Since the signatures of ITER divertor Procurement Arrangements, material purchases, process qualification as well as manufacturing of full-scale prototypes have progressed. This paper provides a ...brief summary of the ITER divertor materials, the requirements for these materials, and the requirements for manufacturing and inspection of the divertor components. Experiences to be acquired through the prototype manufacturing activities are also discussed.
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.
Budget restrictions have forced the ITER Organization to reconsider the baseline divertor strategy, in which operations would begin with carbon (C) in the high heat flux regions, changing out to a ...full-tungsten (W) variant before the first nuclear campaigns. Substantial cost reductions can be achieved if one of these two divertors is eliminated. The new strategy implies not only that ITER would start-up on a full-W divertor, but that this component should survive until well into the nuclear phase. This paper considers the risks engendered by such an approach with regard to known W plasma-material interaction issues and briefly presents the current status of a possible full-W divertor design.
In order to develop and validate the high performance tungsten monoblock technology, the full-tungsten divertor qualification program was defined. As the first step, small-scale mock-ups were ...manufactured and successfully tested under the required high heat flux loads. The test results demonstrated that the technology is available in Japan and Europe. Post-tests observation of the loaded W monoblocks showed generation of self-castellation – a crack along coolant tube axis. The cause of the self-castellation was discussed and a tungsten material characterization program is being developed with the objective to understand mechanical properties that influence the occurrence of the self-castellation.
Physics basis for the first ITER tungsten divertor Pitts, R.A.; Bonnin, X.; Escourbiac, F. ...
Nuclear materials and energy,
August 2019, 2019-08-00, 2019-08-01, Letnik:
20, Številka:
C
Journal Article
Recenzirano
Odprti dostop
•Reviews the fundamental physics aspects of the first ITER W divertor and defines the required operational lifetime within the Staged Approach.•Uses the ITER divertor SOLPS simulation database to ...establish the target peak heat flux and neutral pressure burning plasma operating domain.•Assesses consequences of narrow SOL heat flux channels, fluid drifts, component shaping and 3D magnetic fields for ELM control.•Uses W recrystallization to define an operational budget and shows that heat fluxes ∼50% higher than previously assumed may be acceptable.•Shows that Ne and N should be equally good as seed impurities and suggests that very strong ELM mitigation will be required at high performance.•Provides a list of key outstanding R&D areas to consolidate the divertor physics basis in the period up to ITER operation.
On the eve of component procurement, this paper discusses the present physics basis for the first ITER tungsten (W) divertor, beginning with a reminder of the key elements defining the overall design, and outlining relevant aspects of the Research Plan accompanying the new “staged approach” to ITER nuclear operations which fixes the overall divertor lifetime constraint. The principal focus is on the main design driver, steady state power fluxes in the DT phases, obtained from simulations using the 2-D SOLPS-4.3 and SOLPS-ITER plasma boundary codes, assuming the use of the low Z seeding impurities nitrogen (N) and neon (Ne). A new perspective on the simulation database is adopted, concentrating purely on the divertor physics aspects rather than on the core-edge integration, which has been studied extensively in the course of the divertor design evolution and is published elsewhere. Emphasis is placed on factors which may increase the peak steady state loads: divertor target shaping for component misalignment protection, the influence of fluid drifts, and the consequences of narrow scrape-off layer heat flux channels. All tend to push the divertor into an operating space at higher sub-divertor neutral pressure in order to remain at power flux densities acceptable for the target material. However, a revised criterion for the maximum tolerable loads based on avoidance of W recrystallization, sets an upper limit potentially ∼50% higher than the previously accepted value of ∼10 MW m−2, a consequence both of the choice of material and the finalized component design. Although the simulation database is currently restricted to the 2-D toroidally symmetric situation, considerable progress is now also being made using the EMC3-Eirene 3-D code suite for the assessment of power loading in the presence of magnetic perturbations for ELM control. Some new results for low input power corresponding to the early H-mode operation phases are reported, showing that even if realistic plasma screening is taken into account, significant asymmetric divertor heat fluxes may arise far from the unperturbed strike point. The issue of tolerable limits for transient heat pulses is an open and key question. A new scaling for ELM power deposition has shown that whilst there may be more latitude for operation at higher current without ELM control, the ultimate limit is likely to be set more by material fatigue under large numbers of sub-threshold melting events.
