# Diagnostics: Chapter 8 of the special issue: on the path to tokamak burning plasma operation

D Mazon, G Vayakis, M Walsh, G Yun, S-H Hong, B Peterson, M H

Aumeunier, A Bultel, C Klepper, D Rasmussen, et al.

► To cite this version:

D Mazon, G Vayakis, M Walsh, G Yun, S-H Hong, et al. Diagnostics: Chapter 8 of the special issue: on the path to tokamak burning plasma operation. Nuclear Fusion, 2025, 65, 10.1088/1741-4326/adfc7c hal-05408558

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# Diagnostics: Chapter 8 of the special issue: on the path to tokamak burning plasma operation

To cite this article: D. Mazon *et al* 2025 *Nucl. Fusion* **65** 113001

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# Diagnostics

## Chapter 8 of the special issue: on the path to tokamak burning plasma operation

D. Mazon<fnref n="1" />,*, G. Vayakis<fnref n="2" />, M. Walsh<fnref n="2" />, G. Yun<fnref n="3" />, S.-H. Hong<fnref n="4" />, B. Peterson<fnref n="5" />,
M.H. Aumeunier<fnref n="1" />, A. Bultel<fnref n="6" />, C. Klepper<fnref n="7" />, D. Rasmussen<fnref n="7" />, H. Choi<fnref n="8" />, C. Grisolia<fnref n="1" />,
K. Kim<fnref n="8" />, S. Oh<fnref n="8" />, C. Sun<fnref n="8" />, M. Scholz<fnref n="9" />, B. Esposito<fnref n="2" />, D. Marocco<fnref n="10" />, F. Belli<fnref n="10" />,
L. Bertalot<fnref n="2" />, B. Coriton<fnref n="2" />, V. Ginv<fnref n="2" />, D. Gin<fnref n="2" />, J. Dankowski<fnref n="2" />, A. Hjalmarsson<fnref n="11" />,
V. Krasilnikov<fnref n="2" />, G. Ericsson<fnref n="11" />, M. Tardochi<fnref n="12" />, D. Rigamonti<fnref n="12" />, M. Nocente<fnref n="12" />,
M. Garcia-Munoz¹³, M. Ishikawa¹⁴, M. Cheon<fnref n="8" />, J. Jo¹⁵, S. Zoletnik¹⁶, O. Asztalos¹⁶,¹⁷,
M. Bandyopadhay¹⁸, P. Bharathi¹⁸, M. de Bock<fnref n="2" />, O. Ford¹⁹, M.G. von Hellermann²⁰,
D.W. Johnson²¹, J. Ko<fnref n="8" />, S. Menmuir²², Ph. Mertens²⁰, A.H. Nielsen²³, G.I. Pokol¹⁶,¹⁷,
S.V. Serov<fnref n="2" />, M.J. Singh¹⁸, S.N. Tugarinov²⁴, G.L. Vyas¹⁸, M.G. O'Mullane²⁵, Ling Zhang²⁶,
R. Barnsley<fnref n="2" />, R. Tieulent<fnref n="2" />, D. Colette<fnref n="2" />, V.S. Neverov²⁷, R. Scannell²⁸, H. Liu²⁶, E. Mukhin²⁹,
E. Yatsuka¹⁴, A. Gorbunov²⁷, L. Giudicotti³⁰, G. Kurskiev²⁹, J. Chen<fnref n="10" />, M.A. Van Zeeland<fnref n="4" />,
D. Finkenthal³¹, R. Imazawa³², D. Brower³³, A. Sirinelli<fnref n="2" />, T. Akiyama<fnref n="4" />, T. Carlstrom<fnref n="4" />,
M. Lesher<fnref n="4" />, C. Watts<fnref n="2" />, M. Bassan<fnref n="2" />, M. Austin³⁴,<fnref n="4" />, S.B. Korsholm³⁵, Y. Liu<fnref n="2" />,²⁶, S. Danani¹⁸,
C. Muscatello<fnref n="2" />, W.L. Rowan³⁴, V. Vershkov²⁴, G. Wang³³, J.L. Xie³⁶, M. Zerbini<fnref n="10" />, Y.L. Zhu³⁷,
I. ˇDuran³⁸, A. Gusarov³⁹, K. Vukolov²⁷, A. Litnovsky⁴⁰, L. Moser<fnref n="2" />,⁴¹, N. Babinov²⁹,
A. Dmitriev⁴¹,²⁹, B. Kim<fnref n="5" />, L. Marot⁴¹, A. Razdobarin²⁹, A. Rogov⁴², D. Samsonov²⁹,
C. Seon⁴³, K. Soni⁴¹, R. Yan²⁶, M.R. de Baar⁴⁴,⁴⁵, L. Zabeo<fnref n="2" />, M. Schneider<fnref n="2" />, Th. Blanken⁴⁴,⁴⁵,
Th. Bosman⁴⁴, T. Ravensbergen<fnref n="2" />, B. van de Boorn⁴⁴, C. Orrico⁴⁵, R. Fischer⁴⁶, A. Bock⁴⁶,
S.S. Denk<fnref n="4" />, A. Medvedeva⁴⁷, M. Salewski⁴⁸, D. Stieglitz⁴⁶,
the ASDEX Upgrade Teamª and the WEST Teamᵇ

<figure><img src="image_8.png" /></figure>

<footnote n="1">¹ CEA, IRFM F-13108, Saint Paul Lez Durance, France</footnote>

<footnote n="2">² ITER Organization, Route de Vinon sur Verdon, CS 90 046-13067 Saint Paul Lez Durance Cedex, France</footnote>

<footnote n="3">³ Department of Physics and Division of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang, Gyeongbuk 37673, Korea, Republic Of</footnote>

<footnote n="4">⁴ General Atomics, San Diego, CA, United States of America</footnote>

<footnote n="5">⁵ National Institute of Fusion Science, Toki, Japan</footnote>

<footnote n="6">⁶ CORIA, Saint-Etienne du Rouvray Cedex, France</footnote>

<footnote n="7">⁷ Oak Ridge National Laboratory, Oak Ridge, United States of America</footnote>

<footnote n="8">⁸ Korea Institute of Fusion Energy, Daejeon, Korea, Republic Of</footnote>

<footnote n="9">⁹ Institute of Nuclear Physics Polish Academy of Sciences, Krakow, Poland</footnote>

<footnote n="10">¹⁰ Associazione Euratom-ENEA sulla Fusione, Frascati, Roma, Italy</footnote>

<footnote n="11">¹¹ Department of Physics and Astronomy, Uppsala University, Uppsala, Sweden</footnote>

<footnote n="12">¹² Institute for Plasma Science and Technology, National Research Council, Milan 20125, Italy</footnote>

<footnote>ª See Zohm et al 2024 (https://doi.org/10.1088/1741-4326/ad249d) for the ASDEX Upgrade Team.</footnote>

<footnote>ᵇ http://west.cea.fr/WESTteam.</footnote>

<footnote>* Author to whom any correspondence should be addressed.</footnote>
13 Department of Atomic, Molecular and Nuclear Physics, University of Seville, 41012 Seville, Spain

14 Fusion Research and Development Directorate, Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan

15 Department of Energy Systems Engineering, Seoul National University, Seoul, Korea, Republic Of

16 Centre for Energy Research, Budapest, Hungary

17 Budapest University of Technology and Economics, Budapest, Hungary

18 ITER-India, Institute for Plasma Research, Gandhinagar, India

19 Max-Planck Institut für Plasmaphysik, Greifswald, Germany

20 Forschungszentrum Jülich, Jülich, Germany

21 Princeton Plasma Physics Laboratory, Princeton, NJ, United States of America

22 KTH Royal Institute of Technology, Stockholm, Sweden

23 Technical University of Denmark, Kgs. Lyngby, Denmark

24 Institution 'Project Center ITER', Moscow, Russian Federation

25 Department of Physics, University of Strathclyde, Glasgow G4 0NG, United Kingdom of Great Britain and Northern Ireland

26 Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, China

27 National Research Centre 'Kurchatov Institute', Moscow 123182, Russian Federation

28 UKAEA, Culham Science Centre, Abingdon OX14 3DB, United Kingdom of Great Britain and Northern Ireland

29 Ioffe Physical Technical Institute, 26 Polytechnikeskaya, 194021 St Petersburg, Russian Federation

30 Department of Physics and Astronomy, Padova University, Via Marzolo 8, 35131 Padova, Italy

31 Palomar Scientific Instruments, San Marcos, CA 92069, United States of America

32 National Institutes for Quantum and Radiological Science and Technology, Naka, Japan

33 Department of Physics and Astronomy, University of California Los Angeles, Los Angeles, CA 90095, United States of America

34 Institute for Fusion Studies, University of Texas at Austin, Austin, TX 78712, United States of America

35 Department of Physics, Technical University of Denmark, 2800 Kgs. Lyngby, Denmark

36 University of Science and Technology of China, Hefei, China

37 Department of Electrical and Computer Engineering, University of California, Davis, CA 95616, United States of America

38 Institute of Plasma Physics of the Czech Academy of Sciences, Prague, Czech Republic

39 SCK CEN, B-2400 Mol, Belgium

40 Forschungszentrum Jülich GmbH, Institut für Energie und Klimaforschung, 52425 Jülich, Germany

41 University of Basel, Klingelbergstrasse 82, CH-4056 Basel, Switzerland

42 Fusion Centre, 123182 Moscow, Russian Federation

43 National Institutes for Quantum Science and Technology, Naka, Ibaraki 801-1, Japan

44 NWO Institute DIFFER, De Zaale 20, Eindhoven, Eindhoven, Netherlands

45 Technische Universiteit Eindhoven, Eindhoven, Netherlands

46 Max-Planck-Institut für Plasmaphysik, Garching, Germany

47 Aix Marseille University, CNRS, Centrale Marseille, M2P2, Marseille, France

48 Department of Physics, Technical University of Denmark, Kgs. Lyngby, Denmark

E-mail: didier.mazon@cea.fr

Received 6 January 2025, revised 29 July 2025

Accepted for publication 18 August 2025

Published 10 October 2025

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CrossMark

Abstract

This chapter presents the activity conducted by the ITPA topical group (TG) on Diagnostics over about the last 15 years. Following a general introduction of the ITER Diagnostics led by their measurement roles, the document is organized in several subchapters detailing the design support, research and development activity conducted by each of the specialist working groups (WGs) of the TG. Please note that the magnetic diagnostics were supported at the TG without a specific WG. Their status is included in the general introduction. In the following some highlights of the subchapter's contents are provided. Recent advances in ITER first wall (FW) diagnostics for the measurements of plasma-metallic wall interaction in support of the ITER research plan are reported. An InfraRed imaging Video Bolometer for ITER has been developed and tested on several tokamaks to measure the radiated power loss. A laser-induced breakdown spectroscopy (LIBS) technique which utilizes a pulsed laser beam to ablate locally by
forming a crater, will measure local tritium inventory in the FW material. Real-time Residual Gas Analyzers will measure the neutral gas composition in a divertor port and an equatorial port during plasma operation. Due to the full metallic FW environment, the plasma-wall interaction in ITER will face several challenges such as the compromised radiated power and divertor heat flux measurements by reflection. Ray tracing and analysis codes have been developed to eliminate and correct the effects of reflection in the measurements. The characteristics of the reflecting surfaces depending on the roughness and angle of the incidence have been measured by dedicated experiments, and the results were applied to the reflection elimination. For the measurement of the metallic impurity radiation induced by eroded metallic atoms, a vacuum ultraviolet spectrometer has been developed and tested. An extensive thermonuclear diagnostic suite will be required to support the operation of ITER and the planned experimental program for future burning plasma experiments. Due to the harsh environmental conditions, the implementation of diagnostic systems in ITER is a major challenge. These conditions include high levels of neutron and gamma fluxes, neutron heating, particle bombardment. Therefore, the selection and design of diagnostic systems must take into account a number of phenomena previously unseen in diagnostic design. For this reason, the measurement of neutrons and confined or lost fast ions, with particular emphasis on alpha particles, is critical to ITER. The diagnostics associated with these measurements will be important for future plasma-burning experiments at ITER. The high neutron emission and very large plasma size in ITER make neutron diagnostics the main diagnostic method used to measure plasma parameters such as fusion power, fusion power density, ion temperature, energy of fast ions and their spatial distributions in the plasma core. Active spectroscopy techniques are methods where a neutral particle beam is injected into the plasma and information on plasma parameters is extracted from the measurement of line emission resulting from the beam-plasma interaction, either by plasma ions or by beam atoms. Spatial localization is achieved by crossing the beamline and multiple observation lines. The ITER plasma will be a high temperature, moderately dense, fully ionized collisional plasma. The plasma facing surfaces are principally metallic being fashioned from beryllium or tungsten but many other elements, arising from either structural or from operational needs, may enter this plasma. The energy range of the emitted photons range from meV (infra-red) to multi keV (x-rays) and originate from all areas of the plasma volume. The primary role of passive emission diagnostics is to identify what is in the plasma from spectral signatures. Extracting quantitative information from these measurements such as impurity content, ion temperature, rotation, degree of detachment and radiated power depends on calibrated instruments, a physics model of the atomic and molecular processes and plasma transport and an analysis workflow that takes into account environmental effects such as reflections. The particular needs for ITER have prompted a multi-machine, many-year effort to address all these aspects and this chapter reviews the work on diagnostic design, experiments and new analysis techniques. An overview of the laser diagnostics to be implemented on ITER is also provided in this paper. This includes descriptions of the Thomson scattering in the core, edge and divertor regions, polarimetry and interferometry diagnostics used for measuring plasma density and also measurements of helium density in the divertor using Laser Induced Flourescence. Techniques which can allow improvements on current measurements are also addressed in particular expanding poloidal polarimetry measurements to measure field fluctuations and proposed use of dispersion interferometery which has a number of advantages over existing methods. This paper identifies particular areas where further research and testing on existing tokamaks is useful even at this advanced stage to inform the design of diagnostics for ITER. Outstanding areas of concern for the implementation of laser diagnostics, in particular with a view to reliable operation are identified. An overview of the latest developments of microwave diagnostic systems and techniques is given. The primary focus is the contributions for ITER—the next step burning plasma experiment—which is supplemented by describing recent progress of techniques applicable for fusion experiments beyond ITER. The contributions are intentionally kept concise, and are being supplemented by a rich list of references for further studies. Radiation induced effects are receiving continuous and well-deserved attention of the ITER diagnostic community and they are in many cases one of the primary design drivers of the ITER diagnostic systems. The paper summarizes recent progress in this area focusing primarily on the ITER diagnostics but in some cases provides also outlook for the possible solutions for even more demanding radiation environment of fusion reactors beyond ITER. Despite advancements in the area of modeling and simulation of various radiation induced effects,
experimental testing in a nuclear environment as close as possible to the target one is still seen
as unavoidable for proper qualification of particular diagnostic functional elements. Recent
advancement within three diagnostic areas: optical diagnostics, magnetics and bolometers is
covered. Encouraging results on qualification of silica glass vacuum window assemblies are
presented. In the area of magnetic sensors, progress of irradiation tests performed on ITER
in-vessel LTCC inductive sensors is presented with outlook for novel technological approaches
to inductive sensors utilizing thick printing and photolithography technologies being
highlighted. Summary of advancements in the area of steady state magnetic field sensors based
on Hall effect is given. New results of neutron irradiation test of the ITER borosilicate glass
inserts for vacuum electrical feedthroughs are summarized finding negligible swelling at target
level of neutron fluence. Off-line irradiation tests of fiber optic current sensors for plasma
current measurement demonstrated that both for gamma doses up to 5 MGy and a total neutron
fluence up to 10¹⁵ cm⁻², radiation induced changes are still compatible with required
measurement accuracy on ITER. The ITER bolometers are given as an example how
considering radiation effects may influence the diagnostic design. Finally, outlook for future
main R&D directions is outlined. All optical and laser-based diagnostics in ITER will be using
mirrors to guide plasma radiation toward detectors, cameras and sensors. In the hostile plasma,
radiation and particle environment the optical characteristics of diagnostic mirrors will degrade
directly affecting the entire performance of involved diagnostic systems. An assessment of
factors affecting mirror performance is provided. Among the prime adverse factors are
deposition of plasma impurities, sputtering of mirror surface and steam ingress in the vicinity of
mirrors. Within the International Tokamak Physics Activity with active support by ITER central
team and domestic agencies, the structured research and development (R&D) program on
mitigation of risks for diagnostic mirrors is underway. Within this program the mirror material
development, the passive mitigation of mirror degradation by using diagnostic ducts and
shutters along with an active mirror recovery program comprising the *in-situ* mirror cleaning
and calibration is underway. Recent developments in diagnostic mirror R&D are described in
this Chapter along with an example of their implementation of R&D solutions in ITER Infrared
Thermography diagnostic. An assessment of still open engineering and physics questions,
considerations on mirror risks during an early phase of ITER operation are given along with an
overview of diagnostic mirror evolution in the late ITER operation stage toward the
demonstration fusion power plant. Several crucial areas of diagnostic R&D outlined in ITER
Research Plan are addressed. The basic control groups in a fusion reactor can be broken-down
in five categories: (1) plasma position, magnetic configuration, and plasma current control, (2)
profile control and confinement optimization, (3) MHD control and suppression, (4) edge
dissipation control, radiation and plasma exhaust control and (5) break-down optimization.
These categories are coupled via the physics (a control action in one domain will affect the other
domains) and via shared actuators (e.g. ECRH for impurity accumulation avoidance, current
density distribution control and MHD suppression). Consequently, a supervisory control system
should determine the priority of the various control tasks, their couplings, and the interfaces
with the safety and interlock system. For the systematic development of the various controllers
taking the complexity of the plasma and the control system into account, a model-based
approach is required. A short historical overview is given of the developments in systems and
control theory and control engineering with special emphasis on those developments that are
most relevant for Nuclear Fusion research and operation. An overview is given of the state of
the field of fusion plasma control for the control categories. It will be shown how synthetic
diagnostics are being developed in ITER and how they are used in diagnostic design and design
validation and how they can be in model-based controller synthesis using relatively simple
models. In modern control methods, multiple diagnostics are used to constrain relatively simple
models. The constrained models provide an estimate for the state. This opens the route to state
controllers, such as model predictive control. A major challenge in nuclear fusion research is the
coherent combination of data from heterogeneous diagnostics and modeling codes for machine
control and safety as well as physics studies. Measured data from different diagnostics often
provide information about the same subset of physical parameters. Additionally, information
provided by some diagnostics might be needed for the analysis of other diagnostics. A joint
analysis of complementary and redundant data allows, e.g. to improve the reliability of
parameter estimation, to increase the spatial and temporal resolution of profiles, to obtain
synergistic effects, to consider diagnostics interdependencies and to find and resolve data
inconsistencies. Physics-based modeling and parameter relationships provide additional information improving the treatment of ill-posed inversion problems. A coherent combination of all kind of available information within a probabilistic framework allows for improved data analysis results. The concept of integrated data analysis (IDA) in the framework of Bayesian probability theory is outlined and contrasted with conventional data analysis. Components of the probabilistic approach are summarized and specific ingredients beneficial for data analysis at fusion devices are discussed.

This paper is part of the Special Issue: *On the Path to Tokamak Burning Plasma Operation: A collection of papers prepared by the ITPA Topical Physics Groups reviewing progress in the development of the physics basis for burning plasma operation*.

Keywords: diagnostics, ITER, radiated power loss, neutron irradiation, diagnostic mirrors, real-time, plasma control, integrated data analysis

(Some figures may appear in colour only in the online journal)

## Contents

<table><tr><td>1. Introduction to diagnostic measurement in ITER</td><td>6</td></tr><tr><td>2. Preparation for the measurements of plasma-metallic wall interaction in support of the ITER research plan: divertor power flux, impurity, and tritium retention</td><td>29</td></tr><tr><td>3. Status of the diagnostics of fusion products at ITER</td><td>39</td></tr><tr><td>4. Progress in ITER diagnostics, active spectroscopy</td><td>63</td></tr><tr><td>5. Passive emission diagnostics for high performance metal tokamaks</td><td>78</td></tr><tr><td>6. Laser aided diagnostics</td><td>93</td></tr><tr><td>7. Microwave diagnostics</td><td>110</td></tr><tr><td>8. Influence of radiation effects on the refinement of the ITER design</td><td>124</td></tr><tr><td>9. Diagnostic mirrors for ITER: status and perspectives</td><td>134</td></tr><tr><td>10. Synthetic diagnostics for real time plasma control studies and diagnostic design</td><td>150</td></tr><tr><td>11. Integrated data analysis and validation</td><td>161</td></tr><tr><td>References</td><td>178</td></tr></table>
# 1. Introduction to diagnostic measurement in ITER

Didier Mazon¹, George Vayakis², Michael Walsh² and Gunsu Yun³

## 1.1. Introduction

This summary of the state of burning plasma diagnostics by the International Tokamak Physics Activity (ITPA) Diagnostics topic group (TG) finds ITER and its diagnostics mid-construction. This is a marked change from previous records of the state of ITER and burning plasma measurements and diagnostics. The 1995 Varenna conference saw the first airing of the ITER diagnostic set and its requirements in an integrated form [1]. By the end of the first ITER Engineering Design Activity phase [2, 3] these were consolidated and are visibly correlated with the present designs and requirements. The ‘Progress in the ITER Physics Basis’, in 2007, summarized the ITER diagnostic design as adopted for construction [4]. Since then, diagnostic design and R&D has continued within the ITER design teams with the support of the ITPA, to the point where the details of the design solution and the expected performance are mostly clear. The ITPA diagnostic TG initiated, coordinated and reviewed extensive R&D efforts to support this, through high priority items (HP), Joint Experiments (JEX) and Actions that are summarized in this introductory subchapter and with results elaborated in the subchapters that follow. The adoption of the phased approach to ITER operation [5] resulted in the phasing of diagnostic installation, nevertheless work is ongoing across all systems. In parallel, R&D is still ongoing, as part of the diagnostic supply and as a voluntary effort to support the identified needs in diagnostics.

This subchapter introduces updated roles and requirements for ITER measurements and a summary of the design and R&D basis for these with reference to the subchapters that follow. For magnetics-related measurements, it contains an extensive section on the related diagnostic progress as this topic is not covered elsewhere. Overall, this subchapter summarizes the progress of the key issues identified from the perspective of the ITPA Diagnostics TG, but including progress achieved as part of the ITER construction programme where appropriate.

## 1.2. Requirements for plasma measurements

Plasma measurement in ITER is organized in measurement groups, evolved from the original groupings of [1] to reflect partly a functional (e.g. divertor operation), partly a technique (e.g. passive spectroscopy) and partly a physical process (e.g. halo current) taxonomy. In tracking progress and coverage of the measurement needs, the specific measurement parameters (MPs) of table 10 have proven more useful than the groups. This table lays out for each parameter the key conditions and performance expected. It is not exhaustive: measurement group 2, plasma position and shape, for example, could list many more parameters and more sophisticated versions of the ones shown. The spirit of this specification is that a state-of-the-art diagnostic set designed to meet it should

ensure adequate performance for machine operation parameters related to the one shown.

This table has evolved over time. A full comparison of table 10 to the version used in [4] would show numerous detailed changes in parameter specifications. No ‘To be determined’ (TBD) values remain in the table. Resolutions are now mostly adapted to the final ITER plasma size. In addition, there are some new groups and requirements on dust, tritium and stray electron cyclotron heating (ECH) monitoring as well as the first safety related plasma measurement requirements, on fusion power and total toroidal current.

