Heat Exhaust Capability Research Articles

Acceptance tests of the industrial series manufacturing of WEST ITER-like tungsten actively cooled divertor

The activelly cooled plasma facing units (PFUs) constituting the WEST lower divertor must meet strict technical specifications before their installation into the WEST tokamak. The tests performed at CEA lead mainly: to provide information on the feasibility to attach mechanically PFUs on sectors, to ensure geometrical tolerances for the welding of PFUs to water manifolds, to check the PFU vacuum tightness and to confirm the PFUs heat exhaust capability. Using high heat flux (HHF) test facilities, such as HADES at CEA-Cadarache and GLADIS at IPP-Garching, ∼5% of the PFU production was tested. Infrared thermography (IR) tests were also performed (∼24% of the PFU production tested). We show that PFUs are with a quality in agreement to the requirements and that the assessement of the heat exhaust capability during the series production is needed. Based on statistical approaches, this work also provides information on the methods to assess the quality of tested components using statistic process control.

Improvement on the plasma performances via application of flowing lithium limiters in EAST tokamak

Three generation of flowing liquid lithium (FLiLi) limiters have been developed and successfully tested in EAST since 2014. It was confirmed that liquid Li could be driven by an innovative in-vessel DC electromagnetic J × B pump to form a recirculating loop. The FLiLi design was continuously upgraded, resulting in improvement of the liquid Li surface coverage (>80%), elimination of surface erosion, and the increase of heat exhaust capability. Promising results are obtained during FLiLi operations, including a reduction of impurities and recycling to improve plasma confinement, and a mitigation of ELM activity in H-mode plasmas. The FLiLi limiter program has shown basic compatibility with H-mode discharges with auxiliary heating power up to 8.3 MW. These efforts will expand to flowing liquid Li PFC in high heat flux zones, e.g. near the divertor strike point in reactors.

Thermomechanical simulation of flat tiles mock-ups under high heat flux

In the WEST (W -for tungsten- Environment in Steady-state Tokamak) tokamak, an actively cooled tungsten divertor for testing high heat flux technology is implemented. Beside the presently tested ITER divertor technology (i.e. W-monoblock), the WEST team in collaboration with ASIPP (China) worked on the development of an alternative design for actively cooled W/Cu plasma facing components capable to sustain heat loads close to monoblock design. Within this framework, two small-scale mock-ups based on the W-armoured flat-tile concept were produced by ASIPP and successfully tested in the electron beam facility JUDITH-1 (FZ-Jülich, Germany) under successive thermal cycling at 10, 15 then 20 MW/m². These results showed the capacity of this technology to remove heat loads over several hundreds of cycles at 20 MW/m2 without obvious indication of damage impacting the heat exhaust capability. However, the post-mortem analysis showed crack formation perpendicular to the loaded surface proceeding into the pure Cu interlayer for those tiles tested at 20 MW/m². This contribution focuses on the thermal and structural analysis performed with ANSYS for these components. The convective heat transfer coefficients were calculated using the Nukiyama model and the mechanical properties of pure copper at high temperature were extrapolated using the Ramberg-Osgood model. The thermal results are in good agreement with the experiments. The thermo-mechanical simulation explains partially the defects observed on the mocks up with the selected approach, which leads to suggest improvements for further study by modelling kinematic hardening for tungsten and creep for oxygen-free copper.

High heat flux test results for a thermal break DEMO divertor target and subsequent design and manufacture development

This paper focuses on the development of the water-cooled divertor target concept known as Thermal Break, which was carried out in two phases. In Phase 1, six small scale mock-ups were fabricated and subjected to high heat flux (HHF) testing of up to 25 MW/m2 and thermal cycling of up to 500 cycles at 20 MW/m2. All six mock-ups survived the campaign and maintained 20 MW/m2 heat exhaust capability. Detailed examination of mock-ups was carried out to understand the damage mechanisms. One mock-up, which was tested beyond its design intent at 500 cycles, shows signs of progressive damage. Potential damage modes were identified and influenced subsequent Phase 2 mock-up design. Although there are signs of tungsten surface cracking, the predominant damage mode is not by “deep cracking” but substantial permanent deformation in the interlayer features.Therefore, in Phase 2 the manufacturing procedure was updated, the interlayer grooves were given stress-relieving radii which have significantly reduced the interlayer plastic strain range. Interlayer design parameters were selected following the use of response surface-based design search and optimization. Mock-ups of the Phase 2 design have been manufactured and HHF testing is planned within 2018.

