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Sludge generated in urban wastewater treatment plants is a huge environmental and economic challenge of sustainability. Sludge recovery is an environmental necessity and is a requirement for wastewater treatment plants. Predators have been proposed as a biological solution and simultaneously obtaining by-product of interest. Nematodes are predators that are found naturally in sludge but have never been proposed for sludge reduction. In this work, Caenorhabditis elegans, a nematode broadly used for biomedical studies, was used to grow on urban sludge, reaching a remarkable average population up to 92,668 nematodes/mL, 50.6% of sludge was converted into nematode biomass and up to 21.0% of organic matter removed. To demonstrate the viability of this nematode to grow and reduce urban sludges, different experiments in solid and liquid culture media were performed. The physio-chemical composition of the different types of sludges generated in the urban wastewater treatment plants was determined. The total lipids (4.52%, w/w) and fatty acid profiles obtained in the nematode biomass were determined. This work presents a new method to valorise urban sludge and opens the possibility to obtain by-products with high added value.

Before deuterium-tritium (DTE2), a new tritium campaign (TT) with the objective to validate the activities, dose rates for selected materials and OLTIS container components, as resulting from neutron irradiation at Joint European Torus (JET), is planned. This paper describes the activation and dose rate calculation performed for copper based materials irradiated throughout the foreseen TT plasma operation. Preparatory experiments and calculations are in progress for TT campaign to measure neutron flux spectrum at the irradiation position using dosimetry foils. In addition, the materials to be considered for irradiation during the TT campaign of the OLTIS were identified and analyzed in this study: Al-Bronze and CuCrZr. The neutron-induced activities and dose rates at shutdown were calculated by activation inventory code FISPACT-2010 using the neutron fluxes and spectra provided by the detailed MCNP5 neutron transport calculations. After the irradiation, the activities and dose rates are calculated at several cooling times after the end of irradiation. Highest dose rates the end of irradiation are attributed to Cu-64 radionuclide. Co-58 and Co-60 are the most dominant radionuclides during intermediate and long cooling periods. Highest activities are ascribed to Cu-64 at the beginning and Ni-63 at the end of investigation time.

Quantitative tritium inventory in dust particles from campaigns in the JET tokamak with the carbon wall (2007–2009) and the ITER-like wall (ILW 2011–2012) were determined by the liquid scintillation counter and the full combustion method. A feature of this full combustion method is that dust particles were covered by a tin (Sn) which reached 2100 K during combustion under oxygen flow. The specific tritium inventory for samples from JET with carbon and with metal walls was measured and found to be similar. However, the total tritium inventory in dust particles from the ILW experiment was significantly smaller in comparison to the carbon wall due to the lower amount of dust particles generated in the presence of metal walls.

More than 30% of the world's population are anemic with serious economic consequences including reduced work capacity and other obstacles to national welfare and development. Red blood cell transfusion is the mainstay to correct anemia, but it is also 1 of the top 5 overused procedures. Patient blood management (PBM) is a proactive, patient-centered, and multidisciplinary approach to manage anemia, optimize hemostasis, minimize iatrogenic blood loss, and harness tolerance to anemia. Although the World Health Organization has endorsed PBM in 2010, many hospitals still seek guidance with the implementation of PBM in clinical routine. Given the use of proven change management principles, we propose simple, cost-effective measures enabling any hospital to reduce both anemia and red blood cell transfusions in surgical and medical patients. This article provides comprehensive bundles of PBM components encompassing 107 different PBM measures, divided into 6 bundle blocks acting as a working template to develop institutions' individual PBM practices for hospitals beginning a program or trying to improve an already existing program. A stepwise selection of the most feasible measures will facilitate the implementation of PBM. In this manner, PBM represents a new quality and safety standard.

Thioredoxins (TRX) are traditionally considered as enzymes catalyzing redox reactions. However, redox-independent functions of thioredoxins have been described in different organisms, although the underlying molecular mechanisms are yet unknown. We report here the characterization of the first generated endogenous redox-inactive thioredoxin in an animal model, the TRX-1 in the nematode Caenorhabditis elegans. We find that TRX-1 dually regulates the formation of an endurance larval stage (dauer) by interacting with the insulin pathway in a redox-independent manner and the cGMP pathway in a redox-dependent manner. Moreover, the requirement of TRX-1 for the extended longevity of worms with compromised insulin signalling or under calorie restriction relies on TRX-1 redox activity. In contrast, the nuclear translocation of the SKN-1 transcription factor and increased LIPS-6 protein levels in the intestine upon trx-1 deficiency are strictly redox-independent. Finally, we identify a novel function of C. elegans TRX-1 in male food-leaving behaviour that is redox-dependent. Taken together, our results position C. elegans as an ideal model to gain mechanistic insight into the redox-independent functions of metazoan thioredoxins, overcoming the limitations imposed by the embryonic lethal phenotypes of thioredoxin mutants in higher organisms.

