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Journal articles on the topic 'Heat loads on the divertor'

Author: Grafiati

Published: 8 June 2024

Last updated: 31 July 2025

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1

Barr, William L., and B. Grant Logan. "A Slot Divertor for Tokamaks with High Divertor Heat Loads." Fusion Technology 18, no. 2 (1990): 251–56. http://dx.doi.org/10.13182/fst90-a29297.

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2

Marki, J., R. A. Pitts, J. Horacek, and D. Tskhakaya. "ELM induced divertor heat loads on TCV." Journal of Nuclear Materials 390-391 (June 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, no. 6 (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, and G. Maddaluno. "Disruption heat loads on the JET MkIIGB divertor." Plasma Physics and Controlled Fusion 44, no. 6 (2002): 905–29. http://dx.doi.org/10.1088/0741-3335/44/6/319.

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5

Mavrin, Aleksey A., and Andrey A. Pshenov. "Tolerable Stationary Heat Loads to Liquid Lithium Divertor Targets." Plasma 5, no. 4 (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 redeposit

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6

Dai, S. Y., D. F. Kong, V. S. Chan, L. Wang, Y. Feng, and D. Z. Wang. "EMC3–EIRENE simulations of neon impurity seeding effects on heat flux distribution on CFETR." Nuclear Fusion 62, no. 3 (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

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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, et al. "Surface heat loads on the ITER divertor vertical targets." Nuclear Fusion 57, no. 4 (2017): 046025. http://dx.doi.org/10.1088/1741-4326/aa5e2a.

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9

Abrams, T., M. A. Jaworski, J. Kallman, et al. "Response of NSTX liquid lithium divertor to high heat loads." Journal of Nuclear Materials 438 (July 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., and J. Wesley. "Scaling of Divertor Temperature and Heat Loads for TPX-Class Devices." Fusion Technology 21, no. 3P2A (1992): 1406–15. http://dx.doi.org/10.13182/fst92-a29919.

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12

Gao, Y., Marcin W. Jakubowski, Peter Drewelow, et al. "Methods for quantitative study of divertor heat loads on W7-X." Nuclear Fusion 59, no. 6 (2019): 066007. http://dx.doi.org/10.1088/1741-4326/ab0f49.

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13

Scarabosio, A., C. Fuchs, A. Herrmann, and E. Wolfrum. "ELM characteristics and divertor heat loads in ASDEX Upgrade helium discharges." Journal of Nuclear Materials 415, no. 1 (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, et al. "Characterization of the Crack and Recrystallization of W/Cu Monoblocks of the Upper Divertor in EAST." Applied Sciences 13, no. 2 (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

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15

Oka, Kiyoshi, Satoshi Kakudate, Nobukazu Takeda, Yuji Takiguchi, and Kentaro Akou. "Measurement and Control System for ITER Remote Maintenance Equipment." Journal of Robotics and Mechatronics 10, no. 2 (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

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16

Takeda, Nobukazu, Kiyoshi Oka, Kentaro Akou, and Yuji Takiguchi. "Development of Divertor Remote Maintenance System." Journal of Robotics and Mechatronics 10, no. 2 (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 po

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17

Araki, M., K. Kitamura, K. Urata, and S. Suzuki. "Analyses of divertor high heat-flux components on thermal and electromagnetic loads." Fusion Engineering and Design 42, no. 1-4 (1998): 381–87. http://dx.doi.org/10.1016/s0920-3796(97)00180-4.

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18

Gunn, J. P., S. Carpentier-Chouchana, R. Dejarnac, et al. "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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19

Hong, Suk–Ho, Richard A. Pitts, Hyung-Ho Lee, et al. "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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20

Carli, S., R. A. Pitts, X. Bonnin, F. Subba, and R. Zanino. "Effect of strike point displacements on the ITER tungsten divertor heat loads." Nuclear Fusion 58, no. 12 (2018): 126022. http://dx.doi.org/10.1088/1741-4326/aae43f.

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21

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

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22

Noce, Simone, Davide Flammini, Pasqualino Gaudio, et al. "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, no. 3 (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 s

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23

Riccardi, B., P. Gavila, R. Giniatulin, et al. "Effect of stationary high heat flux and transient ELMs-like heat loads on the divertor PFCs." Fusion Engineering and Design 88, no. 9-10 (2013): 1673–76. http://dx.doi.org/10.1016/j.fusengdes.2013.05.016.

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24

Sizyuk, V., and A. Hassanein. "Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities." Physics of Plasmas 22, no. 1 (2015): 013301. http://dx.doi.org/10.1063/1.4905632.

