The WEST project: Current status of the ITER-like tungsten divertor

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units.The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.     

The WEST project: PFC shaping solutions investigated for the ITER-like W divertor

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST (W Environment for Steady state Tokamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m².     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.     

The WEST project: Testing ITER divertor high heat flux component technology in a steady state tokamak environment

The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20 MW/m² range and ITER-like fluences (1000 s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.     

Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material

The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m⁻². Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m⁻². An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m⁻². Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design.     

Optimizing the overall configuration of a He-cooled W-alloy divertor for a power plant

A number of different He-cooled divertor configurations have been proposed for magnetic fusion energy (MFE) power plant application. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. All these designs utilize tungsten or tungsten alloy as structural material. This paper considers the characteristics of the different divertor configurations and proposes the possibility of optimizing the design by combining different configurations in an integrated design based on the anticipated divertor heat flux profile.     

Thermo-mechanical analysis of a DEMO divertor cooling finger under the EFREMOV test conditions

A divertor component of the forthcoming DEMO fusion reactor should be able to withstand heat flux loads larger than 10 MW/m². Successful design should withstand high flux loads for a number of load cycles since initially the DEMO reactor is expected to operate in a non-steady-state mode. Computations for evaluating the structural response of the divertor published so far have, however, been based on the stationary approach. A combined computational fluid dynamics and structural model for evaluating the structural response of a divertor under the non-stationary load conditions is therefore developed in this work. Heat transfer coefficients between the helium and inner surface of the thimble are calculated first by solving the helium steady-state flow equations. Spatially distributed heat transfer coefficients are then used as a boundary condition in a non-stationary thermo-mechanical analysis of the divertor. This analysis is performed for a number of load cycles under different surface heat flux levels. The model is validated against the EFREMOV test experimental conditions, designed to be close to reactor operation conditions. Good agreement of the highest temperatures on the tile’s top surface with the experimental data is obtained. The results suggest that there are three critical regions in the design where damage could initialize: (a) the thimble’s inner surface with the highest thermal gradients, (b) the tile’s outer surface and (c) the filler layer of the brazed tile-thimble joint where the temperature is higher than permissible. Post-examination data of experimental specimen confirm these conclusions as cracks were observed in the above mentioned areas (a) and (b), while melting of the layer (c) was also observed.     

Thermal control of the liquid lithium divertor for NSTX

The liquid lithium divertor (LLD) to be installed in NSTX has four toroidal panels, each a conical section inclined at 22° like the previous graphite divertor tiles. Each LLD panel is a copper plate clad with 0.25 mm of stainless steel (SS) and a surface layer of flame sprayed molybdenum (Mo) that will host lithium deposited from an evaporator. LITER (evaporators) already used in NSTX will be upgraded for the LLD. Each has twelve 500 W cartridge heaters with thermocouples, 16 other thermocouples, and a channel for helium cooling. During LLD experiments, the LLD will be heated so that the lithium is just above its melting temperature. The length of each shot will be preset to prevent excessive evaporation of lithium from the LLD. This duration depends on the heat load and is likely to be in the range of less than a second to several seconds. Careful thermal control of the LLD is important to maximize the shot times and to guide operation of the LLD. This paper describes the layout of the LLD, its expected thermal performance, the control system, and supporting experiments and analysis. A companion paper in this conference, “Physics design requirements for the national spherical torus experiment liquid lithium divertor,” provides other information.     

Upgrades of the tore Supra water cooling system for the WEST project

In ITER, it is foreseen to use an actively-cooled tungsten (W) divertor likely from the beginning of operation. This Plasma Facing Component (PFC) will be subjected to high energy deposition during the plasma operations that severely limit component lifetime. Tungsten has been less extensively studied in tokamaks than carbon. Unlike most present day short pulse fusion devices, Tore Supra is able to reach the ITER pulse length and provide relevant discharges conditions for tungsten PFCs technology validation.A new upgrade of the machine aiming at testing a W divertor under the steady state heat fluxes is being studied in the framework of the WEST (Tungsten (W) Environment in Steady-state Tokamak) project.As the PFC requirements will change, the PFC Primary Heat transfer System (PHTS) must be upgraded. And, even if the injected power in the plasma will not be significantly increased for the WEST project, an upgrade of the Heat Rejection System (HRS) is also needed. This paper presents the studies carried out for these upgrades and the technical solutions to be implemented.     