Selected set of samples from JET ITER-Like Wall (JET-ILW) divertor tiles exposed in 2015–2016 has been analysed using Thermal Desorption Spectrometry (TDS). The deuterium (D) amounts obtained with TDS were compared with Nuclear Reaction Analysis (NRA). The highest amount of D was found on the top part of inner divertor which has regions with the thickest deposited layers as for divertor tiles removed in 2014. This area resides deep in the scrape-off layer and plasma configurations for the second (ILW-2, 2013–2014) and the third (ILW-3, 2015–2016) JET-ILW campaigns were similar. Agreement between TDS and NRA is good on the apron of Tile 1 and on the upper vertical region whereas on the lower vertical region of Tile 1 the NRA results are clearly smaller than the TDS results. Inner divertor Tile 3 has somewhat less D than Tiles 0 and 1, and the D amount decreases towards the lower part of the tile. The D retention at the divertor inner and outer corner regions is not symmetric as there is more D retention poloidally at the inner than at the outer divertor corner. In most cases the TDS spectra for the ILW-3 samples are different from the corresponding ILW-2 spectra because HD and D2 release occurs at higher temperatures than from the ILW-2 samples indicating that the low energy traps have been emptied during the plasma operations and that D is either in the energetically deep traps or located deeper in the sample.

The impurity concentration in the tokamak divertor plasma is a necessary input for predictive scaling of divertor detachment, however direct measurements from existing tokamaks in different divertor plasma conditions are limited. To address this, we have applied a recently developed spectroscopic N II line ratio technique for measuring the N concentration in the divertor to a range of H-mode and L-mode plasma from the ASDEX Upgrade and JET tokamaks, respectively. The results from both devices show that as the power crossing the separatrix, P-sep, is increased under otherwise similar core conditions (e.g. density), a higher N concentration is required to achieve the same detachment state. For example, the N concentrations at the start of detachment increase from approximate to 2% to approximate to 9% as P-sep, is increased from approximate to 2.5 MW to approximate to 7 MW. These results tentatively agree with scaling law predictions (e.g. Goldston et al.) motivating a further study examining the parameters which affect the N concentration required to reach detachment. Finally, the N concentrations from spectroscopy and the ratio of D and N gas valve fluxes agree within experimental uncertainty only when the vessel surfaces are fully-loaded with N.

The erosion, transport and deposition of tungsten in the outer divertor of JET-ILW has been studied for an H-Mode discharge with low frequency ELMs. For this specific case with an inter-ELM electron temperature at the strike point of about 20 eV, tungsten sputtering between ELMs is almost exclusively due to beryllium impurity and self-sputtering. However, during ELMs tungsten sputtering due to deuterium becomes important and even dominates. The amount of simulated local deposition of tungsten relative to the amount of sputtered tungsten in between ELMs is very high and reaches values of 99% for an electron density of 5E13 cm(-3) at the strike point and electron temperatures between 10 and 30 eV. Smaller deposition values are simulated with reduced electron density. The direction of the B-field significantly influences the local deposition and leads to a reduction if the E x B drift directs towards the scrape-off-layer. Also, the thermal force can reduce the tungsten deposition, however, an ion temperature gradient of about 0.1 eV/mm or larger is needed for a significant effect. The tungsten deposition simulated during ELMs reaches values of about 98% assuming ELM parameters according to free-streaming model. The measured WI emission profiles in between and within ELMs have been reproduced by the simulation. The contribution to the overall net tungsten erosion during ELMs is about 5 times larger than the one in between ELMs for the studied case. However, this is due to the rather low electron temperature in between ELMs, which leads to deuterium impact energies below the sputtering threshold for tungsten.

Selected set of samples from JET TER-Like Wall (JET-ILW) divertor tiles exposed both in 20132014 and 20112014 has been analysed using Thermal Desorption Spectrometry (TDS). The deuterium (D) amounts ob tained with TIS were compared with ion Beam Analysis (TBA) and Secondary ion Mass Spectrometry (SIMS) data. The hig amount of D was found on the top part of inner divertor which has regions with the thickest deposited layers. This area resides deep in the scrape-off layer. Changes in plasma configurations between the first 20112012 and the second 20132014 JET-IL W campaign altered the material migration towards the inner and the outer divertor come increasing the amount of deposition in the shadowed areas of the divertor base tiles. Retention on the outer divertor tiles is clearly smaller than on the inner divertor tiles. Experimental TDS spectra for samples from the top part of inner divertor and from the outer strike point region were modelled using TMAP program. Experimental deuterium profiles obtained with SIMS have been used and the detrapping and the activation energies have been adjusted Analysis of the results of the TMAP simulations enabled to determine the nature of traps in different samples

