Zeitschriftenartikel: „Heat loads on the divertor“

1

Barr, William L., und B. Grant Logan. „A Slot Divertor for Tokamaks with High Divertor Heat Loads“. Fusion Technology 18, Nr. 2 (September 1990): 251–56. http://dx.doi.org/10.13182/fst90-a29297.

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2

Marki, J., R. A. Pitts, J. Horacek und D. Tskhakaya. „ELM induced divertor heat loads on TCV“. Journal of Nuclear Materials 390-391 (Juni 2009): 801–5. http://dx.doi.org/10.1016/j.jnucmat.2009.01.212.

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3

Herrmann, A. „Overview on stationary and transient divertor heat loads“. Plasma Physics and Controlled Fusion 44, Nr. 6 (29.05.2002): 883–903. http://dx.doi.org/10.1088/0741-3335/44/6/318.

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4

Riccardo, V., P. Andrew, L. C. Ingesson und G. Maddaluno. „Disruption heat loads on the JET MkIIGB divertor“. Plasma Physics and Controlled Fusion 44, Nr. 6 (29.05.2002): 905–29. http://dx.doi.org/10.1088/0741-3335/44/6/319.

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5

Mavrin, Aleksey A., und Andrey A. Pshenov. „Tolerable Stationary Heat Loads to Liquid Lithium Divertor Targets“. Plasma 5, Nr. 4 (15.11.2022): 482–98. http://dx.doi.org/10.3390/plasma5040036.

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An 0D model is proposed that makes it possible to estimate the limiting stationary heat loads to the targets covered with liquid lithium (LL) layer, taking into account the effects of vapor shielding by sputtered and evaporated LL and hydrogen recycling. Several models of cooled target substrates are considered in which the LL layer facing the plasma is placed. For the considered substrate models, a parametric analysis of the tolerable stationary heat loads to the target on the substrate thickness, the effective cooling energy per particle of sputtered lithium, and the lithium prompt redeposition factor was carried out. It is shown that, at a small substrate thickness, the choice of the substrate model has a significant impact on the tolerable heat loads. It is also shown that even at unrealistically large values of the effective cooling energy, the dissipation of lithium remains modest. This means that in regimes with a high power coming from the core plasma to the edge, the injection of an additional radiator is required. Finally, it is shown that one of the most effective ways to increase the tolerable stationary heat loads would be to reduce the thickness of the target substrate.

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6

Dai, S. Y., D. F. Kong, V. S. Chan, L. Wang, Y. Feng und D. Z. Wang. „EMC3–EIRENE simulations of neon impurity seeding effects on heat flux distribution on CFETR“. Nuclear Fusion 62, Nr. 3 (01.03.2022): 036019. http://dx.doi.org/10.1088/1741-4326/ac47b5.

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Abstract The numerical modelling of the heat flux distribution with neon impurity seeding on China fusion engineering test reactor has been performed by the three-dimensional (3D) edge transport code EMC3–EIRENE. The maximum heat flux on divertor targets is about 18 MW m−2 without impurity seeding under the input power of 200 MW entering into the scrape-off layer. In order to mitigate the heat loads below 10 MW m−2, neon impurity seeded at different poloidal positions has been investigated to understand the properties of impurity concentration and heat load distributions for a single toroidal injection location. The majority of the studied neon injections gives rise to a toroidally asymmetric profile of heat load deposition on the in- or out-board divertor targets. The heat loads cannot be reduced below 10 MW m−2 along the whole torus for a single toroidal injection location. In order to achieve the heat load mitigation (<10 MW m−2) along the entire torus, modelling of sole and simultaneous multi-toroidal neon injections near the in- and out-board strike points has been stimulated, which indicates that the simultaneous multi-toroidal neon injections show a better heat flux mitigation on both in- and out-board divertor targets. The maximum heat flux can be reduced below 7 MW m−2 on divertor targets for the studied scenarios of the simultaneous multi-toroidal neon injections.

