<figure><img src="image_57.png" /><figcaption>**Figure 45.** LVNC detectors module.</figcaption></figure>

<figure><img src="image_58.png" /><figcaption>**Figure 46.** Section of UVNC in UP Port Plug. 55.B2 detector module is outlined with yellow. Section is done via one half of 55.B2.B0, still LoS for both are visible. Pink area in front of module—DFW with dedicated water cooling lines.</figcaption></figure>

axis. Adjustment is done with the bolts on the fixation points to diagnostic rack. Five detector units of the module observe plasma, while the last one do not have associated collimating holes thus providing data on neutron and/or EM background for measurements.

### 3.5.2.2. *Upper VNC.*
In the figure 46 it is depicted UVNC design in UP18 port plug. Detectors module consists of two boxes independently attached to DSM body. Between two UVNC halves there is a 25 mm slot made for VUV (55.EH) diagnostic system. Each half implements three lines of view, no ‘blind’ channels foreseen for UVNC.

No divertor or blanket elements participating UVNC, but instead DSM and DFW elements. The latter elements are directly bolted together and also closer to detectors than relevant elements of LVNC thus giving no need for complex tolerances analysis here. However other challenges were faced during development of the Upper Camera. So, in the analysis of 2017 it was found for UVNC signal to background (S/B) being <1 for all the channels. Hence optimization was proposed.

<figure><img src="image_59.png" /><figcaption>**Figure 47.** Collimation for UVNC detectors. First iteration extending LoS—upto 13.7° of larger side of rectangle.</figcaption></figure>

<figure><img src="image_60.png" /><figcaption>**Figure 48.** Detector units of UVNC. (a)—electrodes, (b)—detector unit model (section), (c)—detector unit mock-up, (d)—detector unit inside housing structure (implementing SVS enveloping).</figcaption></figure>

First step was changing collimator shape from cylindrical (as it is still in LVNC) to rectangular—extended in the direction of the magnetic surface (see figure 47).

Then the electrodes of the FCs were adjusted accordingly (see figure 48(a)), model of detector unit updated (figure 48(b)), simulation performed and mock up (figure 48(c)) manufactured and tested.

One of the most complicated component of these collimators implementation became development of the DFW cutouts. Indeed, this first PFC needs dense system of cooling channels,
<figure><img src="image_61.png" /><figcaption>**Figure 49.** Collimation for UVNC detectors. Dedicated DFW cutouts geometry.</figcaption></figure>

which can be manufactured only if set of strict restrictions applied, for example, minimum distance between cutouts, their profile. This led to a slight worsening of collimation efficiency: particularly because some cutouts for LoS had to be united, others left only at one half (box)—see figure 49. Still S/B became much better (typically ~3–4) than before making LoS rectangular.

As the next step, detector Unit housing was redone to implement electrodes coming out at the side (figure (48d)) instead of the cylinder basement (as it was considered before). That made Housing shorter gaining addition space for the shielding material with collimators in front of the detectors. More information on the optimizations can be found in [190].

**3.5.3. Measurements requirements and system performance analysis.** Requirements for the primary measurements are given in tables 19 and 20. It is taken from latest approved 55.B2 sSRD (sub system requirement document).

In order to justify suggested design, ability of diagnostic performance to fulfill sSRD measurements requirement checked with modeling, based on multiple MCNP simulations, described in [191]. Green function formalism was applied, as shown in figure 50 cited from given reference.

This approach allows quick calculation of the detector response for different plasma configuration (without need of full scale MCNP modeling for each case), and was applied to demonstrate required diagnostic performance. Statistical errors, inaccuracies of magnetic surface shape, inaccuracies of contribution matrix calculations were analyzed to prove that reconstructed profiles are tolerant to input data errors for the reverse problem under consideration.

Indeed, even in case of large errors (e.g. statistical) median of reconstructed quantity remains close to the true value. For one of the MPs restoration quality demonstrated in the figure 51.

**3.5.4. Shielded cabinets.** Neutron detectors (including FCs and sCDs) produce weak signals in mA range, which requires usage of preamplifier located not farther than dozens meters from detectors. This forces various neutron diagnostics to have corresponding electronics in ITER port cells. Operation conditions there are hostile for electronics, including high radiation and electromagnetic fields [193]. In order to protect preamplifiers dedicated shielding is foreseen.

Despite just recently having last PDR, VNC have one of the most advanced (among ITER systems) designs of the shielded cabinets. Current CAD model for LP14 is given on the figure 52.

In lower port 14 common shielding approach applied with LVNC cabinet hosting both preamplifiers of 55.B2 (red color on left image of figure 52) also of DNFM (red units). Most of the cabinet volume constitutes protection against neutrons.

And using of common shielding approach allows to limit with couple dozen percent's increase of the weight, space and cost for extending solution on two system, while independent cubicles would require doubling these values. Depending on the neighbors in other ports common shielding is planned to be used there as well.

Shielded cabinets of VNC are located in the port cell support structure (PCSS) and consists of neutron shielding made of B4C blocks, gamma shielding made of tungsten block and pure iron plates for magnetic screening. Tungsten block and iron plates are fixed at the stainless steel supporting frame. This frame has doors from the front and back sides to provide access to electronic components.

One of the key analyses justifying shielding solution is MCNP calculations, which for LVNC shielded cabinet confirmed necessary level of neutron attenuations from ~$10^6$ to allowed for non-radiation hard electronics $10^2$ n ($s^{-1}$ cm$^{-2}$). Attenuation of the neutron flux dependence from the point position inside the shielded cabinet for one of its cross sections is given in the figure 53.

## 3.6. NFM

**3.6.1. Introduction.** NFMs are among the most important systems in nuclear fusion devices as they are used to measure total neutron emission rate and then to evaluate fusion power. In ITER, the total neutron emission rate will be measured by three types of the NFMs: the NFM [194], the Divertor NFM (DNFM) [195] and the microfission chamber (MFC) [196].

Requirements for the NFMs are given in table 21. Table 22 shows the role of measurement of the required parameters for the NFMs. They are taken from latest approved SRD-55 (system requirement document) [197]. The NFMs shall measure the total neutron flux and then evaluate the fusion power with time resolution of 1 ms and accuracy of 10% in certain plasma parameter range.

### **3.6.2. Microfission chamber.**

**3.6.2.1. General description.** The MFC is operated as the in-vessel NFM in ITER. An MFC detector is a pencil-sized gas counter (14 mm in diameter and 200 mm in length) containing fissionable material (235U) that can be installed in the VV. Figure 54(a) shows a cross sectional view of the MFC detector used in ITER.

A coating of $UO_2$ covers the outer cylindrical electrode of the MFC. The active length of the device is 76 mm, with a total of 10 mg of 235 U used. The MFC detector is filled with argon gas at 14 atm, which serves as the ionization gas. The
<table><caption>**Table 19.** Fusion power density profile measurement parameter requirements.</caption><thead><tr><td>Parameter</td><td>Range</td><td>Time Res.</td><td>Spatial Res.</td><td>Accuracy</td><td>Role</td></tr></thead><tbody><tr><td>013a. Fusion power density</td><td>1–20 kW·m<sup>-3</sup></td><td>1 ms</td><td>a/10</td><td>20%</td><td>1a.2 BC</td></tr><tr><td>013b. Fusion power density</td><td>20–4000 kW·m<sup>-3</sup></td><td>1 ms</td><td>a/10</td><td>10%</td><td>1a2. BC</td></tr></tbody></table>

<table><caption>**Table 20.** Neutron-and-Alpha-source profile measurement parameter requirements.</caption><thead><tr><td>Parameter</td><td>Range</td><td>Time Res.</td><td>Spatial Res.</td><td>Accuracy</td><td>Role</td></tr></thead><tbody><tr><td>014. Neutron-and-alpha-source profile</td><td>10<sup>14</sup>–1.4·10<sup>18</sup> nm<sup>-3</sup>s<sup>-1</sup></td><td>1 ms</td><td>a/10</td><td>10%</td><td>1b.AC</td></tr></tbody></table>

<figure><img src="image_62.png" /><figcaption>**Figure 50.** DT neutron contribution matrix for fission chambers (sensitivity of particular detectors of LVNC to different point of the plasma cross section).</figcaption></figure>

<figure><img src="image_63.png" /><figcaption>**Figure 52.** LVNC shielding cabinet on PCSS of LP14. Left—internal structure ensuring protection against EM and γ fields, right—external protection from B4C bricks for efficient neutrons attenuation.</figcaption></figure>

<figure><img src="image_64.png" /><figcaption>**Figure 51.** Neutron source profile reconstruction for reference source DT, *I* = 15 MA, *P* = 500 MW.</figcaption></figure>

<figure><img src="image_65.png" /><figcaption>**Figure 53.** Spherical source. Neutron attenuation.</figcaption></figure>

housing material is fabricated from stainless steel 304 l and the electric insulator is composed of alumina (Al₂O₃). The structure of the tri-axial MI cable is shown in figure 54(b). The core is formed of Cu, while SiO₂ is used as an electric insulator between core and inner sheaths and between inner and outer sheaths. Argon gas fills the inner sheath to prevent any discharge between the core and the inner sheath. The MI cable has been designed based on proven MI cables used in several fission reactors. Based on its use in fission reactors and on the results of Monte Carlo N-particle (MCNP [198] neutronic calculations for the ITER conditions, it has been confirmed that the MI cable can operate without any

problems under ITER radiation conditions during the ITER lifetime.

The MFC detectors will be installed between the VV and blanket modules at both the upper and lower outboard positions, as shown in figure 55.

The reason for installing the MFC detectors in such locations is to compensate for the change in sensitivity of each detector when the plasma shape changes [199]. At each proposed location, two MFC detectors will be installed. The deployment of dual MFC detectors at each location improves redundancy. The MI cables are routed from the upper and lower detector positions to the upper port and connected to the
<table><caption>**Table 21.** Fusion power density profile measurement parameter requirements.</caption><thead><tr><th>Parameter</th><th>Range</th><th>Time Res.</th><th>Spatial Res.</th><th>Accuracy</th><th>Role</th></tr></thead><tbody><tr><td>012 Fusion power</td><td>100 kW—0.9 GW</td><td>1 ms</td><td>Integrated</td><td>10%</td><td>1a.2 BC</td></tr><tr><td>015. Total neutron flux</td><td>10<sup>14</sup>–3.2·10<sup>20</sup> ns<sup>-1</sup></td><td>1 ms</td><td>Integrated</td><td>10%</td><td>1a.1 MP</td></tr></tbody></table>

<table><caption>**Table 22.** Fusion power density profile measurement parameter requirements.</caption><thead><tr><th>Parameter</th><th>Primary</th><th>Supplementary</th></tr></thead><tbody><tr><td>012 Fusion power</td><td>55.B4 Neutron flux monitors<br>55.BC Divertor neutron flux<br>Monitors</td><td>55.B3 Microfission<br>Chambers</td></tr><tr><td>015. Total neutron flux</td><td>55.B4 Neutron flux monitors<br>55.BC Divertor neutron flux<br>Monitors</td><td>55.B3 Microfission<br>Chambers</td></tr></tbody></table>

<figure><img src="image_66.png" /><figcaption>**Figure 54.** MFC detection system: (a) Schematic of the MFC for ITER; (b) the structure of the MI cable.</figcaption></figure>

<figure><img src="image_67.png" /><figcaption>**Figure 55.** Arrangement of the poloidal cross section of the in-vessel components of the MFC system on ITER tokamak.</figcaption></figure>

soft cable at the feed through in the upper port. The lengths of the MI cable for the upper and the lower MFCs are about 4 and 11 m, respectively.

The current plans for installing the in-vessel components of the MFC system into the VV require the MFC detectors and MI cables to be installed at different times. As the MFC contains fission material (²³⁵U), it should be installed in the VV after first plasma operations in order to ensure safety. On the other hand, the MI cables are scheduled to be installed in the early phases of the process of assembling the tokamak to allow for their routing under the water manifold and to avoid interface conflicts with other in-vessel components. Therefore, it is necessary to connect the MFC detectors (with short MI cables) to the MI cables inside the VV. Although MFC detectors and MI cables are manufactured as a unit in fission reactors, in ITER the MFCs and the MI cables use a connector which was designed in a previous study. Figure 56 shows the structure of the connector. The core is connected by resistance spot welding using Ni-foil, which has a good track record in neutron detectors for fission reactors. Ceramic shields are

inserted as insulators to avoid any contact of the Ni-foil with the inner sheath as well as to avoid any contact between the inner sheath and the outer sheath. The workability of the connection between the MFC detector and the MI cable in the limited space within the VV has been demonstrated in several tests.

### 3.6.2.2. System performance analysis.
As described above, the MFC detectors will be installed upper and lower outboard positions to compensate for the change in sensitivity of each detector when the plasma shape changes. It has been confirmed by Neutronic calculations using MCNP [198].

Dependence of the detection efficiencies on the horizontal plasma shift is shown in figure 57(a). The detection efficiencies are very weak functions against the horizontal plasma position. Dependence of the detection efficiencies on the vertical plasma shift is shown also in figure 57(b). The detection
<figure><img src="image_68.png" /><figcaption>**Figure 56.** Schematic view of the vacuum tight tri-axial connector of the MI cable; (a) the external appearance, (b) the inner structure of the core region and (c) the inner structure of the side region.</figcaption></figure>

<figure><img src="image_69.png" /><figcaption>**Figure 57.** Dependence of the detection efficiencies on the horizontal plasma shift (a) and the vertical shift (b).</figcaption></figure>

efficiencies of the upper MFC increases monotonically with the vertical plasma shift. On the other hand, that of the lower MFC decreases.

The total neutron emission rate can be represented by the linear combination of the MFC counts as follows,

$$S_n = \frac{w_{up}C_{up}}{\varepsilon_{up}} + \frac{w_{low}C_{low}}{\varepsilon_{low}}, \quad w_{up} + w_{low} = 1$$

where $S_n$ is the total emission rate, $w_i$ is a weight for the linear combination. $C_i$ is a count rate, and $\varepsilon_i$ is detection efficiency for the reference plasma parameters. The weight $w_i$ are determined to minimize the deviation of the total neutron emission rate from the neutronic analysis. The errors of the total neutron emission rate due to change in the plasma parameters are shown in figure 58. It was found that the error due to change in the plasma parameters are less than $\pm 3\%$.

The MFC detectors are installed in a VV and cannot be replaced. Therefore, it is necessary to demonstrate the MFC can properly measure the total neutron emission rate over the operational period of ITER.

A lifetime of the MFC is determined by the change of the sensitivity of the chamber due to the burn-up of the fissile material. $^{235}$U is burned up through mainly fission and neutron capture reactions. So, the number of $^{235}$U atoms, $N_{235U}$, is represented by following equation;

$$\frac{d}{dt}N_{235U}(t) = -N_{235U}(t)\phi(\sigma_{f235U} + \sigma_{C235U}) \quad (3.1)$$

<figure><img src="image_70.png" /><figcaption>**Figure 58.** Error of the detection efficiencies for (a) the horizontal plasma shift and (b) vertical shift.</figcaption></figure>

where $\sigma_{f235U}$ and $\sigma_{C235U}$ are averaged fission and neutron capture cross-sections defined by

$$\sigma = \frac{\int \sigma(E) \phi(E) dE}{\int \phi(E) dE} \quad (3.2)$$

where $\phi(E)$ is the neutron energy spectrum at the micro FC. From equation (3.1),

$$\begin{aligned} N_{235U}(t) &= N_{235U}(0) \exp\{-\phi(\sigma_{f235U} + \sigma_{C235U})t\} \\ &\approx N_{235U}(0) \{1 - \phi(\sigma_{f235U} + \sigma_{C235U})t\} \\ &\approx N_{235U}(0) (1 - \phi\sigma_{f235U}t) \end{aligned}$$

is obtained. For the detector position behind a blanket (typically #12), the burn-up of $^{235}$U atoms is dominated by the fission reaction, because the fission cross-section is about $10^2$ larger than the capture one. The burn-up rate of $\phi(\sigma_{f235U})$ is estimated $2.7 \times 10^{-11}$ s$^{-1}$. The change of the sensitivity is estimated to be only 0.1% behind blankets for the ITER life time which is equivalent to 0.5 GW year. So $^{235}$U chambers can be used without replacement in the ITER lifetime.

The effect of the strong magnetic field on the FC is another problem. In the FC, the electron induced current from ionization by fission fragments is measured. So, the electron drift orbit in the magnetic field was calculated. The electron drift velocity $\mathbf{u}$ is represented by

$$\mathbf{u} = \frac{\mu_e}{1 + \omega_c^2/\nu^2} \left[ E + \frac{\mathbf{E} \times \mathbf{B}}{B} \frac{\omega_c}{\nu} + \frac{(\mathbf{E} \cdot \mathbf{B})}{B^2} \frac{\omega_c^2}{\nu^2} \right] \quad (3.3)$$

where $\mu_e$ is the electron mobility, $\nu$ is the collision frequency of the electron to neutral atoms, and $\omega_c$ is the electron cyclotron (EC) frequency in the magnetic field $\mathbf{B}$. Assuming $E = (E_x, 0, 0)$ and $B = (0, 0, B_z)$,

$$\mathbf{u} = \left( \frac{\mu_e E_X}{1 + \omega_c^2/\nu^2}, \frac{\mu_e E_X}{1 + \omega_c^2/\nu^2} \frac{\omega_c}{\nu}, 0 \right). \quad (3.4)$$

An angle between $\mathbf{u}$ and $\mathbf{E}$, Lorentz angle $\alpha$, is represented by $\tan \alpha = \omega_c/\nu$. In the case of the MFC with 14.6 atm Argon gas, applied voltage of 200 V to 0.5 mm electrode gap, in the magnetic field of 5.7 T, the drift velocity without magnetic field is $u_0 \approx 3 \times 10^3$ m s$^{-1}$. The Lorentz angle is evaluated as $\tan \alpha \approx 0.04 \ll 1$. Thus, the magnetic effect on the electron drift velocity is to be negligible.
**3.6.2.3. Summary.** The MFC system is being developed for installation and plasma operation in ITER. The MFC must withstand the extremely harsh environment inside the VV, such as high heat load and high electromagnetic forces. In addition, the MFC system must be able to accurately measure total neutron emission rate in various noisy environments outside the VV. In order to achieve the ITER requirements, Japan Domestic Agency is intensively advancing the detailed design of the MFC from various perspectives, including structural integrity and noise suppression.

## 3.7 NAS

**3.7.1. Introduction.** The NAS is a diagnostic measuring the absolute neutron flux and fluence on the FW [200]. It utilizes pneumatic post method to send a sample of material close to the plasma, where it gets activated by neutrons. The sample is then retrieved back with the same pneumatic post technique and the activation of the sample is measured with gamma-ray spectrometers [201]. The main goal of the ITER NAS is to evaluate the total neutron production rate from all over the plasma. The measurement accuracy depends on the position and profile of the plasma and the variation of material in front of the irradiation end (IE). It is required to minimize the amount of material and its density variation across the volume between the plasma and the IE. The influence from in-vessel materials such as the DFW, blanket modules, and divertor cassettes, that are located near the IEs is unavoidable. A number of irradiation positions located above and below the plasma as well as on high-field side and low-field side has been selected for the ITER NAS to compensate the strong influence from these components, as well as the plasma shape and position variations.

**3.7.2. Generic description.** The ITER NAS measures gamma radiation from samples activated by fusion neutron flux. Encapsulated samples are transferred between IEs and counting station by the nitrogen driving gas. Tubes of diameter 12.7 mm will be used for the transfer lines of the capsule.

NAS consists of IEs, pneumatic transfer system and the counting station (figure 59) [201]. The counting station includes gas supply, distribution station (Carousel), interstations transfer lines, cooling docks, counting docks with gamma detectors and the ultimate disposal dock. The control system will assure fully automated operations including capsule loading, transmission to and from IE, storage for cooling, gamma activation measurements and transition to disposal dock [202].

Due to the large size and the elongated shape of ITER plasma, multiple positions for the IEs in toroidal section are required for highly reliable measurements. Four IE locations per toroidal section (A, B, C and D in figure 60) are planned for ITER NAS considering the reliability of the measurement and redundancy of the system.

Transfer tubes of the NAS should be bent many times to reach the IEs from the transfer station. To avoid the capsule to get stuck around tube bends, there should be a minimum

<figure><img src="image_71.png" /><figcaption>**Figure 59.** The scheme of neutron activation system for ITER.</figcaption></figure>

bending radius of the tube in designing tube route. All bends of the tube should have a larger radius than this minimum bending radius. Assuming the capsule of outer diameter 8 mm and length 30 mm, and the tube of inner diameter (ID) 9 mm, the minimum bending radius of the tube is about 100 mm. The current design value of the minimum bending radius is 150 mm, with a safety factor of 50% applied.

The port allocation for the NAS is #11 and #18 for the upper port, #11 and #17 for the equatorial port, and #12 and #18 for the lower ports. For points A and B, the IEs will be located inside the port plugs. Other IEs will be installed on the VV wall with the pipelines routed through the lower level ports [203]. In total 12 IEs will be installed.