Waves in the Ion Cyclotron Range of Frequencies (ICRF) enhance local Plasma-Surface Interactions (PSI) near the wave launchers and magnetically-connected objects via Radio-Frequency (RF) sheath rectification. ITER will use 20MW of ICRF power over long pulses, questioning the long-term impact of RF-enhanced localized erosion on the lifetime of its Beryllium (Be) wall. Recent dedicated ICRF-heated L-mode discharges documented this process on JET for different types of ICRF antennas. Using visible spectroscopy in JET ICRF-heated L-mode discharges, poloidally-localized regions of enhanced (by ∼2–4x) Be I and Be II light emission were observed on two outboard limiters magnetically connected to the bottom of the active ITER-Like Antenna (ILA). The observed RF-PSI induced by the ILA was qualitatively comparable to that induced by the JET standard, type-A2 antennas, for similar strap toroidal phasing and connection geometries. The Be II line emission was found more intense when powering the bottom half of the ILA rather than its top half. Conversely, more pronounced SOL density modifications were observed with only top array operation, on field lines connected to the top half of the ILA. So far the near-field modeling of the ILA with antenna code TOPICA (Torino Polytechnic Ion Cyclotron Antenna), using curved antenna model, was partially able to reproduce qualitatively the observed phenomena. A quantitative discrepancy persisted between the observed Be source amplification and the calculated, corresponding increases in E// field at the magnetically connected locations to the ILA when changing from only top to only bottom half antenna operation. This paper revisits these current drive phased and half-ILA powered cases using for the new simulations flat model of the ILA and more realistic antenna feeding to calculate the E// field maps with TOPICA code. Further, the Self-consistent Sheaths and Waves for Ion Cyclotron Heating Slow Wave (SSWICH-SW) code, which couples slow wave

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.

Tritium (T) distributions on tungsten (W)-coated plasma-facing tiles used in the third ITER-like wall campaign (2015-2016) of the Joint European Torus (JET) were examined by means of an imaging plate technique and beta-ray induced x-ray spectrometry, and they were compared with the distributions after the second (2013-2014) campaign. Strong enrichment of T in beryllium (Be) deposition layers was observed after the second campaign. In contrast, T distributions after the third campaign was more uniform though Be deposition layers were visually recognized. The one of the possible explanations is enhanced desorption of T from Be deposition layers due to higher tile temperatures caused by higher energy input in the third campaign.

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during (1) the third IIW campaign (11.W-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, Le LW1 to ILW-3: 20112016 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified Layer on surface has very small amount of impurities such as D, B, C N O and Ni. All mirrors from the divertor (inner, outer, base under the bulk ile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, CN, O, Ni and W. In the inner divertor N reaches 5 x 107 cm2 W is up to 4.3 x 10 cm while the content of Ni is the greatest in the outer divertor 3.8 x 10 cm Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10 cm . The thickness of deposited Layer is in the range or 90 nm to 900 nm. The layer growth rate in the base (2.7 pms !) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrots located at the cassette mouth, flaking of deposits and crosion occurred.

The recently developed Monte-Carlo code ERO2.0 is applied to the modelling of limited and diverted discharges at JET with the ITER-like wall (ILW). The global beryllium (Be) erosion and deposition is simulated and compared to experimental results from passive spectroscopy. For the limiter configuration, it is demonstrated that Be self-sputtering is an important contributor (at least 35%) to the Be erosion. Taking this contribution into account, the ERO2.0 modelling confirms previous evidence that high deuterium (D) surface concentrations of up to similar to 50% atomic fraction provide a reasonable estimate of Be erosion in plasma-wetted areas. For the divertor configuration, it is shown that drifts can have a high impact on the scrape-off layer plasma flows, which in turn affect global Be transport by entrainment and lead to increased migration into the inner divertor. The modelling of the effective erosion yield for different operational phases (ohmic, L- and H-mode) agrees with experimental values within a factor of two, and confirms that the effective erosion yield decreases with increasing heating power and confinement.

Effect of ELMs on fuel retention at the bulk W target of JET ITER-Like Wall was studied with multi-scale calculations. Plasma input parameters were taken from ELMy H-mode plasma experiment. The energetic intra-ELM fuel particles get implanted and create near-surface defects up to depths of few tens of nm, which act as the main fuel trapping sites during ELMs. Clustering of implantation-induced vacancies were found to take place. The incoming flux of inter-ELM plasma particles increases the different filling levels of trapped fuel in defects. The temperature increase of the W target during the pulse increases the fuel detrapping rate. The inter-ELM fuel particle flux refills the partially emptied trapping sites and fills new sites. This leads to a competing effect on the retention and release rates of the implanted particles. At high temperatures the main retention appeared in larger vacancy clusters due to increased clustering rate.