### 1.2.1. MP roles in ITER.
Table 10 defines the specification for the ITER diagnostics. ITER diagnostics designs aim to meet it, sometimes individually, sometimes in combination. In parallel, they target many other requirements including safety, maintainability, reliability and availability. The ITER environment strongly influences the last two, both close to the plasma (particle and radiation emission) and further away (nuclear effects due to neutrons and gammas, access limitations, etc) [6]. The work of the ITPA for radiation effects (section 3.7) and first mirror selection (section 3.8) directly supports the resolution of these reliability and availability issues.

Engineering constraints aside, the measurement role drives the specifications and design of the diagnostic equipment. ITER classifies the *measurement roles* with respect to the importance of each parameter to plasma control in four groups: machine protection (MP), basic control (BC), advanced control (AC) and physics operation (PHY). At the same time, working to meet the measurement specification for each parameter, *diagnostic contributions* can be:

* Primary: the contribution (range, resolution accuracy etc) of the diagnostic meets the parameter specification.

* Backup: the contribution of the diagnostic is a good match to the parameter but there is a limitation (for example, the diagnostic has to be reconfigured to be a good match compromising some other function, or the range of the parameter is not fully covered) which means it is not intended to be routinely used this way.

* Supplementary: the contribution of the diagnostic appears a poor match to the parameter but none the less essential. Examples include calibration of a diagnostic that is primary contributor or having excellent accuracy over a small part of the range.

ITER diagnostic design has, in this context, tried to achieve at least one primary contributor to each parameter, ideally two (to ease availability needs). The tables of section 3, below summarize the primary contributors for each parameter and for each of the measurement areas.

## 1.3. Progress in the basis for plasma measurements

Work in the ITPA diagnostics topic group has been organized around specialist working groups (WGs) that have looked after, in the first instance, functional groups of diagnostics with
similar techniques and/or plasma parameters. Progress relevant to ITER research and constructions is summarized for each group in sections 3.1–3.6 below with more details in corresponding subchapters.

In addition to these groups experts are working on issues across multiple systems: radiation effects, first mirrors, real-time diagnostics, and integrated data analysis (IDA). Sections 3.7–3.10 summarize their progress.

Together these groups, comprising several hundred participants, have worked directly on most of the ITER systems and MPs. An exception is the set of magnetic diagnostics that, because of its tight integration both with the core of ITER and plasma control, has been directly managed by the ITER Organization and the EU Domestic Agency, nonetheless strongly benefitting from ITPA support from the works of the cross-functional groups, in particular the radiation effects group. Section 3.11 summarizes recent progress in this area.

### 1.3.1. First wall (FW) measurements.
The ITER FW is metallic and subject to intense fluxes of all kinds. Its surface will evolve during operation, eroding, accruing deposits, absorbing particles and consequently changing its optical, mechanical and thermal properties. The erosion products in turn become a source of deposits to the wall, but also to first mirrors and to dust accumulating on the floor of the machine together with co-deposited tritium. The ITER design includes a number of diagnostics aimed at the related parameters, either of the FW itself or the plasma areas and fluxes directly affecting it (table 1).

The dedicated subchapter 2 reports on recent advances in ITER FW diagnostics for the measurements of plasma-metallic wall interaction in support of the ITER research plan. An InfraRed imaging Video Bolometer (IRVB) for ITER has been developed and tested on several tokamaks to measure the radiated power loss. A laser-induced breakdown spectroscopy (LIBS) technique that utilizes a pulsed laser beam to ablate locally forming a crater, will measure local tritium inventory in the FW material. Real-time residual gas Analyzers (DRGAs) will measure the neutral gas composition in a divertor port and an equatorial port during plasma operation. For the measurement of the metallic impurity radiation induced by eroded metallic atoms, a vacuum ultraviolet (VUV) spectrometer has been developed and tested.

Due to the full metallic FW environment, the plasma-wall interaction in ITER will face several challenges disentangling reflections from radiated power and divertor heat flux measurements. Ray tracing and analysis codes have been developed to eliminate and correct the effects of reflection in the measurements. The characteristics of the reflecting surfaces depending on the roughness and angle of the incidence have been measured by dedicated experiments, and the results were applied to the reflection elimination.

### 1.3.2. Fusion product (FP) measurements.
For ITER, FP diagnostics contribute to a large number of parameters in a primary role, with the fusion power and fluence being key

indicators of performance. Table 2 shows the primary contributions. Key backup and supplementary (supporting) contributions are detailed in the dedicated subchapter 3.

In the 2007 summary [4], the performance against parameter 014 (n- and α-source profile) was listed as not yet known and the progenitor of parameters 071 & 072 (alpha particle loss) was listed as expected not to be met. In the intervening years, much progress has been made on neutron calibration with dedicated workshops. Better design definition of the calibration and neutron activation systems (NASs) means the measurement performance for the neutron source profile is now known. Gamma ray spectroscopy is also being developed to track the spatial profile of the fast ions directly. As regards alpha particle loss, a dedicated fast ion loss detector is now under detailed design, supported by studies reported in the subchapter 3, but additional work is needed to assess the radiation effects on its scintillator and develop alternatives if it turns out that the lifetime is insufficient.

Additional progress reported in the subchapter includes an integrated design for a high-resolution neutron spectrometer (HRNS) addressing parameter 020 ($n_T/n_D$ in the core) with simultaneous capability for $T_i$ measurements.

### 1.3.3. Active spectroscopy.
ITER uses several diagnostics that rely on beams injected into the plasma interacting with it and emitting light (by plasma ions or beam atoms), from which plasma parameters are extracted from the line emission. Using the charge exchange (CX), beam-emission (BES) and motional Stark effect (MSE) spectroscopic techniques they can extract $Z_{eff}$ for BC, plasma flow and impurity profiles, including He ash for AC as well as information on $Z_{eff}$, $q$ and $T_i$ profiles for physics. For ITER, active spectroscopy diagnostics use both the heating beams and a dedicated diagnostic neutral beam (DNB), but have to operate in a different environment to present machines due to the large size and high density of the ITER plasma. The dedicated subchapter 4 presents the details of their planned implementation and performance. In addition to the primary contributions of these systems (table 3), the subchapter presents opportunities for additional measurements of fast-ion CX for confined and escaping alphas as well as $^3$He-minority heating and slowing down beam ions. It also includes a detailed assessment, including synthetic diagnostic simulation of BES, of the measurement possibilities for edge fluctuation measurement. Finally, it summarizes the state of development of the dedicated 100 keV DNB.

### 1.3.4. Passive spectroscopy.
The primary role of passive emission diagnostics is to identify what is in the plasma from spectral signatures. Extracting quantitative information from these measurements such as impurity content, ion temperature, rotation, degree of detachment and radiated power depends on calibrated instruments, a physics model of the atomic and molecular processes and plasma transport and an analysis workflow that takes into account environmental effects such as reflections. The list of instruments and their primary contributions to the ITER parameter set are listed in table 4.
<table><caption>**Table 1.** Primary contributions of the first wall related diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td rowspan="10">D1: Bolometers</td><td>MP</td><td>009a: P<sub>rad</sub> total</td></tr><tr><td>BC</td><td>007a: P<sub>rad</sub> total, divertor<br>008a: P<sub>rad</sub> total, main plasma<br>083a: P<sub>rad</sub> profile, main plasma + upper X-point</td></tr><tr><td>AC</td><td>008b. P<sub>rad</sub>, top region (upper X-point)<br>082a. Divertor Prad Profile</td></tr><tr><td>PHY</td><td>007b: P<sub>rad</sub> total (transient), divertor<br>008c: P<sub>rad</sub> total (transient), main plasma + upper X-point<br>009b: P<sub>rad</sub> (transient) total<br>082b: Divertor P<sub>rad</sub> (transient) Profile<br>083b: P<sub>rad</sub> profile (transient), main plasma + upper X-point<br>083c: P<sub>rad</sub> vertical line integrals (transient) main plasma + upper X-point</td></tr><tr><td>MP</td><td>039: Max. surface temperature, divertor<br>042: Surface luminance, FW</td></tr><tr><td>BC</td><td>043: Surface temperature, FW<br>079: Γ<sub>d</sub>, Γ<sub>t</sub></td></tr><tr><td>AC</td><td>085: Power load<br>086: Surface temperature</td></tr><tr><td>PHY</td><td>034: E<sub>max</sub> runaway<br>035: I runaway<br>044: Surface temperature during ELMs, FW</td></tr><tr><td>G3: Pressure gauges</td><td>BC</td><td>038: Gas pressure, divertor (P<sub>div</sub>)<br>046: Gas pressure, main (P<sub>main</sub>)<br>048: Gas pressure, duct (P<sub>duct</sub>)</td></tr><tr><td>G4: Residual gas analyzers</td><td>BC</td><td>037: Gas composition (Fuel, He, impurities) (P<sub>div</sub>)<br>045: Gas composition (Fuel, He, impurities), main (P<sub>main</sub>)<br>047: Gas composition (Fuel, He, impurities), duct (P<sub>duct</sub>)</td></tr><tr><td>PHY</td><td>099a. Divertor Surface H, D, T Concentration (Inner baffle)</td></tr><tr><td>G6: IR thermography (divertor)</td><td>AC</td><td>085: Power load<br>086: Surface temperature</td></tr><tr><td rowspan="2">G7: Langmuir probes</td><td>AC</td><td>080b: Divertor Target Parallel Ion Flux</td></tr><tr><td>PHY</td><td>080a: n<sub>e</sub>, divertor target<br>081: T<sub>e</sub>, divertor target</td></tr><tr><td>G8: Erosion monitor</td><td>PHY</td><td>040a. Fine surface metrology—Divertor</td></tr><tr><td>G9: Dust monitor</td><td>PHY</td><td>097a: Surface concentration of mobilizable dust (divertor cassette body under the dome and VV floor)<br>097b: Dust size distribution<br>097c: Viewing of dust area on and under the divertor</td></tr><tr><td rowspan="5">GA: IR cameras: vis/IR TV (Upper)</td><td>MP</td><td>039: Max. surface temperature, divertor<br>042: Surface luminance, FW</td></tr><tr><td>BC</td><td>043: Surface temperature, FW<br>079: Γ<sub>d</sub>, Γ<sub>t</sub></td></tr><tr><td>AC</td><td>085: Power load<br>086: Surface temperature</td></tr><tr><td>PHY</td><td>044: Surface temperature during ELMs, FW</td></tr><tr><td>PHY</td><td>099a. Divertor surface H, D, T Concentration (Inner baffle)</td></tr><tr><td>GC: Tritium monitor</td><td>PHY</td><td>099a. Divertor surface H, D, T Concentration (Inner baffle)</td></tr><tr><td>GD: FW samples</td><td>PHY</td><td>040b. Fine net erosion and re-deposition—first wall<br>099b. First wall surface H, D, T Concentration</td></tr></tbody></table>
<table><caption>**Table 2.** Primary contributions of the fusion product diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td>B1: Radial neutron camera</td><td>AC<br>PHY</td><td>014: Neutron- and α-source profile<br>013: Fusion power density</td></tr><tr><td>B2: Vertical neutron camera</td><td>AC<br>PHY</td><td>014: Neutron- and α-source profile<br>013: Fusion power density</td></tr><tr><td>B4.A0: Neutron flux monitor (NFM) DT Eq#08</td><td>BC</td><td>012: Fusion power</td></tr><tr><td>B4.B0: NFM DT Eq#17</td><td>BC</td><td>012: Fusion power</td></tr><tr><td>B4.C0: NFM DD Eq#01</td><td>BC</td><td>012: Fusion power</td></tr><tr><td>B7: Radial gamma ray spectrometers</td><td>PHY</td><td>034: E<sub>max</sub> runaway<br>035: I runaway<br>068: Alpha density profile</td></tr><tr><td>B8: neutron activation system</td><td>BC</td><td>077: First wall neutron fluence</td></tr><tr><td>B9: Lost alpha monitor</td><td>AC<br>PHY</td><td>072: Fast particle loss flux<br>071: Fast particle loss distribution</td></tr><tr><td>BB: High resolution neutron spectrometer</td><td>BC</td><td>020: n<sub>T</sub>/n<sub>D</sub> core (Integral)</td></tr><tr><td>BC: Divertor neutron flux monitors</td><td>MP<br>BC</td><td>015: Total neutron flux<br>012: Fusion power</td></tr><tr><td>BD: Vertical gamma ray spectrometers</td><td>PHY</td><td>034: E<sub>max</sub> runaway<br>035: I runaway<br>068: Alpha density profile</td></tr><tr><td>E8: Neutral particle analyzer</td><td>BC<br>PHY</td><td>020: n<sub>T</sub>/n<sub>D</sub> Core (Integral)<br>076: n<sub>T</sub>/n<sub>D</sub>, edge<br>075: n<sub>H</sub>/n<sub>D</sub>, edge<br>094: n<sub>H</sub>/n<sub>D</sub>, core</td></tr></tbody></table>

<table><caption>**Table 3.** Primary contributions of the active spectroscopy diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td>E1: CXRS based on DNB (Core)</td><td>BC<br>AC<br>PHY</td><td>029: &lt;Z<sub>eff</sub>&gt;<br>019: v<sub>TOR</sub><br>064: Core T<sub>i</sub> profile<br>066: n<sub>He</sub>/n<sub>e</sub> profile, core<br>067: <sup>3</sup>He concentration profile<br>074: Fractional content profile, Z &le; 10<br>059: Z<sub>eff</sub> profile</td></tr><tr><td>EB: MSE based on heating beam</td><td>AC<br>PHY</td><td>057: r (q = 1.5,2)/a<br>058: r (qmin)/a<br>056: q profile</td></tr><tr><td>EC: CXRS based on DNB (Edge)</td><td>BC<br>AC<br>PHY</td><td>029: &lt;Z<sub>eff</sub>&gt;<br>018: v<sub>POL</sub><br>019: v<sub>TOR</sub><br>067: <sup>3</sup>He concentration profile<br>074: Fractional content profile, Z &le; 10<br>059: Z<sub>eff</sub> profile<br>065: Edge T<sub>i</sub> profile</td></tr><tr><td>EF: Beam emission spectroscopy</td><td>PHY</td><td>065: Edge T<sub>i</sub> profile</td></tr></tbody></table>
<table><caption>**Table 4.** Primary contributions of the passive spectroscopy diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td>E3: VUV survey</td><td>BC</td><td>022: Be, C, O rel. conc.<br>024: Cu rel. conc.<br>026: Extrinsic (Ne, Ar, Kr) rel. conc.<br>028: W rel. conc.</td></tr><tr><td rowspan="3">E4: Impurity influx monitor (Div. Vis/UV)</td><td>MP</td><td>078: Γ<sub>Be</sub>, Γ<sub>c</sub>, Γ<sub>w</sub></td></tr><tr><td>BC</td><td>041: Position of the ionization front, divertor</td></tr><tr><td>PHY</td><td>079: Γ<sub>d</sub>, Γ<sub>t</sub><br>092: T<sub>i</sub> profile, divertor</td></tr><tr><td rowspan="4">E5: Core imaging x-ray spectrometer (CIXS)</td><td rowspan="4">AC</td><td>018: v<sub>POL</sub></td></tr><tr><td>019: v<sub>TOR</sub></td></tr><tr><td>064: Core T<sub>i</sub> profile</td></tr><tr><td>073: Fractional content profile, Z > 10</td></tr><tr><td rowspan="5">E7: Radial x-ray camera</td><td>MP</td><td>017: B<sub>theta</sub>(complex, at wall)/<B<sub>p</sub>></td></tr><tr><td rowspan="4">PHY</td><td>032: ELM temperature transient</td></tr><tr><td>034: E<sub>max</sub> runaway</td></tr><tr><td>035: I runaway</td></tr><tr><td>083b: P<sub>rad</sub> profile (transient), main plasma + upper X-point</td></tr><tr><td rowspan="10">ED: X-ray crystal spec survey</td><td rowspan="5">MP</td><td>021a: C,O influx</td></tr><tr><td>021b: Be influx</td></tr><tr><td>023: Cu Influx</td></tr><tr><td>027: W influx</td></tr><tr><td>022: Be, C, O rel. conc.</td></tr><tr><td rowspan="5">BC</td><td>024: Cu rel. conc.</td></tr><tr><td>025: Extrinsic (Ne, Ar, Kr) influx</td></tr><tr><td>026: Extrinsic (Ne, Ar, Kr) rel. conc.</td></tr><tr><td>028: W rel. conc.</td></tr><tr><td>034: E<sub>max</sub> runaway</td></tr><tr><td>035: I runaway</td></tr><tr><td>EG: Divertor VUV spectroscopy</td><td>MP</td><td>078: Γ<sub>Be</sub>, Γ<sub>c</sub>, Γ<sub>w</sub></td></tr><tr><td>EH: VUV edge imaging</td><td>AC</td><td>073: Fractional content profile, Z > 10<br>074: Fractional content profile, Z ≤ 10</td></tr><tr><td rowspan="2">EI: X-ray crystal spectroscopy edge imaging</td><td>AC</td><td>018: v<sub>POL</sub></td></tr><tr><td>PHY</td><td>065: Edge T<sub>i</sub> profile</td></tr></tbody></table>

The ITER plasma will be a high temperature, moderately dense, fully ionized collisional plasma. The plasma facing surfaces are principally metallic being fashioned from beryllium or tungsten but many other elements, arising from either structural or from operational needs, may enter this plasma. The energy range of the emitted photons range from meV (infrared) to multi keV (x-rays) and originate from all areas of the plasma volume. The particular needs for ITER have prompted a multi-machine, many-year effort to address all these aspects and the dedicated subchapter 5 reviews the work on diagnostic design, experiments and new analysis techniques.

**1.3.5. Laser-aided measurements.** This group of measurements has a wide number of primary roles, in BC of the main plasma density and divertor He concentration, AC of

temperature, density and q profile parameters in the core and edge as well as physics information on divertor density and temperature (table 5).

The specific subchapter 6 provides an overview of these laser-aided diagnostics. The subchapter notes significant progress across these systems, from new components (such as lasers, spectrometers) to new techniques (e.g. LIQ) and methods to work at high temperature (polarimetry corrections, polarization TS). It also addresses techniques that can allow improvements on current measurements, in particular expanding poloidal polarimetry measurements to measure field fluctuations and the proposed use of DI, which has a number of advantages over existing methods. This paper identifies particular areas where further research and testing on existing tokamaks is useful, even at this advanced stage, to inform the design of diagnostics for ITER. It also identifies outstanding
<table><caption>**Table 5.** Primary contributions of the laser-aided diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td rowspan="2">C1: Core plasma Thomson scattering</td><td>AC</td><td>052: Core T<sub>e</sub> profile</td></tr><tr><td>AC</td><td>054: Core n<sub>e</sub> profile</td></tr><tr><td rowspan="2">C2: Thomson scattering—edge</td><td>AC</td><td>055: Edge n<sub>e</sub> profile</td></tr><tr><td>PHY</td><td>053: Edge T<sub>e</sub> profile</td></tr><tr><td rowspan="2">C4: Thomson scattering (Divertor, Outer)</td><td>PHY</td><td>090: n<sub>e</sub> Divertor</td></tr><tr><td>PHY</td><td>091: T<sub>e</sub>, divertor</td></tr><tr><td rowspan="2">C5: Toroidal interferometer/ Polarimeter</td><td>BC</td><td>011: ∫n<sub>e</sub> dI/∫dl</td></tr><tr><td>PHY</td><td>063: TAE δN/n, δT/T</td></tr><tr><td rowspan="4">C6: Polarimeter poloidal</td><td>AC</td><td>057: r (q = 1.5,2)/a</td></tr><tr><td>AC</td><td>058: r (q<sub>min</sub>)/a</td></tr><tr><td>PHY</td><td>056: q profile</td></tr><tr><td>PHY</td><td>056: q profile</td></tr><tr><td>EA: Laser-induced fluorescence</td><td>BC</td><td>087: n<sub>He</sub>, divertor</td></tr><tr><td rowspan="2">FA: Density interferometer polarimeter (DIP)</td><td>BC</td><td>011: ∫n<sub>e</sub> dI/∫dl</td></tr><tr><td>PHY</td><td>090: n<sub>e</sub> profile, divertor</td></tr></tbody></table>

areas of concern for the implementation of laser diagnostics, in particular with a view to reliable operation.

### 1.3.6. Microwave measurements.
Microwave diagnostics are routinely used for core and edge density and temperature profile measurements and on ITER they certainly aim to fulfill that role. In comparing table 6 with the state of this part of the ITER measurement suite in 2007 [4] some key differences are apparent. The plasma position reflectometer (PPR) is absent. It proved impossible to integrate the number of sightlines sufficient to correct the full separatrix (around nine are needed) and, at the same time, progress in magnetics R&D described in section 1.3.11 was sufficient to enable hour-long pulses. A new system, the in-vessel ECH detectors, essentially bolometers for the mm-wave range, measures stray ECH using hundreds of distributed sensors. The dedicated subchapter 7 lays out the design progress with this and all key microwave systems: There have been advances in microwave systems which all lead to more accurate and reliable measurements of key plasma parameters. Refinements in theory, techniques, and hardware bring the diagnostics to full readiness for burning plasma conditions. The systems have been hardened against damage due to intense stray ECH and all the long transmission lines have been optimized for transmission, including the effects of atmospheric humidity. Bursting emission from low-collisionality plasmas remains an operational risk, however.

### 1.3.7. Radiation effects.
The specialized WG on radiation effects has been working on radiation related issues affecting multiple groups of diagnostics, steadily since the inception of the ITPA. In the ITER case, radiation affects primarily ceramics and glasses to the boundary of the vacuum vessel (VV) and all manner of composite materials outside it. Since the previous update, the work of the group has shifted from reporting on basic single material studies and mineral insulated cable

assemblies [4] to composite devices functionally similar to those used in ITER and exposed to radiation. These include windows (diffusion-bonded silica metal assemblies), bolometers, coils and hall probes (multi-metal and glass/ceramic layers), optical fibers (concentric glasses and metal or polymer sheath construction) and feedthroughs (metal and glass composite assemblies). The dedicated subchapter 8 summarizes the progress in this area.

### 1.3.8. First mirror selection.
This topic has generated its own specialized WG, as the material aspects affecting the first mirrors exposed to plasma and radiation are similar across multiple systems. The group strived to produce a guide to the diagnostics designer in the optical domain to allow mirror selection taking into account plasma impurities, sputtering of the mirror surface and, more recently, steam ingress from coolant leaks inside the device. The recipe to maintain good reflectivity includes good placement and surround design, material selection for endurance of the conditions and active recovery features including *in situ* cleaning. This last topic is particularly active.

The dedicated subchapter 9 gives a concise overview of the main results of the structured and coordinated international R&D in the field of diagnostic mirrors and outlines remaining work to resolve open issues for ITER and to prepare the work for DEMO.

