the maximum achievable neutron fluence to determine the influence of thermonuclear neutrons on the physical process under study.

Activation, transmutation and He production are radiation effects, which are extremely difficult to reproduce in a way

representative for ITER. The cross sections of nuclear reactions are extremely sensitive to the neutron spectrum. The ratio fast/thermal neutron flux in a material testing fission reactor depends on the location, but in most of cases it is lower than in ITER. As a result, it is virtually impossible to achieve the representative values of the displacement damage, transmutation fraction, and amount of He in a single thermal fission reactor irradiation. Usually, obtaining the desired value of the dpa, in a fission reactor experiment means that the transmutation effect is significantly higher than it will take place in ITER. Therefore, separate tests should be considered to study displacement damage, transmutation, and He production. The neutron energy distribution for the tests should be properly scaled to be representative to the spectrum at the location of the test sample in the ITER.

The numerical simulation of these effects can be considered for ITER conditions as a sufficiently accurate way to analyze radiation damage, activation and transmutation while taking into account the energy spectrum of neutrons and gammas. The Monte-Carlo N-particle (MCNP) transport code is officially adopted for calculations in ITER [198]. Neutron calculations are performed using the official reactor model 'C-Model R181031 REV. 190110 issued 10/01/2019' [672]. This is a three-dimensional CAD model of a toroidal 40° ITER sector with its main structures. The neutron source is given for the basic ITER scenario with a maximum thermonuclear D-T reaction power of 500 MW at plasma current of 15 MA [673].

For the neutronic analysis, the diagnostic components must be added to the C-Model with the geometry of their placement and corresponding materials contained. The MCNP code calculates the neutron spectra at each interaction of a neutron with the material. The statistical error of the calculated neutron flux is about a few percent, but the actual accuracy of the calculated neutron spectra depends on the correctness of the model geometry and the material composition.

The activation, transmutation and degree of radiation damage of diagnostic components can be estimated on the basis of data on neutron fluxes, taking into account neutron spectra, as well as using libraries of micro-constants of nuclear reaction cross sections.

Nevertheless, radiation testing of diagnostic elements is absolutely necessary to justify their suitability, and therefore available facilities, even though they cannot provide a fully representative environment, have to be used. This means that compromises are required to choose an installation, which allows for the best evaluation of the most important performance parameters. Often, measurements of properties of irradiated materials are difficult due to their high radioactivity. Safety requirements exist for such type of measurements that make them very complicated and sometimes even impossible.

## 8.3. Optical diagnostics

Optical diagnostics are widely used in modern thermonuclear devices. They can measure in situ the main characteristics of the plasma, such as temperature, density of electrons and ions, concentration of impurities, local electric and magnetic fields, etc. Optical diagnostics are essential for tokamak
physics studies, but available designs cannot be directly trans-
ferred to ITER without improvement of the radiation hardness,
especially since ITER sets high requirements for the measure-
ment accuracy. The ITER optical elements will change proper-
ties under gamma-neutron radiation emitted by the hot plasma.
Transparent glass elements (windows, lenses, prisms, filters
and optical fibers) that are needed to collect radiation from
the plasma and to guide it to detectors will quickly lose trans-
mittivity (effect of radiation induced absorption—RIA), espe-
cially in the UV spectral range. Besides that, some glasses
radiate light due to radio-luminescence and/or Cherenkov
effects under gamma-neutron irradiation. Detectors (PMTS,
video cameras, etc) placed close to the plasma also can be dam-
aged by the irradiation.

The vacuum window assembly is an example of a com-
ponent containing optical material subject to strong radiation.
The plasma and the first wall diagnostic system in ITER will
use more than 100 diagnostic windows required for optical,
microwave, and x-ray diagnostic systems. In addition to trans-
mitting the diagnostic signal, they must also maintain the
safety barrier between the radioactive plasma and the external
atmosphere.

Window assemblies are located on the VV port plugs and
the port interspace seal plates. They are difficult to replace and
are inaccessible during machine operation and during move-
ment of main in-vessel components for the duration of a shut-
down. The assemblies must be available continuously during
the full operating life of ITER. Therefore a high reliability of
the window assemblies is required.