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 cam paigns of ITER-Like-Wall (LW) in 2011201211 WI) and 20132014 ILW 2) 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 surface of the beryllium samples. Experimental results revealed that tritium content in the samples is in range of 2101 210 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 1.910 atoms/cm in II.W1 campaign and 2410" atoms/cm in 11W2). The lowest content of tritium was found in the upper part of the vacuum vessel 2010 atoms/cm² and 2.010 atoms/cm in ILWI and IIW2, 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

Tungsten-coated divertor tiles exposed during the third JET ITER-Like Wall (ILW) campaign in 2015–2016 (ILW3) were studied with Secondary Ion Mass Spectrometry (SIMS). ILW-3 campaign contained more high-power plasma discharges and longer plasma time than the earlier ILW campaigns. Measurements showed increased beryllium (Be) deposition on the upper inner divertor, whereas on the outer divertor, Be deposition was lower than during the second campaign in 2013–2014 (ILW-2). Increased intensities of nickel, molybdenum and tungsten were observed at the surface layer of the inner divertor Be dominated deposits. These layers are probably formed during the high-power plasma discharge phase near the end of the ILW-3 campaign. Compared to the earlier campaigns, D retention on the upper inner divertor was observed to on a similar level than after ILW-2, whereas at the lower inner divertor and most parts of the outer divertor, D retention was lower for ILW-3 than ILW-2. D retention was increased at lower part of outer divertor Tile 7, where Be deposition was slightly increased. Probable reason for the reduction is the higher surface temperature of the tiles due to higher powers used.

ERO is a 3D Monte-Carlo impurity transport and plasma-surface interaction code. In 2011 it was applied for the ITER first wall (FW) life time predictions [1] (critical blanket module BM11). After that the same code was significantly improved during its application to existing fusion-relevant plasma devices: the tokamak JET equipped with an ITER-like wall and linear plasma device PISCES-B. This has allowed testing the sputtering data for beryllium (Be) and showing that the “ERO-min” fit based on the large (50%) deuterium (D) surface content is well suitable for plasma-wetted areas (D plasma). The improved procedure for calculating of the effective sputtering yields for each location along the plasma-facing surface using the recently developed semi-analytical sheath approach was validated. The re-evaluation of the effective yields for BM11 following the similar revisit of the JET data has indicated significant increase of erosion and motivated the current re-visit of ERO simulations.

Preliminary results investigating the microstructure, bonding and effect of beryllium oxide formation on retention in the JET ITER-like wall beryllium tiles, are presented. The tiles have been investigated by several techniques: Scanning Electron Microscopy (SEM) equipped with Energy Dispersive X-ray (EDX), Transmission Electron microscopy (TEM) equipped with EDX and Electron Energy Loss Spectroscopy (EELS), Raman Spectroscopy and Thermal Desorption Spectroscopy (TDS). This paper focuses on results from melted materials of the dump plate tiles in JET. From our results and the literature, it is concluded, beryllium can form micron deep oxide islands contrary to the nanometric oxides predicted under vacuum conditions. The deepest oxides analyzed were up to 2-micron thicknesses. The beryllium Deuteroxide (BeOxDy) bond was found with Raman Spectroscopy. Application of EELS confirmed the oxide presence and stoichiometry. Literature suggests these oxides form at temperatures greater than 700 °C where self-diffusion of beryllium ions through the surface oxide layer can occur. Further oxidation is made possible between oxygen plasma impurities and the beryllium ions now present at the wall surface. Under Ultra High Vacuum (UHV) nanometric Beryllium oxide layers are formed and passivate at room temperature. After continual cyclic heating (to the point of melt formation) in the presence of oxygen impurities from the plasma, oxide growth to the levels seen experimentally (approximately two microns) is proposed. This retention mechanism is not considered to contribute dramatically to overall retention in JET, due to low levels of melt formation. However, this mechanism, thought the result of operation environment and melt formation, could be of wider concern to ITER, dependent on wall temperatures.

Recent achievements in studies of the effects of ICRF (Ion Cyclotron Range of Frequencies) power on the SOL (Scrape-Off Layer) and PWI (Plasma WallInteractions) in ASDEX Upgrade (AUG),Alcator CMod,and JETIL Ware reviewed. Capabilities to diagnose and model the effect of DCbiasingand associated impurity production atactive antenna sandon magnetic field connections to antennas are described. The experiments show that ICRFnearfields can lead not only to E×B convection, but also to modifications of the SOL density,which for Alcator CMod are limited to anarrow regionne arantenna. On the other hand, the SOL density distribution along with impurity sources can be tail or edusing localg as injection in AUG and JET-ILW with a positive effect on reduction of impurity sources.The technique of RFimage current cancellation atantennalimiters was successfully applied in AUG using the 3-strap AUG antenna and extended to the 4-strap Alcator C-Mod field-aligned antenna.MultipleobservationsconfirmedthereductionoftheimpactofICRFontheSOLandontotalimpuritypro-ductionwhentheratioofthepowerofthecentralstrapstothetotalantennapowerisintherange0.6

The identification of the sources of atomic tungsten and the measurement of their radiation distribution in front of all plasma-facing components has been performed in JET with the help of two digital cameras with the same two-dimensional view, equipped with interference filters of different bandwidths centred on the W I (400.88 nm) emission line. A new algorithm for the subtraction of the continuum radiation was successfully developed and is now used to evaluate the W erosion even in the inner divertor region where the strong recombination emission is dominating over the tungsten emission. Analysis of W sputtering and W redistribution in the divertor by video imaging spectroscopy with high spatial resolution for three different magnetic configurations was performed. A strong variation of the emission of the neutral tungsten in toroidal direction and corresponding W erosion has been observed. It correlates strongly with the wetted area with a maximal W erosion at the edge of the divertor tile.