### 1.3.9. Real time operation of diagnostics.
ITER has long prioritized diagnostic measurements with respect to control, and table 10: Specification for plasma measurements on ITER, reflects this philosophy. The implementation of the relevant controllers is within the plasma control system (PCS) and indeed plasma control has its own dedicated topic group and Subchapter 10. However, the preparation of models relies on good models of the diagnostics and the preparation of
<table><caption>**Table 6.** Primary contributions of the microwave diagnostics by operational role and parameter.</caption><thead><tr><th>System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td>C7: Collective Thomson scattering</td><td>PHY</td><td>068: Alpha density profile<br>069: Alpha energy spectrum<br>070: p, D, T, <sup>3</sup>He energy spectrum</td></tr><tr><td>F1 ECE</td><td>AC</td><td>052: Core T<sub>e</sub> profile<br>061: NTM δT/Te. (complex; 100 ms integration time)</td></tr><tr><td>F2: Reflectometer (Main Plasma, LFS)</td><td>AC<br>PHY</td><td>055: Edge ne profile<br>031: ELM density transient<br>063: TAE δn/n, δT/T</td></tr><tr><td>F9: Reflectometer (Main Plasma, HFS)</td><td>AC<br>PHY</td><td>054: Core ne profile<br>031: ELM density transient<br>063: TAE δn/n, δT/T</td></tr><tr><td>GB: In-vessel ECH detectors</td><td>BC</td><td>103: In-vessel ECH stray radiation intensity</td></tr></tbody></table>

such models relies on, amongst other things a good idea of the use case and combination to which the data will be put. Furthermore, the controllers need real-time data of sufficient quality and the definition of sufficient and the optimal way to derive it from the diagnostics again relies on the use case and the overall control implementation.

A dedicated WG is now working towards the systematic development of the controllers and underlying synthetic systems and the corresponding subchapter 10 lays out the progress in this area.

### 1.3.10. IDA.
ITER measurements for control and physics understanding have formal availability targets to ensure high efficiency operation of the physics program in the ITER research plan (IRP [5]). To meet the availability, redundancy is inbuilt to the design of individual systems and, for many measurements, more than one diagnostic has a primary contribution, e.g. TS and ECE to core electron temperature. In turn, both depend on the equilibrium reconstruction. In this situation, analysis of the errors and biases of a particular diagnostic can be difficult. For a particular output, such as $T_e$, IDA based on a Bayesian probabilistic approach can be used to combine the information in a rigorous way, improve the reliability of parameter estimation and find inconsistencies. The corresponding work of the recently formed IDA WG, see subchapter 11, shows ITER and other examples of synergy between two diagnostics, three-way profile information combination, improved magnetic equilibrium reconstruction and velocity-space tomography. This work and its benchmarking on running devices will enable to get the most out of ITER operation.

### 1.3.11. Magnetic measurements.
Magnetic diagnostics in ITER interact strongly with the details of internal component engineering for the ITER load assembly. Their design is now almost complete and although it has evolved since the physics basis was last revised [4], the system is recognizably similar [7]. Sensors contribute to multiple measurements, sometimes

via multiple software modules, in a manner similar to that described in [8, 9]. Specific modules include real-time equilibrium reconstruction, plasma current, vertical speed, loop voltage, energy, instability measurements, toroidal field variation and halo reconstruction (figure 1). The modules are configurable and this makes mapping all the potential uses of the

<figure><img src="image_10.png" /><figcaption>**Figure 1.** Main contributions of the magnetics sensor groups (A1 ... A9 and AA ... AO, explained in table 7 with the exception of A5/A6, the steady state sensors, A8, the fiber optic current sensor, AN, the divertor Rogowskis and AO and AC, the toroidal coil sets) to the software modules in preparation. Each module is configurable to accept inputs from most sensors, including backup groups (not shown). Dashed lines indicate expected information flow for long pulse correction.</figcaption></figure>
<table><caption>**Table 7.** Primary contributions of the magnetics by operational role and parameter.</caption><thead><tr><th>Sub-System</th><th>Role</th><th>Parameter</th></tr></thead><tbody><tr><td>A1: Continuous external Rogowskis</td><td>MP</td><td>001b:Itor</td></tr><tr><td>A9: Diamagnetic compensation coils (Outer)</td><td>BC</td><td>051. Toroidal Magnetic Field BT</td></tr><tr><td>AA: Tangential coils (Inner)</td><td>MP<br>BC</td><td>001a: Ip<br>003: DZ/dt of current centroid<br>004: Main plasma gaps, Dsep</td></tr><tr><td>AB: Normal coils (Inner)</td><td>BC</td><td>004: Main plasma gaps, Dsep</td></tr><tr><td>AD: Partial flux loops</td><td>MP<br>BC</td><td>003: DZ/dt of current centroid<br>004: Main plasma gaps, Dsep</td></tr><tr><td>AE: Continuous flux loops (Inner)</td><td>BC</td><td>005: V loop</td></tr><tr><td>AF: Diamagnetic main loop</td><td>MP</td><td>006: βp</td></tr><tr><td>AG: Diamagnetic compensation coils (Inner)</td><td>MP</td><td>006: βp</td></tr><tr><td>AI: Magneto hydro dynamic saddles</td><td>MP</td><td>016a: Br/&lt;Bp&gt; RWM<br>016b: Br/&lt;Bp&gt; Error Field</td></tr><tr><td>AJ: High frequency sensors</td><td>MP<br>PHY</td><td>017: B<sub>theta</sub>(complex, at wall)/&lt;Bp&gt;<br>060: Fishbone B<sub>theta</sub>(mode)/&lt;Bp&gt;<br>062: TAE B<sub>theta</sub>(complex)/&lt;Bp&gt;</td></tr><tr><td>AL: Divertor equilibrium coils</td><td>BC</td><td>002: Channel location, divertor (r dir.)</td></tr><tr><td>AP: Rogowskis (Blanket)</td><td>MP</td><td>050a: Poloidal current in one sector</td></tr></tbody></table>

groups of sensors very complex, but the primary contributions are listed in table 7, after [7] and are discussed below grouped by software module.

The modules reside in a plant system controller originally conceived to run in the diagnostic hall but recently relocated to the main server room, with the exception of the interlock systems running directly in the acquisition aggregator boards [10]. Multiple instances of the modules can be running. The system feeds multiple clients; the data flows are summarized in figure 2. Specially developed integrators are used to accommodate all inductive signals, based on the concept developed for W7X [11], but redesigned *ab initio* to accommodate voltages up to 1 kV and parallel streaming of up to 30 modules into a single high bandwidth aggregator board [12]. Presently, an end-to-end prototype system is under test on WEST.

### 1.3.11.1. Plasma current and related parameters.
Plasma current ($I_p$) measurements for plasma control and investment protection are derived by the in-vessel pickup coils (IVCs). Six sets of equilibrium IVCs (Tangential coils AA) provide independent estimates from as many sectors with a basic bandwidth in the kHz range and an overall error below 1%. An FPGA-based high reliability system [12] provides a real-time 'Best-Ip' estimate to the investment protection ('interlock', CIS) system; simultaneously real-time feedback within the PCS ensures the plasma stays within Central Interlock System (CIS) limits. Fiber-optic current sensors (FOCS, A8)

<figure><img src="image_11.png" /><figcaption>**Figure 2.** Context data flow diagram of the magnetics diagnostic. The system interfaces to all the central I&C networks, feeding the acquired data to algorithms. Reprinted from [10], Copyright (2015), with permission from Elsevier.</figcaption></figure>

can restrict, on a slower timescale, the total toroidal current ($I_{tor}$). Together with the Continuous External Rogowskis (CER, A1) embedded in the Toroidal Field Coils (TFCs), they can be absolutely calibrated on demand using the ITER in-vessel radial field coils and can correct other sensors for integration drift.

IVCs (AA, but also normal coils, AB, and toroidal coils, AC, of similar construction) are in manufacture, following a development path to mitigate radiation effects, in particular radiation-induced currents [13]. After numerous engineering studies [14–17], optimized platforms are in production, using extensively optimized low temperature co-fired ceramic (LTCC) coils with a silver winding in a glass-ceramic matrix (figures 3 and 4). Two variants are included in the
<figure><img src="image_12.png" /><figcaption>**Figure 3.** X-ray of an ITER LTCC sensor showing layers from the side (a) and from the top (b). The size is approx. 40 × 30 × 8 mm. Reprinted from [18], Copyright (2022), with permission from Elsevier.</figcaption></figure>

<figure><img src="image_13.png" /><figcaption>**Figure 4.** Construction of a remote handling-compatible ITER in-vessel pick-up coil incorporating an LTCC coil (blue), placed on a CuCrZr base plate with an Inconel shield designed to allow a DC-15 kHz pass band. In the production run, the shield is copper-coated to reflect electron cyclotron frequencies. Reprinted from [16], Copyright (2019), with permission from Elsevier.</figcaption></figure>

manufacturing run designed to null radiation induced currents in different ways. A specially designed calibration rig and process records the effective axis (orientation) of each sensor to ~0.1 mrad. Calibration of the effective area with frequency is in the preparation stage.

Production of CERs is nearing completion. They are based on a composite design using a two-layer helically wound cable with fiberglass and polyimide insulation. The main TFC impregnation process also impregnates the CER and sets it in its channels within the structure. The performance of these coils has been analyzed [19, 20] and verified experimentally [19, 21]. All coils (three for use and one in the spare TFC) are already installed in their respective TFCs [22]; only on-site connection activities remain.

The FOCS is conceptually like the CER with the strong advantage of immunity from drift and electromagnetic interference [22–24]. Materials selection and interface design is complete and manufacturing of components trapped within the cryostat and building is proceeding [25, 26].

### 1.3.11.2. Plasma shape and speed.
By the definition of ITER, plasma vertical speed, $\dot{z}$, a diagnostic performance parameter (table 10, parameter 003) is linked both to the plasma current and the absolute position of the current centroid, as the in-vessel coils measure directly the rate of change of the $zI_p$ product and $I_p$ so that $\dot{z}I_p = (z\dot{I}_p) - z\dot{I}_p$. The estimate of $z$, the plasma centroid height, in turn requires time-integrated measurements, not just of field tangential to the vessel but also of the normal components. For this reason, the sensors that enter the shape and speed algorithms include both poloidal field axes. The speed measurement employs optimized summation of the signals for the coils and saddle loops and is available in real-time for plasma control.

The tangential (AA), normal (AB) and divertor (AL) coils, together with the saddle loops (AD) provide data to derive equilibria [7] using the P-EFIT code [27, 28]. The P-EFIT code is also used to derive estimates of other quantities (e.g. $\beta_{eq}$, $q$) as well as to broadcast an equilibrium grid for other users and estimates of the plasma wall gaps.

Shunts (AM, not shown in figure 1) are also included in the divertor measurements. They use the cassette body (CB) as the shunt element, to resolve cassette currents similarly to the Halo Rogowskis described in a later section. If sufficiently large, these currents can influence the shape reconstruction. They can also be used for detachment identification [29].

The construction of the in-vessel coils has been discussed in the preceding section, with the exception of the divertor coils (AL) that operate in a special high temperature and high neutron flux environment but with lower frequency requirements. A special CuCrZr monoblock heatsink houses pairs of these coils set at 90° to measure both poloidal field components in the poloidal plane (figure 5).

In ITER, flux loops are being installed both inside and outside the VV, with different design constraints and measurement requirements. They are mounted on the inner and outer walls of the vessel, in all nine sectors. In total, there are 237 flux loops of varying sizes, which represents ~1/5th of all magnetic diagnostic sensors.

The flux loops on the inner skin include partial flux loops (AD) that are of relatively modest size. In addition, there are the rather larger magneto-hydrodynamic (MHD, Al) flux loops. Finally, there are the full poloidal diamagnetic loops
<figure><img src="image_14.png" /><figcaption>**Figure 5.** A two-axis coil prototype (55.AL) about to undergo thermal and vibration tests. Bolts through the LTCC sensors, in combination with copper foam ensure the *in-vacuo* heat transfer to the main block representing the ITER divertor. Frequency response is a few tens of Hz.</figcaption></figure>

<figure><img src="image_15.png" /><figcaption>**Figure 6.** Trial assembly of a saddle loop showing the main features: Clips (numbered 1–30); junction box (lower center); cable relief (top; actual installed loops use smaller relief areas). Inset: Clip #13 (top left of main picture) welded shut.</figcaption></figure>

(AF) and the full toroidal voltage loops (AE). On the outer skin there exist only the full toroidal voltage loops (A7). Due to the high radiation dose rates and temperatures inside the VV, the flux loops are made from Mineral Insulated (MI) cables. These cables comprise a stainless-steel outer sheath, of order 0.3 mm thick, encasing aluminum oxide (alumina, $Al_2O_3$) insulation which surrounds a single, central oxygen free copper (OFC) or copper alloy (CuNi2Si, UNS C70260) conductor. The diameter is adjusted to the expected voltage and the terminations are tested for the expected pressure conditions (UHV to 100 Pa up to 1 kV) [30].

Special clips and terminations are developed. The clips are designed for short cycle stud welding onto any metal surface; the loops are then formed (including intentional undulations to relieve relative thermal strain and resultant fatigue) and the clips resistance-welded shut (figure 6). The process is suitable for the industrial assembly process of the several tens of thousands of clips necessary.

For long pulses, drift arising in the shape measurements can be compensated in principle by measurements of the field components on the external VV skin, either inductively (Outer vessel tangential and normal coils, A3 and A4) or using steady state sensors (tangential, A5 and normal A6, respectively. This correction can be applied in real time if needed [31, 32].

The outer vessel coil design takes advantage of the substantially lower $dB/dt$ outside the ITER VV to implement a large effective area coil of $O(2m^2)$ with modest output voltage and high immunity to parasitic pickup areas and, to some extent, radiation-induced currents [33]. It uses polyimide insulated winding packs (figure 7) and is attached to the ITER VV in a permanent fashion, using a high-strength sprung support system and internal strain relief, designed to cope with rapid cooling to 70 K (from potential He cooling leaks) and baking to 200 °C (figure 8).

The steady state sensors are much smaller but present unique challenges. Early work [34, 35] focused on strongly-doped InSb sensors, the idea being that the relatively large number of carriers would confer higher radiation immunity. Later, the work focused on Bismuth (Bi) sensors that exhibit much higher radiation immunity but have certain other difficulties, in particular non-linear behavior, relatively large temperature sensitivity requiring temperature monitoring to 0.3 K and a tendency to oxidize rapidly on exposure to air.

The sensors' construction, mounting and on-board temperature monitoring are all critical in the ITER environment [36]. The sensing elements (figure 9) are thin cross-shaped Bi layers deposited on a ceramic/copper circuit board with specially tapered contact regions and passivated by additional ceramic coating. Two such elements set in one housing to measure both poloidal field components. The entire unit is calibrated against temperature and field and, to ensure the temperature calibration is usable to <0.3 K, an on-board indium (In) cavity provides a fixed reference to the triple point of In at every ITER baking cycle [37].

These Bi-based sensors were qualified for ITER and will be installed on one out of the three sectors reserved for this use. The high attrition rate of the Bi layers in manufacture is a risk in service so, for the remaining sectors, antimony (Sb) sensors are undergoing trials. These offer better linearity, slightly lower temperature sensitivity and improved resistance to oxidation with lower but acceptable sensitivity. The ultimate radiation hardness of similar sensors should approach that of ceramic/metal induction loops based on mineral insulate cable (several dpa) so that, in hybrid operation as suggested in [31], should last as long as a reactor-grade breeding blanket exchange time. This offers the prospect of steady-state magnetic control of a reactor grade tokamak.
<figure><img src="image_16.png" /><figcaption>**Figure 7.** Winding process of an ITER outer vessel coil. Similar to the prototypes [33], the former is alumina ceramic and the winding is made of polyimide-coated wire.</figcaption></figure>

#### 1.3.11.3. Loop voltage, plasma energy and toroidal field.
Loop voltage is a BC measurement for monitoring current drive that requires relatively little processing. In ITER, in-vessel continuous (AE) and outer vessel (A7) flux loops, constructed similarly to the partial flux loops (AD) but running toroidally at multiple locations and with thicker cable rated for higher voltage, supply eight such values in-vessel and four for the outer vessel. Values at other locations can be computed using the partial flux loops and/or the similar but larger MHD loops (AI, functional description in next section) by poloidal fitting. In this way, approximately 20 locations can be monitored in-vessel without heavy computational requirements and there is sufficient redundancy to ensure availability for the ITER lifetime.

Plasma energy is a BC and stability check tool and in ITER is computed two ways: as a product of the P-EFIT equilibrium calculation ($\beta_{eq}$ or $\beta_{MHD}$) and directly by a compensated full poloidal flux loop system at three toroidal locations. The loops are implemented with the same technology as the saddle loops (AD, described in the shape and speed section), but use a two-turn configuration to allow symmetric deviations around obstacles. Basic vacuum field compensation is by means of toroidal coils (AG) with large effective area ($2 \text{ m}^2$) optimized for low thermoelectric voltage generation at the expense of speed. These coils are supplemented with measurements outside the VV that are slower still but offer improved immunity from radiation (A9).

The AG coils are under pre-series manufacture. They are wound on a CuCrZr bobbin with thin mineral insulated wire,

then fully vacuum brazed to the heatsink to ensure thermal gradients in service below 2 K. Prototypes are undergoing calibration trial (figure 11).

For energy measurements, the basic accuracy is limited by the accuracy of the compensation coils, and by any large non-axisymmetric eddy currents, as in disruptions [38]. In most conditions, the requirements (Parameter 04 of table 10) can be met.

The AG and A9 coils also measure toroidal field variation that requires a basic accuracy below 1 mT (Parameter 51 of table 10), like that needed for plasma energy.

#### 1.3.11.4. Measurement of instabilities.
Many of the coils and loops mentioned so far contribute to instability measurement. In most cases they have a backup or supplementary role to the dedicated instability detectors, that is, the MHD saddle loops, AI, visible in figure 10 and replicated in all manufacturing sectors and the high frequency (HF) coils, AJ, that are organized in a non-uniform but approximately symmetric set as described in [7].

MHD loops are set to detect mainly low ($m,n$) MHD modes, Parameter 08 of table 10. They are constructed similarly to the saddle loops (AD) but are much larger. As defined the plane of the VV, they can achieve detection to ($m, n$) ~(4, 4), sufficient to capture the field structure emerging from a mode locking before disruptions, a resistive wall mode (RWM) or a large error field. Their sensitivity is limited only by integrator drift (~0.1 mT equivalent for a 1 h pulse). Finer poloidal
<figure><img src="image_17.png" /><figcaption>**Figure 8.** Top: ITER VV Sector 6, the inboard outer vessel pickup coils tangential (top row) and normal (bottom row) clearly visible. Bottom: Close-up of a tangential coil. Inconel plates and fiberglass preload the winding pack into the Inconel case; Four welds to vacuum vessel bosses hold the sprung legs in place to allow an operating range 70–573 K without rotation.</figcaption></figure>

<figure><img src="image_18.png" /><figcaption>**Figure 9.** Construction of the steady state sensors in ITER (left, ~8 × 8 cm) and the final version of the Bi-based sensor chip for ITER (right, ~4 × 4 mm).</figcaption></figure>

<figure><img src="image_19.png" /><figcaption>**Figure 10.** A sector where MHD loops (AI, red) coexist with partial flux (saddle) loops (rectangles, blue) and full toroidal flux loops (lines, blue).</figcaption></figure>

(and toroidal) peaking can be sensed by a combination of their measurements with the saddle loops but with higher noise floor (due to their smaller average area).

For high frequency modes (over a few tens of Hz, Parameters 09 and 27 of table 10) covering everything up to the highest toroidal Alfvén eigenmodes (TAEs) expected in ITER, the VV eddy penetration time reduces the sensitivity of the loops and the main signals come from the 207 high frequency coils (figure 12). Their distribution captures, in quasi-regular sampling, nominal n-numbers up to

$n \sim 9$ and $m \sim 10$. Port arrays can locally sample even higher $n$-numbers, but given the structure of the eddy current signature
<figure><img src="image_20.png" /><figcaption>**Figure 11.** Pre-series toroidal compensation coil (AG) undergoing calibration trials in a Helmholtz rig (outside the field of view). The coil is in a special support cradle, designed to determine the coil axis to better than 0.1 mrad. The overall accuracy is 0.1%.</figcaption></figure>

<figure><img src="image_21.png" /><figcaption>**Figure 12.** Distribution of the HF coils (55.AJ) in ITER. There are two higher density belts (arrays, green dots) running toroidally on the low field side ($\theta = 0^\circ$) and two others on the high field side ($\theta = -180^\circ$), as well as six arrays in the poloidal plane (red) and three sets of high-density measurements near the edges of ports (blue). All coils measure poloidal field except for the high-density sensors below the equatorial ports (toroidal field).</figcaption></figure>

from the blanket modules, cannot resolve the true local mode structure.

These coils fit into the same platform used for the plasma shape and current measurement coils, with the key difference being the substitution of a tungsten-coated aluminum nitride (AlN) protection shield for the copper-coated Inconel 718 shield of the former. This shifts the -3 dB point of the case from ~30 kHz to ~500 kHz while allowing for extremely high ECH local heat load. Special supports are used for the arrays near the ports (figure 13). Further construction and performance details can be found in [39].

1.3.11.5. *Measurement of halo currents.* Halo current measurements are needed to understand the disruption process and as an input to calculations of VV forces and strains. The related

parameter (50 in table 10) requires resolving the halo in each of the nine ITER sectors. In practice, the key measurement of the halo in ITER is the current injected in the FW panels, measured by Rogowskis (AP) surrounding the earth straps of the modules. Around 30% of all the panels are sampled (figure 14). These measurements are complimented by Rogowskis around divertor cassette components (AN) that can resolve the current entering and leaving the cassette at six toroidal locations, as well as toroidal coils to measure the flux compression during the vertical displacement events (VDEs, AC at the top of the VV and AO on six cassettes) (figure 15). The halo current injection map is incomplete, so considerable effort went into the method of reconstruction of the full pattern [40–43] and to demonstrate that the measurement requirements can be met.

Construction of the Rogowskis has evolved but from the beginning (figure 16) [14] the designs employed mineral insulated cable construction. Prototyping of integrated (brazed) versions of the sensors in underway.

1.4. *Summary of progress with open issues and outlook for diagnostic operation in ITER*

The ITPA Diagnostic TG has operated across a broad range of topics organized in high priority areas (HP, table 8) and (usually linked) joint experiments (JEX). JEX are indexed in table 9. The work centered on the new challenges offered by ITER. For example, long pulses with superconducting radial field coils require close control of the noise in the related feedback system to keep eddy current losses low (DIAG-1). After an assessment within this and the MHD topic group, the related specification was captured in the specification of parameter 3 of table 10.

Other closed topics since the last IPB update include DIAG-3, the discrepancy between TS and ECE measurements. The trigger for this activity was a report of an apparent discrepancy at JET [49]. Following work at C-MOD [50], correction stop
<figure><img src="image_22.png" /><figcaption>**Figure 13.** High frequency sensor design for the port arrays (top) and early prototype without (lower left) and with (lower right) the W-on-AlN ECH shield.</figcaption></figure>

<figure><img src="image_23.png" /><figcaption>**Figure 14.** Distribution of the halo measurement Rogowskis on the ITER first wall panels (twin red dots per panel). The empty belt at around -120° poloidal angle corresponds to the divertor region, where six cassettes are sampled, indicated by (x).</figcaption></figure>

the JET calibrations [51] and review at ITPA of JET campaigns available in 2016, the JEX was closed. More recent work on JET data shows a non-random discrepancy at high T with systematic behavior [52] and the topic (though not the JEX) is followed again closely.

Active topics include the first mirror tests and strategy, now very mature and discussed in section 3.8 and in subchapter 9 and wall reflections with the latest developments in section 3.1 and in subchapter 2 and a record of incremental progress is in the additional references of table 9.

More recently and as the ITER diagnostic set matures, the Topic Group has increased emphasis on preparation for

operation and exploitation of ITER. Additional WGs for real-time operation, see subchapter 10, and IDA, see subchapter 11, grapple with the optimization of the data flows and data treatment needed to operate ITER with high efficiency and to ensure high quality data. These two WGs manage DIAG-12, whose objective is to develop, customize and validate synthetic diagnostic models, along with associated test cases within the IMAS framework [53]. These models should be made available prior each phase of the ITER Research Plan.

