[Show abstract][Hide abstract] ABSTRACT: The CXRS (Charge-eXchange Recombination Spectroscopy) diagnostic for the core plasma of ITER will be designed to provide observation of the dedicated diagnostic beam (DNB) over a wide radial range, roughly from a normalised radius r/a = 0.7 to close to the plasma axis. The collected light will be transported through the Upper Port Plug #3 (UPP3) to a bundle of fibres and ultimately to a set of remote spectrometers. The design is particularly challenging in view of the ITER environment of particle, heat and neutron fluxes, temperature cycles, electromagnetic loads, vibrations, expected material degradation and fatigue, constraints against tritium penetration, integration in the plug and limited opportunities for maintenance. Moreover, a high performance (étendue × transmission, dynamic range) is expected for the port plug system since the beam attenuation is large and the background light omnipresent, especially in terms of bremsstrahlung, line radiation and reflections. The present contribution will give an overview of the current status and activities which deal with the core CXRS system, summarising the investigations which have taken place before entering the actual development and design phase.
[Show abstract][Hide abstract] ABSTRACT: For the interpretation of the line radiation observed from laser induced ablation spectroscopy (LIAS) such parameters as the density and temperature of electrons within very compact clouds of atoms and singly charged ions of ablated material have to be known. Compared to the local plasma conditions prior to the laser pulse, these can be strongly changed during LIAS since new electrons are generated by the ionisation of particles ejected from the irradiated target. Because of their transience and spatial inhomogeneity it is technically difficult to measure disturbances induced in the plasma by LIAS. To overcome this uncertainty a numerical model has been elaborated, providing a self-consistent description for the spreading of ablated particles and accompanying modifications in the plasma. The results of calculations for LIAS performed on carbon-containing targets in Ohmic and additionally heated discharges in the tokamak TEXTOR are presented. Due to the increase in the electron density the 'ionisation per photon' ratio, S/XB factor, is significantly enhanced compared to unperturbed plasma conditions. The impact of the amount of material ablated and of the plasma conditions before LIAS on the level of the S/XB-enhancement is investigated.
[Show abstract][Hide abstract] ABSTRACT: At first a detailed fast shutter design was finalized for the ITER core charge exchange recombination spectroscopy (CXRS) diagnostic. The shutter has approximately 70 kg of mass and a length of 2.1 m. It operates in fractions of a second (0.7 s) protecting critical optical components against degradation and providing means of calibration for the optical system. The shutter structure is driven by a bidirectional frictionless helium actuator, with forces and axial strokes of 3.4 kN and 2 mm respectively. The shutter structure consists of: (a) two blades made of CuCrZr and stainless steel, calibration surfaces (currently Al2O3) on the top and on the bottom a protective TZM (Mo–0.5Ti–0.08Zr) screens, (b) two arms interconnected that form one cooling circuit including the blades, (c) a bumper system to limit the arms movement, and (d) a support. A description of these components and their functions are given in this paper, followed by some issues, and their corresponding solutions or ongoing investigations, encountered during the design work. Detailed manufacturing drawings have been developed as the deliverable final product of this design stage, and are used in the prototyping phase which includes testing, numerical benchmarking, and validation of the shutter concept.
[Show abstract][Hide abstract] ABSTRACT: Wendelstein 7-X being the most advanced stellarator is currently prepared for commissioning at Greifswald. Forschungszentrum Jülich is preparing a research programme in the field of plasma wall interactions (PWI) by developing a dedicated set of diagnostic systems. The specific interest at Wendelstein 7-X is to understand PWI processes in presence of a 3D plasma boundary of an island divertor. Furthermore, for the first time steady state plasma at high density and low temperature in the divertor region will be available. Since PWI only could be understood in conjunction with the edge plasma properties the aim of the setup is to observe both the edge plasma as well as surface processes. For optimum combination of different diagnostic methods the edge diagnostic systems are aligned toroidally along one out of five magnetic islands. Main systems are a multipurpose fast probe manipulator, two gas boxes in opposite divertor modules together with two endoscopes each observing the divertor regions, a poloidal correlation reflectometer, a dispersion interferometer in the divertor, and VUV and X-ray spectroscopy in the plasma core. The concept of the diagnostic setup is presented in this paper.