The transfer system distributes capsules to the appropriate locations such as IE, counting station, or disposal dock. It consists of capsule loader and distribution machine—Carousel. When a capsule is loaded on the Carousel from the loader, the platter inside the Carousel rotates to place capsule to the point connected to the designated place.

The counting station is located outside the tokamak building, where the neutron flux effect on the detectors is negligible. Detectors such as HPGe or NaI will be used to count gamma-rays from the activated samples. The required parameters for the NAS such as neutron fluence will be evaluated from the gamma spectrum considering the location of the IE, sample material and its mass, and irradiation and cooling time.

**3.7.3. Evaluation of measurement accuracy.** The measurement accuracy of the NAS with 12 IEs is estimated using MCNP calculations. The response of each irradiation location is evaluated by changing the location and the profile of the neutron source (see figure 60).

The evaluated result of the neutron source displacement effect (top part of figure 60) shows that the upper port is the best position for the irradiation due to its lowest sensitivity to the plasma position. The induced error due to the vertical displacement can be even lower when it is compensated with the measurement at the divertor position, as long as the IE at the divertor is well characterized during the plasma operation. It is estimated that the induced error from the neutron source displacement can be ~±1% even without compensation from other diagnostics, from the simultaneous measurement from
<figure><img src="image_72.png" /><figcaption>**Figure 60.** Evaluation of the effect of neutron source position and broadening.</figcaption></figure>

<figure><img src="image_73.png" /><figcaption>**Figure 61.** Examples of NAS in-vessel clamps CADs and their prototypes.</figcaption></figure>

the upper and divertor position, when the displacement range is within ±20 cm vertically and radially. The equatorial port position can be used for backup when the data are compensated from other diagnostics.

The effect of neutron source broadening (bottom part of figure 61) on the measurement, which cannot be estimated during the in-vessel calibration, was evaluated. The result also indicates that the upper port is the best position, because it has the lowest effect from the neutron source broadening, and shows good characteristic of depending only on the vertical broadening. It is interesting to note that the equatorial port position shows symmetric measurement with the upper port position. Therefore, the simultaneous measurements from the upper and equatorial port position are expected to provide the total neutron production with the broadening error of ~1% without compensation from other diagnostics, when the profile peaking factor is in the range of 3 < $\alpha$ < 7.

The calculations show that the combination of measurements from the upper port, equatorial port, and divertor region can provide a relatively good evaluation of the total neutron production in the plasma. In spite of the low reliability of the measurement from the inboard midplane position, it is reasonable to keep these IEs, as they are the only ones capable of providing the absolute value of the neutron flux coming to the inboard side.

### 3.7.4. Design of the NAS components for ITER.
Thermal analysis has made significant impact on the design of the NAS front-end components, as all NAS components installed inside the VV shall follow the design guideline SDC-IC (Structural Design Criteria for ITER In-vessel Components), which requires the maximum temperature of the components to be less than about 400 °C. In order to guarantee stable
<figure><img src="image_74.png" /><figcaption>**Figure 62.** Components of NAS IEs in equatorial port plug 11.</figcaption></figure>

temperature of the in-vessel components, these tubes and IEs are clamped to the VV with a maximum pitch of 20 cm. The clamps (figure 61) are particularly designed to maximize heat sink and provide cover from stray plasma radiation where necessary. The material of clamps and IEs is CuCrZn, while the tubes and the boxes, used for attachment to the VV, are stainless steel SS316.

Port plug IEs represent coaxial structures aimed at minimizing dimensions and maximizing cooling features. (figure 62). Most components will be fabricated from SS316L. The front part of the IE is enclosed with the coolant housing, which is connected with the coolant tubing. Two guiding rings are attached on the outside of the coolant housing for a smooth insertion of the IE into the DSM. The feed-throughs will be welded on the closure plate of the port plugs.

Transfer station consists of many moving components such as a servo-motor, linear actuators and many solenoid or gas-driven valves. Pneumatic properties of the transfer system for transferring the capsule are as in table 23:

<table><caption>**Table 23.** Pneumatic properties of the transfer system for transferring the capsule.</caption><tr><td>Pressure of gas in reservoir:</td><td>~8 bars max</td></tr><tr><td>Pressure of driving gas</td><td>1-8 bar</td></tr><tr><td>OD of sample transfer tube</td><td>12.7 mm</td></tr><tr><td>Thickness of sample transfer tube</td><td>1.25 mm</td></tr><tr><td>OD of retrieving gas tube</td><td>12.7 mm</td></tr><tr><td>Thickness of retrieving gas tube</td><td>1.25 mm</td></tr><tr><td>Diameter of capsule</td><td>~8 mm</td></tr><tr><td>Length of capsule</td><td>~20 mm</td></tr></table>

Samples will be transferred to the designated position by the action of a distribution machine—Carousel. A rotating platter inside the Carousel will transfer the sample to the loading position that is connected to the designated IE. When the samples are ready, the valves behind are opened to shoot them to the designated positions. A programmable logic controller will control the operation of the transfer system.

<figure><img src="image_75.png" /><figcaption>**Figure 63.** Radioactivity of samples (a) at the equatorial port irradiation end by 1 s irradiation, (b) at the divertor irradiation end by 100 s irradiation.</figcaption></figure>

Counting station measures gamma-rays from the activated samples. It consists of gamma-ray detector, signal processing electronics such as high voltage supply, preamplifier, amplifier, and multi-channel analyzer, and data analyzing software. Many gamma-ray detectors such as gas chambers, scintillators, and semiconductor detectors are commercially available. Amongst these detectors NaI detectors and HPGe detectors are the most commonly used ones for neutron activation analysis, but other types of detector are also considered. Appropriate detectors will be chosen for the proper operation of ITER NAS considering state of the art.

**3.75. Performance assessment.** The primary role of the NAS is to provide the FW fluence. Considering the plasma shot based operational method of NAS, it is also important to have a shot-integrated measurement ability regardless of the performance of the plasma such as the discharge time and the total neutron yield. NAS is expected to provide the measurement data in the discharge time of 1 s (neutron flux with higher temporal resolution can be measured by time-resolved neutron diagnostics such as NFM)—a few hundred sec, and in the neutron yield of $10^{14}-3.2 \times 10^{20} \text{ s}^{-1}$. Neutron transport and activation calculations using MCNP and FISPACT code showed that NAS can measure the whole neutron yield range thanks to the freedom of selecting the irradiation locations, kinds and mass of samples, cooling time, etc.

In order to provide necessary statistics and avoid saturation, the radioactivity of the irradiated samples should be in the region of $2 \times 10^3-2 \times 10^5 \text{ Bq. Preliminary performance assessment has demonstrated (figure 63) that it is possible to select necessary activation foil material and size for all required values of neutron yields [204].}$

**3.76. Summary.** The ITER NAS that has been briefly presented hereabove is under development by the Korean Domestic Agency of ITER. Despite the challenges driven by ITER's unprecedented thermal, electromagnetic and nuclear loads, high radiation and activation in full-power operation leading to very limited personnel access and highest safety and reliability requirements [6], despite all these aspects the presented NAS design proves to be suitable to satisfy ITER's measurement requirements.
<table><caption>**Table 24.** Measurements and their requirements for a HRNS system at ITER, taken from [149].</caption><thead><tr><th>Measurement</th><th>Parameter</th><th>Parameter range</th><th>Condition</th><th>Time resolution (ms)</th><th>Accuracy (%)</th></tr></thead><tbody><tr><td>Fuel ion ratio in the core</td><td>n<sub>t</sub>/n<sub>d</sub></td><td>0.01–10</td><td>DT plasma if P<sub>fus</sub> >0.5 MW<br>NBI Present if n<sub>t</sub>/n<sub>d</sub> <0.1<br>NBI absent T<sub>i</sub> known</td><td>100</td><td>20</td></tr><tr><td>Ion temperature in the core</td><td>Core T<sub>i</sub></td><td>5–40 keV</td><td>P<sub>fus</sub> >0.5 MW</td><td>100</td><td>20</td></tr></tbody></table>

### 3.8. Neutron spectrometers

Neutron spectroscopy is one of the few methods with an established capability to determine ion density ratios, $n_t/n_d$, in the plasma core. The ITER requirement for the determination of $n_t/n_d$ is an accuracy of 20% within a time window of 100 ms (see table 24). Moreover, this should be accomplished for a fusion power range of approximately three orders of magnitude, from pure deuterium to 50/50 DT plasmas. To achieve these requirements, recent developments in neutron spectroscopy techniques, DAQ and data analysis methods have been included in a conceptual design [149] and should to be further developed in a final design.

The main role of a HRNS system at ITER is to measure the fuel ion ratio, $n_t/n_d$, in the ITER core in DT operations. A secondary role is to contribute to measurements of the fuel ion temperature in the core. In addition, a HRNS could also contribute to measurements of other operational and physics parameters, such as the fusion power, fast ions distribution functions and heating effects and efficiency.

The ability of neutron spectrometry to provide measurements of $n_t/n_d$ has been reported in a number of publications. For the thin-foil techniques, data from JET DTE1 have been analyzed using measurements from the Magnetic Proton Recoil spectrometer, MPR [205, 206]. Moreover, rather recently JET DTE1 data from the TANDEM neutron spectrometer [207–209] were used to make $n_t/n_d$ estimates using analysis methods similar to those used for the MPR. This is of ITER relevance since TANDEM uses a neutron detection technique closely resembling the thin-foil proton recoil (TPR) technique suggested for the HRNS system. For the time of flight (TOF) techniques, in recent JET DT operations it has been shown that $n_t/n_d$ can be estimated in minority T plasmas, D(T) and majority T plasmas, (D)T.

First we give a brief review of the main points of the proposed ITER HRNS system, then we will report on some pertinent recent and ongoing instrumental and method developments.

The case for and a preliminary design of a HRNS at ITER has been made in the article ‘Conceptual design of the high resolution neutron spectrometer for ITER’ [149]. Due to the required performance over a wide range of fusion power, no single instrument has been found to fulfill the requirements. Therefore, the proposed HRNS system is composed of a system of four spectrometers, employing different detection techniques. From previous experience with other projects and implementations of neutron spectrometers in fusion, the paper

concluded that there are several spectroscopic techniques that have a good, proven record for fusion applications and thus should be considered for ITER. The proposed ITER HRNS therefore consists of (i) a set of TPR spectrometers, (ii) a matrix of neutron diamond detectors (NDDs), (iii) a back-scatter TOF (bTOF) spectrometer and (iv) a forward scattering TOF (fTOF) spectrometer; the latter is intended for D and D(T) plasma operations. These four neutron spectrometers can fulfill the measurement requirements for an ITER power range of 0.5–500 MW. The conceptual design of the system is described in some detail in the article, and will only be briefly mentioned here.

The planned location of the HRNS system is in ITER’s equatorial port cell #1, and figure 5 in [149] shows the layout of the complete system. This allows for the use of a single radial LOS into the plasma core. The HRNS shares this Port Cell with some other neutron/gamma systems, namely the RNC and the gamma ray camera.

#### 3.8.1. TPR technique.
The TPR spectrometer system is intended for the highest performance plasmas, in the fusion power range, $P_{fus}$, 100–500 MW. It is based on the measurement of forward scattered protons from elastic (n,p) scattering on hydrogen in a passive, thin CH$_2$ foil. A set of two (radially) segmented Si detectors measure the recoil protons in coincidence in a restricted angular range. The two Si detectors are arranged as a dE-E pair (thin dE, thick E detector) in order to improve proton identification and reduce background. TPR spectrometers are inherently low-efficiency devices, and therefore, in order to boost count rates, the proposed TPR system consists of three identical sets of CH$_2$ converter foil and double detector arrangements placed one after another in a vacuum tight cylinder. Figure 6 in [149] shows a more detailed picture of the TPR system in the HRNS design. The vacuum chamber should maintain a pressure level of about $10^{-4}$ mbar or better, as this is sufficient to avoid detrimental energy loss of the recoil protons in transit from the scattering foil to the detectors.

Silicon detectors are a well-proven detection technique, used extensively in high-energy physics experiments, and also in fusion research. An energy resolution FWHM <2% is achievable, providing a performance that is sufficient for neutron spectroscopy applications. The estimated count rate of the Si detectors at the highest fusion power is about 250 kHz, which when spread out over several segments can be reduced to comfortable levels of a few 10 s kHz per segment. With
suitably fast preamplifiers and similarly fast backend DAQ electronics, processing of signals due to the direct neutron induced events should not pose any serious challenge. However, a more thorough investigation of background events due to scattered neutrons and gammas will be necessary for the final design. An extension of the present, rather thin, radiation shielding around the TPR system is probably necessary.

**3.8.2. NDD.** Diamond detectors for fusion neutrons in the 14 MeV region (DT plasmas) rely primarily on the $^{12}$C(n,α)$^{9}$Be reaction, which produces two charged particles in the final state. The Q value of the reaction is 5.702 MeV and it is thus only suited for neutrons from T and DT plasmas. A matrix of NDDs is proposed as part of the ITER HRNS system, and a picture can be found in [149], in order to provide redundancy in measurements in the higher end of fusion power, thereby providing complementary measurements to the TPR and TOF instruments. It also provides data from an alternative technique, which could be important to understand specific issues associated with the other techniques.

The NDD is to be placed inside the TPR vacuum chamber, behind that system. The resolution of NDDs is about 1% at 14 MeV and allows high resolution 14 MeV neutron spectroscopy measurements. For a single NDD element (radius 2 mm, thickness 0.5 mm) a total count rate of around 100 kHz is estimated for a $P_{\text{fus}} \sim 50$ MW. To handle higher $P_{\text{fus}}$ a reduction of the NDD element size is required to maintain a total count rate of approximately 100 kHz in a single detector. To collect sufficient counting statistics a suitable number of NDD elements in a matrix are therefore needed. Similarly to the Si detectors of the TPR, this can be handled with modern, standard pre-amplifiers and fast, digital DAQ systems.

As they are installed in essentially the same location as the TPR, also the NDD system would benefit from a more extensive radiation shield around the TPR/NDD vacuum chamber.

NDD have been installed in many fusion devices but quite few relevant physics results have so far been published from their use in high-performance DT operations. Therefore, the high-performance DT operations at JET in 2021 should provide much valuable experience in doing spectroscopy on the intended (n,α) peak in the NDD pulse height spectrum. It should be noted that the performance of the NDD devices as spectrometers of DT fusion neutrons has been convincingly shown in for example characterization of DT neutron generators (NGs). For ITER applications, the challenge rather lies in the expected high count rates, operations over extended periods of time, high levels of background etc. It is also important to investigate the effect of the $^{13}$C content in NDD elements for estimating $n_{T}/n_{D}$.

Further developments of the NDD technique for an ITER HRNS system should include, e.g. careful assessment of their performance in high-performance DT operations (JET 2021), tests of detectors, signal electronics and acquisition at the highest rates envisaged at ITER and neutron transport calculations to design an optimal radiation shield for the HRNS TPR/NDD systems.

**3.8.3. TOF technique.** TOF systems for fusion rely on elastic scattering of a fusion neutron on hydrogen in two sets of detectors (proton/proton or deuteron/proton), one giving a start and the other a stop signal for the flight-time measurement. With a known distance between the detectors of the two sets, and relying on two-body kinematics, the neutron velocity (energy) can be determined. Since it is the neutron that is traversing the system and being responsible for both start and stop signals, the in-beam material of the TOF system's start detectors can be substantially thicker than in the TPR case, thereby offering much higher efficiency. TOF systems can therefore offer good performance in plasmas of relatively modest neutron emission rates. A resolution of FWHM = 5.8% is obtained for the TOFOR system at JET for 2.5 MeV neutrons [210]. In recent D(T) experiments at JET, together with simulations, it has been possible to evaluate the performance of the TOFOR spectrometer in a range of minority T operations. It has then been estimated that DD neutron spectroscopy can be performed with TOFOR for a tri-tium concentration of up to approximately 20%, provided that the count rate capability for TOFOR is not exceeded, corresponding roughly to a signal to background ratio (SBR) for the DD peak on the order of 1. Such performance is possibly sufficient for the intended use of TOF systems at ITER, but further investigations should be made.

Two TOF systems are envisaged for the ITER HRNS: a conventional forward scattering system mainly intended for D and D(T) operations, and a back-scatter TOF system for the lower power ranges in DT operations. To reduce the volume of the full HRNS system, the two TOF systems are intertwined with each other, as shown in figure 8 in [149]. In order to extend the operational range of the back TOF (bTOF) it is proposed to install a flexible collimator in front of the system, in order to reduce the neutron flux on the in-beam detectors in high-power operations.

Extensive experience with the conventional forward TOF (fTOF) technique has been obtained at JET in the last decade through the TOFOR system. The design and operational experience from TOFOR has shown that TOF systems for fusion can be reliably modeled using state-of-the-art nuclear simulation codes like GEANT4 [211] and MCNP [212]. Experiments show that important operational parameters, like the influence of scattered neutrons and gamma background, the level of random events in the data (in high power operations mainly due to detection of non-correlated neutrons in the detectors of the system) and the details of individual detector response are well understood using modern simulation tools. This builds confidence in designing both the fTOF and bTOF systems for ITER. With the implementation of modern digital DAQ electronics, it is possible to somewhat reduce the influence of random coincidences through kinematic cuts [213, 214].

The backscatter TOF (bTOF) builds on the successful implementation of fTOF in fusion, notably the TOFOR spectrometer at JET.

The bTOF technique relies on single elastic scattering of a neutron on a deuterium nucleus in the start detector and subsequently detecting the same neutron in the stop detector,
placed at a well-known distance in the backwards direction. This arrangement has a two-fold advantage for detection of neutrons in the 14 MeV region: first, it gives a fairly low energy to the backscattered neutron, reducing its velocity, and thereby increasing the flight time of the neutron before it reaches the stop detector; second, it gives a large energy to the recoil deuteron, thereby improving the possibility to accurately measure the timing and energy of that particle. Deuterated scintillators, in either liquid or solid form, are envisaged as start detectors. Thus, a bTOF system can be expected to exhibit many performance parameters similar or better compared to a conventional fTOF: the size and flight times would be similar, and thereby also the achievable time resolution. The high energy of the recoil deuteron would improve the possibilities to perform accurate timing and kinematical cuts. Performance studies using Monte-Carlo (MC) methods of bTOF systems are presented in [215].

Further development of the fTOF system should include assessment of the operational limits in D(T) for performing $n_t/n_d$ measurements. Another line of investigation could focus on improving the performance of the start detectors. The solid case for a bTOF system notwithstanding, development of a pilot bTOF system and tests in a 14 MeV facility would be important to gain experimental experience with the technique [216]. In addition, the space-saving arrangement of intertwined TOF systems should be simulated in more detail to assess the possible interference the systems might cause to each other. Finally, a detailed, overall assessment of the background radiation environment for the TOF systems should be made, including both the background from the system's own beam dump and background entering the system area from outside.

**3.8.4. Recent developments.** In the following we will report on some recent relevant developments with significance to the HRNS ITER mission. This includes: (i) method development of density measurements using the TOF technique at JET; (ii) recent work on the TPR technique; (iii) $T_i$ measurements with fTOF in thermal (ohmic heating only) plasmas at JET.

**3.8.4.1. Density measurements using the TOF technique in D(T) and (D)T at JET.** The operations at JET with significant amounts of tritium in 2020–2021, ending with a high-performance campaign in full 50:50 DT, offers ample opportunities to develop the $n_t/n_d$ method with the available neutron spectrometers at JET. Some results have so far been presented for the fTOF technique, at JET represented by the TOFOR spectrometer [217]. The fTOF method has so far been applied to D(T) and (D)T plasmas with trace amounts of up to a few percent of either tritium in a D dominated plasma, D(T), or deuterium in a T dominated plasma, (D)T. The results confirm previous assessments and are consistent, within uncertainties and other experimental circumstances, with other JET diagnostic systems [217]. An example of TOFOR time resolved results of $n_t/(n_t + n_d)$ in a (D)T plasma, JET discharge #99093, is shown in figure 64.

<figure><img src="image_76.png" /><figcaption>**Figure 64.** Time evolution of $n_t/(n_t + n_d)$ for JET discharge #99093 as derived from TOFOR (points with error bars) and gas analyzer KT5P (dashed line with shadowed uncertainty band). Reproduced from [217]. CC BY 4.0.</figcaption></figure>

<figure><img src="image_77.png" /><figcaption>**Figure 65.** TOFOR time-of-flight spectrum integrated in a time interval of 2 s (JET time 11–13 s), used to estimate $n_t/(n_t + n_d)$. Model components of NBI (DT) and NBI (TT) and scattered neutron are used for the fit to measurement data. Reproduced from [217]. CC BY 4.0.</figcaption></figure>

The result is for a neutron production dominated by tritium neutral beam ions reacting with the bulk plasma, mainly consisting of T ions. A roughly constant beam T ion power of 20 MW was injected into the plasma between 10 and 18 s resulting in a fairly constant $n_t/(n_t + n_d)$ result for TOFOR. Comparing the TOFOR result with the gas analyzer KT5P [218, 219], an optical diagnostic used in combination with a Penning gauge located below the divertor, gives a consistent picture, although small systematic differences have been observed for other pulses. Some differences are to be expected, considering the different locations of the active measurements of the two diagnostics.

