Titanium activation foils for neutron studies were exposed inside the JET tokamak vessel during the entire 15 months long experimental campaign with deuterium fuelling. Afterwards, it was found that the structure of some foils was only slightly affected, while others were totally or partly disintegrated into dust: flakes and powder. The analyses indicated that defects produced in materials by cutting during the sample preparation eventually lead to the degradation of the titanium crystal structure upon the exposure in JET to neutral particles including hydrogen. In some cases, a passive oxide layer on the foil surface effectively prevented the access of hydrogen, and by this protected the metal against partial or complete transformation into the titanium hydride powder.

The EUROfusion Safety and Environmental Work Package (WPSAE) has the scope to progress the safety studies for the future EU DEMO reactor. A Generic Site Safety Report (GSSR) has been prepared to include all the steps necessary to cover the safety issues of the nuclear fusion plant, from the definition of principles and requirements, through the detection of the source terms at risk, selecting the postulated initiating events, analyzing the accidents, quantifying the doses to the population and investigating waste production and its management. Eleven GSSR volumes collate the studies performed. The final goal is to prepare safety documentation, as complete as possible, to initiate a preliminary safety report when the plant site is selected.In parallel, the safety studies are supporting the maturation of the design of DEMO, providing feedback on the technical choices for the machine, the selection of materials, the use of space, the equipment necessary to correctly manage the safety risks through continuous collaboration with the design teams of other DEMO Work Packages. The GSSR focused on safety and environmental issues, it did not encompass security issues that could impact the design and hence the safety performance of DEMO. Consideration of the interactions between safety and security will be needed in the development of the DEMO detailed design and its associate Preliminary Safety Analysis Report (PSAR).The main achievements of the Pre-Concept Phase are presented in this paper, by summarizing the contents of GSSR volumes. Completion of the work is foreseen during the future Concept Design Phase (2021-2027).

A model for simulating charge exchange (CX) of fast ions with background atoms in magnetically confined fusion plasmas has been implemented in the ASCOT orbit-following code. The model was verified by comparing simulated reaction mean free paths to analytical values across a range of fusion-relevant parameters. ASCOT was used to simulate beam ions slowing down in the presence of CX reactions in a MAST-U target scenario. ASCOT predicts the CX-induced loss of beam power to be 22% , which agrees to within 15% with the TRANSP prediction. Due to CX, plasma heating and current drive by beam ions are strongly reduced towards the edge. However, an overall lower but noticeable increase of up to 20% in current drive is predicted closer to the core. The simulated deposition of fast CX atoms on the wall is concentrated around the outer midplane, with estimated peak power loads of 70–80 kW m−2 on the central poloidal field coils (P5) and the vacuum vessel wall between them. This analysis demonstrates that ASCOT can be used to simulate fast ions in fusion plasmas where CX reactions play a significant role, e.g. in spherical tokamaks and stellarators.

A design integration study (KDII4) was conducted in the DEMO Pre-Concept Design Phase with the primary goal to develop a pre-concept feasibility design and concept of operation for the Vertical Blanket Segment Architecture. The primary goal was to develop two workable variants for removal of large in-vessel components. This, prompted by two earlier studies, highlighting the integration challenge finding a self-consistent DEMO design point (KDII) and secondly the complexity of operation required to ensure compatibly with the proposed port-based maintenance schemes. Initially, only a single null (SN) divertor configuration was considered, but due to additional identified issues and technical challenges centred on the extraction of the BB segments, a double null (DN) alternative variant has also been investigated. This was prompted by other KDII's studies, especially KDII1 (Design, performance and feasibility of wall protection limiters during plasma transients) and KDII3 (Advanced Magnetic Configurations). Alternative forms of vertical maintenance architecture have been investigated by breaking down the study into Ports, In-vessel Components, Operations and Safety. SN and DN configurations were studied with split and full blankets numbering seven variants in total. Each variant was studied in some detail, comparatively assessing each on its merits. However, no viable solution has been found for down-selection. Nevertheless, the question of vertical maintenance is now better understood. This paper will describe the findings of KDII4 in conjunction with the Remote Maintenance (RM) technology work package, which suggests that the current ‘reference’ design envelope available for RM is too constrained.

