ITER Divertor Research Articles

Usage of Ray Tracing Transfer Matrix to Mitigate the Stray Light for ITER Spectroscopy

AbstractStray light formed by the reflection of photons on inner wall from a bright divertor region can be a serious issue in spectroscopic measurement systems in ITER. In this study, we propose a method to mitigate the influence of stray light using a ray tracing analysis. Usually, a ray tracing simulation requires a time consuming runs. We constructed transfer matrices based on the ray tracing simulation results and used them to demonstrate the influence of stray light. It is shown that the transfer matrix can be used to reconstruct the emission profile by considering the influence of the stray light without any additional ray tracing runs. Mitigation of the stray light in ITER divertor impurity monitor was demonstrated, and a method of prediction of the stray light level for the scrape off layer spectroscopy from divertor region was proposed. (© 2016 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Development of the in Situ Calibration Method for ITER Divertor IR Thermography

An important issue for ITER divertor infrared (IR) thermography (IRTh) is that changes in the emissivity of tungsten divertor targets resulting from depositions; erosions; and dependences on temperature, wavelength, and surface roughness affect the temperature measurement, which requires an accuracy of 10%. Therefore, we investigated the emissivity dependences of tungsten samples in ITER-grade tungsten and validated the proposed in situ calibration method for emissivity evaluation by using an IR laser in laboratory experiments. The emissivity of the tungsten samples had a strong dependence on surface roughness of 1.0 to 5.9 μm. In the two-color method, by measuring the radiances of the tungsten sample in two wavelengths of 3.35 and 4.67 μm, the change of the ratio of the emissivities did not satisfy the measurement requirement. Thus, an in situ calibration method of emissivity is needed. The emissivity evaluated using the in situ calibration method was in good agreement with the emissivity evaluated from the radiance for tungsten samples at temperatures of 22°C, 100°C, and 400°C. Consequently, the in situ calibration method for emissivity evaluation using an IR laser was successfully validated. More work is needed for the application in IRTh.

Surface damage and structure evolution of recrystallized tungsten exposed to ELM-like transient loads

Surface damage and structure evolution of the full tungsten ITER divertor under transient heat loads is a key concern for component lifetime and plasma operations. Recrystallization caused by transients and steady-state heat loads can lead to degradation of the material properties and is therefore one of the most serious issues for tungsten armor. In order to investigate the thermal response of the recrystallized tungsten under edge localized mode-like transient thermal loads, fully recrystallized tungsten samples with different average grain sizes are exposed to cyclic thermal shocks in the electron beam facility JUDITH 1. The results indicate that not only does the microstructure change due to recrystallization, but that the surface residual stress induced by mechanical polishing strongly influences the surface cracking behavior. The stress-free surface prepared by electro-polishing is shown to be more resistant to cracking than the mechanically polished one. The resulting surface roughness depends largely on the loading conditions instead of the recrystallized-grain size. As the base temperature increases from room temperature to 400 °C, surface roughening mainly due to the shear bands in each grain becomes more pronounced, and sub-grains (up to 3 μm) are simultaneously formed in the sub-surface. The directions of the shear bands exhibit strong grain-orientation dependence, and they are generally aligned with the traces of {1 1 2} twin habit planes. The results suggest that twinning deformation and dynamic recrystallization represent the predominant mechanism for surface roughening and related microstructure evolution.

Melt damage to the JET ITER-like Wall and divertor

In October 2014, JET completed a scoping study involving high power scenario development in preparation for DT along with other experiments critical for ITER. These experiments have involved intentional and unintentional melt damage both to bulk beryllium main chamber tiles and to divertor tiles. This paper provides an overview of the findings of concern for machine protection in JET and ITER, illustrating each case with high resolution images taken by remote handling or after removal from the machine. The bulk beryllium upper dump plate tiles and some other protection tiles have been repeatedly flash melted by what we believe to be mainly fast unmitigated disruptions. The flash melting produced in this way is seen at all toroidal locations and the melt layer is driven by j × B forces radially outward and upwards against gravity. In contrast, the melt pools caused while attempting to use MGI to mitigate deliberately generated runaway electron beams are localized to several limiters and the ejected material appears less influenced by j × B forces and shows signs of boiling. In the divertor, transient melting of bulk tungsten by ELMs was studied in support of the ITER divertor material decision using a specially prepared divertor module containing an exposed edge. Removal of the module from the machine in 2015 has provided improved imaging of the melt and this confirms that the melt layers are driven by ELMs. No other melt damage to the other 9215 bulk tungsten lamellas has yet been observed.