An extension of the SolEdge2D-EIRENE code package, named D-WEE, has been developed to add the dynamics of thermal desorption of hydrogen isotopes from the surface of plasma facing materials. To achieve this purpose, D-WEE models hydrogen isotopes implantation, transport and retention in those materials. Before launching auto-consistent simulation (with feedback of D-WEE on SolEdge2D-EIRENE), D-WEE has to be initialised to ensure a realistic wall behaviour in terms of dynamics (pumping or fuelling areas) and fuel content. A methodology based on modelling is introduced to perform such initialisation. A synthetic plasma pulse is built from consecutive SolEdge2D-EIRENE simulations. This synthetic pulse is used as a plasma background for the D-WEE module. A sequence of plasma pulses is simulated with D-WEE to model a tokamak operation. This simulation enables to extract at a desired time during a pulse the local fuel inventory and the local desorption flux density which could be used as initial condition for coupled plasma-wall simulations. To assess the relevance of the dynamic retention behaviour obtained in the simulation, a confrontation to post-pulse experimental pressure measurement is performed. Such confrontation reveals a qualitative agreement between the temporal pressure drop obtained in the simulation and the one observed experimentally. The simulated dynamic retention during the consecutive pulses is also studied.

Post mortem analysis shows that mid and high atomic number metallic impurities are present in deposits on JET plasma facing components with the highest amount of Ni and W, and therefore the largest sink, being found at the top of the inner divertor. Sources are defined as "continuous" or "specific", in that "continuous" sources arise from ongoing erosion from plasma facing surfaces and specific" are linked with specific events which decrease over time until they no longer act as a source. This contribution evaluates the sinks and estimates sources and the balance gives an indication of the dominating processes. Charge exchange neutral erosion is found to be the main source of nickel, whereas erosion of divertor plasma facing components is the main source of tungsten. Specific sources are shown to have little influence over the global mid- and high-Z impurity concentrations in deposits.

Beryllium oxide (BeO) and deuteroxide (BeOxDy) have been found on the melted zone of a beryllium tile extracted from the upper dump plate of JET-ILW (2011-2012 campaign). Results have been obtained using Raman microscopy, which is sensitive to both the chemical bond and crystal structure, with a micrometric lateral resolution. BeO is found with a wurtzite crystal structure. BeOxDy is found as three different types which are not the beta-phase but behaves as molecular species like Be(OD)(2), O(Be-D)(2) and DBeOD. The presence of a small amount of trapped D2O is also suspected. Our results therefore strongly suggest that D trapping occurs after melting through the formation of deuteroxides. The temperature increase favors the formation of crystal BeO which favors deuterium trapping through OD bonding.

The MHD response and the penetration of a deuterium shattered pellet into a JET plasma is investigated via the non-linear reduced MHD code JOREK with the neutral gas shielding (NGS) ablation model. The dominant MHD destabilizing mechanism by the injection is identified as the local helical cooling at each rational surface, as opposed to the global current profile contraction. Thus the injected fragments destabilize each rational surface as they pass through them. The injection penetration is found to be much better compared to MGI, with the convective transport caused by core MHD instabilities (e.g. 1/1 kink) contributing significantly to the core penetration. Moreover, the injection with realistic JET SPI system configurations is simulated in order to provide some insights into future operations, and the impact on the total assimilation and penetration depth of varying injection parameters such as the injection velocity or fineness of shattering is assessed. Further, the effect of changing the target equilibrium temperature or q profile on the assimilation and penetration is also investigated. Such analysis will form the basis of further investigation into a desirable configuration for the future SPI system in ITER.

Tungsten (W) melting due to transient power loads, for example those delivered by edge localised modes (ELMs), is a major concern for next step fusion devices. A series of experiments has been performed on JET to investigate the dynamics of Type-I ELM-induced transient melting. Following initial exposures in 2013 of a W-lamella with sharp leading edge in the bulk W outer divertor, new experiments have been performed in 2016-2017 on a protruding W-lamella with a 15 degrees slope, allowing direct and spatially resolved (0.85 mm/pixel) observation of the top surface using the IR thermography system viewing from the top of the poloidal cross-section. Thermal and IR analysis have already been conducted assuming the geometrical projection of the parallel heat flux on the W-lamellas, thus ignoring the gyro-radius orbit of plasma particles. Although it is well justified during L-mode or inter-ELM period, the hypothesis becomes questionable during ELM when the ion Larmor radius is larger. The goal of this paper is to extend the previous analysis based on the forward approach to the H-mode discharges and investigate in particular the gyro-radius effect during the Type-I ELMs, those used to achieve transient melting on the slope of the protruding W-lamella. Surface temperatures measured by the IR camera are compared with reconstructed synthetic data from 3D thermal modelling using heat loads derived from optical projection of the parallel heat flux (ignoring the gyro-radius orbit), 2D gyro-radius orbit and particle-in-cell (PIC) simulations describing the influence of finite Larmor-radius effects and electrical potential on the deposited power flux. Results show that the ELM power deposition behaves differently than the optical projection of the parallel heat flux, contrary to the L-mode observations, and may thus be due to the much larger gyro-orbits of the energetic ELM ions in comparison to L-mode or inter-ELM conditions.