Figure 65 shows the TOF spectrum for the same plasma discharge after subtraction of the random background. Features in the spectrum can be associated with neutron emission from DT and TT reactions, allowing for the detailed time resolved analysis presented here. The data from JET should make it possible to assess the limit of the $n_t/n_d$ method using fTOF
which in previous simulation studies has been estimated at $n_t/n_d < 0.2$ [220].

### 3.8.4.2. Recent work on the TPR technique.
Proton range telescopes based on silicon detectors is a well-established technique in nuclear physics and applications. However, apart from the now mothballed TANDEM instrument at JET, the technique has not been used at fusion devices, presumably due to the rather scarce opportunity of DT experiments. Here we can report on two lines of recent progress in the use of the TPR technique in fusion: first, we have tested the $n_t/n_d$ method on data from the JET TANDEM spectrometer with good results; second, we have investigated a pilot set-up similar to that proposed for the ITER HRNS TPR implementation [221].

Analysis of TPR (TANDEM) data from the JET DTE1 campaign in 1997 [221] has corroborated the previously published results from the MPR spectrometer [215]. These A dE-E setup of two segmented silicon detectors has been tested at a 14 MeV neutron source at Lund University, Sweden [221]. The detectors were of somewhat poor quality, and only a few of the detectors radial segments could be equipped with signal processing and DAQ electronics. Also, further improvements in pre-amplifier electronics are certainly required. Still, the results corroborate the main conclusions from earlier simulation studies that coincidence dE-E measurements will be able to obtain the spectroscopic energy resolution needed together with background elimination to estimate $n_t/n_d$ for ITER scenarios investigated in [149].

### 3.8.4.3. Ti measurements with fTOF in thermal (ohmic heating only) plasmas at JET.
Measurement of the ion temperature is a supplementary function of the HRNS system. The measurement relies on the identification and analysis of the pure thermal neutron emission from a D, T or DT plasma. However, in most present (pre ITER) fusion experiments, the thermal neutron emission is only a small fraction of the total emission, which is normally dominated by emission from fusion reactions between fuel ions from the heating populations and the thermal bulk. In a few cases when only ohmic heating was applied has the Ti been possible to determine. This was for example reported in [222] using the MPR in the DTE1 experiment at JET, where an ion temperature of about 3 keV was reported for ohmic plasmas. More recently, using the fTOF technique, thermal temperatures of ohmic plasmas at JET in D operations were reported [223]. The reported values were in the range $T_i = 2.5-3.2$ keV and quite consistent with the JET crystal x-ray spectrometer and with the dependence ofreactivity on $T_i$.

### 3.8.5. Conclusion.
The proposed (enabled) HRNS system for ITER is based on well-established neutron spectrometry techniques. The required performance of the included subsystems, in terms of energy resolution, count rate etc, has for the most part already been achieved in systems installed at JET. The ability of neutron spectroscopy to perform its main ($n_t/n_d$) and secondary (Ti) functions has been established in a number

<figure><img src="image_78.png" /><figcaption>**Figure 66.** Compact neutron generator VNIIA NG-24.</figcaption></figure>

of published studies using data from D, T and DT operations at JET. Some recent developments in JET DTE2 and in instrumental techniques corroborate these earlier studies and thus strengthens the case for installing the proposed HRNS system on ITER.

### 3.9. In-situ neutron diagnostic calibration

ITER neutron detectors need to be absolutely calibrated in their ultimate positions in order to eliminate systematic errors associated with scattering features of the machine. In-Situ calibrations are routinely performed at nuclear installations where high precision of neutron yield measurements are required. ITER project requirements (PRs) state the value of 10% as the maximum allowable neutron yield (and thus fusion power) measurement uncertainty [224].

To perform in-situ calibration, a neutron source (permanent radioactive source or a NG) is introduced at different positions inside the vessel or moved around the torus to form a quasi-ring-source. The aim is to simulate plasma neutron emission as realistically as possible. In ITER, only the most sensitive detectors can be calibrated with compact neutron sources in a reasonable time. These diagnostics are:

* NFMs (in equatorial ports #1 and #7),

* divertor NFMs,

* micro-FCs and NAS.

The less sensitive detectors will be cross-calibrated during plasma operation. Compact NG with yields of up to $10^{11}$ and $10^9$ n s$^{-1}$ od DT and DD neutrons, respectively, will be employed at ITER. This decision is driven by performance (high neutron yield) and safety with respect to a $^{252}$Cf source.

In order to minimize the calibration time, the most powerful commercially available compact NG VNIIA NG-24 (figure 66) has been chosen.
<figure><img src="image_79.png" /><figcaption>**Figure 67.** Conceptual design of the IVNC-1 NG manipulator.</figcaption></figure>

Three calibration campaigns are foreseen for ITER:

1. In-vessel neutron calibration campaign 1 (IVNC1) with DD NG with strength of $10^9 \text{ n s}^{-1}$, just before the pre-fusion power operation 1 (PFPO-1) where neutron yield measurements are required by the Project. The DD NG will be deployed in limited areas in front of ports 18, 1 and 2.

2. IVNC-2 campaign is required before the start of PFPO-2 due ITER configuration change. This recalibration will be done with DD NG aiming to minimize machine activation in the process. Unlike IVNC-1, a full toroidal scan is planned for the IVNC-2 campaign.

3. The IVNC-3 campaign will be performed with 14 MeV DT NG just before the full power operation (FPO) Phase.

Each of the three campaigns is scheduled to last for 2 weeks of net NG in-vessel operation time, preceded and followed by preparation, assembly and disassembly periods. The IVNC-1 will be performed right after the partial installation of the blanket modules; therefore human access will be possible for the preparation, assembly and disassembly procedures. Conceptual design of the IVNC-1 NG manipulator is shown in figure 67.

The IVNC-2 and 3 will be performed after the FW modules have been exposed to plasma. Human access will be prohibited in the VV due to radiation and beryllium hazards. These campaigns will be performed with the use of a remote handling tool called the agile remote transporter (ART). The same tool that will be used for blanket modules replacement and optical diagnostic calibrations. At the current stage of the project, the ART is in the conceptual design phase (figure 68).

Studies have been launched with the following purposes:

* development of the high-accuracy NG yield monitoring system for DT and particularly for DD NGs;

* minimization of the NG yield anisotropy;

* development of the remote-handling capabilities of the NG unit with the attached yield monitoring system.

* Optimization of the calibration campaigns.

<figure><img src="image_80.png" /><figcaption>**Figure 68.** Conceptual design of the agile remote transporter (ART).</figcaption></figure>

### 3.10. Alternative diagnostic techniques to fuel products measurements

In addition to neutron diagnostic systems, several other diagnostic systems considered for ITER can provide information on some of the plasma physical parameters listed in table 12. Systems discussed in this section include neutral particle analyzers (NPA), recombination CX spectroscopy (CXRS), and collective TS (CTS). This section is not intended to provide a complete assessment of the capabilities of these systems. Here we will only give a brief summary of the techniques and some of their problems and limitations. The full description of these diagnostic techniques is presented in other parts of the chapter dedicated to ITER-specific diagnostics.

#### 3.10.1. CXRS.
It is anticipated that CXRS diagnostics will provide information on plasma parameters such as plasma rotation, ion temperature profile, impurity density profile and alpha particle density profile. Active CXRS spectroscopy requires neutral beams. In the case of ITER, two types of such beams are envisaged, the first being a DNB providing deuterons with an energy of 100 keV and a second neutral heating beam (HNB) providing deuterons with an energy of 1 MeV.

Probing the plasma with the DNB beam is believed to be insufficient for high-quality measurements of plasma parameters in the core. On the other hand, measurements with high-energy HNB, due to better penetration of the beam, can sample the distribution functions of the confined $\alpha$ particles up to significant energies.

For $\alpha$ particle measurements, it seems that CXRS for DNB will be able to provide information on the distribution of alpha particles slowed down to an energy of 0.4 MeV, and when using HNB to an energy of 0.6 MeV. The limitation of this diagnostics may be the required small measurement angle.
Moreover, the presence of the high-energy ³He population due to RF heating may further complicate these measurements.

Results for plasma rotation and ion temperature profile will depend on plasma impurities measurements. Differences between measurements for plasma fuel ions by neutron diagnostics and CXRS may occur for fast transients, have been observed for fast transients, for example at the onset of beam heating.

### 3.10.2. CTS.
CTS diagnostics probes the ion VDF at the intersection of the probing and receiving 'beams' and in the direction of the difference between received and the emitted radiation (the detailed description could be found in subchapter 7, section 7.6). The CTS consists of one or more combinations of probe beams and receiving antennas in the microwave frequency range of 100 GHz, e.g. for the ASDEX tokamak, while the proposed system for ITER will be based on 60 GHz.

An important feature of the CTS technique is that the signal strength is proportional to the ion charge in the square $(Z_i^2)$, which makes CTS a good candidate for alpha particle measurements $(Z_i^2 = 4)$, even when such ions have a low concentrations. However, the disadvantage of this system will be the low relative efficiency of fuel ion measurements. $(Z_i^2 = 1)$.

CTS is therefore a good candidate for diagnostics that gives information on the distribution function of fast ion, in particular for the fast alpha particle population.

### 3.10.3. NPA.
NPA systems have been installed in many major fusion devices around the world and have a long history of providing experimental data. The rate of neutralized ions reaching the NPA system is a function of essentially the following factors (integrated at all points within the NPA line of side (LoS)):

* the local energy distribution of ions with velocities in the direction of NPA (it is usually considered that the capture of electrons does not significantly change the speed of the ion),

* the probability of local neutralization,

* plasma transparency for the outgoing atom.

Using measurements and modeling of plasma ion density (and impurities), temperature profiles, as well as data from atomic physics, it is possible to compare the measured streams of neutralized ions with the calculated ones and thus draw conclusions about the state of fuel ions in the plasma. To some extent, the interpretation process is similar to that of plasma neutron emission analysis, with the significant difference that for neutrons plasma transparency is equal to one.

Due to the fairly complex plasma modeling required for the interpretation of NPA data, this diagnostic seems to provide mainly qualitative data, where differences in the neutral particle flux for different plasma conditions are taken as an indication of presence or absence of tails of high energy particle in the plasma. This extensive and complex modeling so far seems to preclude the NPA from delivering more quantified results on a routine basis on every discharge, although

quite detailed results are certainly obtained when special analytical efforts are made.

Typically, two NPA systems gathering integrated data from the same LoS are considered. One Low Energy NPA (LENPA) system covering an energy range of 10–200 keV with the ability to simultaneously measure neutralized hydrogen (H), deuterium (D) and tritium (T) ions. While the second high energy NPA (HENPA) should measure $\alpha$ particles with energies of 0.1–4 MeV, i.e. over their entire slowing-down energy range. The LENPA low energy system should provide information on the energy distribution functions of the neutralized fuel ions (deuterium and tritium) as well as protons and the fuel density ratio $n_d/n_t$ in the plasma core. In contrast, the main purpose of the HENPA system is to provide information on the number density and energy distribution function of the confined $\alpha$ particles. Such a tandem system was planned for ITER. However, both systems have their drawbacks. In the case of the LENPA low energy system, there is a problem of data analysis when we are dealing with very flat ionic plasma temperature profiles and when a clear presence of hydrogen isotope composition in the plasma core cannot be expected. In addition, this system is limited by poor penetration of neutral particles from the core into the analyzer.

In turn, for the high-energy system (HENPA), a specific problem is that energetic $\alpha$ particles with energy $E_\alpha$ are indistinguishable from energetic deuterons with energy $E_d = E_\alpha/2$; it means that deuterons of the NB beam with an energy of 1 MeV limit the usefulness of HENPA for $E_\alpha < 2$ MeV. It may also be a problem using this system to measure high energy deuteron populations due to the heating of the ion cyclotron resonance heating (ICRH). In addition, the neutralization of $\alpha$ particles depends on the impurity ions (such as $Be^{2+}$ and $C^{4+}$) that act as electron donors, and the density profiles of these impurities must be known (or modeled) with high accuracy in order to correctly interpret NPA measurements.

The limitation of this system seems to be the poor ability to distinguish between ions of different kinds, so that in practice only the population of fast alpha particles can be measured. Preliminary evaluations show that the measurement requirements for ITER can be met precisely for alpha particles.

## 3.11. Summary and conclusions

The ITER FP diagnostics provide key information on the ion populations in burning plasmas. Several measurement techniques have been considered aiming to measure among:

(a) the total neutron emission rate, i.e. fusion power; (FP is the only diagnostic system able to measure directly the fusion power/ total neutron emission)

(b) neutron emission profile from different plasma/magnetic configuration/scenarios, as well the spatial distribution of ions populations, confined alpha and lost alphas;

(c) the energy features: thermal and fast energies distributions of the neutrons and ions populations, confined alpha and lost alphas;

(d) accuracy 10%: particular attention is given also to the development of the neutron calibration strategy whose
main scope is to achieve the required accuracy of 10% for
the measurement of fusion power.

ITER is now advancing through the construction phase.
Most of the FP systems are under FDRs and some FP are
already in the manufacturing phase.

This paper provides an overview of the FPs diagnostic sys-
tems and the associated engineering issues.

4. Progress in ITER diagnostics, active spectroscopy

S. Zoletnik¹⁶, O. Asztalos¹⁶,¹⁷, M. Bandyopadhy¹⁸,

P.Bharathi¹⁸, M. de Bock², O. Ford¹⁹,

M. G. von Hellermann²⁰, D. W. Johnson²¹, J. Ko⁸,

S. Menmuir²², Ph. Mertens²⁰, A.H. Nielsen²³, G.I. Pokol¹⁶,¹⁷,

S.V. Serov², M.J. Singh³, S.N. Tugarinov²⁴ and G. L. Vyas¹⁸

4.1. *Active spectroscopy diagnostics*

Charge exchange (CX) between the beam atoms and the
plasma ions results in excited plasma ions and the spectral
broadening and shift of their radiation represents the ion tem-
perature and velocity. If the CX occurs on impurity ions the
emission intensity can be used to calculate the impurity ion
density. These measurements are called CXRS. Line radi-
ation by the beam atoms contains information on the exciting
plasma electron density (beam emission spectroscopy, BES),
and the Stark split of the spectra by the $v \times B$ electric field
can be used for calculating the magnetic field value and/or
direction (MSE). On present-day fusion experiments CXRS is
one of the standard methods for the measurement of ion tem-
perature, flow velocity and impurity concentrations, while the
plasma current distribution is measured by MSE. Density pro-
files and fluctuations are measured by BES [225]. The beams
can be either Heating Neutral Beams (HNBs) or specialized
DNB of the plasma species, or alkali atoms for the plasma edge
measurement.

In all cases the key elements of the setup are the atomic
beam with sufficient penetration depth and particle flux, and a
high etendue optical observation system together with a suit-
able detection system. The location of the observation sys-
tem is also of high importance due to two factors: first the
extent of the intersection area between the beam and observa-
tion channels should be limited, to keep the spatial resolution
at a reasonable value. This would call for near-perpendicular
beam-observation angle and view along the magnetic field
lines. The second factor is spectral discrimination between the
passive plasma radiation and active emission either from the
beam or the plasma. This separation is possible for the beam
emission measurement by utilizing the Doppler shift result-
ing from beam atom velocity and requires sufficiently non-
perpendicular observation.

On ITER active spectroscopy diagnostics are being
developed on both the heating beams and a DNB. However, the
situation will be quite different from present-day devices. The
first problem is the large size and high density of the plasma
which renders alkali metal (Li, Na) beams unsuitable for edge

measurement due to their shallow penetration. Also, heating
beams must operate around MeV energy to deposit power
deep enough in the plasma. The CX cross-sections typically
have a maximum around 30–100 keV beam energy therefore
these high energy beams have too low CX cross-section and
cannot be used for CXRS. The high beam velocity is a problem
for BES as well, due to the long lifetime of the excited states,
causing spatial smearing. Due to these problems a decision
has been taken to install a DNB with ~100 keV energy for
CXRS on the He ash and other impurities. For MSE both the
HNB and DNB are considered.

ITER also provides limited choice on the observation dir-
ections as all observation systems must be located in a limited
number of ports. Even more problematic is the strong radi-
ation background and expected damage to optical components
by plasma sputtering and deposition. These problems are not
specific to active spectroscopy diagnostics therefore discussed
in subchapter 4 of this paper.

The planned active spectroscopy diagnostics have been out-
lined and their difficulties listed in Donné et al [4]: MSE
is planned on the heating beams for the plasma current pro-
file (safety factor) measurement, while multiple CXRS obser-
vation setups are planned for a dedicated diagnostic hydro-
gen beam. These latter have multiple measurement aims: He
and other impurity concentrations, ion temperature, flow velo-
cities. Additionally, to these primary roles the CXRS setups
might also contribute to other measurements as well. The
plans since the publication of [4] remain basically unchanged,
but problems and possibilities have been understood and
the designs have made considerable progress. The following
sections describe the progress made in key areas of the ITER
active spectroscopy diagnostics since the publication of [4].

4.1.1. MSE. It was recognized early that the traditional MSE line polarization (MSE-LP) technique [226] would be difficult to adapt to ITER, due to the sensitivity of these measurements to spurious effects on polarization from even very thin mirror coatings. Due to the time required for effective mirror cleaning (see subchapter 9) of this paper and the nonuniformity of the cleaning, such coatings will accumulate in nonuniform spatial distributions between cleaning campaigns. Calibration schemes to compensate for the impact of such coatings do not exist.

An alternative approach uses high-resolution fitting of the
line splitting and intensity ratios (MSE-LS) in the Stark spec-
trum of $D_\alpha$ to make a precision measurement of the local value
of the motional electric field magnitude at many positions
along the beam trajectory through the ITER plasma [227].
The initial design concept for MSE on ITER featured both LP
and LS techniques. There have been demonstrations of MSE-
LS and comparisons of both techniques on several devices
(JET [228], ASDEX-Upgrade [229], DIII-D [230, 231], C-
Mod [232], MST [233, 234] and most recently KSTAR [235,
236]) that measured that line splitting data are consistent with
expectations based on equilibria constrained by conventional
MSE-LP data. These studies feature only a few spatial pos-
itions, and spectra in which the line widths are comparable
to the line splitting. (On ITER there is better separation, with width ~0.5 × splitting).

The ITER MSE diagnostic design strategy called for a decision point on the viability of both techniques. Leading up to the decision, there was effort to identify a means to calibrate MSE-LP measurements, to assess the use of MSE-LS data to constrain equilibrium reconstructions, and to evaluate sources of systematic error in the MSE-LS approach.

MSE-LP calibration typically relies on acquiring data with ‘beam into gas’, with varying magnetic field configurations. This procedure is not permitted on ITER, hence ‘built-in’ calibration schemes were considered. At the position of the shutter near the FW, a rotating polarizer or a localized plasma source would be needed. In the latter case, the polarization direction of the Zeeman $\pi$-line would be analyzed with Fabry-Perot filters. Both concepts involved front-end components with high reliability risk in the harsh FW environment and were hence rejected.

There were two studies assessing the use of MSE-LS data to constrain equilibrium reconstructions. The first used the ESC code in the context of ITER parameters to simulate the equilibrium reconstruction variance when using combinations of MSE-LP and LS data in the determination of $q(r)$ and $p(r)$ [237]. A second study investigated the EFIT reconstruction code modified to implement either simulated or experimental MSE-LS data on DIII-D [238]. The studies indicate that at least 20 spatial positions are needed with $\delta E/E < 1\%$ in order to achieve 5% error on $q(r)$ with MSE-LS alone, and that MSE-LS alone is a better constraint on $p(r)$ than MSE-LP alone.

An ITER MSE simulation code was used to address systematic errors in the interpretation of ITER MSE-LS data. Included were errors in the beam energies, aperture locations, and beam and sightline shifts and tilts. The plan was to look at HNB1 and HNB2, as well as the DNB. Modeling was done for all of these. This study estimated the accuracies required in the above parameters to achieve $\delta E/E < 1\%$. Beam energies need to be known to <1%, aperture and object locations to <10 mm, and sightline and beam angles to <2 mrad. Concepts for obtaining and cross-checking this data exist, but design details have yet to be developed.

MSE-LP systems are considered essential diagnostics on magnetic fusion devices. An MSE-LS system with enough spatial coverage to be able to constrain an equilibrium reconstruction does not yet exist. However, experimental data from many devices and detailed simulations indicate that such a system can meet ITER’s measurement requirements, provided the geometry and beam energy can be accurately known. Hence, it was decided to proceed with MSE-LS alone, with a design featuring adequate spatial resolution [239].

