Water-cooled Plasma Facing Components Research Articles

Overview of the first Wendelstein 7-X long pulse campaign with fully water-cooled plasma facing components

After a long device enhancement phase, scientific operation resumed in 2022. The main new device components are the water cooling of all plasma facing components and the new water-cooled high heat flux divertor units. Water cooling allowed for the first long-pulse operation campaign. A maximum discharge length of 8 min was achieved with a total heating energy of 1.3 GJ. Safe divertor operation was demonstrated in attached and detached mode. Stable detachment is readily achieved in some magnetic configurations but requires impurity seeding in configurations with small magnetic pitch angle within the edge islands. Progress was made in the characterization of transport mechanisms across edge magnetic islands: Measurement of the potential distribution and flow pattern reveals that the islands are associated with a strong poloidal drift, which leads to rapid convection of energy and particles from the last closed flux surface into the scrape-off layer. Using the upgraded plasma heating systems, advanced heating scenarios were developed, which provide improved energy confinement comparable to the scenario, in which the record triple product for stellarators was achieved in the previous operation campaign. However, a magnetic configuration-dependent critical heating power limit of the electron cyclotron resonance heating was observed. Exceeding the respective power limit leads to a degradation of the confinement.

Observation of thermal events on the plasma facing components of Wendelstein 7-X

Long pulse operation of present and future magnetic fusion devices requires sophisticated methods for protection of plasma facing components from overheating. Typically, thermographic systems are being used to fulfill this task. Steady state operation requires, however, autonomous operation of the system and fully automatic detection of abnormal events. At Wendelstein 7-X (W7-X), a large advanced stellarator, which aims at demonstrating the capabilities of the stellarator line as a future fusion power plant, significant efforts are being undertaken to develop a fully automatic system based on thermographic diagnostics. In October 2018, the first divertor-based experimental campaign has been finished. One of the goals of this operation phase (named OP1.2) was to study the capabilities of the island divertor concept using an uncooled test divertor made of fine-grain graphite tiles. Throughout this campaign, it was possible to test the infrared imaging diagnostic system, which will be used to protect the actively water-cooled plasma facing components (PFCs) during the steady-state operation in the next experimental campaign. An overview of the most relevant thermal events on the PFCs that were detected in OP1.2 using this system are presented. This includes events that limited operation during the campaign, like baffle hot spots and divertor overloads, events that are potentially critical in steady state operation like leading edges, events caused by the ECRH and NBI heating systems (shine-through hot spots and fast particle losses) and other events which are a common source of false alarms like surface layers. The detected thermal events are now part of an important and extensive image database which will be used to further automate the system by means of computer vision and machine learning techniques in preparation for steady-state operation, when the system must be able to detect dangerous events and protect the machine in real-time.

Thermal-hydraulics modeling for prototype testing of the W7-X high heat flux scraper element

Long-pulse operation of the Wendelstein 7-X (W7-X) stellarator experiment is scheduled to begin in 2020. During this operational phase the vessel will be equipped with water-cooled plasma facing components to allow for longer pulse durations. Certain simulated plasma scenarios have been shown to produce heat fluxes that could surpass the technological limits of the divertor target elements during steady-state operation. To reduce the heat load on the target elements, the addition of a “scraper element” (SE) is under investigation. The SE is composed of 24 water-cooled carbon fiber reinforced carbon composite monoblock units. Multiple full-scale prototypes have been tested in the GLADIS high heat flux test facility. Previous computational studies revealed discrepancies between the simulations and experimental measurements. In this work, single-phase thermal-hydraulics modeling was performed in ANSYS CFX to identify potential causes for such discrepancies. Possible explanations investigated were the effects of a non-uniform thermal contact resistance and a potential misalignment of the monoblock fibers. While the difference between the experimental and computational results was not resolved by a non-uniform thermal contact resistance, the computational results provided insight into the potential performance of a W7-X monoblock unit. Circumferential temperature distributions highlighted the expected boiling regions of such a unit. Furthermore, simulations revealed that modest angles of fiber misalignment in the monoblocks result in asymmetries at the unit edges and provide temperature differences similar to the experimental results.