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25

Hayashi, Y., M. Kobayashi, K. Mukai, S. Masuzaki, and 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

Zhuang, Qing, Lei Cao, Nanyu Mou, et al. "Study on the effect of EAST divertor geometric accuracy on heat load distribution." Journal of Instrumentation 18, no. 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

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27

Li, Xiangyu, Guanghuai Wang, Yun Guo, and Songwei Li. "Critical heat flux analysis of divertor cooling flow channel in fusion reactor with CFD method." Thermal Science, no. 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 a

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28

Miloshevskii, G. V., and G. S. Romanov. "Evaluation of Heat Loads in Graphite Divertor Plates Acted by a Magnetized Electron Flux." Heat Transfer Research 33, no. 7-8 (2002): 9. http://dx.doi.org/10.1615/heattransres.v33.i7-8.60.

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29

Silburn, S. A., G. F. Matthews, C. D. Challis, et al. "Mitigation of divertor heat loads by strike point sweeping in high power JET discharges." Physica Scripta T170 (October 24, 2017): 014040. http://dx.doi.org/10.1088/1402-4896/aa8db1.

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30

You, J. H., H. Bolt, R. Duwe, J. Linke, and H. Nickel. "Thermomechanical behavior of actively cooled, brazed divertor components under cyclic high heat flux loads." Journal of Nuclear Materials 250, no. 2-3 (1997): 184–91. http://dx.doi.org/10.1016/s0022-3115(97)00240-7.

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31

VAHALA, GEORGE, LINDA VAHALA, JOSEPH MORRISON, SERGEI KRASHENINNIKOV та 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, № 1 (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.

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32

WU Yanghai, DU Hailong, XUE Lei, LI Jiaxian, XUE Miao, and ZHENG Guoyao. "Machine Learning-Based Prediction of Heat Load on Tokamak Divertor Target Plates." Acta Physica Sinica 74, no. 13 (2025): 0. https://doi.org/10.7498/aps.74.20250381.

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The SOLPS-ITER edge plasma simulation code has become a primary tool for divertor physics design and target plate heat load prediction in fusion research. However, SOLPS-ITER- based divertor design requires not only substantial computational time but also intensive hardware resources, which fundamentally limits its application in advancing divertor optimization, particularly in large-scale fusion reactor divertor design. In this paper, the machine learning method is used for the first time to predict the plasma parameters of the divertor target plate for HL-3, which provides a theoretical basi

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33

KAWASHIMA, Hisato, Kazuya UEHARA, Nobuhiro NISHINO, et al. "A Comparison between Divertor Heat Loads in ELMy and HRS H-Modes on JFT-2M." Journal of Plasma and Fusion Research 80, no. 11 (2004): 907–8. http://dx.doi.org/10.1585/jspf.80.907.

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34

Huang, Shenghong, and Shimin Liu. "Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads." Journal of Fusion Energy 37, no. 4 (2018): 177–86. http://dx.doi.org/10.1007/s10894-018-0164-3.

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35

Jachmich, S., Y. Liang, G. Arnoux, et al. "Effect of external perturbation fields on divertor particle and heat loads during ELMs at JET." Journal of Nuclear Materials 390-391 (June 2009): 768–72. http://dx.doi.org/10.1016/j.jnucmat.2009.01.204.

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36

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, no. 4 (2015): 5–33. http://dx.doi.org/10.21517/0202-3822-2015-38-4-5-33.

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37

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

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38

Engels, Dion, Samuel A. Lazerson, Victor Bykov, and 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, no. 3 (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

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39

Park, In Sun, In Je Kang, and Kyu-Sun Chung. "Experimental Estimation of Dust Generation Under ELM-Like Transient Heat Loads in Divertor Plasma Simulator-2." Fusion Science and Technology 77, no. 6 (2021): 429–36. http://dx.doi.org/10.1080/15361055.2021.1929759.

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40

Li, C., H. Greuner, Y. Yuan, et al. "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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41

Nagata, Masayoshi, Yusuke Kikuchi, and 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, no. 4 (2009): 518–22. http://dx.doi.org/10.1002/tee.20438.

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42

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, no. 7 (2016): 1137–62. http://dx.doi.org/10.1134/s106377881607005x.