Other recent JEX support techniques whose implementation on ITER came relatively late, such as LIBS for FW survey of T (LIBS, section 3.1 and subchapter 2), now possible to install on the planned Agile Robot Transporter (similar to [54]) and the Fast Ion Loss Detector [55] now being integrated in ITER. In more detail:

* **DIAG-13: LIBS for in-situ T retention measurement.** This is in support of the IRP activity A.9 [56]. The main objective is to measure D/T concentration in the ITER FW and to constrain H trapping modeling codes.

* **DIAG-14: laser induced desorption spectroscopy (LIDS) for tritium quantification.** This proposal is in support of IRP A.7 [56] and the JEX activities include two different methods: LID-QMS (ms pulse) and LIDS (ms pulse; in the presence of background plasma) vs. LIA-QMS and LIBS.

* **DIAG-15: Density control with pellets.** This proposal is in support of HP#3 (table 8). The proposed activities include systematic controller design through system identification using frequency domain analysis (e.g. of N II front control with N2 seeding) and multi-input multi-output (MIMO) control of radiative loss processes.
<figure><img src="image_24.png" /><figcaption>**Figure 15.** Distribution of Rogowskis on the sampled cassettes (top, blue belts) and simplified electrical network showing the plasma currents to be resolved (bottom, yellow).</figcaption></figure>

<figure><img src="image_25.png" /><figcaption>**Figure 16.** View of the back side of the blanket module showing the location of the two Rogowski coils surrounding the grounding straps (left); picture of a brazed Rogowski prototype (left) employing 0.5 mm MI cable and four welding tabs (D).</figcaption></figure>

Diagnostic measurements for ITER benefitted enormously from the work of the ITPA topic group. ITER is now progressing towards operation. In this phase, diagnostics are being installed and tested and in parallel, plans for the research plan are being refined. This process will continue to throw up

questions of diagnostic performance, efficacy and interoperability as the measurements will have to come together to support ITER's evolving plans. The commitment of this group to support ITER in this phase remains essential to prepare ITER to carry out its mission effectively.
<table><caption>**Table 8.** List of currently active high priority items for the ITPA TG.</caption><thead><tr><td>Item</td><td>Description</td><td>Covered by</td></tr></thead><tbody><tr><td>HP#1</td><td>Optimize lifetime of plasma facing mirrors used in optical systems</td><td>DIAG-2</td></tr><tr><td>HP#2</td><td>Assess impact of microwaves on diagnostic systems</td><td>All SWG</td></tr><tr><td>HP#3</td><td>Assess plasma control system measurement requirements</td><td>DIAG-11</td></tr><tr><td>HP#4</td><td>Develop diagnostic, calibration techniques and strategies compatible with the burning plasma environment</td><td>All SWG</td></tr><tr><td>HP#5</td><td>Develop inversion reconstruction techniques (tomography, equilibrium...)</td><td>All SWG</td></tr></tbody></table>

<table><caption>**Table 9.** List of diagnostic joint experiments (involving at least two organizations excluding ITER IO) and status.</caption><thead><tr><td>JEX</td><td>Start</td><td>Status</td><td>Activity</td><td>References</td></tr></thead><tbody><tr><td>DIAG-1</td><td>2000</td><td>Closed</td><td>Assessment of the effect of noise on vertical velocity measurement</td><td>See text</td></tr><tr><td>DIAG-2</td><td>2000</td><td>Active</td><td>Environmental tests on first mirrors</td><td>Section 3.8, subchapter 9</td></tr><tr><td>DIAG-3</td><td>2007</td><td>Closed</td><td>Resolving the discrepancy between ECE and TS at high Te</td><td>See text</td></tr><tr><td>DIAG-4</td><td>2009</td><td>Closed</td><td>Test of capacitance micro-balances in tokamak</td><td>[44]</td></tr><tr><td>DIAG-5</td><td>2010</td><td>Closed</td><td>Field test of an activation probe</td><td>[45, 46]</td></tr><tr><td>DIAG-6</td><td>2010</td><td>Closed</td><td>Comparisons of charge exchange recombination (CXRS) spectroscopy and x-ray imaging crystal spectroscopy (XICS)</td><td>[45]</td></tr><tr><td>DIAG-7</td><td>2010</td><td>Closed</td><td>Distributed monitoring of microwave power density</td><td>[45-47]</td></tr><tr><td>DIAG-8</td><td>2010</td><td>Active</td><td>Benchmark of wall reflections</td><td>Section 3.1, subchapter 2, [45, 46, 48]</td></tr><tr><td>DIAG-9</td><td></td><td>Closed</td><td>Spectral MSE (MSE-LS) experiments as design driver for ITER MSE</td><td></td></tr><tr><td>DIAG-10</td><td></td><td>Closed</td><td>Minimizing microwave absorption in vacuum windows</td><td></td></tr><tr><td>DIAG-11</td><td>2017</td><td>Active</td><td>Determination of the runaway electron distribution function by spectral gamma ray Bremsstrahlung measurements during disruptions</td><td></td></tr><tr><td>DIAG-12</td><td>2020</td><td>Active</td><td>Development of synthetic diagnostics for ITER</td><td>See text</td></tr><tr><td>DIAG-13</td><td>2021</td><td>Active</td><td>Development of the LIBS technique in view of its implementation in ITER</td><td>See text</td></tr><tr><td>DIAG-14</td><td>2023</td><td>Active</td><td>LIDT studies for T quantification in ITER</td><td>See text</td></tr><tr><td>DIAG-15</td><td>2023</td><td>Active</td><td>Observation and control of the electron density and impurity distribution using small pellets</td><td>See text</td></tr></tbody></table>
<table><caption>Table 10. Specification for plasma measurements on ITER, derived from [57].</caption><thead><tr><th rowspan="2">Measurement group</th><th rowspan="2">Parameter</th><th rowspan="2">Range</th><th rowspan="2">Condition</th><th colspan="3">Resolution</th></tr><tr><th>Time or Freq.</th><th>Spatial or Wave No.</th><th>Accuracy</th></tr></thead><tbody><tr><td rowspan="5">01. Plasma current</td><td rowspan="2">001a: Ip</td><td>0–1 MA</td><td>Default</td><td>1 ms</td><td>Integral</td><td>10 kA</td></tr><tr><td>1–17.5 MA</td><td>Default</td><td>1 ms</td><td>Integral</td><td>1%</td></tr><tr><td rowspan="3">001b: Itor<br>100: Plasma current for CSS [Central Safety System]</td><td>20–0 MA</td><td>Ip quench</td><td>0.1 ms</td><td>Integral</td><td>30% + 10 kA</td></tr><tr><td>1–20 MA</td><td>Default</td><td>10 ms</td><td>Integral</td><td>1% + 10 kA</td></tr><tr><td>5–20 MA</td><td>Normal operation</td><td>1 s</td><td>Integrated</td><td>1% (2 σ)</td></tr><tr><td rowspan="4">02. Plasma position and shape</td><td>002: Channel location, divertor (r dir.)</td><td></td><td>Ip >3 MA</td><td>10 ms</td><td>—</td><td>10 mm</td></tr><tr><td rowspan="2">003: DZ/dt of current centroid</td><td rowspan="2">0–5 m s<sup>-1</sup></td><td>Ip quench</td><td>10 ms</td><td>—</td><td>20 mm</td></tr><tr><td>Default</td><td>1 ms</td><td>Sum of 3 or more toroidally symmetric positions</td><td>10 mm s<sup>-1</sup> (system noise) + 50% (depending on plasma scenario) + plasma noise</td></tr><tr><td>004: Main plasma gaps, Dsep</td><td></td><td>Ip >2 MA, full bore</td><td>10 ms</td><td>—</td><td>10 mm except gap 5 (25 mm)</td></tr><tr><td>03. Loop voltage</td><td>005: V loop</td><td>0–1000 V<br>0–30 V</td><td>Ip quench<br>Default</td><td>1 ms<br>1 ms</td><td>20 points<br>20 points</td><td>10% + 5 mV<br>5 mV</td></tr><tr><td rowspan="2">04. Plasma energy</td><td rowspan="2">006: βp</td><td rowspan="2">0.01–5</td><td>Ip >3 MA</td><td>10 ms</td><td>Integral</td><td>0.05 + 5%<br>30%</td></tr><tr><td>Ip quench</td><td>0.1 ms</td><td>Integral</td><td></td></tr><tr><td rowspan="8">05. Radiated power</td><td>007a: Prad total, divertor</td><td>1–300 MW</td><td>Stationary</td><td>10 ms</td><td>Integral</td><td>10%</td></tr><tr><td>007b: Prad total (transient), divertor</td><td>25 MW–100 GW</td><td>Transients</td><td>1 ms</td><td>Integral</td><td>20%</td></tr><tr><td>008a: Prad total, main plasma</td><td>1–300 MW</td><td>Stationary</td><td>10 ms</td><td>Integral</td><td>10%</td></tr><tr><td>008b. Prad, top region (upper X-point)</td><td>0.1–20 MW</td><td>Stationary</td><td>10 ms</td><td>Integral</td><td>10%</td></tr><tr><td>008 c: Prad total (transient), main plasma + upper X-point</td><td>50 MW–200 GW</td><td>Transients</td><td>1 ms</td><td>Integrated, 4 toroidal locations</td><td>20%</td></tr><tr><td>009a: Prad total</td><td>1–300 MW</td><td>Stationary</td><td>10 ms</td><td>Integral</td><td>10%</td></tr><tr><td>009b: Prad (transient) total</td><td>50 MW–200 GW</td><td>Transients</td><td>1 ms</td><td>Integrated, 4 toroidal locations</td><td>20%</td></tr><tr><td>011: Line-averaged electron density</td><td>1 × 10<sup>18</sup>–4 × 10<sup>20</sup> m<sup>-3</sup><br>8 × 10<sup>20</sup>–2 × 10<sup>22</sup> m<sup>-3</sup></td><td>Default</td><td>1 ms</td><td>Integrated</td><td>1%</td></tr><tr><td></td><td></td><td>DMS</td><td>1 ms</td><td>Integrated</td><td>100%</td></tr></tbody></table>

(Continued.)
<table><caption>Table 10. (Continued.)</caption><thead><tr><th rowspan="2">Measurement group</th><th rowspan="2">Parameter</th><th rowspan="2">Range</th><th rowspan="2">Condition</th><th colspan="3">Resolution</th></tr><tr><th>Time or Freq.</th><th>Spatial or Wave No.</th><th>Accuracy</th></tr></thead><tbody><tr><td rowspan="6">07. Neutron flux and emissivity</td><td>012: Fusion power</td><td>100 kW—0.9 GW</td><td></td><td>1 ms</td><td>Integrated</td><td>10%</td></tr><tr><td>013: Fusion power density</td><td>1 kW.m<sup>-3</sup>—<br>4 MW.m<sup>-3</sup></td><td></td><td>1 ms</td><td>a/10</td><td>10%</td></tr><tr><td>014: Neutron- and α-source profile</td><td>1 × 10<sup>14</sup>—<br>-1.4 × 10<sup>18</sup></td><td></td><td>1 ms</td><td>a/10</td><td>10%</td></tr><tr><td>015: Total neutron flux</td><td>1 × 10<sup>14</sup>—<br>3.2 × 10<sup>20</sup></td><td></td><td>1 ms</td><td>integrated</td><td>10%</td></tr><tr><td>102: Total fusion power for safety relevant measurements determination</td><td>70 MW—0.9 GW</td><td></td><td>1 s</td><td>None</td><td>10% + 10 MW</td></tr><tr><td>08. Error field, locked mode and RWM identification</td><td>016a: Br/&lt;Bp&gt;<br>RWM</td><td>1 × 10<sup>-4</sup>—<br>2 × 10<sup>-2</sup></td><td></td><td>1 ms</td><td>{0,1} &lt; {m,n} &lt; {3,2}</td><td>30%</td></tr><tr><td rowspan="2">09. Low (m,n) MHD modes, sawteeth, locked modes, and disruption precursors</td><td>016b: Br/&lt;Bp&gt;<br>Error Field</td><td>1 × 10<sup>-4</sup>—<br>2 × 10<sup>-2</sup></td><td></td><td>1 ms</td><td>{0,1} &lt; {m,n} &lt; {3,1}</td><td>30%</td></tr><tr><td>016c: Br/&lt;Bp&gt;<br>Locked Mode</td><td>1 × 10<sup>-4</sup>—<br>2 × 10<sup>-2</sup></td><td></td><td>1 ms</td><td>{0,1} &lt; {m,n} &lt; {10,3}</td><td>30%</td></tr><tr><td rowspan="2">10. Plasma rotation</td><td>017: Btheta/&lt;Bp&gt;</td><td>1 × 10<sup>-4</sup>—<br>2 × 10<sup>-2</sup></td><td></td><td>10 Hz—<br>35 KHz</td><td>{0,1} &lt; {m,n} &lt; {10,7}</td><td>10%</td></tr><tr><td>018: vPOL<br>019: vTOR</td><td>1–50 km s<sup>-1</sup><br>1–200 km s<sup>-1</sup></td><td></td><td>10 ms</td><td>a/30*</td><td>30%</td></tr><tr><td>11. Fuel ratio in plasma core</td><td>020. nT/nD Core</td><td>0.01–10</td><td>r/a &lt;0.85</td><td>100 ms</td><td>Integral</td><td>20%</td></tr><tr><td rowspan="10">12. Impurity species monitoring</td><td>021a: C, O influx</td><td>1 × 10<sup>16</sup>–5 × 10<sup>19</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>021b: Be influx</td><td>1 × 10<sup>16</sup>–5 × 10<sup>19</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>021c: Be influx distribution</td><td>1 × 10<sup>16</sup>–5 × 10<sup>19</sup></td><td>Be influx distribution</td><td>10 ms</td><td>Local, 15–40 mm</td><td>10% (rel)</td></tr><tr><td>022: Be, C, O Relative Concentration</td><td>1 × 10<sup>-4</sup>—<br>5 × 10<sup>-2</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>023: Cu Influx</td><td>1 × 10<sup>15</sup>–5 × 10<sup>18</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>024: Cu rel. conc.</td><td>1 × 10<sup>-5</sup>—<br>5 × 10<sup>-3</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>025: Extrinsic (Ne, Ar, Kr) influx</td><td>1 × 10<sup>16</sup>–2 × 10<sup>19</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>026: Extrinsic (Ne, Ar, Kr) Relative Concentration</td><td>1 × 10<sup>-4</sup>—<br>2 × 10<sup>-2</sup></td><td></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr></tbody></table>

(Continued.)
<table><caption>Table 10. (Continued.)</caption><tbody><tr><td></td><td>027: W influx</td><td>1 × 10<sup>14</sup>–5 × 10<sup>17</sup><br/>–2 s<sup>–1</sup></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td></td><td>028: W Relative<br/>Concentration</td><td>1 × 10<sup>–6</sup>–<br/>5 × 10<sup>–4</sup></td><td>10 ms</td><td>Integral</td><td>10% (rel)</td></tr><tr><td>13. Zeff<br/>(line-averaged)</td><td>029. Line-averaged<br/>Zeff</td><td>1–5</td><td>10 ms</td><td>Integral</td><td>20%</td></tr><tr><td rowspan="4">14. H-mode,<br/>ELMs and L-H<br/>mode transition<br/>indicator</td><td>030: ELM<br/>visible radiation<br/>bursts</td><td></td><td>0.02 ms</td><td>Main plasma<br/>integral, Divertor<br/>one site</td><td>—</td></tr><tr><td>031: ELM<br/>density transient</td><td>5 × 10<sup>18</sup>–<br/>3 × 10<sup>20</sup> m<sup>–3</sup></td><td>r/a >0.85</td><td>10 ms</td><td>5 mm</td><td>5%</td></tr><tr><td>032: ELM<br/>temperature<br/>transient</td><td>0.05–10 keV</td><td>r/a >0.85</td><td>0.1 ms</td><td>5 mm</td><td>10%</td></tr><tr><td>033: L-H Dα<br/>step</td><td></td><td></td><td>0.02 ms</td><td>Main plasma<br/>integral, Divertor<br/>one site</td><td>—</td></tr><tr><td rowspan="3">15. Runaway<br/>electrons</td><td>034: Emax<br/>runaway</td><td>1–100 MeV</td><td>10 ms</td><td>—</td><td>20%</td></tr><tr><td>035: I runaway</td><td>(0.05–0.7) × Ip</td><td>After thermal<br/>quench</td><td>10 ms</td><td>—</td><td>30% (rel)</td></tr><tr><td></td><td>0–5 MA</td><td>Failed breakdown</td><td>10 ms</td><td>—</td><td>50 kA</td></tr><tr><td rowspan="7">16. Divertor<br/>operational<br/>parameters</td><td>037: Gas<br/>composition<br/>(Fuel, He,<br/>impurities) (P<br/>div)</td><td>(1 × 10<sup>–4</sup>–1)<br/>× Pdiv</td><td>1 s</td><td>Several points</td><td>20%</td></tr><tr><td>038: Gas<br/>pressure,<br/>divertor (Pdiv)</td><td>1 × 10<sup>–4</sup>–20 Pa</td><td>50 ms</td><td>Several points</td><td>20%</td></tr><tr><td>039: Max.<br/>surface<br/>temperature,<br/>divertor</td><td>200 °C–3600 °C</td><td>2 ms</td><td>—</td><td>10%</td></tr><tr><td>040a. Fine<br/>surface<br/>metrology—<br/>Divertor</td><td>0–1 mm</td><td>Default</td><td>&lt;1 per<br/>discharge</td><td>1 mm</td><td>10 μm</td></tr><tr><td>040b. Fine net<br/>erosion and<br/>redeposition—<br/>first<br/>wall</td><td>0–100 μm</td><td>LTM [Long Term<br/>Maintenance]</td><td>Sample<br/>retrieval<br/>possible only<br/>during LTM</td><td>50 locations on<br/>700 m2 [of first<br/>wall]</td><td>1 μm</td></tr><tr><td>041: Position of<br/>the ionization<br/>front, divertor</td><td>0–1 m</td><td></td><td>1 ms</td><td>100 mm</td><td>—</td></tr><tr><td>17. First wall<br/>(FW) visible<br/>image &amp; wall<br/>temperature</td><td>042: Surface<br/>luminance, FW</td><td>40–1 × 10<sup>5</sup> cd<br/>m<sup>–2</sup></td><td>All heating ports,<br/>upper strike region,<br/>dome, baffle.<br/>&gt;50% of rest,<br/>evenly distributed.</td><td>10 ms</td><td>3 mm</td><td>30% absolute, 1%<br/>relative</td></tr></tbody></table>

(Continued.)
<table><caption>**Table 10. (Continued.)**</caption><thead><tr><th rowspan="2">Measurement group</th><th rowspan="2">Parameter</th><th rowspan="2">Range</th><th rowspan="2">Condition</th><th colspan="3">Resolution</th></tr><tr><th>Time or Freq.</th><th>Spatial or Wave No.</th><th>Accuracy</th></tr></thead><tbody><tr><td rowspan="2"></td><td>043: Surface temperature, FW</td><td>200 °C–3600 °C</td><td rowspan="2">All heating ports, upper strike region, dome, baffle. &gt;50% of rest, evenly distributed.</td><td>10 ms</td><td>5 mm</td><td>20 °C</td></tr><tr><td>044: Surface temperature during ELMs, FW</td><td>400 °C–3600 °C</td><td>20 us</td><td>5 mm</td><td>20 °C</td></tr><tr><td rowspan="2">18. Gas pressure and composition in main chamber</td><td>045: Gas composition (Fuel, He, impurities), main (Pmain)</td><td>(1 × 10<sup>-4</sup>–1) Pmain</td><td rowspan="2"></td><td>10 ms</td><td>Several points</td><td>50%</td></tr><tr><td>046: Gas pressure, main (Pmain)</td><td>1 × 10<sup>-4</sup>–1 Pa</td><td>1 s</td><td>Several points</td><td>20%</td></tr><tr><td rowspan="2">19. Gas pressure and composition in vacuum ducts</td><td>047: Gas composition (Fuel, He, impurities), duct (Pduct)</td><td>(1 × 10<sup>-4</sup>–1) Pduct</td><td rowspan="2"></td><td>1 s</td><td>Several points</td><td>20%</td></tr><tr><td>048: Gas pressure, duct (Pduct)</td><td>1 × 10<sup>-4</sup>–20 Pa</td><td>100 ms</td><td>Several points</td><td>20%</td></tr><tr><td>20. In-vessel inspection</td><td>049: Wall image</td><td>100% coverage of FW and divertor</td><td></td><td></td><td>1 mm</td><td></td></tr><tr><td rowspan="2">21. Halo currents</td><td>050a: Poloidal current in one sector</td><td>0.01–0.1 Ip</td><td>In disruption</td><td>1 ms</td><td>9 sectors, between injection points to VV</td><td>20%</td></tr><tr><td>050b: Current distribution in divertor cassette</td><td>0–190 kA</td><td>In disruption</td><td>1 ms</td><td>6 cassette targets, dome, VV connections</td><td>20%</td></tr><tr><td>22. Toroidal magnetic field</td><td>051: Toroidal Magnetic Field BT</td><td>−5.5–+5.5 T</td><td>Variation in pulse</td><td>1 s</td><td>2 locations</td><td>1 mT + 0.1%</td></tr><tr><td rowspan="2">23. Electron temperature profile</td><td>052: Core Te profile</td><td>0.5–40 keV</td><td>r/a &lt;0.85</td><td>10 ms</td><td>a/30</td><td>10%</td></tr><tr><td>053: Edge Te profile</td><td>0.05–10 keV</td><td>r/a &gt;0.85</td><td>10 ms</td><td>5 mm</td><td>10%</td></tr><tr><td rowspan="2">24. Electron density profile</td><td>054: Core ne profile</td><td>3 × 10<sup>19</sup>–3 × 10<sup>20</sup> m<sup>-3</sup></td><td>r/a &lt;0.85</td><td>10 ms</td><td>a/30</td><td>5%</td></tr><tr><td>055: Edge ne profile</td><td>5 × 10<sup>18</sup>–3 × 10<sup>20</sup> m<sup>-3</sup></td><td>r/a &gt;0.85</td><td>10 ms</td><td>5 mm</td><td>5%</td></tr><tr><td rowspan="3">25. Current profile</td><td>056: q profile</td><td>0.5–5</td><td>Physics study</td><td>10 ms</td><td>a/20</td><td>10%</td></tr><tr><td>057: Normalized Position of q = 1.5,2 surfaces</td><td>5–9</td><td>Physics study</td><td>10 ms</td><td>a/20</td><td>50%</td></tr><tr><td></td><td>0.3–0.9</td><td>NTM Feedback</td><td>10 ms</td><td>—</td><td>50 mm/a</td></tr></tbody></table>