Design and concept validation of the new solid tungsten divertor for ASDEX Upgrade

Div-III, a divertor with solid tungsten target tiles for ASDEX Upgrade is designed and tested and will be installed in 2013. It is a further step in exploring tungsten as material for plasma facing components. It avoids the restrictions of tungsten coatings on graphite and realizes an operation range up to 50 MJ energy removing capability in the outer divertor. In addition, it allows physics investigation such as erosion and deuterium retention as well as effects of castellation and target tilting. The design of the target itself and the attachment was optimized with FE-analysis and was intensively high heat tested up to a double overload. Cyclic tests reveal that the target and the attachment can be operated with the design load of 50 MJ without any damage. Even a twofold overload results in local recrystallization and minor cracks but the targets did not fail during operation. The redesign of the divertor structure was used to increase the conductance between the cryo-pump and the divertor region. The impact of the changed pumping efficiency was investigated with SOLPS/Eirene modeling. The modeling results are an indication for an easier access to lower SOL densities as expected for a higher pumping efficiency in the main chamber.     

Design optimisation of the ITER divertor magnetic probes using FEM analyses

This paper is focused on the design, simulation and optimisation of the ITER divertor magnetic tangential coils. The most critical issue for the divertor coils is to minimise RITES [G. Vayakis, et al., Radiation-induced thermoelectric sensitivity (RITES) in ITER prototype magnetic sensors, Rev. Sci. Instrum. 75 (10) (2004) 4324-4327] and TIEMF [R. Vila, E.R. Hodson, Thermally induced EMF in unirradiated MI cables, J. Nucl. Mater. 367-370 (Part 2) (2007) 1044-1047] by combining a proper choice of conductor with low temperature variation in the coil. Instead of mineral insulated cable (MIC), which was foreseen as the preferred winding, a winding made of ceramic-coated steel wire was recently proposed [G. Chitarin, L. Grando, S. Peruzzo, C. Tacconet, Design developments for the ITER in-vessel equilibrium Halo current sensors, 24th SOFT Conference, Warsaw, Poland, September 2006, Fusion Eng. Design, in press]. It is thought that, for this wire, maintaining a temperature variation in the wiring below 10 K will be sufficient to allow long-pulse operation. Variations of the divertor coil design have been simulated with the help of ANSYS. The aim was to keep the temperature variation in the winding pack within this limit. The optimisation of the coil based only on a cooling by conduction was not sufficient to meet the 10 K target. Therefore, an actively water-cooled coil was designed which finally met these requirements.     

Design and performance study of the helium-cooled T-tube divertor concept

The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m², and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also.     

Design and analysis of the W7-X divertor scraper element

Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m². However, the edge elements of this divertor are limited to only 5 MW/m², and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions.     

Mechanical analysis of the joint between Wendelstein 7-X target elements and the divertor frame structure

The target elements of the actively cooled high heat flux (HHF) divertor of Wendelstein 7-X are made of CFC (carbon fiber-reinforced carbon composite) tiles bonded to a CuCrZr heat sink and are mounted onto a support frame. During operation, the power loading will result in the thermal expansion of the target elements. Their attachment to the support frame needs to provide, on the one hand, enough flexibility to allow some movement to release the induced thermal stresses and, on the other hand, to provide enough stiffness to avoid a misalignment of one target element relative to the others. This flexibility is realized by a spring element made of a stack of disc springs together with a sliding support at one of the two or three mounting points. Detailed finite element calculations have shown that the deformation of the heat sink leads to some non-axial deformation of the spring elements. A mechanical test was performed to validate the attachment design under cyclic loading and to measure the deformations typical of the expected deformation of the elements. The outcome of this study is the validation of the design selected for the attachment of the target elements, which survived experimentally the applied mechanical cycling which simulates the thermal cycling under operation.     