Although there is growing evidence that cortistatin regulates several functions in different tissues, its role in the endocrine pancreas is not totally known. Here, we aim to study the effect of cortistatin on pancreatic beta-cells and glucose-stimulated insulin secretion (GSIS). Exposure of isolated mouse islets to cortistatin inhibited GSIS. This effect was prevented using a somatostatin receptor antagonist. Additionally, cortistatin hyperpolarized the membrane potential and reduced glucose-induced action potentials in isolated pancreatic beta-cells. Cortistatin did not modify ATP-dependent K+ (KATP) channel activity. In contrast, cortistatin increased the activity of a small conductance channel with characteristics of G protein-coupled inwardly rectifying K+ (GIRK) channels. The cortistatin effects on membrane potential and GSIS were largely reduced in the presence of a GIRK channel antagonist and by down-regulation of GIRK2 with small interfering RNA. Thus, cortistatin acts as an inhibitory signal for glucose-induced electrical activity and insulin secretion in the mouse pancreatic beta-cell.

Parallel heat flux calculations at the JET divertor have been based on the assumption that all incoming heat is due to the projection of the heat flux parallel to the magnetic line, q , plus a constant background. This simplification led to inconsistencies during the analysis of a series of dedicated tungsten melting experiments performed in 2013, for which infrared (IR) thermography surface measurements could not be recreated through simulations unless the parallel heat flux was reduced by 80% for L-mode and 60% for H-mode. We give an explanation for these differences using a new IR inverse analysis code, a set of geometrical corrections, and most importantly an additional term for the divertor heat flux accounting for non-parallel effects such as cross-field transport, recycled neutrals or charge exchange. This component has been evaluated comparing four different geometries with impinging angles varying from 2 to 90°. Its magnitude corresponds to 1.2%–1.9% of q , but because it is not affected by the magnetic projection, it accounts for up to 20%–30% of the tile surface heat flux. The geometrical corrections imply a further reduction of 24% of the measured heat flux. In addition, the application of the new inverse code increases the accuracy of the tile heat flux calculation, eliminating any previous discrepancy. The parallel heat flux computed with this new model is actually much lower than previously deduced by inverse analysis of IR temperatures—40% for L-mode and 50% for H-mode—while being independent of the geometry on which it is measured. This main result confirms the validity of the optical projection as long as a non-constant and non-parallel component is considered. For a given total heating power, the model predicts over 10% reduction of the maximum tile surface heat flux compared to strict optical modelling, as well as a 30% reduced sensitivity to manufacturing and assembling tolerances. These conclusions, along with the improvement in the predictabil

This paper deals with a divertor coil currents optimized procedure to design High Flux Expansion (HFE) configurations in the JET tokamak aimed to study the effects of flux expansion variation on the radiation fraction and radiated power re-distribution. A number of benefits of HFE configuration have been experimentally demonstrated on TCV, EAST, NSTX and DIII-D tokamaks and are under investigation for next generation devices, as DEMO and DTT. The procedure proposed here exploits the linearized relation between the plasma-wall gaps and the Poloidal Field (PF) coil currents. Once the linearized model is provided by means of CREATE-NL code, the divertor coils currents are calculated using a constrained quadratic programming optimization procedure, in order to achieve HFE configuration. Flux expanded configurations have been experimentally realized both in ohmic and heated plasma with and without nitrogen seeding. Preliminary results on the effects of the flux expansion variation on total power radiation increase will be also briefly discussed.

The impact of the finite grid size in SOLPS-ITER edge plasma simulations is assessed for JET H-mode discharges with a metal wall. For a semi-horizontal divertor configuration it is shown that the separatrix density is at least 30% higher when a narrow scrape-off layer (SOL) grid width is chosen in SOLPS-ITER compared to the case for which the SOL grid width is maximised. The density increase is caused by kinetic neutrals being not confined inside the divertor region because of the reduced extent of the plasma grid. In this case, an enhanced level of reflections of energetic neutrals at the low-field side (LFS) metal divertor wall is observed. This leads to a shift of the ionisation source further upstream which must be accounted for as a numerical artefact. An overestimate in the cooling at the divertor entrance is observed in this case, identified by a reduced heat flux decay parameters lambda(div)(q). Otherwise and further upstream the mid-plane heat decay length lambda(q) parameter is not affected by any change in divertor dissipation. This confirms the assumptions made for the ITER divertor design studies, i.e. that lambda(q) upstream is essentially set by the assumptions for the ratio radial to parallel heat conductivity. It is also shown that even for attached conditions the decay length relations lambda(ne)>lambda(Te)>lambda(q) hold in the near-SOL upstream. Thus for interpretative edge plasma simulations one must take the (experimental) value of lambda(ne) into account, rather than lambda(q), as the former actually defines the required minimum upstream SOL grid extent.