Design of the ITER MSE-LS system is progressing, with present effort focused on integration of port plug components. The above decision eases constraints on the design. For example, the viewing apertures can be smaller, simplifying integration of viewing optics and shutters. The back-end design is also simpler, since fewer fiber bundles are needed and polychromators and polarimeters are eliminated. The beams are viewed with independent optical systems located in two

equatorial port plugs (EPPs). Both are equipped with mirror cleaning systems. The total number of spatial positions is expected to be about 30.

Meanwhile, researchers continue to develop the MSE-LS technique. One example is recent data from KSTAR, described in the next section.

### 4.1.2. Example of MSE-LS data on KSTAR.
The KSTAR tokamak has been operating the conventional PEM MSE polarimeter, MSE-LP, system since 2015 [240, 241]. Along with this, the feasibility of the MSE-LS approach has been also investigated based on the spectroscopic measurements. These activities have been motivated by the need for a test-bed for ITER MSE research and development [235, 236]. It is noted that the MSE spectrum measurements at the KSTAR tokamak have favorable conditions in that the viewing angles and the beam energies are such that the full energy component is well separated from the rest of the beam emission components, which means not only the adjacent components like the half energy MSE multiplets, but also the background-like components such as thermal and fast-ion deuterium Balmer-$\alpha$ (FIDA) emissions have little effect on the spectral fitting.

Figure 69 shows an example of the MSE spectra with the fitting on them to calculate the magnitude of the total magnetic field at two time slices of a plasma discharge—one with a low plasma density ($1.8 \times 10^{19} \text{ m}^{-3}$) and the other with a medium density ($3.5 \times 10^{19} \text{ m}^{-3}$). The spectrum measurements are made through three channels—core, mid-minor radius, and edge—out of 25 full MSE channels, the rest being connected to the MSE-LP polarimeter, so that both LS and LP measurements are simultaneously possible. A multi-Gaussian fit model on the full energy component takes into account the channel-dependent asymmetry around the $\sigma 0$ component, clearly observed in the measured spectra in figure 69. The background that includes the FIDA component is assumed to be linear in wavelength. The Stark splitting and the relative intensities among the MSE components are treated as free parameters. The fitting works reasonably well with both low and medium plasma densities.

The vertical magnetic field can be inferred from the Stark splitting with a fixed toroidal field, and this is compared with that from the LP method. This comparison is given in figure 70 where the radial profiles of the vertical field measured from the MSE-LP polarimetry are plotted along with those from the spectral fit method illustrated in figure 69 for three MSE channels (Chs 2, 15, and 23). The profiles from two different time points correspond to the two time slices given in figure 69 (low and medium densities). It is shown that the vertical fields inferred from the LS approach are reasonably aligned with these LP-base profiles. The comparison using two profiles also implies that with slight offsets, the vertical fields obtained from the MSE-LS exhibit similar sensitivities as those from the MSE-LP over two different densities.

Several issues still remain regarding the spectral MSE research in KSTAR. The spectral fit procedure would be considerably complicated in the cases with multiple ion-source
<figure><img src="image_81.png" /><figcaption>**Figure 69.** Example of the MSE-LS approach in KSTAR—the multi-Gaussian fits for the full energy components of the beam emission at two different time points (integrated over 0.5 sec)—low (left column) and medium (right column) densities—through three MSE lines of sight (from top to bottom plots). Chs 2, 15 and 23 are the core, mid-minor radius, and edge channels, respectively.</figcaption></figure>

<figure><img src="image_82.png" /><figcaption>**Figure 70.** Comparison of the MSE-LP (polarimeter) and MSE-LS (spectrometer) approaches in the measurement of the vertical field at two different densities. The vertical fields from channels 2, 15, and 23 are from the spectral fits given in figure.</figcaption></figure>

injections due to the spectral overlaps. The sensitivity of the spectral approach also needs to be tested further for discharges where non-monotonic pitch angle, thereby $q(r)$, profiles are expected such as the cases with internal transport barriers or off-axis current drive. Such topics are being investigated at KSTAR associated with the conventional and routine MSE-LP measurements.

### 4.1.3. CXRS on ITER.
CXRS has been considered already in the first ITER design [3] as a possible diagnostic for impurity concentrations, thermalized Helium ash, poloidal rotation measurement using a dedicated diagnostic beam. First feasibility studies for ITER in [3] showed that local measurements of light impurity densities, ion temperatures, and rotation profiles will be possible only outside of $r/a = 0.5$ with a 0.1 s time resolution for the highest density plasmas considered. In Tugarinov et al [242] it was shown that optimized viewing geometry and spectroscopic equipment will allow increasing the SNR significantly. The feasibility of CXRS measurements

was demonstrated from the edge to $r/a = 0.3$ for the highest plasma density $n_e = 1.4 \times 10^{20} \text{ m}^{-3}$. Further development for the CXRS diagnostic for ITER, including design optimization and modeling, is described in [243–247].

According to the ITER requirements [4] the CXRS diagnostic should measure ion temperature radial profiles, impurity density profiles, and rotation velocity profiles with a 100 ms time resolution and a 10%–30% accuracy. The spatial resolution should be 20 mm for the plasma boundary ($r/a > 0.85$) and 66 mm for the central part of the plasma column ($0 \le r/a \le 0.85$). Because of the modulation of the DNB, the time resolution for the ITER CXRS diagnostic is relaxed to 200 ms when the DNB is operated in ‘burst’ mode (3 s at 5 Hz with a 20 s cooling down period) and to 1.5 s when the DNB is operated in ‘control mode’ (100 ms DNB pulse every 1.5 s).

Three CXRS optical views are foreseen [4]; one for the core and two for the edge/pedestal region. This basic layout has been kept and the progress on them is described below.

### 4.1.4. Progress in ITER core CXRS measurement setup.
The optical layout of the core CXRS diagnostic is now mature, at the technical level of a PDR. It comprises the following features (see figure 71 for the optical layout):

In the port plug:

- a front mirror unit (FMU) which protects and contains the entrance aperture, the first mirror M1 and its cleaning system;

- a set of two pairs of mirrors to transport the light to the vacuum windows assembly (VWA). They are arranged in the form of two doglegs to allow for appropriate neutron shielding;

- a separate embedded unit which contains the fast actively cooled frictionless shutter and its actuator;

- a calibration system to monitor the degradation of M1 by illuminating the back of the shutter blade which, in normally closed position, is located at the entrance aperture (cf section on Calibration).
<figure><img src="image_83.png" /><figcaption>**Figure 71.** Core CXRS optical layout and general setup in UPP03, interspace and port cell.</figcaption></figure>

Ex-vessel components:

- a telescope in the interspace, a ‘dogleg’ in the bioshield and a couple of lenses in the port cell to feed the long glass fiber bundle to the spectrometers in the tritium building;

- ancillaries to the port plug systems, like a beam finder to check the line-of-sight in absence of light from the torus (no plasma operation, closed shutter) and the pneumatic helium drive for the shutter.

A full account of the current status is given in [248, 249]. Both references present an overview of the system.

### 4.1.5. Optical performance.
More than in the case of CXRS in the edge plasma, the biggest challenge of CXRS applied to the core plasma is the detection of signals which become radially weaker to the plasma center with attenuation of the observed DNB. Accordingly, the expected photon yield of the whole diagnostic requires a maximized etendue and transmission $T$. The latter reaches about 40% for the plug components including the vacuum window assembly, 70% from the window to the fiber, 82% for the fiber and 35% for a typical spectrometer. For the complete system, the estimated $T$ is thus slightly in excess of 8%.

Similarly, the image is sharp enough for the requirements on spatial resolution to be fulfilled: 67 mm wide channels in the radial direction for the range $r/a \in [0,0.6]$ with a possible extension to $r/a = 0.7$, 250 mm in focal depth, within which the specified $a/30$ resolution can be realized. A radial overlap with the edge CXRS system is achieved. A minimal modulation transfer function (MTF > 0.6) and a maximum spot size at the coupling to the transfer fiber were specified. The risk of vignetting through deformations of the port plug and displacement of the optical beam at the plug output window is minimized by adding a margin >10 mm to the free window aperture. In connection to this, the optical design was selected to be as robust as can be against distortions of the

plug optics under loads of all types. It covers the spectral ranges:

- $\lambda$460 nm band (He, Be)—around 468 nm which enables observation of He II, Be IV;

- $\lambda$520 nm band (C, Ne, Ar, Kr)—around 527 nm, for Ne IV, Be II, Ne X and Ar XVII/XVI; and

- $\lambda$656 nm band (Balmer $\alpha$) with $H_\alpha$, $D_\alpha$, $T_\alpha$ at 656 nm.

The system is thus designed for the visible spectral range with the blue region, down to about 450 nm, the most difficult to achieve but also the most important one.

### 4.2. Other main drivers

A safety requirement and an important design driver of the diagnostic is the shielding efficiency of the in-vessel optical layout against neutrons. It is optimized by the second mirror ‘dogleg’ at the back, close to the windows assembly at the plug closure plate. This allows to protect the magnetic coils and stay in the interspace below the maximum tolerable shutdown dose rate.

For investment protection and a long measurement phase without intervention, the FMU ducts are designed to keep the front mirror as far as possible from the entrance aperture, which was reduced to ø35 mm. The FMU is designed to accommodate the insertion of a shutter blade which is fast enough (0.7 s opening or closing time) to follow the main modulation of the DNB (duty cycle: 3 s on/20 s off). Furthermore, a cleaning device is installed for the first mirror to remove deposits by impurities which would have reached the reflecting surface (see section on mirrors).

The remote handling criteria for maintainability are attained by designing different embedded units as recommended by the Port Integrator. Each of them can be extracted and exchanged upon need. Note that no scheduled maintenance is foreseen, only inspection of the windows and of ex-vessel
<figure><img src="image_84.png" /><figcaption>**Figure 72.** Shutter embedded unit of the core CXRS system.</figcaption></figure>

components, regularly, plus monitoring of the optics transmission in the plug by means of the calibration system.

### 4.2.1. Hardware selection and overall performance.
The first mirror, made of rhodium [250], has a reflectance $R \ge 73\%$. It resists corrosion and sputtering by impinging particles from the main plasma (up to $10^{18} m^{-2}s^{-1}$) or ions in the RF cleaning discharge ($\le 0.2 \text{ mA cm}^{-2}$). A thin monocrystalline Rh plate seems preferable for a more uniform physical sputtering. See subchapter 9 on mirrors for more details.

The shutter is frictionless to avoid seizing and driven by a pneumatic actuator which requires only two pipes for the feedthrough at the plug closure plate. It is a compact embedded unit in one piece [251], see figure 72. The very first version, which had two arms and was less compact, was successfully tested to $10^6$ cycles. It gives confidence about the ruggedness of this piece of equipment since the lifetime expected for ITER amounts to about $6.6 \times 10^5$ cycles.

The fibers in the bundle are defined with a numerical aperture $NA = 0.22$ and a core of ø 600 µm. Unlike the edge CXRS system they start in the port cell, are therefore less vulnerable to radiation effects so they are considered a lifetime component. The selection of adequate spectrometers is under investigation, with candidates as described in the Section below on high etendue spectrometers (HESs).

### 4.2.2. Progress in ITER edge and pedestal CXRS measurement setup.
With the aim of providing optimum radial resolution as well as the means for complementary toroidal and poloidal velocity measurements, two observation periscopes have been designed for the CXRS Edge system in Eq-port #3: lower periscope for observation in radial range $0.5 \le r/a \le 1$ (Edge observation), and upper periscope for CXRS observation in radial range $0.85 \le r/a \le 1$ (Pedestal observation). Figure 73 shows the present layout for the ITER CXRS diagnostic LOS: CXRS Edge and Pedestal are mostly toroidal systems with a $\pm 10^\circ$ angle to the horizontal plane. Similarly to the core CXRS diagnostic the layout of the edge and pedestal CXRS diagnostic is now mature, at the technical level of a PDR.

Components of the ITER CXRS Edge diagnostic are distributed throughout the ITER tokamak complex. The front-end optics are installed in Equatorial Port 3 (EP3), viewing the DNB. The collected light reaches the EP3 Interspace

through vacuum windows. Then it is transmitted along fiber bundles to the diagnostic room in the adjacent Tritium building, accommodating the back end of the CXRS Edge system. That includes HESs, detectors and Instrumentation & Control (I&C) cubicles.

The light collection system produces an image of the DNB in the image plane, located in the EP3 interspace. The plasma facing element of the light collection system is the flat First Mirror. Hot plasma particles can significantly modify the optical properties on mirrors due to erosion and deposition. That is why a sandwich design is used: single crystal molybdenum segments are fixed on poly-crystal molybdenum substrate by hot isostatic pressing (HIP). Molybdenum has a better sputtering resistance than the rhodium proposed for the CXRS core, but a lower reflectivity. The latter is acceptable due to the higher signal levels available at the plasma edge. To extend mirror lifetime during DNB off time a protective shutter is used, and a mirror cleaning system is proposed for occasionally removing plasma deposits.

The light collected by the first mirrors is transmitted through the port plug by mirrors and crosses the vacuum barrier through optical windows with the biggest available clear aperture diameter of 160 mm. Each window consists of two Fused Silica wedged disks, which are also tilted 2° with respect to the optical axis. This avoids any parallel surfaces to mitigate parasitic reflections. Fused silica material was chosen because of its high radiation resistance, high optical transmission, considerable experience in manufacturing and relatively low cost.

The ex-vessel lens assemblies produce an image of the observed area of the DNB. The fiber bundle heads are placed into the image plane of the light collection system. Special fiber head alignment systems ensure proper alignment onto the image. The light signal is transmitted from the port cell via the interspace fiber bundles to the optical connector on the outside of the bioshield. Then light is transmitted to spectrometers in the diagnostic area of the tritium building via the main fiber bundles.

To meet the radial resolution requirements the CXRS Edge diagnostic system has 26 spatial channels: 10 core channels ($0.5 < r/a < 0.85$) and 16 edge channels ($r/a > 0.85$), while the pedestal optics has 16 channels. Because core channels have smaller active signal intensity, they cover a larger area of the DNB than edge channels. Distances between these strips were chosen to provide a spatial resolution of 20 mm at the plasma edge and 67 mm closer to the center. As an example, the Pedestal system layout is shown in figure 74. The integrated fiber bundle assembly includes: the fiber bundle channels as per radial locations (for CXRS pedestal, 16 active channels and 4 passive channels), the bundle to bundle coupling connections (ISS fiber to main fiber coupling after the bio shield), the required terminations (for image coupling, for spectrometer entrance slit coupling), and arrangement of the calibration fibers (to inject calibration light signal into the light collection and transmission path as shown in figure 76). The optical fibers are organized into bundles containing four optical channels. For ITER-measurements, five bundles to cover the 20 radial locations are needed. The fiber head of each bundle has a
<figure><img src="image_85.png" /><figcaption>**Figure 73.** ITER CXRS edge and pedestal diagnostic periscopes layout looking at the DNB.</figcaption></figure>

<figure><img src="image_86.png" /><figcaption>**Figure 74.** CXRS-pedestal light collection fiber image on the DNB and arrangement in the image plane.</figcaption></figure>

10 × 4 matrix of fibers. Each column has ten fibers and denoted as one channel in ITER-CXRS-pedestal diagnostics. In each channel, nine fibers are for light collection, and one is reserved for calibration.

In the interspace, fiber bundles are subjected to neutron and gamma radiation influence. These bundles contain radiation hard fibers (low OH content, H diffused fused silica [252]), but nevertheless may acquire damages and lose transmission. Partly, light transmission can be restored by annealing the optical fibers. For this, the design provides an integrated heating tape (at temperatures up to 200 °C–300 °C) to allow annealing of the radiation damage.

The measurement performance assessment of the Edge CXRS diagnostic can be found in Serov *et al* [253]. Several scenarios, covering all expected baseline operation regimes for ITER, were used. The modeling was done for various low-Z impurity (He, Be, C, and Ne) concentrations for the whole spatial range. Statistical errors for ion temperature, density, and rotation velocity were calculated. Minimum concentrations, allowing measurements with the required accuracy,

were obtained. Effort was made to review other non-statistical error sources for the CXRS measurements, including wall reflections, multiple tungsten lines in CX spectra, effects on the active CX line shape, calibration, and positioning uncertainties. Consideration of the major error sources showed that the CXRS Edge system will be able to meet all ITER measurement requirements, apart from the requirements for the poloidal rotation velocity measurements.

### 4.3. Ion temperature measurement in a carbon-free machine

The transition in ITER wall material away from carbon to a beryllium–tungsten combination has had implications for the CXRS measurement of ion temperature ($T_i$) and rotation. Previously the CVI $n = 8–7$ CX line at 529.0 nm was routinely used and benefitted from being the dominant line in a comparatively clear spectral region. With a metallic FW the CXRS situation is less ideal; the CX lines of Beryllium may not be the dominant feature and the presence of many additional spectral lines give a more complex background spectrum. The analysis
of the CX to determine $T_i$ must be adapted and developed to cope with the new circumstances for which the JET experiment offered the best possibilities.

The beryllium and tungsten material combination of the JET ILW saw the introduction of additional impurities into the plasma and many spectral lines are observed throughout the CX spectral regions of interest. These are mainly from tungsten and include narrow emission lines of neutral W [254] as well as magnetic dipole transitions of various W charge states [255] and W CX emissions. The lines observed depend on the conditions in the plasma and the location viewed. Spectral lines from other impurities are also seen for example due to transient events. When the CX line is dominant in the spectrum then a good determination of $T_i$ and its uncertainty can be obtained. However, when the CX intensity is less favorable compared to the background nuisance lines then the $T_i$ and the quality of $T_i$ obtained from the multi-Gaussian fitting of the spectral region can be affected. There can also be an interplay in the fit between the intensity and width of the CX and nuisance lines. To reduce the influence of the nuisance lines it is valuable to have a good understanding of what these lines are; additional constraints can also improve the fit and assessment of uncertainty. Having both red and blue Doppler-shifted views through the plasma at JET has been greatly beneficial in identifying and constraining the background spectral lines we observe. This knowledge can also then feed forward into the CX analysis on other devices such as on ITER where the different viewing geometry will not offer this opportunity.

The influence of the nuisance lines in the core-CX analysis can also be reduced by using beam modulation or notching, where short off notches are introduced in one or more of the neutral beam injections (NBIs) seen by the CX lines-of-sight. For the CX analysis we subtract the spectrum measured in the beam notch from that measured with the beam on, with the aim to reduce the background and leave a simplified spectrum to fit. This can bring us back to the situation where the CX is again dominant in the spectra. It is most effective where the background lines have not changed, such as in stationary conditions, close to the notch, or using selective matching of the active and notch camera frames. Including notches in the NBI at JET requires additional effort by the scientific teams and it can be a compromise on what is acceptable to the scenario in terms of disturbing the injected power. Any CX contributions of non-notched beams are also subtracted, which at JET is a factor to consider in the use of this technique with the impurity-CX where the basic intensity may already be low. The extent and quality of the $T_i$ measurement ultimately depends on having enough CX intensity available. Use of a single dedicated diagnostic beam in ITER will simplify the situation as it does not require notches in the main injected beam power as well as removing the complication of multiple beams contributing to the CX, it is planned that the ITER diagnostic beam will be modulated.

The switch to the ILW at JET reduced the intrinsic C levels by at least an order of magnitude compared to the C-Wall era. Whilst the CVI-CX line is still observed it is difficult to obtain good $T_i$ measurements from the CVI-CX alone in

most plasma conditions, in particular, in the core region. The intrinsic beryllium offers the possibility of using the BeIV-CX lines at 465.8 nm $n = 6–5$ and 468.5 nm $n = 8–6$ to derive $T_i$. These are routinely measured; however, it is challenging to get an independent $T_i$ from the BeIV-CX and so this is mainly used for ion density. The Be intrinsic concentration is fairly low and the spectral region of these CX lines again has multiple W nuisance lines which need to be taken into account. The BeIV-CX analysis can also be further complicated by the HeII $n = 4–3$ CX transition at 468.5 nm when there is Helium present. The HeII-CX itself also has the potential for deriving $T_i$ if the line is strong enough and the spectral region can be fitted well.

Injecting a small amount of another impurity into the plasma gives us back more control of the CX intensity than having to rely on the intrinsic emissions. By adding Neon, the NeX-CX $n = 11–10$ line at 524.9 nm can be exploited for $T_i$. The amount of Neon required for good measurement to the core depends on the plasma conditions, and the inclusion of this extra radiating species must be acceptable to the experimental scenario. However, even with low concentration, an improved $T_i$ can be obtained to the core. Tungsten nuisance lines are also present in the NeX-CX spectral region of interest and can influence the CX analysis unless successfully included in the fitting or reduced through beam notches. The proximity of the NeX-CX to the CVI-CX in the spectra means that any residual CIV-CX present can be used to add constraint to the analysis. The use of neon for CX has become routine in many JET experiments [255]. Adding an impurity species such as neon will be even more important for ITER where there will be even less or no carbon CX available.