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7

Hassanein, Ahmed. „Analysis of sweeping heat loads on divertor plate materials“. Journal of Nuclear Materials 191-194 (September 1992): 499–502. http://dx.doi.org/10.1016/s0022-3115(09)80095-0.

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8

Gunn, J. P., S. Carpentier-Chouchana, F. Escourbiac, T. Hirai, S. Panayotis, R. A. Pitts, Y. Corre et al. „Surface heat loads on the ITER divertor vertical targets“. Nuclear Fusion 57, Nr. 4 (08.03.2017): 046025. http://dx.doi.org/10.1088/1741-4326/aa5e2a.

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9

Abrams, T., M. A. Jaworski, J. Kallman, R. Kaita, E. L. Foley, T. K. Gray, H. Kugel, F. Levinton, A. G. McLean und C. H. Skinner. „Response of NSTX liquid lithium divertor to high heat loads“. Journal of Nuclear Materials 438 (Juli 2013): S313—S316. http://dx.doi.org/10.1016/j.jnucmat.2013.01.057.

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10

HASSANEIN, A. „Analysis of sweeping heat loads on divertor plate materials*1“. Journal of Nuclear Materials 191-194 (September 1992): 499–502. http://dx.doi.org/10.1016/0022-3115(92)90815-3.

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11

Hogan, J. T., und J. Wesley. „Scaling of Divertor Temperature and Heat Loads for TPX-Class Devices“. Fusion Technology 21, Nr. 3P2A (Mai 1992): 1406–15. http://dx.doi.org/10.13182/fst92-a29919.

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12

Gao, Y., Marcin W. Jakubowski, Peter Drewelow, Fabio Pisano, Aleix Puig Sitjes, Holger Niemann, Adnan Ali und Barbara Cannas. „Methods for quantitative study of divertor heat loads on W7-X“. Nuclear Fusion 59, Nr. 6 (26.04.2019): 066007. http://dx.doi.org/10.1088/1741-4326/ab0f49.

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13

Scarabosio, A., C. Fuchs, A. Herrmann und E. Wolfrum. „ELM characteristics and divertor heat loads in ASDEX Upgrade helium discharges“. Journal of Nuclear Materials 415, Nr. 1 (August 2011): S877—S880. http://dx.doi.org/10.1016/j.jnucmat.2010.10.062.

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14

Xi, Ya, Gaoyong He, Xiang Zan, Kang Wang, Dahuan Zhu, Laima Luo, Rui Ding und Yucheng Wu. „Characterization of the Crack and Recrystallization of W/Cu Monoblocks of the Upper Divertor in EAST“. Applied Sciences 13, Nr. 2 (05.01.2023): 745. http://dx.doi.org/10.3390/app13020745.

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The microstructure of and damage to the upper divertor components in EAST were characterized by using metallography, EBSD, and SEM. Under the synergistic effect of heat load and plasma irradiation, cracking, recrystallization, and interface debonding were found in the components of the upper divertor target. The crack propagates downward from the heat loading surface along the heat flux direction, and the crack propagation mode is an intergranular fracture. The thermal loads deposited on the edge of monoblocks raise the temperature higher than the recrystallization temperature of pure tungsten, and the microstructure changes from being in a rolled state to being recrystallized. Additionally, cracks exist in both recrystallized and rolled areas. EBSD boundary maps show that the range of the recrystallization area is determined via the heat flux distribution. The Cu/CuCrZr interface of the cooling components near the thermal loading area is debonded, and the structural integrity is destroyed.

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15

Araki, M., K. Kitamura, K. Urata und S. Suzuki. „Analyses of divertor high heat-flux components on thermal and electromagnetic loads“. Fusion Engineering and Design 42, Nr. 1-4 (September 1998): 381–87. http://dx.doi.org/10.1016/s0920-3796(97)00180-4.

