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Sample records for inboard divertor target

  1. Electron beam facility for divertor target experiments

    International Nuclear Information System (INIS)

    Anisimov, A.; Gagen-Torn, V.; Giniyatulin, R.N.

    1994-01-01

    To test different concepts of divertor targets and bumpers an electron beam facility was assembled in Efremov Institute. It consists of a vacuum chamber (3m 3 ), vacuum pump, electron beam gun, manipulator to place and remove the samples, water loop and liquid metal loop. The following diagnostics of mock-ups is stipulated: (1) temperature distribution on the mock-up working surface (scanning pyrometer and infra-red imager); (2) temperature distribution over mocked-up thickness in 3 typical cross-sections (thermo-couples); (3) cracking dynamics during thermal cycling (acoustic-emission method), (4) defects in the mock-up before and after tests (ultra-sonic diagnostics, electron and optical microscopes). Carbon-based and beryllium mock-ups are made for experimental feasibility study of water and liquid-metal-cooled divertor/bumper concepts

  2. Alternative divertor target concepts for next step fusion devices

    Science.gov (United States)

    Mazul, I. V.

    2016-12-01

    The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

  3. Effect of low density H-mode operation on edge and divertor plasma parameters

    International Nuclear Information System (INIS)

    Maingi, R.; Mioduszewski, P.K.; Cuthbertson, J.W.

    1994-07-01

    We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation

  4. Enhancing the DEMO divertor target by interlayer engineering

    International Nuclear Information System (INIS)

    Barrett, T.R.; McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M.; Rieth, M.; Reiser, J.

    2015-01-01

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m"2. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m"2 surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m"2.

  5. Enhancing the DEMO divertor target by interlayer engineering

    Energy Technology Data Exchange (ETDEWEB)

    Barrett, T.R., E-mail: tom.barrett@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M. [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, M.; Reiser, J. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany)

    2015-10-15

    Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m{sup 2}. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m{sup 2} surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m{sup 2}.

  6. Narrow power deposition profiles on the JET divertor target

    International Nuclear Information System (INIS)

    Lingertat, J.; Laux, M.; Monk, R.

    2001-01-01

    One of the key unresolved issues in the design of a future fusion reactor is the power handling capability of the divertor target plates. Earlier we reported on the existence of narrow power deposition profiles in JET, obtained mainly from Langmuir probe measurements. We repeated these measurements in the MkI, MkII and MkIIGB divertor configurations with an upgraded probe system, which allowed us to study the profile shape in more detail. The main results of this study are: In NB heated discharges the electron temperature and power flux at the outer target show a distinct peak of ∼5 mm half-width near the separatrix strike point. The corresponding profiles on the inner target do not show a similar feature. The height of the narrow peak increases with NB heating power and decreases with deuterium and impurity gas puffing. Ion orbit losses are suggested as a possible explanation of the observed profile shape

  7. Simulation of divertor targets shielding during transients in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, Sergey, E-mail: serguei.pestchanyi@kit.edu [KIT, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen (Germany); Pitts, Richard; Lehnen, Michael [ITER Organization,Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2016-11-01

    Highlights: • We simulated plasma shielding effect during disruption in ITER using the TOKES code. • It has been found that vaporization is unavoidable under action of ITER transients, but plasma shielding drastically reduces the divertor target damage: the melt pool and the vaporization region widths reduced 10–15 times. • A simplified 1D model describing the melt pool depth and the shielded heat flux to the divertor targets have been developed. • The results of the TOKES simulations have been compared with the analytic model when the model is valid. - Abstract: Direct extrapolation of the disruptive heat flux on ITER conditions predicts severe melting and vaporization of the divertor targets causing their intolerable damage. However, tungsten vaporized from the target at initial stage of the disruption can create plasma shield in front of the target, which effectively protects the target surface from the rest of the heat flux. Estimation of this shielding efficiency has been performed using the TOKES code. The shielding effect under ITER conditions is found to be very strong: the maximal depth of the melt layer reduced 4 times, the melt layer width—more than 10 times and vaporization region shrinks 10–15 times due to shielding for unmitigated disruption of 350 MJ discharge. The simulation results show complex, 2D plasma dynamics of the shield under ITER conditions. However, a simplified analytic model, valid for rough estimation of the maximum value for the shielded flux to the target and for the melt depth at the target surface has been developed.

  8. Analysis of particle transport in a gas target divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ohtsu, Shigeki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

    1996-10-01

    2-dimensional modelling of divertor plasma was performed with three types of the divertor geometry configuration. Pumping is effective to reduce neutral recycling to core region in the configuration without baffle. In baffle configuration, a good shielding of neutrals in the divertor region can be achieved. The dome configuration reduces plasma density near the null region and flow shear near the separatrix. (author)

  9. Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

    International Nuclear Information System (INIS)

    Zhang, Chuanjia; Chen, Bin; Xing, Zhe; Wu, Haosheng; Mao, Shifeng; Luo, Zhengping; Peng, Xuebing; Ye, Minyou

    2016-01-01

    Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D 2 gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m 2 is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

  10. Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Chuanjia; Chen, Bin [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Xing, Zhe [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Haosheng [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Mao, Shifeng, E-mail: sfmao@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Luo, Zhengping; Peng, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

    2016-11-01

    Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D{sub 2} gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m{sup 2} is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

  11. Inboard seal mounting

    Science.gov (United States)

    Hayes, John R. (Inventor)

    1983-01-01

    A regenerator assembly for a gas turbine engine has a hot side seal assembly formed in part by a cast metal engine block having a seal recess formed therein that is configured to supportingly receive ceramic support blocks including an inboard face thereon having a regenerator seal face bonded thereto. A pressurized leaf seal is interposed between the ceramic support block and the cast metal engine block to bias the seal wear face into sealing engagement with a hot side surface of a rotary regenerator matrix.

  12. Assessment of X-point target divertor configuration for power handling and detachment front control

    Directory of Open Access Journals (Sweden)

    M.V. Umansky

    2017-08-01

    Full Text Available A study of long-legged tokamak divertor configurations is performed with the edge transport code UEDGE (Rognlien et al., J. Nucl. Mater. 196, 347, 1992. The model parameters are based on the ADX tokamak concept design (LaBombard et al., Nucl. Fusion 55, 053020, 2015. Several long-legged divertor configurations are considered, in particular the X-point target configuration proposed for ADX, and compared with a standard divertor. For otherwise identical conditions, a scan of the input power from the core plasma is performed. It is found that as the power is reduced to a threshold value, the plasma in the outer leg transitions to a fully detached state which defines the upper limit on the power for detached divertor operation. Reducing the power further results in the detachment front shifting upstream but remaining stable. At low power the detachment front eventually moves to the primary X-point, which is usually associated with degradation of the core plasma, and this defines the lower limit on the power for the detached divertor operation. For the studied parameters, the operation window for a detached divertor in the standard divertor configuration is very small, or even non-existent; under the same conditions for long-legged divertors the detached operation window is quite large, in particular for the X-point target configuration, allowing a factor of 5–10 variation in the input power. These modeling results point to possibility of stable fully detached divertor operation for a tokamak with extended divertor legs.

  13. Feasibility of ''gas target'' mode of divertor operation in ITER

    International Nuclear Information System (INIS)

    Kukushkin, A.S.

    1994-01-01

    Power load upon the divertor target remains one of the most critical issues for a tokamak reactor. Simple estimates, confirmed by 2D modelling, together with some indications from tokamak experiments, showed that the profile of power flow gets narrower along with increase of the reactor power, because strong temperature dependence of the parallel heat conductance, χ parallel αΤ 5/2 , favours parallel heat transport in competition with the cross-field one. This leads to unacceptable peak loads and makes one to look for a means to spread the power more evenly across the magnetic field. The scope of the present paper is to show the results of the modelling studies and to discuss the physical and computational issues which are still missing or are insufficiently developed. I must apologize for partiality for my own calculations with the DDC83 code, but there are some reasons justifying this: they have been the first calculations on this issue, they seem to be the most extensive, and they are certainly the most familiar to me. (orig.)

  14. Technologies for ITER divertor vertical target plasma facing components

    International Nuclear Information System (INIS)

    Schlosser, J.; Escourbiac, F.; Merola, M.; Fouquet, S.; Bayetti, P.; Cordier, J.J.; Grosman, A.; Missirlian, M.; Tivey, R.; Roedig, M.

    2005-01-01

    The ITER divertor vertical target has to sustain heat fluxes up to 20 MW m -2 . The concept developed for this plasma facing component working at steady state is based on carbon fibre composite armour for the lower straight part and tungsten for the curved upper part. The main challenges involved in the use of such components include the removal of the high heat fluxes deposited and mechanically and thermally joining the armour to the metallic heat sink, despite the mismatch in the thermal expansions. Two solutions based on the use of a CuCrZr hardened copper alloy and an active metal casting (AMC (registered) ) process were investigated during the ITER EDA phase: the first one called 'flat tile geometry' was mainly developed for the Tore Supra pumped limiter, the second one called 'monoblock geometry' was developed by the EU Participating Team for the ITER project. This paper presents a review of these two solutions and analyses their assets and drawbacks: pressure drop, critical heat flux, surface temperature and expected behaviour during operation, risks during the manufacture, control of the armour defects during the manufacture and at the reception, and the possibility of repairing defective tiles

  15. Acceptance criteria for the ITER divertor vertical target

    International Nuclear Information System (INIS)

    Fouquet, S.; Schlosser, J.; Merola, M.; Durocher, A.; Escourbiac, F.; Grosman, A.; Missirlian, M.; Portafaix, C.

    2006-01-01

    In the frame of the toroidal pump limiter fabrication for Tore Supra, CEA developed a large experience of infrared test for acceptance of high heat flux components armoured with carbon fibre composite flat tiles. The test is based on a thermal transient induced by an alternative hot/cold water flow in the heat sink structure. The tile surface temperature transients are compared with those of a reference element, the maximum difference for each tile leading to a value called ΔT ref m ax . This method is proposed for the commissioning of plasma facing components for the ITER divertor vertical target. This paper describes the determination of the best acceptance criteria for the 'monoblock' geometry of the carbon part. First, it has been shown that the location and the extension of the defects could reliably be determined by monitoring both top and lateral surfaces during the test. Second, it was possible to fix an acceptance method based on ΔT ref m ax . Samples with calibrated defects are now under fabrication to validate the results

  16. Design study of ITER-like divertor target for DEMO

    International Nuclear Information System (INIS)

    Crescenzi, Fabio; Bachmann, C.; Richou, M.; Roccella, S.; Visca, E.; You, J.-H.

    2015-01-01

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m"−"2, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  17. Design study of ITER-like divertor target for DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Crescenzi, Fabio, E-mail: fabio.crescenzi@enea.it [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bachmann, C. [EFDA, Power Plant Physics and Technology, Boltzmannstraße 2, 85748 Garching (Germany); Richou, M. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Roccella, S.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); You, J.-H. [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-10-15

    Highlights: • ‘DEMO’ is a near-term Power Plant Conceptual Study (PPCS). • The ITER-like design concept represents a promising solution also for DEMO plasma facing units. • The optimization of PFUs aims to enhance the thermo-mechanical behaviour of the component. • The optimized geometry was evaluated by ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). - Abstract: A near-term water-cooled target solution has to be evaluated together with the required technologies and its power exhaust limit under ‘DEMO’ conditions. The ITER-like design concept based on the mono-block technology using W as armour material and the CuCrZr-IG as structural material with an interlayer of pure copper represents a promising solution also for DEMO. This work reports the design study of an “optimized” ITER-like Water Cooled Divertor able to withstand a heat flux of 10 MW m{sup −2}, as requested for DEMO operating conditions. The optimization of plasma facing unit (PFU) aims to enhance the thermo-mechanical behaviour of the component by varying some geometrical parameters (monoblock size, interlayer thickness and, tube diameter and thickness). The optimization was performed by means of the multi-variable optimization algorithms using the FEM code ANSYS. The coolant hydraulic conditions (inlet pressure, temperature and velocity) were fixed for simplicity. This study is based on elastic analysis and 3 dimensional modelling. The resulting optimized geometry was evaluated on the basis of the ITER SDC-IC criteria and in terms of low cycle fatigue (LCF). The margin to the critical heat flux (CHF) was also estimated. Further design study (taking into account the effect of neutron radiation on the material properties) together with mock-up fabrication and high-heat-flux (HHF) tests are foreseen in next work programmes.

  18. Study of high-Z target plate materials in the divertor of ASDEX-Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, S; Asmussen, K; Engelhardt, W; Field, A R; Fussmann, G; Lieder, G; Naujoks, D; Neu, R; Radtke, R; Wenzel, U [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1994-12-31

    The reduction of divertor tile erosion is a challenging problem in present and future tokamaks. Until now, graphite has almost exclusively been used for divertor plates, and it is estimated that unacceptable amounts of material would be eroded under reactor relevant conditions where power fluxes to the target plates as high as 20 MW/m{sup 2} are expected. In a high-recycling divertor with relatively low temperature (5 eVtarget erosion problem. The reason is that the sputtering rates for these materials are extremely low under low temperature conditions. In addition, at high density the ionization lengths can be smaller than the gyro-radius leading to a high probability for prompt redeposition of the eroded ions. To provide a test of the suitability of high-Z materials for the divertor plates, in-situ studies of the erosion of various divertor target materials have been performed by means of passive spectroscopy. From our spectroscopic observations we infer that under high density divertor conditions the advantages of high-Z materials become fully efficient. (author) 6 refs., 2 figs.

  19. Particle and power deposition on divertor targets in EAST H-mode plasmas

    International Nuclear Information System (INIS)

    Wang, L.; Xu, G.S.; Guo, H.Y.; Chen, R.; Ding, S.; Gan, K.F.; Gao, X.; Gong, X.Z.; Jiang, M.; Liu, P.; Liu, S.C.; Luo, G.N.; Ming, T.F.; Wan, B.N.; Wang, D.S.; Wang, F.M.; Wang, H.Q.; Wu, Z.W.; Yan, N.; Zhang, L.

    2012-01-01

    The effects of edge-localized modes (ELMs) on divertor particle and heat fluxes were investigated for the first time in the Experimental Advanced Superconducting Tokamak (EAST). The experiments were carried out with both double null and lower single null divertor configurations, and comparisons were made between the H-mode plasmas with lower hybrid current drive (LHCD) and those with combined ion cyclotron resonance heating (ICRH). The particle and heat flux profiles between and during ELMs were obtained from Langmuir triple-probe arrays embedded in the divertor target plates. And isolated ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM-free period. It was demonstrated that ELM-induced radial transport predominantly originated from the low-field side region, in good agreement with the ballooning-like transport model and experimental results of other tokamaks. ELMs significantly enhanced the divertor particle and heat fluxes, without significantly broadening the SOL width and plasma-wetted area on the divertor target in both LHCD and LHCD + ICRH H-modes, thus posing a great challenge for the next-step high-power, long-pulse operation in EAST. Increasing the divertor-wetted area was also observed to reduce the peak heat flux and particle recycling at the divertor target, hence facilitating long-pulse H-mode operation. The particle and heat flux profiles during ELMs appeared to exhibit multiple peak structures, and were analysed in terms of the behaviour of ELM filaments and the flux tubes induced by modified magnetic topology during ELMs. (paper)

  20. Copper matrix composites as heat sink materials for water-cooled divertor target

    Directory of Open Access Journals (Sweden)

    Jeong-Ha You

    2015-12-01

    Full Text Available According to the recent high heat flux (HHF qualification tests of ITER divertor target mock-ups and the preliminary design studies of DEMO divertor target, the performance of CuCrZr alloy, the baseline heat sink material for DEMO divertor, seems to only marginally cover the envisaged operation regime. The structural integrity of the CuCrZr heat sink was shown to be affected by plastic fatigue at 20 MW/m². The relatively high neutron irradiation dose expected for the DEMO divertor target is another serious concern, as it would cause significant embrittlement below 250 °C or irradiation creep above 350 °C. Hence, an advanced design concept of the divertor target needs to be devised for DEMO in order to enhance the HHF performance so that the structural design criteria are fulfilled for full operation scenarios including slow transients. The biggest potential lies in copper-matrix composite materials for the heat sink. In this article, three promising Cu-matrix composite materials are reviewed in terms of thermal, mechanical and HHF performance as structural heat sink materials. The considered candidates are W particle-reinforced, W wire-reinforced and SiC fiber-reinforced Cu matrix composites. The comprehensive results of recent studies on fabrication technology, design concepts, materials properties and the HHF performance of mock-ups are presented. Limitations and challenges are discussed.

  1. Surface heat loads on the ITER divertor vertical targets

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Carpentier-Chouchana, S.; Escourbiac, F.; Hirai, T.; Panayotis, S.; Pitts, R.A.; Corre, Y.; Dejarnac, Renaud; Firdaouss, M.; Kočan, M.; Komm, Michael; Kukushkin, A.; Languille, P.; Missirlian, M.; Zhao, W.; Zhong, G.

    2017-01-01

    Roč. 57, č. 4 (2017), č. článku 046025. ISSN 0029-5515 Institutional support: RVO:61389021 Keywords : ITER * divertor * ELM heat load * inter-ELM heat load * tungsten Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa5e2a

  2. D III-D divertor target heat flux measurements during Ohmic and neutral beam heating

    International Nuclear Information System (INIS)

    Hill, D.N.; Petrie, T.; Mahdavi, M.A.; Lao, L.; Howl, W.

    1988-01-01

    Time resolved power deposition profiles on the D III-D divertor target plates have been measured for Ohmic and neutral beam injection heated plasmas using fast response infrared thermography (τ ≤ 150 μs). Giant Edge Localized Modes have been observed which punctuate quiescent periods of good H-mode confinement and deposit more than 5% of the stored energy of the core plasma on the divertor armour tiles on millisecond time-scales. The heat pulse associated with these events arrives approximately 0.5 ms earlier on the outer leg of the divertor relative to the inner leg. The measured power deposition profiles are displaced relative to the separatrix intercepts on the target plates, and the peak heat fluxes are a function of core plasma density. (author). Letter-to-the-editor. 11 refs, 7 figs

  3. Divertor detachment

    Science.gov (United States)

    Krasheninnikov, Sergei

    2015-11-01

    The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

  4. Experience gained with the 3D machining of the W7-X HHF divertor target elements

    International Nuclear Information System (INIS)

    Junghanns, P.; Boscary, J.; Peacock, A.

    2015-01-01

    Highlights: • The Wendelstein 7-X surface of the actively cooled divertor is built up of 890 individually 3D machined target elements. • To date 300 target elements have been 3D machined with an accuracy of ±0.015 mm. • Copper discovered on the surface of few elements is no risk to operation. - Abstract: The high heat flux (HHF) divertor of W7-X consists of 100 target modules assembled from 890 actively water-cooled target elements protected with CFC tiles. The divertor surface will be built up of individually 3D machined target elements with 89 individual element types. To date 300 of the 890 target elements have been 3D machined with a very good accuracy. To achieve this successful result, a prototyping phase has been conducted to qualify the manufacturing route and to define the acceptance criteria with measures taken to minimize the risk of unacceptable damage during the manufacturing. After the 3D-machining, during the incoming inspection, copper infiltration from the interface between the CFC tiles and the CuCrZr heat sink to the plasma facing surface was detected in a small number of elements.

  5. European DEMO divertor target: Operational requirements and material-design interface

    Directory of Open Access Journals (Sweden)

    J.H. You

    2016-12-01

    Full Text Available Recently, an integrated program of conceptual design activities for the European DEMO reactor was launched in the framework of the EUROfusion Consortium, where reliable power handling capability was identified as one of the most critical scientific as well as technological challenges for a DEMO reactor. The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The DEMO divertor target will have to withstand extreme thermal loads where the local peak heat flux is expected to reach up to 20 MW/m2 during slow transient events in DEMO. To assure sufficient heat removal capability of the divertor target against normal and transient operational scenarios under expected cumulative neutron dose of up to 13 dpa is one of the fundamental engineering challenges imposed on target design. To develop the design of the DEMO divertor and related technologies, an R&D work package ‘Divertor’ has been set up in this consortium. The subproject ‘Target Development’ is devoted to the development of the conceptual design and the core technologies of the plasma-facing target. Devising and implementing novel structural heat sink materials (e.g. W/Cu composites to advanced target design concepts is one of the major objectives of this subproject. In this paper, the underlying design requirements imposed by the envisaged power exhaust goal and the prominent material-design interface issues are discussed. In addition, the candidate design concepts being currently considered are presented together with the related material issues. Finally, the first results achieved so far are presented.

  6. NSTX Tangential Divertor Camera

    International Nuclear Information System (INIS)

    Roquemore, A.L.; Ted Biewer; Johnson, D.; Zweben, S.J.; Nobuhiro Nishino; Soukhanovskii, V.A.

    2004-01-01

    Strong magnetic field shear around the divertor x-point is numerically predicted to lead to strong spatial asymmetries in turbulence driven particle fluxes. To visualize the turbulence and associated impurity line emission near the lower x-point region, a new tangential observation port has been recently installed on NSTX. A reentrant sapphire window with a moveable in-vessel mirror images the divertor region from the center stack out to R 80 cm and views the x-point for most plasma configurations. A coherent fiber optic bundle transmits the image through a remotely selected filter to a fast camera, for example a 40500 frames/sec Photron CCD camera. A gas puffer located in the lower inboard divertor will localize the turbulence in the region near the x-point. Edge fluid and turbulent codes UEDGE and BOUT will be used to interpret impurity and deuterium emission fluctuation measurements in the divertor

  7. Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

    OpenAIRE

    中村 博文; 西 正孝

    2003-01-01

    Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium transport properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authors' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evalua...

  8. Structural impact of creep in tungsten monoblock divertor target at 20 MW/m2

    Directory of Open Access Journals (Sweden)

    Muyuan Li

    2018-01-01

    Full Text Available In order to increase erosion lifetime of the divertor target, in the 2nd design phase of R&D work package ‘Divertor’ for European DEMO, armor thickness of tungsten monoblock divertor target is increased from 5 mm to 8 mm. By increasing armor thickness, surface temperature increases nearly linearly, which makes effect of creep no longer negligible at slow transients of 20 MW/m2. In this work, structural impact of creep in tungsten monoblock divertor target is for the first time quantitatively analyzed with the aid of finite element method. The numerical simulations have revealed that creep results in an increase of inelastic strain accumulation. With increasing armor thickness, tensile surface stress along x-axis (the longer edge at the plasma-facing surface of tungsten monoblock reduces, while surface stress along z-axis (axial direction of the cooling tube changes from tensile to compressive. Creep will accelerate this change. With increasing grain size, creep strain accumulation at loading surface increases due to higher creep rates, while plastic strain accumulation decreases. Creep can mitigate the risk of deep cracking by reducing the driving force for crack opening, and has a positive impact for preventing the contact between the upper parts of neighboring monoblocks in high heat flux tests.

  9. Tritium permeation evaluation through vertical target of divertor based on recent tritium transport properties

    International Nuclear Information System (INIS)

    Nakamura, Hirofumi; Nishi, Masataka

    2003-11-01

    Re-evaluation of tritium permeation through vertical target of divertor under the ITER operation condition was carried out using tritium properties in the candidate materials such as the diffusion coefficient and the trapping factors in tungsten for armor, and the surface recombination coefficient on copper for the heat sink obtained by authours' recent investigation (authors' data), which simulated the plasma-facing conditions of ITER. Evaluation with the data set of previous evaluation was also carried out for comparison (previous data). The permeation analysis was carried out individually by classifying into the armor region (Carbon Fiber Composites and tungsten) and the slit region without armor (3% of armor surface area) assuming the incident flux and temperature for each region. As the results of the permeation analysis, estimated permeation amount with the authors' data was one order less than that with the previous data at the end of lifetime of the divertor due to authors' small diffusion coefficient of tritium in tungsten. It also indicated the possibility that permeation through the slit region of the armor tiles could dominate total permeation through the vertical target, since tritium permeation amount through tungsten armor with the authors' data was estimated to be reduced drastically smaller than that with the previous evaluation data. The result of a little tritium permeation amount through the vertical target with the authors' data ensured the conservatism of the current evaluation of tritium concentration in the primary cooling water in ITER divertor, as it indicated the possibility of direct drainage of the divertor primary cooling water. (author)

  10. Modelling of surface evolution of rough surface on divertor target in fusion devices

    International Nuclear Information System (INIS)

    Dai, Shuyu; Liu, Shengguang; Sun, Jizhong; Kirschner, A.; Kawamura, G.; Tskhakaya, D.; Ding, Rui; Luo, Guangnan; Wang, Dezhen

    2015-01-01

    Highlights: • We study the surface evolution of rough surface on divertor target in fusion devices. • The effects of gyration motion and E × B drift affect 3D angular distribution. • A larger magnetic field angle leads to a reduced net eroded areal density. • The rough surface evolution affects the physical sputtering yield. - Abstract: The 3D Monte-Carlo code SURO has been used to study the surface evolution of rough surface on the divertor target in fusion devices. The edge plasma at divertor region is modelled by the SDPIC code and used as input data for SURO. Coupled with SDPIC, SURO can perform more sophisticated simulations to calculate the local angle and surface evolution of rough surface. The simulation results show that the incident direction of magnetic field, gyration and E × B force has a significant impact on 3D angular distribution of background plasma and accordingly on the erosion of rough surface. The net eroded areal density of rough surface is studied by varying the magnetic field angle with surface normal. The evolution of the microscopic morphology of rough surface can lead to a significant change in the physical sputtering yield

  11. Divertor target profiles and recycling studies in TCV single null lower standard discharges

    International Nuclear Information System (INIS)

    Pitts, R.A.; Nieswand, C.; Weisen, H.

    1996-05-01

    A 'standard', single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of T e and n e before and after He glow. (author) 5 figs., 7 refs

  12. Fabrication of the wing and vertical target dummy armour prototypes of the ITER divertor

    International Nuclear Information System (INIS)

    Grattarola, M.; Bet, M.; Biagiotti, B.; Gandini, G.; Merola, M.; Ottonello, G.B.; Riccardi, B.; Vieider, G.; Zacchia, F.

    2000-01-01

    The dummy armour prototypes are identical to the reference components in terms of geometry, cooling circuit and material except for the armour material, which is replaced by an equivalent thickness of copper alloy. The main objectives of the dummy armour prototypes are the demonstration of the overall engineering concept of the Divertor, the integration in a 3 deg. cassette together with components manufactured by the other ITER Home Teams and the successive thermo-hydraulic tests on the whole Divertor module. This paper describes the realization of both the wing and the vertical target dummy armour prototypes focusing on the critical aspects of the fabrication and their impact on a further industrialization of the components

  13. Fabrication of the wing and vertical target dummy armour prototypes of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Grattarola, M. E-mail: gratta@ari.ansaldo.it; Bet, M.; Biagiotti, B.; Gandini, G.; Merola, M.; Ottonello, G.B.; Riccardi, B.; Vieider, G.; Zacchia, F

    2000-11-01

    The dummy armour prototypes are identical to the reference components in terms of geometry, cooling circuit and material except for the armour material, which is replaced by an equivalent thickness of copper alloy. The main objectives of the dummy armour prototypes are the demonstration of the overall engineering concept of the Divertor, the integration in a 3 deg. cassette together with components manufactured by the other ITER Home Teams and the successive thermo-hydraulic tests on the whole Divertor module. This paper describes the realization of both the wing and the vertical target dummy armour prototypes focusing on the critical aspects of the fabrication and their impact on a further industrialization of the components.

  14. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    International Nuclear Information System (INIS)

    Visca, Eliseo; Roccella, S.; Candura, D.; Palermo, M.; Rossi, P.; Pizzuto, A.; Sanguinetti, G.P.; Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G.

    2015-01-01

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m 2 but the capability to remove up to 20 MW/m 2 during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  15. HRP facility for fabrication of ITER vertical target divertor full scale plasma facing units

    Energy Technology Data Exchange (ETDEWEB)

    Visca, Eliseo, E-mail: eliseo.visca@enea.it [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Roccella, S. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Candura, D.; Palermo, M. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Rossi, P.; Pizzuto, A. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy); Sanguinetti, G.P. [Ansaldo Nucleare S.p.A., Corso Perrone 25, IT-16152 Genova (Italy); Mancini, A.; Verdini, L.; Cacciotti, E.; Cerri, V.; Mugnaini, G.; Reale, A.; Giacomi, G. [Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, IT-00044 Frascati (Roma) (Italy)

    2015-10-15

    Highlights: • R&D activities for the manufacturing of ITER divertor high heat flux plasma-facing components (HHFC). • ENEA and Ansaldo have jointly manufactured several actively cooled monoblock mock-ups and prototypical components. • ENEA and ANSALDO NUCLEARE jointly participate to the European program for the qualification of the manufacturing technology for the ITER divertor IVT. • Successful manufacturing by HRP (Hot Radial Pressing) of first full-scale full-W armored IVT qualification prototype. - Abstract: ENEA and Ansaldo Nucleare S.p.A. (ANN) have being deeply involved in the European development activities for the manufacturing of the ITER Divertor Inner Vertical Target (IVT) plasma-facing components. During normal operation the heat flux deposited on the bottom segment of divertor is 5–10 MW/m{sup 2} but the capability to remove up to 20 MW/m{sup 2} during transient events of 10 s must also be demonstrated. In order to fulfill ITER requirements, ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP). The last challenge is now to fabricate full-scale prototypes of the IVT, aimed to be qualified for the next step, i.e. the series production. On the basis of the experience of manufacturing hundreds of small mock-ups, ENEA designed and installed a new suitable HRP facility. The objective of getting a final shaped plasma facing unit (PFU) that satisfies these requirements is an ambitious target because tolerances set by ITER/F4E are very tight. The setting-up of the equipment started with the fabrication of full scale and representative ‘dummies’ in which stainless steel instead of CFC or W was used for monoblocks. The results confirmed that dimensions were compliant with the required tolerances. The paper reports a brief description of the innovative HRP equipment and the dimensional check results after HRP of the first full-scale full-W PFU.

  16. A supersonic gas target for a bundle divertor plasma

    International Nuclear Information System (INIS)

    Chang, F.R.; Fisher, J.L.

    1982-01-01

    A novel gas target concept for recovering both energy and particles from a high-energy plasma stream is presented. This concept includes the maintenance of a pressure discontinuity by a normal shock and a very high mass flow rate in a relatively small system. The pressure discontinuity allows the exhaust plasma stream to minimize backflow into the plasma, by interacting with the target in a low-pressure region; the high mass flow rate allows exit temperatures that are reasonable from a materials viewpoint and suitable for energy recovery. (author)

  17. Low cycle fatigue behavior of ITER-like divertor target under DEMO-relevant operation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); You, Jeong-Ha, E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2015-01-15

    Highlights: • LCF behavior of the cooling tube and the interlayer of an ITER-like divertor target is studied. • For the cooling tube, LCF failure will not be an issue under an HHF load of up to 18 MW/m{sup 2}. • Plastic strain in the interlayer is concentrated at the free surface edge of the bond interface. • The predicted LCF lifetime of the interlayer may not meet the design requirement. - Abstract: In this work the low cycle fatigue (LCF) behavior of the copper alloy cooling tube and the copper interlayer of an ITER-like divertor target is reported for nine different combinations of loading and cooling conditions relevant to DEMO divertor operation. The LCF lifetime is presented as a function of loading and cooling conditions considered here by means of cyclic plasticity simulation and using LCF data of materials relevant for ITER. The numerical predictions indicate, that fatigue failure will not be an issue for the copper alloy tube under a high heat flux (HHF) load of up to 18 MW/m{sup 2} as long as it preserves its initial strength. In contrast, the copper interlayer exhibits significant plastic dissipation at the free surface edge of the bond interface adjacent to the cooling tube, where the LCF lifetime is predicted to be below 3000 load cycles for HHF loads higher than 15 MW/m{sup 2}. Most of the bulk region of the copper interlayer away from the free surface edge does not experience severe plastic fatigue and hence does not pose any critical concern as the LCF lifetime is predicted to be at least 7000 load cycles. LCF lifetime decreases as HHF load is increased or coolant temperature is decreased.