The main ion (H,D,T) $n = 3–2$ CX emission at 656 nm is now also routinely measured at JET [256]. The relative position of the CX line and beam emission spectra influence where good $T_i$ can be obtained. Where there is a mix of fuel isotopes the contributions from all isotopes must also be included. The beam notch analysis technique to simplify the spectra, typically by selectively matching active and notch frames, is required for analysis of this CX. At JET beam notches may not be suitable or wanted in the experimental scenario which can limit the availability of this $T_i$ measurement; with its dedicated modulated diagnostic beam ITER will not have this constraint.

Without carbon we can no longer rely on the old standard CVI-CX for core ion temperature measurement. The flexibility of having multiple systems and techniques available at JET has proven beneficial to maximize the potential for and reliability of obtaining good $T_i$. Injecting a small amount of neon as an additional impurity gives more control of the CX intensity than relying on the intrinsic impurities present and can enable good $T_i$ measurement into the core. Having a sufficient intensity of the CX line is crucial. The use of notches in the NBI to help separate the CX from the background spectra and exploiting the main-ion CX for $T_i$ have also proven valuable. It has been essential to develop the CX analysis to take into account the increased complexity of the spectral regions of the CX lines, to reduce the influence of the nuisance spectral lines which could otherwise affect the $T_i$ obtained.
### 4.3.1. Advanced spectral evaluation methods.

ITER CXRS measurements become complicated because of the low SNR due to strong attenuation of the DNB, caused by the large size and high density of the plasma. Other complicating factors are the presence of other spectral lines due to CX reactions with the edge neutrals [257] (passive CX line) and due to the electron impact excitation [258] (edge lines) and multiple tungsten lines [254].

Therefore, modeling is important for the development of the CXRS diagnostic. For this purpose, the simulation of spectra (SOS) code [259] could be utilized. It was developed for predicting beam-induced spectroscopy diagnostic performance on fusion devices. In particular, it has models for all main CXRS spectrum components and could also be used for spectra fitting and error estimation. The spectrum is constructed from the following components, modeled separately: active CX line, passive CX line, edge electron impact excitation lines, bremsstrahlung continuum, and statistical noise. The obtained spectrum is fitted using least-squares minimization procedure. Intensity, temperature, and rotation velocity errors are calculated from the active CX line fitting 95% confidence intervals (corresponding to ±2σ). The main source of atomic data is the ADAS database [260, 261].

The SOS models were validated using experimental data from several tokamaks, e.g. JET [261], TEXTOR [262] and EAST [263]. The benchmarking process will not be discussed in this paper. Results of the benchmarking against JET experimental data [264] demonstrate that the SOS code allows reasonable CXRS spectra predictions and could be used for active spectroscopy diagnostic development for ITER.

### 4.3.2. High-etendue spectrometers.

Due to the low active light intensity and high background the spectrometers for the CXRS diagnostics should have high etendue and high transmission combined with the necessary spectral resolution in a few limited spectral ranges. These requirements cannot be fulfilled with commercial spectrometers; therefore, special devices need to be developed.

To provide plasma parameter measurements according to ITER requirements the CXRS edge diagnostic system spectrometer should be able to collect spectra in wavelength regions indicated above. To fulfill the above requirements the high etendue, high resolution spectrometer (HES) of the CXRS Edge system will have three channels, covering all the lines listed: blue (468 ± 5 nm), green (529 ± 5 nm) and red (656 ± 6 nm).

A multi-channel HES spectrometer prototype, based on transmission gratings, was designed and built for the CXRS Edge system. The principle scheme of the HES spectrometer is shown in figure 75, while table 25 lists the technical specifications obtained during the laboratory tests. These laboratory tests were done without the light traps/viewing dump, which are shown in figure 74. Light traps and shields will be installed for further spectrometer properties improvements. In addition, interference filters will be installed in front of the image plane in each spectral channel in order to reduce the stray light level. One more future step is the improvement of

<figure><img src="image_87.png" /><figcaption>**Figure 75.** Principle scheme of the optimized multi-channel HES spectrometer: entrance slit (1), collimator objective lens (2), holographic transmission gratings for 468 ± 5 nm, 529 ± 5 nm and 656 ± 6 nm spectral ranges respectively (3), camera objective lenses (4), sCMOS XIMEA cameras (5), light traps/viewing dump (6).</figcaption></figure>

<table><caption>**Table 25.** Technical specifications of the prototype HES spectrometer.</caption><tr><td>Dispersion (blue/green/red),</td><td>3.4/3.6/5.0 Å mm—</td></tr><tr><td>Transmission (blue/green/red)</td><td>35/34/25%</td></tr><tr><td>Max. spectral resolution</td><td>~0.1 Å</td></tr><tr><td>F-number</td><td>3</td></tr><tr><td>Magnification</td><td>1</td></tr><tr><td>Entrance slit width</td><td>0.4–1 mm</td></tr><tr><td>Entrance slit height</td><td>20 mm</td></tr><tr><td>Image plane size</td><td>20 × 25 mm</td></tr><tr><td>Contrast</td><td>>750 000</td></tr></table>

the diffraction and transmission efficiency of the holographic gratings. It is expected that the diffraction efficiency could be increased up to 50% and the transmission efficiency up to 85–95%. More details about the CXRS edge spectrometer can be found in Tugarinov *et al* [265].

For the spectrometer design of the core CXRS diagnostic a different approach was adapted. A prototype spectrometer was designed by TNO company, Forschungszentrum Jülich and the Eindhoven University of Technology [266]. As designed, it meets the ITER design requirements of 1 mm² SR etendue for a single optical channel of the core CXRS optics, accepting light from NA = 0.22 fibers at the input. A single large lens is used for both collimation and focusing and a single grating used for all three spectral channels, providing measurements of hydrogen and its isotopes, carbon, neon, helium and beryllium simultaneously. The spectrometer was tested initially at TEXTOR and then moved to ASDEX Upgrade where it was used predominantly for helium measurements [267] including a feasibility assessment of measuring fast-helium from fusion reactions on ITER [268]. More recently, the spectrometer was moved to Wendelstein 7-X where it is now the primary CXRS spectrometer [269]. The spectrometer was notably easy to reassemble and to align due to the excellent mechanical design and clean separation of the input stage, slits and of each spectral channel. This results from the particular optical design which also separates the focusing, making it easy to achieve good wavelength resolution across the
<figure><img src="image_88.png" /><figcaption>**Figure 76.** The scheme of the developed technique for transmission and wavelength calibration of an optical circuit. Indicated: shutter with retroreflector (1), first mirror (2), vacuum window (3), out of vacuum optics (4), image plane (5), main fiber bundle with an additional fiber (6), spectrometer (7), illumination system optics (8), calibration light source (9), neon lamp (10), detector for light source control (11), movable Fabry-Pérot interferometer (12).</figcaption></figure>

entirety of all three wavelength bands and across all spatial channels. The separation allows the slit width to be independ- ently changed for each spectral channel with ease. A small amount of cross-talk between spatial channels is observed at W7-X due to their dense packing but this would likely not be significant in the ITER case. Experience has shown the spec- trometer to be very stable, with a drift of only 10 pm over a day due to temperature changes, making pre- and post-shot calibrations sufficient even for long pulse operation.

Both spectrometer designs for the ITER CXRS diagnostics basically fulfill ITER requirements on etendue, spectral res- olution and throughput, which demonstrates that construction of suitable spectrometers is possible. However, the necessary total etendue of all observation channels can be fulfilled only by multiple devices, therefore the decision on the final spec- trometers will surely be affected by cost limits.

## 4.4. Calibration methods

Three calibrations can be considered for the spectroscopy dia- gnostics: spatial, wavelength and intensity. The spatial calibra- tion of the diagnostics is assumed to rely on optical tests before installation and device metrology during installation, no spe- cial spatial calibration procedures are assumed to be built into the diagnostics.

For intensity calibration one should consider that the CX signal intensity $I_{CXRS}$ of an impurity line depends on the product of local beam and impurity density. Calculating beam attenuation from plasma parameter profiles the impurity dens- ity can be obtained from the measured line intensity, there- fore a method for transmission calibration is necessary to compensate for the change in the reflectivity of the first mir- ror, transmission of lenses and most importantly the optical fibers. The calibration procedure should test the whole optical

chain from the first mirror to the spectrometer and should be repeatable during the ITER lifetime without the need for access to the diagnostic. The calibration requirements are somewhat relaxed by the fact that the beam light emission sig- nal ($I_{BES}$, $H_\alpha$ for the DNB) depends on the product of the beam intensity and the electron density. If the BES light is collected from exactly the same volume as the impurity radiation the relative density of the impurity becomes [246]:

$$ \frac{n_Z}{n_e} = \frac{I_{CXRS} Q_{BES}}{I_{BES} Q_{CXRS}} $$

where $Q_{BES}$ and $Q_{CXRS}$ are atomic emission rates dependent on plasma parameters. Therefore, the calibration process need not be absolute, but only relative between the BES and CXRS wavelengths.

It is unlikely that there will be any means of positioning a calibrated uniform source, typically an integrating sphere, within the VV after operations have commenced – possibly not even during installation and adjustment where it is highly desirable. Other schemes like bremsstrahlung-based calibra- tion, branching ratio calibration (between different Balmer lines) were considered but were not found to be suitable as the main calibration tool. In the two CXRS systems a somewhat different setup is considered although both are based on col- lection of a calibration light transmitted through the collection system.

### 4.4.1. Calibration in the edge CXRS system.
For the edge CXRS systems the calibration principle is based on the use of a light reflecting element (retroreflector or mirror), which is placed on the back side of the first mirror’s shutter plate. The scheme of the developed technique for transmission and wavelength calibration of the optical circuit is presented in figure 76. An additional optical fiber is added to the main fiber
<figure><img src="image_89.png" /><figcaption>**Figure 77.** Schematic view of the calibration system of the core-plasma CXRS. A-object region with diagnostic neutral beam; B-aperture stop (entrance pupil) and shutter blade with diffuser coating on the backside; C: main optics for observation; D-overlap of illumination and radiometer systems in the front mirror unit; E-illumination path in vessel; F-idem ex-vessel; G-radiometric optical path.</figcaption></figure>

bundle (6). The main fiber bundle is connected directly to the entrance slit of the spectrometer (7), which is placed in a diagnostic area in tritium building. The additional optical fiber is illuminated by a light source (9) in spatial angle according to the spectrometer's light transmission ability. Light from the additional fiber passes through the entire aperture of the light collection system (5–4–3–2) and reflects from the retroreflector (1). Then reflected light goes back through the entire optical circuit and illuminates main fiber bundle in the image plane (5).

The idea is that by illuminating one fiber from the diagnostic area we could illuminate all the other fibers using reflected light, which will be transported to a spectrometer where it can be detected and analyzed. This scheme will allow periodical transmission calibrations for the system and could also be used to perform a wavelength calibration of the spectrometer. For this purpose, a special calibration technique, described in Shabashov et al [270], could be applied. It uses a combination of the fringe pattern created with a Fabry–Pérot interferometer (12) and a neon lamp (10) spectrum for accurate wavelength calibration.

The possibility of this calibration scheme was demonstrated using both modeling and laboratory tests.

### 4.4.2. Calibration in the core CXRS system.
The calibration scheme for the core CXRS system aims at measuring the transmission of the optical system, it provides the conversion factor between detector signal levels and spectral radiance (a photon intensity in units similar to $photon/m^2sSrnm$). As a minimum, the calibration must provide the relative sensitivity as a function of wavelength but an absolute calibration is fully advisable.

The original proposed concept uses a white light source located at a distance of ~5.2 m from a diffuser located on the back of the diagnostic shutter, as shown in figure 77. Some of the scattered light is collected by a radiometer which thereby measures the absolute spectral radiance of the diffuser. The calibration scheme uses a bent light path, in fact a single mirror in a tightly baffled light path, to introduce light onto the shutter/diffuser equally in both its open and closed positions.

When open, the diffuser radiance is measured by the directly viewing radiometer and we rely on this radiance being the same when the shutter is in the closed (calibration) position to collect the light through the full observation path of the diagnostic. A slight tilt of the diffuser plate in the order of 20° with respect to the plane of the shutter blade, in the direction of the illumination and radiometer channels, increases the spectral radiance of the diffuser by almost a factor of 3.

The initially planned location for the source and the radiometer was the interspace (IS), close to the two windows through the closure plate of the port plug. However, concerns were raised regarding access to the IS for maintaining the system. As a result, a scheme for relocation of the light source and radiometer in the port cell was developed on the basis of either optical fiber or mirror relay optics. Both options are conceivable. Furthermore, a bare tungsten-halogen lamp without intervening optics between it and the shutter could be used but it is anticipated that alternative light sources will provide considerably higher irradiance levels at the shutter. Such sources, for instance high-brightness laser-driven light sources, are worth considering and ought to be tested. A 1:2 prototype of the whole system is proposed for assessment.

With the proposed calibration system, which intrinsically represents an absolute calibration, we gain — over the pure ratio between the beam-emission and charge-exchange (CX) wavelengths — the potential to measure the bremsstrahlung or $Z_{eff}$ profile, impurity influxes, measure the DNB attenuation, monitor the rate of deterioration of the first mirror, with the possibility of limiting the use of the diagnostic, if this is severe, and evaluate the efficacy of in-situ mirror cleaning. This calibration method is, however, more challenging to technically integrate and its implementation is under review.

### 4.5. Additional measurement possibilities with active beam techniques

The active spectroscopy diagnostics on ITER aim at a few main purposes, but the optics and beams would provide the possibility for some additional measurements which are being used on present-day devices. Although they are not part of the
<figure><img src="image_90.png" /><figcaption>**Figure 78.** Atomic emission rate for fast ion CXRS on He2+ and D+.</figcaption></figure>

approved ITER diagnostic set, but they are already included or proposed as a supplementary use of the diagnostics.

### 4.5.1. Fast ion CXRS (FICXS) for confined and escaping alpha particles.
Four separate FICXS cases are usually addressed by spectral modeling codes: Confined fusion α particles, escaping α particles, $ ^3He^{+2} $ minority heating and finally, slowing- down beam ions (also known as Fast-Ion D-Alpha or FIDA).

The common feature is the procedure of calculating the observed CX spectrum from the convolution of a three-dim VDF with a collision velocity dependent CX emission rate function. Recently revised ADAS [260] emission rates are shown in figure 78. The VDFs can be isotropic as for the case of slowing-down fusion alpha particles or anisotropic for slowing-down beam ions [271]. The main-goal of FICXS modeling is the deduction of fast ion densities, verification of slowing-down velocity distribution features, assessment of fast ion losses and associated anisotropic issues.

Absolute fast ion density modeling is linked to the calculation of a fast ion source rate which in the case of confined fusion alpha particles is based on toroidally symmetric profiles for $n_d$, $n_t$ and $T_i$. By contrast, an anisotropic Fokker-Planck beam injection model (see [271]) applies for slowing-down beam ions. Numerical treatment of Source rate calculations may differ significantly in different simulation codes such as TRANSP [272] or SOS [259]. A particularity for the Beam Fast Ion CXRS modeling of a broad-band Balmer Alpha spectrum (or FIDA, Fast Ion Dalpha spectrum) is the dual role of the beam as a source of fast ions and at the same time as a probe beam for the diagnosing CXRS process. Figure 79 through figure 82 illustrate characteristic features for the four types of FICXS spectra: confined and escaping alpha particles, He3 minority heating and NBI slowing down ions.

### 4.5.2. Measurement possibilities for turbulence and MHD waves.
Currently, millimeter-wave reflectometry is envisioned as the primary diagnostic for the measurement of plasma density fluctuations [273]. In the SOL high amplitude fluctuation (blobs) are an essential element in setting the divertor heat load [274]. At the high fluctuation amplitudes interpretation of reflectometry measurements is complicated [275] and the lifetime of Langmuir probes is questionable.

<figure><img src="image_91.png" /><figcaption>**Figure 79.** Confined alpha particle FICX spectrum for ITER core CXRS diagnostic in Eport-3 at $r/a = 0.5$, DNB, and the spectra plotted vs particle energy.</figcaption></figure>

<figure><img src="image_92.png" /><figcaption>**Figure 80.** Escaping alpha particle study for ITER core CXRS diagnostic in Eport-3-Upper showing thermal and FI spectrum for outermost line of sight ($r/a = 0.98$), and the spectra plotted vs particle energy. Spectral noise level is determined by underlying continuum.</figcaption></figure>

<figure><img src="image_93.png" /><figcaption>**Figure 81.** FICXS modeling of ICRH He3 minority heating assuming a minority concentration $n(He3)/n(He4)$ of 10%. The minority birth energy is 1 MeV.</figcaption></figure>

<figure><img src="image_94.png" /><figcaption>**Figure 82.** FICXS modeling of beam-ion slowing-down energy spectrum for ITER 9 MA-steady-state case including thermal, MSE and fast ions plus halo contributions. Only far end of red-wing fast ion spectrum visible just above continuum noise level.</figcaption></figure>

Therefore, at the edge and SOL it is worthwhile considering fluctuation BES as an alternative to measure blobs [276] and MHD waves. Such a fast BES diagnostic could use the CXRS periscopes viewing the DNB. Two observation geometries were considered for fluctuation BES: the pedestal and
<figure><img src="image_95.png" /><figcaption>**Figure 83.** Modeled channel arrangement for feasibility study of fluctuation BES. (A) The studied arrangement for the pedestal system shows dedicated channels for BES (blue) between the CXRS channels (green). (B) The studied arrangement for the core BES system is proposed to piggy-back on the existing CXRS field of view (blue).</figcaption></figure>

the core observation. The fluctuation BES system could either piggy-back on existing optical fibers [277] or rely on extra fibers that can be installed to utilize the part of the beam image not used by the CXRS systems. The analysis presented here employed the first approach for the core system and the second approach for the pedestal system for ITER scenario 2 as shown in figure 83. It has to be noted, that in this configuration observation of the DNB is limited to $r/a < 1$, therefore SOL turbulence cannot be measured in this scenario. However, a slight modification of the plasma configuration might allow study of at least near-SOL turbulence.

For any fluctuation BES diagnostic three factors are important: separation of beam light from edge plasma radiation, SNR and spatial resolution resulting from the intersection of beam and observation.

The fundamental challenge with ITER BES observations lies in optimizing the narrow wavelength filter to maximize SNR and the SBR. The sheer size of ITER and harsh conditions result in observation geometries with small etendue and complicated optical systems with low optical transmission rates. ITER is also expected to feature high background plasma radiation foreshadowing challenges regarding the SBR. To study possibilities the SOS code [259] was used to model the expected light spectra for each detector channel. To achieve the desired filter spectral profile for each channel, a realistic filter characteristic resembling a three-cavity optical filter was shifted in wavelength. The flat-top region of the filter profile was kept at 90% throughput. Examples of the modeled spectra and the corresponding optimized filter characteristics are shown in figure 84.

By applying the optical filters on the simulated spectra (figure 84) one arrives to the filtered beam emission and corresponding background components, which in turn can be used to calculate the SBR of each channel (figure 85). The filter efficiency is calculated as the ratio of filtered beam emission to the unfiltered beam emission for each channel. The core CXRS features a perpendicular view of the DNB at $\rho = 0.15$, so in the central region the optimized filter can only utilize a side peak of the MSE spectrum to avoid the edge background radiation. This, combined with the large continuous background, results in an unfavorable SBR $\ll 1$. For the outer channels of the core system, the Doppler shift is sufficient (as seen in figure 84(a)), and thus the SBR rises to a tolerable value (figure 85). The

<figure><img src="image_96.png" /><figcaption>**Figure 84.** Observed and filtered spectra modeled by the SOS code: beam emission (MSE) spectrum (orange), total (continuous + edge) background (green) and their sum (blue). The right axis applies to the filter characteristic (dashed blue). (A) Outermost core field of view. (B) Outermost pedestal field of view.</figcaption></figure>

pedestal observation geometry features a decreasing Doppler shift towards the LCFS, resulting in a significant drop of filter efficiency. However, even the worst-case filtering (as shown in figure 84(b)) yields a sufficiently high SBR due to the steadily decreasing background. SBR values in general are sub-optimal, and necessitate a careful methodology for background subtraction. The forthcoming analysis introduces two approaches to background correction. The ‘optimistic’ background scenario considers a background that is steady on the time scale of the studied fluctuations, and therefore it can be measured with a negligible statistical uncertainty. The ‘conservative’ approach assumes a fluctuating background that needs to be measured by a parallel fast detector system with a measurement uncertainty comparable to that of the fluctuating BES signal.

