Beryllium limiter marker tiles were exposed to plasma in the Joint European Torus to diagnose the erosion of main chamber wall materials. A limiter marker tile consists of a beryllium coating layer (7-9 mu m) on the top of bulk beryllium, with a nickel interlayer (2-3 mu m) between them. The thickness variation of the beryllium coating layer, after exposure to plasma, could indicate the erosion measured by ion beam analysis with backscattering spectrometry. However, interpretations from broad beam backscattering spectra were limited by the non-uniform surface structures. Therefore, micro-ion beam analysis (mu-IBA) with 3 MeV proton beam for Elastic back scattering spectrometry (EBS) and PIXE was used to scan samples. The spot size was in the range of 3-10 mu m. Scanned areas were analysed with scanning electron microscopy (SEM) as well. Combining results from mu-IBA and SEM, we obtained local spectra from carefully chosen areas on which the surface structures were relatively uniform. Local spectra suggested that the scanned area (approximate to 600 mu m x 1200 mu m) contained regions with serious erosion with only 2-3 mu m coating beryllium left, regions with intact marker tile, and droplets with 90% beryllium. The nonuniform erosion, droplets mainly formed by beryllium, and the possible mixture of beryllium and nickel were the major reasons that confused interpretation from broad beam EBS.

The toroidal field (TF) coil system is one of the most mechanically stressed systems in a tokamak. Structural integrity of the system must be maintained on the global and on the local scale, where the stress state in each conductor jacket as well as in the insulation is to be within structural allowable limits. Solving this task head-on leads to very high computational demands. In this work a methodology for a finite element (FE) detailed analysis of a graded TF coil winding pack (WP) for EU DEMO is presented. A graded WP design option was analyzed in detail in critical locations of the coil, which were identified by an approximate, but very fast approach developed earlier. A comparison with the said approach was made to verify its applicability and limitations. The presented procedure proved to be accurate requiring only moderate computational efforts. The parametric approach allows fast implementation of design changes into the model. The procedure can be used to speed up the assessment of future design iterations of the DEMO TF coil system.

The toroidal field (TF) coils of a tokamak are subjected to very high operational electromagnetic loads. The shape of a TF coil plays an important role for structural performance since an arbitrary shape is subjected to considerable bending. To mitigate this issue, the TF coils were designed with special "bending free" (constant tension) shapes derived analytically using a thin filament approximation. However, this approximation ignores the stiffnesses of the TF coil case, the central support, and the intercoil structures. This issue was pointed out by other authors and is studied in this paper using FEA. Based on the 2015 EU DEMO baseline design, a finite element model of a constant tension "Princeton D" TF coil was developed. Electromagnetic and structural modeling of the initial and the developed designs were conducted and the in-plane bending in the windings was analyzed. It was shown that a coil case can significantly change the bending moment distribution in the winding pack and even introduce additional bending to a "bending free" shape. On the other hand, such a shape imposes additional constraints on the magnet system design making the usage of these shapes questionable. An alternative strategy of TF coil case reinforcement is discussed.

Forschungszentrum Jülich and partners are developing the ITER core charge exchange spectroscopy (cCXRS) hosted by upper port plug#3 (UPP). The system should withstand loading due to the plasma transients resulting in electromagnetic (EM) forces acting on the UPP and its onboard components. On the other hand, the eddy currents and shock dynamic EM loads occur in the vacuum vessel (VV) and make it vibrate. As a result, it excites the UPP and, consequently, its onboard components. The computational problem is a superposition of these solutions: 1) the deterministically calculated time-history UPP response to the applied EM forces and 2) the UPP response to the VV excitations that are specified as floor response spectra (FRS) at the port stub (attachment to the VV). On the top of this, the EM loads should be combined with the seismic ones which are also specified as an FRS. This paper considers a potential methodology for combining plasma transients with seismic events. A numerical modeling of the UPP hosting a specific cCXRS component is presented. This methodology gives a direct and transparent engineering way to estimate the mechanical strength of the UPP onboard components. The analysis uses port stub FRS input and does not depend on a spectra-to-spectra recalculation procedure (from VV to component attachment) that is well established for the seismic-type response spectra but is not validated for the FRS due to the plasma transients.