  18. Testing candidate interlayers for an enhanced water-cooled divertor target

    International Nuclear Information System (INIS)

    Hancock, David; Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William; Rieth, Michael; Reiser, Jens

    2015-01-01

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  19. Testing candidate interlayers for an enhanced water-cooled divertor target

    Energy Technology Data Exchange (ETDEWEB)

    Hancock, David, E-mail: david.hancock@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Barrett, Tom; Foster, James; Fursdon, Mike; Keech, Gregory; McIntosh, Simon; Timmis, William [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, Michael; Reiser, Jens [Karlsruhe Institute of Technology, IAM-AWP, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2015-10-15

    Highlights: • We introduce an optimised divertor target concept: the “Thermal Break”. • We suggest a candidate interlayer material for this concept: FeltMetal. • We describe a bespoke rig for testing the thermal conductivity of this material. • We present preliminary results for a number of samples. - Abstract: The design of a divertor target for DEMO remains one of the most challenging engineering tasks to be overcome on the path to fusion power. Under the European DEMO programme, a promising concept known as Thermal Break has been developed at CCFE. This concept is a variation of the ITER tungsten divertor in which the pure Copper interlayer between Copper Chrome Zirconium coolant pipe and Tungsten monoblock armour is replaced with a low thermal conductivity compliant interlayer, with the aim of reducing the thermal mismatch stress between the armour and structure. One candidate material for this interlayer is FeltMetal™ (Technetics Group, USA). This material consists of an amorphous matrix of fine copper wires which are sintered onto a thin copper foil, creating a sheet of approximately 1 mm thickness. FeltMetal has been successfully used for many years to provide compliant sliding electrical contacts for the MAST TF coils and on ALCATOR C-Mod and extensive material testing has therefore been undertaken to quantify thermal and mechanical properties. These tests, however, have not been performed under vacuum or DEMO-relevant conditions. A bespoke experimental test rig has therefore been designed and constructed with which to measure the interlayer thermal conductance as a function of temperature and pressure under vacuum conditions. The design of this apparatus and the results of experiments on FeltMetal as well as other candidate interlayers are presented here. In parallel, joint mockups using the candidate interlayers have been prepared and Thermal Break divertor target mockups have been manufactured, requiring the development of a dedicated

  20. Damage evaluation under thermal fatigue of a vertical target full scale component for the ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Durocher, A.; Bobin-Vastra, I.; Schedler, B.

    2007-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a Full Scale Vertical Target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses thermographic examination and thermal fatigue testing results obtained on this component. The study includes thermal analysis, with a tentative proposal to evaluate with finite element approach the location/size of defects and the possible propagation during fatigue cycling

  1. Infrared thermography inspection methods applied to the target elements of W7-X divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France)], E-mail: marc.missirlian@cea.fr; Traxler, H. [PLANSEE SE, Technology Center, A-6600 Reutte (Austria); Boscary, J. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching (Germany); Durocher, A.; Escourbiac, F.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint Paul Lez Durance (France); Schedler, B.; Schuler, P. [PLANSEE SE, Technology Center, A-6600 Reutte (Austria)

    2007-10-15

    The non-destructive examination (NDE) method is one of the key issues in developing highly loaded plasma-facing components (PFCs) for a next generation fusion devices such as W7-X and ITER. The most critical step is certainly the fabrication and the examination of the bond between the armour and the heat sink. Two inspection systems based on the infrared thermography methods, namely, the transient thermography (SATIR-CEA) and the pulsed thermography (ARGUS-PLANSEE), are being developed and have been applied to the pre-series of target elements of the W7-X divertor. Results obtained from qualification experiences performed on target elements with artificial calibrated defects allowed to demonstrate the capability of the two techniques and raised the efficiency of inspection to a level which is appropriate for industrial application.

  2. Infrared thermography inspection methods applied to the target elements of W7-X divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Traxler, H.; Boscary, J.; Durocher, A.; Escourbiac, F.; Schlosser, J.; Schedler, B.; Schuler, P.

    2007-01-01

    The non-destructive examination (NDE) method is one of the key issues in developing highly loaded plasma-facing components (PFCs) for a next generation fusion devices such as W7-X and ITER. The most critical step is certainly the fabrication and the examination of the bond between the armour and the heat sink. Two inspection systems based on the infrared thermography methods, namely, the transient thermography (SATIR-CEA) and the pulsed thermography (ARGUS-PLANSEE), are being developed and have been applied to the pre-series of target elements of the W7-X divertor. Results obtained from qualification experiences performed on target elements with artificial calibrated defects allowed to demonstrate the capability of the two techniques and raised the efficiency of inspection to a level which is appropriate for industrial application

  3. Numerical simulation of CFC and tungsten target erosion in ITER-FEAT divertor

    International Nuclear Information System (INIS)

    Filatov, V.

    2003-01-01

    Physical, chemical and thermal surface erosion for water-cooled target armoured by CFC and tungsten is simulated by numerical code ERosion OF Immolated Layer (EROFIL-1). Some calculation results on the CFC and tungsten vertical target (VT) erosion in the ITER-FEAT divertor are presented for various operation modes (normal operations, slow transients, ELMs and disruptions). The main erosion mechanisms of CFC armour are the chemical and sublimation ones. Maximum erosion depth per 3000 cycles during normal operations and slow transients is of 2.7 mm at H phase and of 13.5 mm at DT phase. An evaluation of VT tungsten armour erosion per 3000 cycles of H and DT operations shows that no physical or chemical erosion as well as no melting are expected for tungsten armour at normal operations and slow transients. The tungsten armour melting at 2x10 6 ELMs is not allowable. The 300 disruptions are not dangerous in view of evaporation

  4. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    International Nuclear Information System (INIS)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D.; Driemeyer, D.E.; Kubik, D.L.; Slattery, K.T.; Hellwig, T.H.

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles

  5. Critical heat flux performance of hypervapotrons proposed for use in the ITER divertor vertical target

    Energy Technology Data Exchange (ETDEWEB)

    Youchison, D.L.; Marshall, T.D.; McDonald, J.M.; Lutz, T.J.; Watson, R.D. [Sandia National Labs., Albuquerque, NM (United States); Driemeyer, D.E. Kubik, D.L.; Slattery, K.T.; Hellwig, T.H. [McDonnell Douglas Aerospace, St. Louis, MO (United States)

    1997-09-01

    Task T-222 of the International Thermonuclear Experimental Reactor (ITER) program addresses the manufacturing and testing of permanent components for use in the ITER divertor. Thermalhydraulic and critical heat flux performance of the heat sinks proposed for use in the divertor vertical target are part of subtask T-222.4. As part of this effort, two single channel, medium scale, bare copper alloy, hypervapotron mockups were designed, fabricated, and tested using the EB-1200 electron beam system. The objectives of the effort were to develop the design and manufacturing procedures required for construction of robust high heat flux (HHF) components, verify thermalhydraulic, thermomechanical and critical heat flux (CHF) performance under ITER relevant conditions, and perform analyses of HHF data to identify design guidelines and failure criteria and possibly modify any applicable CHF correlations. The design, fabrication, and finite element modeling of two types of hypervapotrons are described; a common version already in use at the Joint European Torus (JET) and a new attached fin design. HHF test data on the attached fin hypervapotron will be used to compare the CHF performance under uniform heating profiles on long heated lengths with that of localized, highly peaked, off nominal profiles.

  6. Manufacturing and testing of a prototypical divertor vertical target for ITER

    Science.gov (United States)

    Merola, M.; Plöchl, L.; Chappuis, Ph; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G.

    2000-12-01

    After an extensive R&D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m 2, 1000 cycles at 10 MW/m 2 and more then 1000 cycles at 15-20 MW/m 2. The final critical heat flux test reached a value in excess of 30 MW/m 2.

  7. Manufacturing and testing of a prototypical divertor vertical target for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Merola, M. E-mail: merolam@ipp.mpg.de; Ploechl, L.; Chappuis, Ph.; Escourbiac, F.; Grattarola, M.; Smid, I.; Tivey, R.; Vieider, G

    2000-12-01

    After an extensive R and D activity, a medium-scale divertor vertical target prototype has been manufactured by the EU Home Team. This component contains all the main features of the corresponding ITER divertor design and consists of two units with one cooling channel each, assembled together and having an overall length and width of about 600 and 50 mm, respectively. The upper part of the prototype has a tungsten macro-brush armour, whereas the lower part is covered by CFC monoblocks. A number of joining techniques were required to manufacture this component as well as an appreciable effort in the development of suitable non-destructive testing methods. The component was high heat flux tested in FE200 electron beam facility at Le Creusot, France. It endured 100 cycles at 5 MW/m{sup 2}, 1000 cycles at 10 MW/m{sup 2} and more then 1000 cycles at 15-20 MW/m{sup 2}. The final critical heat flux test reached a value in excess of 30 MW/m{sup 2}.

  8. Testing of improved CFC/Cu bondings for the W7-X divertor targets

    International Nuclear Information System (INIS)

    Greuner, H.; Buswirth, B.; Boscary, J.; Tivey, R.; Plankensteiner, A.; Schedler, B.

    2007-01-01

    Full text of publication follows: Extensive high heat flux (HHF) testing of pre-series divertor targets was performed to establish the industrial process for the manufacturing of 890 targets, which will be needed for the installation of the Wendelstein 7-X (W7-X) divertor. The target design consists of flat tiles of CFC NB31 as plasma facing material bonded by an Active Meta] Casting copper (AMC) interlayer onto a water-cooled CuCrZr structure. This design is required by the specific geometrical requirements of the W7-X divertor. The heat removal capability of this target concept has been demonstrated for the envisaged operational power load of 10 MW/m 2 in previous test series of more than 30 full-scale elements. No large detachment or loss of CFC tiles occurred during cyclic loading tests at 10.5 and 13 MW/m 2 , but growing local de-bonded zones at the free edges of several CFC tiles were observed. Therefore a detailed analysis of the system of CFC/Cu bonding was carried out with respect to a further reduction of the stress at the CFC/Cu interface. Based on the results of the 3/D non-linear thermomechanical FEM analysis of the CFC/Cu interface a set of 17 additional pre-series elements was manufactured by PLANSEE SE. Three types of design variations have been investigated: - adopting an additional plastically compliant Cu interlayer between the cooling structure and the AMC region, - reduced size of CFC tiles, - arrangement of tiles with 90 deg. rotation of the CFC fibre plane. HHF tests were performed in the ion beam test facility GLADIS at IPP Garching with up to 3000 cycles at 10.5 MW/m 2 on this elements. The aim of these tests is to investigate the crack propagation between CFC/Cu and to define the acceptable defect size after 100 HHF cycles as an acceptance criterion for the series manufacturing. The applied criterion should allow the selection of elements for W7-X expected to achieve a suitable operational life time. Finally, the design variant with the

  9. Assessment of erosion of the ITER divertor targets during type I ELMs

    Science.gov (United States)

    Federici, G.; Loarte, A.; Strohmayer, G.

    2003-09-01

    This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties. The cases analysed clarify the influence of several ELM parameters on the heat transfer and erosion processes at the target (i.e. characteristic plasma ELM energy loss from the pedestal, fraction of the energy reaching the divertor, broadening of the strike-points during ELMs, duration and waveform of the ELM heat load) and design/material parameters (i.e. inclination of the target, type and thickness of the armour material, and for tungsten only, fraction of the melt layer loss). Comparison is made between cases where all ELMs are characterized by the same fixed averaged parameters, and cases where instead the characteristic parameters of each ELM are evaluated in a random fashion by using a standard Monte Carlo technique, based on distributions of some of the variables of interest derived from experiments in today's machines. Although uncertainties rule out providing firm quantitative predictions, the results of this study are useful to illustrate trends. Based on the results, the implications on the design and operation are discussed and priorities are determined for the R&D needed to reduce the remaining uncertainties.

  10. Assessment of erosion of the ITER divertor targets during type I ELMs

    International Nuclear Information System (INIS)

    Federici, G; Loarte, A; Strohmayer, G

    2003-01-01

    This paper presents the results of a preliminary assessment conducted to estimate the thermal response and erosion lifetime of the ITER divertor targets clad either with carbon-fibre composite or tungsten during type I ELMs. The one-dimensional thermal/erosion model, used for the analyses, is briefly described. It includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the bulk thermal response, and it is based on an implicit finite-difference scheme, which allows for temperature-dependent material properties. The cases analysed clarify the influence of several ELM parameters on the heat transfer and erosion processes at the target (i.e. characteristic plasma ELM energy loss from the pedestal, fraction of the energy reaching the divertor, broadening of the strike-points during ELMs, duration and waveform of the ELM heat load) and design/material parameters (i.e. inclination of the target, type and thickness of the armour material, and for tungsten only, fraction of the melt layer loss). Comparison is made between cases where all ELMs are characterized by the same fixed averaged parameters, and cases where instead the characteristic parameters of each ELM are evaluated in a random fashion by using a standard Monte Carlo technique, based on distributions of some of the variables of interest derived from experiments in today's machines. Although uncertainties rule out providing firm quantitative predictions, the results of this study are useful to illustrate trends. Based on the results, the implications on the design and operation are discussed and priorities are determined for the R and D needed to reduce the remaining uncertainties

  11. Results and analysis of high heat flux tests on a full-scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Merola, M.; Bobin-Vastra, I.; Schlosser, J.; Durocher, A.

    2005-01-01

    After an extensive R and D development program, a full-scale divertor target prototype, manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full monoblock geometry. The lower part (CFC armour) and the upper part (W armour) of each monoblock were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. This paper summarises and analyses the main test results obtained on this prototype

  12. Turbine airfoil having outboard and inboard sections

    Science.gov (United States)

    Mazzola, Stefan; Marra, John J

    2015-03-17

    A turbine airfoil usable in a turbine engine and formed from at least an outboard section and an inboard section such that an inner end of the outboard section is attached to an outer end of the inboard section. The outboard section may be configured to provide a tip having adequate thickness and may extend radially inward from the tip with a generally constant cross-sectional area. The inboard section may be configured with a tapered cross-sectional area to support the outboard section.

  13. Design improvement of the target elements of Wendelstein 7-X divertor

    International Nuclear Information System (INIS)

    Boscary, J.; Peacock, A.; Friedrich, T.; Greuner, H.; Böswirth, B.; Tittes, H.; Schulmeyer, W.; Hurd, F.

    2012-01-01

    Highlights: ► Improvement of the cooling structure design. ► Improvement of the CFC tile arrangement at the element end. ► Design and fabrication validated with high heat flux testing. ► Selected solution removes stationary heat load of 5 MW/m 2 and 2 MW/m 2 on the top and on the side facing the pumping gap of the element, respectively. - Abstract: The actively cooled high-heat flux divertor of the Wendelstein 7-X stellarator consists of individual target elements made of a water-cooled CuCrZr copper alloy heat sink armored with CFC tiles. The so-called “bi-layer” technology developed in collaboration with the company Plansee for the bonding of the tiles onto the heat sink has reliably demonstrated the removal of the specified heat load of 10 MW/m 2 in the central area of the divertor. However, due to geometrical constraints, the loading performance at the ends of the elements is reduced compared to the central part. Design modifications compatible with industrial processes have been made to improve the cooling capabilities at this location. These changes have been validated during test campaigns of full-scale prototypes carried out in the neutral beam test facility GLADIS. The tested solution can remove reliably the stationary heat load of 5 MW/m 2 and 2 MW/m 2 on the top and on the side of the element, respectively. The results of the testing allowed the release of the design and fabrication processes for the next manufacturing phase of the target elements.

  14. Spectroscopic measurement of target plate erosion in the ASDEX Upgrade divertor

    Energy Technology Data Exchange (ETDEWEB)

    Filed, A R; Garcia-Rosales, C; Lieder, G; Pitcher, C S; Radtke, R [Association Euratom-Max-Planck-Institut fuer Plasmaphysik, Garching (Germany); ASDEX Upgrade Team

    1996-02-01

    The erosion of the graphite divertor plates in the ASDEX Upgrade tokamak is measured spectroscopically. Spatial profiles of the D{sup 0} and C{sup +} influxes across the outer target plate are determined from measured absolute line intensities. Plasma parameters (n{sub e}, T{sub e}) at the target, which are required to determine the appropriate photon emission efficiencies for these lines, are obtained from an in-vessel reciprocating Langmuir probe above the target plate. Yields for the erosion of the graphite by the incident D{sup +} flux are determined from the ratio of the measured C{sup +} to D{sup 0} fluxes. Over a range of moderate densities the measured yields of {<=} 4% are explicable in terms of physical sputtering alone. Chemical sputtering by low energy Franck-Condon neutrals probably contributes, however, to the total erosion. At higher densities detachment of the plasma from the targets occurs owing to formation of a MARFE near the X point. Under these conditions localized physical sputtering of the targets ceases. The impurity level (Z{sub eff}) is, however, maintained following detachment, indicating a corresponding maintenance of carbon influx, perhaps due to chemical erosion of the total graphite surface and/or an improvement in particle confinement in the detached state. (author). 26 refs, 14 figs, 1 tab.

  15. Development of residual thermal stress-relieving structure of CFC monoblock target for JT-60SA divertor

    Energy Technology Data Exchange (ETDEWEB)

    Tsuru, Daigo, E-mail: tsuru.daigo@jaea.go.jp; Sakurai, Shinji; Nakamura, Shigetoshi; Ozaki, Hidetsugu; Seki, Yohji; Yokoyama, Kenji; Suzuki, Satoshi

    2015-10-15

    Highlights: • We carried out numerical simulations on residual thermal stress of targets for the JT-60SA divertor. • We developed three measures to reduce residual thermal stress. • We proposed two structures of CFC monoblock target for the JT-60SA divertor. • We confirmed the effectiveness of the structure by infrared thermography inspection and high heat flux test. - Abstract: Carbon fibre-reinforced carbon composite (CFC) monoblock target for JT-60SA divertor is under development towards the mass-production. CFC monoblocks, WCu interlayers and a CuCrZr cooling tube at the centre of the monoblocks were bonded by vacuum brazing in a high temperature, to a target. If residual thermal stress due to difference of thermal expansions between CFC and CuCrZr exceeds the maximum allowable stress of the CFC after the bonding, cracks are generated in the CFC monoblock and heat removal capacity of the target degrades. In this paper, new structures of the targets were proposed, to reduce residual thermal stress and to mitigate the degradation of heat removal capacity of the targets. Some measures, including slitting of the CFC monoblock aside of the cooling tube, replacement of the interlayer material and shifting the position of the cooling tube, were implemented. The effectiveness of the measures was evaluated by numerical simulations. Target mock-ups with the proposed structures were manufactured. Infrared thermography inspection and high heat flux test were carried out on the mock-ups in order to evaluate the heat removal capacity.

  16. Plasma shape control calculations for BPX divertor design

    International Nuclear Information System (INIS)

    Strickler, D.J.; Neilson, G.H.; Jardin, S.C.; Pomphrey, N.

    1991-01-01

    The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs

  17. Thermographic observation of the divertor target plates in the stellarators W7-AS and W7-X

    International Nuclear Information System (INIS)

    Hildebrandt, D.; Gadelmeier, F.; Grigull, P.; McCormick, K.; Naujoks, D.; Suender, D.

    2003-01-01

    Thermography is applied on the stellarator W7-AS to monitor the thermal load of the recently installed divertor targets. A three dimensional numerical code was developed to evaluate power fluxes arriving at the targets from the measured temporal evolution of the surface temperature distribution. Values of the thermal conductivity of the used CFC-target material for all three directions are required for this evaluation and determined by observing the propagation of controlled heat pulses applied by an infrared laser. The evaluation of the thermographic measurements during plasma operation shows characteristic spatial and temporal features of the arrived heat fluxes. Significant features in high density regimes like plasma detachment from the divertor target plates or strongly enhanced localised plasma radiation (MARFE) has been observed by the installed infrared cameras. The implications of these observations for the thermographic system for W7-X are shortly addressed

  18. Infrared thermography inspection methods applied to the target elements of W7-X Divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Durocher, A.; Schlosser, J.; Farjon, J.-L.; Vignal, N.; Traxler, H.; Schedler, B.; Boscary, J.

    2006-01-01

    As heat exhaust capability and lifetime of plasma-facing component (PFC) during in-situ operation are linked to the manufacturing quality, a set of non-destructive testing must be operated during R-and-D and manufacturing phases. Within this framework, advanced non-destructive examination (NDE) methods are one of the key issues to achieve a high level of quality and reliability of joining techniques in the production of high heat flux components but also to develop and built successfully PFCs for a next generation of fusion devices. In this frame, two NDE infrared thermographic approaches, which have been recently applied to the qualification of CFC target elements of the W7-X divertor during the first series production will be discussed in this paper. The first one, developed by CEA (SATIR facility) and used with successfully to the control of the mass-produced actively cooled PFCs on Tore Supra, is based on the transient thermography where the testing protocol consists in inducing a thermal transient within the heat sink structure by an alternative hot/cold water flow. The second one, recently developed by PLANSEE (ARGUS facility), is based on the pulsed thermography where the component is heated externally by a single powerful flash of light. Results obtained on qualification experiences performed during the first series production of W7-X divertor components representing about thirty mock-ups with artificial and manufacturing defects, demonstrated the capabilities of these two methods and raised the efficiency of inspection to a level which is appropriate for industrial application. This comparative study, associated to a cross-checking analysis between the high heat flux performance tests and these inspection methods by infrared thermography, showed a good reproducibility and allowed to set a detectable limit specific at each method. Finally, the detectability of relevant defects showed excellent coincidence with thermal images obtained from high heat flux

  19. The MAST improved divertor

    International Nuclear Information System (INIS)

    Darke, A.C.; Hayward, R.J.; Counsell, G.F.; Hawkins, K.

    2005-01-01

    The Mega Amp Spherical Tokamak (MAST) at Culham is one of the leading world machines studying the spherical tokamak (ST) concept. At the time of the initial construction in 1998 little was known about the sort of divertor structures that would be required in an ST. The machine was therefore provided with relatively rudimentary structures that were designed mostly to protect important components from the hot plasma. While these have served the machine well it was accepted that they might not be suitable when operating MAST to its full potential. The years of experience of operating MAST have led to the design, manufacture and now installation of a new divertor, the MAST improved divertor (MID), that should be able to cope with the full performance of the machine. The design is based on imbricated (fan-shaped) disks of tiles at the top and bottom of the machine for the outer strike points, giving an excellent compromise between power handling and diagnostic access, with substantial new centre column strike point armour and a shaped plate in between. High purity graphite is chosen as the plasma facing material in preference to CFC since in this case it has a better balance of performance and cost. The lower imbricated disk is insulated in alternate sectors for studies of divertor biasing and extensive diagnostics and additional inboard gas injection are included

  20. Engineering design of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1995-10-01

    A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

  1. Estimation of the tritium retention in ITER tungsten divertor target using macroscopic rate equations simulations

    Science.gov (United States)

    Hodille, E. A.; Bernard, E.; Markelj, S.; Mougenot, J.; Becquart, C. S.; Bisson, R.; Grisolia, C.

    2017-12-01

    Based on macroscopic rate equation simulations of tritium migration in an actively cooled tungsten (W) plasma facing component (PFC) using the code MHIMS (migration of hydrogen isotopes in metals), an estimation has been made of the tritium retention in ITER W divertor target during a non-uniform exponential distribution of particle fluxes. Two grades of materials are considered to be exposed to tritium ions: an undamaged W and a damaged W exposed to fast fusion neutrons. Due to strong temperature gradient in the PFC, Soret effect’s impacts on tritium retention is also evaluated for both cases. Thanks to the simulation, the evolutions of the tritium retention and the tritium migration depth are obtained as a function of the implanted flux and the number of cycles. From these evolutions, extrapolation laws are built to estimate the number of cycles needed for tritium to permeate from the implantation zone to the cooled surface and to quantify the corresponding retention of tritium throughout the W PFC.

  2. Comparative studies of liquid metals for an alternative divertor target in a fusion reactor

    Science.gov (United States)

    Tabarés, F. L.; Oyarzabal, E.; Tafalla, D.; Martin-Rojo, A. B.; Pastor, I.; Ochando, M. A.; Medina, F.; Zurro, B.; McCarthy, K. J.; the TJ-II Team

    2017-12-01

    Two liquid metals (LM), Li and LiSn (20:80 at), presently considered as alternative materials for the divertor target of a fusion reactor, have been exposed to the plasma in a capillary porous system (CPS) arrangement in TJ-II. A negligible perturbation of the plasma has been recorded in both cases, even when stellarator plasmas are particularly sensitive to high Z elements due to the tendency to central impurity accumulation. The surface temperature of the LM CPS samples (made of a tungsten mesh impregnated in SnLi or Li) has been measured during the plasma pulse with ms resolution by pyrometry and the thermal balance during heating and cooling has been used to obtain the thermal parameters of the SnLi and Li CPS arrangements. Temperatures as high as 1150 K during TJ-II plasma exposure were observed for the LiSn solid case. Strong changes in the thermal conductivity of the alloy were recorded in the cooling phase at temperatures close to the nominal melting point. The deduced values for the thermal conductivity of the LiSn alloy/CPS sample were significantly lower than those predicted from their individual components.

  3. Scrape-off layer ion temperature measurements at the divertor target during type III ELMs in MAST measured by RFEA

    Energy Technology Data Exchange (ETDEWEB)

    Elmore, S., E-mail: Sarah.Elmore@ccfe.ac.uk [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Allan, S.Y.; Fishpool, G.; Kirk, A. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Kočan, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Tamain, P. [Association Euratom-CEA, CEA/DSM/IRFM, CEA-Cadarache, F-13108 St Paul-lez-Durance Cedex (France); Thornton, A.J. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2015-08-15

    Edge-localised modes (ELMs) can carry significant fractions of their energy as far as main chamber plasma-facing components in divertor tokamaks. Since in future devices (e.g. ITER, DEMO) these energies could cause issues for material lifetime and impurity production, the energy and temperature of ions in ELMs needs to be investigated. In MAST, novel divertor measurements of T{sub i} during ELMs have been made using the divertor retarding field energy analyser (RFEA) probe. These measurements have shown instantaneous ion energy distributions corresponding to an effective T{sub i} at 5 cm from the strike point at the target that can be as high as 60 eV and that this decreases with time after the ELM start. This is consistent with the hottest, fastest ions arriving at the target first by parallel transport, followed by the lower end of the ion energy distribution. This analysis will form a basis for future data analysis of fast swept measurements of ion distributions in ELMs.

  4. Exploring the engineering limit of heat flux of a W/RAFM divertor target for fusion reactors

    Science.gov (United States)

    Mao, X.; Fursdon, M.; Chang, X. B.; Zhang, J. W.; Liu, P.; Ellwood, G.; Qian, X. Y.; Qin, S. J.; Peng, X. B.; Barrett, T. R.; Liu, P.

    2018-06-01

    The design and development of a fusion reactor divertor plasma facing component (PFC) is one of the many challenging issues on the road to commercial use of fusion energy. The divertor PFC is expected to exhaust steady state heat loads in the region of 10 MW m‑2 while keeping temperatures and thermo-mechanical stresses in its structure within the allowable limits. For ITER (International Thermo-Nuclear Experimental Reactor) a water cooled W/CuCrZr divertor PFC concept has been developed. However, this concept is not necessarily assured for use in future fusion reactors mainly because the neutron radiation dose would be at least an order magnitude higher, resulting in limited thermo-mechanical performance and considerably more activated waste products. In the present study, a water cooled divertor PFC using reduced activation ferritic-martensitic (RAFM) steel as the heat sink pipe has been designed with pressurised water reactor-like cooling conditions (pressure of 15.5 MPa, velocity of 10–20 m s‑1 and temperature of 300 °C). The PFC is made up of a number of rectangular tungsten tiles, each with an inner circular hole (so-called monoblocks), joined onto a RAFM steel pipe with copper interlayers. The thermo-mechanical performance of the PFC has been studied in detail. The heat transfer coefficient between the RAFM pipe inner surface and the water was calculated using published correlations. Geometric parameters and water velocity were optimized with finite element (FE) thermal analysis, to achieve acceptable temperatures in the structure given the target exhaust heat load of 10 MW m‑2. Under this heat load and the optimised thermal design parameters, the structure of the PFC was further assessed by mechanical analysis. We find that under these conditions the RAFM steel pipe experiences cyclic plasticity, and fails the common linear elastic ratchetting (3 Sm) rule. Nevertheless, the designed W/RAFM divertor PFU can withstand 10 MW m‑2 heat load, albeit

  5. Particle and power deposition on divertor targets in EAST H-mode plasmas

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.

    2012-01-01

    ELMs were chosen for analysis in order to reduce the uncertainty resulting from the influence of fast electrons on Langmuir triple-probe evaluation during ELMs. The power deposition obtained from Langmuir triple probes was consistent with that from the divertor infra-red camera during an ELM...

  6. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A.; Bobin-Vastra, I.

    2004-01-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m 2 without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m 2 . The W mono-block endured ∼600 cycles at 10 MW/m 2 . This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m 2 . A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  7. Results and analysis of high heat flux tests on a full scale vertical target prototype of ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M.; Escourbiac, F.; Schlosser, J.; Durocher, A. [Association Euratom-CEA, Centre d' Etudes de Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [EFDA Close Support Unit, Garching (Germany); Bobin-Vastra, I. [Framatome, 71 - Le Creusot (France)

    2004-07-01

    After an extensive development program, a Full-Scale Divertor Target prototype (VTFS) manufactured with all the main features of the corresponding ITER divertor, was intensively tested in the high heat flux FE200 facility. The prototype consists of four units having a full mono-block geometry. The lower part (CFC armour) and the upper part (W armour) of each mono-block were joined to the solution annealed, quenched and cold worked CuCrZr tube by HIP technique. The CFC mono-block was successfully tested up to 1000 cycles at 23 MW/m{sup 2} without any indication of failure. This value is well beyond the ITER design target of 300 cycles at 20 MW/m{sup 2}. The W mono-block endured {approx}600 cycles at 10 MW/m{sup 2}. This value of flux is one order of magnitude higher than the ITER design target for the upper part of the vertical target. Fatigue damage is observed when pursuing the cycling up to 15 MW/m{sup 2}. A first stress analysis seems to predict these factual results. However, macro-graphic examinations should bring a better damage valuation. Meanwhile, the fatigue testing will continue on the W healthy part of the VTFS prototype with castellation located on the heated surface (reducing the stresses close to the W-Cu interface). (authors)

  8. Structural impact of armor monoblock dimensions on the failure behavior of ITER-type divertor target components: Size matters

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; You, Jeong-Ha, E-mail: you@ipp.mpg.de

    2016-12-15

    Highlights: • Quantitative assessment of size effects was conducted numerically for W monoblock. • Decreasing the width of W monoblock leads to a lower risk of failure. • The Cu interlayer was not affected significantly by varying armor thickness. • The predicted trends were in line with the experimental observations. - Abstract: Plenty of high-heat-flux tests conducted on tungsten monoblock type divertor target mock-ups showed that the threshold heat flux density for cracking and fracture of tungsten armor seems to be related to the dimension of the monoblocks. Thus, quantitative assessment of such size effects is of practical importance for divertor target design. In this paper, a computational study about the thermal and structural impact of monoblock size on the plastic fatigue and fracture behavior of an ITER-type tungsten divertor target is reported. As dimensional parameters, the width and thickness of monoblock, the thickness of sacrificial armor, and the inner diameter of cooling tube were varied. Plastic fatigue lifetime was estimated for the loading surface of tungsten armor and the copper interlayer by use of a cyclic-plastic constitutive model. The driving force of brittle crack growth through the tungsten armor was assessed in terms of J-integral at the crack tip. Decrease of the monoblock width effectively reduced accumulation of plastic strain at the armor surface and the driving force of brittle cracking. Decrease of sacrificial armor thickness led to decrease of plastic deformation at the loading surface due to lower surface temperature, but the thermal and mechanical response of the copper interlayer was not affected by the variation of armor thickness. Monoblock with a smaller tube diameter but with the same armor thickness and shoulder thickness experienced lower fatigue load. The predicted trends were in line with the experimental observations.

  9. Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

    2016-07-01

    An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

  10. Finite Element Based Design Optimization of WENDELSTEIN 7-X Divertor Targets

    International Nuclear Information System (INIS)

    Plankensteiner, A.; Leuprecht, A.; Schedler, B.; Scheiber, K.; Greuner, H.