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16

Gunn, J. P., S. Carpentier-Chouchana, R. Dejarnac, F. Escourbiac, T. Hirai, M. Komm, A. Kukushkin, S. Panayotis und R. A. Pitts. „Ion orbit modelling of ELM heat loads on ITER divertor vertical targets“. Nuclear Materials and Energy 12 (August 2017): 75–83. http://dx.doi.org/10.1016/j.nme.2016.10.005.

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17

Hong, Suk–Ho, Richard A. Pitts, Hyung-Ho Lee, Eunnam Bang, Chan-Soo Kang, Kyung-Min Kim und Hong-Tack Kim. „Inter-ELM heat loads on tungsten leading edge in the KSTAR divertor“. Nuclear Materials and Energy 12 (August 2017): 1122–29. http://dx.doi.org/10.1016/j.nme.2017.02.005.

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18

Carli, S., R. A. Pitts, X. Bonnin, F. Subba und R. Zanino. „Effect of strike point displacements on the ITER tungsten divertor heat loads“. Nuclear Fusion 58, Nr. 12 (11.10.2018): 126022. http://dx.doi.org/10.1088/1741-4326/aae43f.

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19

Li, Muyuan, Francesco Maviglia, Gianfranco Federici und Jeong-Ha You. „Sweeping heat flux loads on divertor targets: Thermal benefits and structural impacts“. Fusion Engineering and Design 102 (Januar 2016): 50–58. http://dx.doi.org/10.1016/j.fusengdes.2015.11.026.

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20

Oka, Kiyoshi, Satoshi Kakudate, Nobukazu Takeda, Yuji Takiguchi und Kentaro Akou. „Measurement and Control System for ITER Remote Maintenance Equipment“. Journal of Robotics and Mechatronics 10, Nr. 2 (20.04.1998): 139–45. http://dx.doi.org/10.20965/jrm.1998.p0139.

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ITER in-vessel components such as blankets and divertors are categorized as scheduled maintenance components because they are subjected to severe plasma heat and particle loads. Blanket maintenance requires remote handling equipment and tools able to handle Heavy payloads of about 4 tons within a 2mm precision tolerance. divertor maintenance requires remote replacement of 60 cassettes with a dead weight of about 25 tons each. In the ITER R&D program, full-scale remote handling equipment for blanket and divertor maintenance has been designed and assembled for demonstration tests. This paper reviews the measurement and control system developed for full-scale remote handling equipment, the Japan Home Team contribution.

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21

Takeda, Nobukazu, Kiyoshi Oka, Kentaro Akou und Yuji Takiguchi. „Development of Divertor Remote Maintenance System“. Journal of Robotics and Mechatronics 10, Nr. 2 (20.04.1998): 88–95. http://dx.doi.org/10.20965/jrm.1998.p0088.

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The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R&D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.5. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team.

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22

Riccardi, B., P. Gavila, R. Giniatulin, V. Kuznetsov, R. Rulev, N. Klimov, D. Kovalenko, V. Barsuk, V. Koidan und S. Korshunov. „Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs“. Fusion Engineering and Design 88, Nr. 9-10 (Oktober 2013): 1673–76. http://dx.doi.org/10.1016/j.fusengdes.2013.05.016.

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23

Noce, Simone, Davide Flammini, Pasqualino Gaudio, Michela Gelfusa, Giuseppe Mazzone, Fabio Moro, Francesco Romanelli, Rosaria Villari und Jeong-Ha You. „Neutronics Assessment of the Spatial Distributions of the Nuclear Loads on the DEMO Divertor ITER-like Targets: Comparison between the WCLL and HCPB Blanket“. Applied Sciences 13, Nr. 3 (29.01.2023): 1715. http://dx.doi.org/10.3390/app13031715.