RAMI analysis for EAST upper full tungsten divertor water cooling system

EAST (Experimental Advanced Superconducting Tokamak) device is a D-shaped full superconducting tokamak with actively water cooled plasma facing components. As the EAST project technical risk control and possible failure mitigation approach, RAMI (Reliability, Availability, Maintainability and Inspectability) analysis was required for the upper full tungsten divertor water cooling system, which insures the system to cool the PFCs with high efficient and safety without leakage. At the same time the cooling water could be at suitable pressure, temperature and velocity, water parameters of cooling circuit also could be monitored. A functional breakdown was prepared and all functions are described using the IDEFØ (Integration DEfinition Function – language Ø) method. FMECA (Failure Mode, Effects and Criticality Analysis) was performed to the system initial risk. There are 31 risks for the initial state, including 7 major risks. No major risk remains after taking into account all the actions. RBDs (Reliability Block Diagrams) were prepared to calculate the reliability and availability of each function. The initial availability of upper divertor cooling water system was 93.63%, after optimization the expected availability was 98.56% over 11,520h. It was assessed that the upper full tungsten divertor water cooling system RAMI analysis results meet the EAST project requirement during this design phase and the results will be qualified further when the system design is updated.

FEM-DBEM approach to analyse crack scenarios in a baffle cooling pipe undergoing heat flux from the plasma

Wendelstein 7-X is the world’s largest nuclear fusion experiment of stellarator type, in which a hydrogen plasma is confined by a magnet field generated with external superconducting coils, allowing the plasma to be heated up to the fusion temperature. The water-cooled Plasma Facing Components (PFC) protect the Plasma Vessel (PV) against radiative and convective heat from the plasma. After the assembly process of heat shields and baffles, several cracks were found in the braze and cooling pipes. Due to heat load cycles occurring during each Operational Phase (OP), thermal stresses are generated in the heat sinks, braze root and cooling pipes, capable to drive fatigue crack-growth and, possibly, a water leak through the pipe thickness. The aim of this study is to assess the most dangerous initial crack configurations in one of the most critical baffles by using numerical models based on a FEM-DBEM approach.

Results of high heat flux qualification tests of W monoblock components for WEST

One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m−2. These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected.

Tungsten fibre-reinforced composites for advanced plasma facing components

The European Fusion Roadmap foresees water cooled plasma facing components in a first DEMO design in order to provide enough margin for the cooling capacity and to only moderately extrapolate the technology which was developed and tested for ITER. In order to make best use of the water cooling concept copper (Cu) and copper-chromium-zirconium alloy (CuCrZr) are envisaged as heat sink whereas as armour tungsten (W) based materials will be used. Combining both materials in a high heat flux component asks for an increase of their operational range towards higher temperature in case of Cu/CuCrZr and lower temperatures for W. A remedy for both issues- brittleness of W and degrading strength of CuCrZr- could be the use of W fibres (Wf) in W and Cu based composites. Fibre preforms could be manufactured with industrially viable textile techniques. Flat textiles with a combination of 150/70 µm W wires have been chosen for layered deposition of tungsten-fibre reinforced tungsten (Wf/W) samples and tubular multi-layered braidings with W wire thickness of 50 µm were produced as a preform for tungsten-fibre reinforced copper (Wf /Cu) tubes. Cu melt infiltration was performed together with an industrial partner resulting in sample tubes without any blowholes. Property estimation by mean field homogenisation predicts strongly enhanced strength of the Wf/CuCrZr composite compared to its pure CuCrZr counterpart. Wf /W composites show very high toughness and damage tolerance even at room temperature. Cyclic load tests reveal that the extrinsic toughening mechanisms counteracting the crack growth are active and stable. FEM simulations of the Wf/W composite suggest that the influence of fibre debonding, which is an integral part of the toughening mechanisms, and reduced thermal conductivity of the fibre due to the necessary interlayers do not strongly influence the thermal properties of future components.

