An overview is presented of the progress since 2021 in the construction and scientific programme preparation of the Divertor Tokamak Test (DTT) facility. Licensing for building construction has been granted at the end of 2021. Licensing for Cat. A radiologic source has been also granted in 2022. The construction of the toroidal field magnet system is progressing. The prototype of the 170 GHz gyrotron has been produced and it is now under test on the FALCON facility. The design of the vacuum vessel, the poloidal field coils and the civil infrastructures has been completed. The shape of the first DTT divertor has been agreed with EUROfusion to test different plasma and exhaust scenarios: single null, double null, X-divertor and negative triangularity plasmas. A detailed research plan is being elaborated with the involvement of the EUROfusion laboratories.

The fusion reaction between deuterium and tritium, D(T,n)4 He is the main source of energy in future thermonuclear reactors. Alpha-particles (4 He-ions) born with an average energy of 3.5 MeV transferring energy to the thermal plasma during their slowing down, should provide the self-sustained D-T plasma burn. The adequate confinement of alpha-particles is essential to provide efficient heating of the bulk plasma and steady burning of a reactor plasma. That is why the fusion-born alpha-particle studies have been a priority task in the second D-T experiments (DTE2) on the Joint European Torus (JET) to understand the main mechanisms of their slowing down, redistribution and losses and to develop optimal plasma scenarios. JET with Be-wall and W-divertor, enhanced auxiliary heating systems and improved energetic-particle diagnostic capabilities, producing significant population of alpha-particles, provided the possibility for comprehensive studying of the alpha-particle behaviour. Selected results of the confined and lost alpha-particle measurements, evidence of alpha-particle self-heating and assessments of the fusion performance are presented in this paper giving an opportunity for further modelling and extrapolation to the International Thermonuclear Experimental Reactor and burning plasma reactors.

Within the 9th European Framework programme, since 2021 EUROfusion is operating five tokamaks under the auspices of a single Task Force called ‘Tokamak Exploitation’. The goal is to benefit from the complementary capabilities of each machine in a coordinated way and help in developing a scientific output scalable to future largre machines. The programme of this Task Force ensures that ASDEX Upgrade, MAST-U, TCV, WEST and JET (since 2022) work together to achieve the objectives of Missions 1 and 2 of the EUROfusion Roadmap: i) demonstrate plasma scenarios that increase the success margin of ITER and satisfy the requirements of DEMO and, ii) demonstrate an integrated approach that can handle the large power leaving ITER and DEMO plasmas. The Tokamak Exploitation task force has therefore organized experiments on these two missions with the goal to strengthen the physics and operational basis for the ITER baseline scenario and for exploiting the recent plasma exhaust enhancements in all four devices (PEX: Plasma EXhaust) for exploring the solution for handling heat and particle exhaust in ITER and develop the conceptual solutions for DEMO. The ITER Baseline scenario has been developed in a similar way in ASDEX Upgrade, TCV and JET. Key risks for ITER such as disruptions and run-aways have been also investigated in TCV, ASDEX Upgrade and JET. Experiments have explored successfully different divertor configurations (standard, super-X, snowflakes) in MAST-U and TCV and studied tungsten melting in WEST and ASDEX Upgrade. The input from the smaller devices to JET has also been proven successful to set-up novel control schemes on disruption avoidance and detachment.

In the frame of the design activities of the Divertor Tokamak Test (DTT) facility the development of a comprehensive set of neutron and gamma-ray diagnostics is on-going in order to enable measurements of: neutron yield, neutron yield rate, neutron emissivity over a poloidal section through the plasma; neutron emission spectrum; runaway electrons induced bremsstrahlung radiation and gamma-ray emission from reactions between fast ions and plasma impurities. The present paper provides an overview of the DTT neutron and gamma-ray diagnostics and describes the present status of their design including main components and interfaces, detector types and performances.

The Radial Neutron Camera (RNC) is an ITER diagnostic system devoted to the radial measurement of the plasma neutron emissivity during ITER operation. In particular, plasma core measurements will be performed using 48 detectors located in the Ex-Port RNC subsystem and viewing the plasma through 16 lines-of-sight (LOS). Since discrepancies concerning the position of the LOS and the related optical path can occur during RNC installation and ITER operation, a Position Monitoring System (PMS) is foreseen to evaluate the misalignment and provide a correction tool for the neutrons count rate. During the design of the PMS for the RNC, optic fiber-based displacement sensors have been identified as the best solution to measure the displacements from reference positions along the radial, vertical and toroidal directions. Potential issues that could arise during the installation and operation of such sensors have been evaluated; MCNP calculations have been performed to determine the radiation level at the position of the PMS components. Finally, mechanical tests have been carried out at ENEA to evaluate the performance of the sensors: results indicate compliance with the specifications provided by the producer and the ITER RNC design requirements.