Evaluation of performance for the EAST upgraded divertor targets during type I ELMy H-mode

The long-pulse high-confinement (H-mode) plasma regime is considered to be a preferable scenario in future fusion devices, and in the period of normal operation during H-mode, edge-localised modes (ELMs) are one of the most serious threats to the performance and capability of divertor targets. The EAST recently achieved a variety of H-mode regimes with ELMs. For the purpose of studying the performance of the EAST upgraded divertor during type I ELMs, a series of simulations were performed by using three-dimensional (3D) finite element code. To make a visible outcome of the direct ELM impact on the divertor targets, a preliminary evaluation system with three indices to exhibit the influence has been developed. The indices that comprise temperature evolution, thermal penetration depth and crack initiation life, which could reveal the process of micro-crack formation, are calculated in both low and high-power scenarios for type I ELMs. The initial results indicate that the transient heat load has a significant influence in a very short thickness layer along the direction perpendicular to the plasma-facing surface throughout its duration. The conclusion could offer a pertinent guide to the next-step high-power long-pulse operation in EAST and would also be helpful for scientifically studying the damage and fatigue mechanism of the divertor in ITER and future fusion power reactors.

Modeling of complex gas distribution systems operating under any vacuum conditions: Simulations of the ITER divertor pumping system

An integrated software tool for modeling and simulation of complex gas distribution systems operating under any vacuum conditions is presented and validated. The algorithm structure includes (a) the input geometrical and operational data of the network, (b) the definition of the fundamental set of network loops and pseudoloops, (c) the formulation and solution of the mass and energy conservation equations, (d) the kinetic data base of the flow rates for channels of any length in the whole range of the Knudsen number, supporting, in an explicit manner, the solution of the conservation equations and (e) the network output data (mainly node pressures and channel flow rates/conductance). The code validity is benchmarked under rough vacuum conditions by comparison with hydrodynamic solutions in the slip regime. Then, its feasibility, effectiveness and potential are demonstrated by simulating the ITER torus vacuum system with the six direct pumps based on the 2012 design of the ITER divertor. Detailed results of the flow patterns and paths in the cassettes, in the gaps between the cassettes and along the divertor ring, as well as of the total throughput for various pumping scenarios and dome pressures are provided. A comparison with previous results available in the literature is included.

Pre-conceptual design activities for the materials plasma exposure experiment

The development of next step fusion facilities such as DEMO or a Fusion Nuclear Science Facility (FNSF) requires first closing technology gaps in some critical areas. Understanding the material-plasma interface is necessary to enable the development of divertors for long-pulse plasma facilities. A pre-conceptual design for a proposed steady-state linear plasma device, the Materials Plasma Exposure Experiment (MPEX), is underway. A helicon plasma source along with ion cyclotron and electron Bernstein wave heating systems will produce ITER divertor relevant plasma conditions with steady-state parallel heat fluxes of up to 40MW/m2 with ion fluxes up to 1024/m2s on target. Current plans are for the device to use superconducting magnets to produce 1–2T fields. As a steady-state device, active cooling will be required for components that interact with the plasma (targets, limiters, etc.), as well as for other plasma facing components (transport regions, vacuum tanks, diagnostic ports). Design concepts for the vacuum system, the cooling system, and the plasma heating systems have been completed. The device will include the capability for handling samples that have been neutron irradiated in order to consider the multivariate effects of neutrons, plasma, and high heat-flux on the microstructure of divertor candidate materials. A vacuum cask, which can be disconnected from the high field environment in order to perform in-vacuo diagnosis of the surface evolution is also planned for the facility.