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The Plasma Facing Components (PFCs) of the divertor target contribute to the fundamental functions of heat removal and particle exhaust during fusion operation, being subjected to a very hostile and complex loading environment characterized by intense particles bombardment, high heat fluxes (HHF), varying stresses loads and a significant neutron irradiation. The development of a well-designed divertor target, which represents a crucial step in the realization of DEMO, needs the assessment of all these loads as accurately as possible, to provide pivotal data and indications for the design and structural performance prediction of the PFCs. In a particular way, this study is fully devoted to the comprehension of the distributions on the divertor target of the main nuclear loads due to neutron irradiation, performed for the first time using an extremely detailed approach. This work has been carried-out considering the latest configuration of the DEMO reactor, including the updated design of the divertor and ITER-Like PFCs geometry, varying the blanket layout (Water Cooled Lithium Lead—WCLL and Helium Cooled Pebble Bed—HCPB), thus evaluating the impact of the different blanket concept on the above-mentioned distributions. Neutronics analyses have been performed with MCNP5 Monte Carlo code and JEFF3.3 nuclear data libraries. 3D DEMO MCNP models have been created, focusing in particular on a thorough representation of the divertor and PFCs, allowing for the assessment of the distributions of the main nuclear loads: radiation damage (dpa/FPY), He-production rate (appm/FPY) and nuclear heating density (W/cm3) and total nuclear power deposition (MW). These results are presented by means of 2D maps and plots for each PFCs sub-component both for WCLL and HCPB blanket case: W-monoblocks, Cu-interlayers\CuCrZr-pipe and PFC-CB (Cassette Body) supports made of Eurofer steel.

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24

Sizyuk, V., und A. Hassanein. „Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities“. Physics of Plasmas 22, Nr. 1 (Januar 2015): 013301. http://dx.doi.org/10.1063/1.4905632.

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25

Hayashi, Y., M. Kobayashi, K. Mukai, S. Masuzaki und T. Murase. „Divertor heat load distribution measurements with infrared thermography in the LHD helical divertor“. Fusion Engineering and Design 165 (April 2021): 112235. http://dx.doi.org/10.1016/j.fusengdes.2021.112235.

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26

Miloshevskii, G. V., und G. S. Romanov. „Evaluation of Heat Loads in Graphite Divertor Plates Acted by a Magnetized Electron Flux“. Heat Transfer Research 33, Nr. 7-8 (2002): 9. http://dx.doi.org/10.1615/heattransres.v33.i7-8.60.

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27

Silburn, S. A., G. F. Matthews, C. D. Challis, D. Frigione, J. P. Graves, M. J. Mantsinen, E. Belonohy et al. „Mitigation of divertor heat loads by strike point sweeping in high power JET discharges“. Physica Scripta T170 (24.10.2017): 014040. http://dx.doi.org/10.1088/1402-4896/aa8db1.

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28

You, J. H., H. Bolt, R. Duwe, J. Linke und H. Nickel. „Thermomechanical behavior of actively cooled, brazed divertor components under cyclic high heat flux loads“. Journal of Nuclear Materials 250, Nr. 2-3 (Dezember 1997): 184–91. http://dx.doi.org/10.1016/s0022-3115(97)00240-7.

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29

Zhuang, Qing, Lei Cao, Nanyu Mou, Qianqian Lin, Xiyang Zhang, Xianke Yang, Le Han, Pengfei Zi, Tiejun Xu und Damao Yao. „Study on the effect of EAST divertor geometric accuracy on heat load distribution“. Journal of Instrumentation 18, Nr. 01 (01.01.2023): P01025. http://dx.doi.org/10.1088/1748-0221/18/01/p01025.

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Abstract In the EAST experiment, the extent of damages of the divertor is different in toroidal direction. One of the reasons is uneven of heat load of toroidal distribution, which may be caused by geometric errors of the divertor surface. The EAST lower divertor is cooled by 8 toroidal active water-cooling branches, and calorimetric system estimates the heat load and its distribution by measuring the cooling water temperature difference and flow rate. The non-uniformity of the heat load of 8 branches is -3.5% ∼ 4.5%. Besides, using the Leica AT960 / AT401 laser tracker to measure the profile deviation of the upgraded lower divertor, the non-uniformity of the geometric accuracy is -15% ∼ 25%. Besides, it's found that the annular heat load distribution is positively correlated with the comprehensive deviation of the divertor surface. The correlation coefficient is about 15%, and at least seven of the eight divertor regions meet this characteristic.