The Radial Neutron Camera (RNC) is a diagnostic system located in ITER Equatorial Port #1 providing several spatial and time-resolved parameters for the fusion power estimation, plasma control and physics studies. The 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 Lines of Sight. Signals from RNC detectors (fission chambers, single Crystal Diamonds and scintillators) need preamplification because of their low amplitude. These preamplifiers have to be as close as possible to the detectors in order to minimize signal degradation and must be protected against fast and thermal neutrons, gamma radiation and electromagnetic fields. The solution adopted is to host the preamplifiers in a shielded cabinet located in a dedicated area of the Port Cell, behind the Bioshield Plug. The overall design of the cabinet must ensure the necessary magnetic, thermal and nuclear shielding and, at the same, satisfy weight and allocated volume constraints and maintain its structural integrity. The present paper describes the nuclear design of the shielded cabinet, performed by means of 3D particle transport calculations (MCNP), taking into account the radiation streaming through the Bioshield penetrations and the cross-talk effect from the neighboring Lower and Upper Ports. We present the assessment of its nuclear shielding performances and analyze the compliancy with the alert thresholds for commercial electronics in terms of neutron flux and cumulated ionizing dose.

The paper presents an overview of the design status of the Radial Neutron Camera (RNC), that, together with the Vertical Neutron Camera, will provide, through reconstruction techniques applied to the measured line-integrated neutron fluxes, the time resolved measurement of the ITER neutron and α-source profile (i.e. neutron emissivity, neutrons emitted per unit time and volume). The RNC is composed of two subsystems, the In-Port RNC and Ex-Port RNC located, respectively, inside and outside the Plug of Equatorial Port #01. The In-Port subsystem is in a more advanced design stage since it has recently undergone the Final Design Review in the ITER procurement process. The paper describes the diagnostic layout, the interfaces, the measurement capabilities and the main challenges in its realization. Prototyping and testing of neutron detectors and electronics components were carried out and led to the choice of the component solutions that can match the environmental and operational constraints in terms radiation hardness, high temperature and electromagnetic compatibility. The performance of the RNC in terms of neutron emissivity measurement capability was assessed through 1D and 2D reconstruction analysis. It is proven that the neutron emissivity can be reconstructed in real-time within the measurement requirements: 10% accuracy, 10 ms time resolution and a/10 (a = plasma minor radius) space resolution.

Since the 2018 IAEA FEC Conference, FTU operations have been devoted to several experiments covering a large range of topics, from the investigation of the behaviour of a liquid tin limiter to the runaway electrons mitigation and control and to the stabilization of tearing modes by electron cyclotron heating and by pellet injection. Other experiments have involved the spectroscopy of heavy metal ions, the electron density peaking in helium doped plasmas, the electron cyclotron assisted start-up and the electron temperature measurements in high temperature plasmas. The effectiveness of the laser induced breakdown spectroscopy system has been demonstrated and the new capabilities of the runaway electron imaging spectrometry system for in-flight runaways studies have been explored. Finally, a high resolution saddle coil array for MHD analysis and UV and SXR diamond detectors have been successfully tested on different plasma scenarios.

This paper studies the Electromagnetic Compatibility (EMC) compliance of a diagnostic system intended for the nuclear fusion reactor ITER. Specifically, a neutron detection chain is analyzed, where the electrical signals emitted by a diamond matrix hit by neutrons. A non-conventional design of such an electronic chain is requested, to meet the specific requirements within the particular constraints imposed for such an application. This results in specific EMC issue related to the effect of common mode noise on the signal integrity. The EMC analysis is carried out on a mock-up system, and solutions to face the above issues are designed, implemented and experimentally validated.

Single Crystal Diamonds (sCD) have been selected as neutron detectors for the ITER Radial Neutron Camera (RNC). This paper investigates experimentally the hardness of these detectors to 14 MeV neutrons in order to have indications of their limits of applicability in the RNC. The experiment has been performed at the Frascati Neutron Generator (FNG) facility using three sCDs of different thickness (500, 300,100 mu m). The degradation of the sCD Charge Collection Efficiency (CCE) with increasing 14 MeV neutron fluence has been found to be lower with decreasing sCD thickness. Strong polarization effects produced by the trapping of charge carriers in the defects produced by the neutron irradiation are observed already at about 4. 10(14) n/cm(2) in the 100 mu m thick sCD. A software correction of the polarization effect is proposed, which enables to extend the sCD range of operation up to higher fluences. Taking into account such correction, a 50 mu m thick sCD is expected to reach 10(15) n/cm(2) with a CCE greater than 0.7.