(Continued.)
<table><caption>**Table 10. (Continued.)**</caption><tbody><tr><td></td><td>058: Normalized Position of qmin</td><td>0.3–0.7</td><td>NTM Feedback</td><td>1 s</td><td>—</td><td>50 mm/a</td></tr><tr><td>26. Zeff profile</td><td>059: Zeff profile</td><td>1.0–5.0</td><td>Default Transients</td><td>100 ms<br>10 ms</td><td>a/10<br>a/10</td><td>10%<br>20%</td></tr><tr><td>27. High frequency instabilities (MHD, NTMs, AEs, turbulence)</td><td>060: Fishbone Btheta(mode)/<br>&lt;Bp&gt;<br>061: NTM δT/Te<br>062: TAE Btheta/&lt;Bp&gt;<br>063: TAE δN/n, δT/T</td><td>1 × 10<sup>-4</sup>–<br>5 × 10<sup>-2</sup><br>(0.1–5) × 1 × 10<sup>-2</sup><br>1 × 10<sup>-6</sup>–<br>4 × 10<sup>-4</sup><br>5 × 10<sup>-6</sup>–<br>5 × 10<sup>-4</sup></td><td>Te &gt; 1 keV,</td><td>0.1–10 kHz<br>0.1–10 kHz<br>30 kHz—<br>0.5 MHz<br>30 kHz—<br>0.5 MHz</td><td>(m,n) = (1,1)<br>Island width<br>&gt;20 mm<br>N = 2–50<br>n = 2–50</td><td>30%<br>1 × 10<sup>-3</sup><br>30%<br>30% (rel)</td></tr><tr><td>28. Ion temperature profile</td><td>064: Core Ti<br>065: Edge Ti</td><td>0.5–40 keV<br>0.05–10 keV</td><td>r/a &lt;0.85<br>r/a &gt;0.85</td><td>100 ms<br>100 ms</td><td>a/30<br>10–20 mm</td><td>10%<br>10%</td></tr><tr><td>29. Core He density</td><td>066: Core nHe/ne<br>067: 3He concentration profile</td><td>1%–20%<br>1%–10%</td><td>r/a &lt;0.85<br>r/a &lt;0.85</td><td>100 ms<br>100 ms</td><td>a/10<br>a/10</td><td>10%<br>10%</td></tr><tr><td>30. Confined alphas and fast ions</td><td>068: Alpha Density profile<br>069: Alpha energy spectrum<br>070: p, D, T He3 energy spectrum</td><td>(0.1–2) E18 m<sup>-3</sup><br>0.1–3.5 MeV<br>0.1–1 MeV</td><td></td><td>100 ms<br>100 ms<br>100 ms</td><td>a/10<br>a/10<br>a/20</td><td>20%<br>20%<br>20%</td></tr><tr><td>31. Escaping alphas and fast ions</td><td>071: Fast particle loss distribution<br>072: Fast particle loss flux</td><td>±1 vll/v<br>0.1–5 MeV<br>0.2–20 MW m<sup>2</sup></td><td>Steady-state and transient<br>Steady-state and transient<br>Steady-state and transient</td><td>5 μs<br>5 μs<br>5 μs</td><td>Discrete measurement regions<br>Discrete measurement regions<br>Discrete measurement region</td><td>20%<br>20%<br>10%</td></tr><tr><td>32. Impurity density profile</td><td>073: Fractional content, Z &gt; 10<br>074: Fractional content, Z ≤ 10</td><td>0.01–0.3%<br>0.5%–20%</td><td>r/a &lt;0.85<br>r/a &gt;0.85<br>r/a &lt;0.85<br>r/a &gt;0.85</td><td>100 ms<br>100 ms<br>100 ms<br>100 ms</td><td>a/10<br>50 mm<br>a/10<br>50 mm</td><td>20%<br>c<br>20%<br>20%</td></tr><tr><td>33. Fuel ratio in edge</td><td>075: n<sub>H</sub>/n<sub>D</sub>, edge<br>076: n<sub>T</sub>/n<sub>D</sub>, edge</td><td>0.01–0.1<br>0.01–10</td><td>r/a &gt;0.85<br>r/a &gt;0.85</td><td>100 ms<br>100 ms</td><td>Radial integral<br>Radial integral</td><td>20%<br>20%</td></tr></tbody></table>

(Continued.)
<table><caption>Table 10. (Continued.)</caption><thead><tr><th rowspan="2">Measurement group</th><th rowspan="2">Parameter</th><th rowspan="2">Range</th><th rowspan="2">Condition</th><th colspan="3">Resolution</th></tr><tr><th>Time or Freq.</th><th>Spatial or Wave No.</th><th>Accuracy</th></tr></thead><tbody><tr><td rowspan="2">34. Neutron fluence</td><td>077: First wall neutron fluence</td><td>0.1–1 MWa m<sup>-2</sup></td><td rowspan="2">Normal operation</td><td rowspan="2">1 d</td><td>Several poloidal points</td><td>10%</td></tr><tr><td>101: Neutron fluence for lifetime</td><td>0–0.4 MWa m<sup>-2</sup></td><td>Equatorial midplane DFW</td><td>0.03 Mwa m<sup>-2</sup> (2σ)</td></tr><tr><td rowspan="2">35. Impurity and D, T influx in divertor</td><td>078: Be, C, W influx in divertor</td><td>1 × 10<sup>17</sup>–1 × 10<sup>22</sup> at s<sup>-1</sup></td><td rowspan="2"></td><td rowspan="2">1 ms</td><td>50 mm</td><td>30%</td></tr><tr><td>079: D2, T2 influx in divertor</td><td>1 × 10<sup>19</sup>–1 × 10<sup>25</sup> at s<sup>-1</sup></td><td>50 mm</td><td>30%</td></tr><tr><td rowspan="3">36. Plasma parameters at the divertor targets</td><td>080a: ne, divertor target</td><td>1 × 10<sup>18</sup>–1 × 10<sup>22</sup> m<sup>-3</sup></td><td rowspan="3"></td><td rowspan="3">1 ms</td><td>3 mm resolvability, 24 mm interval</td><td>30%</td></tr><tr><td>080b: Divertor Target Parallel Ion Flux</td><td>1 × 10<sup>4</sup>–1 × 10<sup>7</sup> Am<sup>-2</sup></td><td>3 mm resolvability, 24 mm interval</td><td>30%</td></tr><tr><td>081: Te, divertor target</td><td>1 eV–150 keV</td><td>3 mm resolvability, 24 mm interval</td><td>30%</td></tr><tr><td rowspan="6">37. Radiation profile</td><td>082a. Divertor Prad Profile</td><td>0.1–300 MWm<sup>-3</sup></td><td>Stationary</td><td>10 ms</td><td>50 mm</td><td>10%</td></tr><tr><td>082b: Divertor Prad (transient) Profile</td><td>0.03–20 GWm<sup>-3</sup></td><td>Transient</td><td>1 ms</td><td>50 mm</td><td>20%</td></tr><tr><td>083a: Prad profile, main plasma + upper X-point</td><td>0.001–100 MWm<sup>-3</sup></td><td>Stationary</td><td>10 ms</td><td>100 mm</td><td>20%</td></tr><tr><td>083b: Prad profile (transient), main plasma + upper X-point</td><td>10 MWm<sup>-3</sup>–1 GWm<sup>-3</sup></td><td>Transient</td><td>1 ms</td><td>100 mm</td><td>20%</td></tr><tr><td>083 c: Prad vertical line integrals (transient) main plasma + upper X-point</td><td>30 MW m<sup>-2</sup>–8 GW m<sup>-2</sup></td><td>Transient</td><td>1 ms</td><td>3–5 vertical line integrals at 4 toroidal locations</td><td>20%</td></tr><tr><td>085: Power load profile in divertor</td><td>0.02–5 GW m<sup>-2</sup></td><td>Disruption</td><td>0.02 ms</td><td>3 mm</td><td>20%</td></tr><tr><td rowspan="2">38. Heat loading profile in divertor</td><td>086: Surface temperature</td><td>0.1–25 MW m<sup>-2</sup> 1000 °C–3600 °C 200 °C–1000 °C</td><td>Default</td><td>2 ms</td><td>3 mm</td><td>10%</td></tr><tr><td></td><td></td><td></td><td>0.02 ms</td><td>3 mm</td><td>10%</td></tr><tr><td rowspan="2">39. Divertor Helium density</td><td>087: nHe, divertor</td><td>1 × 10<sup>17</sup>–1 × 10<sup>19</sup> m<sup>-3</sup></td><td rowspan="2"></td><td rowspan="2">1 ms</td><td>2 ms</td><td>10%</td></tr><tr><td></td><td></td><td></td><td></td><td></td></tr><tr><td rowspan="2">40. Fuel ratio in divertor</td><td>088: nH/nD, divertor</td><td>0.01–0.1</td><td rowspan="2"></td><td rowspan="2">100 ms</td><td>Integral</td><td>20%</td></tr><tr><td>089: nT/nD, divertor</td><td>0.01–10</td><td>Integral</td><td>20%</td></tr><tr><td rowspan="2">41. Divertor electron parameters</td><td>090: ne Divertor</td><td>1 × 10<sup>19</sup>–1 × 10<sup>22</sup> m<sup>-3</sup></td><td rowspan="2"></td><td rowspan="2">1 ms</td><td>50 mm × 3 mm</td><td>30%</td></tr><tr><td>091: Te, divertor</td><td>0.3–200 eV</td><td>50 mm × 3 mm</td><td>20%</td></tr></tbody></table>

(Continued.)
<table><caption>**Table 10. (Continued.)**</caption><tr><td>42. Ion temperature in the divertor</td><td>092: Ti profile, divertor</td><td>0.3–200 eV</td><td>1 ms</td><td>50 mm × 3 mm</td><td>20%</td></tr><tr><td>43. Divertor flow</td><td>093: Vp Divertor</td><td>1 × 10<sup>3</sup>–1 × 10<sup>5</sup> ms<sup>-1</sup></td><td>1 ms</td><td>50 mm × 3 mm</td><td>20%</td></tr><tr><td>44. nH/nD ratio in plasma core</td><td>094: nH/nD, core</td><td>0.01–0.1</td><td>r/a < 0.85</td><td>100 ms</td><td>20%</td></tr><tr><td>45. Neutral density between plasma and first wall</td><td>095. D/T Influx in Chamber</td><td>1 × 10<sup>18</sup>–1 × 10<sup>20</sup> at m<sup>-2</sup> s<sup>-1</sup></td><td>100 ms</td><td>Several positions</td><td>30%</td></tr><tr><td rowspan="3">46. Dust monitoring</td><td>097a: Surface Concentration of mobilizable dust (divertor cassette body under the dome and VV floor)</td><td>0.01 ~ 40 kg m<sup>-2</sup></td><td>Once every LTM or on demand</td><td>Integral</td><td>± factor 5</td></tr><tr><td>097b: Dust size distribution</td><td>1–200 µm</td><td>6 bins per decade (E6)</td><td>Once every LTM or on demand</td><td>±20% (relative)</td></tr><tr><td>097c: Viewing of dust area on and under the divertor</td><td>4 mm × 2 m</td><td>Once every LTM or on demand</td><td>100 µm</td><td>±50%</td></tr><tr><td rowspan="2">47. Tritium Monitoring</td><td>099a. Divertor Surface H, D, T Concentration (Inner baffle)</td><td>1 × 10<sup>21</sup>–2 × 10<sup>24</sup> m<sup>-2</sup></td><td>Outside plasma pulses under vacuum condition</td><td>On demand</td><td>Individual points 5 mm lateral resolution</td><td>50% abs, 20% relative</td></tr><tr><td>099b. First Wall Surface H, D, T Concentration</td><td>1 × 10<sup>18</sup>–2 × 10<sup>23</sup> m<sup>-2</sup></td><td>LTM</td><td>On request during LTM</td><td>50 locations on 700 m<sup>2</sup></td><td>20%</td></tr><tr><td rowspan="2">48. In-vessel ECH stray radiation</td><td>103: In-vessel ECH stray radiation intensity</td><td>Breakdown: Up to 3 MW m<sup>-2</sup> during 5.5 sec<br/>Normal operation: Up to 1.25 MW m<sup>-2</sup> during pulse</td><td>10 ms</td><td>5–9 sectors. Up to 20 sensors/sector</td><td>100 kW m<sup>-2</sup></td></tr><tr><td></td><td></td><td>10 ms</td><td>5–9 sectors. Up to 20 sensors/sector</td><td>100 kW m<sup>-2</sup></td></tr></table>

<footnote>(*) a/X is generally taken as the nominal minor radius, 2 m X<sup>-1</sup>, a/10 = 200 mm etc.</footnote>
## 2. Preparation for the measurements of plasma-metallic wall interaction in support of the ITER research plan: divertor power flux, impurity, and tritium retention

S.-H. Hong⁴, B. Peterson⁵, M.H. Aumeunier¹, A. Bultel⁶, C. Klepper⁷, D. Rasmussen⁷, H. Choi⁸, C. Grisolia¹, K. Kim⁸, S. Oh⁸ and C. Sun⁸

### 2.1. Introduction

To achieve the research goals of ITER, governing challenges of plasma-wall interaction are inevitable, especially achieving $Q = 10$ scenarios with a minimal amount of metallic impurities inside the core [58]. Furthermore, very high heat flux to the tungsten divertor is expected as well as neutron wall loading originating from the D–T reaction [59]. Detached plasma operation scenarios with high-precision measurements of the divertor power flux and impurity contents in the core and SOL are required, which will provide sophisticated and accurate input for plasma control [60]. Furthermore, ITER is licensed as a nuclear facility which needs to follow very strict license control on the amount of tritium inside the machine as well as on-site [61].

The heat flux on the divertor will be measured by an in-vessel infrared (IR) camera system looking down from the top slanted ports. The IR camera system needs to be calibrated for the measurements of the temperature of the tungsten diver-tor surface (monoblocks). The global power balance including plasma heating and current drive can be evaluated by the meas-urements of the plasma stored energy and total radiated power from the plasma.

Note that, the heat flux measurements by an IR camera strongly depend on the condition of the surface [62]. Change of the surface emissivity will result in wrong measurements of the monoblock temperature leading to wrong heat flux estimates. If the emissivity changes as a function of time, the changes must be followed: erosion of the tungsten surface by ion bom-bardment and corresponding prompt redeposition of tungsten atoms change the emissivity of the surface as well as the sput-tering threshold. Furthermore, the redeposition of eroded FW materials (beryllium and tungsten) onto the FW mirror has to be mitigated [63].

As the FW of ITER will be a full metal wall consisting of beryllium and tungsten, there will be a large portion of stray radiation caused by reflection from the FW [64]. Such reflection caused by the wall gives wrong information to the local temperature measurements, the location of hot spots, and finally to the global power balance. Therefore, a dedicated technique to identify and remove the reflected patterns from the measurements based on various cameras installed at ITER is essential.

Due to the use of tritium in the fusion plasma operation phase, active monitoring of tritium inside the VV is mandatory [65]. A LIBS and a LIDS are under development.

In this subchapter, we will discuss the progress of plasma-surface interaction diagnostics, including specific design issues, and their limitations. In section 2, the development

of an image bolometer system for ITER will be described in detail. In section 3, the development of the ITER prototype VUV spectrometer will be discussed. The status of the development of LIBS and LIDS for ITER will be followed in section 4. In section 5, technical challenges in the measurements of the plasma-wall interaction at ITER, i.e. the development of reflection elimination techniques and the systematic measurements of emissivity and the bi-directional refraction distribution function (BRDF) of tungsten, and the angular dependence of IR measurements will be described. The subchapter will be summarized in section 6.

### 2.2. Measurements of radiated power loss by imaging bolometers

For the purpose of the measurement of radiated power from ITER, the resistive bolometer (RB), which has a long history in fusion studies [66–69], is the primary detector being considered and prepared for installation. The RB is relatively compact, which permits it to be installed in the FW and diver-tor surrounding the plasma at one poloidal cross-section. This enables tomographic inversion of the line-averaged data to produce a local measurement of the radiated power density [69, 70]. One of the main drawbacks of the RB is the lamellar construction of the detector head, consisting of a radiation-absorbing metal foil, an insulating membrane, and a metallic meander that forms the resistor. Intrinsic stresses from the manufacturing process paired with different thermal expansion coefficients for absorber and membrane led to high localized stresses and breaking during the fatigue cycle under thermal cycling [71] for the prototypes based on Pt absorbers on a SiN membrane. However, this problem has been solved using an Au absorber layer instead of Pt [72]. Prototypes with different configurations and 20 µm thick Au absorbers, as necessary for ITER, have been shown to withstand thermal cycling up to 400 °C. Much research has been performed in the development of these detectors in the 14 years since the publication of the Progress in the ITER Physics Basis [4] and a diagnostic review [69]. The work on the RB has been recently reviewed [72] and therefore, the associated design and procurement activities for ITER will not be covered in this article.

Another type of detector that has seen much development in the last 14 years is the IRVB [73, 74]. The IRVB consists of one large metal foil held in a metal frame that absorbs the radiation incident through an aperture. The resulting temperature increase of the foil is measured by an IR camera located outside of the bioshield. IR optics bring the IR signal out of the bioshield and through the vacuum interface, avoiding the use of signal wires. The IRVB is being considered for application to ITER and future fusion reactors due to its use of inherently reactor relevant materials and its lack of wired feedthroughs, avoiding the risks of leaks. In contrast, the resistive bolometers require four wires for each channel to be brought through the vacuum interface. In comparison to the RB, the relative disadvantages of the IRVB are the large size and lower sensitivity. However, the sensitivity advantage of the RB is lost to some extent in the presence of radio-frequency-wave-induced noise if the detectors are not adequately shielded, which does
not plague an IRVB since all electrical wiring is located in the IR camera which can be properly shielded [75]. The RBs also have a well-developed *in-situ* calibration system using ohmic heating of the foil which allows calibration of the foils between discharges [76]. The large size of the IRVB and the necessity that all of the channels for one IRVB would view the plasma through the same aperture, require that it be installed in a port plug with a tangential view, relying on the assumption of toroidal symmetry in order to perform the tomographic inversion. Considering the differences between the two concepts in foil temperature measurement, lines of sight (LOS), and possible failure modes, they represent complementary approaches to the measurement of radiated power, which would provide some redundancy and assure the uninterrupted measurement of radiated power in ITER in case either would fail. Also, the differing LOS of the two concepts could be combined to enhance the tomography and the study of toroidal asymmetries in radiation.

In this section, we will summarize the development of the IRVB and discuss the possibility of applying it to ITER. The primary developments that will be reviewed are of a uniform and reproducible blackening of the absorbing foil, of an *in-situ* calibration system [77, 78] and the successful application of IRVBs to LHD [79], JT-60U [80], KSTAR [81], C-Mod [75], and MAST-U [82]. Finally, we will evaluate the viability of the application to ITER in terms of the estimated SNR.

### 2.2.1. Development of the IRVB.
Significant progress has been made in the development of the IRVB detector foil through the procurement of a vacuum vapor deposition facility for blackening the metal absorber foil with graphite. Blackening of the foil is necessary on the plasma side to increase absorption in the visible spectrum and on the IR camera side to improve the IR emissivity of the foil. Previously an aerosol graphite colloid called Aerogdag was used to hand spray a graphite coating on the foil. This resulted in a thick coating on the order of 5–8 µms [83] which turned out to be quite non-uniform and also not reproducible on both sides of the foil. The nonuniformity and irreproducibility could be compensated for in the calibration but may cause problems in the analysis and is better to be avoided. Therefore, a vacuum vapor evaporation facility was purchased and adapted to the coating of graphite on the metal foils [78]. This resulted in the ability to reproducibly control the thickness of the coating to within about 20 nm and allowed the determination of the optimal thickness of the coating to be 160 nm. Observation with a scanning electron microscope (SEM) also showed the coated surface to be much more uniform than in the handsprayed case.

Other progress in the development of the IRVB involved the improvement in calibration techniques and the development of an *in-situ* calibration system. Calibration of the IRVB foil involves the determination of the local values of the thermal and optical properties of the foil by using a HeNe laser or UV diodes. These properties include the thermal diffusivity and conductivity of the foil and the blackbody emissivity of the graphite coating. These parameters are then used in the

two-dimensional heat diffusion equation to solve for the incident radiated power from the temperature distribution on the foil measured by the IR camera. In order to take into account nonuniformities in the parameters and in the foil thickness, each point of the foil must be calibrated with a spatial resolution equivalent to that of the IRVB. In other words, each IRVB pixel must be individually calibrated. With the IRVB, calibration of the foil is done by heating the foil with a HeNe laser [73] or UV diodes [84] and measuring the change in the resulting temperature profile on the foil. In the laboratory, we move the foil keeping the IR camera and laser fixed in order to sample each part of the foil [84]. For *in-situ* calibration however, the laser must be moved from spot to spot on the foil. This type of system was realized on JT-60U [85, 86] and LHD [77, 78].

### 2.2.2. Application of the IRVB to various devices.
IRVBs have been applied to numerous devices. The first IRVB was installed on LHD [87] and up to four IRVBs have been simultaneously operated on LHD providing radiation images for comparison of synthetic images from EMC3-Eirene [88] and in an attempt at three-dimensional tomography [89]. Two different versions of the IRVB were tested on JT-60U. The first used a relatively low-cost, low-performance IR camera with a 2.5 µm thick × 9 cm × 7 cm gold foil with a semi-tangential view of the plasma from a 5 mm × 5 mm aperture [90]. This IRVB was the first to demonstrate two-dimensional radiation profiles in a tokamak through tomographic inversion and clearly showed the increase in core radiation during the introduction of iron impurities into the plasma [91, 92]. Just prior to the shutdown of JT60-U this IRVB was upgraded with a 5 µm Ta foil and an advanced IR camera with periscope optics. This resulted in a 30-fold improvement in sensitivity compared to the previous version [80]. This IRVB was later transferred to the KSTAR experiment and after upgrading the IR camera [81], it routinely provided total radiated power measurements and 2D profiles through tomographic inversion [92] as the only bolometer diagnostic on KSTAR. The third application of the IRVB to a tokamak took place on the C-mod experiment. This experiment showed good agreement between the RBs and IRVB and demonstrated the noise-free advantage of the IRVB compared to the RBs during radio-wave heating [75]. Recently the IRVB has been installed in the MAST-U experiment, demonstrating the advantages of the two dimensional view of the plasma in filling in gaps in the RB fields of view of the super-x divertor [82].

### 2.2.3. Investigating the viability of an IRVB for ITER.
During the last six years, investigations have been made into the viability of an IRVB for ITER. An example of a tangential field of view of an IRVB for ITER is shown in figure 17. The 7 cm × 9 cm Pt foil center is located at a major radius of 8.56 m and a height of 0.44 m above the midplane in an arbitrary location in an equatorial port. The aperture is 6 mm × 6 mm and is located 7.8 cm in front of the foil. The aperture and foil have a downward tilt of 4 degrees with respect to the horizontal. An IR camera and IR optical system with 1024 × 1280 pixels,
<figure><img src="image_26.png" /><figcaption>**Figure 17.** Field of view for an IRVB in ITER.</figcaption></figure>

105 $f\,s^{-1}$ and a sensitivity of the foil temperature of 15 mK is assumed. Using location and photon energy spectrum dependent models of the radiation from SOLPS (edge) and SANCO (core) we determine that the Pt foil thickness necessary to stop 90% of the incident power from the highest energy channel to be 36 µm. Using this thickness, we achieve an IRVB with 15 × 20 channels, a sensitivity of 3.8 $W\,m^{-2}$ and a SNR of 65 for the maximum signal of 250 $W\,m^{-2}$ from synthetic images shown in figure 18 calculated from the SOLPS and SANCO models [93]. In addition a feasibility study involving two tangentially viewing IRVBs (one viewing the entire crossection and one viewing the divertor) located in existing VIS/IR diagnostic locations in ITER was carried out including analysis of SNRs and nuclear heating [94]. The IRVBs each had 15 × 20 pixels and thicknesses of 30 and 10 µms, respectively, which should stop 95% of the radiated power incident on the foil. This resulted in SNRs of 77 and 59, respectively. Monte Carlo Nuclear Particle code analysis showed the neutron heating of the foil to be below the estimated signal noise level. However, heating by secondary gammas was on the order of half of the expected signal level, indicating the need for some sort of countermeasures.