•The optimized ITER divertor design is presented.•Shaping of vertical target design was validated by 3D field line tracing calculation and thermos-mechanical analysis.•At the monoblock level, 0.5 mm ...deep toroidal bevel was implemented and a reduction of the thickness down to 6 mm was demonstrated to be acceptable.
The shaping of the ITER divertor vertical targets has been refined as a consequence of manufacturing and engineering considerations during the prototype manufacturing activities. In this paper, the optimized ITER divertor design is presented together with design validation by 3D field line tracing calculation and thermo-mechanical analysis by finite element calculations. Furthermore, the reduction of W monoblock armour thickness to 6 mm is also discussed.
During the qualification program of the tungsten divertor vertical targets, the tested mock-ups (250 tested tungsten monoblocks in total) successfully demonstrate their thermal performances and ...structural integrity. However, some of the tested monoblocks, in average 30%, showed macro-cracks in the tungsten.
This paper presents the results of 3D elastic-plastic thermo-mechanical analysis of the tungsten monoblock under stationary thermal loads for two cases of tungsten material properties: (1) stress-relieved and (2) recrystallized. The comparison pointed out that the recrystallized tungsten monoblock accumulated more damages at the loaded than stress-relieved tungsten. In addition, recrystallization may lead to early development of cracks on the monoblock either due to progressive deformation or fatigue while the non-recrystallized monoblock has a much smaller probability to develop cracks, as long as the exposed surfaces are free from defects.
•ASIPP manufactured six W monoblock mock-ups that were tested at IDTF for full-tungsten divertor qualification program.•Ultrasonic test was performed to investigate the defects of ...interface.•Destructive test was performed on three mock-ups to analysis the damage.•FEA was performed to study the temperature and stress distribution of monoblocks.
In 2015, as part of ITER Full-Tungsten Divertor Qualification Program, Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) manufactured six small-scale Tungsten (W) monoblock mock-ups that were tested at the electron beam facility, ITER Divertor Test Facility (IDTF, St Petersburg, RF). The high heat flux (HHF) tests consisted of 5000 cycles at 10MW/m2, followed by 300 cycles at 20MW/m2 and additional 700 cycles at 20MW/m2. All mock-ups fulfilled the requirements of high heat flux performance successfully. One (WTC-5) of the six mock-ups was then selected to carry out critical heat flux (CHF) test with local heat flux up to 37–39MW/m2. After HHF test, ultrasonic test (UT) and destructive tests were performed to characterize the damages of monoblocks. The debonding of the Cu/CuCrZr interface was detected by both nondestructive and destructive tests. Intergranular rupture of W was observed by Scanning Electron Microscope (SEM). The recrystallization was found in the W monoblocks and the recrystallized depth were analyzed by metallography and Vickers hardness measurement.
•Detailed design development plan for the ITER tungsten divertor.•Latest status of the ITER tungsten divertor design.•Brief overview of qualification program for the ITER tungsten divertor and status ...of R&D activity.
In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R&D activity is summarized in this paper.
•Focus is on protection of leading edges between toroidally adjacent monoblocks.•Summarizes conclusions of coordinated, multi-device ITPA Divertor and SOL task.•Leading loading found to be well ...described by optical approximation.•Shaping required to prevent deep edge melting during steady state and ELMs.•Shaping implies reduced parallel heat flux to avoid material crystallization.
The key remaining physics design issue for the ITER tungsten (W) divertor is the question of monoblock (MB) front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM) or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA) and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.
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