In the framework of the DEMOnstration fusion power plant (DEMO) design coordinated by the EUROfusion consortium, a pre-conceptual design of the superconducting magnet system has been developed. For the toroidal field coils (TFCs), three winding pack (WP) options have been proposed; exploring different winding approaches (pancakes vs. layers), and manufacturing techniques (react & wind vs. wind & react Nb3Sn). Thermal-hydraulic and mechanical analyses on the three WPs have produced encouraging results, with some critical issues to be solved in future studies and optimizations. The experimental tests on TF prototype short sample conductors have demonstrated a limited performance degradation with electro-magnetic cycles and significantly lower effective strains than most of the large-size Nb3Sn conductors reported in literature. The toroidal field quench protection circuit has been studied, starting from different topologies and focusing on the most promising one. Two designs are also presented for the central solenoid magnet, with preliminary evaluations on the AC losses during the plasma breakdown. Finally, the design of a TF winding pack based on HTS conductors and the experimental tests on "fusion-relevant" HTS cables are illustrated.

The assessment of the Shutdown Dose Rate (SDR) due to neutron activation is a major safety issue for fusion devices and in the last decade several benchmark experiments have been conducted at JET during Deuterium-Deuterium experiments for the validation of the numerical tools used in ITER nuclear analyses. The future Deuterium-Tritium campaign at JET (DTE2) will provide a unique opportunity to validate the codes under ITER-relevant conditions through the comparison between numerical predictions and measured quantities (C/E). For this purpose, a novel SDR experiment, described in the present work, is in preparation in the frame of the WPJET3-NEXP subproject within EUROfusion Consortium. The experimental setup has been accurately designed to reduce measurement uncertainties; spherical air-vented ionization chambers (ICs) will be used for on-line ex-vessel decay gamma dose measurements during JET shutdown following DT operations and activation foils have been selected for measuring the neutron fluence near ICs during operations. Active dosimeters (based on ICs) have been calibrated over a broad energy range (from about 30 keV to 1.3 MeV) with X and gamma reference beam qualities. Neutron irradiation tests confirmed the capability of active dosimeters of performing on-line decay gamma dose rate measurements, to follow gamma dose decay at the end of neutron irradiation as well as insignificant activation of the ICs. (C) 2017 The Authors. Published by Elsevier B.V.

A set of quartz crystal microbalances (QMB) was used at JET with full carbon wall to monitor mass erosion/deposition rates in the remote areas of the divertor. After introduction of the ITER- like wall (ILW) in JET with beryllium main wall and tungsten divertor, strong reduction of the material deposition and accompanied fuel retention was observed. Therefore the existing QMB electronics have been modified to improve the accuracy of frequency measurements by a factor of ten down to 0.1 Hz which corresponds to 1.4 ng cm−2. The averaged deposition rates of 1.2–3 ng cm−2 s−1 and erosion rates of 5.6–8.1 ng cm−2 s−1 were observed in the inner divertor of JET -ILW with the inner strike point positions close to the bottom edge of vertical tile 3 and at the horizontal tile 4 respectively. The erosion with averaged rates of ≈2.1 ng cm−2 s−1 and ≈120 ng cm−2 s−1 were observed in the outer divertor for the outer strike point positions at tile 5 and tile 6 respectively.

After completing the main construction phase of Wendelstein 7-X (W7-X) and successfully commissioning the device, first plasma operation started at the end of 2015. Integral commissioning of plasma start-up and operation using electron cyclotron resonance heating (ECRH) and an extensive set of plasma diagnostics have been completed, allowing initial physics studies during the first operational campaign. Both in helium and hydrogen, plasma breakdown was easily achieved. Gaining experience with plasma vessel conditioning, discharge lengths could be extended gradually. Eventually, discharges lasted up to 6 s, reaching an injected energy of 4 MJ, which is twice the limit originally agreed for the limiter configuration employed during the first operational campaign. At power levels of 4 MW central electron densities reached 3 x 10(19) m(-3), central electron temperatures reached values of 7 keV and ion temperatures reached just above 2 keV. Important physics studies during this first operational phase include a first assessment of power balance and energy confinement, ECRH power deposition experiments, 2nd harmonic O-mode ECRH using multi-pass absorption, and current drive experiments using electron cyclotron current drive. As in many plasma discharges the electron temperature exceeds the ion temperature significantly, these plasmas are governed by core electron root confinement showing a strong positive electric field in the plasma centre.

The mechanical concept of the cCXRS frictionless shutter is based on elastic lateral bending of its arms caused by pneumatic actuation and is limited by special bumpers. The shutter blocks or opens the path of light emitted by the plasma, which is under analysis, to cover the diagnostic's first mirror. A thorough understanding of the shutter's dynamic behaviour is indispensable to adjust the mechanical structure for the final design. Particular attention is paid to the arm rebounds that are dependent on the arms impact kinetic energy and the arms' preloading on the bumpers. To optimise the shutter structure the simplified parametric mock-up was developed which allows variation of the arms' stiffness, the mass, the arms preloading as well as the time evolution of the actuator pressure. The experimental results discussed in the paper are in accordance with the analytical and numerical predictions that proved to be a very effectiye tool for further design needs. (C) 2017 Elsevier B.V. All rights reserved.