Progress of ITER full tungsten divertor technology qualification in Japan: Manufacturing full-scale plasma-facing unit prototypes

Japan Atomic Energy Agency (JAEA) is in progress for technology demonstration toward Full-tungsten (W) ITER divertor outer vertical target (OVT), especially, W monoblock technology that needs to withstand the repetitive heat load as high as 20MW/m2 for 10s. Under the framework of the W divertor qualification program developed ITER organization, JAEA as Japanese Domestic Agency (JADA) manufactured seven full-scale plasma-facing unit (PFU) prototypes with the Japanese industries. Four prototypes that have 146W monoblock joint with casted copper (Cu) interlayer passed successfully the ultrasonic testing. In the other three prototypes that have the different W/Cu interlayer joint, joint defects were found. The dimension measurements reveal the requirements of the gap between W monoblocks and the surface profile of PFU are feasible.

磁化同軸プラズマガン装置におけるCTプラズマの特性評価とパルス高熱負荷模擬試験への適用

A magnetized coaxial plasma gun (MCPG) device at Univ. of Hyogo was used as a simulation facility for pulse heat loads such as ELMs (Edge Localized Modes) onto plasma-facing materials in magnetically confined fusion devices. In the present study, characterization of the compact toroid (CT) plasmas produced by the MCPG device was performed. Dependences of the CT plasma parameters of line-averaged electron density, ion temperature and flow velocity on the gun discharge voltage and the bias magnetic flux were measured. As a result, it was observed that the pulse length, the line-averaged electron density, and the ion temperature and flow velocity were approximately 200 µs, approximately 6×1021 m-3, approximately 30 eV, and approximately 50 km s-1, respectively. The surface energy density measured with a tungsten calorimeter was approximately 0.4 MJ m-2. Thus, it was shown that the ELM-like pulsed plasma heat loads with the energy density and the pulsed length predicted in the ITER divertor can be produced using the MCPG device.

Investigation of cat's eye mirror with seven reflective surfaces for in-situ spectral calibration system on divertor impurity monitor for ITER

Abstract The divertor impurity influx monitor (DIM) for ITER is spectroscopic system to observe lights emitted from thermonuclear plasma from near ultra violet to infrared region. The spectral characteristics of the DIM system will change during operation due to the harsh environment, and then an in-situ spectral calibration system has been designed for the ITER divertor impurity monitor. A cat's eye mirror with seven reflective surfaces was suggested to apply the system, and then the validation was evaluated in the full-scale ITER DIM optical system. The returning ratios of the cat's eye mirror and micro retroreflector array in the full-scale system were observed using four visible lasers of different wavelengths. The cat's eye mirror is validated for the ITER DIM system, and the potential of improvement by reduction of the reflections were suggested.

Study on the characteristics of thermal structural response of full tungsten divertor under ELM-like heat pulse

Damage cause and mechanism of the tungsten specimen under ELM-like heat pulse were examined by using a Nd:YAG laser experiment assuming the ITER divertor allowable heat load, 0.5MJ/m2 and finite-element analysis in terms of thermal mechanical response. The thermal stress distribution with the heat load area was compared by FEM and the stress distribution by a Nd:YAG laser corresponded to that expected in the actual divertor target. The crack growth direction was changed from a perpendicular to parallel and this was caused by the thermal stress distribution in isotropic microstructure under ELM-like heat pulse. In addition the change of crack growth direction caused the deterioration of thermal conduction near the surface and it led to the local surface melting near cracks even at 0.51MJ/m2 with 3000 heat load cycles.

Impact of combined transient plasma/heat loads on tungsten performance below and above recrystallization temperature

The influence of recrystallization on thermal shock resistance has been identified as an issue that may influence the long term performance of ITER tungsten (W) divertor components. To investigate this issue a unique series of experiments has been performed on ITER divertor W monoblock mock-ups in three EU high heat flux facilities: GLADIS (neutral beam), JUDITH 2 (electron beam) and Magnum-PSI (plasma beam). To simulate ITER mitigated edge localised modes, heat fluxes between 0.11 and 0.6 GW m−2 were applied for Δt < 1 ms. Two different base temperatures, Tbase = 1200 °C and 1500 °C, were chosen on which ~18 000/100 000 transient events were superimposed representing several full ITER burning plasma discharges in terms of number of transients and particle fluence. An increase in roughening for both e-beam and plasma loaded surfaces was observed when loading during or after recrystallization and when loading at higher temperature. However, regarding the formation of cracks and microstructural modifications the response was different for e-beam and plasma loaded surfaces. The samples loaded in Magnum-PSI did not crack nor show any sign of recrystallization, even at Tbase = 1500 °C. This could be a dynamic hydrogen flux effect, because pre-loading of samples with hydrogen neutrals (GLADIS) or without hydrogen (e-beam JUDITH 2) did not yield this result. These results show clearly that the loading method used when investigating and qualifying the thermal shock performance of materials for ITER and future fusion reactors can play an important role. This should be properly accounted for and in fact should be the subject of further R&D.