Two types of silica glass KU-1 and KS-4V with hydroxyl content OH < 1000 ppm and OH < 0.1 ppm, respectively, were investigated under ITER relevant conditions during round-robin irradiation tests [674]. A water-pool nuclear reactor, a Co-60 gamma source, and a NG were used for this. It was shown that both types of silica glass have good radiation hardness in the visible spectral range and can be used as window materials for ITER optical diagnostics.

An initial window assembly design considers using brazing to attach the transparent element to the housing, figure 155. Irradiation of such assemblies was performed in the BR1 reactor of SCK●CEN in Mol at 150 °C during 5.5 months up to a total fast (E > 0.1 MeV) fluence of ~2 × 10¹⁶ n cm⁻² [675]. Post-irradiation thermo-mechanical and He-leak tests were performed in a hot cell. The examination was concluded with a destructive push-out test. Different damage patterns were observed for different window assemblies. A general conclusion was that after thermal cycling and reactor irradiation the windows remain sufficiently strong to sustain mechanical loads relevant for the ITER operation.

The current window assembly design considers the use of
two fused silica 2 cm thick plane plates [676]. The first plate
is tilted by 5 degrees around the horizontal axis, the second
surface is tilted by 5 degrees around the vertical axis in order to
reduce multiple reflection effect arising between the innermost
window surfaces. MCNP calculations yield that neutron fluxes
at the window can reach 10⁸ n (cm⁻² s⁻¹) in spite of effective
neutron shielding located in the port-plug. Radiation testing of
such assemblies is planned for the period 2022–24.

<figure><img src="image_167.png" /><figcaption>**Figure 155.** Brazed prototype window assemblies, UKAEA design.</figcaption></figure>

Fiber optic bundles are intended to transfer plasma light from the ITER Port areas to detectors in the diagnostic room [677]. They will be exposed to gamma-neutron radiation along significant fraction of their length (near 10 m) and it is better to locate them as far as possible from the plasma to decrease the effect of RIA. In present, suitable types of radiation-resistant silica fibers have been identified and their production technologies have been developed. The silica optical fiber doped with H₂ and coated by a thin carbon film has a high radiation resistance and their transmittance can be recovered during pauses between ITER plasma pulses by exposing them to gamma irradiation [252, 678]. Therefore, such fibers are candidates for the manufacture of fiber bundles.

8.4. Magnetic diagnostics

Successful operation of ITER will rely on the use of a large
set of magnetic diagnostics [4]. From data obtained with those
systems, fundamental parameters such as plasma position,
shape, and current will be derived to allow for real-time plasma
control by a set of poloidal coils as well as for MP.

8.4.1. *Cables and connectors.* The ITER magnetic diagnostic systems usually require detection of low currents or voltages. Therefore, mitigation of radiation and thermal effects induced within the sensor's cable path, such as radiation induced conductivity (RIC), radiation induced electromagnetic force (RIEMF), temperature induced electromotive force (TIEMF), radiation induced thermo electric sensitivity (RITES), and radiation induced electric degradation (RIED) presents a very significant challenge. These effects were addressed in numerous publications. It is concluded that the internally twisted mineral insulated cables often offer the best intrinsic balance for most of these effects and for rejection of most long wavelength external effects, leaving only the defect-driven TIEMF as the main source of spurious voltage. However, the problem that still requires significant attention is the cable termination and the connection to the delicate wiring of a particular sensors.

In addition, dedicated experiment was performed in order
to assess potential dimensional changes of glass to metal
sealing (GTMS) inserts within ITER electrical feedthroughs
(figure 156) due to neutron radiation. Dimensional changes of
<figure><img src="image_168.png" /><figcaption>**Figure 156.** Example of ITER electrical feedthroughs with borosilicate GTMS inserts.</figcaption></figure>

these inserts could potentially compromise vacuum tightness of the feedthroughs.