DEMO is a key part of the EU fusion roadmap, and the programme reaches the end of the pre-conceptual phase with a gate review in 2020. As part of the work to complete this phase, eight Key Design Integration Issues (KDII's) have been identified as critical to the programme. Two of these KDII's identified a requirement for a more detailed architectural study. Within KDII#3 (advanced divertor configurations) a double null configuration has been developed; in parallel KDII#4 (vertical segment architecture) identified a need to evaluate a split breeding blanket architecture. These two requirements were combined into a single study that assesses the invessel architecture for a double null DEMO featuring split breeding blankets. This paper presents the configurations developed and an evaluation of their feasibility with respect to integration of key in-vessel components (breeding blankets, divertors), port hardware (service pipes, shielding, vacuum pumping), and maintenance strategy. Furthermore, wider considerations such as the impact of the port orientation on the building architecture are also included. Finally, this paper will identify the risks and further work required to advance the double null configuration.

Analyses of radiation fields resulting from a deuterium-tritium (DT) plasma in fusion devices is a critical input to the design and validation of many aspects of the reactor design, including, shielding, material lifetime and remote maintenance requirements/scheduling. Neutronics studies, which perform in-depth analysis are typically performed using radiation transport codes such as MCNP, TRIPOLI, Serpent, FLUKA and OpenMC. The Serpent 2 Monte-Carlo code, developed by VTT in Finland, is the focus of this work which seeks to benchmark the code for fusion applications. The application of Serpent 2 in fusion specific analysis requires validation of the codes performance in an energy range, and a geometrical description, which significantly differs to conventional nuclear fission analysis, for which the code was originally developed. A Serpent model of the Frascati Neutron Generator (FNG) Helium Cooled Pebble Bed (HCPB) mock up experiment has been prepared and the calculated results compared against experimental data, as well as the reference Monte Carlo code MCNP. The analysis is extended to a model of DEMO with HCPB blanket concept. For this model, the flux, nuclear heating, tritium production and DPA are calculated, all of which are integral nuclear responses in fusion reactor analysis. In general, a very good agreement is demonstrated for both of the benchmarks, with any discrepancies pinpointed to different physics models implemented.

Spectrum unfolding is a key tool used together with diagnostics in the determination of nuclear fields that are associated with a range of nuclear technologies spanning fusion, fission, nuclear medicine and accelerator domains. The underlying process requires a mathematical method for solving the Fredholm integral equation of the first kind. This paper discusses the development, testing and comparison of the modern combined framework of methods for performing neutron spectrum unfolding SPECTRA-UF, which includes the UF_G and UF_M subroutines, based on the underlying mathematics of the GRAVEL and MAXED methods respectively, along with a custom parameterised subroutine, UF_P. We compared the behaviour of each method using a set of synthetic data. We discuss the challenges associated with unfolding fusion spectra, and the behaviour of each subroutine along with the feasibility of using general parameterised spectra as initial a priori spectra. The UF_M, UF_G and UF_P methods showed reasonable agreement where good a priori was supplied and all improved on the a priori spectrum given, but behaved poorly where less accurate a priori was provided, with UF_G showing itself to rely more heavily on the a priori spectrum given. The UF_M subroutine performed most favourably, producing the lowest mean fractional deviation across the majority of spectra. The UF_P was able to represent the fusion peaks and relatively smooth epi-thermal regions, but performed less well where the flux spanned many orders of magnitude. The modelling of the down-scatter component of the fusion peaks was also challenging to reliably model using simple distributions.

The way to arrive at a licensing phase for a nuclear fusion installation is not straightforward mainly because of the lack of operating experience and of dedicated nuclear regulations. In fact, only small/medium experimental facilities exist with limited licensing processes and only one large experiment, ITER, has obtained a construction license. Therefore, the safety assessment and the preparation of the preliminary safety report is almost a first of a kind for DEMO. Taking advantage of the fission power plants experience and considering to the maximum extent the ITER safety studies, the preparation of a Generic Site Safety Report (GSSR) has begun. It will require some years to be completed; however currently, at the starting point, the strategy to develop it is clear and well defined. This paper considers all the safety issues that will be included in the GSSR because they have been clarified in the frame of the European Workprogramme for DEMO from 2014 up to 2018, and should not be modified in the future, such as the safety requirements for the plant and the systems, the tools to be used for the safety assessment, the procedures for the selection of the reference accidents, and so on. Together with these topics, considered as goals achieved, there are others for which an additional effort is necessary because they do not cover all the expected requirements of the likely licensing procedures applicable for DEMO. A complete spectrum of Design Basis Accidents and Beyond Design Basis Accidents that can determine the risk of releases from the main systems of the power plant is still incomplete together with the safety classification of most of the Structures, Systems and Components, and the feasibility and analyses of some accident mitigation systems. The outcome of this study is the quantification, when possible, of the gap between the results achieved and the goals established in the power plant guidelines. It will help also to qualify the effort required in terms of studies, experiments and human resources to reach a good stage for successful DEMO licensing.