Interpretation of the deep cracking phenomenon of tungsten monoblock targets observed in high-heat-flux fatigue tests at 20 MW/m2

The HHF qualification tests conducted on the ITER divertor target prototypes showed that the tungsten monoblock armor suffered from deep cracking due to fatigue, when the applied high-heat-flux load approaches 20MW/m2. In spite of the critical implication of the deep cracking of armor on the structural integrity of a whole target component, no rigorous interpretation has been given to date. In this paper, a theoretical interpretation of the observed deep cracking feature is presented. A two-stage modeling approach is employed where deep cracking is thought to be a consecutive process of crack initiation and crack growth, which is assumed to be caused by plastic fatigue and brittle facture, respectively. The fatigue lifetime to crack initiation on the armor surface and the crack tip load of brittle fracture are assessed as a function of crack length and heat flux loads. The potential mechanisms of deep cracking are discussed for a typical slow transient high-heat-flux load cycle. It is shown that the quantitative predictions delivered in this study agree well with the observed findings offering insight into the nature of tungsten armor failure.

Impurity radiation in DEMO systems modelling

For fusion reactors with ITER divertor technology, it will be imperative to significantly reduce the heat flux into the divertor e.g. by seeded impurity radiation. This has to be done without affecting the accessibility of a high performance scenario. To assess the implications of seeded plasma impurities on DEMO design, we have developed an impurity radiation model for radiation inside the separatrix. Evaluating the validity of our model, we find the assumption of a local ionisation equilibrium to be appropriate for our purposes and the assumption of flat impurity profiles – even though not satisfactory – to represent the best currently possible. Benchmarking our model against other codes highlights the need to use up to date atomic loss function data.From the impurity radiation perspective, the main uncertainties in current DEMO design stem from the lack of confinement and L-H-threshold scalings that can be robustly extrapolated to highly radiative DEMO scenarios as well as the lack of appropriate models for the power flow from the separatrix into the divertor that include radiation in the scrape off layer. Despite these uncertainties in the model we can exclude that significant fuel dilution through seeded impurities (with Z≥ZAr) will be an issue for DEMO, but the controllability of highly radiative scenarios still needs to be coherently shown.

Aberrations in preliminary design of ITER divertor impurity influx monitor

Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

A review on two previous divertor target concepts for DEMO: mutual impact between structural design requirements and materials performance

Development of a diverter target with a sufficient capability of power exhaust is a crucial prerequisite for the realization of a fusion power plant. While the design and technology for divertor target has been successfully developed for ITER, the applicability of this concept is not necessarily assured yet for DEMO mainly because the neutron irradiation dose expected for the DEMO divertor will be an order of magnitude higher than that of the ITER divertor. The possible embrittlement of structural heat sink materials due to irradiation is likely to restrict the structural performance and the operational flexibility of a target component to a considerable extent. For judgment of design feasibility of a target concept a quantitative evaluation of the thermal and structure mechanical performance is needed. In this article, a review on two representative target design concepts considered for the DEMO divertor is presented. Emphasis is put on the mutual impact between the design requirements and the performance of structural materials. Water-cooled and helium-cooled concepts are discussed considering two baseline heat sink materials, CuCrZr alloy and tungsten, respectively. Conclusions are derived from the critical features of the heat sink performance in terms of structural reliability, design/material interface and further R&D needs.