    2006-01-01

    In the fusion experiment WENDELSTEIN 7-X divertor plasma facing components have to withstand severe loading conditions. In general thermally induced mechanical stressing turns out to be most critical with respect to life time predictions of the component. In the specific case flat tiles of CFC grade NB31 are joined to the precipitation hardened CuCrZr heat sink by employing an active metal cast (AMC)-Cu as an interlayer between CFC and CuCrZr. Residual stresses resulting from the manufacturing process act as initial stresses in the subsequent operational heat flux loading. For the latter loading regime these stresses intrinsically are generated due to the large contrast in the CTE for CFC and Cu. Different design variants of those CFC flat tile armoured target elements have been analysed via the finite element package ABAQUS aiming at derivation of an optimized component design. The numerical study comprises variants with different degrees of tessellation of the CFC flat tile section, orientation of the CFC, lamellar design of the AMC-interlayer, and different designs of the cooling channels. The thermo-mechanical material characteristics are accounted for the finite element models with elastic-plastic properties being assigned to the metallic sections CuCrZr and AMC-Cu, respectively, and orthotropic nonlinear-elastic properties being used to the CFC section. The latter has been realized in form of a user-defined material subroutine that is used at the integration point level of the finite element model. In particular, twelve scalar-type damage parameters obeying their own evolution equations with respect to the loading history account for specific stress-strain relationships in the three principal material directions and planes with six damage parameters being used for normal loading under tensile and compressive stress states, respectively, and six parameters being used for shear loading. For the aim of model verification calculated surface temperatures, global

  11. Ion orbit modelling of ELM heat loads on ITER divertor vertical targets.

    Czech Academy of Sciences Publication Activity Database

    Gunn, J. P.; Carpentier-Chouchana, S.; Dejarnac, Renaud; Escourbiac, F.; Hirai, T.; Komm, Michael; Kukushkin, A.; Panayotis, S.; Pitts, R.A.

    2017-01-01

    Roč. 12, August (2017), s. 75-83 ISSN 2352-1791. [International Conference on Plasma Surface Interactions 2016, PSI2016 /22./. Roma, 30.05.2016-03.06.2016] Institutional support: RVO:61389021 Keywords : ITER * Divertor * ELM heat loads Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) http://www.sciencedirect.com/science/article/pii/S2352179116302745

  12. Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

    Energy Technology Data Exchange (ETDEWEB)

    Budaev, V. P., E-mail: budaev@mail.ru [National Research Centre Kurchatov Institute (Russian Federation)

    2016-12-15

    Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production of submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.

  13. 'EU divertor celebration day'

    International Nuclear Information System (INIS)

    Merola, M.

    2002-01-01

    The meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria was held on the occasion of the completion of manufacturing activities of a complete set of near full-scale prototypes of divertor components including the vertical target, the dome liner and the cassette body. About 30 participants attended the meeting including Dr. Robert Aymar, ITER Director, representatives from EFDA, CEA, ENEA, IPP and others

  14. Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Courtois, X., E-mail: xavier.courtois@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, F. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint-Paul-Lez-Durance (France); Richou, M.; Cantone, V. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Constans, S. [AREVA-NP, Le Creusot (France)

    2013-10-15

    Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear. Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.

  15. Engineering design of a Radiative Divertor for DIII-D

    International Nuclear Information System (INIS)

    Smith, J.P.; Allen, S.L.; Anderson, P.M.; Baxi, C.B.; Chin, E.; Fenstermacher, M.E.; Hill, D.N.; Hollerbach, M.A.; Hyatt, A.W.; Junge, R.; Mahdavi, M.A.; Porter, G.D.; Redler, K.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.

    1995-01-01

    A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D 2 ) or the core Z eff (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (>0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important, as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined. (orig./WL)

  16. Thermal-hydraulic tests on net divertor targets using swirl tubes

    International Nuclear Information System (INIS)

    Schlosser, J.; Chappuis, P.; Deschamps, P.; Massmann, P.; Falter, H.D.; Deschamps, G.H.

    1991-01-01

    Thermal-hydraulic tests have been carried out in collaboration between NET, CEA Cadarache and JET in order to find a cooling method capable of removing the high heat fluxes expected for the NET/ITER divertor. The goal was to evaluate by experiments the critical heat flux (CHF) and heat transfer in the subcooled boiling regime using twisted tapes as turbulence promoters and testing them under relevant thermal-hydraulic conditions. The CEA 200 kW Electron Beam (EB) facility and the 10 MW JET Neutral Beam (NB) test bed have been used to heat up the NET relevant test sections (TS) consisting of rectangular copper elements with circular internal channels. The TS have been exposed to the electron or ion beams under normal incidence. This paper reports the results of the experiments and of thermal analyses performed in support of the tests. The experimental CHF values have been benchmarked with the Tong-75 correlation

  17. Fracture mechanical analysis of tungsten armor failure of a water-cooled divertor target

    Energy Technology Data Exchange (ETDEWEB)

    Li, Muyuan; Werner, Ewald [Lehrstuhl für Werkstoffkunde und Werkstoffmechanik, Technische Universität München, Boltzmannstr. 15, 85748 Garching (Germany); You, Jeong-Ha, E-mail: you@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, 85748 Garching (Germany)

    2014-11-15

    Highlights: • The FEM-based VCE method and XFEM were employed for computing K{sub I} (or J-integral) and predicting progressive cracking, respectively. • The most probable pattern of crack formation is radial cracking in the tungsten armor block. • The most probable site of cracking is the upper interfacial region of the tungsten armor block adjacent to the top position of the copper interlayer. • The initiation of a major crack becomes likely, only when the strength of tungsten armor block is significantly reduced from its original strength. - Abstract: The inherent brittleness of tungsten at low temperature and the embrittlement by neutron irradiation are its most critical weaknesses for fusion applications. In the current design of the ITER and DEMO divertor, the high heat flux loads during the operation impose a strong constraint on the structure–mechanical performance of the divertor. Thus, the combination of brittleness and the thermally induced stress fields due to the high heat flux loads raises a serious reliability issue in terms of the structural integrity of tungsten armor. In this study, quantitative estimates of the vulnerability of the tungsten monoblock armor cracking under stationary high heat flux loads are presented. A comparative fracture mechanical investigation has been carried out by means of two different types of computational approaches, namely, the extended finite element method (XFEM) and the finite element method (FEM)-based virtual crack tip extension (VCE) method. The fracture analysis indicates that the most probable pattern of crack formation is radial cracking in the tungsten armor starting from the interface to tube and the most probable site of cracking is the upper interfacial region of the tungsten armor adjacent to the top position of the copper interlayer. The strength threshold for crack initiation and the high heat flux load threshold for crack propagation are evaluated based on XFEM simulations and computations

  18. A review of progress towards radiative divertor

    International Nuclear Information System (INIS)

    Zagorski, Roman

    1997-07-01

    A solution of the problem of the power and particle exhaust from the next step tokamaks, will require new techniques which redistribute the power entering the SOL onto much larger surface area than conventional divertor design permits, while maintaining good impurity retention in divertor volume and allowing for efficient helium pumping. Progress made in developing such techniques is discussed. Status of the modelling studies of dynamic gas target divertor and impurity seeded radiating divertors is presented. Recent results of experiments on radiative and gas target divertors are reviewed

  19. The influence of plasma-surface interaction on the performance of tungsten at the ITER divertor vertical targets

    Science.gov (United States)

    De Temmerman, G.; Hirai, T.; Pitts, R. A.

    2018-04-01

    The tungsten (W) material in the high heat flux regions of the ITER divertor will be exposed to high fluxes of low-energy particles (e.g. H, D, T, He, Ne and/or N). Combined with long-pulse operations, this implies fluences well in excess of the highest values reached in today’s tokamak experiments. Shaping of the individual monoblock top surface and tilting of the vertical targets for leading-edge protection lead to an increased surface heat flux, and thus increased surface temperature and a reduced margin to remain below the temperature at which recrystallization and grain growth begin. Significant morphology changes are known to occur on W after exposure to high fluences of low-energy particles, be it H or He. An analysis of the formation conditions of these morphology changes is made in relation to the conditions expected at the vertical targets during different phases of operations. It is concluded that both H and He-related effects can occur in ITER. In particular, the case of He-induced nanostructure (also known as ‘fuzz’) is reviewed. Fuzz formation appears possible over a limited region of the outer vertical target, the inner target being generally a net Be deposition area. A simple analysis of the fuzz growth rate including the effect of edge-localized modes (ELMs) and the reduced thermal conductivity of fuzz shows that the fuzz thickness is likely to be limited by the occurrence of annealing during ELM-induced thermal excursions. Not only the morphology, but the material mechanical and thermal properties can be modified by plasma exposure. A review of the existing literature is made, but the existing data are insufficient to conclude quantitatively on the importance and extent of these effects for ITER. As a consequence of the high surface temperatures in ITER, W recrystallization is an important effect to consider, since it leads to a decrease in material strength. An approach is proposed here to develop an operational budget for the W material, i

  20. Upgraded acceptance criteria from transient thermography control for the W7-X divertor target elements

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)], E-mail: marc.missirlian@cea.fr; Boscary, J. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching (Germany); Guigon, R.; Schlosser, J.; Durocher, A. [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Greuner, H. [Max-Planck-Institut fuer Plasmaphysik, Euratom Association, Boltzmannstr. 2, D-85748 Garching (Germany)

    2009-06-15

    The commissioning of plasma-facing component fields needs advanced non-destructive methods to detect in a reliable way the defects, which can impair the component performances and/or integrity during operation. Within this framework, CEA developed a dedicated non-destructive examination method based on active infrared thermography (SATIR facility) to inspect the bonding between armour material and metallic heat sink. Used with successful in the commissioning of the toroidal pump limiter of Tore Supra, this technique was applied in the frame of the pre-series activities of the Wendelstein 7-X high heat flux divertor elements to assess the bonding quality of the delivered components. This paper presents the methodology adopted to define an acceptance criterion based on SATIR test bed possibly applied for a serial inspection of the Wendelstein 7-X elements. Using the well-tried acceptance test based on the DTref{sub m}ax parameter, the new method includes advanced data post-processing techniques from thermo-signal SATIR and a data merging method to help the decision-making and to optimise the reliability of the binary response expected for a final decision in terms of acceptance test.

  1. Upgraded acceptance criteria from transient thermography control for the W7-X divertor target elements

    International Nuclear Information System (INIS)

    Missirlian, M.; Boscary, J.; Guigon, R.; Schlosser, J.; Durocher, A.; Greuner, H.

    2009-01-01

    The commissioning of plasma-facing component fields needs advanced non-destructive methods to detect in a reliable way the defects, which can impair the component performances and/or integrity during operation. Within this framework, CEA developed a dedicated non-destructive examination method based on active infrared thermography (SATIR facility) to inspect the bonding between armour material and metallic heat sink. Used with successful in the commissioning of the toroidal pump limiter of Tore Supra, this technique was applied in the frame of the pre-series activities of the Wendelstein 7-X high heat flux divertor elements to assess the bonding quality of the delivered components. This paper presents the methodology adopted to define an acceptance criterion based on SATIR test bed possibly applied for a serial inspection of the Wendelstein 7-X elements. Using the well-tried acceptance test based on the DTref m ax parameter, the new method includes advanced data post-processing techniques from thermo-signal SATIR and a data merging method to help the decision-making and to optimise the reliability of the binary response expected for a final decision in terms of acceptance test.

  2. Cyclic heat load testing of improved CFC/Cu bonding for the W 7-X divertor targets

    International Nuclear Information System (INIS)

    Greuner, H.; Boeswirth, B.; Boscary, J.; Chaudhuri, P.; Schlosser, J.; Friedrich, T.; Plankensteiner, A.; Tivey, R.

    2009-01-01

    Extensive high heat flux cycling testing of pre-series targets was performed in the neutral beam facility GLADIS to establish the industrial process for the manufacturing of 890 targets, which will be needed for the installation of the WENDELSTEIN 7-X divertor. The targets are manufactured of flat tiles of CFC NB31 as plasma facing material bonded by an Active Metal Casting copper interlayer onto a water-cooled CuCrZr structure. Based on the results of the 3D thermo-mechanical FEM analysis of the CFC/Cu interface, an additional set of 17 full-scale pre-series elements including three design variations was manufactured by PLANSEE SE. The insertion of an additional plastically compliant copper interlayer between the cooling structure and the Active Metal Casting interlayer showed the best results. No critical tile detachment was observed during >5000 cycles at 10 MW/m 2 . These results demonstrated the sufficient life time of the component for the expected heat load in operation.

  3. The effects of field reversal on the Alcator C-Mod divertor

    International Nuclear Information System (INIS)

    Hutchinson, I.H.; LaBombard, B.; Goetz, J.A.; Lipschultz, B.; McCracken, G.M.; Snipes, J.A.; Terry, J.L.

    1995-01-01

    Imbalances between the inboard and outboard legs of the single null divertor in tokamak Alcator C-Mod are observed to reverse when the direction of the toroidal field is reversed. These imbalances are measured by embedded probes in the target plates, tomographic reconstructions of bolometry and line radiation, and visible imaging. Density imbalances of about a factor of ten at the targets are observed at moderate density, decreasing as the density is raised until they are almost balanced. The data indicate that the electron pressure is not imbalanced, thus arguing against momentum imbalance as the cause of these drift-induced effects. Instead, power flux imbalance caused by E r ''and'' B convection, and enhanced by radiation, is suggested as the underlying cause. (Author)

  4. High temperature divertor plasma operation

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi.

    1991-02-01

    High temperature divertor plasma operation has been proposed, which is expected to enhance the core energy confinement and eliminates the heat removal problem. In this approach, the heat flux is guided through divertor channel to a remote area with a large target surface, resulting in low heat load on the target plate. This allows pumping of the particles escaping from the core and hence maintaining of the high divertor temperature, which is comparable to the core temperature. The energy confinement is then determined by the diffusion coefficient of the core plasma, which has been observed to be much lower than the thermal diffusivity. (author)

  5. Scrape-off layer and divertor theory meeting: Proceedings

    International Nuclear Information System (INIS)

    1994-03-01

    This report contains viewgraphs on the following topics: fluid modelling of neutrals in the SOL and divertor; instabilities of gas-fueled divertors: theory and adaptive simulations; stability of ionization fronts of gaseous divertor plasmas; monte carlo calculation of heat transport; reduced charge model for edge impurity flows; thermally collapsed solutions for gaseous/radiative divertors; adaptive grid methods in transport simulation; advanced numerical solution algorithms applied to the multispecies edge plasma equations; two-dimensional edge plasma simulation using the multigrid method; neutral behavior and the effects of neutral-neutral and neutral-ion elastic scattering in the ITER gaseous divertor; particle throughput in the TPX divertor; marfes in tokamaks; a comparative study of the limiter and divertor edge plasmas in TEXT-U; issues of toroidal tokamak-type divertor simulators; ASDEX upgrade; the ITER divertor; the DIII-D divertor program and TPX divertor; DEGAS 2: a transmission/escape probabilities model for neutral particle transport: comparison with DEGAS 2; a collisional radiative model of hydrogen for high recycling divertors; comparison of fluid and non- fluid neutral models in B2.5; DIII-D radiative divertor simulations; 3-D fluid simulations of turbulence from conducting wall mode; turbulence and drifts in SOL plasmas; recent results for 1 1/2-D ITER gas target divertor modelling; evaluation of pumping and fueling in coupled core, SOL, and divertor chamber calculations; and ITER gas target divertors: comparison of volume recombination and large radial transport scenarios using DEGAS

  6. Comparative studies of inner and outer divertor discharges and a fueling study in QUEST

    Energy Technology Data Exchange (ETDEWEB)

    Mitarai, O., E-mail: omitarai@ktmail.tokai-u.jp [Kumamoto Liberal Arts Education Center, Tokai University, 9-1-1 Toroku, Higashi-ku, Kumamoto 862-8652 (Japan); Nakamura, K.; Hasegawa, M.; Onchi, T.; Idei, H.; Fujisawa, A.; Hanada, K.; Zushi, H.; Higashijima, A.; Nakashima, H.; Kawasaki, S. [Research Institute for Applied Mechanics, Kyushu University, 6-1 Kasugakoen, Kasuga 816-8580 Japan (Japan); Matsuoka, K. [National Institute for Fusion Science, 322-6 Oroshi-cho, Toki 509-5292 (Japan); Koike, S.; Takahashi, T. [Division of Electronics and Informatics, Faculty of Science and Technology, Gunma University, 1-5-1 Tenjin-cho, Kiryu, Gunma 376-8515 (Japan); Tsutsui, H. [Research Laboratory for Nuclear Reactors, Tokyo Inst. Tech, 2-12-1 Ookayama, Tokyo 152-8550 (Japan)

    2016-11-01

    Highlights: • Central solenoid has a small flux in QUEST. • Large plasma current is obtained when the position is shifted to the inboard side. • Two types of divertor operation are compared. • Novel merging fueling methods are proposed. • Coaxial helicity injection (CHI) fueling was examined in QUEST divertor configuration. - Abstract: As QUEST has a small central solenoid (CS), a larger Ohmic discharge current has been obtained when the plasma shifts to the inboard side. This tendency restricts a divertor operation to the smaller plasma current regime. As the inner divertor coil has a smaller mutual inductance, it would be expected that its utilization seems to be better for easier plasma current ramp-up for a divertor operation. In this work, we made comparative studies on the plasma current ramp-up for two divertor coils. It is found that while the inner divertor coil with smaller mutual inductance needs a larger coil current, the outer divertor coil with larger mutual inductance needs a smaller coil current for divertor operation. Thus we have found that the plasma current ramp-up characteristics are almost similar for both configurations. We also propose a new fueling method for spherical tokamak (ST) using the coaxial helicity injection (CHI). The main plasma current would be generated at first, and then the CHI plasma current is created between bottom two electrode plates and merged into the main plasma current for fueling.

  7. The flush-mounted rail Langmuir probe array designed for the Alcator C-Mod vertical target plate divertor

    Science.gov (United States)

    Kuang, A. Q.; Brunner, D.; LaBombard, B.; Leccacorvi, R.; Vieira, R.

    2018-04-01

    An array of flush-mounted and toroidally elongated Langmuir probes (henceforth called rail probes) have been specifically designed for the Alcator C-Mod's vertical target plate divertor and operated over multiple campaigns. The "flush" geometry enables the tungsten electrodes to survive high heat flux conditions in which traditional "proud" tungsten electrodes suffer damage from melting. The toroidally elongated rail-like geometry reduces the influence of sheath expansion, which is an important effect to consider in the design and interpretation of flush-mounted Langmuir probes. The new rail probes successfully operated during C-Mod's FY2015 and FY2016 experimental campaigns with no evidence of damage, despite being regularly subjected to heat flux densities parallel to the magnetic field exceeding ˜1 GW m-2 for short periods of time. A comparison between rail and proud probe data indicates that sheath expansion effects were successfully mitigated by the rail design, extending the use of these Langmuir probes to incident magnetic field line angles as low as 0.5°.

  8. Characterizations of power loads on divertor targets for type-I, compound and small ELMs in the EAST superconducting tokamak

    DEFF Research Database (Denmark)

    Wang, L.; Xu, G.S.; Guo, H.Y.

    2013-01-01

    -III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98(y,2) ∼ 1). A significant loss......The Experimental Advanced Superconducting Tokamak (EAST) has recently achieved a variety of H-mode regimes with different edge-localized mode (ELM) dynamics, including type-I ELMs, compound ELMs, which are manifested by the onset of a large spike followed by a sequence of small spikes on Dα......-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MW m−2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may...

  9. VUV Spectroscopy in DIII-D Divertor

    International Nuclear Information System (INIS)

    Alkesh Punjabi; Nelson Jalufka

    2004-01-01

    The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report

  10. Two-dimensional numerical study of ELMs-induced erosion of tungsten divertor target tiles with different edge shapes

    International Nuclear Information System (INIS)

    Huang, Yan; Sun, Jizhong; Hu, Wanpeng; Sang, Chaofeng; Wang, Dezhen

    2016-01-01

    Highlights: • Thermal performance of three edge-shaped divertor tiles was assessed numerically. • All the divertor tiles exposed to type-I ELMs like ITER's will melt. • The rounded edge tile thermally performs the best in all tiles of interest. • The incident energy flux density was evaluated with structural effects considered. - Abstract: Thermal performance of the divertor tile with different edge shapes was assessed numerically along the poloidal direction by a two-dimensional heat conduction model with considering the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma. The energy flux density distribution arriving at the castellated divertor tile surface was evaluated by a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo Collisions code and then the obtained energy flux distribution was used as input for the heat conduction model. The simulation results showed that the divertor tiles with any edge shape of interest (rectangular edge, slanted edge, and rounded edge) would melt, especially, in the edge surface region of facing plasma poloidally under typical heat flux density of a transient event of type-I ELMs for ITER, deposition energy of 1 MJ/m"2 in a duration of 600 μs. In comparison with uniform energy deposition, the vaporizing erosion was reduced greatly but the melting erosion was aggravated noticeably in the edge area of plasma facing diveror tile. Of three studied edge shapes, the simulation results indicated that the divertor plate with rounded edge was the most resistant to the thermal erosion.

  11. Energy transport to the divertor plates of ASDEX-Upgrade during ELMy H-mode phases

    International Nuclear Information System (INIS)

    Herrmann, A.; Laux, M.; Coster, D.; Neuhauser, J.; Reiter, D.; Schneider, R.; Weinlich, M.

    1995-01-01

    The energy flux to the ASDEX-Upgrade divertor plates is routinely measured by themography and Langmuir probes. The thermographically observed power decay length at the target plate is about 1 cm near the inboard separatrix. During an edge localized mode (ELM) of type I the density profiles are significantly, changed; an additional contribution occurs characterized by a power decay length in the order of 10 cm outside the separatrix and additional power is deposited into the private flux region. It is supposed that this is due to the changing, contribution of energy conduction versus convection. Results of ELM-modelling using the coupled B2-EIRENE code reproduce the main features of the experimental observations. The sheath transmission factor is calculated by combining themography and Langmuir probe data. ((orig.))

  12. Studies of impurity deposition/implantation in JET divertor tiles using SIMS and ion beam techniques

    International Nuclear Information System (INIS)

    Likonen, J.; Lehto, S.; Coad, J.P.; Renvall, T.; Sajavaara, T.; Ahlgren, T.; Hole, D.E.; Matthews, G.F.; Keinonen, J.

    2003-01-01

    At the end of C4 campaign at JET, a 1% SiH 4 /99% D 2 mixture and pure 13 CH 4 were injected into the torus from the outer divertor wall and from the top of the vessel, respectively, in order to study material transport and scrape-off layer (SOL) flows. A set of MkIIGB tiles was removed during the 2001 shutdown for surface analysis. The tiles were analysed with secondary ion mass spectrometry (SIMS) and time-of-flight elastic recoil detection analysis (TOF-ERDA). 13 C was detected in the inner divertor wall tiles implying material transport from the top of the vessel. Silicon was detected mainly at the outer divertor wall tiles and very small amounts were found in the inner divertor wall tiles. Si amounts in the inner divertor wall tiles were so low that rigorous conclusions about material transport from divertor outboard to inboard cannot be made

  13. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  14. Interpretation of ion flux and electron temperature profiles at the JET divertor target during high recycling and detached discharges

    International Nuclear Information System (INIS)

    Monk, R.D.

    1997-01-01

    Detailed experiments have been carried out with the JET Mark I pumped divertor to characterise high recycling and detached plasma regimes. This paper presents new measurements of high resolution divertor ion flux profiles that identify the growth of additional peaks during high recycling discharges. These ion flux profiles are used in conjunction with Dα and neutral flux measurements to examine the physics of divertor detachment and compare against simple analytic models. Finally, problems are highlighted with conventional methods of single and triple probe interpretation under high recycling conditions. By assuming that the single probe behaves as an asymmetric double probe the whole characteristic may be fitted and significantly lower electron temperatures may be derived when the electron to ion saturation current ratio is reduced. The results from the asymmetric double probe fit are shown to be consistent with independent diagnostic measurements. (orig.)

  15. Lithium vapor trapping at a high-temperature lithium PFC divertor target

    Science.gov (United States)

    Jaworski, Michael; Abrams, T.; Goldston, R. J.; Kaita, R.; Stotler, D. P.; de Temmerman, G.; Scholten, J.; van den Berg, M. A.; van der Meiden, H. J.

    2014-10-01

    Liquid lithium has been proposed as a novel plasma-facing material for NSTX-U and next-step fusion devices but questions remain on the ultimate temperature limits of such a PFC during plasma bombardment. Lithium targets were exposed to high-flux plasma bombardment in the Magnum-PSI experimental device resulting in a temperature ramp from room-temperature to above 1200°C. A stable lithium vapor cloud was found to form directly in front of the target and persist to temperature above 1000°C. Consideration of mass and momentum balance in the pre-sheath region of an attached plasma indicates an increase in the magnitude of the pre-sheath potential drop with the inclusion of ionization sources as well as the inclusion of momentum loss terms. The low energy of lithium emission from a surface measured in previous experiments (Contract DE-AC02-09CH11466.

  16. Manufacturing, testing and post-test examination of ITER divertor vertical target W small scale mock-ups

    International Nuclear Information System (INIS)

    Visca, Eliseo; Cacciotti, Emanuele; Komarov, Anton; Libera, Stefano; Litunovsky, Nikolay; Makhankov, Alexey; Mancini, Andrea; Merola, Mario; Pizzuto, Aldo; Riccardi, Bruno; Roccella, Selanna

    2011-01-01

    ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R and D activities. During the last years ENEA has set up and widely tested a manufacturing process, named Hot Radial Pressing (HRP), suitable for the construction of high heat flux plasma-facing components, such as the divertor targets. In the frame of the EFDA contract six mock-ups were manufactured by HRP in the ENEA labs using W monoblocks supplied by the Efremov Institute in St. Petersburg, Russian Federation and IG CuCrZr tubes. According to the technical specifications the mock-ups were examined by ultrasonic technique and after their acceptance they were delivered to the Efremov Institute TSEFEY-M e-beam facility for the thermal fatigue testing. The test consisted in 3000 cycles of 15 s heating and 15 s cooling at 10 MW/m 2 and finally 1000 cycles at 20 MW/m 2 . After the testing the ultrasonic non-destructive examination was repeated and the results compared with the investigation performed before the testing. A microstructure modification of the W monoblock material due to the overheating of the surfaces and the copper interlayer structure modification were observed in the high heat flux area. The leakage points of the mock-ups that did not conclude the testing were localized in the middle of the monoblock while they were expected between two monoblocks. This paper reports the manufacturing route, the thermal fatigue testing, the pre and post non destructive examination and finally the results of the destructive examination performed on the monoblock small scale mock-ups.

  17. A large divertor manipulator for ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, Albrecht, E-mail: albrecht.herrmann@ipp.mpg.de; Jaksic, Nikola; Leitenstern, Peter; Greuner, Henri; Krieger, Karl; Marné, Pascal de; Oberkofler, Martin; Rohde, Volker; Schall, Gerd

    2015-10-15

    Highlights: • A large divertor manipulator for ASDEX Upgrade is developed and tested. • It allows replacing a relevant part of the divertor by dedicated targets and probes. • Modified solid standard targets. • Electrical and mechanical probes for dedicated investigations. • Test of actively cooled component. - Abstract: In 2013 a new bulk tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG). During the concept and design phase of Div-III the option of adaptable divertor instrumentation and divertor modification as contribution for divertor investigations in preparation of ITER was given a high priority. To gain flexibility for the test of divertor modifications without affecting the operational space of AUG, the large divertor manipulator, DIM-II, was designed and installed. DIM-II allows to retract 2 out of 128 outer divertor target tiles including the water cooled support structure into a target exchange box and to replace these targets without breaking the vacuum of the AUG vessel. DIM-II is based on a carriage-rail system with a driving rod pushing a front-end with the target module into the divertor position for plasma operation. Three types of front-ends are foreseen for physics investigations: (i) modified standard targets clamped to the standard cooling structure, (ii) dedicated front-ends making use of the whole available volume of about 230 × 160 × 80 mm{sup 3} and (iii) actively cooled/heated targets for cooling water temperatures up to 230 °C. This paper presents the DIM-II design including the FEM calculations for the modified divertor support structure and the front-end options, as well as the test procedure and operation mode.

  18. Advanced divertor configurations with large flux expansion

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V.A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Bell, R.E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Menard, J.E.; Paul, S.F.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Raman, R. [University of Washington, Seattle, WA (United States); Ryutov, D.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Scotti, F.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mueller, D.M.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Reimerdes, H.; Canal, G.P. [Ecole Polytechnique Fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, Association Euratom Confédération Suisse, Lausanne (Switzerland); and others

    2013-07-15

    Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m{sup 2} to 0.5–1 MW/m{sup 2} was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX

  19. CIT divertor conceptual design

    International Nuclear Information System (INIS)

    Wesley, J.C.; Sevier, D.L.

    1988-06-01

    A conceptual design of the divertor target assembly for the 1.75-m CIT baseline device has been developed. The divertor target assembly consists of four toroidal arrays of pyrolytic graphite plates that cover the inside surface of the ends of the vacuum vessel in the locations where the magnetic separatrices of the plasma intersect the vessel wall. During the course of the plasma discharge, the currents on the poloidal field coils that establish the plasma equilibrium are varied to sweep the separatrix strike locations across the divertor targets. This spreads the plasma heat loading over sufficient area to keep the peak target surface temperature within allowable limits. The required magnetic sweep (/+-/5 cm for the inside strike location and /+-/12 cm for the outside strike location) can be affected by programming either the external poloidal strike location) can be effected by programming either the external poloidal field (PF) coils or the internal PF control coils plus the external PF solenoid coils (PF1 and PF2). The ensuing variations in the elongation and triangularity of the plasma are modest, and fall within the ranges of plasma elongation and triangularity specified in the CIT General Requirements Document. 17 figs., 13 tabs

  20. Assessment of the effect of parallel temperature gradients in the JET SOL on Te measured by divertor target Langmuir probes

    NARCIS (Netherlands)

    Ďuran, I.; Ješko, K.; Fuchs, V.; Groth, M.; Guillemaut, C.; Gunn, J.P.; Horáček, J.; Pitts, R.A.; Tskhakaya, D.

    2015-01-01

    Abstract Higher than expected electron temperatures (Te) are often measured by divertor Langmuir probes (LP) in high recycling and detached regimes in JET and other tokamaks. As a possible mechanism to explain this discrepancy, we investigate the effect of penetration of fast, almost collisionless

  1. Qualification of high heat flux components: application to target elements of W7-X divertor

    International Nuclear Information System (INIS)

    Missirlian, M; Durocher, A; Grosman, A; Schlosser, J; Boscary, J; Escourbiac, F; Cismondi, F

    2007-01-01

    The development of actively cooled plasma-facing components (PFC) represents one of fusion's most challenging engineering efforts. In this frame, a high-quality bonding between the refractory armour and the heat sink is essential to ensure the heat removal capability and the thermal performances of PFC. Experience gained during manufacturing of Tore Supra actively cooled PFC led to the establishment of a qualification methodology and provided a large experience of acceptance criteria using an active infrared thermography (systeme d'acquisition de traitement infra-rouge, SATIR). This paper presents the application of this qualification process to the W7-X pre-series components, with the objective of assessing and defining workable acceptance criteria that enable reliable predictions of performance at the nominal heat flux requirements in W7-X. Finally, to check the reliability of the non-destructive examination (NDE) method by transient infrared thermography, the newly defined acceptance criteria were applied to W7-X pre-series target elements (batch no. 3). The SATIR results, benchmarked with HHF tests performed on the GLADIS ion beam facility were discussed to assess the ability to detect critical defects at the interface between tiles and heat sink

  2. Detached divertor plasmas in JET

    Energy Technology Data Exchange (ETDEWEB)

    Horton, L D; Borrass, K; Corrigan, G; Gottardi, N; Lingertat, J; Loarte, A; Simonini, R; Stamp, M F; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P C [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

    1994-07-01

    In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

  3. Advanced divertor concepts

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Sagara, A.; Suzuki, H.; Morisaki, T.; Masuzaki, S.; Watanabe, T.; Noda, N.; Motojima, O.

    1996-01-01

    LHD divertor development program has generated various innovative divertor concepts and technologies which will help to improve the plasma performance in both helical and tokamak devices. They are two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. 17 refs., 8 figs

  4. Small angle slot divertor concept for long pulse advanced tokamaks

    Science.gov (United States)

    Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

    2017-04-01

    SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

  5. A new divertor plates design concept for the double null NET configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Renda, V.; Federici, G.; Papa, L.

    1986-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper. (author)

  6. A new divertor plates design concept for the double null net configuration

    International Nuclear Information System (INIS)

    Farfaletti-Casali, F.; Iop, O.; Renda, V.; Federici, G.; Papa, L.