Performance assessment of high heat flux W monoblock type target using thin graded and copper interlayers for application to DEMO divertor

For the development of a divertor for DEMO, the European WPDIV project is underway since 2014. The first phase of the project (2014–2016) aims to provide mock-ups adapted to DEMO operation requirements and the second phase aims to furnish mock-ups, with standardized geometry, which fulfill phase 1 requirements. Within the WPDIV project, several options are under development. One of these aims is to replace the thick copper interlayer, used for ITER divertor components, with a very thin coat (functional gradient material or pure copper) for armor-to-pipe joining. One of the benefits is related to armor temperature which is decreased as the distance of heat conduction path is shortened. Some blocks equipped with thin functional gradient material as interlayer proved, in 2016, to handle high cycling performances without any degradation (no surface change aspect and no decrease of thermal heat exhaust capability). This article gives a brief overview on the recent achievements of the development of thin interlayer concept focusing on the design, mock-up production, inspection, high heat flux (HHF) qualification testing and post-examinations.

Experiments of continuously and stably flowing lithium limiter in EAST towards a solution for the power exhaust of future fusion devices

Liquid lithium (Li) can partly ameliorate lifetime and power-exhaust issues of plasma facing components (PFCs) by enabling a self-healing, self-replenishing surface with a reduced susceptibility to neutron damage in future fusion devices. To assess operational stability and heat-exhaust capability under tokamak exposure, two generations of continuously flowing liquid Li (FLiLi) limiters on the concept of a thin flowing Li film have been successfully designed and tested in high performance discharges in EAST. The design uses a circulating Li layer with a thickness of <0.1 mm and a flow rate ∼2 cm3s−1. In addition, the limiter employs a novel electro-magnetic pump to drive liquid Li flow from a collector at the bottom of the limiter into a distributor at its top. Free surface gravitational flow closes the loop for a continuously flowing liquid Li film on the wetted PFC. Here we summarize key FLiLi limiter development and experimental results in H-mode plasmas.

Effect of induction heating on flaw generation and expansion

During acceptance tests phase of plasma facing components, main performance requirements are checked. For divertor application, it is important to achieve an adequate heat exhaust capability. For ITER divertor, defects localized at tungsten to compliant layer interface are defects impacting such capability. For this reason, acceptance tests, of ITER divertor components, mainly consist in using ultrasonic testing (US) and high heat flux (HHF) tests. If a defect is detected by US tests, it is mandatory to validate that detected defect as no detrimental impact on component heat exhaust capability. At present, HHF tests can assess this performance. In this paper, we describe a test, called STING (thermomechanical STamping by Induction heatiNG), which aims to evaluate, before HHF tests, the heat exhaust evolution of plasma facing components after a moderate thermal solicitation. This paper is focused on the effect of STING solicitation on flaw generation and expansion.

Qualification and post-mortem investigation of actively cooled tungsten flat-tile mock-ups for WEST divertor

The WEST (W – for tungsten – Environment in Steady-state Tokamak) project has provided the opportunity of to developing and qualifying, in collaboration with ASIPP (China) and FZJ (Germany) a newly attractive W/Cu actively cooled PFCs technology able to sustain high heat loads close to ITER divertor ones.Within this framework, ASIPP has produced two small-scale mock-ups based on the W-armoured flat-tile concept. The results presented herein focus on the qualification program, performed by CEA, which consisted in a preliminary non-destructive testing (NDT) control based on infrared thermography to assess the joint interface quality followed by cyclic high heat flux (HHF) testing performed at FZJ. The HHF testing, performed by successive thermal cycling, allowed demonstrating the thermal performance of this technology to remove heat loads over several hundreds of cycles at 20 MW/m2 without obvious indication of damage impacting the heat exhaust capability for both mock-ups.Furthermore, the subsequent post-mortem microstructural analysis allowed confirming the issues regarding the induced fatigue damages, especially the tungsten crack formation perpendicular to the incident surface appeared during cycling at 10 MW/m2 and plastic deformation inducing microstructure evolution in the pure Cu-interlayer during cycling beyond 10 MW/m2 for both mock ups.

