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.
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.
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.
•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.
•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.
This paper is concerned with investigation of an erosion of the ITER-like divertor castellated targets of pure tungsten, lanthanum tungsten and CFC under plasma heat loads expected during the Type I ...ELMs and disruptions in ITER. These experiments were carried out on a plasma gun QSPA-T at the SRC RF TRINITI under EU/RF collaboration. The targets were exposed by series repeated plasma pulses with heat loads in the range of 0.2–2.5
MJ/m
2 and a pulse duration of 0.5
ms. The erosion value as a function of pulse number and energy density were obtained. The erosion of lanthanum tungsten started at the lower energy density as compared with pure tungsten and was mainly due to a melt layer movement and a droplets ejection. Characteristics of ejected droplets were measured. The erosion of CFC macrobrushes under ELM and disruption heat loads was determined mainly by damage of PAN-fibers.
Abstract The success of reproductive technologies is facilitated by the cryopreservation of embryos and gametes. In Italy, where legislation prohibits zygote and embryo cryopreservation, clinics have ...extensively introduced oocyte cryopreservation. Two different strategies of oocyte cryopreservation are available: slow freezing or ultrarapid cooling (vitrification). Although the results are very encouraging with both methods, there is still controversy regarding both the procedure itself and the most suitable method to use. This study reports the routine application of the two different oocyte cryopreservation methods in programmes running in two consecutive periods. The study centre carried out 286 thawing cycles for a total of 1348 thawed oocytes cryopreserved by the slow-freezing method and 59 warming cycles for a total of 285 warmed oocytes cryopreserved by vitrification. Comparison of the outcomes obtained with the slow-freezing method versus vitrification in women who underwent IVF for infertility showed survival, fertilization, pregnancy and implantation rates of 57.9% versus 78.9% ( P < 0.0001), 64.6% versus 72.8% ( P = 0.027), 7.6% versus 18.2% ( P = 0.021) and 4.3% versus 9.3% ( P = 0.043) respectively. These results suggest that oocyte vitrification is associated with a better outcome than the slow-freezing method.
In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the ...design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline.