A tritium imaging plate technique (TIPT) in combination with an electron-probe microscopic analysis (EPMA) were applied to examine tritium (T) retention characteristics in individual dust particles collected in the Joint European Torus with the ITER-like Wall (JET-ILW) after the first campaign in 2011-2012. A lot of carbon (C)-dominated dust particles were found, which would be pre-existing carbon deposits in the JET-C or released carbon particles from the remaining carbon-fiber components in the JET-ILW. Most of T was retained at the surface of and/or in the C-dominated dust particles. The retention in tungsten, beryllium and other metal-dominated dust particles is relatively lower by a factor of 10-100 in comparison with that in the Cdominated particles.

In the past at JET, with the MkI divertor, a systematic study of the influence of X-point height and poloidal flux expansion has been conducted [1,2] showing minor differences in the radiation distribution, whereas in [3] experiment and simulations have shown enhancement of detachment as the flux expansion was increased. More recently at JET, equipped with the ITER-like wall (ILW), radiative seeded scenarios have been studied and a maximum radiation fraction 75% has been achieved. EDGE2D-EIRENE [4–6] simula- tions [7,8] have already shown that the divertor heat fluxes can be reduced with N2 injection, qualita- tively consistent with experimental observations [9] , by adjusting the impurity injection rate to reproduce the measured divertor radiation. In this paper we will present edge predictive simulations on the assess- ment of effects of poloidal flux expansion and recycling on radiation distribution and X-point peaking on JET-ILW nitrogen seeded plasmas.

Neutronics benchmark experiments are conducted at JET in the frame of WPJET3 NEXP within EUROfu-sion Consortium for validating the neutronics codes and tools used in ITER nuclear analyses to predictquantities such as the neutron flux along streaming paths and dose rates at the shutdown due to acti-vated components. The preparation of neutron streaming and shutdown dose rate experiments for thefuture Deuterium-Tritium operations (DTE2 campaign) are in progress. This paper summarizes the statusof measurements and analyses in progress in the current Deuterium–Deuterium (DD) campaign and theefforts in preparation for DTE2.

Tritium depth profile and its temperature programmed desorption rate were determined for selected samples cut out of beryllium tiles removed from the Joint European Torus vacuum vessel during the 2012 shut down. A beryllium dissolution method under controlled conditions was used to determine the tritium depth profile in the samples, whereas temperature programmed desorption experiments were performed to assess tritium release pattern. Released tritium was measured using a proportional gas flow detector. Prior to desorption and dissolution experiments, the plasma-facing surfaces of the samples were studied by scanning electron microscopy and energy dispersive X-ray spectroscopy. Experimental results revealed that > 95% of the tritium was localized in the top 30 –45 μm of the ‘plasma-facing’ surface, however, possible tritium presence up to 100 μm cannot be excluded. During tem- perature programmed desorption at 4.8 K/min in the flow of purge gas He + 0.1% H 2 the tritium release started below 475 K, the most intense release occurred at 725 –915 K and the degree of detritiation of > 91% can be obtained upon reaching 1075 K. The total tritium activity in the samples was in range of 2 –32 kilo Becquerel per square centimetre of the plasma-facing surface area.

For simulation of plasma-facing component erosion in fusion experiments, an analytical expression for the ion velocity just before the surface impact including the local electric field and an optional surface biasing effect is suggested. Energy and angular impact distributions and the resulting effective sputtering yields were produced for several experimental scenarios at JET ILW mostly involving PFCs exposed to an oblique magnetic field. The analytic solution has been applied as an improvement to earlier ERO mod- elling of localized, Be outer limiter, RF-enhanced erosion, modulated by toggling of a remote, however magnetically connected ICRH antenna. The effective W sputtering yields due to D and Be ion impact in Type-I and Type-III ELMs and inter-ELM conditions were also estimated using the analytical approach and benchmarked by spectroscopy. The intra-ELM W sputtering flux increases almost 10 times in comparison to the inter-ELM flux.

The JET tangential gamma-ray spectrometer is undergoing an extensive upgrade in order to make itcompatible with the forthcoming deuterium-tritium (DT) experiments. The paper presents the resultsof the design for the main components for the upgrade of the spectrometer beam-line: tandem collima-tors, gamma-ray shields, and neutron attenuators. The existing tandem collimators will be upgraded byinstalling two additional collimator modules. Two gamma-ray shields will define the gamma-ray field-of-view at the detector end of the spectrometer line-of-sight. A set of three lithium hydride neutronattenuators will be used to control the level of the fast neutron flux on the gamma-ray detector. Thedesign of the upgraded spectrometer beam-line has been supported by extensive radiation (neutron andphoton) transport calculations using both large volume and point radiation sources.