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30

Li, Xiangyu, Guanghuai Wang, Yun Guo und Songwei Li. „Critical heat flux analysis of divertor cooling flow channel in fusion reactor with CFD method“. Thermal Science, Nr. 00 (2021): 203. http://dx.doi.org/10.2298/tsci210216203l.

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Situated at the bottom of the vacuum vessel, the divertor extracts heat and ash produced by the fusion reaction, minimizes plasma contamination, and protects the surrounding walls from thermal and neutronic loads. The vertical targets of divertor are designed to be able for up to 20 MW/m2 high heat flux. It is a great ordeal for both the material performance and the cooling ability. Critical heat flux (CHF) margin is very crucial during the design of divertor. ANSYS FLUENT is used in this paper to predict the CHF on a monoblock structure with a twisted tape inside the tube. Numerical results are validated with the corresponding sets of experimental results. In this paper, CFD method used to predict CHF of divertor cooling channel was first introduced. On the other hand, influence of inlet subcooling on CHF is studied in detail. The inlet subcooling affect the CHF much complicated for the single- side heated and swirl flow channel. Whether the influencing trend or the locations of CHF occurrence are different under different inlet subcooling. The derivations between the simulation and experimental results were no more than 32%. This study proves the CFD tools can provide efficient help on the understanding of the CHF phenomenon of complex construction.

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31

VAHALA, GEORGE, LINDA VAHALA, JOSEPH MORRISON, SERGEI KRASHENINNIKOV und DIETER SIGMAR. „K–ε compressible 3D neutral fluid turbulence modelling of the effect of toroidal cavities on flame-front propagation in the gas-blanket regime for tokamak divertors“. Journal of Plasma Physics 57, Nr. 1 (Januar 1997): 155–73. http://dx.doi.org/10.1017/s0022377896005235.

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Recent experiments and 2D laminar plasma–fluid simulations have indicated that plasma detachment from the divertor plate is strongly tied to plasma recombination. With plasma recombination, a neutral gas blanket will form between the divertor plate and the plasma frame front. Because of plasma-neutral coupling, the plasma flow along the field lines will drive neutral gas flow with Mach number [ges ]1 and Reynolds number [ges ]1000. A compressible set of conservation and transport equations are solved with 2D mean toroidal flow and 3D turbulence effects over various toroidal cavity geometries. The radial structure of the temperature profile is determined for both turbulent and laminar flow as the flame front propagates down the toroidal cavity. Quantitative results are obtained for the increased heat transfer to the toroidal walls due to turbulence as well as radial profiles for the transport coefficients. It is found that heat loads to the toroidal walls can be increased by factors of 5–20 over that for laminar flow for the cavity geometries studied here. This increased heat transfer to the toroidal walls will lead to decreased levels of heat flux impinging on the divertor plate.

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32

KAWASHIMA, Hisato, Kazuya UEHARA, Nobuhiro NISHINO, Kensaku KAMIYA, Kazuhiro TSUZUKI, Bakhtiari MOHAMMAD, Yoshihiko NAGASHIMA et al. „A Comparison between Divertor Heat Loads in ELMy and HRS H-Modes on JFT-2M“. Journal of Plasma and Fusion Research 80, Nr. 11 (2004): 907–8. http://dx.doi.org/10.1585/jspf.80.907.

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33

Huang, Shenghong, und Shimin Liu. „Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads“. Journal of Fusion Energy 37, Nr. 4 (15.06.2018): 177–86. http://dx.doi.org/10.1007/s10894-018-0164-3.