Shutdown dose rate calculations provide an essential input to the design and research of fusion power plant technology. They allow the estimation of dose to personnel and equipment during planned and unplanned maintenance. The mesh coupled rigorous 2 step (MCR2S) methodology used at Culham Centre for Fusion Energy (CCFE) was originally developed to link the MCNP particle transport code and the FISPACT-II inventory code. As new particle transport codes are developed there is a strong motivation to move towards a code agnostic approach. This paper details the integration of MCR2S with the FISPACT-II API and two other transport codes, Serpent 2 and OpenMC. Two benchmarks, the FNG shutdown dose rate experimental benchmark and the ITER computational benchmark, have been performed and compared to results produced with MCNP. In general, the results show that MCNP, Serpent 2 and OpenMC give shutdown dose rate results similar to the FNG experiment for both experimental campaigns. However, all codes appeared to slightly overestimate the dose rates for Campaign 1 (all results had a C/E between 1 and 1.5) and underestimate the dose rates for Campaign 2 (all results had a C/E between 0.6 and 1). Differences were seen between OpenMC and MCNP for the ITER port plug benchmark, where the lack of variance reduction in OpenMC meant that the neutron flux estimates at the rear of the model were not converged. This led to differences of up to 13% in the shutdown dose rates. It was shown that Serpent 2 and MCNP, where variance reduction was used, gave shutdown dose rates within 3% of each other. Although some areas for development of the Serpent 2 and OpenMC transport codes have been highlighted, overall the comparisons give confidence that the implementation of these two transport codes into the MCR2S work-flow has been carried out successfully.

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.

For future fusion plants, there is a great need for detectors capable to accurately monitor neutrons under the harsh fusion environment. In particular, detectors required in Test Blanket Modules must be capable to accurately measure neutron fluence under high and variable neutron fluxes, high gamma background, high temperature, high and variable magnetic fields. The noVEl neutRon Detector for fusion (VERDI) project aims to develop a detector which will provide a robust approach for neutron detection in fusion plants. The detector comprises a low activation capsule containing a defined concentration of added metallic elements. The neutron fluence and energy spectrum will be derived by the analysis of the multiple gamma lines produced by the metallic elements activation. In this work, the candidate metallic elements for the VERDI detector are defined. A benchmark experiment with a number of prototype detectors was performed at the ENEA Frascati Neutron Generator under DT operation to demonstrate the feasibility of VERDI detectors to measure neutron fluence under a reference fusion relevant field. The results of product isotopes measurements are compared to respective calculated data using the FISPACT-II code and very good agreement is observed. Preliminary unfolding of the neutron spectrum was also performed.

A novel multiphysics methodology for the computation of realistic plasma neutron sources has been developed. The method is based on state-of-the-art plasma transport and neutron spectrum calculations, coupled with a Monte Carlo neutron transport code, bridging the gap between plasma physics and neutronics. In the paper two JET neutronics tokamak models are used to demonstrate the application of the developed plasma neutron sources and validate them. Diagnostic data for the record JET D discharge 92436 are used as input for the TRANSP code, modelling neutron emission in two external plasma heating scenarios, namely using only neutral beam injection and a combination of the latter and ion cyclotron resonance heating. Neutron spectra, based on plasma transport results, are computed using the DRESS code. The developed PLANET code package is employed to generate plasma neutron source descriptions and couple them with the MCNP code. The effects of using the developed sources in neutron transport calculations on the response of JET neutron diagnostic systems is studied and compared to the results obtained with a generic plasma neutron source. It is shown that, although there are significant differences in the emissivity profiles, spectra shape and anisotropy between the neutron sources, the integral response of the time-resolved ex-vessel neutron detectors is largely insensitive to source changes, with major relative deviations of up to several percent. However it is calculated that, due to the broadening of neutron spectra as a consequence of external plasma heating, larger differences may occur in activation of materials which have threshold reactions located at DD neutron peak energies. The PLANET plasma neutron source computational methodology is demonstrated to be suitable for detailed neutron source effect studies on JET during DT experiments and can be applied to ITER analyses.