Performance of electro-plated and joined components for divertor application

A general challenge in divertor development, independently of design type and cooling medium water or helium, is the reliable and adapted joining of components. Depending on the design variants, the characteristics of the joints will be focused on functional or structural behavior to guarantee e.g. good thermal conductivity and sufficient mechanical strength. All variants will have in common that tungsten is the plasma facing material. Thus, material combinations to be joined will range from Cu base over steel to tungsten. Especially tungsten shows lacks in adapted joining due to its metallurgical behavior ranging from immiscibility over bad wetting up to brittle intermetallic phase formation. Joining assisted by electro-chemical deposition of functional and filler layers showed that encouraging progress was achieved in wetting applying nickel interlayers. Nickel proved to be a good reference material but alternative elements (e.g. Pd, Fe) may be more attractive in fusion to manufacture suitable joints.Replacing of Ni as activator element by Pd for W/W or W/steel joints was achieved and joining with Cu-filler was successfully performed. Manufactured joints were characterized applying metallurgical testing and SEM/EDX analyses demonstrating the applicability of Pd activator. First shear tests showed that the joints exhibit mechanical stability sufficient for technical application.     

FEM investigation and thermo-mechanic tests of the new solid tungsten divertor tile for ASDEX Upgrade

A new solid tungsten divertor for the fusion experiment ASDEX Upgrade is under construction at present. A new divertor tile design has been developed to improve the thermal performance of the current divertor made of tungsten coated fine grain graphite. Compared to thin tungsten coatings, divertor tiles made of massive tungsten allow to extend the operational range and to study the plasma material interaction of tungsten in more detail. The improved design for the solid tungsten divertor was tested on different full scale prototypes with a hydrogen ion beam. The influence of a possible material degradation due to thermal cracking or recrystallization can be studied. Furthermore, intensive Finite Element Method (FEM) numerical analysis with the respective test parameters has been performed. The elastic–plastic calculation was applied to analyze thermal stress and the observed elastic and plastic deformation during the heat loading. Additionally, the knowledge gained by the tests and especially by the numerical analysis has been used to optimize the shape of the divertor tiles and the accompanying divertor support structure.This paper discusses the main results of the high heat flux tests and their numerical simulations. In addition, results from some special structural mechanic analysis by means of FEM tools are presented. Finally, first results from the numerical lifecycle analysis of the current tungsten tiles will be reported.     

Evaluation of heat transfer by sublimation for the application to the divertor heat sink for high fusion energy conversion

Thermal and structural responses of divertor target were evaluated by using finite element method. High heat flux simulating ELMs at the level of 100 MW/m² was assumed onto the tungsten armor, and surface temperature profile was obtained. When dynamic heat load over 100 MW/m² was applied, the maximum surface temperature exceeded 1300 °C, and it caused recrystallization of tungsten regardless of the heat transfer below it. The result was used to conduct dynamic heat load experiment on tungsten, and material behavior of tungsten was evaluated under dynamic heat load. This study also proposed new concept of divertor heat sink which can distribute high heat flux and transfers the heat to high temperature medium. It consists of tungsten armor, composite enhanced with high thermal conductivity fiber, and heat transport system applying phase transition. High heat flux simulating ELMs was also applied to target surface of the divertor, temperature gradient, thermal stress of tungsten and composite were evaluated. Based on the results of analysis, thermal structural requirement was considered.     

Geometry and expected performance of the solid tungsten outer divertor row in JET

At JET new plasma-facing components for the main chamber wall and the divertor are being designed and built to mimic the expected ITER plasma wall conditions in the deuterium-tritium operation phase. The main wall elements at JET will be made of beryllium and the divertor plasma-facing surface will be made of tungsten. Most of the divertor tiles will consist of tungsten-coated Carbon Fibre Composite (CFC) material. However one toroidal row in the outer divertor will be made of solid, inertially cooled tungsten. The geometry of these solid tungsten divertor components is optimized within the boundary conditions of the interfaces and the constraints given by the electrodynamical forces. Shadowing calculations as well as rough field line penetration analysis is used to define the geometry of the tungsten lamella stacks. These calculations are based on a set of magnetic equilibria reflecting the operation domain of current JET plasma scenarios. All edges in poloidal and toroidal direction are shadowed to exclude near perpendicular field line impact. In addition, the geometry of the divertor structure is being optimized so that the fraction of the plasma wetted surface is maximised. On the basis of the optimized divertor geometry, performance calculations are done with the help of ANSYS to assess the maximum power exhaust possible with this inertially cooled divertor row.