Considering the photon current values computed by applying the optimized filter throughputs on the simulated spectra, two detector types were considered for the ITER fluctuation measurements. An avalanche photodiode (APD) for the core observation geometry and multi pixel photon counter for the pedestal observation geometry, both at 500 kHz sampling frequency. To assess the expected noise, the SNR dependency on the photon current was used for the above stated detectors [278, 279]. An optimistic SNR estimate is calculated by assuming negligible background light fluctuation, while a conservative estimate assumes similar relative amplitude fluctuation for the background than for the beam light. Figure 85(b) shows the expected SNR values, where the realistically achievable values probably lie in between. The SNR range for the pedestal observation geometry is not very favorable, as the values range from 2 to 10 depending on the location and the above approaches to estimation. The SNR estimates for the core system outside of $\rho = 0.4$ are much better. Figure 85(c) shows the relative amplitude threshold for
<figure><img src="image_97.png" /><figcaption>**Figure 85.** Expected performance parameters for both fluctuation BES systems plotted against the normalized minor radius. (a) Filter efficiency and SBR for each channel. (b) Range of expected SNR values. (c) Range of expected detection limits for individual fluctuation events in relative amplitude.</figcaption></figure>

detection of individual transient events, as calculated from the corresponding SNR values. The core observation geometry is predicted to assess relative amplitudes as far as below 10%, while the pedestal observation scenario is predicted to be able to detect relative amplitudes of 10% in the best-case scenario or above. The actual detection limit greatly improves for stationary fluctuations, where statistical averaging or spectral methods can be applied.

Regarding the spatial resolution, the pedestal system should be capable of resolving the 2D structure of field-aligned perturbations, as the LOS are well aligned to the magnetic field lines across the beam [280]. The quantitative measure of the emission smearing resulting from the slight misalignment between magnetic field lines and LOS is characterized in figure 86. The smearing effect decreases from the LCFS towards the core, from values in the range of 15 mm, down to 5 mm. Furthermore, emission smearing due to the finite lifetime of excited states for the pedestal diagnostic was calculated to be 5 mm at the pedestal and 3 mm towards the core. Finally, accounting for the 10 mm radial width of the detector projections, the system's effective radial resolution is valued at 18 mm at the innermost edge, while 30 mm close to the LCFS. Similarly, an overall poloidal resolution range of 22 mm to 24 mm was estimated. Despite the considerable beam width, the proposed setup presents a reasonable spatial resolution,

<figure><img src="image_98.png" /><figcaption>**Figure 86.** Spatial resolution of the pedestal fluctuation measurement for field-aligned perturbations. Poloidal projection of the proposed detector array and the extension of a field-aligned perturbation due to misalignment of LOS and magnetic field lines.</figcaption></figure>

<figure><img src="image_99.png" /><figcaption>**Figure 87.** Poloidal snapshot of edge density turbulence simulated by HESEL, and the poloidal projections of possible fluctuation BES LOS of the pedestal CXRS viewing the scrape-off layer in inward-shifted scenario.</figcaption></figure>

with the capacity to resolve plasma density fluctuations and filaments in a width range starting from 1–2 cm.

The close-to-vertical LOS of the core system view result in a poloidal resolution in the range of 100–150 mm, and a 35–50 mm of radial resolution in the outer region, where the signal quality was shown to be favorable. Therefore, a fluctuation BES system on the core CXRS periscope will not be able to detect turbulence, but it can be considered for the detection of MHD waves in the outer core region.

#### 4.5.3. Synthetic BES diagnostic.
In order to assess the measurement possibilities of a fluctuation BES system a synthetic diagnostics workflow was set up to provide synthetic signals [281] for SOL/edge turbulence using the HESEL code [282], and the synthetic fluctuation BES signal provided by RENATE [283]. Figure 87 shows a snapshot of the simulated turbulent 2D poloidal density field for Scenario 2 plasma, and the channel projections of the pedestal fluctuation BES system (as shown in figure 83(a)) overlaid, assuming a slightly inwards shifted plasma. Emergent SOL filaments, as predicted by HESEL, will be smoothened as a result of the spatial smearing of the diagnostic system, but are still sufficiently large to be spatially resolved.
<figure><img src="image_100.png" /><figcaption>**Figure 88.** Turbulent density fluctuations in the SOL simulated by HESEL, and the corresponding response in BES signal amplitudes.</figcaption></figure>

This conclusion is illustrated in figure 88, which shows the time evolution of the density along the central detector array of figure 87, and the time evolution of the corresponding BES signals without detector noise added. Tilted structures emerging from the LCFS are clear indicators of radial filament propagation, although the resolution might not suffice for accurate velocity measurements, it does suffice for the assessment of SOL filament statistical features as well as filament identification as long as they are sufficiently large to pass the detection threshold.

Measurement possibilities have been further analyzed by the introduction of realistic simulated measurement noise, by considering the deterministic signal as expected values [284]. First analysis shows that the low SNR values predicted for the pedestal fluctuation BES system significantly affect the signal statistics, therefore comparison of measurements with synthetic diagnostic signals is essential [285].

#### 4.5.4. Diagnostic beam development.
As described in section 1, for measuring active spectroscopy signals with good SNR and good spatial resolution, a 100 keV hydrogen DNB is considered for ITER [4]. The beam parameters desired to meet the ITER-CXRS diagnostics requirements [246, 252, 287] are listed in table 26. The injector consists of a 100 keV RF based negative ion source which produces energetic beams with controlled optics using a three-grid extractor and accelerator system, coupled to the source. The stringent requirement of beam divergence $\le 7$ mrad is met by using optimized aperture shapes, inter grid spacing, voltage tuning and appropriate control on the beamlet deflections due to various factors like the magnetic and filter fields, beamlet-beamlet interactions, etc [288]. Considering 60% neutralization efficiency of negative ions, transport losses over lengths of 21.7 m due to beam divergence, deflections and reionization losses, the desired extracted current from the RF source is 60 A @ 100 keV. However, for achieving a good SNR of CXRS signals, a neutral beam current of $\sim 20$ A at 100 keV beam energy will be injected into ITER plasmas either with a modulation frequency of 5 Hz, 3 s ON/20 s OFF duty cycle (burst mode), or as a 100 ms pulse every 1.5 s continuous throughout the

<table><caption>**Table 26.** Parameters of the ITER DNB [286].</caption><thead><tr><td>Parameter</td><td>Value</td></tr></thead><tbody><tr><td>Beam energy</td><td>100 keV</td></tr><tr><td>Beam species</td><td>H</td></tr><tr><td>Extraction area</td><td>$\sim 0.2$ m<sup>2</sup></td></tr><tr><td>Beam divergence</td><td>$< 7$ mrad</td></tr><tr><td>Accelerated ion current</td><td>60 A</td></tr><tr><td>Accelerated current density</td><td>30 mA cm<sup>-2</sup></td></tr><tr><td>Beam power at $\sim 20.7$ m</td><td>2 MW</td></tr><tr><td>Modulation</td><td>5 Hz during 3 s ON time</td></tr><tr><td>Duty cycle</td><td>3 s on, 20 s off</td></tr></tbody></table>

discharge (control mode). The passive signals recorded during beam off period will be used for an efficient background correction to retrieve the active signal which are otherwise convoluted with a strong passive signal.

The extracted current passes through a 3 m neutralizer placed at 1 m from the source. The mixture of ion and neutrals is separated in the 1 m long residual ion dump (RID) placed at 0.75 m from the neutralizer. The neutral beam exiting from the RID is characterized with a calorimeter having two panels arranged in a V shape that can open and close under evacuated conditions. Considering the huge power and power densities of $\sim 10$ MW m$^{-2}$, the panels of the calorimeter are made of a stack of CuCrZr elements working on the hypervapotron concept. Once characterized, the two arms of the calorimeter are opened to allow transport of the beam to the plasma.

There are large cryopumps (pumping speed: $10^6$ 1 s$^{-1}$) installed, beginning from the neutralizer exit to the end of the injector, to pump out the Hydrogen gas. The reionization losses from the exit of the RID to the point at which the beam enters the ITER machine are estimated to be $\sim 20\%$. The ions born out of reionization of neutrals and deflected by the stray magnetic fields are collected on the frontend components and the duct which are lined with suitable liners. The design of the liner surface takes into account the heat loads and maximum power densities of the incident deflected ions on the surfaces due to focusing effects of the stray fields.

In order to establish the operation and performance of the DNB injector prior to its onsite commissioning at ITER, and to provide the beam operational database for supporting modeling and expected S/N ratios during CXRS measurements at ITER, an INdian Test facility (INTF) is conceptualized and is currently under commissioning at ITER-India laboratory, Bhat Gadhinagar. This is a voluntary task undertaken by INDA with a unique facility of transporting and characterizing ITER like beams over 21.5 m, similar to the beam transport path length at ITER. The operation of beam is optimized, and the performance is assessed by deploying several invasive and non-invasive diagnostics. An overview of different diagnostics, planned for INTF operation is shown schematically in figure 89 and a more detailed description can be found in [289–291].

The beam divergence and beam uniformity are measured by a Doppler shift spectroscopy (DSS) diagnostic. The negative ion density in the source and its uniformity near the large
<figure><img src="image_101.png" /><figcaption>**Figure 89.** (a) Deployment of various diagnostics in INTF for characterizing the DNB. Figure 1(b) shows the details of the line of sights of Doppler shift spectroscopy diagnostics and the placement of the CFC based inertial calorimeter for the beam divergence and beam inhomogeneity measurements. Reproduced courtesy of IAEA. Figure from [291]. Copyright (2019) IAEA.</figcaption></figure>

extraction plane will be monitored by cavity ring down spectroscopy. The beam power, power density and beam spatial divergence are measured using a water cooled V shaped calorimeter placed at ~5.9 m and at 20.67 m from the Grounded Grid (GG), an inertial calorimeter (placed at ~1 m from GG) made up of carbon fiber composite (CFC) with an IR camera will be used for imaging the beam footprint. Near to the beam duct and beam focusing location, arrays of visible cameras are placed to obtain a tomographic reconstruction of the beam emissivity [229, 230]. These beam characterization experiments will essentially aid the ITER-CXRS diagnostics to benchmark the actual DNB current that will be available to them for measurements when the DNB is operational.

### 4.5.5. Present status of ITER- DNB.
The Indian efforts to operate and characterize beams for ITER follow a parallel path of beam and technology development. The road map for DNB development involves experiments on two additional test beds, ROBIN and TWIN.

The ROBIN test bed is established under a memorandum of understanding (MoU) with IPP Garching. It has a single RF driver source test bed where H- beams of 2 A with extracted current densities of 27 mA cm⁻² have been established in a surface assisted Cesiated source [292, 293]. In the recent years ITER like H- current densities ≥30 mA cm⁻² with e-/ion ratios <1 have been achieved. At present the ROBIN test bed has been upgraded to a neutral beam test bed to understand and operate neutral beams using ITER like electrostatic ion deflector (ERID) for separating the ionic component of the beam from the neutrals post neutralization.

The TWIN test bed is 1/4th of the DNB source in terms of its extraction area. It consists of a two-RF driver based indigenously developed ion source, using a coil configuration and connection to the RF generator similar to the one envisaged for the DNB source. TWIN is designed to operate in both air (where the RF drivers are exposed to air) and vacuum (where the entire ion source is inside a vacuum chamber, similar to ITER configuration) modes [294]. The aim is to learn power

coupling to two drivers from a single RF generator, simulating thereby to a large extent the experiments with the full-sized source on the INTF and later at ITER. RF power to the tune of ~70 kW have been coupled to produce plasma in the source. The source also has a 3-grid extractor and accelerator system with optics similar to the DNB 1 segment 3 grid extractor and accelerator system. The same has been manufactured and currently under assembly at the test bed.

The INTF is a replica of the ITER DNB beam line with a 21 m path length with an aim to generate database related to the availability of the neutral current at the location of the measurement using CXRS [287]. The facility consists of a VV, an 8 m long, 4.5 m diameter one with double O ring sealed openable top lid, and is coupled to a 12 m tapered duct consisting of cylindrical sections with reducing diameters. Ten modules of cryopumps each having a pumping speed of 10⁵ 1 s⁻¹ have been installed in the vessel. The beam source and the beam line components viz the neutralizer, ERID and the calorimeter shall be housed in the vessel to carry out the desired experiments of neutral beam production for ITER like parameters. The beam line components have been manufactured and installed in the VV. The source assembly and integration is underway at the factory with the delivery expected by the end of 2025. The 2nd calorimeter housed at the end of the beam line to determine the neutral beam characteristics after a traversal path length of 21 m has been procured and is currently under installation. Interfacing of the desired auxiliary systems like the hydraulic, gas and multichannel DAQ and control system along with the desired multi megawatt power supplies is underway. Table 27 provides a comparison of the important parameters related to the source and the beam characteristics.

## 4.6. Summary

Although active spectroscopy diagnostic on ITER will be possible only from the Pre-Fusion Power Operation Phase 2 (PFPO-2) good progress has been achieved in all active spectroscopy diagnostics. Physics, modeling, and technology background of all techniques have been clarified, but for more
<table><caption>**Table 27.** Parameters of the ITER DNB test devices.</caption><thead><tr><th>Design Parameters</th><th>INTF</th><th>TWIN</th><th>ROBIN</th></tr></thead><tbody><tr><td>Number of RF plasma drivers</td><td>8</td><td>2</td><td>1</td></tr><tr><td>Maximum RF power @ 1 MHz (kW)</td><td>800</td><td>180</td><td>100</td></tr><tr><td>Beam extraction area (cm²)</td><td>~1970</td><td>~492</td><td>~73</td></tr><tr><td>Number of beamlets (exposed to plasma)</td><td>1280</td><td>320</td><td>146</td></tr><tr><td>Plasma grid aperture diameter (cm)</td><td>1.4</td><td>1.4</td><td>0.8</td></tr><tr><td>Beam energy (keV)</td><td>100</td><td>50</td><td>45</td></tr><tr><td>Beam current (A)</td><td>60</td><td>15</td><td>2</td></tr><tr><td>Extracted electron-to-ion ratio</td><td>&lt;1</td><td>&lt;1</td><td>&lt;1</td></tr><tr><td>Negative ion beam current density (mA cm<sup>-2</sup>)</td><td>&gt;30</td><td>&gt;30</td><td>&gt;30</td></tr></tbody></table>

details measurements are still being done on existing tokamaks. Detailed design of the observation and detection systems is ongoing and is expected to be finished on time. A key element is the diagnostic beam, of which laboratory tests are progressing towards the final parameters.

In recent years it has been recognized that active spectra on ITER will contain more information than originally considered, therefore modeling has started for determining the possibility of extracting information on fast particles and turbulence. The results indicate that with modest extensions of the detection systems these measurements could be implemented.

## 5. Passive emission diagnostics for high performance metal tokamaks

M. G. O'Mullane²⁵, Ling Zhang²⁶, R. Barnsley², R. Tieulent², D. Mazon¹, D. Colette¹ and V. S. Neverov²⁷

### 5.1. Importance of passive spectroscopy

Building theories and models to interpret the emission from plasmas has a long history and is still the only feasible approach for understanding the behavior of non-terrestrial sources. The quantum theory of the atom was uncovered to explain the spectra from both emission sources, stars, arcs, filaments, flames etc, and the absorption of gases [295]. Increasing sophistication in the instruments for measuring emission in different regions of the electromagnetic spectrum validated the theories and enabled more precise measurements of plasma parameters, which in turn led to ways of controlling plasmas based on a robust physical picture of the plasma. Magnetically confined plasmas for fusion rely on a controllable plasma with characteristics bounded in $n\tau-T$ space, an extension of the Lawson criterion [296], where $n$ is the plasma density, $T$ the temperature and $\tau$ is the confinement time. For ignition this means $n\tau \ge 10^{20}m^{-3}s$ at $T > 10$ keV. Such operational requirements also set the size of the plasma, with the current largest machine JET having a plasma volume of ~80 m³ and ITER a factor of ten larger at ~830 m³. The size of the machine also determines the ease of access for diagnostic systems. Most current fusion devices are relatively accessible and their suite of diagnostics evolved in an ad hoc fashion. Newer

machines incorporated diagnostic provision into the design of the machine, e.g. W7-X [297], and notably ITER has considered its diagnostics as an integral part of the design process [3]. The complexity introduced by the blanket and cryostat, the deleterious effects of the neutron flux from fusion DT products and the reliability requirements of licensing a nuclear facility introduce constraints on diagnostics not present in current devices. The accessibility constraints for DEMO may be even more restrictive [298] so it is important to maximize the capabilities and interpretation methodology of the diagnostic data from ITER.

A wide angle view of the JET plasma in the visible part of the spectrum, shown in figure 90, illustrates the scope and complications of using passive spectroscopy to diagnose fusion plasmas. All regions of the plasma, core, edge, SOL and divertor, emit radiation. The local plasma conditions in these regions vary widely and in the edge and divertor the gradients of the plasma temperature and density profiles are large. The degree of ionization and the energy of the local, background electrons is the main determinant of the type, and hence wavelength, of the emission. Generally visible light arises from the more loosely bound electrons of the neutral to a few times ionized ions, and therefore is more likely to occur in the lower temperature edge and divertor regions. VUV, SXR and x-rays are more characteristic of the confined plasma inside the last closed flux surface (LCFS). The central electron temperature of a high performance ITER plasma will be at ~25 keV, sufficiently high to fully ionize argon, iron and xenon. The ionization potential of tungsten is 80.8 keV so will always retain bound electrons, which has negative implications if the operating scenario does not prevent impurity accumulation. For safe operation and maximum scientific return, measurements must be taken from all these regions of the plasma. Figure 90 also demonstrates that all, purely passive, instruments will give a measurement that is integrated along its LOS. To obtain profile information, multiple LOS or some form of scanning of a single LOS must be engineered in addition to an analysis workflow which can invert the measured signals, e.g. [299] for a visible system and [300] for an x-ray instrument. However single LOS instruments can be simpler to build and are more robust in use and their measurements do have value. In particular the primary role of some survey spectro meters is to detect what impurities are in the plasma, so just their presence or absence of a signal above the noise floor is
<figure><img src="image_102.png" /><figcaption>**Figure 90.** Interior of JET illuminated by its plasma emission, contaminated following a sodium influx. The yellow color shows that post-pulse analysis may not always be necessary. The complex interior surfaces and the sharp outline of the last closed flux surface can be seen.</figcaption></figure>

a sufficient indicator. The vessel walls will emit black-body radiation but this review is confined to emission originating from the plasma.

ITER will be a well diagnosed machine in keeping with its research remit and may well be the last burning plasma device with an extensive suite of general purpose spectroscopy systems. The choice of instruments is pragmatic and is based on the roles of MP and BC, AC and for physics understanding, in that order of importance. A set of measurement requirements determined the instrumentation as described in [4, 6]. Passive spectroscopy has a primary role in detecting and quantifying impurity content and an important secondary role, to provide redundancy and resilience, in many other measurements. Figure 91 shows the location and observation cones of ten systems based on visible, VUV or x-ray emission. Complete coverage of the plasma, in all wavelength regions, is impossible but a mixture of imaging and narrow LOS will sample emission from the core, edge and divertor plasma with time resolutions of the order of tens of milliseconds and spatial resolution in the centimeter range. The measurement expectations given in the early overview of the spectroscopic systems [301] are still relevant but, 20 years on, there are many detailed papers on the design and implementation of the various instruments. Table 28 lists the suite of passive spectroscopy diagnostics and some of the publications of note.

This is not an exhaustive review of passive emission diagnostics for a fusion machine; there are many general reviews and some specialized subsets dedicated to diagnostics in different wavelength regions e.g. [315–318], but rather reports on developments of interest to large tokamaks with metal plasma facing surfaces. The choice of ITER for a beryllium FW and tungsten tiles in the divertor [319], along with excluding carbon as an in-vessel material, prompted many investigations into the emission of tungsten at ITER conditions. The ILW at JET is the closest analogue [320] to the ITER concept but there have been numerous supplementary studies on different aspects of the behavior in a plasma, deficiencies in atomic data, investigation of transport and

<figure><img src="image_103.png" /><figcaption>**Figure 91.** The port location, in upper (UP), lower (LP) or equatorial (EQ) ports, of the passive spectroscopy instruments of ITER with the extent of the cone of view for each instrument. See table 28 for measurement roles.</figcaption></figure>

measurement/instrument challenges. The following sections review the current knowledge of tungsten spectral emission from neutral to highly-ionized systems, the investigation of radiation hard energy-resolved detectors and the emission model used to aid the design of many of the passive emission diagnostics in ITER. Finally the inverse problem of turning these measurements into useful engineering and operational quantities is outlined.

## 5.2. The spectroscopy of tungsten

ITER will be a metallic tokamak with a beryllium FW [321] and a tungsten divertor [322]. Beryllium getters oxygen and, as a low Z element, a significant quantity can be tolerated in the plasma. Tungsten can accept very high power loads and does not retain tritium [323] which is the principal reason for excluding carbon which would otherwise be an excellent PFC. Tungsten was used in the aperture limiters of early tokamaks and the first tungsten spectral feature reported was a ‘quasicon tinuum’ (QC) x-ray feature at $\lambda \sim 5$ nm, from ORMAK [324]. This was attributed to $4f^{10}4f^{11} - 4d^{9}4f^{n+1}$, $4d^{10}5d^{n}$ transitions of Ag-like to Rh-like tungsten ions. Figure 92 shows that this is a characteristic signature of tungsten in all tokamak plasmas and that spectrometers with high spectral resolution have enabled contributions from individual ions to be identified [325].

The map of the strongest tungsten lines as a function of ionization stage in figure 93 shows that emission from tungsten is primarily in the VUV and x-ray spectral regions, which has the implication that any spectroscopic instruments for ITER must be vacuum coupled and designed to minimize the flux of neutrons onto any optical components and detectors.