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed.

The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating He-3 in the early phase of the discharge; for the present design of DEMO-with a static magnetic field strength of B-o = 5.855 T-that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna-plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.

A synthetic diagnostic has been developed for the JET lost alpha scintillator probe, based on the ASCOT fast ion orbit following code and the AFSI fusion source code. The synthetic diagnostic models the velocity space distribution of lost fusion products in the scintillator probe. Validation with experimental measurements is presented, where the synthetic diagnostic is shown to predict the gyroradius and pitch angle of lost DD protons and tritons. Additionally, the synthetic diagnostic reproduces relative differences in total loss rates in multiple phases of the discharge, which can be used as a basis for total loss rate predictions.

Convolutional neural networks (CNNs) have found applications in many image processing tasks, such as feature extraction, image classification, and object recognition. It has also been shown that the inverse of CNNs, so-called deconvolutional neural networks, can be used for inverse problems such as plasma tomography. In essence, plasma tomography consists in reconstructing the 2D plasma profile on a poloidal cross-section of a fusion device, based on line-integrated measurements from multiple radiation detectors. Since the reconstruction process is computationally intensive, a deconvolutional neural network trained to produce the same results will yield a significant computational speedup, at the expense of a small error which can be assessed using different metrics. In this work, we discuss the design principles behind such networks, including the use of multiple layers, how they can be stacked, and how their dimensions can be tuned according to the number of detectors and the desired tomographic resolution for a given fusion device. We describe the application of such networks at JET and COMPASS, where at JET we use the bolometer system, and at COMPASS we use the soft X-ray diagnostic based on photodiode arrays.

Predictability of burning plasmas is a key issue for designing and building credible future fusion devices. In this context, an important effort of physics understanding and guidance is being carried out in parallel to JET experimental campaigns in H and D by performing analyses and modelling towards an improvement of the understanding of DT physics for the optimization of the JET-DT neutron yield and fusion born alpha particle physics. Extrapolations to JET-DT from recent experiments using the maximum power available have been performed including some of the most sophisticated codes and a broad selection of models. There is a general agreement that 11–15 MW of fusion power can be expected in DT for the hybrid and baseline scenarios. On the other hand, in high beta, torque and fast ion fraction conditions, isotope effects could be favourable leading to higher fusion yield. It is shown that alpha particles related physics, such as TAE destabilization or fusion power electron heating, could be studied in ITER relevant JET-DT plasmas.

In future fusion power plants, such as DEMO, the D-T neutron emission rate is predicted to exceed 1 x 10(21) n s(-1). Accurately monitoring neutron energies and intensities will be the primary method for estimating fusion power, and calculating key nuclear parameters, including the tritium breeding ratio and nuclear heating. The noVel nEutRon Detector for fusIon (VERDI) project, implemented under the EUROFusion Enabling Research 2017 program, aims to develop a detector capable of withstanding the harsh environment of a future fusion power plant. The VERDI detector is based on the foil activation technique, which relies on neutron spectrum unfolding methods to process the convolution of gamma-ray measurement and detector response function to infer the neutron energy spectrum. This paper details the experimental method and results collected using six prototype VERDI detectors during an initial experiment performed in July 2017 at the ENEA Frascati Neutron Generator (FNG) under D-T neutrons (14 MeV). The measured activities of product isotopes are compared with equivalent data calculated using the FISPACT-II code to provide an average C/E-act agreement of 1.05 +/- 0.13. Experimental results from the FNG have been applied to neutron spectrum unfolding techniques using established unfolding codes, MAXED and GRAVEL.

For many years, machine learning tools have proved to be very powerful disruption predictors in tokamaks.On the other hand, the vast majority of the techniques deployed assume that the input data is independent and is sampled from exactly the same probability distribution for the training set, the test set and the final real time deployment.This hypothesis is certainly not verified in practice, since the experimental programmes evolve quite rapidly, resulting typically in ageing of the predictors and consequent suboptimal performance.This paper describes various adaptive training strategies that have been tested to maintain the performance of disruption predictors in non-stationary conditions.The proposed approaches have been implemented using new ensembles of classifiers, explicitly developed for the present application.The improvements in performance are unquestionable and, given the difficulties encountered so far in translating predictors from one device to another, the proposed adaptive methods from scratch can therefore be considered a useful option in the arsenal of alternatives envisaged for the next generation of devices, particularly at the very beginning of their operation.