    1987-01-01

    A new divertor plate design concept for the Double Null NET configuration (NET-DN) is presented in this paper. This concept applies to the plasma configuration of NET and takes advantage by the maintenance scheme of the internal components adopted in NET. According to this maintenance approach, which uses the top loading of the internal segments, 48 inboard removable segments, 3 for each of the 16 reactor sectors, act as simple protective panels, gathering together in only one piece the plates of both the upper and lower divertor regions and the intermediate portion of the inboard first wall. They are cooled by water flowing inside a set of hairpin-shaped, stainless steel tubes, arranged in poloidal direction inside a copper heat sink, and fed by supply lines at the top of the reactor. The surface facing the plasma is covered by a tungsten alloy layer. In such a way, the maintenance of the two divertor regions and of the inboard first wall can be easily achieved by removing the inboard panels from the top of the reactor. The layout of the cooling system and preliminary thermohydraulics and thermomechanical calculations, carried out for assessing the feasibility of the proposed system for the NET reference configuration, are reported in this paper

  7. Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

    Science.gov (United States)

    Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

    2018-05-01

    Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

  8. Divertor design through shape optimization

    International Nuclear Information System (INIS)

    Dekeyser, W.; Baelmans, M.; Reiter, D.

    2012-01-01

    Due to the conflicting requirements, complex physical processes and large number of design variables, divertor design for next step fusion reactors is a challenging problem, often relying on large numbers of computationally expensive numerical simulations. In this paper, we attempt to partially automate the design process by solving an appropriate shape optimization problem. Design requirements are incorporated in a cost functional which measures the performance of a certain design. By means of changes in the divertor shape, which in turn lead to changes in the plasma state, this cost functional can be minimized. Using advanced adjoint methods, optimal solutions are computed very efficiently. The approach is illustrated by designing divertor targets for optimal power load spreading, using a simplified edge plasma model (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  9. Evaluating Stellarator Divertor Designs with EMC3

    Science.gov (United States)

    Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

    2013-10-01

    In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

  10. The comparison of heat flux pattern on lower divertor in KSTAR

    International Nuclear Information System (INIS)

    Bang, Eunnam; Hong, Suk-Ho; Bak, JunGyo; Kim, Kyungmin; Kim, Hongtack; Kim, Hakkun; Yang, H.L.

    2015-01-01

    Highlights: • The heat flux on the lower divertor is higher than upper divertor. • The heat flux on OD is decreased with IVCP. • The heat flux on CD is decreased with RMP, but that on OD is increased. • Because the strike point was shifted from CD toward OD due to the RMP. - Abstract: The heat flux in KSTAR is estimated for various discharge conditions by using thermocouple arrays. The heat flux on the divertor is higher than that on inboard limiter or passive stabilizer by a factor of 2. Although the plasma configuration in KSTAR has been set to a double-null configuration, the heat flux on lower divertor is higher than that on upper divertor by 3–8 times, indicating a lower-single-null-like configuration. It is observed that the operation of the in-vessel cryo-pump (IVCP) changes the heat flux pattern significantly: When the IVCP was not operated, the heat fluxes on inboard divertor (ID), central divertor (CD) and outboard divertor (OD) were similar, but when the IVCP was operated, the heat fluxes on ID and CD were increased slightly and that on OD was decreased by 2–3 times. The heat flux on divertor was decreased from 35 to 26 kW/m"2 with the use of the resonant magnetic perturbation (RMP), especially that on CD was decreased by 2–4 times, while that on OD is increased by 2–3 times than without RMP. For the longest H-mode pulse of 22 s shot, the heat flux on lower OD was 73 kW/m"2, which is the maximum heat flux among the shots obtained in 2013 campaign.

  11. Upgraded divertor Thomson scattering system on DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

    2016-11-15

    A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

  12. Divertor radiation in the ASDEX upgrade tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Sehmer, Till; Bernert, Matthias; Koll, Juergen; Meister, Hans; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, 52425 Juelich (Germany); Collaboration: The ASDEX Upgrade Team

    2016-07-01

    To reduce in ITER the expected power flux density onto the divertor target, the plasma-wall interaction in the divertor needs to be strongly reduced. The fundamental path to achieve this is using radiation from seeded impurities, whereas the localization of this radiation (e.g. inside/outside confined region), which could have an impact onto the power balance, is a key challenge. The absolute radiated power distribution can be measured by foil bolometers. To study at the ASDEX Upgrade tungsten divertor the localization and quantification of radiation, the respective line of sight density of the bolometers has been improved by two additional cameras. The divertor radiation enhanced by nitrogen (N{sub 2}) seeding has been investigated, using variations of (1) the external heating power or (2) the N{sub 2} seeding rate. While in both cases the inner divertor stays fully detached, measurements indicate that the region of dominant radiation moves from the inner divertor through the X-Point into the confined region. In the outer divertor however, the measurements indicate either an immediate upwards shift or a continuous movement of the radiation away from the target, depending on experimental conditions.

  13. Application of the radiating divertor approach to innovative tokamak divertor concepts

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

    2015-08-15

    We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

  14. LHD helical divertor

    International Nuclear Information System (INIS)

    Ohyabu, N.; Watanabe, T.; Ji Hantao

    1993-07-01

    The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experiment, high density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with temperature of a few kev, generated by efficient pumping, expects to lead to significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way. (author)

  15. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  16. A solid tungsten divertor for ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A; Greuner, H; Jaksic, N; Böswirth, B; Maier, H; Neu, R; Vorbrugg, S

    2011-01-01

    The conceptual design of a solid tungsten divertor for ASDEX Upgrade (AUG) is presented. The Div-III design is compatible with the existing divertor structure. It re-establishes the energy and heat receiving capability of a graphite divertor and overcomes the limitations of tungsten coatings. In addition, a solid tungsten divertor allows us to investigate erosion and bulk deuterium retention as well as test castellation and target tilting. The design criteria as well as calculations of forces due to halo and eddy currents are presented. The thermal properties of the proposed sandwich structure are calculated with finite element method models. After extensive testing of a target tile in the high heat flux test facility GLADIS, two solid tungsten tiles were installed in AUG for in-situ testing.

  17. Design of the inboard passive stabilizer for TPX

    International Nuclear Information System (INIS)

    Hoffmann, E.; Boonstra, R.; Baxi, C.B.; Chin, E.; Drees, L.; Lee, W.; Redler, K.L.; Reis, E.E.; Bialek, J.

    1995-01-01

    The Inboard Passive Stabilizer (IPS) is part of the plasma stabilizing system built into the TPX. Its purpose is to provide passive stabilization of the plasma vertical instability on short time scales. With carbon fiber composite (CFC) armor tiles it serves as a startup limiter, protects the vacuum vessel from radiation heat load during steady state operation and also functions as Neutral Beam armor. The inboard passive stabilizer is a saddle coil, constructed of a ring of copper plates, armored with CFC tiles, that surrounds the inner vacuum vessel at the midplane. The design of the plates, the support structure, cooling lines, CFC tiles and tile attach method is described. Tiles that experience only the normal heat load of 0.4 MW/m 2 are attached with mechanical fasteners. Tiles in the neutral beam shine through area are exposed to as much as 1.7 MW/m 2 and are brazed to the IPS. Significant forces are generated in the plates by the stabilization currents as well as during the frequent bakeout cycles. These plates are required to be fully remotely handled, including tile replacement, and the influence of this requirement on the design is discussed

  18. First results of closed helical divertor experiment in LHD

    International Nuclear Information System (INIS)

    Morisaki, T.; Masuzaki, S.; Kobayashi, M.

    2012-11-01

    The baffle-structured closed Helical Divertor (CHD) is being constructed in LHD to actively control the edge plasma, which consists of ten discrete modules installed on inboard side of the torus. At this stage, two of ten modules have been constructed. In the initial experiments, performance of CHD was experimentally investigated, comparing with numerical expectations. During the continuous gas puffing discharge, it was observed the neutral pressure in the CHD was more than 10 times higher than that in the open HD, which agrees well with the numerical simulation. In the high density regime, indication of the divertor detachment was observed in CHD, which was caused by the high recycling and high density state in CHD. With a Penning discharge diagnostics, the neutral particle behaviour with different species was investigated. Little difference between hydrogen and helium was observed in transport property. (author)

  19. Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

    International Nuclear Information System (INIS)

    Gotoh, Y.; Tanabe, T.; Ishimoto, Y.; Masaki, K.; Arai, T.; Kubo, H.; Tsuzuki, K.; Miya, N.

    2006-01-01

    Erosion and redeposition profiles of carbon tiles used in the W-shaped divertor of JT-60U with all carbon plasma facing wall are studied. The inner divertor is mostly covered by carbon redeposited layers, while the outer divertor mostly eroded. In the dome region, the erosion dominates on the inner dome-wing, while redeposited on the outer dome-wing. The redeposited layers on the outer dome-wing show very clear columnar structures indicating local carbon transport form the outer divertor to the outer dome-wing. The weight gain by the redeposition extrapolated to the whole divertor area is 0.55 kg. Since the extrapolated total erosion is about 0.33 kg, the remaining 0.22 kg must be originated from the main chamber erosion. Significant amount of the redeposition is caused locally by multiple processes of erosion, ionization and prompt redeposition toward inboard direction owing to gyration along magnetic filed line. This inboard transport is one of the reasons for small redeposition on the plasma shadowed area of the W shaped divertor of JT-60U with pumping slots placed at the bottoms side

  20. Dissipative divertor operation in the Alcator C-Mod tokamak

    International Nuclear Information System (INIS)

    Lipschultz, B.; Goetz, J.; LaBombard, B.; McCracken, G.M.; Terry, J.L.; Graf, M.; Granetz, R.S.; Jablonski, D.; Kurz, C.; Niemczewski, A.; Snipes, J.

    1995-01-01

    The achievement of large volumetric power losses (dissipation) in the Alcator C-Mod divertor region is demonstrated in two operational modes: radiative divertor and detached divertor. During radiative divertor operation, the fraction of SOL power lost by radiation is P R /P SOL ∼0.8 with single null plasmas, n e 20 m -3 and I p e,div ≤6x10 20 m -3 . As the divertor radiation and density increase, the plasma eventually detaches abruptly from the divertor plates: I SAT drops at the target and the divertor radiation peak moves to the X-point region. Probe measurements at the divertor plate show that the transition occurs when T e ∼5 eV. The critical n e for detachment depends linearly on the input power. This abrupt divertor detachment is preceded by a comparatively long period ( similar 1-200 ms) where a partial detachment is observed to grow at the outer divertor plate. ((orig.))

  1. Variation of particle exhaust with changes in divertor magnetic balance

    International Nuclear Information System (INIS)

    Petrie, T.W.; Allen, S.L.; Brooks, N.H.

    2006-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B x ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B x ∇B ion particle drift direction. Our data suggests that the presence of B x ∇B and E x B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks

  2. Characteristics of the Secondary Divertor on DIII-D

    Science.gov (United States)

    Watkins, J. G.; Lasnier, C. J.; Leonard, A. W.; Evans, T. E.; Pitts, R.; Stangeby, P. C.; Boedo, J. A.; Moyer, R. A.; Rudakov, D. L.

    2009-11-01

    In order to address a concern that the ITER secondary divertor strike plates may be insufficiently robust to handle the incident pulses of particles and energy from ELMs, we performed dedicated studies of the secondary divertor plasma and scrape-off layer (SOL). Detailed measurements of the ELM energy and particle deposition footprint on the secondary divertor target plates were made with a fast IR camera and Langmuir probes and SOL profile and transport measurements were made with reciprocating probes. The secondary divertor and SOL conditions depended on changes in the magnetic balance and the core plasma density. Larger density resulted in smaller ELMs and the magnetic balance affected how many ELM particles coupled to the secondary SOL and divertor. Particularly striking are the images from a new fast IR camera that resolve ELM heat pulses and show spiral patterns with multiple peaks during ELMs in the secondary divertor.

  3. Modeling detachment physics in the NSTX snowflake divertor

    Energy Technology Data Exchange (ETDEWEB)

    Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

    2015-08-15

    The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

  4. Role of molecular effects in divertor plasma recombination

    Directory of Open Access Journals (Sweden)

    A.S. Kukushkin

    2017-08-01

    Full Text Available Molecule-Activated Recombination (MAR effect is re-considered in view of divertor plasma conditions. A strong isotopic effect is demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included in 2D edge plasma models, is negligible. However, in this case the other branch, through D−, usually neglected in modelling, becomes relatively strong. The overall share of MAR in divertor plasma recycling stays within 20%. The operational parameters of the divertor plasmas, such as the peak power loading on the divertor targets or the pressure limit for partial detachment of the divertor plasma, are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the divertor diagnostics.

  5. The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

    Directory of Open Access Journals (Sweden)

    J. Uljanovs

    2017-08-01

    Full Text Available Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane.

  6. Models for poloidal divertors

    International Nuclear Information System (INIS)

    Post, D.E.; Heifetz, D.; Petravic, M.

    1982-07-01

    Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done

  7. Models for poloidal divertors

    Energy Technology Data Exchange (ETDEWEB)

    Post, D.E.; Heifetz, D.; Petravic, M.

    1982-07-01

    Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done.

  8. Controlling marginally detached divertor plasmas

    Science.gov (United States)

    Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

    2017-06-01

    A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

  9. Plasma flow in the DIII-D divertor

    International Nuclear Information System (INIS)

    Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

    1998-07-01

    Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor

  10. Divertor design for the TITAN reversed-field-pinch reactor

    International Nuclear Information System (INIS)

    Cooke, P.I.H.; Bathke, C.G.; Blanchard, J.P.; Creedon, R.L.; Grotz, S.P.; Hasan, M.Z.; Orient, G.; Sharafat, S.; Werley, K.A.

    1987-01-01

    The design of the toroidal-field divertor for the TITAN high-power-density reversed-field-pinch reactor is described. The heat flux on the divertor target is limited to acceptable levels (≤ 10 MW/m 2 ) for liquid-lithium cooling by use of an open divertor geometry, strong radiation from the core and edge plasma, and careful shaping of the target surface. The divertor coils are based on the Integrated-Blanket-Coil approach to minimize the loss in breeding-blanket coverage due to the divertor. A tungsten-rhenium armour plate, chosen for reasons of sputtering resistance, and good thermal and mechanical properties, protects the vanadium-alloy coolant tubes

  11. The DIII-D Radiative Divertor Project: Status and plans

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Bozek, A.S.

    1996-10-01

    New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

  12. Textor bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

    1982-01-01

    This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

  13. TEXTOR bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

    1982-01-01

    This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

  14. Divertor cooling device

    International Nuclear Information System (INIS)

    Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

    1993-01-01

    Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

  15. Physics design and experimental study of tokamak divertor

    International Nuclear Information System (INIS)

    Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

    2007-06-01

    The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

  16. T-12 divertor experiment

    Energy Technology Data Exchange (ETDEWEB)

    Bortnikov, A V; Brevnov, N N; Gerasimov, S N; Zhukovskii, V G; Kuznetsov, N V; Naftulin, S M; Pergament, V I; Khimchenko, L N [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

    1981-01-01

    In designing tokamak devices and reactors, in the last few years, the use of elongated-cross-section plasma discharges has been proposed to improve the economic and physical parameters. Application of a quadrupole poloidal magnetic field necessary for sustaining the elongated discharge cross-section serves, in this case, to create the magnetic configuration of an axisymmetric poloidal divertor. To-day, the creation of such a combination, including an elongated plasma cross-section and a divertor and using the outer poloidal magnetic field coils, seems to be the most reasonable approach, from the point of view of design and technology. Such a divertor was produced and studied at the T-12 tokamak. A stable equilibrium configuration of a finger-ring tokamak with a divertor has been produced by superposing the magnetic fields of the plasma current, the external quadrupole coils and the copper shell currents; the reactor blanket can fulfil the function of the latter. It is shown that both a symmetric magnetic configuration with two divertors and a droplet configuration with a single divertor may be realized by controlling the plasma column position with respect to the equatorial plane. The stability of the plasma column against vertical displacement depends on this position and the distance between the separatrix points. Vertical instability stabilization has been observed. The divertor layer efficiently screens the plasma from the impurity influx from the wall and unloads the wall from particle and energy fluxes. The results obtained from the tokamak T-12 experiment have demonstrated the capability of a system with outer poloidal field coils and a copper shell providing an elongated-cross-section plasma column with poloidal divertors.

  17. EU R and D on divertor components

    International Nuclear Information System (INIS)

    Merola, M.; Daenner, W.; Pick, M.

    2005-01-01

    Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R and D on the divertor components. A number of activities have been completed and new ones have been launched. The present paper gives an update of the works carried out by the EU Participating Team in support of the development of the divertor, which is one of the most challenging components of the next-step ITER machine. The following topics are covered: (1) the further development and consolidation of suitable technologies for the production of high heat-flux components, which culminated with the successful manufacturing and testing of a full-scale vertical target prototype; (2) the completion of the post-irradiation testing of divertor mock-ups and samples; (3) the preparation for the hydraulic and assembly tests of a complete set of full-scale divertor components; (4) the on-going R and D on the definition of workable acceptance criteria for the procurement of ITER high heat-flux components; (5) the activities in support of the divertor design

  18. Particle control in the DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Lippmann, S.I.; Mahdavi, M.A.; Petrie, T.W.; Stambaugh, R.D.; Hogan, J.; Klepper, C.C.; Mioduszewski, P.; Owen, L.; Hill, D.N.; Rensink, M.; Buchenauer, D.

    1991-11-01

    A new, electrically biasable, semi-closed divertor was installed and operated in the D3-D lower outside divertor location. The semi-closed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. (The planned cryogenic pumping is not yet installed). Electrical bias controls the distribution of particle recycle between the inner and outer divertors by rvec E x rvec B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduce the sensitivity of plenum pressure to separatrix position. In particular, rvec E x rvec B drifts in the D3-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density. 5 refs., 8 figs

  19. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

    Merola, M.; Tivey, R.; Martin, A.; Pick, M.

    2006-01-01

    The procurement of the ITER divertor is planned to start in 2009. On the basis of the present common understanding of the sharing of the ITER components, the Japanese Participating Team (JAPT) will supply the outer vertical target, the Russian Federation (RF) PT the dome liner and will perform the high heat flux testing, the EU PT will supply the inner vertical targets and the cassette bodies, including final assembly of the divertor plasma-facing components (PFCs). The manufacturing of the PFCs of the ITER divertor represents a challenging endeavor due to the high technologies which are involved, and due to the unprecedented series production. To mitigate the associated risks, special arrangements need to be put in place prior to and during procurement to ensure quality and to keep to the time schedule. Before procurement can start, an ITER review of the qualification and production capability of each candidate PT is planned. Well in advance of the assumed start of the procurement, each PT which would like to contribute to the divertor PFC procurement, should first demonstrate its technical qualification to carry out the procurement with the required quality, and in an efficient and timely manner. Appropriate precautions, like subdivision of the procurement into stages, are also to be adopted during the procurement phase to mitigate the consequences of possible unexpected manufacturing problems. In preparation for writing the procurement specification for the vertical targets, the topic of setting acceptance criteria is also being addressed. This activity has the objective of defining workable acceptance criteria for the PFC armour joints. A complete set of analyses is also in progress to assess the latest design modifications against the design requirements. This task includes neutronic, shielding, thermo-mechanical and electromagnetic analyses. More than half of the ITER plasma parameters that must be measured and the related diagnostics are located in the

  20. Plans of LHD divertor experiment

    International Nuclear Information System (INIS)

    Ohyabu, Nobuyoshi; Komori, Akio; Sagara, Akio; Noda, Nobuaki; Motojima, Osamu

    1996-01-01

    Scenarios of the LHD divertor experiment are presented. In the LHD divertor experimental program, various innovative divertor concepts and technologies, developed during its design phase will be utilized to improve the plasma performance. Two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement) are among them. Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. (author)

  1. Innovative divertor concepts for LHD

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Akaishi, K.

    1994-07-01

    We are developing various innovative divertor concepts which improve the LHD plasma performance. These are two divertor magnetic geometries (helical and local island divertors), three operational scenarios (radiative cooling in the high density, cold boundary, confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode like confinement improvement) and technological development of new efficient hydrogen pumping schemes. (author)

  2. Overview of the divertor design and its integration into RTO/RC-ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Tivey, R.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Heidl, H.; Ibbott, C.; Martin, E.

    2000-01-01

    The design of the divertor and its integration into the reduced technical objectives/reduced cost-international thermonuclear energy reactor (RTO/RC-ITER) is based on the experience gained from the 1998 design of international thermonuclear energy reactor (ITER) and on the research and development performed throughout the engineering design activities (EDA). This paper gives an overview of the layout and functional design of the RTO/RC-ITER divertor, including the integration into the machine and the remote replacement of the divertor cassettes. Design guidelines are presented which have allowed quick preparation of divertor layouts suitable for further study using the B2-EIRENE edge plasma code. As in the 1998 design, the divertor is segmented into cassettes, and the segmentation, which is three per sector, is driven by access through the divertor level ports. Maintaining this access and avoiding interference with poloidal field coils means that the divertor level ports need to be inclined (7 deg.). This opens up the possibility of incorporating inboard and outboard baffles into the divertor cassettes. The cassettes are transported in-vessel by making use of the toroidal rails onto which the cassettes are finally clamped in position. Significant reduction of the space available between the X-point and the vacuum vessel results in re-positioning of the toroidal rails in order to retain sufficient depth for the inner and outer divertor legs. This, in turn, requires some changes to the remote handling (RH) concept. Remote handling (RH) is now based on using a cantilevered articulated gripper during the radial movement of the cassettes inside the RH ports. However, the principle to use a cassette toroidal mover (CTM) for in vessel handling is unchanged, hence maintaining the validity of previous EDA research and development. The space previously left below the cassettes for RH was also used for pumping. Elimination of this space has led to re-siting of the pumping

  3. Physical study of experimental fusion breeder FEB divertor

    International Nuclear Information System (INIS)

    Zhu Yukun; Zhou Xiaobing; Huang Jinhua; Feng Kaiming; Deng Peizhi; Huo Tiejun

    1999-10-01

    The physical study of FEB divertor is presented. In order to improve the impurity control and increase ion-neutral interactions in the divertor, the configuration of the divertor is optimized to be the close type in the engineering design activity compared with the open type in the early conceptual activity. The operation mode of the divertor is designed to be partial detached plasma mode under conditions of combination gas-puffing with impurity injection. The position of gas-puffing is optimized to be at the torus mid-plane with NEWT1D code from the viewpoint of impurity retention and radiation in the scrape-off layer/divertor region. Boron is chosen as the injected impurity. The effect of boron impurity injection is evaluated from the reduced heat load on the divertor target. The plasma pressure drop along the scrape-off layer/divertor region is estimated with the two-point transport model and impurity radiation model in the dynamic gas target concept. The simulation results show that the plasma pressure drop factor f p is not only related to the radiation fraction f rad but also related greatly to the stagnation point density n s

  4. Versator divertor experiment: preliminary designs

    International Nuclear Information System (INIS)

    Wan, A.S.; Yang, T.F.

    1984-08-01

    The emergence of magnetic divertors as an impurity control and ash removal mechanism for future tokamak reactors bring on the need for further experimental verification of the divertor merits and their ability to operate at reactor relevant conditions, such as with auxiliary heating. This paper presents preliminary designs of a bundle and a poloidal divertor for Versator II, which can operate in conjunction with the existing 150 kW of LHRF heating or LH current drive. The bundle divertor option also features a new divertor configuration which should improve the engineering and physics results of the DITE experiment. Further design optimization in both physics and engineering designs are currently under way

  5. Tritium analysis of divertor tiles used in JET ITER-like wall campaigns by means of β-ray induced x-ray spectrometry

    Science.gov (United States)

    Hatano, Y.; Yumizuru, K.; Koivuranta, S.; Likonen, J.; Hara, M.; Matsuyama, M.; Masuzaki, S.; Tokitani, M.; Asakura, N.; Isobe, K.; Hayashi, T.; Baron-Wiechec, A.; Widdowson, A.; contributors, JET

    2017-12-01

    Energy spectra of β-ray induced x-rays from divertor tiles used in ITER-like wall campaigns of the Joint European Torus were measured to examine tritium (T) penetration into tungsten (W) layers. The penetration depth of T evaluated from the intensity ratio of W(Lα) x-rays to W(Mα) x-rays showed clear correlation with poloidal position; the penetration depth at the upper divertor region reached several micrometers, while that at the lower divertor region was less than 500 nm. The deep penetration at the upper part was ascribed to the implantation of high energy T produced by DD fusion reactions. The poloidal distribution of total x-ray intensity indicated higher T retention in the inboard side than the outboard side of the divertor region.

  6. Snowflake Divertor Configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  7. 'Snowflake' divertor configuration in NSTX

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.

    2011-01-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  8. "Snowflake" divertor configuration in NSTX

    Science.gov (United States)

    Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

    2011-08-01

    Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

  9. Thermal effects of divertor sweeping in ITER

    International Nuclear Information System (INIS)

    Wesley, J.C.

    1992-01-01

    In this paper, thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the International Thermonuclear Fusion Reactor (ITER) are calculated. For the 0. 2 Hz x ± 12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by ∼ 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5 s sweep period is long relative to the 1. 4 s thermal time constant of the divertor target. An increase of the sweep frequency to ∼ 1 Hz is suggested: this increase would provide a power handling margin of ∼ 25% relative to present operational criteria

  10. Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

    Science.gov (United States)

    Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

    2016-10-01

    The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

  11. Evaluation of divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Ferrari, M.; Giancarli, L.; Kleefeldt, K.; Nardi, C.; Roedig, M.; Reimann, J.; Salavy, J.F.

    2001-01-01

    In the frame of the preliminary study of plants suitable for the energy production from the fusion power, particular emphasis has been given on the divertor studies. Since a significant percentage of the power generated from the fusion process is absorbed in the divertor, the thermal efficiency of the power conversion cycle requires a high coolant outlet temperature of the divertor, leading to solutions that are different from those adopted for the present experimental fusion plants. Therefore, copper alloys having extremely high thermal conductivity, cannot be used as structural material for this kind of devices. The most suitable coolants to be used in the divertor are water, helium and liquid metals. A conceptual design study has been developed for each of these three fluids, with the aim to evaluate the maximum allowable thermal flux at the divertor target plate and the R and D requirements for each solution. While a water-cooled divertor can be designed with a limited R and D effort, the development of helium or liquid metal cooled divertors requires a more engaging R and D program

  12. Operating conditions of the BPX divertor

    International Nuclear Information System (INIS)

    Hill, D.N.; Milovich, J.; Rognlien, T.; Braams, B.J.; Brooks, J.N.; Campbell, R.; Haines, J.; Knoll, D.; Prinja, A.; Stotler, D.P.; Ulrickson, M.

    1991-01-01

    In this paper we discuss the expected operating conditions at the divertor of the BPX tokamak (Burning Plasma Experiment), the next- step US tokamak proposed for the study of self-heated plasmas at Q ≅ 5 to ignition. In this double-null device (κ ≅ 2), the predicted first-wall loading is high because of is compact size (R = 2.6m, α = 0.8m, I p = 10.6 MA, and B T ) and its high projected fusion power output (100--500 MW with up to 20 MW of ICRH). Present designs call for inertially cooled carbon-based target plate material and X-point sweeping to handle the divertor heat flux during the 3--5 s flat-top at full power. The X-point is maintained about 15--20 cm off the target plates (a distance of ∼5m along field lines), which represents a reasonable compromise between lowering the divertor electron temperature (T e,d ) by increasing the connection length, and lowering the peak divertor heat flux (q d ) by increasing the magnetic flux expansion (which is about 15--20 in this case). It is planned for the BPX device to operate with H-mode confinement; ELMs are expected because of the relatively high power flow through the edge plasma (P sep ≅ 0.6 MW/m 2 for P fus = 500 MW). The ELMs will help reduce the impurity concentration in the core plasma (Z eff ≅ 1.7) and keep the density down, but should not add significantly to the divertor heat flux since their measured contribution to the global power balance drops with increasing input power

  13. Divertor design for the Tokamak Physics Experiment

    International Nuclear Information System (INIS)

    Hill, D.N.; Braams, B.

    1994-05-01

    In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

  14. High heat flux thermal-hydraulic analysis of ITER divertor and blanket systems

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Ioki, K.; Tivey, R.; Krassovski, D.; Kubik, D.

    1998-01-01

    Three separate cooling systems are used for the divertor and blanket components, based mainly on flow routing access and on grouping together components with the highest heat load levels and uncertainties: divertor, limiter/outboard baffle, and primary first wall/inboard baffle. The coolant parameters for these systems are set to accommodate peak heat load conditions with a reasonable critical heat flux (CHF) margin. Material temperature constraints and heat transport system space and cost requirements are also taken into consideration. This paper summarises the three cooling system designs and highlights the high heat flux thermal-hydraulic analysis carried out in converging on the design values for the coolant operating parameters. Application of results from on-going high heat flux R and D and a brief description of future R and D effort to address remaining issues are also included. (orig.)

  15. Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

    Energy Technology Data Exchange (ETDEWEB)

    Wang, L., E-mail: lwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dalian University of Technology, Dalian 116024 (China); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); General Atomics, P. O. Box 85608, San Diego, CA 92186 (United States); Li, J.; Wan, B.N.; Gong, X.Z.; Zhang, X.D.; Hu, J.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Association EURATOM-FZJ, D-52425 Jülich (Germany); Xu, G.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, X.L. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Maingi, R.; Menard, J.E. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Luo, G.N.; Gao, X.; Hu, L.Q.; Gan, K.F.; Liu, S.C.; Wang, H.Q.; Chen, R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

    2015-08-15

    Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m{sup 2} and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust.

  16. Divertor power load studies for attached L-mode single-null plasmas in TCV

    NARCIS (Netherlands)

    Maurizio, R.; Elmore, S.; Fedorczak, N.; Gallo, A.; Reimerdes, H.; Labit, B.; Theiler, C.; Tsui, C. K.; Vijvers, W. A. J.; TCV team,; MST1 Team,

    2018-01-01

    This paper investigates the power loads at the inner and outer divertor targets of attached, Ohmic L-mode, deuterium plasmas in the TCV tokamak, in various experimental situations using an Infrared thermography system. The study comprises variations of the outer divertor leg length and target flux

  17. Variation of Particle Control with Changes in Divertor Geometry

    International Nuclear Information System (INIS)

    Petrie, T W; Allen, S L; Brooks, N H; Fenstermacher, M E; Ferron, J R; Greenfield, C M; Groth, M; Hyatt, A W; Leonard, A W; Luce, T C; Mahdavi, M A; Murakami, M; Porter, G D; Rensink, M E; Schaffer, M J; Wade, M R; Watkins, J G; West, W P; Wolf, N S

    2004-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx(divergent)B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx(divergent)B ion particle drift direction. Our data suggests that the presence of Bx(divergent)B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED / n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks

  18. Variation of particle control with changes in divertor geometry

    International Nuclear Information System (INIS)

    Petrie, T.W.; Allen, S.L.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Porter, G.D.; Rensink, M.E.; Wolf, N.S.; Ferron, J.R.; Greenfield, C.M.; Hyatt, A.W.; Leonard, A.W.; Luce, T.C.; Mahdavi, M.A.; Schaffer, M.J.; West, W.P.; Murakami, M.; Wade, M.R.; Watkins, J.G.

    2005-01-01

    Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx∇B ion particle drift direction. Our data suggests that the presence of Bx∇B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED /n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks. (author)

  19. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

    1998-08-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T e > 40 eV) ELMing plasmas, and detached (T e 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y ≤ 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates at the OSP of an attached plasma (∼ 10 microm/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  20. Divertor erosion in DIII-D

    International Nuclear Information System (INIS)

    Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.

    1998-05-01

    Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D + ) ≤ 2.0 x 10 -3 ). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

  1. Divertor characterization experiments

    International Nuclear Information System (INIS)

    Porter, G.D.; Allen, S.; Fenstermacher, M.; Hill, D.; Brown, M.; Jong, R.A.; Rognlien, T.; Rensink, M.; Smith, G.; Stambaugh, R.; Mahdavi, M.A.; Leonard, A.; West, P., Evans, T.

    1996-01-01

    Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate

  2. Compatibility of detached divertor operation with robust edge pedestal performance

    Energy Technology Data Exchange (ETDEWEB)

    Leonard, A.W., E-mail: leonard@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Osborne, T.H.; Snyder, P.B. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States)

    2015-08-15

    The compatibility of detached radiative divertor operation with a robust H-mode pedestal is examined in DIII-D. A density scan produced low temperature plasmas at the divertor target, T{sub e} ⩽ 2 eV, with high radiation leading to a factor of ⩾4 drop in peak divertor heat flux. The cold radiative plasma was confined to the divertor and did not extend across the separatrix in X-point region. A robust H-mode pedestal was maintained with a small degradation in pedestal pressure at the highest densities. The response of the pedestal pressure to increasing density is reproduced by the EPED pedestal model. However, agreement of the EPED model with experiment at high density requires an assumption of reduced diamagnetic stabilization of edge Peeling–Ballooning modes.