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34

Jachmich, S., Y. Liang, G. Arnoux, T. Eich, W. Fundamenski, H. R. Koslowski und R. A. Pitts. „Effect of external perturbation fields on divertor particle and heat loads during ELMs at JET“. Journal of Nuclear Materials 390-391 (Juni 2009): 768–72. http://dx.doi.org/10.1016/j.jnucmat.2009.01.204.

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35

Budaev, V. P. „RESULTS OF HIGH HEAT FLUX TUNGSTEN DIVERTOR TARGET TESTS UNDER ITER AND REACTOR TOKAMAK-RELEVANT PLASMA HEAT LOADS (REVIEW)“. Problems of Atomic Science and Technology, Ser. Thermonuclear Fusion 38, Nr. 4 (2015): 5–33. http://dx.doi.org/10.21517/0202-3822-2015-38-4-5-33.

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36

Ishitsuka, E., M. Uchida, K. Sato, M. Akiba und H. Kawamura. „High heat load tests of neutron-irradiated divertor mockups“. Fusion Engineering and Design 56-57 (Oktober 2001): 421–25. http://dx.doi.org/10.1016/s0920-3796(01)00347-7.

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37

Engels, Dion, Samuel A. Lazerson, Victor Bykov und Josefine H. E. Proll. „Investigating the n = 1 and n = 2 error fields in W7-X using the newly accelerated FIELDLINES code“. Plasma Physics and Controlled Fusion 64, Nr. 3 (21.01.2022): 035003. http://dx.doi.org/10.1088/1361-6587/ac43ef.

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Abstract No fusion device can be created without some uncertainty; there is always a slight deviation from the geometric specification. These deviations can add up create a deviation of the magnetic field. This deviation is known as the (magnetic) error field. Correcting these error fields is desired as they cause asymmetries in the divertor loads and can thus cause damage to the device if they grow too large. These error fields can be defined by their toroidal (n) and poloidal number (m). The correction of the n = 1 and n = 2 fields in Wendelstein 7-X (W7-X) is investigated in this work. This investigation focuses on field line diffusion to the divertor, a proxy for divertor heat flux. Such work leverages the 25× speedup obtained through the implementation of a new particle-wall collision model. The n = 1 and n = 2 error fields of the as-built coils model of W7-X are corrected by scanning phase and amplitude of the trim and control coils. Reductions in the divertor load asymmetry by factors of four are demonstrated using error field correction. It is found that the as-built coils model has a significantly lower m / n = 1 / 1 error field than found in experiments (Bozhenkov et al 2018 Nucl. Fusion 59 026004).

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38

Park, In Sun, In Je Kang und Kyu-Sun Chung. „Experimental Estimation of Dust Generation Under ELM-Like Transient Heat Loads in Divertor Plasma Simulator-2“. Fusion Science and Technology 77, Nr. 6 (04.08.2021): 429–36. http://dx.doi.org/10.1080/15361055.2021.1929759.

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39

Li, C., H. Greuner, Y. Yuan, S. X. Zhao, G. N. Luo, B. Böswirth, B. Q. Fu, Y. Z. Jia, X. Liu und W. Liu. „Surface modifications of W divertor components for EAST during exposure to high heat loads with He“. Journal of Nuclear Materials 463 (August 2015): 223–27. http://dx.doi.org/10.1016/j.jnucmat.2014.10.063.

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40

Nagata, Masayoshi, Yusuke Kikuchi und Naoyuki Fukumoto. „Application of Magnetized Coaxial Plasma Guns for Simulation of Transient High Heat Loads on ITER Divertor“. IEEJ Transactions on Electrical and Electronic Engineering 4, Nr. 4 (Juli 2009): 518–22. http://dx.doi.org/10.1002/tee.20438.

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41

Budaev, V. P. „Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)“. Physics of Atomic Nuclei 79, Nr. 7 (Dezember 2016): 1137–62. http://dx.doi.org/10.1134/s106377881607005x.