### 2.2.4. Summary.
In conclusion, recent progress has shown that the IRVB should be applicable to ITER and future fusion reactors as a primary radiation diagnostic or one that is complementary to the conventional RBs and provides necessary redundancy. Future work will include tomographic modeling to determine the necessary foil thickness and spatial resolution. Future diagnostic design should include consideration of the necessary cooling of the foil frame and the IR optics

<figure><img src="image_27.png" /><figcaption>**Figure 18.** Synthetic image of radiation form SOLPS and SANCO models from the IRVB with the field of view shown in figure 17.</figcaption></figure>

(needed to bring the IR signal from the foil to the IR camera located behind the bioshield, similar to VIS/IR diagnostics) to insure adequate sensitivity of the diagnostic and effects of thermal radiation from hot components on the measurement signal in addition to ITER relevant prototype testing.

### 2.3. Impurity measurements by ITER VUV spectrometer

Most of the line emissions from plasmas in ITER are expected to be in the VUV or soft x-ray (SXR) wavelength ranges of 0.05–200 nm. These line emissions mainly originate from intrinsic impurities such as Be (the FW), O (oxidized layer on the FW), He (the residual gas after glow discharge or He ash), Cu (the hidden layer of the FW or divertor), Fe (diagnostic FW (DFW)), Ni (DFW) and W (divertor). Furthermore, line emission from extrinsically injected atoms for operations of a radiative divertor (Ar, Ne, and Kr) are also expected. Monitoring of these impurity species is required to diagnose the overall machine conditions and impending fault conditions during long-pulse operation.

In ITER, three sets of distinct VUV spectrometers in different places are employed to facilitate impurity monitoring with full coverage of different regions of plasmas in core, edge, and divertor as seen in figure 19 [95–97].

The design of the VUV core survey spectrometer is optimized for the monitoring of impurity species of all relevant impurity ions over a broad spectral range of 2.4–160 nm [95, 98]. The required sensitivity of the spectrometer is derived from the calculated photon emissivities of impurity species from plasma emission modeling of ITER using the SANCO impurity transport code [99] or SOLPS-ITER code [100]. High spectral resolution at 10 ms time resolution is accomplished
<figure><img src="image_28.png" /><figcaption>**Figure 19.** Overview of three ITER VUV spectrometers, VUV core survey spectrometer, divertor VUV spectrometer located in equatorial port, and VUV edge imaging spectrometer located in upper port.</figcaption></figure>

by implementing a five-channel spectrometer for the broad wavelength range of 2.4–160 nm. A spectral resolution, $\lambda/\Delta\lambda$, of several hundred (~500) with a high throughput and etendue of ~1 × 10⁻⁴ mm²sr resulted as the output of the design optimization for geometry parameters of a spectrometer using a toroidal shaped grating [101]. To verify the design of the ITER VUV spectrometer with regard to alignment and efficiency, a prototype VUV spectrometer has been developed [95, 96]. The prototype VUV core survey spectrometer with a wavelength ranging from 14.6 nm to 32 nm was installed at KSTAR after a calibration test in the laboratory [95, 96]. This custom-designed prototype spectrometer with toroidal grating and back-illuminated CCD detector was tested successfully, especially, in the tungsten injection experiment during the KSTAR experimental campaign of 2016.

The primary role of the ITER divertor VUV spectrometer is to measure radiation from impurity ions, especially tungsten ions, in the divertor region. In the previous design for ITER-98 in 1998, the divertor cassette in the ITER lower port was supposed to accommodate the ITER divertor VUV spectrometer [102]. However, it was found that the expected deposition on the field mirror would be severe in the lower port near the divertor plasma compared to the equatorial or the upper ports. The available space in the lower port is also very limited. For this reason, the equatorial port #11 of ITER accommodates the present ITER divertor VUV spectrometer, with one cylindrical field mirror in the port plug and one ellipsoidal collimation mirror after the bio-shield, as seen in figure 19 [96]. The wavelength range was chosen to be from 14.6 nm to 32 nm, targeting the emission from tungsten ions, which would be the main impurity species in divertor plasmas.

The VUV edge imaging spectrometer system located in the upper port is optimized for monitoring of the one-dimensional impurity distribution profile in the plasma edge region of 0.85 < r/R < 1.03 [97]. The wavelength range of 17–32 nm was chosen for the VUV edge imaging spectrometer to contain important spectral lines. About ten spatial-resolving (imaging)

points are to be monitored along a ~1000 mm long line across the flux surfaces in the plasma edge region, as seen in figure 19. The required time resolution for the VUV edge imaging spectrometer is 100 ms.

To minimize risks related to the degradation of the field mirror (the first mirror) due to deposition, erosion, and heating, the field mirror is set back from the plasma as far as possible while maintaining the required poloidal view. Ray-tracing analysis for the collection optics was conducted to allow the mirror solid angle to the plasma to be minimized, with a small entrance pupil at the shutter location. The wavelength range has been chosen so as to minimize reflectivity changes due to deposits of plasma impurities.

Since a full tungsten divertor was included in the baseline design of the ITER divertor in 2013, a wide range of research has been performed for the tungsten divertor, not only for engineering purposes but also to investigate the effect of tungsten ions on plasma performance. To study the effect of the tungsten ions on plasma performance, VUV spectrometers have been used to measure tungsten emission lines in various magnetic fusion devices such as JET (ITER-like wall (ILW)), ASDEX-U, WEST, LHD, EAST, HL-2A, and KSTAR. Recently dedicated experiments using tungsten pellets or injected tungsten dust have been performed to investigate tungsten emission lines in LHD, EAST, and KSTAR [96, 103, 104]. Through these experiments, emission lines from tungsten ions (around from W +24 to W +45) could be identified in the wavelength range of 2–7 nm. In the wavelength range of 15–30 nm, overall unidentified line arrays (quasi-continuum) are typically observed from low charge state ions of tungsten (e.g. W +6), and only a limited number of lines could be identified. Based on these results, the wavelength channel 2–7 nm of ITER VUV core survey spectrometer is expected to be the main channel to measure tungsten ions emission from main plasmas. Due to the limitation of the grazing angle of the first mirror ~16 degree for ITER divertor VUV spectrometer from the space constraints, this
<figure><img src="image_29.png" /><figcaption>**Figure 20.** Global view of the LIBS technique implementation.</figcaption></figure>

spectrometer was designed to monitor the low charge state lines of tungsten in the wavelength range of 15–30 nm.

## 2.4. Tritium monitoring

### 2.4.1. Laser induced breakdown spectrometer (LIBS).

LIBS is a technique for non-contact measurement of the multi-elemental atomic composition of a sample based on the production of a laser-induced plasma [105]. For applications dedicated to nuclear fusion, only the analysis of solid surfaces is developed [106, 107]. Heated by the absorption of photons on the diameter of a pulsed laser beam (typically picosecond or nanosecond), the material is ablated locally by forming a crater and is found in the state of plasma while emitting radiation during its relaxation (cf figure 20). This radiation first appears as a purely thermal continuum, then as first ionic and then atomic lines before taking the form of molecular bands depending on the surrounding conditions (nature of the gas and level of pressure) [108].

The measurement consists in triggering the acquisition at a time sufficiently long from the deposition of the laser energy to allow the observation of plasma spectral lines. The acquisition lasts long enough in relation to the typical time scales of the plasma life time in order to observe spectral lines of low intensity. The intensity of the observed spectral lines depends on the presence of the corresponding elements in the matrix. It also depends on how the sample is ablated. The relative size of the laser spot compared to that of the material grains, their composition, the internal composition gradients as well as the surface condition of the sample play an important role in the observation of the spectral lines. These dependencies are grouped under the term of ‘matrix’ effects [109].

The observed lines then allow the determination of the atomic composition of the sample in two ways. We can first

use calibrated samples of known composition and that are validated by alternative analysis techniques providing ‘calibration curves’. By simple comparison, the composition of the material can then be deduced [110]. This approach cannot work for samples of unknown composition or for samples that cannot be subject to a priori compositional control. It is therefore very limited.

The second approach, therefore more universal, consists in ensuring that the plasma is in local thermodynamic equilibrium (LTE). This is the ‘calibration free’ method [111]. This can be achieved if the plasma has had time to relax, that takes few hundred of ns. In addition, this state of equilibrium is better obtained if the plasma relaxes under a confinement obtained with a dedicated gas. The pressure of this gas can be lower or equal to $10^5$ Pa. This keeps its internal collision frequencies at sufficiently high levels. Working in argon is very often effective in this respect. Under these conditions, it is possible to reconstruct the spectra using spectral databases and compare them to the experimental one to derive the sample composition.

The LIBS lateral spatial resolution depends on the size of the laser spot on the sample and is improved by the use of a top-hat laser beam. The depth resolution that can be achieved with the LIBS technique is limited by the thermal diffusion in the sample after irradiation which depends on the duration of the pulse. Using an ns laser pulse, the ablated depth is often on the order of a few µm. If ps laser pulses are used, this depth can be reduced to a few hundred nm, but with a significant reduction in the intensity of the lines of interest. This is then very limiting because a shot-to-shot analysis must be carried out in order to measure the depth concentration gradient.

The activities developed around the LIBS technique applied to fusion materials analysis are developed in the frame of the ITPA coordinated activity along two major themes: laboratory
studies concerning the development of LIBS on beryllium relevant ITER samples including modeling activities and the integration of LIBS probes on remote handling arms for tokamak tests [112]. These developments bring together institutes, laboratories and universities as CORIA-CEA, ENEA, VTT, Comenius University, University of Tartu, IPPLM, INFLPR. In what follows, we will concentrate on beryllium-related activities. However, it should be pointed out that LIBS was already used in the mid 2000s to study the composition of codeposited carbon layers [113, 114]. It can also be used to assess the composition of boron codeposited layers as such obtain after boronization of a metal tokamak operating with W, for example [115].

### 2.4.1.1. Laboratory activities.
Beryllium is constituting the FW of the ITER tokamak, and is tested as a priority. It is of major importance to get a reliable technique able to measure with a good accuracy the concentration of hydrogen in the ITER vessel walls. An embedded LIBS operating at atmospheric pressure is considered as one of the possible diagnostic solutions. During LIBS experiments on Be bulk materials, micro-nanoparticles of Be are released due to the ablation process. LIBS experiments must therefore be carried out in a very strict framework given the toxicity of these dusts. Tests are also carried out on other materials considered to be substitutes for beryllium and easier to handle. The samples studied can be loaded with hydrogen (D, H, and T), N, He and the experiments are carried out in rare gases. The objectives focus on the experimental conditions allowing the spectral resolution of the α lines of the Balmer series of hydrogen isotopes (the most intense) [116], in particular between Dα and Tα separated by 0.054 nm. These measurements are carried out in ns and ps mode [117], in accumulated or shot-to-shot mode in order to test the profilometric measurement capacities. The measurement must be coupled with ex situ determinations of the depth ablated by pulse. The results are compared with those obtained by other surface analysis methods as RBS, NRA, TOF-ERDA, SIMS, etc [118]. The agreement is very satisfactory. It should be noted that the first LIBS measurement of the tritium content of a sample was carried out in 2022 by CORIA-CEA. Tests were also carried out on deuterated tiles that had stayed in JET and made available to other partners. Deuterium and hydrogen are measured, as well as metallic pollution [119]. The measurement of helium in the sample is also tested [120]. Helium can be implanted in tokamak at the surface of the material due to plasma implantation. It can also be produced in the bulk material due for instance to tritium decay. However, helium measurement turns out to be very difficult due to the high volatility of this element and the difficulty in exciting it electronically. Double-pulse experiments are currently being carried out to test the heating of the plasma produced by a first pulse in order to increase the population density of the upper level of the transition, and therefore the intensity of the emission line [121]. This is done in picosecond regime in order to keep the capacity of profilometric studies [113].

These experiments are carried out in parallel with modeling studies aimed at reproducing the results and allowing their

optimization. Two axes are developed. The first one consists of calculating the expected spectrum under LTE conditions and verifying that the necessary databases are sufficiently accurate. This reconstruction of the spectra is based on the resolution of the radiative transfer equation [122]. The second axis exploits the capabilities of artificial intelligence AI [123]. An extensive database is acquired under controlled conditions over wide ranges of parameters of interest such as hydrogen content, plasma temperature and electron density, etc. Then the AI learns to recognize from this base the characteristics of the plasma emitting the new spectrum that it can observe during an additional experiment. The measurement is thus optimized and made significantly faster.

### 2.4.1.2. LIBS integration in tokamak.
LIBS tests were carried out manually in the EAST tokamak in particular to detect metallic impurities on the limiter. W, Mo and Fe were observed [124, 125]. A compact LIBS probe has been specifically developed to perform these experiments.

Other tests have been carried out by exploiting the capacity of remote handling using robot arms. A solution was tested by ENEA in the FTU tokamak [126]. A compact ns laser source is positioned at the end of a remote arm. The analysis is being carried out by a spectrometer which is also external. A suitable metal cone is positioned on the optical axis in the immediate vicinity of the surface to be analyzed in order to sweep the area with an argon flow allowing the production of laser-induced plasmas in LTE. Metallic pollution has been analyzed as well as the presence of hydrogen and deuterium.

Another solution was tested in the WEST tokamak. A LIBS probe was designed and implemented by CEA-CORIA to equip the end of the WEST articulated inspection arm; the AIA [127]. The probe works at atmospheric pressure. The laser source is ns type. It is positioned outside the tokamak. Two optical fibers are used. The first fiber carries the laser pulse to the head of the probe equipped with adapted optics allowing it to be focused without loss on the wall to be analyzed. The second fiber collects part of the light emitted by the LIBS plasma which leads it to the spectrometer also placed outside the tokamak for analysis. Measurements were made on the tokamak divertor. The metallic impurities have been also measured.

As part of the ITPA activities, the teams intend to carry out the validation of the measurement by LIBS of the quantity of tritium contained in ITER-relevant beryllium samples in the form of deposits but also massive materials.

### 2.4.2. Real-time residual gas analysis.
Residual gas analysis in real-time, i.e. including during plasma operations, will be provided on ITER by DRGAs. DRGAs will measure the neutral gas composition in a divertor port and an equatorial port during plasma operation. Plasma-relevant response time is enabled by assuring most practical and safe proximity to the regions to be sampled (i.e. divertor pumping ducts and main chamber, respectively) and plasma operations-compatibility is assured by selection of radiation resilient sensor heads and the removal of electronics to the diagnostics building.
a)

<figure><img src="image_30.png" /><figcaption>**Figure 21.** (a) RGA integrated into the equatorial 11 port cell support structure [115] and (b) analysis chamber.</figcaption></figure>

b)

<figure><img src="image_31.png" /></figure>

US-DA has procured two DRGA subsystems, which comprise a connection from the vacuum chamber to an analysis chamber in the port cell (figure 21(a)). The analysis station includes not only quadrupole mass spectrometers (commonly called RGAs) but also optical gas analyzers, OGAs, which use a low temperature plasma to provide the excitation of the neutrals (figure 21(b)). A combination of the two technologies is critical for resolving fusion fuel cycle-relevant gases. Unlike other Type-2 diagnostics (e.g. the Survey VUV system) the pumping system that is integrated with the DRGA is also a critical part of the measurement. A detailed description of DRGAs, as well as the physics basis for the design and challenges in its completion, is provided in Klepper et al [128].

DRGAs at the divertor will play the largest role in the ITER Research Plan and, for this reason, a scope expansion is under consideration to provide two additional systems for the divertor for redundancy. By accessing the sub-divertor region, these systems will be most important for providing real-time information on the hydrogen isotopic content (i.e. H/D/T) in the main plasma, and on the D-T burn via detection of $^4$He produced as fusion ash and then being channeled by the divertor into the pumping ducts. As such, they provide a global measurement that correlates well with the main plasma content, in plasma-wall equilibration timescales (a few seconds) [129].

DRGAs will also play an important role in analyzing impurities detached from the inner wall, especially important for various wall conditioning techniques: baking, glow discharge cleaning (GDC), ion cyclotron wall conditioning (ICWC)/electron cyclotron wall conditioning (ECWC). Baking of the VV and blanket/FW at 200 °C and 240 °C will remove water and part of hydrogen isotopes, while H or He ions bombard the inner wall to desorb/desaturate surfaces and remove medium Z impurities. Because the toroidal magnetic field will be on during the plasma campaign, the use of GDC as the intershot wall conditioning technique will be limited. ICWC and ECWC need to be developed for the conditioning of the FW surfaces between pulses. In order to optimize the efficiency of those wall conditioning techniques, qualitative

and quantitative assessment of the exhaust gas must be performed. Sampling the sub-divertor through a <~1 mm diameter pressure reducing orifice, the divertor DRGA is optimized for high neutral gas pressure measurements (~0.1–10 Pa). Therefore, for new applications, such as LID-QMS for T monitoring between shots (Sub-section 4.2) the Equatorial Port DRGA (EP-DRGA, integrated into EP-11) will likely be used, as its sampling pipe is less constricted, designed to dynamically sample main chamber neutral pressures in the 0.001–0.1 Pa range, albeit with ~10 s response time. The EP-DRGA design is less advanced than the divertor DRGA, but the interface between DRGA and T-monitor (i.e. LID-QMS) is currently in work between the two diagnostic teams.

## 2.5. Technical challenges in the measurements of the plasma-wall interaction at ITER

2.5.1. *Elimination of reflection caused by metallic wall.* The performances of plasmas strongly depend on the ability to simultaneously monitor and control the plasma-wall interaction for protecting the walls from excessive heat loads. Infra-red (IR) measurement is a key diagnostic to provide thermal images of the plasma facing components (PFC) under plasma exposure. Nevertheless, with the introduction of all-metal walls in today's fusion devices, additional difficulties result in the interpretation of IR measurements through disturbance phenomena such as reflections and/or emissivity variation with surface temperature and plasma exposition (surface erosion/deposition on monitored PFCs). This can lead to inaccurate PFC surface temperature estimation and interpretation of 'false hot spots'. Currently there are no experimental techniques mature enough to overcome these phenomena. A numerical approach is being developed to predict and solve unknown emissivity and reflections features. Such approach is based on an end-to-end simulation (or so-called synthetic diagnostic) able to model all physical phenomena involved in the IR measurement chain (plasma-wall interaction, thermic, photonic and optic).
<figure><img src="image_32.png" /><figcaption>**Figure 22.** Infrared experimental image (left) and simulated image (right) of ITER-like wide angle tangential view of WEST tokamak. The IR ray tracer is able to separate the thermal events from false hot spots caused by multiple reflections and provide the origin of reflections. Reproduced from [133]. CC BY 4.0.</figcaption></figure>

Ray tracing simulation is also a remarkable tool to intuitively anticipate photon behavior in complex geometries with changing optical materials properties. The ray tracing code developed by the company ANSYS-OPTIS has already been applied on several devices (Tore Supra, WEST, JET, AUG, W7X, ITER) for improving the understanding of IR images in a fully reflective and radiative environment [130–132]. Applied to WEST and AUG, this allowed discriminating both the real heat load deposition from false hot spots (figure 22), while demonstrating the potential impact of a variable emission model on plasma-facing units (PFUs) [133, 134]. Another open-source ray tracing code, Raysect, has been used for modeling volume sources and tested on the AUG bolometry system [135, 136]. $W_{\text{ray}}$ and $W_{\text{ana}}$ are package simulators being newly developed for the ray tracing and reflection analysis for KSTAR [137]. The technique is being built on the simple idea that the reflections can be discriminated from the true signal when all ray information is given since an optical system (reflection and refraction) is an intrinsically linear system. $W_{\text{ray}}$ is a ray tracing code finding all input ray paths, while $W_{\text{ana}}$ is a reconstruction code finding the true signal without the contributions of reflections. $W_{\text{ray}}$ utilizes bound volume hierarchy (BVH) and multi-thread processing for high speed. The code is based on ‘Unity’ to reduce the coding load outside ray-tracing, and the photo image generated by $W_{\text{ray}}$ is shown in figure 23.

Once all ray tracing roots are given, the reflection elimination in a real geometry can be examined. The algorithm is applied to the temperature profiling of the KSTAR PFCs with assumptions of uniform temperature over a single PFC and toroidal symmetry. Figure 24 shows that the reflection is gradually eliminated through error-back-propagations.

In parallel, several laboratory experiments have been conducted to establish a comprehensive model of the material emission and reflectance as a function of roughness, temperature, wavelength, and machining. Before being exposed, it is found that the tungsten emissivity ranges around 0.1–0.2 at 4 µm with a low temperature dependence [138, 139].

These laboratory experiments have proven the angular dependency and a highly specular response of reflectance, which has been confirmed by comparing simulated and experimental images in WEST and AUG. Moreover, these measurements also reveal that the behavior of emission and reflectance varies significantly with sample roughness and machining process. This has been confirmed by experiments in WEST and AUG, showing that thermal optical properties of materials change as the plasma operation progresses [140]. The challenge is now how to monitor these modifications without direct access to in-vessel components. Simulations coupled with dedicated experiments are essential. In WEST, a method based on the double heating process was developed to estimate *in-situ* the emissivity of the W-coated graphite divertor tiles in the WEST tokamak [140]. This method takes advantage of the divertor temperature increase during the plasma operation and uses photonic calculation to disentangle the emitted and reflected parts in the measured radiance. The results show a strong variation of the emissivity along the divertor W surfaces with a factor 4 variation after the experimental campaigns. An alternative method is also currently explored in to identify the emissivity or temperature profile on the divertor PFCs by solving an inverse problem. The method consists in solving a least squares problem that minimizes iteratively the difference between the measured IR image and a modeled image built with a forward radiative model in which all reflections are taken into account [141]. From a numerical prototype, it is shown that target emissivity value can be recovered from a controlled vessel baking scene (with uniform and known temperature) using an inverse thermography method for filtering reflections with an error of 6%. When the emissivity profile is identified, the same method is used to recover the temperature with an accuracy better than 3% after removing the reflections [142]. Recent new methods based on machine learning techniques have also achieved encouraging results in filtering reflections and retrieving the surface temperatures with high accuracy (better than 10%) without knowing emissivity [143]. The next step is to validate all these new numerical approaches on controlled experiments in laboratory before they are automatically used on tokamaks. For the KSTAR tokamak, an experimental setup, $W_{\text{emi}}$, for the measurement of PFC surface properties is constructed and measured the BRDF as shown in figure 25. $W_{\text{emi}}$ is equipped with two surfaces mimicking the reflection by W surface near the light source (e.g. target strike point region), where two surfaces are located very closely. For instance, the V-shaped neutral reflector plates at the bottom of the ITER divertor would be examples. In $W_{\text{emi}}$, the second surface can make an angle between 45° to 90° and W surfaces with various roughness are ready to measure.