Four samples of borosilicate GTMS inserts with outer diameter ~7.1 mm were irradiated in LVR-15 experimental fission reactor in CVŘ Rež, Czech Republic. The samples were irradiated inside cadmium shielded container in order to increase fast to thermal neutron ratio to better approximate actual neutron spectrum the feedthroughs will be exposed to on ITER. The total accumulated neutron fluence during this irradiation as measured by embedded activation foils reached 6 × 10¹⁷ cm⁻² (*E* <0.1 MeV) and 5 × 10¹⁷ cm⁻² (*E* > 0.1 MeV). The outer diameter of the samples was measured before irradiation using optical measurement system Vertex 251 located in the LVR-15 hot chambers with measurement accuracy ~1 µm. The same measurement of outer diameter was done on irradiated samples. Measured differences between pre-irradiated and post-irradiated samples were comparable with measurement accuracy of the Vertex 251 system, i.e. ~1 µm. It was concluded that radiation induced dimensional changes of these borosilicate glass GTMS inserts pose no risk with respect to the potential loss of vacuum accident on ITER.

### 8.4.2. Inductive sensors.
Until now, magnetic measurements in tokamaks are successfully performed using inductive sensors, where the signal (voltage) is proportional to the derivative of the magnetic flux through the sensor. To find the plasma current or magnetic field an integration step is necessary. Currently used analogue integrators have known intrinsic problems. Two most important are the integrator drift, which introduces an absolute error increasing with the integration time, and saturation of the integrator input stage in the case of fast flux variation, for example during plasma disruptions.

In ITER during the long stationary phase, the signals to be integrated are essentially zero, and for relevant changes of the plasma equilibrium the resulting magnetic signals are small. This makes the magnetic diagnostics sensitive to any spurious voltages that may arise from RIEMF and TIEMF [6].

The inner vessel sensors, which include different types of coils and loops, are the main magnetic set. They are distributed

just behind the blanket modules or mounted on the instrumented divertor cassettes. Because of their in-vessel location, they are subject to significant radiation loads, in particular in the divertor region.

The measurement accuracy of magnetic sensors defines margins to the machine engineering limits and the operational flexibility. The coils are the most critical element, because their effective area must be rather small, which limits acceptable parasitic voltages to values below 0.5 µV for long-pulse operation. On the other hand, the size is still big to allow for temperature gradients, which may influence the sensor accuracy.

An option for magnetic sensors selected for ITER is the use of LTCC coils. The LTCC sensor consist of thin ceramic layers, on which metallic tracks are printed and then the unit is sintered. Such sensors were proposed as the most promising solution to meet the requirements imposed by the ITER operating conditions [679].

To check that LTCC sensors will perform satisfactorily under the ITER radiation environment two types of prototype ITER LTCC coils were irradiated up to 72 MGy using a Co60 source at ~50 °C [680]. LTCC-Ag/10 had 10 ceramic layers with silver tracks, and LTCC-Au/30 had 30 ceramic layers with printed gold conductors. The continuity and the insulation resistance were measured *in situ*. The leak currents for the coils during irradiation were 70 nA and 35 nA at 50 V, for Ag and Au coils respectively, which corresponds to an insulation resistance in the GOhm range. One LTCC-Ag/10 coil was broken during the rig assembly, likely due to a non-uniform compression of the coil. Also, the fragility of the connectors was identified as a problem, which is actually relevant for various diagnostics at ITER.

A neutron irradiation of LTCC sensors has also been performed at the LVR-15 nuclear reactor [15]. Four samples, including LTCC-Ag/10 from the above gamma-irradiation experiment, were exposed. The accumulated fast (*E* > 0.1 MeV) neutron fluence was ~10²⁰ cm⁻², exceeding the expected ITER lifetime fluence for these sensors. Due to the high activation, only visual inspection of the sensors was possible after the irradiation. The sensor pre-exposed to gamma radiation which was also subject to highest thermal gradient during neutron irradiation was cracked and three others maintained their physical integrity. The failure was attributed to the thermal stresses which occurred due to radiation heating and exceeded LTCC ceramic material limits. Effect of possible gamma-radiation induced ageing was concluded to be negligible. Without electrical measurements being possible due to high activation of the samples, the question about changes of performance characteristics of the exposed sensors remains open.