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42

Gago, Mauricio, Arkadi Kreter, Bernhard Unterberg und Marius Wirtz. „Bubble Formation in ITER-Grade Tungsten after Exposure to Stationary D/He Plasma and ELM-like Thermal Shocks“. Journal of Nuclear Engineering 4, Nr. 1 (21.02.2023): 204–12. http://dx.doi.org/10.3390/jne4010016.

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Plasma-facing materials (PFMs) in the ITER divertor will be exposed to severe conditions, including exposure to transient heat loads from edge-localized modes (ELMs) and to plasma particles and neutrons. Tungsten is the material chosen as PFM for the ITER divertor. In previous tests, bubble formation in ITER-grade tungsten was detected when exposed to fusion relevant conditions. For this study, ITER-grade tungsten was exposed to simultaneous ELM-like transient heat loads and D/He (6%) plasma in the linear plasma device PSI-2. Bubble formation was then investigated via SEM micrographs and FIB cuts. It was found that for exposure to 100.000 laser pulses of 0.6 GWm−2 absorbed power density (Pabs), only small bubbles in the nanometer range were formed close to the surface. After increasing Pabs to 0.8 and 1.0 GWm−2, the size of the bubbles went up to about 1 µm in size and were deeper below the surface. Increasing the plasma fluence had an even larger effect, more than doubling bubble density and increasing bubble size to up to 2 µm in diameter. When using deuterium-only plasma, the samples showed no bubble formation and reduced cracking, showing such bubble formation is caused by exposure to helium plasma.

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43

López-Galilea, I., G. Pintsuk, C. García-Rosales und Jochen Linke. „High Heat Flux Testing of TiC-Doped Isotropic Graphite for Plasma Facing Components“. Advanced Materials Research 59 (Dezember 2008): 288–92. http://dx.doi.org/10.4028/www.scientific.net/amr.59.288.

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The technical design solution for the future thermonuclear fusion reactor, ITER, must guarantee a reasonable lifetime from a safety and economical point of view. Carbon fibre reinforced carbon (CFC) is envisaged as a corrective material solution for the strike point area of ITER divertor due to its high thermal shock resistance necessary to withstand excessive heat loads during transient thermal loads; in particular plasma disruptions that can deposit energy densities of several ten MJm-2 with a typical timescale in the order of milliseconds. In this work, as potential alternative to CFCs new finely dispersed TiC-doped isotropic graphites with high thermal conductivity and mechanical strength, manufactured using synthetic mesophase pitch “AR” as raw material, have been evaluated under typical disruption conditions using an energetic electron beam at the JUDITH facility.

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Si, Hang, Rui Ding, Ilya Senichenkov, Vladimir Rozhansky, Pavel Molchanov, Xiaoju Liu, Guozhang Jia et al. „SOLPS-ITER simulations of high power exhaust for CFETR divertor with full drifts“. Nuclear Fusion 62, Nr. 2 (01.02.2022): 026031. http://dx.doi.org/10.1088/1741-4326/ac3f4b.

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Abstract One of the major challenges for the Gigawatt-class Chinese Fusion Engineering Testing Reactor (CFETR) is to efficiently handle huge power fluxes on plasma-facing components , especially the divertor targets. This work investigates the effects of two candidate radiation impurity species, argon (Ar) and neon (Ne), with two different divertor geometries (baseline and long leg divertor geometry) on the reduction of steady-state power load to divertor targets in CFETR by using the SOLPS-ITER code package with full drifts and kinetic description of neutrals. The modeling results show clearly that increasing the seeding rate of Ar or Ne with fixed fueling gas D2 injection rate reduces the target electron temperature and heat flux density for the baseline divertor geometry, which can be reduced further by higher D2 injection rate. With a high impurity seeding rate, partial detachment with steady-state power load at the divertor target below the engineering limit of 10 MW m−2 is demonstrated. In addition, the radiation efficiency for Ar is better than that for Ne. Increasing the divertor leg length reduces the electron temperature and heat load at the targets. This modeling, therefore, suggests that a long leg divertor design with Ar seeding impurity is appropriate to meet the CFETR divertor requirements.