Figure 26 shows examples of the emissivity and reflectance measurements by $W_{\text{emi}}$. The temperature of the primary tungsten tile (at 90°) is varied from room temperature up to 700 °C. The emissivity at the temperature of 250 °C was 0.15 and it increases to 0.35 at 500 °C. After that, the emissivity seems to be saturated, but this needs to be confirmed. The reflectance from room temperature up to about 500 °C was about 1 %, and then it increased exponentially to 5.3% at a temperature
<figure><img src="image_33.png" /><figcaption>**Figure 23.** A typical photo image generated with 'W<sub>ray</sub>'.</figcaption></figure>

<figure><img src="image_34.png" /><figcaption>**Figure 24.** The algorithm test for the elimination of the reflections with error back propagations.</figcaption></figure>

<figure><img src="image_35.png" /><figcaption>**Figure 25.** The setup for the emissivity and reflectance measurement of PFCs.</figcaption></figure>
<figure><img src="image_36.png" /></figure>

<figure><img src="image_37.png" /><figcaption>**Figure 26.** (a) The emissivity and (b) reflectance measurement of W PFC.</figcaption></figure>

<figure><img src="image_38.png" /></figure>

<figure><img src="image_39.png" /><figcaption>**Figure 27.** The temperature measurements affected by adjacent tiles.</figcaption></figure>

of 700 °C. Both measurements show that the emissivity and reflectance have been changed as a function of the tungsten temperature.

Figure 27 shows the temperature measurements of two tiles which make an angle of 45° with respect to each other. The temperatures of both tiles were measured by both thermocouples and an IR camera looking at the tilted tile. The bottom tile was heated from room temperature up to 650 °C, while the tilted tile was not heated directly. Nevertheless, the temperature of the tilted tile was increased by radiated heat from the bottom tile. The temperature measured by the thermocouple inserted in the bottom tile was used as X-axis. As the temperature of the bottom tile increases, both temperatures measured by the thermocouple and IR show different behavior. The temperature measured by the thermocouple increases slowly and is saturated at a temperature around 300 °C, while the temperature measured by the IR camera increases further up to 600 °C. It is evident that the IR camera did not measure the temperature of the tilted tile, but it measured the temperature of the bottom tile by reflection. Further measurements and modeling of the IR signature will be performed.

### 2.5.2. Deposition of impurities on first mirrors.
The ITER VUV spectrometers use mirrors to extract the impurities' emission light from divertor and edge plasma regions to the detector. The first mirror (field mirror) is the front-end

element of the plasma-viewing optical system and directly exposed to deposition by neutral particles from the wall (subchapter 8.2). The deposition on this mirror is the main cause of the degradation that is detrimental to accurate measurement of the optical signal. Therefore, passive mitigation of impurity deposition was adopted as the mitigation strategy for the first mirror of the ITER VUV spectrometer, due to difficulty to apply a mirror cleaning technique considering the required roughness value of VUV wavelength range (sub-nanometer).

As will be described in subchapter 8.2, applying a baffled duct is one of the representative deposition mitigation techniques. Simulations as well as various experiments with different geometries of the baffled ducts have been performed in LHD, and TEXTOR. Especially, systematic mitigation efficiency tests by baffled duct geometries are performed with three differently shaped baffled ducts (short, long, short + expanded volume) in KSTAR.

The results are summarized in table 11. Here, the degradation of the mirror was measured ex-situ by measuring the reflectivity using the Dα line (654.9 nm) reflectance on a visible mirror.

The investigation of the *in-situ* deposition or erosion effect on the mirror surface with the baffled duct was conducted by using quartz crystal microbalances (QCMs) installed in KSTAR. The detailed result of *in-situ* deposition measurements is shown in chapter 2.
<table><caption>**Table 11.** Measured reflectivity of gold coated mirror after deposition of the mirrors at KSTAR corresponding to each geometry of baffle duct in front of the mirror.</caption><thead><tr><th>Baffle geometry</th><th>Absolute reflectance (Relative Reflectance %)</th></tr></thead><tbody><tr><td>Short baffled duct</td><td>44.3 (46.8)</td></tr><tr><td>Long baffled duct</td><td>59.6 (63.0), 56.9 (60.1)</td></tr><tr><td>Short baffled duct + expanded volume</td><td>90.7 (95.8)</td></tr><tr><td>Reference mirror</td><td>94.6 (100)</td></tr></tbody></table>

## 2.6. Summary

In this subchapter, we have briefly reported the recent advances in diagnostics for the measurements of plasma-wall interaction in support of the ITER research plan. Other than describing the progress of the preparation of basic and conventional diagnostics for the plasma-wall interaction such as Langmuir probe arrays or visible spectroscopy, the research was focused on the new developments in diagnostics and measurement techniques required to achieve the ITER research goals.

Radiated power can be accurately measured by the IRVB, which has been developed and tested on several tokamaks. With sophisticated techniques of ray tracing and reflection correction algorithms, the critical issue for the estimation of divertor heat flux caused by the reflection of plasma radiation under the metallic FW device will be solved. The characteristics of the reflecting surfaces depending on the roughness and angle of incident have been studied by a dedicated experiments and the results were applied to the developed codes for the reflection elimination.

For the measurement of the metallic impurity radiation induced by eroded metallic atoms, a VUV spectrometer has been developed.

Tritium inventory on the FW will be monitored by the LIBS technique, and DRGAs will measure the neutral gas composition in a divertor port and an equatorial port.

Those FW diagnostics will play a key role in the ITER operation and are essential for the successful execution of the ITER research plan.

# 3. Status of the diagnostics of fusion products at ITER

M. Scholz⁹, B. Esposito², D. Marocco¹⁰, F. Belli¹⁰,
L. Bertalot², B. Coriton², V. Gin², D. Gin², J. Dankowski²,
A. Hjalmarsson¹¹, V. Krasilnikov², G. Ericsson¹¹,
M. Tardochi¹², D. Rigamonti¹², M. Nocente¹²,
M. Garcia-Munoz¹³, M. Ishikawa¹⁴, M. Cheon⁸ and J. Jo¹⁵

## 3.1. Introduction

ITER will be the first tokamak in which the ratio of fusion power to plasma heating power will be greater than $Q = 1$ and finally should reach $Q = 10$. ITER will be the first fusion device with burning plasma, in which the collective behavior of alpha particles and of other fast and thermal ions will

occur. For this reason, measurements of neutrons and confined or lost fast ions, with an emphasis on alpha particles, are crucial for ITER. Diagnostics related to these measurements will be important for future experiments in burning plasma at ITER. High neutron emission and the very large plasma size in ITER make the neutron diagnostics the main diagnostics used to measure plasma parameters such as the fusion power, the fusion power density, the ion temperature, the fast ion energy and their spatial distributions in the plasma core.

The set of neutron diagnostics for ITER should meet the specific requirements of long life under the influence of large neutron fluxes and fluences, estimated to between 30 and $10^4$ times greater than those occurring during the maximum DT discharge at JET. This means that the neutron diagnostic systems will work under a strong radiation load, which will require the use of massive radiation shielding. Thus, neutron diagnostics in ITER such as the neutron cameras and spectro-meters will have limited angular fields of view, which challenges their absolute calibration.

The neutron diagnostics under development for ITER are in various stages of readiness and include: the radial (RNC) [144] and vertical (VNC) [145] neutron cameras, the internal [146], external [147] and divertor neutron flux monitors (NFM), the NAS [148] and the HRNS [149].

As in the case of neutron emission measurements, a very important issue in large tokamaks such as ITER is the measurement of confined and lost alpha particles. The problem will be significant for ITER. For confined alpha particles, various diagnostic proposals are considered based on CX recombination spectroscopy (CXRS) or on the measurement of the neutron high-energy tail produced by alpha knock-on energetic ions [150] using HRNS. A passive method that can be used is the measurement of alpha particles by gamma radiation spectroscopy using the reactions $^9\text{Be}(\alpha, n\gamma)^{12}\text{C}$ and $^{10}\text{B}(\alpha, p\gamma)^{13}\text{C}$ [151].

This subchapter on diagnostics of FPs for the ITER experiment covers in the next section the measurement specifications and the operational role of each diagnostic for FPs. Diagnostics of alpha particles and neutrons are covered in the following sections (from 3 to 8). Section 9 describes the in-situ calibration strategy for neutron diagnostics. In section 10 we give a short description of alternative diagnostic techniques to measure fast particles excluding gamma spectroscopy and FILD detectors, which are described in this subchapter (see section 3). The last section of this subchapter is a summary.

## 3.2. Operational role and measurement specifications of FP diagnostic subsystems

The burning plasma parameters that will be measured in ITER with the required accuracy, measurement ranges and resolutions are presented in table 12. All these parameters are divided in three categories (see column 2 in the table) according to their following role:

* (1a) measurements for MP and BC,

* (1b) for AC;

* (2) for performance evaluation and physics.
<table><caption>**Table 12.** Parameters measured by ITER fusion product diagnostics, their role and requirements.</caption><thead><tr><th>Parameters</th><th>Role</th><th>Range</th><th>Time resolution</th><th>Accuracy</th></tr></thead><tbody><tr><td>Fusion power</td><td>1a</td><td>0.1–900 MW</td><td>1 ms</td><td>10%</td></tr><tr><td>Fusion power density</td><td>2</td><td>1–4 10<sup>3</sup> kWm<sup>-3</sup></td><td>1 ms</td><td>10%</td></tr><tr><td>Total neutron flux</td><td>1a</td><td>10<sup>14</sup>–3.2 · 10<sup>20</sup> ns<sup>-1</sup>m<sup>-3</sup></td><td>1 ms</td><td>10%</td></tr><tr><td>Neutron-and-alpha source profile</td><td>1b</td><td>10<sup>14</sup>–1.4 · 10<sup>18</sup> ns<sup>-1</sup>m<sup>-3</sup></td><td>1 ms</td><td>10%</td></tr><tr><td>Core ion temperature</td><td>1b</td><td>0.5–40 keV</td><td>100 ms</td><td>10%</td></tr><tr><td>n<sub>T</sub>/n<sub>D</sub> fuel ratio in plasma core</td><td>1a</td><td>0.01–10</td><td>100 ms</td><td>20%</td></tr><tr><td>Neutron fluence on the first wall</td><td>1b</td><td>0.1–1 MWm<sup>-2</sup></td><td>10 s</td><td>10%</td></tr><tr><td>Confined α particle energy and spatial distribution</td><td>2</td><td>0.1–1 MWm<sup>-2</sup> (0.1–2) · 10<sup>18</sup> m<sup>-3</sup></td><td>100 ms</td><td>20%</td></tr><tr><td>Fast ion energy and spatial distribution</td><td>2</td><td>0.1–1 MeV</td><td>100 ms</td><td>20%</td></tr></tbody></table>

<table><caption>**Table 13.** Fusion product diagnostic systems.</caption><thead><tr><th>Sub-system</th><th>Types</th><th>Meas. parameter no. (primary role in bold)</th></tr></thead><tbody><tr><td>Micro-fission chamber (In Vessel)<br>External neutron flux monitor (In Port)<br>Divertor neutron flux monitor</td><td>A-Neutron flux monitor (NFM)</td><td>1,2 (and 5 for external NFM only)<br>(MFC has no primary role)</td></tr><tr><td>Neutron activation system</td><td>B-NAS</td><td>1,2,5,8,9</td></tr><tr><td>Radial neutron camera<br>Vertical neutron camera</td><td>C-Neutron profile monitors (NPM)</td><td>1,2,3,4,5,6,7,8,9</td></tr><tr><td>High resolution neutron spectrometer</td><td>D-Neutron spectrometer (HRNS)</td><td>1,4,6,7</td></tr><tr><td>Gamma ray spectrometer</td><td>E-Radial gamma ray spectrometer (RGRS)</td><td>2,4,6,7</td></tr><tr><td>Lost ions probes</td><td>F-Fast-ion loss detectors (FILD)</td><td>6,7</td></tr></tbody></table>

The category 1a and 1b means that the machine cannot be operated without working diagnostics providing every 1a parameters and 1b parameters, respectively.

The FP diagnostics of ITER are presented in table 13. They are grouped into six types A—F as shown in table 13.

The neutron diagnostics play an important role in measuring plasma parameters in burning plasma experiments. The main parameters measured by this system are the reaction rate, i.e. fusion power, the neutron source/alpha profiles and the fuel ratio ($n_T/n_D$). Each of these parameters is measured by one or two primary systems and should be supported by a different secondary subsystem to ensure high reliability of the measurements. Primary diagnostic means that this system is well-suited to performing a given measurement. The relation between the above mentioned parameters and the subsystem is shown in table 14.

For example, from table 14, it can be seen that the fusion output can be determined by a combination of a NFM with

the required time resolution (A—primary), but also its measurement can be assisted by the NAS (B), the neutron profile monitors (C) and the neutron spectrometer (D). NFMs can be calibrated *in situ* because their sensitivity can change during ITER lifetime. The activation system is highly reliable and free from electromagnetic contamination, although it is not time resolved. Correction for plasma position and neutron source profile changes will be provided by the profile monitors (C—supporting).

Additionally, the High Resolution Neutron Spectrometer (HRNS, B—supporting) can independently provide the absolute fusion output by sampling a line—integrated volume near the center [152]. This example shows that such an organization of the whole neutron system for ITER (see tables presented) should allow for the cross-checking of the mentioned subsystems and their complementarity. This will allow to obtain reliable results of neutron measurements with a given accuracy in the case of burning plasma experiments in ITER.
<table><caption>**Table 14.** The relation between parameters to be measured and neutron subsystem to be used.</caption><thead><tr><th></th><th colspan="4">The neutron subsystem to be used.</th></tr></thead><tbody><tr><td>Parameters to be measured</td><td>A-Neutron flux monitor (NFM)</td><td>B-Neutron activation system (NAS)</td><td>C-Neutron profile monitors (NPM)</td><td>D-High-resolution neutron spectrometer (HRNS)</td></tr><tr><td>Fusion power</td><td>Primary</td><td>Supporting</td><td>Supporting</td><td>N/A</td></tr><tr><td>Neutron source profiles</td><td>Supporting</td><td>Supporting</td><td>Primary</td><td>Supporting</td></tr><tr><td>The fuel ratio (n<sub>T</sub>/n<sub>D</sub>).</td><td></td><td></td><td>Supporting</td><td>Primary</td></tr></tbody></table>

More detailed discussions about the status and development issues of individual neutron diagnostic systems are included in the following sections.

### 3.3. Alpha particle measurements

#### 3.3.1. Gamma ray spectroscopy.
Gamma-ray spectroscopy is a recent technique based on the detection of gamma-rays in the energy range of MeV. In a fusion plasma, gamma-rays are mainly emitted from the de-excitation of a nucleus as the result of nuclear reactions occurring between impurities and fast ions. In order to have a significant gamma emission, the ions need to be energetic, typically with energies higher than 0.5 MeV. The spectroscopic measurement of the emitted gamma-rays allows to study the energy distribution of the fast ions and their slowing down in the plasma. When the gamma-rays are measured along many LOS, the spatial profile of the fast ions can be inferred. ITER will explore for the first time the physics of a burning plasma in which the heating fraction released by the slowing down of the fusion α particles exceed that of the external heating system. For this reason, the study of the α particles will be crucial. This can be done by observing the 4.44 MeV gamma-ray line emitted by the $^{12}$C(n, α)$^{9}$Be reactions that occur in the plasma between the fusion α particles and the $^{9}$Be impurities that are present due to the erosion of the ITER FW. The measurement of the intensity and the shape of the 4.44 MeV peak requires the use of dedicated instrumentation with energy resolution better than 1.5% and with counting rate capability higher than 1 MHz. The conceptual design of the Radial Gamma-Ray Spectrometer (RGRS) system for ITER has been studied and will be finalized by the end of the 2026. The system consists of a set of gamma-ray spectrometers placed along the radial line of sights (LOSs) and are integrated in the radial neutron camera (RNC) (figure 37, section 3.4) which is described in section 4. The design of the system took advantage on the experience gained at the Joint European Torus (JET) which is the fusion experiment equipped with the most advanced set of gamma-ray diagnostics. The selected scintillator crystals are large LaBr$_{3}$ (3" × 6", diameter, height) coupled to photomultiplier (PMT) tubes provided with active voltage dividers. To reduce the direct 14 MeV neutron background, the RGRS will be provided with 120 cm long lithium hydride (LiH) attenuators [153] placed in front of each detector. To partially reduce the load of the gamma-ray background induced by neutrons interacting with the tokamak structure and the surrounding materials,

<figure><img src="image_40.png" /><figcaption>**Figure 28.** 4.44 MeV γ-ray (right) emissivities (particles m$^{-3}$ s$^{-1}$) in the poloidal plane for the ITER 500 MW DT scenario
Reproduced from [156]. © 2017 IAEA, Vienna. All rights reserved.</figcaption></figure>

a shielding based on iron and concrete has been designed to attenuate the radiation load coming from the side of the detectors. The RGRS system consists of three dedicated LOS of 40 mm diameter and a shared one with the RNC whose diameter is 25 mm (see figure 37).

The expected 4.44 MeV emissivity for the full power 500 MW ITER scenario has been calculated with the GENESIS code [154] (see figure 28) while the expected pulse height spectrum measured by the LaBr$_{3}$ scintillators has been simulated with an MCNP code (see figure 29) [155]. Calculations revealed a count rate of 12 kHz of the 4.44 MeV signal when a LoS of 40 mm is used, while a count rate of 1.8 kHz is calculated for the 25 mm diameter collimator.

#### 3.3.2. FILD detectors.
Energetic particle losses are routinely measured in present tokamaks using charged particle collectors located at the edge of the plasma. Most Fast-Ion Loss Detectors (FILD) [157] use, as active component, Faraday Cups (FCs), scintillators or a combination of both [55, 157–169]. Charged particle collectors are often complemented with IR measurements of the heat load caused by the escaping ions on the PFCs. While charged particle collectors give velocity-space resolved measurements of escaping ions with Alfvenic temporal resolution at a certain position, IR-measurements can cover the entire 3D wall. Just IR-measurements would not be sufficient to diagnose fast-ion losses as they cannot discriminate between thermal and
<figure><img src="image_41.png" /><figcaption>**Figure 29.** Expected pulse height spectrum on LaBr3 (40 mm diameter collimator) for the 500 MW DT scenario at ITER. Reproduced from [156]. © 2017 IAEA, Vienna. All rights reserved.</figcaption></figure>

<figure><img src="image_42.png" /><figcaption>**Figure 30.** CAD view of FILD head with some escaping ion trajectories hitting the detector. In blue particle trajectories which enter into the scintillator chamber and in red particle trajectories blocked by the graphite protection. The inset on the right-bottom part of the figure shows the collimating process. Reprinted from [157], with the permission of AIP Publishing.</figcaption></figure>

fast-ion losses, give no information on the velocity-space of the losses or typically have limited temporal resolution well below the Alfvénic time-scale.

### 3.3.2.1. FILD operational principle in-port RNC detectors.

FILD systems work as magnetic spectrometers dispersing the escaping ions by means of a collimator on a scintillator plate. The hit points of the impinging ions on the scintillator plate are given by the particle energy (gyroradius) and pitch-angle ($v_{par}/v_{tot}$). Figure 30 shows the head of a FILD system mounted on the midplane manipulator of the ASDEX Upgrade tokamaks. The inset illustrates the FILD operational principle with particles with different trajectories hitting the scintillator plate at different positions. Scintillator-based FILD systems often use FCs embedded in scintillator plates to obtain absolute measurements of escaping ions. Figure 31 shows an AUG

<figure><img src="image_43.png" /><figcaption>**Figure 31.** (a) Lateral view of detector head without graphite protection. Trajectories of particles with three different pitch angles, 87°, 70°, and 50°, and two different gyroradii, 3 and 6 cm, entering through the collimator and hitting the scintillator plate are also shown. (b) Inside view of the detector head with the simulated trajectories of particles with pitch angles 87°, 70°, and 50°. On the bottom the thermoelement is indicated. The particle trajectories were calculated forward and backward from the first aperture slit with the starting point randomly distributed over the whole slit. Reprinted from [157], with the permission of AIP Publishing.</figcaption></figure>

<figure><img src="image_44.png" /><figcaption>**Figure 32.** Pitch-angle—gyroradius strike map with PMT array layout overlaid. Reprinted from [157], with the permission of AIP Publishing.</figcaption></figure>

FILD detector head without graphite protection with the main components of the detector head indicated in red.

Most present FILD devices measure simultaneously the escaping ions hitting the scintillator plate with three systems; a FC embedded in a scintillator plate, a CCD camera to look at the escaping ion velocity-space and a PMT tube array with Alfvénic temporal resolution to obtain MHz measurements of escaping ions fluxes. The layout of the PMT array covering the entire scintillator plate of an AUG FILD is shown in figure 32.

### 3.3.2.2. Measurement requirements for ITER.
Following this operational principle, the ideal ITER FILD system should fulfill the following measurement requirements;

#### A. temporal resolution;

FILD with Alfvénic temporal resolution is highly recommended in order to identify the MHD fluctuations responsible for the measured fast-ion losses in a sea of MHD fluctuations. A MHz temporal resolution is set by the Alfvénic time, $\tau_A = R_0/v_A \sim 1$ µs. This is obtained for a standard ITER DT case with an Alfvén speed $v_A = 8 \times 10^6$ m s$^{-1}$
<figure><img src="image_45.png" /><figcaption>**Figure 33.** ASCOT simulations of alpha particle heat load in ITER on a 3D first wall due to an externally applied $n = 4$ RMP. The location of the FILD head is indicated with a white box. Reprinted from [55], with the permission of AIP Publishing.</figcaption></figure>

calculated with $B_T = 5.2$ T, a 50%-50% D–T mixture and an electron density, $n_e = 8 \times 10^{19} \text{ m}^{-3}$

### B. spatial resolution;

The heat load spatial pattern shown in figure 33 gives an estimate of the spatial resolution the FILD should have in ITER. A spatial resolution similar to the gyroradius of the fusion born alpha-particles, $r_L \sim 7$ cm, seems to be required to be able to resolve the expected heat load wall structures.

### C. Velocity-space range and resolution;

Wave-particle resonances can occur over almost the entire fast-ion phase-space. In order to identify the wave-particle resonances responsible for the actual fast-ion transport/loss, a wide velocity-space coverage and reasonably good energy and pitch-angle resolution are thus necessary.

Figure 34 shows the velocity-space of the escaping ions collected at the head of the FILD system with its aperture located 11 cm outside of the DFW in the ASCOT simulations presented here. The losses appear over the entire energy range, from ~6.5 cm (corresponding to the alpha-particle birth energy) down to the thermal energy at large pitch-angles, ~80°. Similarly, a broad escaping ion pitch-angle range, from 40° to 85°, with an energy similar to the birth energy, appears in the simulated losses at the FILD system. As in recent experimental observations, clear structures at certain pitch-angles are observed.

#### 3.3.2.3. ITER FILD synthetic measurement.
The ITER FILD system will likely be a charged particle collector composed of both a scintillator and a FCs' matrix, located in the 55.B9 port plug. Recently, a conceptual design of a reciprocating FILD system with outstanding velocity-space resolution has been developed [157]. Full orbit ASCOT simulations have been used to estimate the flux of fusion-born alpha particles on a realistic 3D wall. Externally applied resonant magnetic perturbations (RMPs) were used to mimic the effect of symmetry

<figure><img src="image_46.png" /><figcaption>**Figure 34.** ASCOT simulations. Velocity-space of alpha particle losses hitting the FILD detector head in the presence of an externally applied $n = 4$ RMP. Reprinted from [157], with the permission of AIP Publishing.</figcaption></figure>

<figure><img src="image_47.png" /><figcaption>**Figure 35.** A synthetic FILD measurement of alpha particle losses induced by the applied $n = 4$ RMP. Reprinted from [157], with the permission of AIP Publishing.</figcaption></figure>

breaking 3D magnetic perturbation on the ITER alpha particle population.