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43

Gago, Mauricio, Arkadi Kreter, Bernhard Unterberg, and 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, no. 1 (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 c

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44

Si, Hang, Rui Ding, Ilya Senichenkov, et al. "SOLPS-ITER simulations of high power exhaust for CFETR divertor with full drifts." Nuclear Fusion 62, no. 2 (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 clearl

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45

Huang, Xiaoxuan, Jianjun Wei, Zongbiao Ye, and Fujun Gou. "Evolution Mechanism and Mechanical Response of Tungsten Surface Damage Under Pulsed Heat Load and Helium Plasma Irradiation." Processes 13, no. 6 (2025): 1711. https://doi.org/10.3390/pr13061711.

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This study investigates the synergistic effects of pulsed heat load and helium plasma irradiation on the surface damage evolution of high-purity tungsten, a candidate plasma-facing material (PFM) for future fusion reactors. Using a self-developed linear plasma device, tungsten samples were exposed to controlled single-pulse heat loads (32–124 MW·m−2) and helium plasma fluxes (7.76 × 1022–2.40 × 1023 ions·m−2·s−1). SEM and XRD analyses revealed a progressive damage mechanism involving helium bubble formation, pit collapse, coral-like nanostructure evolution, and melting-induced restructuring. T

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46

Adebayo-Ige, P. O., K. F. Gan, C. J. Lasnier, R. Maingi, and B. D. Wirth. "Divertor heat load estimates on NSTX and DIII-D using new and open-source 2D inversion analysis code." Nuclear Fusion 64, no. 9 (2024): 096006. http://dx.doi.org/10.1088/1741-4326/ad60dd.

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Abstract A thermography inversion algorithm has been developed in the open-source Python-based computer code, HYPERION, to calculate the heat flux incident on plasma-facing components (PFCs) in axisymmetric tokamaks. The chosen mesh size at the surface significantly affects the calculated transient heat flux results. The calculated transient heat flux will exceed the real value when the mesh size tends to zero but will underestimate the real value when the mesh size is large. A criterion for determining the appropriate mesh size for the transient heat flux calculation will be discussed. The nu

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47

López-Galilea, I., G. Pintsuk, C. García-Rosales, and Jochen Linke. "High Heat Flux Testing of TiC-Doped Isotropic Graphite for Plasma Facing Components." Advanced Materials Research 59 (December 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

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48

Wang, Fuqiong, Xiang Gu, jiankun hua, et al. "Divertor heat flux challenge and mitigation in the EHL-2 spherical torus." Plasma Science and Technology, January 23, 2025. https://doi.org/10.1088/2058-6272/adadb8.

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Abstract The divertor design is critical to heat-loads handling and thus to achievements of high-performance plasma operations in EHL-2 (ENN He-Long 2) tokamak. This paper presents the design of an X-point target (XPT) divertor, featuring a conventional inner divertor and an XPT outer divertor, aimed at an effective controlling of heat loads, which may be extremely high during the high ion-temperature scenarios. The divertor target plates have been considered to be made from carbon-based materials, which can handle heat loads up to 5 MW/m². Divertor performances, including the head load contro

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49

Gao, Yu, Yuhe Feng, Michael Bernd Sebastian Endler, et al. "Improvement in the simulation tools for heat distribution predictions and control of baffle and middle divertor loads in Wendelstein 7-X." Nuclear Fusion, December 29, 2022. http://dx.doi.org/10.1088/1741-4326/acaf0e.

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Abstract In the first divertor campaign in Wendelstein 7-X (W7-X), unexpected significant heat loads were observed at particular plasma-facing components (e.g. baffle tiles and middle divertor part) which were not designed to receive high heat flux. In a prior investigation, it was concluded that the previous diffusive field line tracing model (DFLT) used for divertor design in W7-X can not reproduce these loads, due to the missing physics in simulating the heat transport in the shaded flux tubes. To tackle this issue, two new efficient codes (DFLT rev and EMC3-Lite) are introduced and validat

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50

Xie, Tian, Hang Li, Wei Zhang, et al. "EMC3-EIRENE simulations of edge plasma and impurity transport by toroidally localized argon seeding on CFETR X-divertor." Nuclear Fusion, December 9, 2024. https://doi.org/10.1088/1741-4326/ad9bc9.

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Abstract Three-dimensional (3D) Edge Monte Carlo transport code EMC3-EIRENE has been employed to study edge plasma and impurity transport with toroidally localized argon seeding on Chinese fusion engineering testing reactor (CFETR) X-divertor configuration. The argon impurity seeded at different poloidal locations has been investigated to evaluate the varied profile of the main plasma in the scrape-off layer (SOL) and on the divertor targets, which shows a strong dependence on the poloidal position of argon gas puffing. The argon impurity seeded at the upstream SOL regions can result in a toro

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