[Show abstract][Hide abstract] ABSTRACT: The chemical kinetics of the reaction of thin films with reactive gases is investigated. The removal of thin films using thermally activated solid-gas to gas reactions is a method to in-situ control deposition inventory in vacuum and plasma vessels. Significant scatter of experimental deposit removal rates at apparently similar conditions was observed in the past, highlighting the need for understanding the underlying processes. A model based on the presence of reactive gas in the films bulk and chemical kinetics is presented. The model describes the diffusion of reactive gas into the film and its chemical interaction with film constituents in the bulk using a stationary reaction-diffusion equation. This yields the reactive gas concentration and reaction rates. Diffusion and reaction rate limitations are depicted in parameter studies. Comparison with literature data on tokamak co-deposit removal results in good agreement of removal rates as a function of pressure, film thickness and temperature.
[Show abstract][Hide abstract] ABSTRACT: An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s−1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas.
[Show abstract][Hide abstract] ABSTRACT: Thermo-chemical removal (TCR), or baking in reactive gases, is a candidate method to control the co-deposit related tritium inventory in fusion devices. TCR can be understood as reaction–diffusion processes in a porous material. O2-TCR was applied to 150–550 nm thick a-C:D layers with similar textures. A linear relation between the integral TCR rate and the layer thickness, as predicted by the understanding, was observed in the experiment, i.e. the time to remove the hydrogen inventory is independent of its initial amount. TCR with nitrogen dioxide (NO2) at temperatures of 200–350 °C was conducted with a set of a-C:D and W–C–H layers. At 350 °C NO2 removed ~ 15% porosity a-C:D within 3 min. The O retention in remaining a-C:D was ≈ 1017 O cm−2. An activation energy of ≈ 0.78 eV for reactions of NO2 with D and C was determined. The results were applied for predictions of the TCR effectivity in ITER. The treatment of W–C–H led to O uptake (O/W ≈ 2–3), while W and C contents remained unchanged.
[Show abstract][Hide abstract] ABSTRACT: The ITER-like wall recently installed in JET comprises solid beryllium limiters and a combination of bulk tungsten and tungsten-coated carbon fibre composite divertor tiles without active cooling. During a beryllium power handling qualification experiment performed in limiter configuration with 5 MW neutral beam injection input power, accidental beryllium melt events, melt layer motion and splashing were observed locally on a few beryllium limiters in the plasma contact areas. The Lorentz force is responsible for the observed melt layer movement. To move liquid beryllium against the gravity force, the current flowing from the plasma perpendicularly to the limiter surface must be higher than 6 kA m−2. The thermo-emission current at the melting point of beryllium is much lower. The upward motion of the liquid beryllium against gravity can be due to a combination of the Lorentz force from the secondary electron emission and plasma pressure force.
[Show abstract][Hide abstract] ABSTRACT: Laser induced ablation spectroscopy (LIAS) is a diagnostic to provide temporally and spatially resolved in situ measurements of tritium retention and material migration in order to characterize the status of the first wall in future fusion devices. In LIAS, a ns-laser pulse ablates the first nanometres of the first wall plasma-facing components into the plasma edge. The resulting line radiation by plasma excitation is observed by spectroscopy. In the case of the full ionizing plasma and with knowledge of appropriate photon efficiencies for the corresponding line emission the amount of ablated material can be measured in situ. We present the photon efficiency for the deuterium Balmer alpha-line resulting from ablation in TEXTOR by performing LIAS on amorphous hydrocarbon (a-C:D) layers deposited on tungsten substrate of thicknesses between 0.1 and 1.1 mu m. An experimental inverse photon efficiency of [D/XB](D alpha (EXP))(a-C:D -> D)(LIAS) = 75.9 +/- 23.4 was determined. This value is a factor 5 larger than predicted values from the ADAS database for atomic injection of deuterium under TEXTOR plasma edge conditions and about twice as high, assuming normal wall recycling and release of molecular deuterium and break-up of D-2 via the molecular ion which is usually observed at the high temperature tokamak edge (T-e > 30 eV).