The ITER Radial Neutron Camera is a diagnostic whose objective is measuring neutron emissivity and fusion power density through an array of detectors placed in collimating structures. The RNC is composed of two subsystems (In-Port RNC and Ex-Port RNC), located in the Equatorial Port 01 of the ITER tokamak. Although the measurements from the RNC are required for ITER D-T phase, its In-Port components must be ready for Assembly phase 2. Consequently, the two subsystems will be delivered at different times. At the current status of the design the In-Port RNC interfaces must be defined, if not fully specified, in order to allow for the subsystem integration in the Port Plug. A thorough assessment of the interfaces of the subsystem with all the diagnostics, plants and services in the port has been made, taking into account the concurrent development of the Equatorial Port 01 and the progress in the design of some of the subsystem components that may affect the identification of interfacing Plant Systems. This paper deals with the process that led to definition of the internal and external interfaces of the In-Port RNC, highlighting the main issues and the solutions adopted to perform integration within the Equatorial Port Plug 01.

Single Crystal Diamond (sCD) are candidate neutron detectors for the ITER Radial Neutron Camera (RNC) due to their capability of performing both neutron counting and spectroscopy while withstanding high levels of radiation. This paper presents a study on methods of separation between neutron and gamma events, through the analysis of the pulse shapes of a 500μm thick sCD. Methods for gamma rejection may be beneficial especially during ITER deuterium–deuterium operation to extract the 2.5 MeV neutron signal. Four neutron/gamma separation algorithms for sCD neutron detectors are presented. Three algorithms rely on pulses shape information and one on the frequency analysis. The final experimental implementation based on multi-step separation approach achieved above 90% of correctly discriminated pulses.

Transport modelling of Joint European Torus (JET) dimensionless collisionality scaling experiments in various operational scenarios is presented. Interpretative simulations at a fixed radial position are combined with predictive JETTO simulations of temperatures and densities, using the TGLF transport model. The model includes electromagnetic effects and collisions as well as E ⃗ × B ⃗ shear in Miller geometry. Focus is on particle transport and the role of the neutral beam injection (NBI) particle source for the density peaking. The experimental 3-point collisionality scans include L-mode, and H-mode (D and H and higher beta D plasma) plasmas in a total of 12 discharges. Experimental results presented in (Tala et al 2017 44th EPS Conf. ) indicate that for the H-mode scans, the NBI particle source plays an important role for the density peaking, whereas for the L-mode scan, the influence of the particle source is small. In general, both the interpretative and predictive transport simulations support the experimental conclusions on the role of the NBI particle source for the 12 JET discharges.

Fusion performance in tokamaks hinges critically on the efficacy of the edge transport barrier (ETB) in suppressing energy losses. The new concept of 'fingerprints' is introduced to identify the instabilities that cause transport losses in the ETBs of many of today's experiments, from among widely posited candidates. Analysis of the gyrokinetic-Maxwell equations and gyrokinetic simulations of experiments reveals that each mode type produces characteristic ratios of transport in the various channels: density, heat, and impurities. This, together with experimental observations of transport in some channel or of the relative size of the driving sources of channels, can identify or determine the dominant modes causing energy transport. In multiple H-mode cases with edge-localized modes that are examined, these fingerprints indicate that magnetohydrodynamic (MHD)-like modes are apparently not the dominant agent of energy transport; rather, this role is played by micro-tearing modes (MTMs) and electron temperature gradient (ETG) modes, and in addition, possibly by ion temperature gradient/ trapped electron modes (ITG/TEM) on JET (Joint European 'Torus). MHD-like modes may dominate the electron particle losses. Fluctuation frequency can also be an important means of identification, and is often closely related to the transport fingerprint. The analytical arguments unify and explain previously disparate experimental observations on multiple devices, including DIII-D, JET, and ASDEX-U. Detailed simulations of two DIII-D ETBs also demonstrate and corroborate this.

In this paper, we analyze heat transport in the JET tokamak using data from its high resolution ECE diagnostic and analyses based on the transfer entropy (TE). The analysis reveals that heat transport is not smooth and continuous, but is characterized by 'trapping regions' separated by `minor transport barriers'. Meat may 'jump over' these barriers and when the heating power is raised, this 'jumping' behavior becomes more prominent. To check that our results are relevant for global heat transport, we deduced an effective diffusion coefficient from the TE results. Both its value and overall radial variation are consistent with heat diffusivities reported in literature. The detailed radial structure of the effective diffusion coefficient was shown to be linked to the mentioned minor transport barriers.