A synthetic FILD measurement of alpha particle losses induced by the applied $n = 4$ RMP, including scintillator background emission produced by neutron-driven nuclear reactions in the scintillator coating itself, is shown in figure 35.

The synthetic CCD camera shows several populations of alpha particle losses with well-defined pitch-angles and different energies illustrating the potential of such a diagnostic to contribute to the ITER success.

### 3.4. RNC

#### 3.4.1. Motivation.
The primary role of the RNC is to provide—through reconstruction techniques applied to the line-integrated neutron fluxes—time-resolved measurements of the neutron and $\alpha$-source profile (i.e. Neutron Emissivity, neutrons emitted per unit time and volume ($s^{-1} \text{ m}^{-3}$)) and the fusion power density. Additional measurements provided by the RNC are the core ion temperature and the fuel ratio profiles.
<figure><img src="image_48.png" /><figcaption>**Figure 36.** Layout of Radial Neutron Camera. The two interfacing diagnostics (Radial Gamma Ray Spectrometer (RGRS) and High Resolution Neutron Spectrometer (HRNS)) are also indicated. In-port RNC LOS are depicted in yellow, ex-port RNC LOS in grey, RGRS LOS in blue and HRNS LOS in pink.</figcaption></figure>

### 3.4.2. Design.
The RNC is located in ITER Equatorial Port #01 (EP#01) and is composed of two collimating structures, namely the In-port RNC system and the Ex-port RNC system viewing the plasma radially through apertures in the diagnostic shielding module (DSM) of the port plug (figure 36).

RNC measures the uncollided 14 MeV and 2.5 MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions through an array of neutron flux detectors located in collimated LOS, 6 for the In-Port and 16 for the Ex-Port. Figure 37 shows the LOS through the plasma and their geometrical details.

A design process model based on system engineering has been adopted, leading to a qualitative and quantitative methodology for the identification of the RNC baseline architecture [170].

### 3.4.3. Current status.
The preliminary and final design reviews (PDR, FDR) for the In-port RNC and the PDR for the ex-Port RNC have been successfully completed. On-going work is related to the preparation of the FDR for the ex-port RNC which is planned for mid-2025.

All main interfaces have been identified [171]. An extensive campaign of R&D and testing of prototyped RNC components is also on-going, focusing on neutron detectors (radiation hardness [172–174], technical specifications [175, 176]), electrical feedthroughs, front-end-electronics (electro-magnetic compatibility (EMC)) and assessment of real-time emissivity reconstruction capability.

MCNP Monte Carlo code analysis is extensively used [177] in order to enhance the diagnostic measurement performance and to evaluate the nuclear loads on the structural elements, detectors and associated components.

A calibration procedure has been defined that relies on embedded sources, reference ITER pulses and cross-calibration with ITER fission chambers (FCs) and NASs coupled to Monte Carlo simulations of radiation transport [178].

<figure><img src="image_49.png" /></figure>

<table><thead><tr><th>LOS #</th><th>Location</th><th>LOS angle* (°)</th><th>Collimator diameter (mm)</th><th>Collimator length (m)</th><th>r/a **</th></tr></thead><tbody><tr><td>1</td><td>In-port</td><td>50.0</td><td>32</td><td>944</td><td>0.85</td></tr><tr><td>2</td><td>In-port</td><td>56.0</td><td>32</td><td>1294</td><td>0.76</td></tr><tr><td>3</td><td>In-port</td><td>62.5</td><td>32</td><td>1463</td><td>0.67</td></tr><tr><td>4</td><td>Ex-port</td><td>73.9</td><td>11</td><td>2600</td><td>0.5</td></tr><tr><td>5</td><td>Ex-port</td><td>75.5</td><td>11</td><td>3000</td><td>0.45</td></tr><tr><td>6</td><td>Ex-port</td><td>77.5</td><td>11</td><td>3000</td><td>0.39</td></tr><tr><td>7</td><td>Ex-port</td><td>78.9</td><td>11</td><td>3000</td><td>0.34</td></tr><tr><td>8</td><td>shared RNC and RGRS</td><td>80.7</td><td>11</td><td>3000</td><td>0.28</td></tr><tr><td>9</td><td>Ex-port</td><td>82.3</td><td>11</td><td>3000</td><td>0.23</td></tr><tr><td>10</td><td>Ex-port</td><td>84.1</td><td>11</td><td>3000</td><td>0.17</td></tr><tr><td>11</td><td>exclusive use RGRS</td><td>85.7</td><td>—</td><td>—</td><td>0.12</td></tr><tr><td>12</td><td>Ex-port</td><td>87.5</td><td>11</td><td>3000</td><td>0.07</td></tr><tr><td>13</td><td>exclusive use HRNS</td><td>89.1</td><td>—</td><td>—</td><td>0.02</td></tr><tr><td>14</td><td>Ex-port</td><td>90.9</td><td>11</td><td>3000</td><td>0.05</td></tr><tr><td>15</td><td>exclusive use RGRS</td><td>92.5</td><td>—</td><td>—</td><td>0.1</td></tr><tr><td>16</td><td>Ex-port</td><td>94.3</td><td>11</td><td>3000</td><td>0.16</td></tr><tr><td>17</td><td>exclusive use RGRS</td><td>95.9</td><td>—</td><td>—</td><td>0.21</td></tr><tr><td>18</td><td>Ex-port</td><td>97.7</td><td>11</td><td>3000</td><td>0.26</td></tr><tr><td>19</td><td>Ex-port</td><td>99.3</td><td>11</td><td>3000</td><td>0.31</td></tr><tr><td>20</td><td>Ex-port</td><td>101.1</td><td>11</td><td>3000</td><td>0.37</td></tr><tr><td>21</td><td>Ex-port</td><td>102.7</td><td>11</td><td>2300</td><td>0.42</td></tr><tr><td>22</td><td>Ex-port</td><td>104.5</td><td>11</td><td>1800</td><td>0.48</td></tr><tr><td>23</td><td>Ex-port</td><td>106.1</td><td>8</td><td>1400</td><td>0.54</td></tr><tr><td>24</td><td>In-port</td><td>117.5</td><td>32</td><td>1463</td><td>0.7</td></tr><tr><td>25</td><td>In-port</td><td>124.0</td><td>32</td><td>1294</td><td>0.8</td></tr><tr><td>26</td><td>In-port</td><td>130.0</td><td>32</td><td>944</td><td>0.88</td></tr></tbody></table>

*Clockwise from a vertical line.

**Calculated for Q×5 scenario

<table><caption>**Figure 37.** RNC LOS (top) and geometrical details (bottom).</caption><thead><tr><th rowspan="3"></th><th colspan="3">FOCUS COORDINATES<br>(mm)</th><th rowspan="3">Focus - front collimating unit distance<br>(mm)</th></tr><tr><th>x</th><th>y</th><th>z</th></tr><tr><th></th><th></th><th></th></tr></thead><tbody><tr><td>Ex-port</td><td>11048</td><td>-119</td><td>620</td><td>1500</td></tr><tr><td>In-port Upper</td><td>8410</td><td>446</td><td>-132</td><td>1041</td></tr><tr><td>In-port Lower</td><td>8410</td><td>406</td><td>1372</td><td>1041</td></tr></tbody></table>

### 3.4.4. Measurement requirements.
ITER FP MPs are detailed in table 12 (section 3.2). RNC is designed to measure the fusion power density in the range $1 \text{ kW m}^{-3} - 4 \text{ MW m}^{-3}$ with 10% accuracy. The neutron and alpha source profile shall be measured in the range $10^{14} - 1.4 \cdot 10^{18} \text{ s}^{-1} \text{m}^{-3}$ with 10% accuracy. Both measurements should be performed with time resolutions of 1 ms (off-line) and 10 ms (real-time) and a spatial resolution of $a/10$ ($a$ = tokamak minor radius).

### 3.4.5. Subsystem designs.

#### 3.4.5.1. In-port RNC.
The In-port RNC is located in Drawer #3 of the Port Plug of EP#01 and is devoted to probe the edge of the plasma. The detectors are located in a Cassette
<figure><img src="image_50.png" /><figcaption>**Figure 38.** In-port RNC cassette assembly.</figcaption></figure>

<table><caption>**Table 15.** Characteristics of in-port RNC 238 U fission chamber.</caption><thead><tr><th colspan="2"><sup>238</sup>U fission chamber</th></tr></thead><tbody><tr><td>Diameter sensitive area (mm)</td><td>32</td></tr><tr><td><sup>238</sup>U total mass (mg)</td><td>145</td></tr><tr><td><sup>238</sup>U coating thickness (mm)</td><td>0.45 mm</td></tr><tr><td><sup>238</sup>U coating purity (%)</td><td>Better than 9.99 ± 0.01</td></tr><tr><td>Efficiency (14 MeV) (counts/neutron)</td><td>5.4 × 10<sup>-5</sup></td></tr></tbody></table>

Assembly (3 detector modules in the Upper part of the Cassette and 3 detector modules in the Lower part of the cassette, figure 38).

The LOS traverse the DSM and reach the FW. The In-port RNC Embarked Units are:

* EU02 (DSM 3 optical path): shaped conical voids within the DSM and the DFW corresponding to each LOS.

* EU03 (cassette assembly): sealed container connected to the secondary vacuum (service vacuum system) enclosing the detector modules, the beam dumps & shielding material.

* EU04 (collimator tubes): there is one tube for each LOS and the whole system is connected to the DSM and to the cassette assembly.

The in-port RNC has two sets of dedicated vacuum and electrical feedthroughs, respectively on the cassette assembly and on the closure plate. The detector module associated to each LOS contains a FC and a single crystal diamond (sCD) matrix. The main features of these neutron detectors are summarized in tables 15 and 16.

The two detectors are mounted as shown in figure 38.

**3.4.5.2. Ex-port RNC.** The ex-port RNC is devoted to probe the core of the plasma. Detectors and collimators are distributed on two planes hosted in a massive shielded structure (Shielding Block) located in the port interspace (10 LOS on the left hand side looking at the plasma and 6 LOS on the right hand side). Penetrations in the DSM allow the LOS to reach

<table><caption>**Table 16.** Characteristics of In-Port RNC sCD matrix.</caption><thead><tr><th colspan="2">sCD matrix</th></tr></thead><tbody><tr><td>Diameter sensitive area (mm)</td><td>32</td></tr><tr><td>Form</td><td>≈30 pixels, 4.5 × 4.5 mm<br>arranged in two groups</td></tr><tr><td>Thickness (μm)</td><td>50</td></tr><tr><td>Efficiency (14 MeV)<br>(counts/neutron)</td><td>5.4 × 10<sup>-5</sup></td></tr><tr><td>Embedded alpha calibration<br>source (one per matrix)</td><td>Mixed <sup>244</sup>Cm/<sup>239</sup>Pu/<sup>241</sup>Am<br>~6 kBq</td></tr></tbody></table>

the DFW. The thickness of the closure plate at the level of the ex-port RNC LOS is reduced (22 mm) in order to minimize the interactions with neutrons; the region of the closure plate with reduced thickness is defined as ‘closure plate customization’.

The ex-port RNC has strong physical interfaces with two diagnostics: the HRNS and the RGRS. The ex-port RNC embarked units are:

* EU01 (DSM 2 Optical Path): shaped conical voids within DSM Drawer#2 and DFW corresponding to Export LOS.

* EU11 (IS Equipment): massive Shielding Block structure located on EP01 Interspace Supporting Structure (ISS) and hosting the ex-port RNC flight tubes,

Collimation units, detector modules, and beam dumps. The detector module associated to each LOS contains:

* a <sup>4</sup>He scintillator (internal pressure = 100 bar, length = 30 mm, diameter 44 mm) as main detector for full power DT measurements;

* a sCD matrix (same as for the in-port: 4 pixels, 4 × 4 × 0.1 mm) as backup detector for full power DT measurements;

* a EJ-276 G plastic scintillator (15 mm thick × 15 mm diameter) for low power DT measurements.

PMTs resistant to intense magnetic fields (fine mesh type) are selected for the two scintillators and are located outside the collimating neutron channel in order to avoid direct neutron flux. In this position, PMTs with silica windows withstand the environmental conditions (>200 kGy and >10<sup>14</sup> n cm<sup>-2</sup>).

The sCD is the first detector along the LOS, followed by the <sup>4</sup>He scintillator and by the plastic scintillator that is the last one, in order to minimize effects of scattering on the other detectors due to its constituent hydrogenous materials.

The Shielding Block (figure 39) is a sandwich structure containing and protecting the ex-port neutron diagnostics frame and is designed to minimize the background due to scattered neutrons at the position of RNC detectors. It is based on a self-supporting stainless steel structure for the collimators and panels of a castable reactor shielding material made of a borated hydrogenated mix (low density: 1.68 g cm<sup>-3</sup>, boron content: 5%).
<figure><img src="image_51.png" /><figcaption>**Figure 39.** Sketch of the ex-port shielding block. Reproduced from [179]. CC BY 4.0.</figcaption></figure>

#### 3.4.5.3. System performance analysis.
The performance of the diagnostic system has been assessed through 1D and 2D neutron emissivity reconstruction analysis through: (a) a 1D reconstruction code (Measurement Simulation Software Tool (MSST) using constant emissivity on magnetic flux surfaces as from IO input emissivity profiles) based on Tikhonov regularization with first derivative objective functional [180, 181] and (b) a tomography code based on the Minimum Fisher Regularization (MFR) method that can include the total neutron yield value provided by an independent diagnostic as additional constraint in the tomography procedure [182].

The neutron fluxes at the detector positions as well as the expected uncollided neutron count rates (above selected detectors energy thresholds) used as input for the analysis are shown in figure 40. The two codes were used to reconstruct the DT $Q = 10$ scenario ($\text{yield} = 1.8 \times 10^{20} \text{ n s}^{-1}$) with 10 ms time resolution: 100 synthetic datasets were used in the MFR case and 1000 in the MSST case. The analysis was carried out considering only background and statistical errors and assuming a random Gaussian noise with 10% amplitude as error for total neutron yield.

Figure 41 shows the results of the performance analysis, in terms of emissivity reconstruction and relative accuracy, obtained using the MFR method. The improvement in the RNC 2D reconstruction of the neutron emissivity obtained by using the total neutron yield [183] is also evident: the extension of the spatial region in which the accuracy of the reconstruction is better than 10 %, the better reconstruction of peaked emissivity profiles and the robustness against measurements noise and LOS data loss, picture from [179].

The consistency of the two different analysis methods is shown by figure 42 that compares the accuracy provided by MFR and MSST.

<figure><img src="image_52.png" /><figcaption>**Figure 40.** Neutron fluxes at detector position and count rates. Material attenuation along LoS is taken into account.</figcaption></figure>

<figure><img src="image_53.png" /><figcaption>**Figure 41.** MFR reconstruction results with and without inclusion of total neutron yield constraint. Top: 2D maps of reconstructed neutron emissivity (average value). Center: 2D maps of reconstruction accuracy. Reproduced from [179]. CC BY 4.0.</figcaption></figure>

Table 17 summarizes the ITER scenarios used in the RNC performance analysis simulations.

The associated achievable performances are provided in table 18: the profile regions (in terms of r/a range and
<figure><img src="image_54.png" /><figcaption>**Figure 42.** 1D profiles of average reconstructed emissivity (top) and of reconstruction accuracy with 1% error (bottom); the accuracy from the MSST reconstruction is also shown for comparison.</figcaption></figure>

associated emissivity) in which such performances are achiev-
able in real-time simultaneously are also indicated.

The table is based on MSST 1D simulations that include direct errors (Poisson statistical error, background due to scattered neutrons, random noise on acquired detector pulses, attenuation, detector efficiency, etendue) and indirect errors (equilibrium reconstruction). An implementation in real-time of the MSST reconstruction code demonstrates that the acquisition and processing phases leading to neutron emissivity reconstruction can be carried out respectively in 0.2 ms and 1.5 ms (total time ≪10 ms) [184–187].

Note that the time resolution requirement for measurements
provided not in real-time (1 ms) is anyway achievable thanks
to the use of digital acquisition systems (see e.g. [188]).

**3.5. Vertical neutron camera (VNC)**

**3.5.1. Introduction.** The VNC (55.B2) diagnostic is a system of fan-shape multichannel collimators and neutron detectors. VNC together with 55B1 RNC provides time-resolved measurements of Fusion power density profile and neutron & α source profile. So, the system is primary of measurement of parameters (MPs) 13 and 14 of ITER measurement 07 [189], also it has supplementary or backup role for another 5 ITER MPs.

The system will be functional only for the last phase (DT),
but deliveries for earlier stages are needed for integration com-
ponents into port integrator(s) structures.

The recent publications on the system include [190–192].

The VNC is composed of two fan-shaped collimating struc-
tures: upper (UVNC, 55.B2.B0)—in upper port 18 (UP18)—
predominantly observing plasma area near inductor, and lower
(LVNC, 55.B2.A0)—in lower port 14 (LP14)—mainly col-
lecting signal from R > R₀ region. Totally plasma is observed
vertically by 11 LOS (figure 43).

In scenarios/approach of the symmetry of the neutron emis-
sion on the magnetic surfaces having two chambers increases
reliability and availability of the measurements. However also
two working in parallel subsystems allows getting data on the
symmetry breaking, which can, for example, be crucial for fol-
lowing dynamics of the trapped vs confined fractions of the
fast ions. Finally, application of two systems 55.B1 and 55.B2
can further be considered for both improve of robustness and
for getting advanced details on the emission symmetry break-
ing. Minimizing assumptions on symmetry and providing so
more general tomographic reconstruction data to NFMs allows
to ensure having more precise data on primary measurements
on total fusion power and neutron production.

**3.5.2. Design.** Multiple solutions are shared between LVNC and UVNC. First of all, it includes I&C, so that one dedicated FDR will cover relevant aspect of both subsystems. Further, for cooling and efficient backing of the in-vessel components IBED water circuits are implemented. As soon as FCs have internal volume with 3 Bar pressure, following ITER vacuum requirements it has to be enveloped with guiding volume connected to SVS (service vacuum system). Then detector units share similar configuration with 2 FCs, 2 SCD and also ¹³⁷Cs + ²⁴¹Am sources.

However, LVNC and UVNC features multiple crucial dif-
ferences and sub sections below will more concentrate on
respected individual aspects of designs.

**3.5.2.1. Lower VNC.** On the figure 44 it is presented cross section for the key relevant to LVNC operation elements inside VV.

Conical LoS employed for LVNC. They starts in detector
module in the diagnostic rack (bottom), coming through diver-
tor cassette (DC; blue areas in the middle), have focal point in
the cut outs of the FW/blankets modules. This required par-
ticular design of the DC #40 and mentioned of blanket shield
block and FW (SB&FW #18C14). Aside of collimating holes
in DC #40, also it features absence of the pump slot—in order
to reduce neutron background for the detectors.

In order to ensure required accuracy of measurements non-
obstructed direct view on plasma is needed with well defined
geometry of all elements forming LoS. So dedicated tolerances
studies were initiated. First part included analysis of the tol-
erances values itself. Exact position of the blankets as well
as divertor cassettes relative to VV will be optimized based
on the actual magnetic configuration of the ITER with corre-
ponding shift of up to 20 mm in each of vertical and radial dir-
ections. However all the elements on the chain (also including
diagnostic rack) is conceived to have the same move keeping
thus geometry of the LoS.
<table><caption>**Table 17.** ITER scenarios used in the RNC performance analysis.</caption><thead><tr><th>Code</th><th><i>I</i><sub>p</sub> (MA)</th><th><i>B</i> (T)</th><th>Fuel</th><th>Fusion power (MW)</th><th>Peak neutron emissivity (s<sup>-1</sup> m<sup>-3</sup>)</th><th>Neutron yield (s<sup>-1</sup>)</th></tr></thead><tbody><tr><td>DT <i>Q</i> = 10</td><td>15</td><td>5.3</td><td>DT</td><td>500</td><td>5.7 &times; 10<sup>17</sup></td><td>1.8 &times; 10<sup>20</sup></td></tr><tr><td>DT <i>Q</i> = 5</td><td>15</td><td>5.3</td><td>DT</td><td>255</td><td>3.7 &times; 10<sup>17</sup></td><td>9.8 &times; 10<sup>19</sup></td></tr><tr><td>DT low</td><td>7.5</td><td>2.65</td><td>DT</td><td>23</td><td>4.9 &times; 10<sup>16</sup></td><td>8.4 &times; 10<sup>18</sup></td></tr><tr><td>DD high</td><td>7.5</td><td>5.3</td><td>DD</td><td>1</td><td>3.1 &times; 10<sup>15</sup></td><td>1 &times; 10<sup>18</sup></td></tr></tbody></table>

<table><caption>**Table 18.** Summary of RNC performance for neutron emissivity profile measurements (real-time).</caption><thead><tr><th>Scenario #</th><th>Neutron emissivity range (s<sup>-1</sup>m<sup>-3</sup>)</th><th>Time resolution (ms)</th><th>Accuracy</th><th>Spatial resolution</th><th>r/a</th></tr></thead><tbody><tr><td>DT <i>Q</i> = 10</td><td>1.4 &times; 10<sup>17</sup>–5.7 &times; 10<sup>17</sup></td><td>10</td><td>10%</td><td>a/10</td><td>&lt;0.8</td></tr><tr><td>DT <i>Q</i> = 5</td><td>1.6 &times; 10<sup>17</sup>–3.3 &times; 10<sup>17</sup></td><td>10</td><td>10%</td><td>a/10</td><td>0.25–0.6</td></tr><tr><td>DT low</td><td>3.7 &times; 10<sup>15</sup>–4 &times; 10<sup>16</sup></td><td>100</td><td>10%</td><td>a/10</td><td>0.2–0.8</td></tr><tr><td>DD high</td><td>1.5 &times; 10<sup>15</sup>–3.8 &times; 10<sup>15</sup></td><td>10</td><td>20%</td><td>a/10</td><td>&lt;0.6</td></tr></tbody></table>

<figure><img src="image_55.png" /><figcaption>**Figure 43.** Schematic view show VNC LoS, magnetic field contours. LVNC and UVNC subsystems are rotated to the one Z-X plane for representation purpose.</figcaption></figure>

<figure><img src="image_56.png" /><figcaption>**Figure 44.** LoS arrangement for LVNC. LVNC module roughly outlined with yellow rectangle to separate from Rack area.</figcaption></figure>

After dedicated investigation the misalignments of targets expected based on as-build preliminary analysis was found 0–4 mm in toroidal direction and 0–5 mm in poloidal and radial directions. Second part of the study based on MCNP simulations confirmed that measurements inaccuracy due to geometry uncertainties (also considering other contributors) can be kept within necessary 10% level.

Finally, one more key effect considered is pre compression of the divertor cassettes installation. Indeed, in configuration management models (CMMs), the LoS of the LVNC diagnostic system and the collimating holes in the LVNC CB are aligned in context (while DC can be seen as clashing with

its fixation elements—to describe pre compression). Reverse engineering action was undertaken to consider deformation of the LVNC CB in the area under the proper set of loads, basing on structural analysis to include dead weight, pre-compression, thermal expansion in operation condition, effect of coolant pressure and irradiation effect. At the same time following latest studies collimating holes diameter was increase from 30 to 40 mm in this area—which further improved robustness of solution, accords to recommendations of the manufacturability investigations and was found to further improve signal/noise ratio.

In the figure 45 is given detector module. Its upper part serves as an alignment frame (which ensures mechanical fixation in the rack).

It allows to shift the whole assembly for ±5 mm each direction and rotate for 5± mrad relatively to each coordinate