[Show abstract][Hide abstract] ABSTRACT: The control of the radioactive inventory in the vacuum vessel of ITER is a main safety issue. Erosion of activated plasma-facing components (PFC) and co-deposition of tritiated dust on PFC and in areas below the divertor constitute the main sources of in-vessel radioactive inventory mobilizable in the case of an accident and also during venting of the vessel. To trace the dust and tritium inventory in the machine, the use of collectors in the form of removable samples was evaluated, beside other techniques, since it provides a reliable way to follow the history of the deposits and check critical areas. Four types of removable probes and two optional active diagnostics were selected out of about 30 different options. For all four probes, a conceptual design was worked out and the feasibility was checked with preliminary estimations of thermal and electromagnetic loads, as well as remote handling paths. The highest temperature estimated for the front face of all probes lies in the range 300–500 °C, which is tolerable. Installed in representative places, such removable samples may provide information about the dust and tritium distribution inside the vacuum vessel.
[Show abstract][Hide abstract] ABSTRACT: Cracking thresholds and crack patterns in tungsten targets after repetitive ITER-like edge localized mode (ELM) pulses have been studied in recent simulation experiments by laser irradiation. The tungsten specimens were tested under selected conditions to quantify the thermal shock response. A Nd:YAG laser capable of delivering up to 32 J of energy per pulse with a duration of 1 ms at the fundamental wavelength λ = 1064 nm has been used to irradiate ITER-grade tungsten samples with repetitive heat loads. The laser exposures were performed for targets at room temperature (RT) as well as for targets preheated to 400 °C to measure the effects of the ELM-like loading conditions on the formation and development of cracks. The magnitude of the heat loads was 0.19, 0.38, 0.76 and 0.90 MJ m−2 (below the melting threshold) with a pulse duration of 1 ms. The tungsten surface was analysed after 100 and 1000 laser pulses to investigate the influence of material modification by plasma exposures on the cracking threshold. The observed damage threshold for ITER-grade W lies between 0.38 and 0.76 GW m−2. Continued cycling up to 1000 pulses at RT results in enhanced erosion of crack edges and crack edge melting. At the base temperature of 400 °C, the formation of cracks is suppressed.
[Show abstract][Hide abstract] ABSTRACT: The development of optical diagnostics, like endoscopes, compatible with the ITER environment (metallic plasma facing components, neutron proof optics, etc.) is a challenge, but current tokamaks such as JET provide opportunities to test fully working concepts. This paper describes the engineering aspects of a fully mirrored endoscope that has recently been designed, procured and installed on JET. The system must operate in a very strict environment with high temperature, high magnetic fields up to B = 4 T and rapid field variations (∂B/∂t ∼ 100 T/s) that induce high stresses due to eddy currents in the front mirror assembly. It must be designed to withstand high mechanical loads especially during disruptions, which lead to acceleration of about 7 g at 14 Hz. For the JET endoscope, when the plasma thermal loading, direct and indirect, was added to the assumed disruption loads, the reserve factor, defined as a ratio of yield strength over summed up von Mises stresses, was close to 1 for the mirror components. To ensure reliable operation, several analyses were performed to evaluate the thermo-mechanical performance of the endoscope and a final validation was obtained from mechanical and thermal tests, before the system's final installation in May 2011. During the tests, stability of the field of view angle variation was kept below 1° despite the high thermal gradient on endoscope head (∂T/∂x ∼ 500 K/m). In parallel, to ensure long time operation and to prevent undesirable performance degradation, a shutter system was also implemented in order to reduce impurity deposition on in-vessel mirrors but also to allow in situ transmission calibration. One of the main specifications of the shutter system was high reliability. This was considered as achieved when the prototype was successfully tested to 3000 cycles at a temperature of 300 °C.
No preview · Article · Oct 2013 · Fusion Engineering and Design
[Show abstract][Hide abstract] ABSTRACT: The design of the tile assemblies of the bulk tungsten divertor row in JET was improved in the course of several experiments as far as the power and energy performances are concerned: many prototypes were exposed to high heat fluxes in several electron and ion beam facilities during the development phase. These experiments were carried out in parallel with extensive modelling of the complete tungsten tile assembly in the so-called Global Thermal Model (GTM). The goal was to understand the heat flow from the plasma-facing surface through the supporting structure down to the base plate of the JET MkII divertor sufficiently to be able to later interpret operational data from the torus. Temperatures measured in the torus are in good agreement (−10/+15%) with the model. Some characteristic times show stronger deviations, with no incidence on the highest temperature at all times.