The minimum electron temperature required to ionize the tungsten stages in the $\lambda \sim 5$ nm feature is $\sim 1$ keV and its structure and shape is sensitive to temperature. The long pulse lengths of WEST enabled a spatial scan corresponding to a peak temperature profile change of 2.07–3.35 keV [328] and the analysis confirms that at the higher temperatures the maximum charge state was $W^{46+}$ but the lower temperature emission zone was dominated by $W^{33+}$ and $W^{34+}$ and lower. Radial profiles of $W^{43+}-W^{45+}$ were measured with a space
<table><caption>**Table 28.** The suite of passive spectroscopy diagnostics to be installed on ITER. The internal naming convention (PBS) is given for convenience and the primary role is taken from the latest measurement requirements (accessed October 2022). Each instrument will also have supplementary and backup roles. Their location on the ITER tokamak is shown in figure 91.</caption><thead><tr><th>System</th><th>PBS</th><th>Primary role</th><th>Recent references</th></tr></thead><tbody><tr><td>H-α monitor</td><td>55.E2</td><td>C, O influx<br>Be influx distribution<br>ELM radiation<br>L-H step change<br>D +T chamber influx</td><td>[302, 303]</td></tr><tr><td>Divertor impurity monitor</td><td>55.E4</td><td>Position of ionization front<br>Flux of D, T, Be, C and W<br>T<sub>ion</sub> profile</td><td>[[304], subchapters 6, 7]</td></tr><tr><td>VUV survey</td><td>55.E3</td><td>Be, C, O, Cu and W<br>relative concentration<br>Extrinsic Ne, Ar and Kr relative concentration</td><td>[95, 305]</td></tr><tr><td>VUV divertor</td><td>55.EG</td><td>Flux of Be, C and W</td><td>[96]</td></tr><tr><td>VUV edge</td><td>55.EH</td><td>Fractional impurity content for Z &lt; 10 and Z ≥ 10</td><td>[97]</td></tr><tr><td>Core x-ray</td><td>55.E5</td><td>T<sub>ion</sub> profile<br>Poloidal velocity<br>Toroidal velocity<br>Fractional impurity content<br>for Z &gt; 10</td><td>[306]</td></tr><tr><td>Visible spectroscopy</td><td>55.E6</td><td>Continuum measurement</td><td>[307, 308]</td></tr><tr><td>Reference system<br>Survey x-ray</td><td>55.ED</td><td>Be, C, O, Cu and W influx and relative concentration<br>Extrinsic Ne, Ar and Kr influx and relative concentration</td><td>[309–311]</td></tr><tr><td>Edge x-ray</td><td>55.EI</td><td>Poloidal velocity<br>Edge T<sub>ion</sub> profile</td><td>[312]<sup>a</sup></td></tr><tr><td>Radial x-ray camera</td><td>55.E7</td><td>B<sub>θ</sub>/<i>B</i> MHD mode<br>ELM temperature transient<br>Runaway current<br>Transient radiated power profile</td><td>[313, 314]</td></tr></tbody></table>

resolved VUV instrument at EAST [329] and shows a wide profile for these ions (up to $ρ \sim 0.4$) and profiles which display a variation to core temperatures changes between 2.9 and 3.3 keV. This feature will be a key measurement for ITER but unlike ORMAK, ASDEX Upgrade, WEST, EAST and JET where the emission occurs in the confined plasma this will be an edge, or possibly pedestal, feature and thus may behave more like an influx measure.

The efficiency of tungsten radiation and its consequences for plasma viability were recognized even from simple average-ion calculations [330] so tungsten was not used in the next generation of large tokamaks such as TFTR, JET and JT-60. There were advances in atomic physics calculation capabilities of high-Z ions but experimental comparisons for tungsten came mainly from laser experiments, e.g. [331], or from occasional injection of heavy impurities into tokamaks [332]. A series of experiments at ASDEX Upgrade in the late 1990s [333] suggested that tungsten surfaces in a high density, low temperature divertor had advantageous

sputtering and re-deposition characteristics and that a radiative mantle and ICRH heating could be used to mitigate any accumulation. These findings prompted renewed interest for tungsten in tokamaks leading to significant experimental programmes: the near 90% tungsten FW of ASDEX Upgrade [334], JET installed the ILW with a beryllium FW and tungsten divertor [58] and WEST repurposed Tore Supra to mimic ITER divertor heat flux [335] and EAST is now a full metal wall device with a molybdenum FW and tungsten upper and lower divertors [336]. Since the selection of tungsten as the primary divertor PFC in ITER almost every tokamak has a tungsten related programme.

Until this revival of interest detailed measurements and atomic structure calculations for other ionization stages of tungsten were patchy. Spectra of WI [337] and WII [338] from arc spectra were well known and there were more limited analyses of WIII-WVIII but then there was a gap until WXXVIII, the Ag-like system from the tokamak observations. The most current compilation of evaluated energy levels

<footnote>a Not the principal subject of paper.</footnote>
<figure><img src="image_104.png" /><figcaption>**Figure 92.** The λ ~5 nm tungsten feature from ORMAK (1980), JET (2020) and EAST (2021). The JET example shows a simplified 2 temperature zone fit to the feature—the spectrum was smoothed (Savitzky–Golay filter), with the background removed and normalized before fitting. The EAST spectrum identifies many of the contributing ionization stages. Reprinted figure with permission from [326], Copyright (1980) by the American Physical Society.</figcaption></figure>

<figure><img src="image_105.png" /><figcaption>**Figure 93.** The wavelengths of contributing stages of Ar, Fe and W in the VUV and x-ray spectral ranges as a function of ionization stage. The Δn = 0, 1, 2 transitions are indicated by open squares, open circles and filled circles. The 20 strongest transition per stage were given by the photon emissivity coefficients taken from the calculations used in [327] but there is no ionization balance to rank their relative contribution due to the differing widths of each ion's emission shell.</figcaption></figure>

and wavelengths is the NIST energy levels and lines database
[339] and a comprehensive summary of tungsten data pro-
duced between 1980–2007 is described in [340].

The development of the electron beam ion trap (EBIT)
revolutionized spectroscopy by enabling precise isolation of
individual ionization stages. An early result from the Tokyo
EBIT was a measurement of the neon-like W<sup>64+</sup> spectrum at
a beam trap energy of 20 keV [341]. As will be seen in the next
section some of the seven lines from the 2s<sup>2</sup> 2p<sup>6</sup> S<sub>0</sub> – 2s<sup>2</sup> 2p<sup>5</sup>
3n<sup>1</sup> transitions, spanning 0.1158–0.1492 nm, can be used to
measure ion temperature and rotation velocity profiles [342].
Accurate wavelengths are essential when designing crystal
based instruments since the wavelength bandpass of such
instruments is dependent on the 2d lattice spacing and can
therefore be quite narrow.

There have been many EBIT experiments filling-in the
spectra of the unknown tungsten ions and table 29 shows the
range of results and the different spectral regions required to
make use of tungsten emission as a diagnostic. The EBIT
plasma is very different from that in a tokamak being two
orders of magnitude lower in density and with a very nar-
row range of electron energies, unlike the Maxwellian elec-
tron energy distributions of tokamaks. Even so a collisional-
radiative model is required to interpret the observed intensities
in EBIT spectra [343, 344]. Some metastables, see [345] for an
example in Ti-like W<sup>52+</sup>, can build up significant population
and become extremely long lived. These are unlikely to occur
in tokamak conditions because of the wider range of energies
of the background electrons.

The EBIT observations have been used to disentangle
the quasi-continuum features observed in tokamaks [358,
359] and employ grating and crystal spectrometers which
are broadly similar to the designs for ITER. The EBIT
observations in the visible spectrum are from forbidden M1
<table><caption>**Table 29.** Summary of EBIT results for tungsten ions. Multiple ions are reported in many papers and the wavelength range tabulated may be wider than measured but is given to indicate which diagnostic would be most suitable to observe in a tokamak.</caption><thead><tr><th>Ion stage</th><th>Wavelength (Å)</th><th>Machine</th><th>Year</th><th>References</th></tr></thead><tbody><tr><td>W<sup>5+</sup> – W<sup>7+</sup></td><td>180–280</td><td>EBIT-I</td><td>2015</td><td>[346]</td></tr><tr><td>W<sup>5+</sup> – W<sup>8+</sup></td><td>180–250</td><td>SH-HtscEBIT</td><td>2022</td><td>[347]</td></tr><tr><td>W<sup>7+</sup> – W<sup>8+</sup>, W<sup>10+</sup></td><td>4200–6400</td><td>SH-HtscEBIT</td><td>2015</td><td>[348]</td></tr><tr><td>W<sup>11+</sup> – W<sup>15+</sup></td><td>160–270</td><td>SH-HtscEBIT</td><td>2015</td><td>[349]</td></tr><tr><td>W<sup>26+</sup></td><td>2600–5020</td><td>SH-PermEBIT</td><td>2014</td><td>[350]</td></tr><tr><td>W<sup>28+</sup></td><td>3300–4000</td><td>SH-HtscEBIT</td><td>2014</td><td>[351]</td></tr><tr><td>W<sup>21+</sup> – W<sup>45+</sup></td><td>45–70</td><td>Berlin</td><td>2001</td><td>[352]</td></tr><tr><td>W<sup>24+</sup> – W<sup>33+</sup></td><td>15–35</td><td>CoBIT</td><td>2013</td><td>[103]</td></tr><tr><td>W<sup>39+</sup> – W<sup>47+</sup></td><td>40–200</td><td>NIST</td><td>2007</td><td>[353]</td></tr><tr><td>W<sup>40+</sup> – W<sup>50+</sup></td><td>5–6</td><td>LLNL</td><td>2004</td><td>[354]</td></tr><tr><td>W<sup>45+</sup>, W<sup>46+</sup></td><td>5.19–5.26</td><td>Shanghai</td><td>2018</td><td>[355]</td></tr><tr><td>W<sup>48+</sup> – W<sup>61+</sup></td><td>27–41</td><td>LLNL</td><td>2013</td><td>[356]</td></tr><tr><td>W<sup>64+</sup></td><td>0.9–1.77</td><td>Tokoyo</td><td>1996</td><td>[341]</td></tr><tr><td>W<sup>65+</sup> – W<sup>71+</sup></td><td>6.5–7.7</td><td>SuperEBIT</td><td>2009</td><td>[357]</td></tr></tbody></table>

transitions in the ground state term. Originally predicted in [360] that there may be a Ti-like line at 354.61 nm, thus allowing a simpler visible spectrometer to measure a highly charged tungsten ion, it has not yet been observed in a tokamak plasma. However the iso-electronic lines in krypton, molybdenum and xenon have been measured in LHD [361], although the temperatures required to ionize W<sup>52+</sup> were not achievable. The research programme at LHD has observed other M1 lines from W<sup>26+</sup> and W<sup>27+</sup> (e.g. 4f<sup>2</sup> <sup>3</sup>F<sub>4</sub> – <sup>3</sup>F<sub>3</sub> at 333.7 nm) [362] with a 44-LOS visible spectrometer. The advantages of a visible spectrometer over vacuum coupled instruments are clear for DEMO and assessing the capabilities of using such transitions should be part of the ITER physics programme.

Measuring the influx of impurities into tokamak plasmas has usually relied on observing lines in the visible spectrum and relating this photon flux to a particle flux, typically via an ionizations per photon coefficient, S/XB where S is the ionization rate, X the excitation rate and B is the branching ratio [363]. Properly the S/XB coefficient is formed from the excitation part of the collisional-radiative matrix coefficients [364]. As an example, the erosion profile of a tungsten surface caused by ELMs in DIII-D is measured by converting the emission of the W<sup>0</sup> 400.9 nm via a suitable S/XB coefficient [365]. This line, first identified by dedicated experiments at PSI-I [366] and confirmed in a tokamak setting in ASDEX Upgrade [367], has become the mainstay of tracking the influx of neutral tungsten into tokamak plasmas.

Characterizing this line has been one of the larger recent endeavors in passive spectroscopy, spanning fundamental atomic physics, numerous tokamak experiments and independent measurements of the S/XB coefficient. As seen in figure 94 the line is well isolated but this is not a 'simple' transition. The Grotrian diagram of figure 95 shows that the lower level is not the ground level and this complicates the calculation of an S/XB coefficient. There are other complexities which affect theoretical atomic calculations.

<figure><img src="image_106.png" /><figcaption>**Figure 94.** The WI line at 400.9 nm in a JET discharge during and between ELMs. The lines is well isolated but other tungsten lines are blended with lines from oxygen. Figure from A. Meigs.</figcaption></figure>

Firstly, there is some disagreement in the identification of the upper level. The NIST database shows it as 5d<sup>5</sup>(6S)6p<sup>7</sup>P<sub>4</sub> based on a *raie ultime* argument [368]. However analyzing arc spectra with different isotopes of tungsten did not show the large isotope shift expected from a pure s-p transition [369] and concluded that the upper level is >85% 5d<sup>4</sup>6s6p<sup>7</sup>P<sub>4</sub>. This agrees with later calculations [370, 371] and is also the designation in the ADAS dataset used in [365]. Secondly for W<sup>0</sup> ionization cross sections there are few experimental measurements [372] and unsatisfactory convergence in the theoretical cross sections [373] (there is better agreement for direct ionization but not from excited levels). Finally modeling, with ERO, of W<sup>0</sup> emission profiles in the PSI-2 linear device following sputtering by argon required following the evolution, and interplay, of the 5d<sup>4</sup>6s<sup>2</sup>5D and 5d<sup>5</sup>6s<sup>7</sup>S metastables [374] which indicates that there is not a simple relationship between the emission and particle influx. Practically, this unfinished theoretical understanding of the atomic physics of neutral tungsten, has resulted in empirical
<figure><img src="image_107.png" /><figcaption>**Figure 95.** Grotrian diagram for neutral Wi showing the upper and lower levels of the 400.9 nm lines used for influx measurements.</figcaption></figure>

estimations of the S/XB coefficient to apply to the 400.9 nm line [375-377].

However, such a 'rule of thumb' approach is not ideal for first principles modeling of tungsten behavior. Emission lines of neutral tungsten originating from the ground term are in the near-UV region, ~230-350 nm [378] but many of these lines connect different spin systems. This is also a difficult spectral region for diagnostics because the means for moving the optical components out of a direct LOS are not effective: UV fibers attenuate too quickly and the efficiency of reflectors is very poor at these wavelengths. Nevertheless direct observation of these lines at DIII-D and CTH [379] coupled with improved theoretical calculations [380] may be able to place the 400.9 nm line on a much more secure theoretical basis. A similar approach is underway for W⁺ [381] with new atomic data of similar quality [382].

With increasing charge state the spectral emission tends to lower wavelengths. Figure 96 shows modeled emission of the first three ionization stages of tungsten from VUV to visible (100-700 nm). The atomic structure is calculated with the HFR +CP approach of [383] which gives a reasonable accuracy of better than 5%. There are usable lines in the visible for neutral tungsten, which is confirmed in many tokamak experiments. However W⁺ is mostly in the near-UV and by W²⁺ the dominant spectral region is firmly in the VUV region with no usable visible lines. As shown in [384] it may be important to follow the ionization pathways of W⁰ → W⁺ → W²⁺ → W³⁺ to correctly determine the re-deposition of tungsten following sputtering. Measured emission from W³⁺ → W⁴⁺ may be needed for the higher densities of the ITER divertor.

Any model of the behavior of tungsten, from its influx to possible accumulation in the core, ideally requires measurements from all parts of the plasma. In practice this means

<figure><img src="image_108.png" /><figcaption>**Figure 96.** The spectral region of expected emission from W0–W2 +. The photon emissivities for prominent lines are calculated from ADAS adf04 data based on an atomic structure following the HFR +CP approach of [383]. This method has a wavelength accuracy of better than 5%. Figure from [381].</figcaption></figure>

<figure><img src="image_109.png" /><figcaption>**Figure 97.** The wide variety of shapes of a tungsten spectral feature from the horizontal viewing SPRED VUV spectrometer at JET over a campaign. The background and contaminating nickel and chromium lines have been removed.</figcaption></figure>

that emission from nearly all stages of tungsten must be used. Additionally, knowledge of the composition of these spectral features is essential. Many of these observations will be isolated lines but the quasi-continua also contain significant information. There is a prominent tungsten spectral feature at 150-250 nm which is attributed to tungsten ionization stages below W²⁺ [358]. Its emission zone temperature is lower than the λ ~ 5 nm feature, so it reacts to edge conditions, as seen in the variation of shapes seen over a JET campaign in figure 97. The partially filled 4f subshells of these stages remains a difficult atomic structure problem so precise observations are still required and recent reports from HL-2A [385] and LHD [386] have identified some components of tungsten in this spectral region; lines from W⁵⁺ and W⁵⁺ have been identified. In very low temperature, Tₑ~0.5 keV, plasmas at EAST, where tungsten accumulation is present, spectral lines from W⁵⁺–W⁸⁺ become prominent.
<figure><img src="image_110.png" /><figcaption>**Figure 98.** A part of the VUV spectrum along a line of sight looking through the outer leg of the JET divertor following a tungsten influx. The emission is mainly from WIV–WVII. Full identifications are in [389]. The lower figure shows similar tungsten stages in a low temperature (~0.5 keV) EAST plasma from a line of sight through the core. Reproduced from [389]. © The Author(s). Published by IOP Publishing Ltd. CC BY 4.0.</figcaption></figure>

Probable overlaps in the spectrum arising from stages in different spatial locations, which may be very close in the edge plasma due to large gradients in temperature, has prompted atomic structure calculations, laboratory experiments and detailed analysis of spectra from tokamaks. The line identifications from vacuum arc spectra of W$^{7+}$ between 190 and 500 Å [387, 388] have recently been observed in a divertor spectrum [389] in typical JET operational conditions. 303 tungsten lines, from W$^{3+}$–W$^{7+}$ have been identified, with a further 44 unattributed tungsten lines, in the divertor plasma of JET. Figure 98 shows a small portion of the 150–1325 Å range and detailed spectroscopy such as this is essential in building a set of observations which will be needed to fully understand the behavior of tungsten.

The design choices for the VUV spectrometers of ITER were made before the desirability of measuring emission from the first 5 stages of tungsten was apparent and are constrained by location and the wavelength response of optical elements. The Divertor Impurity Monitor (DIM) [304, 390] covers 200–1000 nm and the suite of main chamber VUV spectrometers have a number of spectral channels covering 2.4–160 nm for a single LOS of the main plasma [95], an imaging system over 17–32 nm in the upper edge plasma [97] and an imaging system over 14.4–32 nm viewing the divertor [96].

### 5.3. A common emission model for diagnostic design

Many of the passive spectroscopy diagnostics for ITER are refinements of existing technology and can be tested on existing machines, e.g. [96, 310]. However present day devices cannot achieve the same conditions as ITER so a flexible model

of the expected emission is essential when developing diagnostics. Predicted signals can then be converted into practical units to show how successfully the system can achieve the measurement requirements. Apart from helium generated from fusion, doped pellet injection and gas puffing, all other impurities will originate from plasma surface interaction. Although sophisticated codes for modeling the edge plasma are used extensively in divertor design and scenario development [391] they are computationally expensive and do not have robust models for transient events. The models also rarely address the richer mix of impurities that will be present in ITER. Furthermore, the diagnostic systems are expected to measure a much wider range of possible conditions not considered in the models.

A common emission model has been used to support the design of x-ray [309, 342], VUV [96], bolometry [72] and SXR cameras [392] diagnostics. All these systems measure an emission integrated along a LOS. In general the intensity of an atomic line from an element is,

$$I(\lambda) = \int_{\text{LOS}} N_e(\xi) [N^{z+}(\xi) P \varepsilon C^{\text{exc}}(\xi, \lambda) + N^{(z+1)}(\xi) P \varepsilon C^{\text{rec}}(\xi, \lambda)] d\xi$$

where $N_e(\xi)$ is the electron density, $N^{z+}(\xi)$ the ionization stage density and PEC the photon emissivity coefficient which is an atomic quantity depending on the local electron temperature and density [364]. There are separate PECs for excitation and recombination and the emission profile is integrated along the LOS $\xi$ of the instrument. Additionally there will be a background from the continuum emission contributed by this element and all other elements radiating in the LOS. The emission depends on the temperature and density profiles as well as the ionization balance at each point along the LOS. The ionization balance profile is determined by plasma transport but theoretical descriptions, although much improved, still give qualitative rather than quantitative predictions for these profiles [393]. Therefore for this diagnostic work, the impurity transport code used to predict spectroscopic line emission has three considered, but imposed, inputs: the time evolution of the influx, the final concentration in the confined plasma and the description of the transport coefficients. Using typical values, based on the experience of other machines and the more complete edge modeling, is sufficient for design needs because the diagnostics signals must not be marginal and adequate headroom to deal with unexpected behavior is also a design criterion.