The analysis of the radiation damage of imaging systems is based on all different types-of aiialoiue/digital cameras with uncooled as well as actively cooled image sensors in the VIS/NIR/MWIR spectral ranges. The Monte Carlo N-Particle (MCNP) code has been used to determine the neutron fluence at different camera locations in JET. An explicit link between the sensor damage and the neutron fluence has been observed. Sensors show an increased dark-current and increased numbers of hot-pixels. Uncooled cameras must be replaced once per year after exposure to a neutron fluence of similar to 1.9-3.2 x 10(12)neutrons/cm(2). Such levels of fluence will be reached after approximate to 14-22 ELMy H-mode pulses during the future D-T campaign. Furthermore, dynamical noise seen as a random pattern of bright pixels was observed in the presence of hard radiation (neutrons and gammas). Failure of the digital electronics inside the cameras as well as of industrial controllers is observed beyond a neutron fluence of about similar to 4 x 10(9) neutrons/cm(2). The impact of hard radiation on the different types of electronics and possible application of cameras during future D-T campaign is discussed.

The assessment of the Shutdown Dose Rate (SDR) due to neutron activation is a major safety issue forfusion devices and in the last decade several benchmark experiments have been conducted at JET duringDeuterium-Deuterium experiments for the validation of the numerical tools used in ITER nuclear analy-ses. The future Deuterium-Tritium campaign at JET (DTE2) will provide a unique opportunity to validatethe codes under ITER-relevant conditions through the comparison between numerical predictions andmeasured quantities (C/E). For this purpose, a novel SDR experiment, described in the present work, isin preparation in the frame of the WPJET3-NEXP subproject within EUROfusion Consortium. The exper-imental setup has been accurately designed to reduce measurement uncertainties; spherical air-ventedionization chambers (ICs) will be used for on-line ex-vessel decay gamma dose measurements duringJET shutdown following DT operations and activation foils have been selected for measuring the neutronfluence near ICs during operations. Active dosimeters (based on ICs) have been calibrated over a broadenergy range (from about 30 keV to 1.3 MeV) with X and gamma reference beam qualities. Neutron irra-diation tests confirmed the capability of active dosimeters of performing on-line decay gamma dose ratemeasurements, to follow gamma dose decay at the end of neutron irradiation as well as insignificantactivation of the ICs.

The erosion of Be and W marker layers was investigated using long-term samples containing marker lay- ers during the second ITER-like wall discharge campaign 2013–2014 (ILW-2). The samples were mounted in Be coated Inconel tiles between the inner wall guard limiters (IWGL). They were analyzed using elas- tic backscattering (EBS) before and after exposure. All samples showed noticeable erosion. The results were compared to the data for Be and W erosion rates for the first 2011–2012 JET ITER-like wall (ILW-1) campaign, and to the data for C erosion during the 20 05–20 09 campaign when JET was operated with a carbon wall. The mean W erosion rates and the toroidal and poloidal distributions of the W erosion were nearly the same for the ILW-1 and ILW-2 campaigns. The mean erosion rate of Be during the ILW-2 campaign was smaller by a factor of about two compared to the ILW-1 campaign.

In the past at JET, with the MkI divertor, a systematic study of the influence of X-point height and poloidal flux expansion has been conducted [1,2] showing minor differences in the radiation distribution, whereas in [3] experiment and simulations have shown enhancement of detachment as the flux expansion was increased. More recently at JET, equipped with the ITER-like wall (ILW), radiative seeded scenarios have been studied and a maximum radiation fraction 75% has been achieved. EDGE2D-EIRENE [4–6] simula- tions [7,8] have already shown that the divertor heat fluxes can be reduced with N2 injection, qualita- tively consistent with experimental observations [9] , by adjusting the impurity injection rate to reproduce the measured divertor radiation. In this paper we will present edge predictive simulations on the assess- ment of effects of poloidal flux expansion and recycling on radiation distribution and X-point peaking on JET-ILW nitrogen seeded plasmas.

New techniques that attempt to more fully exploit spectroscopic diagnostics in the divertor and pedestal region during highly dissipative scenarios are demonstrated using experimental results from recent low-Z seeding experiments on Alcator C-Mod, JET and ASDEX Upgrade. To exhaust power at high parallel heat flux, q ‖ > 1 GW/m 2 , while minimizing erosion, reactors with solid, high-Z plasma facing components (PFCs) are expected to use extrinsic impurity seeding. Due to transport and atomic physics processes which impact impurity ionization balance, so-called ‘non-coronal’ effects, we do not accurately know and have yet to demonstrate the maximum q ‖ which can be mitigated in a tokamak. Radiation enhancement for nitrogen is shown to arise primarily from changes in Li- and Be-like charge states on open field lines, but also through transport-driven enhancement of H- and He-like charge states in the pedestal region. Measurements are presented from nitrogen seeded H-mode and L-mode plasmas where emission from N 1+ through N 6+ are observed. Active charge exchange spectroscopy of partially ionized low-Z impuri- ties in the plasma edge is explored to measure N 5+ and N 6+ within the confined plasma, while passive UV and visible spectroscopy is used to measure N 1+ -N 4+ in the boundary. Examples from recent JET and Alcator C-Mod experiments which employ nitrogen seeding highlight how improving spectroscopic cov- erage can be used to gain empirical insight and provide more data to validate boundary simulations.