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Jakubowski, M. W., T. E. Evans, M. E. Fenstermacher, M. Groth, C. J. Lasnier, A. W. Leonard, O. Schmitz et al. „Overview of the results on divertor heat loads in RMP controlled H-mode plasmas on DIII-D“. Nuclear Fusion 49, Nr. 9 (14.08.2009): 095013. http://dx.doi.org/10.1088/0029-5515/49/9/095013.

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Rieth, Michael, Dave Armstrong, Bernhard Dafferner, Sylvia Heger, Andreas Hoffmann, Mirjam Diana Hoffmann, Ute Jäntsch et al. „Tungsten as a Structural Divertor Material“. Advances in Science and Technology 73 (Oktober 2010): 11–21. http://dx.doi.org/10.4028/www.scientific.net/ast.73.11.

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Refractory materials, in particular tungsten base materials are considered as primary candidates for structural high heat load applications in future nuclear fusion power plants. Promising helium-cooled divertor design outlines make use of their high heat conductivity and strength. The upper operating temperature limit is mainly defined by the onset of recrystallization but also by loss of creep strength. The lower operating temperature range is restricted by the use of steel parts for the in- and outlets as well as for the back-bone. Therefore, the most critical issue of tungsten materials in connection with structural divertor applications is the ductile-to-brittle transition. Another problem consists in the fact that especially refractory alloys show a strong correlation between microstructure and their manufacturing history. Since physical and mechanical properties are influenced by the underlying microstructure, refractory alloys can behave quite different, even if their chemical composition is the same. Therefore, creep and thermal conductivity have been investigated using typical commercial tungsten materials. Moreover, the fracture behavior of different tungsten based semi-finished products was characterized by standard Charpy tests which have been performed up to 1100 °C in vacuum. Due to their fabrication history (powder mixing, pressing, sintering, rolling, forging, or swaging) these materials have specific microstructures which lead different fracture modes. The influence of the microstructure characteristics like grain size, anisotropy, texture, or chemical composition has been studied.

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Sieglin, B., T. Eich, M. Faitsch, A. Herrmann, A. Kirk, A. Scarabosio, W. Suttrop und A. Thornton. „Assessment of divertor heat load with and without external magnetic perturbation“. Nuclear Fusion 57, Nr. 6 (09.05.2017): 066045. http://dx.doi.org/10.1088/1741-4326/aa6c20.

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Arnoux, G., P. Andrew, M. Beurskens, S. Brezinsek, C. D. Challis, P. De Vries, W. Fundamenski et al. „Divertor heat load in ITER-like advanced tokamak scenarios on JET“. Journal of Nuclear Materials 390-391 (Juni 2009): 263–66. http://dx.doi.org/10.1016/j.jnucmat.2009.01.094.

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Ritz, G., T. Hirai, P. Norajitra, J. Reiser, R. Giniyatulin, A. Makhankov, I. Mazul, G. Pintsuk und J. Linke. „Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads“. Physica Scripta T138 (Dezember 2009): 014064. http://dx.doi.org/10.1088/0031-8949/2009/t138/014064.

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Tereshin, V. I., A. N. Bandura, O. V. Byrka, V. V. Chebotarev, I. E. Garkusha, I. Landman, V. A. Makhlaj, I. M. Neklyudov, D. G. Solyakov und A. V. Tsarenko. „Application of powerful quasi-steady-state plasma accelerators for simulation of ITER transient heat loads on divertor surfaces“. Plasma Physics and Controlled Fusion 49, Nr. 5A (29.03.2007): A231—A239. http://dx.doi.org/10.1088/0741-3335/49/5a/s19.

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