Radio-frequency sheaths physics: Experimental characterization on Tore Supra and related self-consistent modeling

During the 2011 experimental campaign, one of the three ion cyclotron resonance heating (ICRH) antennas in the Tore Supra tokamak was equipped with a new type of Faraday screen (FS). The new design aimed at minimizing the integrated parallel electric field over long field lines as well as increasing the heat exhaust capability of the actively cooled screen. It proved to be inefficient for attenuating the radio-frequency (RF)-sheaths on the screen itself on the contrary to the heat exhaust concept that allowed operation despite higher heat fluxes on the antenna. In parallel, a new approach has been proposed to model self-consistently RF sheaths: the SSWICH (Self-consistent Sheaths and Waves for IC Heating) code. Simulations results from SSWICH coupled with the TOPICA antenna code were able to reproduce the difference between the two FS designs and part of the spatial pattern of heat loads and Langmuir probe floating potential. The poloidal pattern is a reliable result that mainly depends on the electrical design of the antenna while the radial pattern is on the contrary highly sensitive to loosely constrained parameters such as perpendicular conductivity that generates a DC current circulation from the private region inside the antenna limiters to the free scrape off layer outside these limiters. Moreover, the cantilevered bars seem to be the element in the screen design that enhanced the plasma potential.

Characterization of heat flux generated by ICRH heating with cantilevered bars and a slotted box Faraday screen

In the framework of the ion cyclotron resonance heating (ICRH) development led at CEA Cadarache, an actively cooled Faraday screen (FS) prototype with cantilevered horizontal bars and a slotted box has been designed to increase the heat exhaust capability (for high-power operation), reduce the parallel RF electric field along long field lines and qualify alternative mechanical solutions for ITER (bars are disconnected from the septum to reduce the stress level). The new FS has been installed on an existing ICRH antenna, and was tested during the 2011 Tore Supra experimental campaign. The antenna hosting the new screen exhibits high sensitivity to the edge plasma condition, some instabilities of electrical matching and improved heat exhaust capabilities in accordance with the thermo-mechanical design. RF-induced heat loads derived from IR thermography have been found to be about five times higher in the equatorial plane with the new design compared with the conventional design. The experimental results show that minimizing the parallel RF electric field along long field lines is not enough to reduce the wave–plasma interaction on the screen. This paper summarizes the experimental RF-induced heat load for several plasma scenarios and edge parameters (plasma current, density and heating power level) with emphasis on RF-sheath rectification and E × B convection generated in front of the antenna through the differential biasing of adjacent field lines.

Physics basis and design of the ITER plasma-facing components

In ITER, as in any tokamak, the first wall and divertor plasma-facing components (PFC) must provide adequate protection of in-vessel structures, sufficient heat exhaust capability and be compatible with the requirements of plasma purity. These functions take on new significance in ITER, which will combine long pulse, high power operation with severe restrictions on permitted core impurity concentrations and which, in addition, will produce transient energy loads on a scale unattainable in today’s devices. The current ITER PFC design has now reached a rather mature stage following the 2007 ITER Design Review. This paper presents the key elements of the design, reviews the physics drivers, essentially thermal load specifications, which have defined the concept and discusses a selection of material and design issues.

Advanced qualification methodology for actively cooled plasma facing components

The use of high heat flux plasma facing components (PFCs) in steady state fusion devices requires high reliability. These components have to withstand heat fluxes in the range 10–20 MW m−2 involving a number of severe engineering constraints. Feedback from the experience of various industrial manufacturings showed that the bonding of the refractory armour material onto the metallic heat sink causes generic difficulties strongly depending on material qualities and specific design. As the heat exhaust capability and lifetime of PFCs during plasma operation are directly linked to the manufacturing quality, a set of qualification activities such as active infrared thermography, lock-in and acoustic measurements were performed during the component development phases following a qualification route. This paper describes the major improvements stemming from better measurement accuracy and refined data processing and analyses recent developments aimed at investigating the capability to qualify the component in situ during its lifetime.

Progress in ergodic divertor operation on Tore Supra

Upgrade of the Tore Supra ergodic divertor (ED) has led to significant progress in ED physics. Pulse durations of 30 s with LHCD have been achieved demonstrating the heat exhaust capability of both the actively cooled technology at hand and of this specific divertor concept. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching two on the magnetic surface q = 2+(3/12). This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q∼2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is associated with an increase of the fluctuating electric drift velocity. This results in an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The three-dimensional response of the temperature field is computed with a fluid code. The code can recover the intrinsic transport barrier at the separatrix, reported experimentally, together with small amplitude temperature modulations in the divertor volume. Experimental evidence for the three density regimes (linear, high recycling and detachment) is reported. The low critical density values for transitions between these regimes indicate that similar parallel physics governs the axisymmetric and ED, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ion cyclotron radiofrequency heating coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of both impurity control and radiation efficiency are reviewed. Global power balance has been analysed in order to account for non-axisymmetric radiation. These results, taken together, confirm the large radiation capability of the ED.