  3. Divertor IR thermography on Alcator C-Moda)

    Science.gov (United States)

    Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

    2010-10-01

    Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

  4. The JET divertor coil

    International Nuclear Information System (INIS)

    Last, J.R.; Froger, C.; Sborchia, C.

    1989-01-01

    The divertor coil is mounted inside the Jet vacuum vessel and is able to carry 1 MA turns. It is of conventional construction - water cooled copper, epoxy glass insulation -and is contained in a thin stainless steel case. The coil has to be assembled, insulated and encased inside the Jet vacuum vessel. A description of the coil is given, together with technical information (including mechanical effects on the vacuum vessel), an outline of the manufacture process and a time schedule. (author)

  5. A study on rotational augmentation using CFD analysis of flow in the inboard region of the MEXICO rotor blades

    DEFF Research Database (Denmark)

    Guntur, Srinivas; Sørensen, Niels N.

    2015-01-01

    This work presents an analysis of data from existing as well as new full-rotor computational fluid dynamics computations on the MEXICO rotor, with focus on the flow around the inboard parts of the blades. The boundary layer separation characteristics on the airfoil sections in the inboard parts...

  6. Island divertor studies on W7-AS

    International Nuclear Information System (INIS)

    Sardei, F.; Feng, Y.; Grigull, P.; Herre, G.; Hildebrandt, D.; Hofmann, J.V.; Kisslinger, J.; Brakel, R.; Das, J.; Geiger, J.; Heinrich, O.; Kuehner, G.; Niedermeyer, H.; Reiter, D.; Richter-Gloetzl, M.; Runov, A.; Schneider, R.; Stroth, U.; Verbeek, H.; Wagner, F.; Wolf, R.

    1997-01-01

    Basic topological features of the island divertor concept for low shear stellarators are discussed with emphasis on the differences to tokamak divertors. Extensive measurements of the edge structures by two-dimensional plasma spectroscopy and by target calorimetry are in excellent agreement with predicted vacuum and equilibrium configurations, which are available up to central β values of ∝1%. For this β value the calculated field-line pitch inside the islands is twice that of the corresponding vacuum case. Video observations of the strike points indicate stability of the island structures for central β values up to ∝3.7%. The interpretation of the complex island divertor physics of W7-AS has become possible by the development of the three-dimensional plasma transport code EMC3 (Edge Monte Carlo 3D), which has been coupled self-consistently to the EIRENE neutral gas code. Analysis of high density NBI discharges gives strong indications of stable high recycling conditions for n e ≥10 20 m -3 . The observations are reproduced by the EMC3/EIRENE code and supported by calculations with the B2/EIRENE code adapted to W7-AS. Improvement of recycling, pumping and target load distribution is expected from the new optimized target plates and baffles to be installed in W7-AS. (orig.)

  7. Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

    Science.gov (United States)

    Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

    2018-05-01

    One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

  8. Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

    International Nuclear Information System (INIS)

    Lieder, G.; Napiontek, B.; Radtke, R.; Field, A.; Fussmann, G.; Kallenbach, A.; Kiemer, K.; Mayer, H.M.

    1993-01-01

    Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs

  9. Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lieder, G; Napiontek, B; Radtke, R; Field, A; Fussmann, G; Kallenbach, A; Kiemer, K; Mayer, H M [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1994-12-31

    Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs.

  10. The effect of charge exchange with neutral deuterium on carbon emission in JET divertor plasmas

    International Nuclear Information System (INIS)

    Maggi, C.; Horton, L.; Summers, H.

    1999-11-01

    High density, low temperature divertor plasma operation in tokamaks results in large neutral deuterium concentrations in the divertor volume. In these conditions, low energy charge transfer reactions between neutral deuterium and the impurity ions can in principle enhance the impurity radiative losses and thus help to reduce the maximum heat load to the divertor target. A quantitative study of the effect of charge exchange on carbon emission is presented, applied to the JET divertor. Total and state selective effective charge exchange recombination rate coefficients were calculated in the collisional radiative picture. These coefficients were coupled to divertor and impurity transport models to study the effect of charge exchange on the measured carbon spectral emission in JET divertor discharges. The sensitivity of the effect of charge exchange to the assumptions in the impurity transport model was also investigated. A reassessment was made of fundamental charge exchange cross section data in support of this study. (author)

  11. Divertor heat and particle control experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Baker, D.R.; Allen, S.L.

    1994-05-01

    In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D 2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models

  12. NSTX plasma operation with a Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

  13. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    1999-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  14. Boundary plasma control with the ergodic divertor

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Beyer, P.

    2001-01-01

    Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

  15. Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

    Science.gov (United States)

    Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

    2015-11-01

    China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

  16. JET with a pumped divertor -- Technical issues and main results

    International Nuclear Information System (INIS)

    Bertolini, E.

    1995-01-01

    The most recent modification to JET has been the installation of a single-null pumped divertor, for active control of plasma impurities. This is to address central physics issues relevant to the design of a next step tokamak. Experiments conducted during the 1994--95 campaign, with plasma currents up to 6MA, have shown that the Mark I divertor, which makes use of strike point sweeping across the target plates, is a suitable tool to control the influx of impurities in the plasma core. The operation of a tokamak with a pumped divertor has been characterized in detail. However the divertor configuration must be optimized to better meet ITER requirements. Therefore an improved (more closed) divertor structure, which may not require sweeping, is under assembly at present (Mark II). It is designed, in addition, to allow divertor tile structures to be fully replaceable by remote handling techniques, following D-T fusion experiments. New types of events involving electromechanical interactions of plasma with the vessel and in-vessel structural components have been encountered, due to plasma vertical instabilities and disruptions (such as toroidal asymmetries of vacuum vessel forces and side-ways vessel displacements). The physics and engineering experimental work performed in JET is primarily dedicated to the finalization of the ITER design

  17. Numerical studies on divertor experiments

    International Nuclear Information System (INIS)

    Ueda, N.; Itoh, K.; Itoh, S.-I.; Tanaka, M.; Hasegawa, M.; Shoji, T.; Sugihara, M.

    1988-04-01

    Numerical analysis on the divertor experiments such as JFT-2M tokamak is made by use of the two-dimensional time-dependent simulation code. The plasma in the scrape-off layer (SOL) and divertor region is solved for the given particle and heat sources from the main plasma, Γ p and Q T . Effect of the direction of the toroidal magnetic field is studied. It is found that the heat flux which is proportional to b vector x ∇T i has influences on the divertor plasmas, but has a small effect on the parameters on the midplane in the framework of the fluid model. Parameter survey on Γ p and Q T is made. The transient response of the SOL/divertor plasma to the sudden change of Γ p and Q T is studied. Time delay in the SOL and divertor region is calculated. (author)

  18. Innovations in the LHD divertor program

    International Nuclear Information System (INIS)

    Ohyabu, N.; Komori, A.; Noda, N.; Morisaki, T.; Sagara, A.; Suzuki, H.; Watanabe, T.; Motojima, O.; Takase, H.

    1995-01-01

    Various innovative divertor concepts have been developed to improve the LHD plasma performance. They are two divertor magnetic geometries (helical divertor configurations with and without n/m=1/1 island) and two operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). In addition, technological development of new efficient hydrogen pumping schemes are being pursued for enhancing the divertor control capability. 16 refs., 4 figs

  19. Design, R&D and commissioning of EAST tungsten divertor

    Science.gov (United States)

    Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

    2016-02-01

    After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

  20. A new scaling for divertor detachment

    Science.gov (United States)

    Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

    2017-05-01

    The ITER design, and future reactor designs, depend on divertor ‘detachment,’ whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P sep/R or P sep B/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-like scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, ‘advanced’ divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.

  1. Module of lithium divertor for KTM tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

    2012-10-15

    Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

  2. Hydrogen recycling and transport in the helical divertor of TEXTOR

    Energy Technology Data Exchange (ETDEWEB)

    Clever, Meike

    2010-07-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm{+-}0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  3. Hydrogen recycling and transport in the helical divertor of TEXTOR

    International Nuclear Information System (INIS)

    Clever, Meike

    2010-01-01

    The aim of this thesis was to investigate the hydrogen recycling at the target plates of the helical divertor in TEXTOR and by this the capability of this divertor configuration to access such favourable operational regimes. In order to study the different divertor density regimes in TEXTOR, discharges were performed in which the total plasma density was increased continuously up to the density limit. The recycling was investigated in a fixed helical divertor structure where four helical strike points with a poloidal width of about 8-10 cm are created at the divertor target plates. The experimental investigation of the hydrogen recycling was carried out using mainly spectroscopic methods supplemented by Langmuir probe, interferometric and atomic beam measurements. In the framework of this thesis a spectroscopic multi camera system has been built that facilitates the simultaneous observation of four different spectral lines, recording images of the divertor target plates and the plasma volume close to the target. The system facilitates the simultaneous measurement of the poloidal and toroidal pattern of the recycling flux at the divertor target without the need for sweeping the plasma structure. The simultaneous observation of different spectral lines reduces the uncertainty in the analysis based on several lines, as the contribution from uncertainties in the reproducibility of plasma parameters in different discharges are eliminated and only the uncertainty of the measurement method limits the accuracy. The spatial resolution of the system in poloidal and toroidal direction (0.8 mm±0.01 mm) is small compared to the separation of the helical strike points, the capability of the measurement method to resolve these structures is therefore limited by the line-of-sight integration and the penetration depth of the light emitting species. The measurements showed that the recycling flux increases linearly with increasing plasma density, a high recycling regime is not

  4. The divertor remote maintenance project

    International Nuclear Information System (INIS)

    Maisonnier, D.; Martin, E.; Akou, K.

    2001-01-01

    Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

  5. The divertor remote maintenance project

    International Nuclear Information System (INIS)

    Maisonnier, D.; Martin, E.; Akou, K.

    1999-01-01

    Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

  6. An analytic model for flow reversal in divertor plasmas

    International Nuclear Information System (INIS)

    Cooke, P.I.H.; Prinja, A.K.

    1987-04-01

    An analytic model is developed and used to study the phenomenon of flow reversal which is observed in two-dimensional simulations of divertor plasmas. The effect is shown to be caused by the radial spread of neutral particles emitted from the divertor target which can lead to a strong peaking of the ionization source at certain radial locations. The results indicate that flow reversal over a portion of the width of the scrape-off layer is inevitable in high recycling conditions. Implications for impurity transport and particle removal in reactors are discussed

  7. A survey of problems in divertor and edge plasma theory

    International Nuclear Information System (INIS)

    Boozer, A.; Braams, B.; Weitzner, H.; Hazeltine, R.; Houlberg, W.; Oktay, E.; Sadowski, W.; Wootton, A.

    1992-01-01

    Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings

  8. Scrape-off layer radiation and heat load to the ASDEX Upgrade LYRA divertor

    International Nuclear Information System (INIS)

    Kallenbach, A.; Kaufmann, M.; Coster, D.P.

    1999-01-01

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region. (author)

  9. Latest status of manufacturing activity of ITER divertor and engineering issues on tungsten divertor

    International Nuclear Information System (INIS)

    Suzuki, Satoshi

    2011-01-01

    Divertors for ITER are now in construction. In the present chapter, the specification and the latest status of manufacturing of ITER divertors are presented. In addition, issues in the development of divertors for the fusion demo reactor are given on the basis of experiences on the ITER divertor development. (J.P.N.)

  10. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-01-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

  11. TCV divertor upgrade for alternative magnetic configurations

    Directory of Open Access Journals (Sweden)

    H. Reimerdes

    2017-08-01

    Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

  12. Theory of Advanced Magnetic Divertors

    Science.gov (United States)

    Kotschenreuther, Michael; Valanju, Prashant; Mahajan, Swadesh; Covele, Brent

    2013-10-01

    The magnetic field structure in the SOL is the most important determinant of divertor physics. A comprehensive analytical and numerical methodology is developed to investigate SOL magnetic fields in the backdrop of two advanced divertor geometries- the X-divertor (XD) proposed and discussed in 2004, and the snowflake divertor (SFD) of 2007-2010. The analysis shows that XD and SFD represent very distinct and readily distinguishable magnetic geometries, epitomized through a differentiating metric, the Divertor Index (DI). In terms of this simple metric, the XD (DI > 1) and the SFD (DI XD flux surfaces are less convergent, in fact, divergent (flaring). These different SOL magnetics imply different physics, particularly with respect to detachment dynamics. It is also shown that some experiments on NSTX and DIII-D match both the prescription and the predictions of the 2004 XD paper. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

  13. The ITER divertor cassette project meeting

    International Nuclear Information System (INIS)

    Merola, M.; Riccardi, B.; Tivey, R.

    1999-01-01

    The Divertor Cassette Project topical meeting was held on May 26-28, 1999 at the ENEA Brasimone Research Centre in Camugnano (Bologna), Italy. Specialists from all the four Parties and the JCT participated in the meeting. It was concluded that the Divertor Cassette Project has significantly contributed to solving a large part of the critical issues of the ITER divertor design

  14. Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

    International Nuclear Information System (INIS)

    Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

    2006-01-01

    The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

  15. Plasma decontamination during ergodic divertor experiments in TORE SUPRA

    International Nuclear Information System (INIS)

    Monier-Garbet, P.; DeMichelis, C.; Fall, T.; Ghendrih, Ph.; Goniche, M.; Grosman, A.; Hess, W.; Mattioli, M.

    1991-01-01

    In Tore Supra an ergodic divertor (ED) has been integrated in the machine design and successfully operated, as already reported. This paper analyses the decontamination effect resulting from the creation of an ergodic boundary zone. Two plasma geometrical configurations (outboard and inboard) are studied, the plasma being limited respectively either, on the low field side (lfs), by an outboard limiter (3 to 5 cm ahead of the ED modules) or, on the high field side (hfs), by the graphite inner wall. Strong decontamination effects have already been reported for the first configuration by observing line emission of the intrinsic (carbon and oxygen) and purposely injected (nitrogen) impurities. When limited by the inner wall, the plasma is several centimeters farther from the ED modules than in the lfs configuration. The magnetic perturbation is then greatly reduced, and much smaller decontamination effects should be expected. In this paper, the hfs configuration data is compared with that from the lfs configuration. Preliminary experiments combining lower hybrid current drive and ED operation in the hfs configuration are also reported. (author) 5 refs., 4 figs

  16. Plasma facing components integration studies for the WEST divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ferlay, Fabien, E-mail: fabien.ferlay@cea.fr; Missirlian, Marc; Guilhem, Dominique; Firdaouss, Mehdi; Richou, Marianne; Doceul, Louis; Faisse, Frédéric; Languille, Pascal; Larroque, Sébastien; Martinez, André; Proust, Maxime; Louison, Céphise; Jeanne, Florian; Saille, Alain; Samaille, Frank; Verger, Jean-Marc; Bucalossi, Jérôme

    2015-10-15

    Highlights: • The divertor PFU integration has been studied regarding existing environment. • Magnetic, electric, thermal, hydraulic, mechanical loads and assembly are considered. - Abstract: In the context of the Tokamak Tore-Supra evolution, the CEA aims at transforming it into a test bench for ITER actively cooled tungsten (ACW) plasma facing components (PFC). This project named WEST (Tungsten Environment in Steady state Tokamak) is especially focused on the divertor target. The modification of the machine, by adding two axisymmetric divertors will make feasible an H-mode with an X-point close to the lower divertor. This environment will allow exposing the divertor ACW components up to 20 MW/m{sup 2} heat flux during long pulse. These specifications are well suited to test the ITER-like ACW target elements, respecting the ITER design. One challenge in such machine evolution is to integrate components in an existing vacuum vessel in order to obtain the best achievable performance. This paper deals with the design integration of ITER ACW target elements into the WEST environment considering magnetic, electric, thermal and mechanical loads. The feasibility of installation and maintenance has to be strongly considered as these PFC could be replaced several times. The ports size allows entering a 30° sector of pre-installed tungsten targets which will be plugged as quickly and easily as possible. The main feature of steady state operation is the active cooling, which leads to have many embedded cooling channels and bulky pipes on the PFC module including many connections and sealings between vacuum and water channels. The 30° sector design is now finalized regarding the ITER ACW elements specifications. No major modifications are expected.

  17. Actively convected liquid metal divertor

    International Nuclear Information System (INIS)

    Shimada, Michiya; Hirooka, Yoshi

    2014-01-01

    The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem. (letter)

  18. Snowflake divertor plasmas on TCV

    International Nuclear Information System (INIS)

    Piras, F; Coda, S; Furno, I; Moret, J-M; Sauter, O; Turri, G; Bencze, A; Duval, B P; Felici, F; Pochelon, A; Zucca, C; Pitts, R A; Tal, B

    2009-01-01

    Starting from a standard single null X-point configuration, a second order null divertor (snowflake (SF)) has been successfully created on the Tokamak a Configuration Variable (TCV) tokamak. The magnetic properties of this innovative configuration have been analysed and compared with a standard X-point configuration. For the SF divertor, the connection length and the flux expansion close to the separatrix exceed those of the standard X-point by more than a factor of 2. The magnetic shear in the plasma edge is also larger for the SF configuration.

  19. The dynamic ergodic divertor in TEXTOR-A novel tool for studying magnetic perturbation field effects

    International Nuclear Information System (INIS)

    Neubauer, O.; Czymek, G.; Finken, K.H.; Giesen, B.; Huettemann, P.W.; Lambertz, H.T.; Schruff, J.

    2005-01-01

    Recently TEXTOR has been upgraded by the installation of the dynamic ergodic divertor (DED). The purpose of the DED is to influence transport parameters in plasma edge and core and to study the resulting effects on heat exhaust, edge cooling, impurity screening, plasma confinement and stability. Alternatively, the DED creates static or rotating multipolar helical magnetic perturbation fields of different mode patterns. A set of 16 helical coils has been installed on the inboard high-field side of the vacuum vessel. Rotating fields of up to 10 kHz can be generated. A novel coil design has been developed which fulfills the various mechanical, electrical, high frequency, thermal, and vacuum requirements. In addition to the various technical aspects of the DED design, implementation and commissioning, highlights of recent experiments will be presented. In particular the impact of the perturbation field on MHD stability and plasma rotation will be addressed

  20. Advanced divertor experiments on DIII-D

    International Nuclear Information System (INIS)

    Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

    1991-04-01

    The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

  1. Computational Investigations of Inboard Flow Separation and Mitigation Techniques on Multi-Megawatt Wind Turbines

    Science.gov (United States)

    Chow, Raymond

    The aerodynamic characteristics of the NREL 5-MW rotor have been examined using a Reynolds-averaged Navier-Stokes method, OVERFLOW2. A comprehensive off-body grid independence study has been performed. A strong dependence on the size of the near-body wake grid has been found. Rapid diffusion of the wake appears to generate an overprediction of power and thrust. A large, continuous near-wake grid at minimum of two rotor diameters downstream of the rotor appears to be necessary for accurate predictions of near-body forces. The NREL 5-MW rotor demonstrates significant inboard flow separation up to 30% of span. This separation appears to be highly three-dimensional, with a significant amount of radial flow increasing the size of the separated region outboard. Both integrated aerodynamic coefficients and detailed wake structures for the baseline NREL 5-MW rotor are in excellent agreement with results by Riso at Uinfinity = 8 and 11 m/s. A simple, continuous full-chord fence was applied at the maximum chord location of the blade, within the region of separation. This non-optimized device reduced the boundary-layer cross-flow and resulting separation, and increased rotor power capture by 0.9% and 0.6% at U infinity = 8 and 11 m/s, respectively. Suction side only fences perform similarly in terms of power capture but reduce the increase in rotor thrust. Fence heights from 0.5% to 17.5% of the maximum chord all demonstrate some level of effectiveness, with fences (1-2.5%cmax) showing similar performance gains to taller fences with smaller penalties in thrust. Performance in terms of power capture is not very sensitive to spanwise location when placed within the separation region. Blunt trailing edge modifications to the inboard region of the blade showed a relatively significant effect on rotor power. Over a large range of trailing edge thicknesses from hTE = 10 to 25%c, power was found to increase by 1.4%. Thrust increased proportionally with the thicknesses examined

  2. One dimensional simulation on stability of detached plasma in a tokamak divertor

    International Nuclear Information System (INIS)

    Nakazawa, Shinji; Nakajima, Noriyoshi; Okamoto, Masao; Ohyabu, Nobuyoshi

    1999-06-01

    The stability of radiation front in the Scrape-Off-Layer (SOL) of a tokamak is studied with a one dimensional fluid code; the time-dependent transport equations are solved in the direction parallel to a magnetic field line. The simulation results show that stable detached solutions exist, where the plasma temperature near the divertor target is ∼2 eV. It is found that whenever such stable detached states are attained, the strong radiation front is contact with or at a small distance from the divertor target. When the energy externally injected into the SOL is decreased below a critical value, the radiation front starts to move towards the X-point, cooling the SOL plasma. In such cases, no stationary solutions such that the radiation front rests in the divertor channel are observed in our parameter space. This qualitatively corresponds to the results of tokamak divertor experiments which show the movement of radiation front. (author)

  3. SOLPS simulations of X-divertor in NSTX-U

    Science.gov (United States)

    Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

    2017-10-01

    The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

  4. Effects of divertor geometry and pumping on plasma performance on DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Hill, D.N.; Porter, G.D.

    1997-06-01

    This paper reports the status of an ongoing investigation to discern the influence of the divertor and plasma geometry on the confinement of both ELM-free and ELMing discharges in DIII-D. The ultimate goal is to achieve a high-performance core plasma which coexists with an advanced divertor plasma. The divertor plasma must reduce the heat flux to acceptable levels; the current technique disperses the heat flux over a wide area by radiation (a radiative divertor). To date, we have obtained our best performance in double-null (DN) high-triangularity (δ ∼ 0.8) ELM-free discharges. As discussed in detail elsewhere, there are several advantages for both the core and divertor plasma with highly-shaped DN operation. Previous radiative-divertor experiments with D 2 injection in DN high-δ ELMing H-mode have shown that this configuration is more sensitive to gas puffing (τ decreases). Moving the X-point away from the target plate (to ∼15 cm above the plate) decreases this sensitivity. Preliminary measurements also indicate that gas puffing reduces the divertor heat flux but does not reduce the plasma pressure along the field line. The up/down heat flux balance can be varied magnetically (by changing the distance between the separatrices), with a slight magnetic imbalance required to balance the heat flux. The overall mission of the Radiative Divertor Project (RDP) is to install a fully pumped and baffled high-δ DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-δ (δ∼ 0.4) plasmas, so much of the divertor physics has been performed in LSN; these results are discussed in Section 2. As part of the first phase of the RDP, we have installed a new high-δ USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case

  5. Progress in ergodic divertor operation on Tore Supra

    International Nuclear Information System (INIS)

    Ghendrih, Ph.; Becoulet, M.; Colas, L.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Azeroual, A.; Basiuk, V.; Beaumont, B.; Becoulet, A.; Bremond, S.; Bucalossi, J.; Capes, H.; Corre, Y.; Costanzo, L.; Michelis, C. de; Devynck, P.; Feron, S.; Friant, C.; Garbet, X.; Giannella, R.; Grisolia, C.; Hess, W.; Hogan, J.; Ladurelle, L.; Laugier, F.; Martin, G.; Mattioli, M.; Meslin, B.; Monier-Garbet, P.; Moulin, D.; Nguyen, F.; Pascal, J.Y.; Pecquet, A.L.; Pegourie, B.; Reichle, R.; Saint-Laurent, F.; Vallet, J.C.; Zabiego, M.

    1999-09-01

    Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 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 in associated with an increase of the fluctuating electric drift velocity. The bottom line is then 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 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, 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 ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

  6. Visible spectroscopy in the DIII-D divertor

    International Nuclear Information System (INIS)

    Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

    1996-06-01

    Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges

  7. Critical need for MFE: the Alcator DX advanced divertor test facility

    Science.gov (United States)

    Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

    2013-10-01

    Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

  8. Fabrication of a full-size mock-up for inboard 10o section of ITER vacuum vessel thermal shield

    International Nuclear Information System (INIS)

    Chung, W.; Nam, K.; Noh, C.H.; Kang, D.K.; Kang, S.M.; Oh, Y.G.; Choi, S.W.; Kang, S.H.; Utin, Y.; Ioki, K.; Her, N.; Yu, J.

    2011-01-01

    A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10 o section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.

  9. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    International Nuclear Information System (INIS)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

    2013-01-01

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes

  10. Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

    Energy Technology Data Exchange (ETDEWEB)

    Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

    2013-10-15

    Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

  11. Divertor plasma modification by divertor biasing and edge ergodization in JFT-2M

    International Nuclear Information System (INIS)

    Shoji, T.; Nagashima, K.; Tamai, H.; Ohdachi, S.; Miura, Y.; Ohasa, K.; Maeda, H.; Ohyabu, N.; Leonard, A.W.; Aikawa, H.; Fujita, T.; Hoshino, K.; Kawashima, H.; Matsuda, T.; Maeno, M.; Mori, M.; Ogawa, H.; Shimada, M.; Uehara, K.; Yamauchi, T.

    1995-01-01

    The effects of divertor biasing and edge ergodization on the divertor plasma have been investigated in the JFT-2M tokamak. Experimental results show; (1) The differential divertor biasing can change the in/out asymmetry of the divertor plasma. It especially changes the density on the ion side divertor plasma. The in/out electron pressure difference has a good correlation with the biasing current. (2) The unipolar divertor biasing can change the density profile of divertor plasma. The radial electric field and shear flow are the cause for this change. (3) The electron temperature of the divertor plasma in the H-mode with frequent ELMs induced by edge ergodization is lower than that of usual H-mode. That is due to the enhancement of the radial particle flux by frequent ELMs, ((orig.))

  12. Fabrication of divertor cassette for ITER

    International Nuclear Information System (INIS)

    Sanguinetti, G.P.

    2008-01-01

    The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

  13. A Lithium Vapor Box Divertor Similarity Experiment

    Science.gov (United States)

    Cohen, Robert A.; Emdee, Eric D.; Goldston, Robert J.; Jaworski, Michael A.; Schwartz, Jacob A.

    2017-10-01

    A lithium vapor box divertor offers an alternate means of managing the extreme power density of divertor plasmas by leveraging gaseous lithium to volumetrically extract power. The vapor box divertor is a baffled slot with liquid lithium coated walls held at temperatures which increase toward the divertor floor. The resulting vapor pressure differential drives gaseous lithium from hotter chambers into cooler ones, where the lithium condenses and returns. A similarity experiment was devised to investigate the advantages offered by a vapor box divertor design. We discuss the design, construction, and early findings of the vapor box divertor experiment including vapor can construction, power transfer calculations, joint integrity tests, and thermocouple data logging. Heat redistribution of an incident plasma-based heat flux from a typical linear plasma device is also presented. This work supported by DOE Contract No. DE-AC02-09CH11466 and The Princeton Environmental Institute.

  14. Development of divertor remote maintenance system

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    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 and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  15. Development of divertor remote maintenance system

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji

    1998-01-01

    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 and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

  16. ELM induced divertor heat loads on TCV

    Science.gov (United States)

    Marki, J.; Pitts, R. A.; Horacek, J.; Tskhakaya, D.; TCV Team

    2009-06-01

    Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor ˜3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying ˜1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

  17. ELM induced divertor heat loads on TCV

    Energy Technology Data Exchange (ETDEWEB)

    Marki, J., E-mail: janos.marki@epfl.c [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Pitts, R.A. [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Horacek, J. [Institute of Plasma Physics, Association EUROATOM-IPP.CR, Za Slovankou 3, 182 00 Prague 8 (Czech Republic); Tskhakaya, D. [Association EURATOM-OAW, Institut fuer Theoretische Physik, A-6020 Innsbruck (Austria)

    2009-06-15

    Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor approx3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying approx1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

  18. Energy and particle transport in the radiative divertor plasmas of DIII-D

    International Nuclear Information System (INIS)

    Leonard, A.W.; Allen, S.L.; Brooks, N.H.

    1997-06-01

    It has been argued that divertor energy transport dominated by parallel electron thermal conduction, or q parallel = -kT 5/2 2 dT e /ds parallel, leads to severe localization of the intense radiating region and ultimately limits the fraction of energy flux that can be radiated before striking the divertor target. This is due to the strong T 5/2 e dependence of electron heat conduction which results in very short spatial scales of the T e gradient at high power densities and low temperatures where deuterium and impurities radiate most effectively. However, we have greatly exceeded this constraint on DIII-D with deuterium gas puffing which reduces the peak heat flux to the divertor plate a factor of 5 while distributing the divertor radiation over a long length

  19. Matted-fiber divertor tagets for sputter resistance

    International Nuclear Information System (INIS)

    Gierszewski, P.J.; Todreas, N.E.; Mikic, B.; Yang, T.F.

    1981-06-01

    Reductions in net sputtering yields can be obtained by altering the surface topography to maximize redeposition of sputtered atoms. A simple analysis is used to indicate a potential reduction by a factor of 2 to 5 for matted fiber divertor targets, relatively independent of incident, reflected and sputtered atom distributions. The fiber temperature is also shown to be acceptable, even up to 10 MW/m 2 , for reasonably combinations of materials, fiber diameter and fiber spacing

  20. Divertor scenario development for NSTX Upgrade

    Science.gov (United States)

    Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

    2012-10-01

    In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

  1. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  2. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  3. Divertor plate for thermonuclear reactor

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro; Sato, Keisuke; Nishio, Satoshi.

    1993-01-01

    In a divertor plate for a thermonuclear reactor, adjacent cooling pipes are electrically insulated from each other and pipes made of a gradient functional material prepared by compositing ceramics having an insulation property and metals are metallurgically joined to at least one portion of each of the cooling pipes. Electric current caused upon occurrence of plasma disruption is interrupted by the insulation portion, so that a large circuit is not formed and electromagnetic force is decreased to such a extent that the divertor plate is not ruptured. Since a header of the cooling pipes can be installed at any optional position, the installation space can be reduced. Further, since inlet and exit collection headers can be disposed on both ends of the cooling pipes, it is possible to shorten the length of the cooling pipe of the divertor plate corresponded to high heat fluxes and reduce the pressure loss on the side of coolants to about 1/2. Further, turn back portions of small radius of curvature of the cooling pipes are eliminated to reduce the cost and extend the lifetime and, in addition, protection tiles can be attached easily. (N.H.)

  4. The lithium vapor box divertor

    International Nuclear Information System (INIS)

    Goldston, R J; Schwartz, J; Myers, R

    2016-01-01

    It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m −2 , implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma. (paper)

  5. Development of a compact W-shaped pumped divertor in JT-60U

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, S.; Hosogane, N.; Masaki, K.; Kodama, K.; Sasajima, T.; Kishiya, K.; Takahashi, S.; Shimizu, K.; Akino, N.; Miyo, Y.; Hiratsuka, H.; Saidoh, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Inoue, M.; Umakoshi, T.; Onozuka, M.; Morimoto, M. [Mitsubishi Heavy Industries, Wadasaki-cho, Hyogo-ku, Kobe-shi, 642 (Japan)

    1998-09-01

    In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.) 11 refs.

  6. X-Divertor Geometries for Deeper Detachment Without Degrading the DIII-D H-Mode

    Science.gov (United States)

    Covele, Brent; Kotschenreuther, M. T.; Valanju, P. M.; Mahajan, S. M.; Leonard, A. W.; Hyatt, A. W.; McLean, A. G.; Thomas, D. M.; Guo, H. Y.; Watkins, J. G.; Makowski, M. A.; Hill, D. N.

    2015-11-01

    Recent DIII-D experiments comparing the standard divertor (SD) and X-Divertor (XD) geometries show heat and particle flux reduction at the divertor target plate. The XD features large poloidal flux expansion, increased connection length, and poloidal field line flaring, quantified by the Divertor Index. Both SD and XD were pushed deep into detachment with increased gas puffing, until core energy confinement and pedestal pressure were substantially reduced. As expected, outboard target heat fluxes are significantly reduced in the XD compared to the SD under similar upstream plasma conditions, even at low Greenwald fraction. The high-triangularity (floor) XD cases show larger reduction in temperature, heat, and particle flux relative to the SD in all cases, while low-triangularity (shelf) XD cases show more modest reductions over the SD. Consequently, heat flux reduction and divertor detachment may be achieved in the XD with less gas puffing and higher pedestal pressures. Further causative analysis, as well as detailed modeling with SOLPS, is underway. These initial experiments suggest the XD as a promising candidate to achieve divertor heat flux control compatible with robust H-mode operation. Work supported by US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-04ER54754, and DE-FG02-04ER54742.