Engineering Studies on the EAST Tungsten Divertor

Experimental advanced superconducting tokamak device is a D-shaped full superconducting tokamak with actively water cooled plasma facing components. To achieve long pulse and high βH-mode plasma, new plasma position and shape are calculated and optimized during the campaign of 2013-2014. New divertors are designed and developed to fit the plasma and endure the heat flux up to 10 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . The divertor is International Thermonuclear Experimental Reactor-like. It bases on monoblock and cassette technology and is composed of plasma facing component (PFC) units, cassettes, and support systems. Monoblock structures are just employed on the PFC units of target plates of the divertors, thus W flat tiles are used on the baffles and dome. At the end of monoblocks, end boxes are applied. Cassettes act not only as the supports of the PFC units, but also as manifolds for the cooling channels of the units. The support systems consist of inner support rails, outer support rails, and auxiliary braces. The support systems are installed on the vacuum vessel before cassettes are put on. To dock the cassette with the support system, it is just lifted and pushed forward. To verify and optimize the structure, Research and Development work has carried out for the divertor. The prototypes of PFC units and cassettes are fabricated with different manufacture processes. The total assembly process is also simulated with these prototypes on mockup facility and the vacuum vessel. The results show that the divertor system can be manufactured successfully according to the requirements of drawings and installed conveniently and precisely on the tokamak. All these works certify that the design of the divertor is all right. The batch production of the divertors can be started.

Actively Water-Cooled Plasma Facing Components of the Wendelstein 7-X Stellarator

The actively water-cooled plasma facing components (PFCs) of the Wendelstein 7-X stellarator consisting of the first wall protection and the divertor systems have a total surface area of about 265m2. The complex 3D geometry of the plasma and plasma vessel with 244 vessel ports dedicated to diagnostics, heating systems and water-cooling pipe-work together with the need to minimize the space taken and the significant heat loads expected on the components presents significant design and manufacturing challenges.The actively water-cooled divertor, made of 100 target modules, has an area of 19 m2. Each target module is formed from target elements made of CFC flat tiles bonded with the bi-layer technology to CuCrZr heat sinks. In total 16,000 tiles are bonded to the 890 target elements. A full-scale target module prototype has been manufactured to validate the design, the selected technological solutions and the inspection methods to be used in the serial module fabrication.About 30% of the target elements have been delivered and the production of the remaining elements should be completed by 2014. The fabrication of the components of the first wall protection, 320 stainless steel panels and 170 heat shields, is almost completed.

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling.This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 °C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

Design and technological solutions for the plasma facing components of WENDELSTEIN 7-X

The operation of W7-X stellarator for pulse length up to 30 min with 10 MW input power requires a full set of actively water-cooled plasma facing components. From the lower thermally loaded area of the wall protection system designed for an averaged load of 100 kW/m 2 to the higher loaded area of the divertor up to 10 MW/m 2, various design and technological solutions have been developed meeting the high load requirements and coping with the restricted available space and the particular 3D-shaped geometry of the plasma vessel. 80 ports are dedicated alone to the water-cooling of plasma facing components and a complex networking of kilometers of pipework will be installed in the plasma vessel to connect all components to the cooling system. An advanced technology was developed in collaboration with industry for the target elements of the high heat flux (HHF) divertor, the so-called “bi-layer” technology for the bonding of flat tiles made from CFC NB31 onto the CuCrZr cooling structure. The design, R&D and the adopted technological solutions of plasma facing components are presented. At present, except the HHF divertor, most of plasma facing components has been already manufactured.

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Lower Hybrid Current Drive in Tore Supra

Since the mission of Tore Supra is to produce quasi-steady-state discharges, the lower hybrid current drive (LHCD) system constitutes the most important method of additional heating and noninductive current drive. A description of the LHCD system is given, including the different launcher designs developed for the Tore Supra long-pulse program. Following the completion of the Composants Internes et Limiteur project, together with the installation of a high-performance LHCD launcher, world record discharges, injected and extracted energy exceeding 1 GJ, were obtained in 2003. With the flexibility of lower hybrid (LH) waves to tailor the current profile, an enhanced performance regime, the so-called LHEP, has been maintained in quasi-steady-state discharges. Detailed measurements of the fast electron distribution have allowed us to constrain LHCD ray-tracing models and to quantify parametric dependencies describing the fast electron tail. Localized heat loads on the LHCD launchers due to interaction with fast particles have been measured and quantified, using infrared imaging and calorimetric measurements on water-cooled plasma facing components. Furthermore, experimental results in the area of LH wave coupling are presented.