The first use of the approach described here was for a design study of a high resolution x-ray spectrometer [394]. The temperature and density profiles were calculated with the ASTRA 1.5D transport code [395] where its self-consistent transport coefficients were adjusted to match the confinement time of what is now the ITPA global H-mode confinement database [396]. The profiles were available for one time-point in the flat-top of a high performance ($Q > 10$) DT discharge. The provenance of these background profiles, essential for diagnostic assessment, was unsatisfactory until the development of the integrated modeling & analysis suite (IMAS) [397] and the
<figure><img src="image_111.png" /><figcaption>**Figure 99.** A range of modeled temperature and density profiles for ITER operational scenarios used as the steady-state background plasma for forward modeled impurity emission in support of diagnostic design.</figcaption></figure>

policy that IMAS is the preferred way to support the development of the ITER research plan and diagnostics [398]. Some of the profiles in this work are available via IMAS and figure 99 shows the range of scenarios considered.

The different plasma scenarios will affect the profiles of the ionization stages giving rise to the emission. Note that this effect is of limited importance to the light elements since they exist at low electron temperatures and should emit irrespective of the core conditions. Figure 100 shows the change for iron, krypton and tungsten profiles between the high performance 15 MA and the half-current 7.5 MA H-mode scenarios. Using the iron H-like Fe$^{25+}$ transition as a diagnostic will work for either H-mode scenario. The increased core electron temperature brings the H-like krypton and Ne-like tungsten stages into greater prominence and hence their usefulness as a diagnostic. What is not shown in the figure is the relative abundance

<figure><img src="image_112.png" /><figcaption>**Figure 100.** Profiles of the ionization stages for krypton, iron and tungsten for 7.5 MA (top) and 15 MA (bottom) H-mode ITER plasmas. The stages which give the strong, isolated lines are highlighted.</figcaption></figure>

between the two scenarios. Figure 101 compares the emission profiles of the most prominent x-ray lines, i.e. those with one or two active electrons: the H- and He-like iron and krypton and Ni- and Ne-like tungsten detailed in table 29. The concentrations, in the confined plasmas inside the LCFS, are set to ‘typical’ values but the profiles will scale linearly with concentration. The behavior of these three elements in an L-mode scenario is markedly different in that the Ne-line W$^{64+}$ stage is not ionized and the Ni-like W$^{46+}$ becomes a core, rather than edge, emitter. The emission zone of the iron and krypton x-ray lines also move inwards but the He-like and H-like lines of these elements still have a reasonable spatial coverage. The tungsten lines show the greatest separation with plasma conditions, not
<figure><img src="image_113.png" /><figcaption>**Figure 101.** X-ray line emissivity profiles for iron, krypton and tungsten in full and half-current H-mode plasmas. The concentration is set to be the same for both scenarios.</figcaption></figure>

<figure><img src="image_114.png" /><figcaption>**Figure 102.** X-ray line emissivity profiles for krypton and xenon in full and half-current H-mode plasmas assuming that each element contributes 1 MW of radiated power in each scenario.</figcaption></figure>

unexpectedly since only the 15 MA plasma attains conditions sufficient to strongly ionize the Ne-like shell. Krypton also shows that the H-like and He-like sample different regions of the core plasma so measuring both lines will be important to ensure full coverage [399–401].

Iron and tungsten are intrinsic impurities of ITER but operational priorities are likely to minimize their pollution of the confined plasma by adjusting gas fuelling which has been shown to reduce tungsten erosion and thus penetration [402]. The resulting concentrations may not be sufficient to be usable by diagnostics. Adding a known quantity of an extrinsic impurity is one method to mitigate emission from impurities falling below the detection threshold. The noble gases offer the most control on shaping the influx either by admixture with the fuel or via a dedicated injection system [403]. The choice of which gas to use is determined by measurement requirements, operating conditions and physics and engineering constraints of the diagnostic. As an example, the x-ray spectrometers have a role in measuring the ion temperature profile. This immediately excludes argon as a suitable impurity to diagnose a wide spatial coverage. Figure 102 compares the radial extent of the original choice of krypton with xenon lines and normalizes the relative intensities by fixing the additional radiation to 1 MW, assuming that this is a reasonable non-perturbing value and that the gas introduction systems are sufficiently responsive to maintain the required concentrations. These are $5 \times 10^{-5}$ and $1.6 \times 10^{-5}$ for Kr and Xe in the 7.5 MA scenario and lower at $1.2 \times 10^{-5}$ and $3.8 \times 10^{-6}$ in the high performance plasma, highlighting the greater efficiency as a radiator of the more ionized charge stages. There are other trade-offs: the factor of 3 lower emissivity of $Xe^{51+}$ to Kr$^{34+}$ in the core ($r/a < 0.4$) can be offset by the factor of 20 higher reflectivity of a germanium crystal at 0.3 nm than at 0.09 nm. The suggestion

for considering xenon [404] has prompted measurements in exciting machines [405] and has informed the design of spectrometers for use in CFETR [406].

Impurities may have deleterious effects by diluting the fuel or, for high-Z elements, by reducing the temperature via radiation with the latter effect dominating with increasing nuclear charge. Such radiation is not always unwelcome. High performance discharges in ITER will rely on neon radiation to remove excess energy in the SOL and divertor [407] and xenon is under consideration for the same role in DEMO [408] where the dissipation needs are greater (40 MW vs 450 MW). Bolometers and SXR cameras are used to measuring and the same emission model used for spectroscopy can be used for energy-integrated radiation. The precision of modeling, and later interpreting, of the radiation measurements depends on the precision of the ionization balance and the power coefficients. For light elements the uncertainty on these atomic data is low [364] and there are efforts to rigorously quantify the uncertainty [409]. This is not the case for the high-Z elements. The choice of tungsten as a PFC for ITER also prompted new fundamental atomic cross section measurements. A storage ring measurement of the dielectronic recombination (DR) rate coefficient of $W^{20+}$ showed that low energy resonances influence the rate at temperatures of its peak abundance in hot, fusion plasmas and that existing calculations of DR underestimated the rate by a factor of 4 [410]. A decade of new measurements [411], empirical adjustments to existing rates based on tokamak ion balance studies [358] and new calculations [412–414] have converged to improve the understanding of tungsten. Including sufficient pathways from stages with open 4d and 4f shells was required for DR calculations and the complex atomic structure has large effects on the radiated power also. Systematic studies of the atomic
configurations needed so as not to underestimate the cooling function of high-Z elements [415, 416] are being used, in conjunction with spectroscopy, to diagnose the metallic plasma content of JET [417]. Figure 103 shows the advance in the theoretical abundances and radiated power of tungsten from the time of the ORMAK spectra to today. There is a relatively small spread in the total (equilibrium) power for the core ions which reduces any atomic uncertainty in the confined plasma predictions. The greatest change among the lower charge ionization stages corresponds to temperatures up to ~3 keV which spans the important SOL and pedestal region for ITER. Verifying the radiation from these stages by independent experiment is more difficult but the identification of spectra from W$^{6+}$–W$^{30+}$ in EBITs and fusion experiments will enable this [418].

The spatial profile of emission is largely determined by the local electron temperature and impurity transport. There is a lesser dependence on the local electron density through the effective atomic coefficients. For the emission model described here the evolution of an ionization stage, $n_z$, is given by the continuity equation

$$ \frac{\partial n_z}{\partial t} = -\frac{1}{\psi} \frac{\partial}{\partial \psi} (\psi \Gamma_z) + n_e (n_{z-1} S_{z-1 \to z} + n_{z+1} \alpha_{z+1 \leftarrow z}) $$

where $\psi$ is the flux surface, $n_e$ the electron density, $S$ the ionization rate and $\alpha$ the recombination rate. The particle flux is given by an empirical ansatz,

$$ \Gamma_z = -D_z \frac{\partial n_z}{\partial \psi} + \nu_z n_z, $$

where $D$ can be considered as a diffusion term and $\nu$ describes a convection process. In the absence of a robust, predictive, model for impurity transport a sensitivity analysis is used to quantify the effect of transport on the predicted emission. Figure 104 shows the modeled emission of iron x-ray and VUV lines for a set of transport coefficients which fall into an envelope of values which should include the ‘real’ value. The $D$ and $\nu$ coefficients are paired to give a set of $n_D \times n_\nu$ combinations and the influx to the simulation is adjusted to give the same concentration, $10^{-4}$, for each transport. This normalization results in a spread of total radiated power between 3.5–8.5 MW for the high performance scenario. Coronal equilibrium, corresponding to no transport, is a hard limiting case and falls within the envelope of possible emission. Since the spatial variation in the resultant emission is not of the same order of magnitude as the transport variation, the dependence of the modeled emission on transport can be taken to be not a primary driver and a generic set of coefficients may be safely used for the purposes of diagnostic design.

The typical transport used for previous emission modeling for diagnostic design, adopts a shaped diffusion profile, consistent with many experimental results, and an edge convection in the outer plasma (shown in red in figure 104). In the core where diffusion dominates ($0 < r/a < 0.7$) there is little difference from the coronal and constant diffusion models, although the radiation is 1%–5% smaller due to the lower diffusion inside $r/a \sim 0.4$. The biggest difference is as the edge

<figure><img src="image_115.png" /><figcaption>Figure 103. The fractional abundance and radiated power (cooling curve) of tungsten from the current ADAS recommendation, the empirical adjustment of ADPAK [358], FLYCHK [419] and the average ion method [329]. The dashed line shows the radiation fraction in the SXR energy band for a diagnostic with a 250 µm beryllium window and a Si-diode detector.</figcaption></figure>

and is wholly due to the inclusion of the convection term. A number of papers analyzing JET transport [420–422] show an outward pinch, or positive convection, in the outer part of the plasma, peaking at $r/a \sim 0.8$. Including this effect gives a different distribution of ionization stages, although the effect on the total volume-average power is modest.

For SXR and bolometry the mixture of impurities is an additional choice since the current designs integrate over energy and therefore cannot discriminate between different impurities. The power coefficients are filtered over any transmission function, foil absorption and reflection in bolometers [423] or combined window and detector response in SXR cameras, similar to the JET system [424]. The emission model
<figure><img src="image_116.png" /><figcaption>**Figure 104.** A transport sensitivity analysis for x-ray and VUV iron lines in a 15 MA high performance scenario. The coefficients in red are generic values used for most of the simulations. The transport was varied and the concentration normalized to $10^{-4}$.</figcaption></figure>

can be extended in the time domain to simulate sawteeth and ELMs. Figure 105 shows the predicted response of a saw-tooth crash on the SXR emission for a typical mix of ITER impurities.

### 5.4. Quasi-spectroscopy of energy-resolved SXRs

A map of the energy-resolved radiation profile, in figure 106, for a likely species mix in ITER shows the potential to extract enhanced diagnostic information if the SXR imaging system was extended in energy space, i.e. moving towards coarse spectroscopy but having a much wider energy overview than the narrow bandwidth of spectrometers. The shape of SXR emission in the core could be used to measure the temperature profile, simplifying multi-energy SXR analyses such as [425] by removing the need to disentangle SXR emission mediated

<figure><img src="image_117.png" /><figcaption>**Figure 105.** The SXR response of the emission from a plasma with 4% Be, 0.01% Fe and 0.0001% W in the steady state phase of a 13.8 MA hybrid ITER plasma. The SXR energy range is set by a 250 µm Be window and a Si-diode response [424].</figcaption></figure>

<figure><img src="image_118.png" /><figcaption>**Figure 106.** The radiation profile for a high power 15 MA ITER plasma with an expected impurity mix. The radiated power is split into 256 energy bins with the energy is distributed in bins of equal spacing in log-space.</figcaption></figure>

by different foil thicknesses. Furthermore, depending on the resolution, dominant spectral lines many be measurable.

The proposed SXR camera for ITER [313] is a conventional design optimized for a fast temporal response when measuring MHD activity, disruption precursors and ELMs. The emission is integrated over a large energy range bounded by a beryllium window which filters emission below ~2 keV and a silicon diode detector which decreases in efficiency with increasing energy from near 100% at 4 keV to ~1% at ~30 keV [426]. Similar systems are in use in present day machines and operate successfully within their, potentially detector damaging, x-ray and gamma ray environments. ITER will have much larger
intensities and detectors will be exposed for much longer
so enhanced radiation hardening may be required. There are
significant advances in detector technology so an upgrade is
enviseed before the DT phase of ITER operations.

The high x-ray luminosity of synchrotrons and particle col-
liders has led to the development of radiation hardened semi-
conductor detectors. The PILATUS3 detector from Dectris
Ltd is an array of 195 × 487 photodiode pixels built on a
450 µm n-doped silicon substrate and is sensitive to x-ray
photons between 1.6 and 30 keV. An imaging, multi-energy,
SXR pinhole camera with this detector has been demonstrated
at MST [427] where line emission and the SXR background,
and even hard x-rays from runaway processes, has been clearly
measured.

In astrophysics a spectral resolution of a few eV in the keV
energy x-ray emission is required to address certain outstand-
ing questions. Framing diagnostic needs, and the precision of
other inputs such as atomic data, in such a fashion is typical
[428]. Space borne instruments need to be radiation hard
and microcalorimeters have been developed to provide the
necessary resolution and robustness. The SXS high-resolution,
high-throughput spectrometer on the ill-fated Hitomi Satellite
had such a detector with an energy resolution of ~7 eV at
6 keV and a range of 0.3–12 keV [327], although this is tuned
to a narrower range of a few 100 eV for each observation.
Microcalorimeter detectors have been proposed for the ITER
core x-ray spectrometer [429] and were prototyped at MST
[430]. However there is a concern that the neutron flux may
cause a temperature rise in the microcalorimeter absorbers and
thus affect the measurement.

In addition to high x-ray photon fluxes, tokamaks during
DT operation will also generate a large flux of neutrons. The
preliminary tritium experiment at JET showed that Si diode
detectors would suffer neutron-induced noise to the signal
and lasting damage [431]. Operating these diodes with a 3 V
reverse bias reduced performance losses arising from radiation
damage but a pinhole arrangement with concrete shielding
[432] was required to enable successful measurements during
the first JET DT experiments [433, 434]. This configuration
was retained for the 2021 JET DT experiments. The DT neut-
ron fluence onto the JET detector (~10<sup>13</sup>ncm<sup>-2</sup> [432]) is a few
orders of magnitude lower than the fluence expected for ITER
(~10<sup>16</sup>ncm<sup>-2</sup> [173]). Therefore, SXR detectors in ITER must
also mitigate neutron damage.

A minimalist approach is to remove the detectors from a
direct LOS to behind extensive shielding but this requires a
method to guide x-rays of energies 0.1–15 keV (0.08–12 nm)
along a light guide. The SXR wavelengths have a length
scale similar to the lattice constants of crystals that can be
considered as transmission guides but polycapillary lenses,
which are thin hollow tubes of borosilicate glass, can be used
to manipulate x-rays via multiple total external reflections
at grazing incidence. However the transmission losses can
exceed >99.9% [435]. The impurity mix in figure 106 results
in a volume average power of 54 MW of which 35 MW is
in the SXR 2–25 keV energy band. This results in a flux of
52kWm<sup>-2</sup> through the LCFS so even though the transmission

<figure><img src="image_119.png" /><figcaption>**Figure 107.** The measured counts per 50 ms from a pinhole imaging camera with GEM detectors on WEST for two shots. The channels start (#1) on the high field side and terminate on the low field side (#85). Reproduced from [440]. CC BY 4.0.</figcaption></figure>

fraction of the polycapillary lens is small, the viewed flux is
substantial. A study of the ITER SXR system with polycapil-
lary x-ray transport concluded that their use in an ITER context
is feasible [436]. However this loses the many advantages of a
multi-energy measurement.

Detectors based on the interaction of tokamak originated
x-rays and pressurized gas chambers such as the low voltage
ionization chamber (LVIC) [437] and Gas Electron Multiplier
(GEM) [438] have the potential to be extremely resistant
to neutron damage. However unlike the semiconductor and
microcalorimeter based detectors, these systems may need
more post-processing to separate the photon energy into con-
venient bins.

GEM detectors operating in a photon counting mode can discriminate on the energy of the impinging photon [439]. An energy resolved measurement from an lower hybrid current drive heated WEST plasma [440] is shown in figure 107. This plasma is dominated by tungsten emission and good energy and spatial resolution has been demonstrated. The robustness of GEM detectors to neutron effects was examined in a post-mortem analysis of a test system at ASDEX Upgrade which showed that internal parts were activated by the neutron flux (DD so many orders of magnitude lower than ITER) and added noise to the signals [441]. All materials are affected to some degree but it may be possible to engineer a device for ITER to minimize this form of instrumental error.

The neutron resilience of LVIC detectors is based on their
simplicity, being a simple tube with gas under pressure (up to
5 bar). Energy resolution may be achieved by adding multiple
anodes [442]. Figure 108 shows the arrangement of 5 anodes
placed along a 150 × 2 mm chamber. The positions at 1, 3, 10,
35 and 100 mm respond at increasingly higher energies. The
response of the deeper chambers is dependent of the preceding
ones viz.,
<figure><img src="image_120.png" /><figcaption>**Figure 108.** The spectral response of an argon filled LVIC with anodes placed at 1, 3, 10, 35, and 100 mm from the entrance which is a 250 µm beryllium layer. Reprinted from [442], with the permission of AIP Publishing.</figcaption></figure>

$$ \eta_i(hv) = T(hv)A_i(hv) \prod_{k=1}^{i-1} (1 - A_k(hv)), $$

where T is the transmission of the beryllium filter and A_i the absorption of the ith chamber (defined in terms of the gas density and photoionization probability at the energy of the incoming x-ray photon $h\nu$ (section III-A of [442]). There are no measurements with multi-anode LVICs of SXR emission from tokamaks but the response of the detector to mono-energetic x-ray emission from a synchrotron has been characterized. The saturation voltage response of a prototype was measured at the Shanghai Synchrotron Radiation Facility [443]. There was good agreement with a model of the device for energies between 15 and 20 keV but the calculated responsivity was 60% overestimated at 10 keV. The results are promising but improvements are possible by reducing leakage current and choosing an optimal gas mixture. Further synchrotron experiments are planned before a prototype diagnostic is tested in a tokamak. This optimization, and future analysis, of LVIC measurements is best undertaken by constructing a synthetic diagnostic with controls over the LOS and the arrangement of the sub-chambers in each detector. This approach is relevant for all diagnostics and such work on passive spectroscopy is addressed in the next section.

## 5.5. Analysis and determining physics quantities

The measured signals must be transformed into usable quantities which for ITER are classified by their role in MP, plasma control or physics evaluation. A comprehensive set of measurement requirements details the desired measurement, resolution (spatial and temporal) and accuracy needed to achieve the measurement [4]. Different levels of sophistication in analysis are needed to achieve the various requirements from the passive spectroscopy diagnostics. For survey spectrometers just the appearance of a spectral line is sufficient to identify an impurity in the plasma. However even this most simple analysis depends on a good wavelength calibration and the tabulation of transitions, preferably previously observed in other

<figure><img src="image_121.png" /><figcaption>**Figure 109.** The photon emissivity coefficients, given as a function of the profile conditions for JET pulse #85 232 (central $T_e \sim 3.5$ keV and $N_e \sim 4 \times 10^{19} m^{-3}$) for the Ni-like $W^{46+}$ line used to determine a tungsten concentration [444].</figcaption></figure>

tokamaks. Apart from tungsten, the plasma facing materials of ITER are well characterized, and the many spectroscopic investigations of tungsten outlined in section 2, have ensured that this analysis is well prepared for ITER plasmas.

Not all spectrometers in use on present day tokamaks are calibrated for intensity. However, a reliable intensity calibration is essential for quantitative measurements and periodic re-calibration may be required to account for changes over time. Possible causes are due to mirrors or windows being coated by plasma detritus, degradation in detectors, optical fibers, cables and electronics from the x-ray and neutrons produced by the plasma (see other articles in this review).

Some of these issues are not present in vacuum-coupled systems, such as x-ray spectrometers, but the accuracy of the result is also determined by the precision of the atomic data used in the analysis. Inner shell x-ray transitions in Ni-like and Cu-like tungsten, have been used to estimate the tungsten concentration in JET ($\sim 10^{-5}$ in ELMy H-mode) [444]. That analysis used photon emissivities calculated with the relativistic distorted wave code FAC [445]. Figure 109 compares the PEC for the $W^{46+}$: $3p^63d^{101}S_0 - 3p^53d^{10}4d(3/2,5/2)J = 1$ line with Breit-Pauli distorted wave and a relativistic R-matrix with radiation damping calculation [401]. The plot follows the style of figure 4 of [444] showing the PECs across the JET plasma profile. The spread in values between the FAC and R-matrix models could be used as an error bar on this part of the analysis, which would suggest a contribution of $\sim 20\%$. However the analysis in [444] is more refined in that it also uses a similar inner shell line from $W^{45+}$ and Nelike $W^{32+}$ lines. More than one ion brings in a dependence on the ionization balance, which introduces a further uncertainty as seen in the different models of fractional abundance in figure 103. It is likely that propagating the uncertainties of the atomic data can be as important, and may be even more dominant, than considering