No preview · Article · Oct 2013 · Fusion Engineering and Design
[Show abstract][Hide abstract] ABSTRACT: In preparation for ITER, JET has been upgraded with a new ITER-like wall (ILW) whereby the main plasma-facing components, previously made of carbon, have been replaced by Be in the main chamber and Win the divertor. A new endoscope with optimised divertor view has been developed in order to survey and monitor the emission of specific impurities such as tungsten, beryllium and the possibly remaining carbon in the tungsten divertor of the JET-ILW. It operates in the wavelength range from 390 nm to 2500 nm with high optical transmittance (>= 30%) as well as high spatial resolution, that is <= 2 mm at the object plane and <= 3 mm over the whole depth of field (+/- 0.7 m). The endoscope is a prototype for testing an ITER relevant design concept based on reflective optics only. The endoscope has an optimised observation in the near ultraviolet and in the blue spectral region to ensure the detection of the W I-emission line at 400.8 nm. In parallel to the new optical design, a new type of ITER-like shutter system based on pneumatic techniques has been developed and integrated in the endoscope head. The new optical design includes options for an in situ calibration of the endoscope transmittance during the experimental campaign.
No preview · Article · Oct 2013 · Fusion Engineering and Design
[Show abstract][Hide abstract] ABSTRACT: JET underwent a transformation from a full carbon-dominated tokamak to a fully metallic device with beryllium in the main chamber and a tungsten divertor. This material combination is foreseen for the activated phase of ITER. The ITER-Like Wall (ILW) experiment at JET shall demonstrate the plasma compatibility with metallic walls and the reduction in fuel retention. We report on a set of experiments (Ip = 2.0 MA, Bt = 2.0-2.4 T, δ = 0.2-0.4) in different confinement and plasma conditions with global gas balance analysis demonstrating a strong reduction in the long-term retention rate by more than a factor of 10 with respect to carbon-wall reference discharges. All experiments are executed in a series of identical plasma discharges in order to achieve maximum plasma duration until the analysis limit of the active gas handling system is reached. The composition analysis shows high purity of the recovered gas, typically 99% D. For typical L-mode discharges (Paux = 0.5 MW), type III (Paux = 5.0 MW) and type-I ELMy H-mode plasmas (Paux = 12.0 MW) a drop of the deuterium retention rate normalized to the operational time in divertor configuration is measured from 1.27 × 1021, 1.37 × 1021 and 1.97 × 1021 D s-1 down to 4.8 × 1019, 7.2 × 1019 and 16 × 1019 D s-1, respectively. The dynamic retention increases in the limiter phase in comparison with carbon-fibre composite, but also the outgassing after the discharge has risen in the same manner and overcompensates this transient retention. Overall an upper limit of the long-term retention rate of 1.5 × 1020 D s-1 is obtained with the ILW. The observed reduction by one order of magnitude confirms the expected predictions concerning the plasma-facing material change in ITER and is in line with identification of fuel co-deposition with Be as the main mechanism for the residual long-term retention. The reduction widens the operational space without active cleaning in the DT phase in comparison with a full carbon device.
[Show abstract][Hide abstract] ABSTRACT: Molecular spectroscopy was used to observe molecular deuterium at the
outer strike point of the new bulk tungsten JET divertor. The rotational
and vibrational populations of the deuterium molecules in the ground
state were determined from the deuterium Q-branches of Fulcher-α
band emission (d3Πu-→a3Σg+) in the
600-640 nm spectral range. For L-mode plasmas in the low recycling
regime the molecular emission maximum is located in the vicinity of the
strike point. The spatial profile of the emission was strongly modified
during plasma detachment in both L- and H-mode plasmas. The rotational
temperature of excited molecules reached 2760 K in L-mode. The
vibrational population has a peculiarity: a remarkably high population
of the d3Πu-(v = 0) vibrational level indicating a
non-Boltzmann vibrational distribution of D2 in tungsten
No preview · Article · Jul 2013 · Journal of Nuclear Materials