In this work we present a method to evaluate the neutron response function of an NE213 liquid scintillator. This method is particularly useful when the proton light yield function of the detector has not been measured, since it is based on a proton light yield function taken from literature, MCNPX simulations, measurements of gammarays from a calibration source and measurements of neutrons from fusion experiments with ohmic plasmas. The inclusion of the latter improves the description of the proton light yield function in the energy range of interest (around 2.46 MeV).We apply this method to an NE213 detector installed at JET, inside the radiation shielding of the magnetic proton recoil (MPRu) spectrometer, and present the results from the calibration along with some examples of application of the response function to perform neutron emission spectroscopy (NES) of fusion plasmas.We also investigate how the choice of the proton light yield function affects the NES analysis, finding that the result does not change significantly. This points to the fact that the method for the evaluation of the neutron response function is robust and gives reliable results. (C) 2017 Published by Elsevier B.V.

Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power) if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW) have reached a maximum radiative fraction (Frad) of ∼75%. Divertor Langmuir probe (LP) measurements in these discharges indicate, however, that less than ∼3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high Frad. In H-mode nitrogen-seeded discharges with Frad increasing from ∼30% to up to ∼75%, the main chamber wall particle fluence rises by a factor ∼3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at Frad ∼ 60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the fraction of the input power transported to the main chamber wall remains below ∼5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached conditions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem.

A novel technique for sawteeth control in tokamak plasmas using ion-cyclotron resonance heating (ICRH) has been developed in the JET-ILW tokamak. Unlike previous ICRH methods, that explored the destabilization of the internal kink mode when the radio-frequency (RF) wave absorption was placed near the q = 1 surface, the technique presented here consists of stabilizing the sawteeth as fast as possible by applying the ICRH power centrally and subsequently induce a sawtooth crash by switching it off at the appropriate instant. The validation of this method in JET-ILW L-mode discharges, including preliminary tests in H-mode plasmas, is presented.

Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components) of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane. (C) 2017 The Authors. Published by Elsevier Ltd.

Beryllium film deposition was studied with cavity samples in remote areas of the inner and outer JET divertor and below divertor tile 5 during the 2011-2012 campaign with the ITER-like wall. Predominantly beryllium films were formed inside the cavities with some additional carbon, the ratio Be/C was > 2. These deposited layers had high D/(Be+C) ratios of about 0.3. The formation of these films is mainly due to sticking of beryllium-containing particles with low sticking coefficients < 0.5. The observed surface loss probabilities depend on the position in the divertor. The particles responsible for film deposition originated from the location of in the divertor strike points. (C) 2016 Elsevier Ltd.

Robust power exhaust schemes employing impurity seeding are needed for target operational scenarios in present day tokamak devices with metallic plasma-facing components (PFCs). For an electricity-producing fusion power plant at power density Psep/R>15MW/m divertor detachment is a requirement for heat load mitigation. 2D plasma edge transport codes like the SOLPS code as well as plasma-wall interaction (PWI) codes are key to disentangle relevant physical processes in power and particle exhaust. With increased quantitative credibility in such codes more realistic and physically sound estimates of the life-time expectations and performance of metallic PFCs can be accomplished for divertor conditions relevant for ITER and DEMO. An overview is given on the recent progress of plasma edge and PWI modelling activities for (carbon-free) metallic devices, that include results from JET with the ITER-like wall, ASDEX Upgrade and Alcator C-mod. It is observed that metallic devices offer an opportunity to progress the understanding of underlying plasma physics processes in the edge. The validation of models can be substantially improved by eliminating carbon from the experiment as well as from the numerical system with reduced degrees of freedom as no chemical sputtering from amorphous carbon layers and no carbon or hydro-carbon transport are present. With the absence of carbon as the primary plasma impurity and given the fact that the physics of the PWI at metallic walls is less complex it is possible to isolate the crucial plasma physics processes relevant for particle and power exhaust. For a reliable 2D dissipative plasma exhaust model these are: cross-field drifts, complete kinetic neutral physics, geometry effects (including main-chamber, divertor and sub-divertor structures), SOL transport reflecting also the non-diffusive nature of anomalous transport, as well as transport within the pedestal region in case of significant edge impurity radiation affecting pedestal pressure and hence Psep.