The risk of the in-vessel sensors failures due to damage induced by radiation effects will be mitigated by installation of additional sensors on the outer surface of the VV, where the neutron flux is an order of magnitude lower [6]. The shielding by the vessel, however, drastically reduces the frequency response of these sensors down to a ~Hz level [681], which means they can only be used for slow plasma position drift correction. An additional design constraint for these sensors is
<figure><img src="image_169.png" /><figcaption>**Figure 157.** Samples of ceramics metallized by copper using thick printing technology before and after irradiation on LVR-15 fission reactor.</figcaption></figure>

a space less than 10 mm in the radial direction, which is the width of the gap between the VV and the thermal shield.

Recently, a few innovative concepts of inductive sensors are appearing and are being developed and qualified primarily with respect to their deployment on future fusion reactors beyond ITER.

One of such new approaches is application of thick printing (TP) technology to manufacture robust and radiation hard inductive sensors [682, 683]. Here, the spiral-like pattern of metallic ink is screen printed onto already sintered ceramic substrate creating effectively a coil-like structure. The substrate with screen printed ink is then fired to create and stabilize final metallic tracks. Various materials can be used both for ceramic substrates, while the most common one is Al₂O₃, and also for metallic ink, where the most common are those based on copper. Stacking of several layers of such metalized ceramic patterns is foreseen creating similar structure as it is in the case of LTCC sensors. The main advantage of TP technology is separation of ceramic sintering procedure from the metallic ink firing procedure. This reduces the risks and removes technological peculiarities associated with LTCC sensors manufacturing [18] where dimensional changes of metallized ceramic tapes during firing can lead to sensor defects. On the other hand, TP technology leads rather to a laminar sensor structure compared to a compact ceramic block typical for LTCC sensors.

Initial study of radiation hardness of TP copper metallization of two types of Al₂O₃ (96% and 99% purity) and also AlN substrates was done at LVR-15 fission reactor, figure 157. Several samples were irradiated by total neutron fluence of 4 × 10¹⁸ cm⁻² and adhesion of the copper layer on irradiated samples was tested. The copper layer adhesion was not significantly affected by the neutron irradiation and in all cases the post irradiation adhesion was better than 35 N mm⁻².

Another new approach to design and manufacturing of local magnetic field inductive sensors is employment of photolithography (PL) technology [684]. Here, the ultrathin ~10 µm conductive golden tracks can be shaped using standard PL technology on SiO₂ wafers. The main advantage of this

<figure><img src="image_170.png" /><figcaption>**Figure 158.** Design layout of the ITER OVSS sensors.</figcaption></figure>

approach is achievement of large total effective area of the resulting coil while keeping its self-inductance small which allows to use the same sensor both for equilibrium magnetic field reconstruction and also for detection of high frequency MHD events. Key aspects of this technology are being developed like stacking and reliable electrical connection of such multilayer coils, using substrate and metallization materials more relevant to harsh fusion environment etc. Initial neutron irradiation tests of the first prototypes are also foreseen in the near future.

8.4.3. *Hall effect sensors.* Hall sensors are included in the ITER design to mitigate the risk of voltage drift in the inductive sensors due to RIEMF and TIEMF. Three ITER VV sectors (8, 5, and 2—in order of installation) will be equipped with full poloidal arrays of outer vessel steady state (OVSS) magnetic field sensors based on bismuth and antimony. Sensors will be mounted within a narrow gap between the outer VV skin and the thermal shield. Each probe will contain two Hall sensors, oriented perpendicularly to each other, to measure poloidal magnetic field and N type thermocouple to properly compensate for temperature dependence of sensors sensitivity.

No maintenance is possible for these sensors due to lack of access after ITER VV assembly. As a result, they will have to survive for the full ITER lifetime, which calls for the extensive qualification and testing including radiation hardness assessment.

Design and material choice for the ITER OVSS see figure 158 was driven by three main criteria:

1. Radiation hardness up to maximum expected ITER life time neutron fluence ~2 × 10¹⁸ cm⁻².

2. Compatibility with long term exposure to ITER VV baking temperature of 220 °C in vacuum.

3. Achieving as high sensitivity as possible to improve SNR and achieve required absolute accuracy with measurement error not exceeding 4 mT.

The first generation of the OVSS sensors [36] employed ~1 µm layer of bismuth prepared by magneto sputtering as sensitive material. AlN ceramic substrates metallized by copper using commercial direct bond copper (DBC) technology