The effect of transient temporal pulse shape on surface temperature and tungsten damage

The plasma-facing components (PFCs) in future fusion devices such as ITER will receive intense transient heat fluxes from plasma instabilities, such as edge localized modes (ELMs) and disruptions, which will limit material lifetime. The energy density and transient pulse duration are typically used to characterize the PFC damage threshold. However, these parameters are not sufficient to define a damage threshold, because different transient pulse shapes with the same energy density and same pulse duration produce different peak surface temperatures (and thus stresses, which ultimately determine material damage). The surface temperature and damage of tungsten (the material to be used for the ITER divertor target armor) in the form of surface roughening and melting are investigated using various temporal pulse shapes from an Nd:YAG laser in the PISCES-B facility. The heat flux factor is examined and shown to be an inadequate parameter to characterize the temperature rise except for square temporal pulse shapes. For ELM-like heat pulses, the long tail in the temporal shape results in a lower peak surface temperature and less damage compared to a symmetric triangle pulse with equal energy density.

Morphological and structural effects on tungsten targets produced by fusion plasma pulses from a table top plasma focus

A table top plasma focus device operating at hundreds of joules was used to simulate an equivalent damage factor than the obtained on the divertor in tokamak experiments. Using the ejected plasma produced after the pinch disruptions, the effects on tungsten targets from 50 cumulative plasma shocks with power fluxes per shot between 2.6 and 9200 kW cm−2 and with a duration time in the order of tens of nanoseconds (damage factor in the order of 100–103 (W cm−2)s1/2) were studied. Morphological analysis shows an increasing appearance of cracked surfaces with holes, fissures and defects, suggesting a potential progression of stress effects and a fast heat load that melts the surface, ending in thermal contractions that recrystallize the surface of the target. A structural analysis demonstrates a compressive stress development and suggests that part of the energy is released in the melting of the surface in case of a plasma shock with a power flux of 9.2 MW cm−2, 75 ns duration pulse, 2.5 × 103 (W cm−2)s1/2 damage factor. How to increase the damage factor by one order of magnitude up to the expected value from type I ELMs on the ITER divertor, i.e. 104 (W cm−2)s1/2 is discussed.

Design and integration of lower ports for ITER diagnostic systems

All around the ITER vacuum vessel, forty-four ports will provide access to the vacuum vessel for remote handling operations, diagnostic systems, heating, and vacuum systems: 18 upper ports, 17 equatorial ports, and 9 lower ports. Among the lower ports, three of them will be used for the remote handling installation of the ITER divertor. Once the divertor is in place, these ports will host various diagnostic systems mounted in the so-called diagnostic racks. The diagnostic racks must allow the support and cooling of the diagnostics, extraction of the required diagnostic signals, and providing access and maintainability while minimizing the leakage of radiation toward the back of the port where the humans are allowed to enter. A fully integrated inner rack, carrying the near plasma diagnostic components, will be an stainless steel structure, 4.2m long, with a maximum weight of 10t. This structure brings water for cooling and baking at maximum temperature of 240°C and provides connection with gas, vacuum and electric services. Additional racks (placed away from plasma and not requiring cooling) may be required for the support of some particular diagnostic components. The diagnostics racks and its associated ex vessel structures, which are in its conceptual design phase, are being designed to survive the lifetime of ITER of 20 years. This paper presents the current state of development including interfaces, diagnostic integration, operation and maintenance, shielding requirements, remote handling, loads cases and discussion of the main challenges coming from the severe environment and engineering requirements.

Development of instrumentation and control systems for the ITER diagnostic systems in JADA

Japan Domestic Agency (JADA) is responsible for six diagnostic systems in the ITER project. We have successfully developed a prototype instrumentation and control (I&C) system for the ITER divertor thermocouples system. The prototype I&C system includes a supervisory function, a sequencing management function, and a data acquisition function. The supervisory function, which was implemented using EPICS (Experimental Physics and Industrial Control System), manages internal operations for measurement such as health check of sensors. We developed a conversion tool from flowcharts to EPICS record source codes. The EPICS records generated by the conversion tool are used to trigger each operation step and to indicate the progress of the sequence. We also developed the sequencing management function that coordinates the execution of operation steps. Both the flowchart conversion tool and sequencing management function were developed to reduce designers’ mistakes and to ensure rapid development. The prototype I&C system for the DTC was tested. We confirmed that the I&C system performed satisfying the requirements. This validated design will be applied to other ITER diagnostic systems procured by JADA.