  7. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 2. Comprehending the divertor structure

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

    2006-01-01

    Divertor is given the largest heat load in the in-vessel components of fusion machine. The functions and conditions of divertor are stated from the point of view of thermal and structural dynamics. The way of thinking of structure design of divertor of JT-60 and the ITER (International Thermonuclear Experimental Reactor) is explained. As the conditions of divertor, the materials for large heat load, heat removal, pressure boundary, control of damage, and thermal stress/strain are considered. The divertor has to be changed periodically. The materials are required the heat removal function for high heat load. CuCrZr will be used to cooling tube and heat sink, and CFC materials for the surface. The cross section of ITER, a part of divertor, heat load of divertor and other components, the thermal conductivity of CFC and metal materials, conditions of cooling water for divertor of BWR, PWR and ITER, the thermal stress produced on rod, vertical target of ITER, structure of cooling tube, distribution of temperature and critical heart flux of inner wall of cooling tube, and fatigue clack of cooling tube are shown. (S.Y.)

  8. Model of divertor biasing and control of scrape-off layer and divertor plasmas

    International Nuclear Information System (INIS)

    Nagasaki, K.; Itoh, K.; Itoh, S.

    1991-02-01

    Analytic model of the divertor biasing is described. For the given plasma and energy sources from the core plasma, the heat and particle flux densities on the divertor plate as well as scrape-off-layer (SOL)/divertor plasmas are analyzed in a slab model. Using a two-dimensional model, the effects of the divertor biasing and SOL current are studied. The conditions to balance the plasma temperature or sheath potential on different divertor plates are obtained. Effect of the SOL current on the heat channel width is also discussed. (author)

  9. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    International Nuclear Information System (INIS)

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m 2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10 19 n/cm 2 . In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H 2 O/LiNO 3 layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is 0 C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of ∼2 were found at the hot spots. 2 refs., 6 figs., 2 tabs

  10. Engineering conceptual design of CFETR divertor

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Xuebing, E-mail: pengxb@ipp.cas.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Mao, Xin [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Chen, Peiming; Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China)

    2015-10-15

    Highlights: • Three divertor structures for two plasma configurations, ITER-like and snowflake. • Property of enlarging wet area for all three divertors is analyzed. • The divertor accommodating with both the plasma configurations is unfeasible. • Divertor cooling system is developed. - Abstract: The China Fusion Engineering Test Reactor (CFETR), which is in conceptual design phase, aims at producing fusion power of 50–200 MW with tritium breeding ratio of ∼1.2 and duty cycle time of 0.3–0.5. Its designed main parameters are major/minor radii of 5.7 m/1.6 m and plasma current of 10 MA. Although the fusion power is lower than the one of ITER, the relative smaller machine dimensions and planed much higher auxiliary heating power of 100–140 MW make that the power exhausting for the CFETR divertor is a very critical issue. To solve this issue, the divertor should be better designed with advanced physical operation mode, advanced configuration/geometry or high efficient cooling structure. In the paper, much effort was put on the divertor configuration and geometry. With designed magnet system, three divertor configurations can be realized, ITER-like, snowflake and super-X. However, considering structural design feasibility and remote handling compatibility, only the first two configurations were selected for the first step of engineering design. Three divertors were designed. They have different first wall geometries to accommodate with different plasma configurations, one for the ITER-like, one for the snowflake and the third one for both the configurations. All three divertors employ the same cassette body as the support and the cooling water manifold for the first wall. This feature simplifies the interface of the divertor to other components in the vacuum vessel. Besides, the cooling structure and the remote maintenance concept are also introduced in the paper.

  11. Moving Divertor Plates in a Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.J.; Zhang, H.

    2009-01-01

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

  12. Optimization of a bundle divertor for FED

    International Nuclear Information System (INIS)

    Hively, L.M.; Rothe, K.E.; Minkoff, M.

    1982-01-01

    Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

  13. Moving Divertor Plates in a Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    S.J. Zweben, H. Zhang

    2009-02-12

    Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

  14. The effect of feedback-controlled divertor nitrogen seeding on the boundary plasma and power exhaust channel width in Alcator C-Mod

    Science.gov (United States)

    Labombard, B.; Brunner, D.; Kuang, A. Q.; McCarthy, W.; Terry, J. L.

    2017-10-01

    The scrape-off layer (SOL) power channel width, λq, is projected to be 0.5 mm in power reactors, based on multi-machine measurements of divertor target heat fluxes in H-mode at low levels of divertor dissipation. An important question is: does λq change with the level of divertor dissipation? We report results in which feedback controlled nitrogen seeding in the divertor was used to systematically vary divertor dissipation in a series of otherwise identical L-mode plasmas at three plasma currents: 0.55, 0.8 and 1.1 MA. Outer midplane profiles were recorded with a scanning Mirror Langmuir Probe; divertor plasma conditions were monitored with `rail' Langmuir probe and surface thermocouple arrays. Despite an order of magnitude reduction in divertor target heat fluxes (q// 400 MW m-2 to 40 MW m-2) and corresponding change in divertor regime from sheath-limited through high-recycling to near-detached, the upstream electron temperature profile is found to remain unchanged or to become slightly steeper in the near SOL and to drop significantly in the far SOL. Thus heat in the SOL appears to take advantage of this impurity radiation `heat sink' in the divertor by preferentially draining via the narrow (and perhaps an increasingly narrow) λq of the near SOL. Supported by USDoE award DE-FC02-99ER54512.

  15. Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

    Directory of Open Access Journals (Sweden)

    Yuki Homma

    2017-08-01

    Full Text Available Erosion reduction of tungsten (W divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

  16. Magnetic field models and their application in optimal magnetic divertor design

    Energy Technology Data Exchange (ETDEWEB)

    Blommaert, M.; Reiter, D. [Institute of Energy and Climate Research (IEK-4), FZ Juelich GmbH, Juelich (Germany); Baelmans, M. [KU Leuven, Department of Mechanical Engineering, Leuven (Belgium); Heumann, H. [TEAM CASTOR, INRIA Sophia Antipolis (France); Marandet, Y.; Bufferand, H. [Aix-Marseille Universite, CNRS, PIIM, Marseille (France); Gauger, N.R. [TU Kaiserslautern, Chair for Scientific Computing, Kaiserslautern (Germany)

    2016-08-15

    In recent automated design studies, optimal design methods were introduced to successfully reduce the often excessive heat loads that threaten the divertor target surface. To this end, divertor coils were controlled to improve the magnetic configuration. The divertor performance was then evaluated using a plasma edge transport code and a ''vacuum approach'' for magnetic field perturbations. Recent integration of a free boundary equilibrium (FBE) solver allows to assess the validity of the vacuum approach. It is found that the absence of plasma response currents significantly limits the accuracy of the vacuum approach. Therefore, the optimal magnetic divertor design procedure is extended to incorporate full FBE solutions. The novel procedure is applied to obtain first results for the new WEST (Tungsten Environment in Steady-state Tokamak) divertor currently under construction in the Tore Supra tokamak at CEA (Commissariat a l'Energie Atomique, France). The sensitivities and the related divertor optimization paths are strongly affected by the extension of the magnetic model. (copyright 2016 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  17. A carbon-carbon panel design concept for the inboard limiter of the Compact Ignition Tokamak (CIT)

    International Nuclear Information System (INIS)

    Mantz, H.C.; Bowers, D.A.; Williams, F.R.; Witten, M.A.

    1989-01-01

    The inboard limiter of the Compact Ignition Tokamak (CIT) must protect the vacuum vessel from the plasma energy. This limiter region must withstand nominal heat fluxes in excess of 10 MW/m 2 and in addition it must be designed to be remotely maintained. Carbon-carbon composite material was selected over bulk graphite materials for the limiter design because of its ability to meet the thermal and structural requirements. The structural design concept consists of carbon-carbon composite panels attached to the vacuum vessel by a hinged rod/retainer concept. Results of the preliminary design study to define this inboard limiter are presented. The design concept is described along with the analyses of the thermal and structural response during nominal plasma operation and during plasma disruption events. 2 refs., 8 figs

  18. Design and analysis of the DII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

    1995-10-01

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electromagnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 degrees C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  19. Design and analysis of the DIII-D radiative divertor water-cooled structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Phelps, R.D.; Redler, K.M.; Reis, E.E.

    1995-01-01

    The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electro-magnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

  20. Rapidly Moving Divertor Plates In A Tokamak

    International Nuclear Information System (INIS)

    Zweben, S.

    2011-01-01

    It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

  1. Reactor application of an improved bundle divertor

    International Nuclear Information System (INIS)

    Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

    1978-11-01

    A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

  2. Comparison between stellarator and tokamak divertor transport

    International Nuclear Information System (INIS)

    Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

    2010-11-01

    The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

  3. Snowflake divertor experiments on TCV

    International Nuclear Information System (INIS)

    Piras, F; Coda, S; Duval, B P; Labit, B; Marki, J; Moret, J-M; Pitzschke, A; Sauter, O; Medvedev, S Yu

    2010-01-01

    An ELMy H-mode 'snowflake' (SF) divertor is established and studied for the first time in the TCV tokamak. The H-mode access and the edge localized mode (ELM) dynamics are compared with a conventional single-null diverted configuration. The SF configuration exhibits 15% higher confinement and 2-3 times lower ELM frequency. Ideal MHD stability analysis suggests enhanced stability of the SF H-mode pedestal to mid- to high-toroidal-mode-number modes. The capability of the SF to redistribute the edge power on the additional strike points has been confirmed experimentally.

  4. Effect of neutrals localized at torus inboard side on the impurity transport in edge stochastic magnetic field layer of LHD

    International Nuclear Information System (INIS)

    Morita, S.; Oishi, T.; Kobayashi, M.; Goto, M.; Kawamura, G.; Zhang, H.M.; Hunag, X.L.; Wang, E.H.

    2014-01-01

    Two-dimensional (2-D) distribution of impurity line emissions has been measured in Large Helical Device (LHD) based on the 2-D extreme ultraviolet (EUV) spectroscopy for studying the edge impurity transport in stochastic magnetic field layer with three-dimensional (3-D) structure. The impurity behavior in the vicinity of two X-points at inboard and outboard sides of torus becomes separately visible with the 2-D measurement. As a result, it is found that the carbon location changes from inboard to outboard X-points when the plasma axis is shifted from R_a_x=3.6 m to 3.75 m. A 3-D simulation with EMC3-EIRENE code agrees with the result at R_a_x=3.75 m but disagreed with the result at R_a_x=3.60 m. The discrepancy between the measurement and simulation at R_a_x=3.60 m is considerably reduced when the effect of neutral hydrogen localized at the inboard side is taken into account, which can modify the density gradient and friction force along the magnetic field. (author)

  5. Operating windows of pebble divertor

    International Nuclear Information System (INIS)

    Matsuhiro, K.; Isobe, M.; Ohtsuka, Y.; Ueda, Y.; Nishikawa, M.

    2001-01-01

    A marked feature of the pebble divertor is an effect by use of functional multi-layer coated pebble, which consists of a surface plasma facing layer, an intermediate tritium permeation barrier layer, and a kernel for heat removal. The dimensions, structure and the irradiation conditions of pebbles are the important issues for the development of the pebble divertor. From the view point of resistance of the induced thermal stress, the pebble is taken as small as possible in size. On the other hand, from the view point of the pumping performance, the suitable irradiation temperature range of the surface layer of pebble was estimated from the experiments and the numerical analysis. The pumping process enhanced by dynamic retention is available to extend the higher allowable irradiation temperature range from 900K to 1100K. As taking the temperature rise limitation due to pumping effect and the fractural strength due to the induced thermal stress limitation, it was found that the diameter of the pebble is possible to be 1-2 mm in about 20 MW/m 2 for the SiC kernel and 2-3 mm in less than 30 MW/m 2 for the graphite kernel. (author)

  6. Multiple equilibria of divertor plasmas

    International Nuclear Information System (INIS)

    Vu, H.X.; Prinja, A.K.

    1993-01-01

    A one-dimensional, two-fluid transport model with a temperature-dependent neutral recycling coefficient is shown to give rise to multiple equilibria of divertor plasmas (bifurcation). Numerical techniques for obtaining these multiple equilibria and for examining their stability are presented. Although these numerical techniques have been well known to the scientific community, this is the first time they have been applied to divertor plasma modeling to show the existence of multiple equilibria as well as the stability of these solutions. Numerical and approximate analytical solutions of the present one-dimensional transport model both indicate that there exists three steady-state solutions corresponding to (1) a high-temperature, low-density equilibrium, (2) a low-temperature, high-density equilibrium, and (3) an intermediate-temperature equilibrium. While both the low-temperature and the high-temperature equilibria are stable, with respect to small perturbations in the plasma conditions, the intermediate-temperature equilibrium is physically unstable, i.e., any small perturbation about this equilibrium will cause a transition toward either the high-temperature or low-temperature equilibrium

  7. Heat removal capability of divertor coaxial tube assembly

    International Nuclear Information System (INIS)

    Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

    1994-05-01

    To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications. (author)

  8. Particle recirculation in the ergodic divertor of Tore Supra

    International Nuclear Information System (INIS)

    Gunn, J.P.; Azeoual, A.; Becoulet, M.

    1999-01-01

    The present paper addresses the issue of particle recirculation in discharges where low energy flux to ergodic divertor target plates is achieved, in highly radiating detached ohmic plasmas. Plasma temperature and particle flux are measured by flush-mounted probes in the divertor plates, and by an upstream fast scanning Mach probe. The scalings with core density of the ion flux and electron temperature are well described by the simple two-point model used in axisymmetric poloidal divertors. The detachment signature is a pressure drop that occurs when the edge temperature falls below 10 eV. The parallel ion flux gradient is always positive, indicating that recombination is unlikely to play an important role in detachment. Visible spectroscopy of a neutralizer plate shows that attainment of cold detached plasmas near the density limit coincides with an abrupt increase of fueling for both deuterium and impurities. A feedback algorithm based on real time Langmuir probe measurements has been developed to monitor detachment and avoid disruptions. (authors)

  9. Design of DIII-D advanced divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thruston, G.

    1989-01-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffing. The Advanced Divertor has two principal components: ( 1) a toroidally symmetric baffle; and (2) a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. 2 refs., 4 figs

  10. Design of DIII-D Advanced Divertor

    International Nuclear Information System (INIS)

    Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thurston, G.

    1989-11-01

    The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffling. The Advanced Divertor has two principal components: a toroidally symmetric baffle; and a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. 2 refs., 4 figs

  11. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

    An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

  12. Status of the ITER full-tungsten divertor shaping and heat load distribution analysis

    International Nuclear Information System (INIS)

    Carpentier-Chouchana, S; Hirai, T; Escourbiac, F; Durocher, A; Fedosov, A; Ferrand, L; Kocan, M; Kukushkin, A S; Jokinen, T; Komarov, V; Lehnen, M; Merola, M; Mitteau, R; Pitts, R A; Sugihara, M; Firdaouss, M; Stangeby, P C

    2014-01-01

    In September 2011, the ITER Organization (IO) proposed to begin operation with a full-tungsten (W) armoured divertor, with the objective of taking a decision on the final target material (carbon fibre composite or W) by the end of 2013. This period of 2 years would enable the development of a full-W divertor design compatible with nuclear operations, the investigation of further several physics R and D aspects associated with the use of W targets and the completion of technology qualification. Beginning with a brief overview of the reference heat load specifications which have been defined for the full-W engineering activity, this paper will report on the current status of the ITER divertor shaping and will summarize the results of related three-dimensional heat load distribution analysis performed as part of the design validation. (paper)

  13. Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)]. E-mail: dagatae@itereu.de; Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr. 2, D-85748 Garching (Germany)

    2005-11-15

    The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints.

  14. Towards the development of workable acceptance criteria for the divertor CFC monoblock armour

    International Nuclear Information System (INIS)

    D'Agata, E.; Tivey, R.

    2005-01-01

    The plasma-facing components (PFCs) of the ITER divertor will be subjected to high heat flux (HHF). Carbon-fibre composite (CFC) is selected as the armour for the region of highest heat flux where the scrape-off layer of the plasma intercepts the vertical targets (VT). Failure of the armour to heat sink joints will compromise the performance of the divertor and could ultimately result in its failure and the shut down of the ITER machine. There are tens of thousands of CFCs to CuCrZr joints. The aim of the PFC design is to ensure that the divertor can continue to function even with the failure of a few joints. In preparation for writing the procurement specification for the ITER vertical target PFCs, a programme of work is underway with the objective of defining workable acceptance criteria for the PFC armour joints

  15. Intermittent Divertor Filaments in the National Spherical Torus Experiment and Their Relation to Midplane Blobs

    International Nuclear Information System (INIS)

    Maqueda, R.J.; Stotler, D.P.

    2010-01-01

    While intermittent filamentary structures, also known as blobs, are routinely seen in the low-field-side scrape-off layer of the National Spherical Torus Experiment (NSTX) (Ono et al 2000 Nucl. Fusion 40 557), fine structured filaments are also seen on the lower divertor target plates of NSTX. These filaments, not associated with edge localized modes, correspond to the interaction of the turbulent blobs seen near the midplane with the divertor plasma facing components. The fluctuation level of the neutral lithium light observed at the divertor, and the skewness and kurtosis of its probability distribution function, is similar to that of midplane blobs seen in D α ; e.g. increasing with increasing radii outside the outer strike point (OSP) (separatrix). In addition, their toroidal and radial movement agrees with the typical movement of midplane blobs. Furthermore, with the appropriate magnetic topology, i.e. mapping between the portion of the target plates being observed into the field of view of the midplane gas puff imaging diagnostic, very good correlation is observed between the blobs and the divertor filaments. The correlation between divertor plate filaments and midplane blobs is lost close to the OSP. This latter observation is consistent with the existence of 'magnetic shear disconnection' due to the lower X-point, as proposed by Cohen and Ryutov (1997 Nucl. Fusion 37 621).

  16. Development of a full-size divertor cassette prototype for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [Sandia National Labs., Albuquerque, NM (United States); Vieider, G.; Pacher, H.D. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany). NET Design Team] [and others

    1996-10-01

    Production of a full-size divertor cassette involves eight major components. All of the components are mounted on the cassette body. Inner divertor channel components for the vertical target design are being provided by the Japan Home Team. Outer divertor channel components for the vertical target design are being provided by the European and United States Home Teams. Gas box liners are being provided by the Russian Home Team. The full-size components manufactured by the four parties will be shipped to the US Home Team for assembly into a full size divertor cassette. The techniques for assembly and maintenance of the cassette will be demonstrated during this process. The assembled cassette will be tested for proper flow distribution and proof of the filling and draining procedures. The testing will include vacuum leak tightness at full temperature and pressure, cyclic heating to 150 {degrees}C, verification of dimensional accuracy of the assembled components, and application of thermal gradients to measure dimensional stability. The development of the divertor for the International Thermonuclear Experimental Reactor (ITER) depends on successful R&D efforts on materials, joining, and plasma materials interactions. Results of the development program are presented. The scale-up of the processes developed in the basic research and development tasks is accomplished by producing and high-heat-flux testing medium and full-scale mock- ups. The design of the mock-ups is discussed.

  17. Design of the Wendelstein 7-X inertially cooled Test Divertor Unit Scraper Element

    Energy Technology Data Exchange (ETDEWEB)

    Lumsdaine, Arnold, E-mail: lumsdainea@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Boscary, Jean [Max Planck Institute for Plasma Physics, Garching (Germany); Fellinger, Joris [Max Planck Institute for Plasma Physics, Greifswald (Germany); Harris, Jeff [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hölbe, Hauke; König, Ralf [Max Planck Institute for Plasma Physics, Greifswald (Germany); Lore, Jeremy; McGinnis, Dean [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Neilson, Hutch; Titus, Peter [Princeton Plasma Physics Lab, Princeton, NJ (United States); Tretter, Jörg [Max Planck Institute for Plasma Physics, Garching (Germany)

    2015-10-15

    Highlights: • The justification for the installation of the Test Divertor Unit Scraper Element is given. • Specially designed operational scenarios for the component are presented. • Plans for the design of the component are detailed. - Abstract: The Wendelstein 7-X stellarator is scheduled to begin operation in 2015, and to achieve full power steady-state operation in 2019. Computational simulations have indicated that for certain plasma configurations in the steady-state operation, the ends of the divertor targets may receive heat fluxes beyond their qualified technological limit. To address this issue, a high heat-flux “scraper element” (HHF-SE) has been designed that can protect the sensitive divertor target region. The surface profile of the HHF-SE has been carefully designed to meet challenging engineering requirements and severe spatial limitations through an iterative process involving physics simulations, engineering analysis, and computer aided design rendering. The desire to examine how the scraper element interacts with the plasma, both in terms of how it protects the divertor, and how it affects the neutral pumping efficiency, has led to the consideration of installing an inertially cooled version during the short pulse operation phase. This Test Divertor Unit Scraper Element (TDU-SE) would replicate the surface profile of the HHF-SE. The design and instrumentation of this component must be completed carefully in order to satisfy the requirements of the machine operation, as well as to support the possible installation of the HHF-SE for steady-state operation.

  18. Development of a full-size divertor cassette prototype for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Vieider, G.; Pacher, H.D.

    1996-01-01

    Production of a full-size divertor cassette involves eight major components. All of the components are mounted on the cassette body. Inner divertor channel components for the vertical target design are being provided by the Japan Home Team. Outer divertor channel components for the vertical target design are being provided by the European and United States Home Teams. Gas box liners are being provided by the Russian Home Team. The full-size components manufactured by the four parties will be shipped to the US Home Team for assembly into a full size divertor cassette. The techniques for assembly and maintenance of the cassette will be demonstrated during this process. The assembled cassette will be tested for proper flow distribution and proof of the filling and draining procedures. The testing will include vacuum leak tightness at full temperature and pressure, cyclic heating to 150 degrees C, verification of dimensional accuracy of the assembled components, and application of thermal gradients to measure dimensional stability. The development of the divertor for the International Thermonuclear Experimental Reactor (ITER) depends on successful R ampersand D efforts on materials, joining, and plasma materials interactions. Results of the development program are presented. The scale-up of the processes developed in the basic research and development tasks is accomplished by producing and high-heat-flux testing medium and full-scale mock- ups. The design of the mock-ups is discussed

  19. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

    1995-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

  20. Stability, divertors and innovative concepts

    International Nuclear Information System (INIS)

    Mirnov, S.

    2003-01-01

    This paper contains a short resume of the sections on 'Stability, Divertors and Innovative Concepts' presented at the 19th IAEA Fusion Energy Conference. The main conclusions are: (1) the problem of type I ELMs in tokamaks seems to be not so dramatic; (2) it was demonstrated that the working pulse length of large thermonuclear devices can achieve 100 s and more; (3) the problem of tritium retention seems to be not so dramatic now; probable approaches of its solution are visible; (4) active methods of plasma instabilities suppression (NTM, RWM, sawteeth, external MHD) work successfully; (5) new methods of mitigation of the disruption consequences were offered. New technological ideas and new ideas on magnetic confinement were presented. (author)

  1. Divertor cassette movers prototypes for ITER

    International Nuclear Information System (INIS)

    Bogusch, E.; Batz, R.; Bieber, O.; Gottfried, R.; Cerdan, G.

    1998-01-01

    Following competitive tendering, in October 1996 Siemens was contracted by the European Commission to design and supply an assembly of four Divertor Cassette Movers Prototypes including the control and command systems for the movers proper. The assembly consisting of one Cassette Toroidal Mover (CTM), one Radial Mover Tractor (TRC), one Second Cassette Carrier (SCC), and one Radial Cassette Carrier (RCC) represents key components of the Divertor Test Platform at Brasimone, one of the seven large R+D projects for ITER. By detailed design, high-precision manufacturing and testing of these devices, Siemens contributed to the verification of an important task within the European R and D program towards ITER construction. Replacement of the divertor cassettes is a scheduled maintenance operation throughout the life of ITER. The successful fabrication and testing of the Divertor Cassette Movers Prototypes is all important milestone to verify this delicate operation. (authors)

  2. Stochasticity about a poloidal divertor separatrix

    International Nuclear Information System (INIS)

    Skinner, D.A.; Osborne, T.H.; Prager, S.C.; Park, W.

    1986-10-01

    The stochasticization of the magnetic separatrix due to the presence of a helical perturbation in a poloidal divertor tokamak is illustrated by a numerical computation which traces magnetic field lines

  3. Simulation of tungsten erosion and transport near the divertor plate during ELMs by a kinetic method

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Zhenyue; Sang, Chaofeng; Hu, Wanpeng; Du, Hailong; Wang, Dezhen, E-mail: wangdez@dlut.edu.cn

    2016-11-01

    Highlights: • A kinetic method is used to simulate tungsten erosion and transport during ELMs. • The erosion of tungsten plate by different species (deuterium and carbon ions) is shown. • The charge states of sputtered tungsten particles are given statistically. - Abstract: Tungsten (W) is fore seen as one of the most important candidates of the plasma-facing materials (PFM) for future fusion devices, due to its beneficial properties. However, the high-Z characteristic makes it a potential contamination to the core plasma. Divertor is the main component that directly contacts the plasma, therefore, it is very important to understand the erosion of W divertor plate and the corresponding transport of the eroded wall impurity, especially during edge localized modes (ELMs). In this work, a one-dimension-in-space and three-dimensions-in-velocity particle-in-cell code (EPPIC1D) is used to simulate the erosion of W divertor plate, and the transport of eroded W impurity near the divertor plate is studied by a Monte Carlo code. Benefiting from the kinetic simulation, energy/particle flux to the target could be calculated accurately, and the erosion of W plate by different species is simulated during ELMs. The trajectories and distributions of eroded W impurity particles are demonstrated, which shows us a basic idea of how these impurity particles are generated and transported. It is found that C{sup 3+} plays a dominated role on the erosion of W divertor plate during ELMs even when its concentration is low. Both W atoms and ions distribute mainly near the divertor plate, indicating only a very small fraction of W impurity particles could escape from divertor region and penetrate into the core plasma.

  4. ARIES-III divertor engineering design

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.; Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S.; Herring, J.S.; Valenti, M.; Steiner, D.

    1992-01-01

    This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m 2 , a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m 2 . The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed

  5. ARIES-III divertor engineering design

    Energy Technology Data Exchange (ETDEWEB)

    Wong, C.P.C.; Schultz, K.R. [General Atomics, San Diego, CA (United States); Cheng, E.T. [TSI Research, Solana Beach, CA (United States); Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering; Brooks, J.N.; Ehst, D.A.; Sze, D.K. [Argonne National Lab., IL (United States); Herring, J.S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Valenti, M.; Steiner, D. [Rensselaer Polytechnic Inst., Troy, NY (United States). Plasma Dynamics Lab.

    1992-01-01

    This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m{sup 2}, a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m{sup 2}. The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed.

  6. Recent advances towards a lithium vapor box divertor

    Directory of Open Access Journals (Sweden)

    R.J. Goldston

    2017-08-01

    Full Text Available Fusion power plants are likely to require near complete detachment of the divertor plasma from the divertor target plates, in order to have both acceptable heat flux at the target to avoid prompt damage and also acceptable plasma temperature at the target surface, to minimize long-term erosion. However hydrogenic and impurity puffing experiments show that detached operation leads easily to x-point MARFEs, impure plasmas, degradation in confinement, and lower helium pressure at the exhaust. The concept of the Lithium Vapor Box Divertor is to use local evaporation and strong differential pumping through condensation to localize low-Z gas-phase material that absorbs the plasma heat flux and so achieve detachment while avoiding these difficulties. The vapor localization has been confirmed using preliminary Navier–Stokes calculations. We use ADAS calculations of εcool, the plasma energy lost per injected lithium atom, to estimate the lithium vapor pressure, and so temperature, required for detachment, taking into account power balance. We also develop a simple model of detachment to evaluate the required upstream density, based on further taking into account dynamic pressure balance. A remarkable general result is found, not just for lithium-vapor-induced detachment, that the upstream density divided by the Greenwald-limit density scales as nup/nGW ∝ (P5/8/B3/8 Tdet1/2/(εcool+γTdet, with no explicit size scaling. Tdet is the temperature just before strong pressure loss, assumed to be ∼ ½ of the ionization potential of the dominant recycling species, and γ is the sheath heat transmission factor.

  7. FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

    International Nuclear Information System (INIS)

    O'NEIL, RC; STAMBAUGH, RD

    2002-01-01

    OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities

  8. Plasma-neutral gas interaction in a tokamak divertor: effects of hydrogen molecules and plasma recombination

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Pigarov, A.Yu.; Soboleva, T.K.; Sigmar, D.J.

    1997-01-01

    We investigate the influence of hydrogen molecules on plasma recombination using a collisional-radiative model for multispecies hydrogen plasmas and tokamak detached divertor parameters. The rate constant found for molecular activated recombination of a plasma can be as high as 2 x 10 -10 cm 3 /s, confirming our pervious estimates. We investigate the effects of hydrogen molecules and plasma recombination on self-consistent plasma-neutral gas interactions in the recycling region of a tokamak divertor. We treat the plasma flow in a fluid approximation retaining the effects of plasma recombination and employing a Knudsen neutral transport model for a 'gas box' divertor geometry. For the model of plasma-neutral interactions we employ we find: (a) molecular activated recombination is a dominant channel of divertor plasma recombination; and (b) plasma recombination is a key element leading to a decrease in the plasma flux onto the target and substantial plasma pressure drop which are the main features of detached divertor regimes. (orig.)

  9. Effects of the New Island Divertor on the Plasma Performance in the W7-AS Stellarator

    International Nuclear Information System (INIS)

    Grigull, P.; McCormick, K.; Baldzuhn, J.; Burhenn, R.; Brakel, R.; Ehmler, H.; Feng, Y.; Gadelmeier, F.; Giannone, L.; Hartmann, D.; Hildebrandt, D.; Hirsch, M.; Jaenicke, R.; Kisslinger, J.; Klinger, T.; Knauer, J.; Koenig, R.; Naujoks, D.; Niedermeyer, H.; Pasch, E.

    2003-01-01

    The island divertor in the W7-AS stellarator enables access to a new NBI-heated, high density operating regime with promising confinement properties. This regime -- the High Density H-Mode -- displays no evident mode activity, is extant above a threshold density and characterized by flat density profiles, high energy- and low impurity-confinement times and edge localized radiation. Impurity accumulation, normally associated with ELM-free H-modes, is avoided. Quasi steady-state discharges with n e up to 4 1020 m-3, edge radiation levels up to 90%, and partial plasma detachment at the divertor targets can be simultaneously realized

  10. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    International Nuclear Information System (INIS)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A; Constans, S; Merola, M; Riccardi, B

    2009-01-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  11. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Science.gov (United States)

    Escourbiac, F.; Richou, M.; Guigon, R.; Constans, S.; Durocher, A.; Merola, M.; Schlosser, J.; Riccardi, B.; Grosman, A.

    2009-12-01

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  12. Modelling of island divertor physics and comparison to W7-AS experimental results

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Grigull, P.; McCormick, K.; Giannone, L.; Kisslinger, J.; Reiter, D.; Igitkhanov, Y.; Wenzel, U.

    2003-01-01

    Extensive parameter studies have been carried out with the EMC3-EIRENE code. Major code predictions, namely the absence of high recycling prior to detachment, additional momentum losses associated with the specific island divertor geometry and the jump of the radiation at detachment transition have been verified by the W7-AS divertor experiments. Measurements and simulations are compared for high density, high power W7-AS divertor discharges and the physics related to rollover and detachment is discussed in detail. Local comparisons with the W7-AS experiment have been started with a new code version accounting for the real open-island geometry. Specifically, the observed asymmetric power unloading of the target plates at detachment transition could now be reproduced and explained. Agreement with the experiment was also found for the unexpected spatial structure of particle deposition by including classical ExB drifts into the code

  13. Definition of acceptance criteria for the ITER divertor plasma-facing components through systematic experimental analysis

    Energy Technology Data Exchange (ETDEWEB)

    Escourbiac, F; Richou, M; Guigon, R; Durocher, A; Schlosser, J; Grosman, A [CEA/IRFM, F-13108, Saint-Paul-lez-Durance (France); Constans, S [AREVA-NP, Le Creusot (France); Merola, M [ITER Organization, Cadarache (France); Riccardi, B [Fusion For Energy, Barcelona (Spain)], E-mail: frederic.escourbiac@cea.fr

    2009-12-15

    Experience has shown that a critical part of the high-heat flux (HHF) plasma-facing component (PFC) is the armour to heat sink bond. An experimental study was performed in order to define acceptance criteria with regards to thermal hydraulics and fatigue performance of the International Thermonuclear Experimental Reactor (ITER) divertor PFCs. This study, which includes the manufacturing of samples with calibrated artificial defects relevant to the divertor design, is reported in this paper. In particular, it was concluded that defects detectable with non-destructive examination (NDE) techniques appeared to be acceptable during HHF experiments relevant to heat fluxes expected in the ITER divertor. On the basis of these results, a set of acceptance criteria was proposed and applied to the European vertical target medium-size qualification prototype: 98% of the inspected carbon fibre composite (CFC) monoblocks and 100% of tungsten (W) monoblock and flat tiles elements (i.e. 80% of the full units) were declared acceptable.