The Neutron Camera (NC) is a Joint European Torus (ET) diagnostic, based on a set of 19 collimated lines of sight equipped with plastic (BC418) and liquid (NE213) scintillators, with the main function of measuring the neutron emissivity profile due to 2.5 MeV (DD) and 14 MeV (DT). Due to several limitations of the present data acquisition system and in view of the JET DT campaign, an enhancement project (Neutron Camera Upgrade, NCU) was launched. The main objective was to improve the measurement capability of NC for 14 MeV neutrons, by application of a high throughput FPGA-based digital acquisition system. The present paper describes the hardware architecture and the FPGA processing selected for the NCU project and the first tests carried out at JET. (C) 2017 Published by Elsevier B.V.

The structural integrity of superconducting magnets that are key elements of a fusion reactor must be ensured. At an early design stage relatively simple calculation tools can greatly facilitate design optimization. The main objective of this paper is the mechanical pre-dimensioning of the tokamak toroidal field coils by simple means prior to the global 3D numerical modeling. A semi-analytical calculation tool that reasonably estimates the static strength of the toroidal field coil under the electromagnetic forces at the critical location (inner leg equatorial plane) is described. The novelty of the approach is that it treats not only the massive coil casing but also the winding pack conductor jacket under an essentially 3D stress state. The calculation tool features pre-optimization of the coil winding for graded layered winding layouts. The minimum space (radial built) required for the coil inboard portion that is a key design parameter is defined after possible winding pre-optimization. The procedure has been successfully benchmarked against numerical solutions and has been used for pre-dimensioning the toroidal coils in the frame of the current 2015 DEMO activity. (C) 2017 The Author(s). Published by Elsevier B.V.

One of the important techniques used at JET for studying fast ions is based on measurements of gamma rays which are produced as a result of nuclear reactions between ions and plasma impurities. The intense neutron and gamma-ray fluxes expected during a DT campaign impose dew requirements on detector characteristics used in such experiments. In addition to good energy resolution, detectors must also be characterized by a high signal-to-noise ratio and allow to perform measurements at high counting rate about 1 Mcps. The scintillators which fulfill these requirements are, among others, LaBr3:Ce, already tested at JET, and CeBr3 with a scintillation decay time of similar to 20 ns. We report on measurements which were performed with a detector module equipped with a 3" x 3" CeBr3 scintillator and with an active voltage divider AVD@NCBJ, designed and constructed at NCBJ. Standard gamma -ray sources, as well as a PuBe source, were used for measurements. The comparison of measured and Monte Carlo simulated spectra is also presented. (C) 2017 Elsevier B.V. All rights reserved.

A series of neon seeded JET ELMy H-mode pulses is considered from the modeling as well as from the experimental point of view. For two different Ne seeding rates and two different D puffing gas levels the heating power, Pheat, is in the range 22–29.5 MW. The main focus is on the numerical reconstruction of the total radiated power (which mostly depends on the W concentration) and its distribution between core and divertor and of Zeff (which mostly depends on the Ne concentration). To model self-consistently the core and the SOL two input parameters had to be adjusted case by case: the SOL diffusivity, DSOL, and the core impurity inward pinch, vpinch. DSOL had to be increased with increasing ΓNe and the level of vpinch had to be changed, for any given ΓNe, according to the level of Pheat: it decreases with increasing Pheat. Since the ELM frequency, fELM, is experimentally correlated with Pheat, (it increases with Pheat) the impurity inward pinch can be seen as to depend on fELM. Therefore, to maintain a low vpinch level (i.e. high fELM) ΓNe/Pheat should not exceed a certain threshold, which slightly increases with the ΓD puffing rate. This might lead to a limitation in the viability of reducing the target heat load by Ne seeding at moderate ΓD, while keeping Zeff at acceptably low level.

A strong effect of divertor configuration on the threshold power for the L-H transition (PLH) was observed in recent JET experiments in the new ITER-like Wall (ILW) [1], [2], [3]. Following a series of EDGE2D-EIRENE code simulations with Be impurity and drifts a possible mechanism for the PLH variation with the divertor geometry is proposed. Both experiment and code simulations show that in the configuration with lower neutral recycling near the outer strike point (OSP), electron temperature (Te) peaks near the OSP prior to the L-H transition, while in the configuration with higher OSP recycling Te peaks further out in the scrape-off layer (SOL) and the plasma stays in the L-mode at the same input power. Code results show large positive radial electric field (Er) in the near SOL under lower recycling conditions leading to a large E × B shear across the separatrix which may trigger earlier (at lower input power) edge turbulence suppression and lower PLH. Suppressed Te's at OSP in configurations with strike points on vertical targets (VT) were observed earlier and explained by a geometrical effect of neutral recycling near this particular position, whereas in configurations with strike points on horizontal targets (HT) the OSP appears to be more open for neutrals (see e.g. review paper [4]).