The ITER Radial Neutron Camera (RNC) Data Acquisition (DAQ) prototype is based on the PCIe protocol as the interface to be used between the I/O unit and the host computer, allowing the scalability of the final RNC DAQ system. The prototype architecture comprises two digitizer modules with two channels in each, installed in the host computer and the maximum produced data throughput is up to 1.6 GB/s per board, allowing a sustainable 2 MHz peak event to cope with the long plasma discharges, up to half an hour. The Linux Device Driver provides the interface between the hardware and the host applications running on a high-performance computer, which receives the acquired data through the Direct Memory Access (DMA) channels. The preliminary tests show that the Linux kernel miss some hardware interrupts when the time between interrupts is in the microsecond scale, which implies data loss if a traditional read technique based on interrupt handling is implemented. The direct usage of the polling mechanisms to retrieve the data is not suitable as the variable event rate over the same discharge does not allow to define an optimal fixed-time to retrieve data and identify missing data packets. This contribution presents the architecture, implementation and test of a different device driver approach using the polling mechanism which improves the performance and reliability. This allows the device driver to automatically check and transfer missing data blocks, recovering the data losses transparently for the host applications. The performance results for tests with different event data rates and duration up to an hour are also presented.

Results from two different sets of JET experiments are presented.In experiments in which toroidicity-induced Alfvén eigenmodes (TAEs) localized at different radial locations had the same frequencies and toroidal mode numbers, the occurrence of enhanced losses after the excitation of TAEs in the core of the plasma was observed.On the contrary, enhanced losses were not observed if the TAEs localized at different radial locations had different frequencies and toroidal mode numbers.Numerical modeling indicates that, in the first set of experiments, the enhanced losses were caused by a combined effect of the TAEs localized at different radial locations.The TAEs localized in the plasma core transported energetic ions from the core to outer regions of the plasma.Then, the TAEs localized in outer regions of the plasma interacted with these ions just transported by the core-localized TAEs causing a further radial displacement of the ions to the plasma edge.This process eventually ends up causing the loss of the resonant ions.In the second set of experiments, it was found that TAEs localized in the plasma core and in outer regions did not interact with the same ions and so no enhanced losses were measured.Sheared profiles of the safety factor combined with flat mass density profiles lead to larger differences on the frequencies of the TAEs localized at different radial locations, eventually avoiding loss of energetic ions through the described mechanism.

In this study, beryllium tiles from Joint European Torus (JET) vacuum vessel wall were analysed and compared regarding their position in the vacuum vessel and differences in the exploitation conditions during two campaigns of ITER-Like-Wall (ILW) in 2011-2012 (ILW1) and 2013-2014 (ILW2) Tritium content in beryllium samples were assessed. Two methods were used to measure tritium content in the samples - dissolution under controlled conditions and tritium thermal desorption. Prior to desorption and dissolution experiments, scanning electron microscopy and energy dispersive x-ray spectroscopy were used to study structure and chemical composition of plasma-facing-surfaces of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2.10(11)-2.10(13) tritium atoms per square centimetre of the surface area with its highest content in the samples from the outer wall of the vacuum vessel (up to 1.9.10(13) atoms/cm(2) in ILW1 campaign and 2.4.10(13) atoms/cm(2) in ILW2). The lowest content of tritium was found in the upper part of the vacuum vessel (2.0.10(12) atoms/cm(2) and 2.0.10(11) atoms/cm(2) in ILW1 and ILW2, respectively). Results obtained from scanning electron microscopy has shown that surface morphology is different within single tile, however if to compare two campaigns main tendencies remains similar.

The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project.

The aim of this paper is to present a signal processing algorithm that, applied to the raw Locked Mode signal, allows us to obtain a disruption indicator in principle exploitable on different tokamaks. A common definition of such an indicator for different machines would facilitate the development of portable systems for disruption prediction, which is becoming of increasingly importance for the next tokamak generations. Moreover, the indicator allows us to overcome some intrinsic problems in the diagnostic system such as drift and offset. The behavior of the proposed indicator as disruption predictor, based on crossing optimized thresholds of the signal amplitude, has been analyzed using data of both JET and ASDEX Upgrade experiments. A thorough analysis of the disruption prediction performance shows how the indicator is able to recover some missed and tardy detections of the raw signal. Moreover, it intervenes and corrects premature or even wrong alarms due to, e.g., drifts and/or offsets.