  14. Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

    International Nuclear Information System (INIS)

    Chen Yiping; Liu Songlin

    2010-01-01

    Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

  15. Snowflake divertor configuration studies in National Spherical Torus Experiment

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

    2012-08-15

    Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

  16. Atomic and molecular processes in JT-60U divertor plasmas

    Energy Technology Data Exchange (ETDEWEB)

    Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1997-01-01

    Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

  17. Snowflake divertor configuration studies for NSTX-Upgrade

    International Nuclear Information System (INIS)

    Soukhanovskii, V.A.

    2011-01-01

    Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

  18. Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

    Science.gov (United States)

    Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

    2018-04-01

    A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

  19. Divertor Design and Physics Issues of Huge Power Handling for SlimCS Demo Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Asakura, N.; Hoshino, K.; Tobita, K.; Someya, Y.; Utoh, H.; Nakamura, M., E-mail: asakura.nobuyuki@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka (Japan); Takizuka, T. [Osaka University, Osaka (Japan)

    2012-09-15

    Full text: Power exhaust scenario for a 3 GW class fusion reactor with the ITER-size plasma has been developed with enhancing the radiation loss from seeding impurity. Transport of plasma, impurity and neutrals was simulated self-consistently, for the first time, under the Demo divertor condition using an integrated divertor simulation code SONIC. The total heat load, q{sub target}, was evaluated including radiation power load and neutral load, in addition to the plasma heat load. It was found that heat and particle diffusion coefficients significantly affect the plasma detachment. For the case of increasing the coefficients by the factor of two, peak q{sub target} is reduced from 18 MW/m{sup 2} to below the engineering design level of 10 MW/m{sup 2}, while the characteristic width of the heat flux at the midplane SOL increases slightly from 2.2 to 2.7 mm. It was also found that that enhancement of the local {chi} and D at the outer SOL affects a reduction in the peak q{sub target} near the separatrix. Effects of the divertor geometry such as the divertor leg were investigated. Outer divertor leg length was extended to 2.7 m, while the magnetic flux expansion at the target is reduced to a half compared to the reference case of 1.8 m. Large radiation volume is shifted further upstream from the target due to a reduction in T{sub e}. The peak q{sub target} decreases to 10 MW/m{sup 2} due to reduction in both the plasma heat load and the radiation power load. (author)

  20. Drift wave turbulence studies on closed and open flux surfaces: effect limiter/divertor plates location

    International Nuclear Information System (INIS)

    Ribeiro, T.; Scott, B.

    2007-01-01

    The field line connection of a tokamak sheared magnetic field has an important impact on turbulence, by ensuring a finite parallel dynamical response for every degree of freedom available in the system. This constitutes the main property which distinguishes closed from open flux surfaces in such a device. In the latter case, the poloidal periodicity of the magnetic field is replaced by a Debye sheath arising where the field lines strike the limiter/divertor plates. This is enough to break the field line connection constraint and allow the existence of convective cell modes, leading to a change in the character of the turbulence from drift wave- (closed flux surfaces) to interchange-type (open flux surfaces), and hence increasing the turbulent transport observed. Here we study the effect of changing the poloidal position of the limiter/divertor plates, using the three-dimensional electromagnetic gyrofluid turbulence code GEM, which has time dependently self consistent field aligned flux tube coordinates. For the closed flux surfaces, the globally consistent periodic boundary conditions are invoked, and for open flux surfaces a standard Debye sheath is used at the striking points. In particular, the use of two limiter positions simultaneously, top and bottom, is in order, such to allow a separation between the inboard and outboard sides of the tokamak. This highlights the differences between those two regions of the tokamak, where the curvature is either favourable (former) or unfavourable (latter), and further makes room for future experimental qualitative comparisons, for instance, on double null configurations of the tokamak ASDEX Upgrade. (author)

  1. Design and Analysis of the Cryopump for the DIII-D Upper Divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Baxi, C.B.; Bozek, A.S.

    1999-01-01

    A cryocondensation pump for the upper inboard divertor on DIII-D is to be installed in the vacuum vessel in the fall of 1999. The cryopump removes neutral gas particles from the divertor and prevents recycling to the plasma. This pump is designed for a pumping speed of 18,000 ell/s at 0.4 mTorr. The cryopump is toroidally continuous to minimize inductive voltages and avoid electrical breakdown during disruptions. The cryopump consists of a 25 mm Inconel tube cooled by liquid helium and is surrounded by nitrogen cooled shields. A segmented ambient temperature radiation/particle shield protects the nitrogen shields. The pump is subjected to a steady state heat load of less than 10 W due to conduction and radiation heat transfer. The helium tube will be subjected to Joule heating of less than 300 J due to induced current and a particle load of less than 12 W during plasma operation. The thermal design of the cryopump requires that it be cooled by 5 g/s liquid helium at an inlet pressure of 115 kPa and a temperature of 4.35 K. Thermal analysis and tests show that the helium tube can absorb a transient heat load of up to 100 W for 10 s and still pump deuterium at 6.3 K. Disruptions induce toroidal currents in the helium line and nitrogen shields. These currents cross the rapidly changing magnetic fields, applying complex dynamic loads on the cryopump. The forces on the pump are extrapolated from magnetic measurements from DIII-D plasma disruptions and scaled to a 3 MA disruption. The supports for the nitrogen shield consist of a racetrack design, which are stiff for reacting the disruption loads, but are radially flexible to allow differential thermal displacements with the vacuum vessel. Static and dynamic finite element analyses of the cryopump show that the stresses and displacements over a range of disruption and thermal loadings are acceptable

  2. A symplectic map for trajectories of magnetic field lines in double-null divertor tokamaks

    Science.gov (United States)

    Crank, Willie; Ali, Halima; Punjabi, Alkesh

    2009-11-01

    The coordinates of the area-preserving map equations for integration of magnetic field line trajectories in tokamaks can be any coordinates for which a transformation to (ψ,θ,φ) coordinates exists [A. Punjabi, H. Ali, T. Evans, and A. Boozer, Phys. Lett. A 364, 140 (2007)]. ψ is toroidal magnetic flux, θ is poloidal angle, and φ is toroidal angle. This freedom is exploited to construct a map that represents the magnetic topology of double-null divertor tokamaks. For this purpose, the generating function of the simple map [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)] is slightly modified. The resulting map equations for the double-null divertor tokamaks are: x1=x0-ky0(1-y0^2 ), y1=y0+kx1. k is the map parameter. It represents the generic topological effects of toroidal asymmetries. The O-point is at (0.0). The X-points are at (0,±1). The equilibrium magnetic surfaces are calculated. These surfaces are symmetric about the x- and y- axes. The widths of stochastic layer near the X-points in the principal plane, and the fractal dimensions of the magnetic footprints on the inboard and outboard side of upper and lower X-points are calculated from the map. This work is supported by US Department of Energy grants DE-FG02-07ER54937, DE-FG02-01ER54624 and DE-FG02-04ER54793.

  3. Integrated core-edge-divertor modeling studies

    International Nuclear Information System (INIS)

    Stacey, W.M.

    2001-01-01

    An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

  4. Technology R&D Activities for the ITER Full-tungsten Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lorenzetto, P.; Bednarek, M.; Gavila, P.; Riccardi, B.; Saibene, G., E-mail: patrick.lorenzetto@f4e.europa.eu [Fusion for Energy, Barcelona (Spain); Escourbiac, F.; Hirai, T.; Merola, M.; Pitts, R. [ITER Organization, St Paul-lez-Durance (France); Suzuki, S. [JAEA, Ibaraki (Japan); Mazul, I. [Efremov Institute, St.Petersburg (Russian Federation)

    2012-09-15

    Full text: The current ITER Baseline foresees the use of carbon fibre composite (CFC) as armour material in the high heat flux strike point regions and tungsten (W) elsewhere in the divertor for the initial non-active phase of operation with hydrogen and helium plasmas. This divertor would then be replaced with a full-W divertor for the nuclear phase with deuterium and deuterium- tritium plasmas. To reduce costs the ITER Organization (IO) has proposed to install a full-W divertor from start of operations and to implement a work programme to develop a full-W divertor design, qualify the corresponding fabrication technology and investigate critical physics and operational issues with support from the R&D fusion community. An extensive R&D programme has been implemented over more than 15 years to develop fabrication technologies for the procurement of ITER divertor components. Significant effort has been devoted to the development of reliable armour/heat sink joining techniques such as Hot Isostatic Pressing (Europe), Hot Radial Pressing (Europe) or brazing (Japan, Russia). In this development programme, established for the CFC/W divertor variant, the design solution for W-armoured components was optimized for the divertor baffle and dome regions, namely for steady state operation conditions at heat flux values of typically 5 MW/m{sup 2} and for slow transient events at heat flux values up to 10 MW/m{sup 2}. A very positive outcome of this R&D work has been that some fabrication technologies mentioned above can achieve much higher performances, close to the expected slow transient conditions for the strike point region (20 MW/m{sup 2} for 10 s). To prepare for the procurement of a full-W divertor, a development work programme has been launched including in particular the manufacturing and high heat flux testing of small-scale mock-ups with improved monoblock geometries and full-W pre-qualification prototypes, and the manufacturing and testing of qualification full

  5. Conceptual design studies for the European DEMO divertor: Rationale and first results

    International Nuclear Information System (INIS)

    You, J.H.; Mazzone, G.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, S.; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.

    2016-01-01

    Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

  6. Conceptual design studies for the European DEMO divertor: Rationale and first results

    Energy Technology Data Exchange (ETDEWEB)

    You, J.H., E-mail: you@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Mazzone, G.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Bachmann, Ch. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Autissier, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Barrett, T. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cocilovo, V.; Crescenzi, F. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Domalapally, P.K. [Research Cnter Rez, Hlavní 130, 250 68 Husinec–Řež (Czech Republic); Dongiovanni, D. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Entler, S. [Institute of Plasma Physics CAS, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Federici, G. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Frosi, P. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fursdon, M. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Hancock, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marzullo, D. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); McIntosh, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Müller, A.V. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Porfiri, M.T. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); and others

    2016-11-01

    Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

  7. The effects of particle recycling on the divertor plasma: A particle-in-cell with Monte Carlo collision simulation

    Science.gov (United States)

    Chang, Mingyu; Sang, Chaofeng; Sun, Zhenyue; Hu, Wanpeng; Wang, Dezhen

    2018-05-01

    A Particle-In-Cell (PIC) with Monte Carlo Collision (MCC) model is applied to study the effects of particle recycling on divertor plasma in the present work. The simulation domain is the scrape-off layer of the tokamak in one-dimension along the magnetic field line. At the divertor plate, the reflected deuterium atoms (D) and thermally released deuterium molecules (D2) are considered. The collisions between the plasma particles (e and D+) and recycled neutral particles (D and D2) are described by the MCC method. It is found that the recycled neutral particles have a great impact on divertor plasma. The effects of different collisions on the plasma are simulated and discussed. Moreover, the impacts of target materials on the plasma are simulated by comparing the divertor with Carbon (C) and Tungsten (W) targets. The simulation results show that the energy and momentum losses of the C target are larger than those of the W target in the divertor region even without considering the impurity particles, whereas the W target has a more remarkable influence on the core plasma.

  8. Divertor power and particle fluxes between and during type-I ELMs in the ASDEX Upgrade

    Science.gov (United States)

    Kallenbach, A.; Dux, R.; Eich, T.; Fischer, R.; Giannone, L.; Harhausen, J.; Herrmann, A.; Müller, H. W.; Pautasso, G.; Wischmeier, M.; ASDEX Upgrade Team

    2008-08-01

    Particle, electric charge and power fluxes for type-I ELMy H-modes are measured in the divertor of the ASDEX Upgrade tokamak by triple Langmuir probes, shunts, infrared (IR) thermography and spectroscopy. The discharges are in the medium to high density range, resulting in predominantly convective edge localized modes (ELMs) with moderate fractional stored energy losses of 2% or below. Time resolved data over ELM cycles are obtained by coherent averaging of typically one hundred similar ELMs, spatial profiles from the flush-mounted Langmuir probes are obtained by strike point sweeps. The application of simple physics models is used to compare different diagnostics and to make consistency checks, e.g. the standard sheath model applied to the Langmuir probes yields power fluxes which are compared with the thermographic measurements. In between ELMs, Langmuir probe and thermography power loads appear consistent in the outer divertor, taking into account additional load due to radiation and charge exchange neutrals measured by thermography. The inner divertor is completely detached and no significant power flow by charged particles is measured. During ELMs, quite similar power flux profiles are found in the outer divertor by thermography and probes, albeit larger uncertainties in Langmuir probe evaluation during ELMs have to be taken into account. In the inner divertor, ELM power fluxes from thermography are a factor 10 larger than those derived from probes using the standard sheath model. This deviation is too large to be caused by deficiencies of probe analysis. The total ELM energy deposition from IR is about a factor 2 higher in the inner divertor compared with the outer divertor. Spectroscopic measurements suggest a quite moderate contribution of radiation to the target power load. Shunt measurements reveal a significant positive charge flow into the inner target during ELMs. The net number of elementary charges correlates well with the total core particle loss

  9. First results from the dynamic ergodic divertor at TEXTOR

    International Nuclear Information System (INIS)

    Lehnen, M.; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O.

    2005-01-01

    Experimental results from the dynamic ergodic divertor (DED) at TEXTOR are given, describing the complex structure of the edge plasma and the properties of the divertor as well as its influence on the plasma rotation

  10. Divertor scaling laws for tokamaks

    International Nuclear Information System (INIS)

    Catto, P.J.; Krasheninnikov, S.I.; Connor, J.W.

    1997-01-01

    The breakdown of two body scaling laws is illustrated by using the two dimensional plasma code UEDGE coupled to an advanced Navier-Stokes neutrals transport package to model attached and detached regimes in a simplified geometry. Two body similarity scalings are used as benchmarks for runs retaining non-two body modifications due to the effects of (i) multi-step processes altering ionization and radiation via the excited states of atomic hydrogen and (ii) three body recombination. Preliminary investigations indicate that two body scaling interpretations of experimental data fail due to (i) multi-step processes when a significant region of the plasma exceeds a plasma density of 10 19 m -3 , or (ii) three body recombination when there is a significant region in which the temperature is ≤1 eV while the plasma density is ≥10 20 m -3 . These studies demonstrate that two body scaling arguments are often inappropriate in the divertor and the first results for alternate scalings are presented. (orig.)

  11. Geometrical properties of a 'snowflake' divertor

    International Nuclear Information System (INIS)

    Ryutov, D. D.

    2007-01-01

    Using a simple set of poloidal field coils, one can reach the situation in which the null of the poloidal magnetic field in the divertor region is of second order, not of first order as in the usual X-point divertor. Then, the separatrix in the vicinity of the null point splits the poloidal plane not into four sectors, but into six sectors, making the whole structure look like a snowflake (hence the name). This arrangement allows one to spread the heat load over a much broader area than in the case of a standard divertor. A disadvantage of this configuration is that it is topologically unstable, and, with the current in the plasma varying with time, it would switch either to the standard X-point mode, or to the mode with two X-points close to each other. To avoid this problem, it is suggested to have a current in the divertor coils that is roughly 5% higher than in an ''optimum'' regime (the one in which a snowflake separatrix is formed). In this mode, the configuration becomes stable and can be controlled by varying the current in the divertor coils in concert with the plasma current; on the other hand, a strong flaring of the scrape-off layer still remains in force. Geometrical properties of this configuration are analyzed. Potential advantages and disadvantages of this scheme are discussed

  12. Engineering and design aspects related to the development of the ITER divertor

    International Nuclear Information System (INIS)

    Dietz, J.; Chiocchio, S.; Antipenkov, A.

    1994-01-01

    Most of the divertor concepts proposed for the Next Step devices relied on the exhaust of the SOL power to target plates which intersect the magnetic field fines. The resulting highly peaked thermal load, together with the concentrated fluxes of energetic particles, posed severe design constraints and ultimately led to unacceptably short target lifetime. The ITER high density gas target divertor concept is based on transferring the nominal power perpendicular to the magnetic field lines from the plasma edge onto large surfaces and on dissipating the particles' energy through atomic and molecular mechanisms. While the basic ideas for this approach have been motivated by recent results in present tokamaks, a full assessment of this concept still requires extensive experimental and modelling work. The paper describes the engineering and design aspects involving the development of the ITER divertor and shows how the physics assumptions translate into engineering requirements, and how the additional existing constraints (such as the limited space, neutron load, electromagnetic effects, compatibility with other components, remote maintainability) have been taken into account for the design definition. The concept developed takes advantage of the spatial separation of the several physics phenomena anticipated to take place in the divertor, thus relaxing the needs to accommodate in the same region opposing requirements

  13. Experimental studies of the snowflake divertor in TCV

    NARCIS (Netherlands)

    Labit, B.; Canal, G. P.; Christen, N.; Duval, B. P.; Lipschultz, B.; Lunt, T.; Nespoli, F.; Reimerdes, H.; Sheikh, U.; Theiler, C.; Tsui, C. K.; Verhaegh, K.; Vijvers, W. A. J.

    2017-01-01

    To address the risk that, in a fusion reactor, the conventional single-null divertor (SND) configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD), are investigated in TCV. The expected benefits of the SFD-minus in terms of

  14. Neutral particle retention in the JET MK I divertor

    International Nuclear Information System (INIS)

    Ehrenberg, J.K.; Campbell, D.J.; Harbour, P.J.; Horton, L.D.; Loarte, A.; McCormick, G.K.; Monk, R.D.; Saibene, G.R.; Simonini, R.; Taroni, A.; Stamp, M.F.

    1997-01-01

    Retention of neutral deuterium and nitrogen in the JET MK I divertor has been investigated. Results show that ohmic plasma detachment reduces deuterium retention, that the magnetic divertor configuration has some influence on the achievable deuterium retention, and that nitrogen in nitrogen-seeded steady state detached H-mode discharges accumulates in the divertor. (orig.)

  15. Thermal effects of runaway electrons in an armoured divertor

    International Nuclear Information System (INIS)

    Stad, R.C.L. van der.

    1993-12-01

    This report describes the results of a numerical thermal analysis of the heat deposition of runaway electrons accompanying plasma disruptions in a armoured divertor. The divertor concepts studied are carbon on molybdenum and beryllium on copper. The conclusion is that the runaway electrons can cause melting of the armour as well as melting of the structure and can damage the divertor severely. (orig.)

  16. Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

    International Nuclear Information System (INIS)

    Du, Hailong; Sang, Chaofeng; Wang, Liang; Bonnin, Xavier; Sun, Jizhong; Wang, Dezhen

    2016-01-01

    Highlights: • The in-out divertor asymmetry is studied using SOLPS. • The discharge operation and the magnetic filed have a great influence on the divertor asymmetry. • The asymmetry is not obvious in low recycling regime as that in high recycling regime. - Abstract: This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

  17. Effects of discharge operation regimes and magnetic field geometry on the in-out divertor asymmetry in EAST

    Energy Technology Data Exchange (ETDEWEB)

    Du, Hailong; Sang, Chaofeng [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Liang [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Bonnin, Xavier [LSPM-CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Sun, Jizhong [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China); Wang, Dezhen, E-mail: wangdez@dlut.edu.cn [Key Laboratory of Materials Modification by Laser, Ion and Electron Beams (Ministry of Education), School of Physics and Optoelectronic Technology, Dalian University of Technology, Dalian 116024 (China)

    2016-11-01

    Highlights: • The in-out divertor asymmetry is studied using SOLPS. • The discharge operation and the magnetic filed have a great influence on the divertor asymmetry. • The asymmetry is not obvious in low recycling regime as that in high recycling regime. - Abstract: This paper aims to investigate the reason why the divertor in-out asymmetry was not obvious as experimentally observed in EAST only considering the classical drifts from previous simulations (Guo et al., J. Nucl. Mater. 438 (2013) 280; Du et al., J. Nucl. Mater. 463 (2015) 485). With consideration of the classical drifts, a series of different typical discharge scenarios in EAST in different magnetic field geometries were simulated by using the SOLPS5.2 code package. The simulated results reveal that the classical drifts make a major contribution to the in-out divertor asymmetry in the high recycling regime (HRR) and partial detachment (one divertor target begins to detach, while the other divertor remains attached) regime. In comparison, in low recycling regime the classical drifts play a much smaller role in the contributions to the in-out divertor asymmetry, which can explain reasonably the reason for it in Guo et al. (J. Nucl. Mater. 438 (2013) 280). In addition, the magnetic field geometry also has a great impact on the classical drifts inducing the asymmetry; it is found that for lower single-null, upper single-null and connected double-null topologies, in HRR the classical drifts play an dominant role in the contribution to the in-out divertor asymmetry, while for a disconnected double null magnetic field configuration, they play a minor role, which is the reason why the in-out asymmetry was unobvious by considering the drifts in Du et al. (J. Nucl. Mater. 463 (2015) 485).

  18. Liquid metal cooled divertor for ARIES

    International Nuclear Information System (INIS)

    Muraviev, E.

    1995-01-01

    A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m 2 , and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed

  19. Constrained ripple optimization of Tokamak bundle divertors

    International Nuclear Information System (INIS)

    Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

    1983-02-01

    Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

  20. Development of integrated SOL/Divertor code and simulation study of the JT-60U/JT-60SA tokamaks

    International Nuclear Information System (INIS)

    Kawashima, H.; Shimizu, K.; Takizuka, T.

    2007-01-01

    extend the operational region from detached to attached plasma, we optimize the divertor geometry of JT-60SA from the view point of the pumping efficiency [..(pumping flux)/(generated neutral flux at the divertor target)]. (author)

  1. Design integration of liquid surface divertors

    International Nuclear Information System (INIS)

    Nygren, R.E.; Cowgill, D.F.; Ulrickson, M.A.; Nelson, B.E.; Fogarty, P.J.; Rognlien, T.D.; Rensink, M.E.; Hassanein, A.; Smolentsev, S.S.; Kotschenreuther, M.

    2004-01-01

    The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied

  2. Design and test program of a simplified divertor dummy coil structure for the WEST project

    Energy Technology Data Exchange (ETDEWEB)

    Doceul, L., E-mail: louis.doceul@cea.fr [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Portafaix, C. [ITER Organization, Route de Vinon-sur-Verdon 13115, St. Paul-lez-Durance (France); Saille, A. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France); Salami, M. [AVANTIS Engineering Groupe, ZI de l’Aiguille 46100, Figeac (France); Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B. [CEA, IRFM, Saint-Paul-Lez-Durance Cedex F-13108 (France)

    2013-12-15

    Highlights: • The mechanical design and integration of the divertor structure has been performed. • The design of the casing and the winding-pack has been finalized. • The coil assembly process has been validated. • The realization of a coil mock-up scale one is in progress. -- Abstract: In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature. In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also

  3. Design and test program of a simplified divertor dummy coil structure for the WEST project

    International Nuclear Information System (INIS)

    Doceul, L.; Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A.; Portafaix, C.; Saille, A.; Salami, M.; Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B.

    2013-01-01

    Highlights: • The mechanical design and integration of the divertor structure has been performed. • The design of the casing and the winding-pack has been finalized. • The coil assembly process has been validated. • The realization of a coil mock-up scale one is in progress. -- Abstract: In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature. In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also

  4. Innovative design for FAST divertor compatible with remote handling, electromagnetic and mechanical analyses

    Energy Technology Data Exchange (ETDEWEB)

    Di Gironimo, Giuseppe, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Cacace, Maurizio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Crescenzi, Fabio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Labate, Carmelenzo [CREATE, University of Naples Parthenope, Via Acton 38, 80133 Napoli (Italy); Lanzotti, Antonio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Lucca, Flavio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Marzullo, Domenico; Mozzillo, Rocco [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Pagani, Irene [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Ramogida, Giuseppe; Roccella, Selanna [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Viganò, Fabio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy)

    2015-10-15

    Highlights: • The conceptual design of FAST divertor has been carried out through a continuous process of requirements refinement and design optimization (V-model approach), in order to achieve a design suited to the needs, RH compatible and ITER-like. • Thermal, structural and electromagnetic analyses have been performed, resulting in requirements refinement. • FAST divertor is now characterized by more realistic, reliable and functional features, satisfying thermo-mechanical capabilities and the remote handling (RH) compatibility. - Abstract: Divertor is a crucial component in Tokamaks, aiming to exhaust the heat power and particles fluxes coming from the plasma during discharges. This paper focuses on the optimization process of FAST divertor, aimed at achieving required thermo-mechanical capabilities and the remote handling (RH) compatibility. Divertor RH system final layout has been chosen between different concept solutions proposed and analyzed within the principles of Theory of Inventive Problem Solving (TRIZ). The design was aided by kinematic simulations performed using Digital Mock-Up capabilities of Catia software. Considerable electromagnetic (EM) analysis efforts and top-down CAD approach enabled the design of a final and consistent concept, starting from a very first dimensioning for EM loads. In the final version here presented, the divertor cassette supports a set of tungsten (W) actively cooled tiles which compose the inner and outer vertical targets, facing the plasma and exhausting the main part of heat flux. W-tiles are assembled together considering a minimum gap tolerance (0.1–0.5 mm) to be mandatorily respected. Cooling channels have been re-dimensioned to optimize the geometry and the layout of coolant volume inside the cassette has been modified as well to enhance the general efficiency.

  5. Fabrication and installation of the DIII-D radiative divertor structures

    International Nuclear Information System (INIS)

    Hollerbach, M.A.; Smith, J.P.

    1997-11-01

    Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 ell/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks

  6. Local island divertor experiments on LHD

    International Nuclear Information System (INIS)

    Morisaki, T.; Masuzaki, S.; Komori, A.; Ohyabu, N.; Kobayashi, M.; Feng, Y.; Sardei, F.; Narihara, K.; Tanaka, K.; Ida, K.; Peterson, B.J.; Yoshinuma, M.; Ashikawa, N.; Emoto, M.; Funaba, H.; Goto, M.; Ikeda, K.; Inagaki, S.; Kaneko, O.; Kawahata, K.; Kubo, S.; Miyazawa, J.; Morita, S.; Nagaoka, K.; Nagayama, Y.; Nakanishi, H.; Ohkubo, K.; Oka, Y.; Osakabe, M.; Shimozuma, T.; Shoji, M.; Takeiri, Y.; Sakakibara, S.; Sakamoto, R.; Sato, K.; Toi, K.; Tsumori, K.; Watababe, K.Y.; Yamada, H.; Yamada, I.; Yoshimura, Y.; Motojima, O.

    2005-01-01

    A local island divertor (LID) experiment has begun on LHD, with the aims of controlling edge recycling and improving the plasma confinement. The fundamental divertor functions of the LID have been demonstrated in the recent experiments. From the particle flux profile measurements on the LID head it was found that the particles diffusing out from the core region are well guided along the island separatrix to the LID head. Owing to the closed configuration around the LID head, evidence of the high efficient pumping was observed, together with a strong capacity to screen impurities. The first results of edge modeling using the EMC3-EIRENE code are also presented

  7. Control of divertor configuration in JT-60

    International Nuclear Information System (INIS)

    Yoshino, R.; Kukuchi, M.; Ninomiya, H.; Yoshida, H.; Tsuji, S.; Hosogane, N.; Seki, S.

    1985-01-01

    The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

  8. Safety characteristics of the monolithic CFC divertor

    International Nuclear Information System (INIS)

    Zucchetti, M.; Merola, M.; Matera, R.

    1994-01-01

    The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also. ((orig.))

  9. Safety characteristics of the monolithic CFC divertor

    Science.gov (United States)

    Zucchetti, M.; Merola, M.; Matera, R.

    1994-09-01

    The main distinguishing feature of the monolithic CFC divertor is the use of a single material, a carbon fibre reinforced carbon, for the protective armour, the heat sink and the cooling channels. This removes joint interface problems which are one of the most important concerns related to the reference solutions of the ITER CDA divertor. An activation analysis of the different coolant options for this concept is presented. It turns out that neither short-term nor long-term activation are a concern for any coolants investigated. Therefore the proposed concept proves to be attractive from a safety stand-point also.

  10. Preliminary analysis of the efficiency of non-standard divertor configurations in DEMO

    Directory of Open Access Journals (Sweden)

    F. Subba

    2017-08-01

    Full Text Available The standard Single Null (SN divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD and the Super X-Divertor (SX, and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW, we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3 and Zeff= 1.3.

  11. Effect of separatrix magnetic geometry on divertor behavior in DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Canik, J.M. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W.; Mahdavi, M.A. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Watkins, J.G. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Ferron, J.R.; Groebner, R.J. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Hill, D.N. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Hyatt, A.W. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Luce, T.C. [General Atomics, P.O. Box 85608, San Diego, CA 92186-5608 (United States); Moyer, R.A. [University of California–San Diego, La Jolla, CA 92093-0417 (United States); Stangeby, P.C. [University of Toronto Institute of Aerospace Studies, Toronto (Canada)

    2013-07-15

    We report on recent experiments on DIII-D that examined the effects that variations in the parallel connection length in the scrape-off layer (SOL), L{sub ||}, and the radial location of the outer divertor target, R{sub TAR}, have on divertor plasma properties. Two-point modeling of the SOL plasma predicts that larger values of L{sub ||} and R{sub TAR} should lower temperature and raise density at the outer divertor target for fixed upstream separatrix density and temperature, i.e., n{sub TAR} ∝ [R{sub TAR}]{sup 2}[L{sub ||}]{sup 6/7} and T{sub TAR} ∝ [R{sub TAR}]{sup −2}[L{sub ||}]{sup −4/7}. The dependence of n{sub TAR} and T{sub TAR} on L{sub ||} was consistent with our data, but the dependence of n{sub TAR} and T{sub TAR} on R{sub TAR} was not. The surprising result that the divertor plasma parameters did not depend on R{sub TAR} in the predicted way may be due to convected heat flux, driven by escaping neutrals, in the more open configuration of the larger R{sub TAR} cases. Modeling results using the SOLPS code support this postulate.

  12. Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

    Science.gov (United States)

    Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

    2017-12-01

    Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

  13. SOLPS-ITER Study of neutral leakage and drift effects on the alcator C-Mod divertor plasma

    Directory of Open Access Journals (Sweden)

    W. Dekeyser

    2017-08-01

    Full Text Available As part of an effort to validate the edge plasma model in the SOLPS-ITER code suite under ITER-relevant divertor plasma and neutral conditions, we report on progress in the modeling of the Alcator C-Mod divertor plasma with the new code. We perform simulations with a complete drifts model and kinetic neutrals, including effects of neutral viscosity, ion-molecule collisions and Lyα-opaque conditions, but assuming a pure deuterium plasma. Through a series of simulations with varying divertor geometries, we show the importance of including neutal leakage paths through the divertor substructure on the divertor plasma solution. Moreover, the impact of drifts on inner-outer target asymmetries is assessed. Including both effects, we achieve excellent agreement between simulations and upstream and outer target Langmuir Probe data. In absence of strong volumetric losses due to e.g. impurity radiation in our simulations, the strong inner target detachment observed experimentally remains elusive in our modeling at present.

  14. Conceptual design of CFETR divertor remote handling compatible structure

    International Nuclear Information System (INIS)

    Dai, Huaichu; Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei

    2016-01-01

    Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

  15. Conceptual design of CFETR divertor remote handling compatible structure

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Huaichu, E-mail: yaodm@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei (China); Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

  16. Self-sustained oscillations in the divertor plasma

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Kukushkin, A.S.; Pistunovich, V.I.; Pozharov, V.A.

    1985-01-01

    A simple analytical model of the edge plasma with high recycling, which relays on the presence of a small parameter - the ratio of the particle flows crossing the magnetic field to those impinging onto the divertor target, is proposed. A concept of the one-dimensional steady state (OSS) is introduced as the zero approximation in the small parameter. The mean number density N-tilde of the particles - ions plus neutrals - in the magnetic flux tube is choosen as the most representative and convenient parameter of the problem. The OSS are shown to be ambiguous in some N-tilde range for sufficiently high values of the energy flow entering the scrape-off layer from the bulk plasma. An equation, that describes a quasi-steady variations in OSS, is derived and a mechanism of exciting the self-sustained oscillations is developed. Results of simulation of the edge plasma oscillations are found to be in a good agreement with this mechanism, which could be responsible for the H-mode oscillations observed in the divertor experiments

  17. An X-point ergodic divertor

    International Nuclear Information System (INIS)

    Chu, M.S.; Jensen, T.H.; La Haye, R.J.; Taylor, T.S.; Evans, T.E.