Tore-Supra infrared thermography system, a real steady-state diagnostic

Tore-Supra Tokamak ( I p = 1.5 MA, B t = 4 T) has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components (PFCs) for high-performance long pulse plasma discharges. When not actively cooled, plasma facing components can only accumulate a limited amount of energy since the temperature increases continuously during the discharge until radiation cooling equals the incoming heat flux. Such an environment is found in the JET Tokamak [JET Team, IAEA-CN-60/A1–3, Seville, 1994] and on TRIAM [M. Sakamoto, H. Nakashima, S. Kawasaki, A. Iyomasa, S.V. Kulkarni, M. Hasegawa, E. Jotaki, H. Zushi, K. Nakamura, K. Hanada, S. Itoh, Static and dynamic properties of wall recycling in TRIAM-1M, J. Nucl. Mater. 313–316 (2003) 519–523] [Y. Kamada, et al., Nucl. Fusion 3 (1999) 1845]. In Tore-Supra, the surface temperature of the actively cooled plasma facing components reach steady state within a second. We present here the Tore-Supra thermographic system, made of seven endoscope bodies equipped so far with eight infrared (IR) cameras. It has to be noted that this diagnostic is the first diagnostic to be actively cooled, as required for steady state. The main purpose of such a diagnostic is to prevent the plasma to damage the actively cooled plasma facing components (ACPFCs), which consist of the toroidal pumped limiter (TPL), 7 m 2, and of five radio-frequency antennae, 1.5 m 2 each.

Infrared surface temperature measurements for long pulse operation, and real time feedback control in Tore-Supra, an actively cooled Tokamak

Tore-Supra is a Tokamak (Ip=1.5MA, Bt=4T) aiming at doing researches in the field of controlled nuclear fusion. It has been constructed with a steady-state magnetic field using super-conducting magnets and water-cooled plasma facing components for high performances long pulse plasma discharges. When not actively cooled, plasma-facing components can only accumulate a limited amount of energy since the temperature increase continuously (T˜sqrt(time)) during the discharge until radiation cooling is equal to the incoming heat flux (T>1800K). Such an environment is found in the most today Tokamaks. In the present paper we report the recent results of Tore-Supra, especially the design of the new generation of infrared endoscopes to measure the surface temperature of the plasma facing components. The Tore-Supra system is described in details, the new JET infrared thermography system is presented and some insight of the ITER set of visible/infrared endoscope is given.

Water-cooled target modules for steady-state operation of the W7-X divertor

The stellarator WENDELSTEIN 7-X (W7-X) includes water-cooled plasma facing components (PFCs) to allow steady-state operation and to provide an efficient particle and power exhaust up to 10 MW for a maximum pulse duration of 30 min. Ten divertor units are arranged along the helical edge of the fivefold periodic plasma column. The three-dimensional shape and positioning of the target surfaces are optimized to address physics issues for a wide range of experimental parameters, which influence the topology of the boundary. The three-dimensional target surfaces are reproduced by a series of consecutive plane target elements as a set of parallel water-cooled elements positioned onto the frameworks of target modules. The design and arrangement of target modules and elements are described.

Ten years of maintenance on Tore Supra actively cooled components

Tore Supra, the only superconducting Tokamak equipped of actively water cooled plasma facing components, has now completed 10 years of operation (1988–1998). The associated primary pressurised water cooling loop (4 MPa at 150°C) is able to exhaust a 20 MW convected or radiated power during long pulses of 200 s. The first part of this paper is devoted to an inventory and analysis of the water leaks from the plasma facing and their consequences, such as the optimisation of the pressurised water loop emptying sequence. The experience gained during 10 years by maintenance operations has also led to an improvement of the water leak detection procedure, in order to identify as accurately as possible the location of the leak. The second part of the paper describes the improvements made on quality control of Plasma Facing Components during these last 10 years. Quality Control has been increased right through the technological developments and the manufacturing process of mock-up, prototypes and series production. A specific helium tightness procedure dedicated to water cooled components has been established, consists mainly of several validation tests under pressure and temperature operating conditions. Hence reliability of the components has been largely increased.