    1991-10-01

    A new ergodic divertor is proposed. It utilizes a system of external (n = 3) coils arranged to generate overlapping magnetic islands in the edge region of a diverted tokamak and connect the randomized field lines to the external (cold) divertor plate. The novel feature in the configuration is the placement of the external coils close to the X-point. A realistic design of the external coil set is studied by using the field line tracing method for a low aspect ratio (A ≅ 3) tokamak. Two types of effects are observed. First, by placing the coils close to the X-point, where the poloidal magnetic field is weak and the rational surfaces are closely packed only a moderate amount of current in the external coils is needed to ergodize the edge region. This ergodized edge enhances the edge transport in the X-point region and leads to the potential of edge profile control and the avoidance of edge localized modes (ELMs). Furthermore, the trajectories of the field lines close to the X-point are modified by the external coil set, causing the hit points on the external divertor plates to be randomized and spread out in the major radius direction. A time-dependent modulation of the currents in the external (n = 3) coils can potentially spread the heat flux more uniformly on the divertor plate avoiding high concentration of the heat flux. 10 refs., 9 figs

  18. The ITER Divertor Cassette Project meeting

    International Nuclear Information System (INIS)

    Akiba, M.; Tivey, R.

    2000-01-01

    The Divertor Cassette Project topical meeting took place on April 5-7, 2000 at the JAERI Naka site in Japan. The meeting focused on the progress made by the three parties under task agreements on the development of carbon-fibre composite and tungsten armored high flux plasma-facing components

  19. Compact poloidal divertor reference design for TNS

    International Nuclear Information System (INIS)

    Yang, T.F.; Lee, A.Y.; Ruck, G.W.; Lange, W.J.

    1977-01-01

    A compact poloidal divertor concept has been developed for TNS tokamaks and its feasibility has been demonstrated by sufficient detailed magnetic, thermal, mechanical and vacuum analyses. This particular divertor is formed by a pair of opposing coil sets which define a magnetic flux slot where the particle burial chamber is located. The magnetic flux in the space between the coil sets is compressed vertically to limit the height and to expand the horizontal width of the particle and energy burial chamber. The intensity of the poloidal field is increased to make the pitch angle of the flux lines very large so that the diverted particles can be intercepted by a large number of panels oriented at a small angle with respect to the flux lines. Large collecting surface areas can be obtained so that the thermal load and particle flux are reduced to a practical level. Flowing lithium film and solid metal panels have been considered as the particle collector and the latter is preferred. This divertor allows for most economical use of the available space inside the TF coils and thus has minor impact on the overall size of the tokamak. The divertor design is essentially independent of the tokamak system, although analyses were performed based on TNS

  20. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.N. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Pestchanyi, S.E. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2007-06-15

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  1. A study of X-divertor in NSTX-U with SOLPS simulations

    Science.gov (United States)

    Chen, Zhong-Ping; Kotschenreuther, Mike; Mahajan, Swadesh; Gerhardt, Stefan

    2018-03-01

    The X-divertor (XD) geometry in NSTX-U is demonstrated, via SOLPS simulations, to perform better than the standard divertor (SD); in particular, it allows detachment at a lower upstream density and stabilizes the detachment front near the target, away from the main X-point. Consequently a stable detached operation becomes possible—the localization near the plate allows a vast reduction of heat fluxes without degrading the core plasma. Indeed, it is confirmed by our simulation that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures, resulting in scrape-off layers that are more favorable for advanced tokamak operation. These advantages are attributed to the unique geometric characteristics of XD—poloidal flaring near the target.

  2. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Science.gov (United States)

    Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

    2007-06-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  3. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    International Nuclear Information System (INIS)

    Bazylev, B.N.; Janeschitz, G.; Landman, I.S.; Loarte, A.; Pestchanyi, S.E.

    2007-01-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated

  4. An analytical erosion model for divertor plates and limiter experiments in CHS

    International Nuclear Information System (INIS)

    Sagara, A.; Noda, N.; Akiyama, R.; Arimoto, H.; Idei, H.; Iguchi, H.; Kaneko, O.; Kohmoto, T.; Kubo, S.; Matsuoka, K.; Morita, S.; Motojima, O.; Nishimura, K.; Okamura, S.; Takahasi, C.; Takita, Y.; Yamada, I.; Matsunami, N.; Rice, J.; Yamada, H.; Shoji, T.; Ueda, M.

    1992-01-01

    A self-consistent analytical solution for net erosion of a divertor plate which is set perpendicular to magnetic field lines is presented. The primary flux profile of hydrogen and impurities except redepositing particles is externally given as well as the return ratio of sputtered atoms to the plate. In the direction along the divertor trace, all conditions are uniform. The ionization mean free path is assumed constant to simplify equations. The analytical solution is compared with net erosion experiments carried out in compact helical system (CHS) by exposing a graphite target to a neutral beam heated plasma column introduced perpendicularly to the target along the magnetic field lines through a 2 cm slit opend on a graphite limiter. After exposure to 98 discharges, the target surface is analyzed with Rutherford backscattering method. Deposition profiles of Ti and O impurities are very well explained with the analytical predictions. (orig.)

  5. Numerical study of the connection lengths for various magnetic configurations in Wendelstein 7-X to optimize the heat load on the divertor

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Priyanjana; Hoelbe, Hauke; Sunn Pedersen, Thomas [Max Planck Institute of Plasma Physics, Greifswald (Germany)

    2016-07-01

    Fusion has the potential to play an important role as a future energy resource. It has the capacity to produce large-scale clean energy. The two main confinement concepts are the tokamak and the stellarator. The W7-X machine is based on stellarator principle and is using special form of coils to achieve steady-state plasma confinement. Divertors are used in tokamaks and stellarator to control the exhaust of waste gases and impurities from the machine. The divertor concept of W7-X is a so-called island divertor. The island chain isolates the confinement core from regions where the plasma-wall interaction takes place. The area of the divertor that receives the main part of the heat loads, the so-called wetted area, increases with the distance along the magnetic field from the outboard midplane to the divertor target. The connection length is relatively short in tokamaks with conventional divertors. In the stellarator island divertor, the connection length can be varied significantly, which should allow for optimization of the wetted area. We present here a numerical study of the achievable connection lengths in various W7-X configurations and discuss the possibilities for running dedicated experiments to understand the physics of what sets the wetted area.

  6. Divertors for Helical Devices: Concepts, Plans, Results, and Problems

    International Nuclear Information System (INIS)

    Koenig, R.; Grigull, P.; McCormick, K.

    2004-01-01

    With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the subdivertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields

  7. Divertors for helical devices: Concepts, plans, results and problems

    International Nuclear Information System (INIS)

    Koenig, R.; Grigull, P.; McCormick, K.

    2003-01-01

    With LHD and W7-X stellarator development is now taking a large leap forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control, and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large stellarators were carefully prepared in smaller scale devices like Heliotron E, CHS and W7-AS. While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller scale experiments like Heliotron-J, CHS and NCSX will be used for the further development of divertor concepts. The two divertor configurations that are presently being investigated, are the helical and the island divertor, as well as the local island divertor (LID), which was successfully demonstrated on CHS and just went into operation on LHD. Presently, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor which will allow quasi continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi steady-state operating scenario in a newly found high density H-mode operating regime, which benefits from high energy and extremely low impurity confinement times, with edge radiation levels of up to 90 % and sufficient neutral compression in the subdivertor region (> 10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios and toroidal asymmetries due to symmetry breaking error fields, etc. will be discussed. (orig.)

  8. Physics conclusions in support of ITER W divertor monoblock shaping

    Directory of Open Access Journals (Sweden)

    R.A. Pitts

    2017-08-01

    Full Text Available 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.

  9. 3D modelling of the island divertor for W7-AS

    International Nuclear Information System (INIS)

    Sardei, F.; Feng, Y.; Kisslinger, J.; Grigull, P.

    1996-01-01

    Island divertors in low-shear stellarators exhibit the same basic topology (X-point diversion of field lines towards target plates) as tokamak divertors. However, the geometry is different. For island divertors, the small distance between the target and the LCFS (∼5cm for W7-AS and 8cm for W7-X) requires higher plasma densities than in comparable tokamaks to effectively decouple the target plasma and the neutrals from the core. These are basic prerequisites to realize high recycling and detachment conditions necessary for exaust. On the other hand, the island SOL can be used to confine recycling particles outside the LCFS, which may result in a density rise inside the islands, and hence in an improved screening of the neutrals. Nonlinear 3D effects are introduced in the transport equations by the non-axisymmetry of the configuration and by the segmentation of the target plates. The resulting toroidal inhomogeneities (variable connection lengths, toroidally localized recycling, poor parallel equilibration at low T) can hardly be approximated by an averaging 2D model. (orig.)

  10. Possible divertor solutions for a fusion reactor. Pt. I. Physical aspects based on present day divertor operation

    International Nuclear Information System (INIS)

    Kallenbach, A.; Bosch, H.-S.; De Pena Hempel, S.; Dux, R.; Kaufmann, M.; Mertens, V.; Neuhauser, J.; Suttrop, W.; Zohm, H.

    1997-01-01

    For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m -2 , far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Z eff . Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δP rad /δZ eff ∝n e 2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)

  11. Development of a radiative divertor for DIII-D

    International Nuclear Information System (INIS)

    Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

    1995-01-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

  12. ITER tungsten divertor design development and qualification program

    Energy Technology Data Exchange (ETDEWEB)

    Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Escourbiac, F.; Carpentier-Chouchana, S.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Kukushkin, A.; Merola, M.; Mitteau, R.; Pitts, R.A.; Shu, W.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Riccardi, B. [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Suzuki, S. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati, Rome (Italy)

    2013-10-15

    Highlights: • 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 and D activity. -- Abstract: 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 and D activity is summarized in this paper.

  13. Divertor, thermonuclear device and method of neutralizing high temperature plasma

    International Nuclear Information System (INIS)

    Ikegami, Hideo.

    1995-01-01

    The thermonuclear device comprises a thermonuclear reactor for taking place fusion reactions to emit fusion plasmas, and a divertor made of a hydrogen occluding material, and the divertor is disposed at a position being in contact with the fusion plasmas after nuclear fusion reaction. The divertor is heated by fusion plasmas after nuclear fusion reaction, and hydrogen is released from the hydrogen occluding material as a constituent material. A gas blanket is formed by the released hydrogen to cool and neutralize the supplied high temperature nuclear fusion plasmas. This prevents the high temperature plasmas from hitting against the divertor, elimination of the divertor by melting and evaporation, and solve a problem of processing a divertor activated by neutrons. In addition, it is possible to utilize hydrogen isotopes of fuels effectively and remove unnecessary helium. Inflow of impurities from out of the system can also be prevented. (N.H.)

  14. Fluid simulation of beryllium transport in the ITER gaseous divertor

    International Nuclear Information System (INIS)

    Knoll, D.A.; Campbell, R.B.; McHugh, P.R.

    1994-01-01

    The transport of either intrinsic or injected impurities will play a crucial role in the energy loss mechanisms in the ITER gaseous/cold plasma target divertor. Both 1-D and 2-D multi-charge state fluid codes are used to model the transport of beryllium in the ITER SOL. Our major conclusion is that in order to model the containment of impurities, the background flow field must be known in detail. Comparing 1-D and 2-D solutions, hydrogen flow reversal plays an important role in the entrainment process. Further, the flow of particles from the core plasma also has a strong impact on the resultant entrainment of the impurities in both 1-D and 2-D. It is imperative that those components of poloidal velocity due to E x B and diamagnetic drifts be included in the models. (orig.)

  15. Is Carbon a Realistic Choice for ITER's Divertor?

    International Nuclear Information System (INIS)

    Skinner, C.H.; Federici, G.

    2005-01-01

    Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives

  16. Towards fully authentic modelling of ITER divertor plasmas

    International Nuclear Information System (INIS)

    Maddison, G.P.; Hotston, E.S.; Reiter, D.; Boerner, P.

    1991-01-01

    Ignited next step tokamaks such as NET or ITER are expected to use a poloidal magnetic divertor to facilitate exhaust of plasma particles and energy. We report a development coupling together detailed computational models for both plasma and recycled neutral particle transport processes, to produce highly detailed and consistent design solutions. A particular aspect is involvement of an accurate specification of edge magnetic geometries, determined by an original equilibrium discretisation code, named LINDA. Initial results for a prototypical 22MA ITER double-null configuration are presented. Uncertainties in such modelling are considered, especially with regard to intrinsic physical scale lengths. Similar results produced with a simple, analytical treatment of recycling are also compared. Finally, a further extension allowing true oblique target sections is anticipated. (author) 8 refs., 5 figs

  17. Design and tests of a simplified divertor dummy coil structure for the WEST project

    International Nuclear Information System (INIS)

    Doceul, L.; Bucalossi, J.; Dougnac, H.; Ferlay, F.; Gargiulo, L.; Keller, D.; Larroque, S.; Lipa, M.; Pilia, A.; Saille, A.; Samaille, F.; Soler, B.; Thouvenin, D.; Verger, J.M.; Zago, B.; Portafaix, C.; Salami, M.

    2015-01-01

    Full text of publication follows. In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target. To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 200 Celsius degrees, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s). The divertor structure will be a complex assembly of 4 meter diameter and 4 meter height representing a total weight of around 20 tonnes. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, mechanical loads (induced by disruptions) and electrical isolation (13 kV test) under high temperature. In order to fully validate the feasibility, the mounting and the performance of this complex component, the production of a scale one dummy coil is in progress. The paper will illustrate, the technical developments performed during 2012 in order to finalise the design for the call for tender phase. The progress and the first results of the simplified dummy coils will be also addressed. (authors)

  18. SLAC divertor channel entrance thermal stress analysis

    International Nuclear Information System (INIS)

    Johnson, G.L.; Stein, W.; Lu, S.C.; Riddle, R.A.

    1985-01-01

    X-ray beams emerging from the new SLAC electron-positron storage ring (PEP) impinge on the entrance to tangential divertor channels causing highly localized heating in the channel structure. Analyses were completed to determine the temperatures and thermally-induced stresses due to this heating. These parts are cooled with water flowing axially over them at 30 0 C. The current design and operating conditions should result in the entrance to the new divertor channel operating at a peak temperature of 123 0 C with a peak thermal stress at 91% of yield. Any micro-cracks that form due to thermally-induced stresses should not propagate to the coolant wall nor form a path for the coolant to leak into the storage ring vacuum. 34 figs., 4 tabs

  19. NSTX Plasma Response to Lithium Coated Divertor

    Energy Technology Data Exchange (ETDEWEB)

    H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

    2011-01-21

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  20. He-cooled divertor development for DEMO

    International Nuclear Information System (INIS)

    Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

    2007-01-01

    Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

  1. NSTX plasma response to lithium coated divertor

    International Nuclear Information System (INIS)

    Kugel, H.W.; Bell, M.G.; Allain, J.P.; Bell, R.E.; Ding, S.; Gerhardt, S.P.; Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.M.; LeBlanc, B.P.; Maingi, Rajesh; Majeski, R.; Maqueda, R.J.; Mansfield, D.K.; Mueller, D.; Nygren, R.E.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.A.; Taylor, C.N.; Timberlake, J.; Wampler, W.R.; Zakharov, L.E.; Zweben, S.J.

    2011-01-01

    NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Z(eff) and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, < 0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

  2. Plasma diagnostics for the DIII-D divertor upgrade (abstract)

    International Nuclear Information System (INIS)

    Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

    1990-01-01

    The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

  3. Comparative divertor-transport study for helical devices

    International Nuclear Information System (INIS)

    Feng, Y.; Sardei, F.; Kobayashi, M.

    2008-10-01

    Using the island divertors (ID) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine-size following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. Revealed is the fundamental role of the low-order magnetic islands in both divertor concepts. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime which is absent from W7-AS and LHD is expected for W7-X. Topics addressed are restricted to the basic function elements of a divertor such as particle flux enhancement and impurity retention. In particular, the divertor function on reducing the influx of intrinsic impurities is examined for all the three devices under different divertor plasma conditions. Special attention is paid to examining the island screening potential of intrinsic impurities which has been predicted for all the three devices under high divertor collisionality conditions. The results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. (author)

  4. Experimental studies of the snowflake divertor in TCV

    Directory of Open Access Journals (Sweden)

    B. Labit

    2017-08-01

    Full Text Available To address the risk that, in a fusion reactor, the conventional single-null divertor (SND configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD, are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized.

  5. Time and space-resolved energy flux measurements in the divertor of the ASDEX tokamak by computerized infrared thermography

    International Nuclear Information System (INIS)

    Mueller, E.R.; Steinmetz, K.; Bein, B.K.

    1984-06-01

    A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 10 13 cm -3 ) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

  6. The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

    2017-08-01

    Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

  7. Divertor experiment for impurity control in DIVA

    International Nuclear Information System (INIS)

    Nagami, Masayuki

    1979-04-01

    Divertor actions of controlling the impurities and the transport of impurity ions in the plasma have been investigated in the DIVA device. Following are the results: (1) The radial transport of impurity ions is not described only by neoclassical theory, but it is strongly influenced by anomalous process. Radial diffusion of impurity ions across the whole minor radius is well described by a neoclassical diffusion superposed by the anomalous diffusion for protons. Due to this anomalous process, which spreads the radial density profile of impurity ions, 80 to 90% of the impurity flux in the plasma outer edge is shielded even in a nondiverted discharge. (2) The divertor reduces the impurity flux entering the main plasma by a factor of 2 to 4. The impurity ions shielded by the scrape-off plasma are rapidly guided into the burial chamber with a poloidal excursion time roughly equal to that of the scrape-off plasma. (3) The divertor reduces the impurity ion flux onto the main vacuum chamber by guiding the impurity ions diffusing from the main plasma into the burial chamber, thereby reducing the plasma-wall interaction caused by diffusing impurity ions at the main vacuum chamber. The impurity ions produced in the burial chamber may flow back to the main plasma through the scrape-off layer. However, roughly only 0.3% of the impurity flux into the scrape-off plasma in the burial chamber penetrates into the main plasma due to the impurity backflow. (4) A slight cooling of the scrape-off plasma with light-impurity injection effectively reduces the metal impurity production at the first wall by reducing the potential difference between the plasma and the wall, thereby reducing the accumulation of the metal impurity in the discharge. Radiation cooling by low-Z impurities in the plasma outer edge, which may become an important feature in future large tokamaks both with and without divertor, is numerically evaluated for carbon, oxygen and neon. (author)

  8. Two-point model for divertor transport

    International Nuclear Information System (INIS)

    Galambos, J.D.; Peng, Y.K.M.

    1984-04-01

    Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime

  9. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    McGuire, K.; Beiersdorfer, P.; Bell, M.

    1985-01-01

    Routine operation in the enhanced energy confinement (or H-mode) regime during neutral beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral beam heated discharges with this limiter show similar confinement times (normalized to tau/sub E//I/sub p/) to average H-mode plasmas. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasicoherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω less than or equal to 0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERP's are characterized by sharp spikes in the divertor plasma density, H/sub α/ emission, and on the x-ray signals they appear as sawtoothlike relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high β/sub T/ in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable β/sub T/. A study of the stability of both the limiter L-mode and divertor H-mode discharges close to the theoretical β boundary, showed that the major disruptions observed there are sometimes caused by a fast growing m/n = 1/1 mode with no observable external precursor oscillations

  10. Divertor and scoop limiter experiments on PDX

    International Nuclear Information System (INIS)

    McGuire, K.; Beirsdorfer, P.; Bell, M.

    1985-01-01

    Routine operation in the enhanced-energy-confinement (or H-mode) regime during neutral-beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral-beam-heated discharges with this limiter show similar confinement times (normalized to tausub(E)/Isub(p)) to average H-mode plasma. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasi-coherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω<=0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERPs are characterized by sharp spikes in the divertor plasma density, Hsub(α) emission, and on the X-ray signals they appear as sawtooth-like relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high βsub(T) in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable βsub(T). A study of the stability of both the limiter L-mode and divertor H-mode discharge close to the theoretical β boundary showed that the major disruptions observed there are sometimes caused by a fast growing m/n=1/1 mode with no observable external precursor oscillations. (author)

  11. The remote exchange of the JET divertor

    International Nuclear Information System (INIS)

    Pick, M.

    1999-01-01

    In 1997 a series of experiments were performed in the JET machine using deuterium-tritium (D-T) mixtures and resulting in discharges with record breaking fusion power and fusion energy. The experiments demonstrated a key technology required for fusion, namely the on-line operation of a tritium fuel re-processing plant. These experiments left the inside of the JET vessel inaccessible to manned access for approximately one year. During this time, the complete Mark IIA divertor, a major system within the torus, was successfully removed and replaced with a new divertor design, the Mark II Gas Box divertor, using only remote handling techniques. This was the first application of the JET remote handling system and a demonstration of a further key ITER technology. The paper explains the methodology and operational approach taken to achieve the results using the remote handling system developed at JET. It describes the remote handling equipment including the force-reflecting servo-manipulator, the specialised tools designed, the facilities needed, and the trials, planning and training carried out to ensure the safe, reliable and rapid completion of the remote handling tasks. The planned tasks are outlined including the execution of the novel procedure for a remote, sub-millimetre precision, dimensional survey of the divertor support structure using digital photogrammetry. Furthermore the paper shows how the adaptability of the system was used to successfully undertake a large number of unplanned tasks including the removal of damaged tiles, a damaged diagnostic system and the vacuum cleaning of diagnostic windows. (author)

  12. Spectroscopic investigation of ELM phenomena in the ASDEX-Upgrade divertor with high time resolution

    International Nuclear Information System (INIS)

    Field, A.R.; Buechl, K.; Fuchs, C.J.; Fussmann, G.; Herrmann, A.; Lieder, G.; Napiontek, B.; Radtke, R.; Wenzel, U.; Zohm, H.

    1993-01-01

    Improved tokamak H-mode confinement is associated with the formation of an insulating zone just within the separatrix. At a critical pressure gradient a sudden burst of MHD activity (an ELM) degrades edge confinement, releasing particles and energy into the scrape-off layer (SOL) which is subsequently transported to the divertor. Here, these phenomena are studied using spectroscopic diagnostics and target plate thermography of high spatial and temporal resolution. (author) 3 refs., 6 figs

  13. Spectroscopic investigation of ELM phenomena in the ASDEX-Upgrade divertor with high time resolution

    Energy Technology Data Exchange (ETDEWEB)

    Field, A R; Buechl, K; Fuchs, C J; Fussmann, G; Herrmann, A; Lieder, G; Napiontek, B; Radtke, R; Wenzel, U; Zohm, H [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    1994-12-31

    Improved tokamak H-mode confinement is associated with the formation of an insulating zone just within the separatrix. At a critical pressure gradient a sudden burst of MHD activity (an ELM) degrades edge confinement, releasing particles and energy into the scrape-off layer (SOL) which is subsequently transported to the divertor. Here, these phenomena are studied using spectroscopic diagnostics and target plate thermography of high spatial and temporal resolution. (author) 3 refs., 6 figs.

  14. Erosion of ITER divertor armour and contamination of sol after transient events erosion products

    International Nuclear Information System (INIS)

    Bazylev, B.N.; Landman, I.S.; Pestchanyi, S.E.

    2005-01-01

    Plasma impact to the divertor expected in the tokamak ITER during ELMs or disruptions can result in a significant surface damage to CFC- and tungsten armours (brittle destruction and melting respectively) as well as in contamination of SOL by evaporated impurities. Numerical investigations for tungsten and CFC targets provide important details of the material erosion process. The simulations carried out in FZK on the material damage, carbon plasma expansion and the radiation fluxes from the carbon impurity are surveyed

  15. Low energy neutral particle fluxes in the JET divertor

    International Nuclear Information System (INIS)

    Reichle, R.; Horton, L.D.; Ingesson, L.C.; Jaeckel, H.J.; McCormick, G.K.; Loarte, A.; Simonini, R.; Stamp, M.F.

    1997-01-01

    First measurements are presented of the total power loss through neutral particles and their average energy in the JET divertor. The method used distinguishes between the heat flux and the electromagnetic radiation on bolometers. This is done by comparing measurements from inside the divertor either with opposite lines of sight or with a tomographic reconstruction of the radiation. The typical value of the total power loss in the divertor through neutrals is about 1 MW. The average energy of the neutral particles at the inner divertor leg is 1.5-3 eV when detachment is in progress, which agrees with EDGE2D/NIMBUS modelling. (orig.)

  16. Divertor heat flux mitigation in the National Spherical Torus Experimenta)

    Science.gov (United States)

    Soukhanovskii, V. A.; Maingi, R.; Gates, D. A.; Menard, J. E.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Bell, M. G.; Bell, R. E.; Boedo, J. A.; Bush, C. E.; Kaita, R.; Kugel, H. W.; Leblanc, B. P.; Mueller, D.; NSTX Team

    2009-02-01

    Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

  17. Towards a physics-integrated view on divertor pumping

    International Nuclear Information System (INIS)

    Day, Chr.; Gleason-González, C.; Hauer, V.; Igitkhanov, Y.; Kalupin, D.; Varoutis, S.

    2014-01-01

    Highlights: • Physics-integrated design approaches are to be preferred over approaches based on simple requirement lists. • A physics-integrated assessment is presented for the divertor vacuum pumping system based on detachment onset conditions for the divertor. • This approach considers density dependent pump albedo to reflect the effects of gas recycling at the divertor and the changes in flow regime with density. • A comparison with DEMO indicates that the divertor pumping system for a pulsed DEMO scales less than linearly with fusion power. - Abstract: One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset

  18. FLP: a field line plotting code for bundle divertor design

    International Nuclear Information System (INIS)

    Ruchti, C.

    1981-01-01

    A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

  19. Influence of stray light for divertor spectroscopy in ITER

    International Nuclear Information System (INIS)

    Kajita, Shin; Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael; Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi

    2015-01-01

    The influence of stray light in the divertor spectroscopy system in ITER is quantitatively investigated using a ray tracing simulation. Simulation results show that the stray light is negligible at positions in the divertor where the plasma emission is strong. However, it is also shown that the stray light can be significantly greater than the real signal if the plasma intensity is low. Deuterium and beryllium emissions are used for the assessment; for beryllium cases in particular, since the emission profile may be non-uniform in the divertor region, the influence of stray light can be non-negligible at some positions, e.g., above the divertor dome

  20. Technological development of the Monobloc Divertor Concept

    International Nuclear Information System (INIS)

    DiPietro, E.; Brossa, M.; Guerreschi, U.; Suresh, D.; Cardella, A.

    1992-01-01

    This paper reports on a technological program devoted to the assessment of the feasibility and the qualification of the Monobloc Divertor Concept for the divertor of the NET/ITER Machine which has been developed with the joint collaboration between ENEA, the NET Team, Ansaldo DNT and Metallwerk Plansee. The basic idea guiding the development of the monobloc divertor consists in obtaining a component suitable to sustain the operation thermal loads, attaining peak values in the range of 15 MW/2 in steady state conditions, by a proper arrangement of refractory tiles (acting as an armour) directly brazed to the cooling pipes. In the first phase the main activities have been devoted to find a reliable joint between the armour and the cooling pipes. A number of candidate armour materials have been investigated chosen among the most promising CFC currently available in combination with molybdenum alloys (T2M and Mo41Re) and dispersion strengthened copper. The most relevant results of the test activity including the comparison of different brazing alloys and techniques and the evaluation of suitable NDE techniques are reported

  1. Westinghouse compact poloidal divertor reference design

    International Nuclear Information System (INIS)

    Yang, T.F.; Lee, A.Y.; Ruck, G.W.

    1977-08-01

    A feasible compact poloidal divertor system has been designed as an impurity control and vacuum vessel first-wall protection option for the TNS tokamak. The divertor coils are inside the TF coil array and vacuum vessel. The poloidal divertor is formed by a pair of coil sets with zero net current. Each set consists of a number of coils forming a dish-shaped washer-like ring. The magnetic flux in the space between the coil sets is compressed vertically to limit the height and to expand the horizontal width of the particle and energy burial chamber which is located in the gap between the coil sets. The intensity of the poloidal field is increased to make the pitch angle of the flux lines very large so that the diverted particles can be intercepted by a large number of panels oriented at a small angle with respect to the flux lines. They are carefully shaped and designed such that the entire surfaces are exposed to the incident particles and are not shadowed by each other. Large collecting surface areas can be obtained. Flowing liquid lithium film and solid metal panels have been considered as the particle collectors. The power density for the former is designed at 1 MW/m 2 and for the latter 0.5 MW/m 2 . The major mechanical, thermal, and vacuum problems have been evaluated in sufficient detail so that the advantages and difficulties are identified. A complete functional picture is presented

  2. THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

    International Nuclear Information System (INIS)

    C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

    2000-01-01

    The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

  3. COMPARISON OF ELM PULSE PROPAGATION IN THE DIII-D SOL AND DIVERTORS WITH AN ION CONVECTION MODEL

    International Nuclear Information System (INIS)

    FENSTERMACHER, ME; PORTER, GD; LEONARD, AW; BROOKS, NH; BOEDO, JA; COLCHIN, RJ; GRAY, DS; GROEBNER, RJ; GROTH, M; HOGAN, JT; HOLLMANN, EM; LASNIER, CJ; OSBORNE, TH; PETRIE, TW; RUDAKOV, DL; SNYDER, PB; TAKAHASHI, H; WATKINS, JG; ZENG, L; DIII-D TEAM

    2003-01-01

    OAK-B135 Results from dedicated ELM experiments, performed in DIII-D with fast diagnostics to measure the evolution of Type-I ELM effects in the SOL and divertor, are compared with a simple ion convection model and with initial time-dependent UEDGE simulations. Delays between ELM effects observed in the inner versus the outer divertor regions in the experiments scale, as a function of density, with the difference in ion convection time along field lines from the outer midplane to the divertor targets. The ELM perturbation was modeled as an instantaneous radially uniform increase of diffusion coefficients from the top of the pedestal to the outer SOL. The perturbation was confined to a low field side poloidal zone ± 40 o from the outer midplane. The delays in the simulations are similar to those observed in the experiments

  4. Heat flux management via advanced magnetic divertor configurations and divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Kolemen, E., E-mail: ekolemen@princeton.edu [Princeton University, Princeton, NJ 08544 (United States); Allen, S.L. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bray, B.D. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Humphreys, D.A.; Hyatt, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Maingi, R.; Nazikian, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Petrie, T.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Unterberg, E.A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

    2015-08-15

    The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

  5. Divertor plasma studies on DIII-D: Experiment and modeling

    International Nuclear Information System (INIS)

    West, W.P.; Brooks, N.H.; Allen, S.L.

    1996-09-01

    In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

  6. Advantages and Challenges of Radiative Liquid Lithium Divertor

    Science.gov (United States)

    Ono, Masayuki

    2017-10-01

    Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

  7. Engineering design of the Aries-IV gaseous divertor

    International Nuclear Information System (INIS)

    Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

    1994-01-01

    ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

  8. Divertor plasma physics experiments on the DIII-D tokamak

    International Nuclear Information System (INIS)

    Mahdavi, M.A.; Allen, S.L.; Evans, T.E.

    1996-10-01

    In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model

  9. Divertor impurity monitor for the International Thermonuclear Experimental Reactor

    Science.gov (United States)

    Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

    1999-01-01

    The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

  10. Effects of neutral gas collisions on the power transmission factor at the divertor sheath

    International Nuclear Information System (INIS)

    Futch, A.H.; Matthews, G.F.; Buchenauer, D.; Hill, D.N.; Jong, R.A.; Porter, G.D.

    1992-01-01

    We show that charge-exchange and other ion-neutral collision can reduce the power transmission factor of the plasma sheath, thereby lowering the ion impact energy and target plate sputtering. The power transmission factor relates the heat flux reaching the divertor target to the plasma density and temperature just in front of the surface: δ=Q surf /J ew k T e . Experimental data from the DIII-D tokamak suggests that δ could be as low as 2-3 near the region of peak divertor particle flux, instead of the 7-8 expected from usual sheath theory. Several effects combine to allow ion-neutral interactions to be important in the divertor plasma sheath. The shallow angle of incidence of the magnetic field (1-3deg in DIII-D) leads to the spatial extension of the sheath from approximately ρ i ∝1 mm normal to the plate to several centimeters along the field lines. Ionization reduces the sheath potential, and collisions reduce the ion impact energy. (orig.)

  11. Simulation of neutral gas flow in a tokamak divertor using the Direct Simulation Monte Carlo method

    International Nuclear Information System (INIS)

    Gleason-González, Cristian; Varoutis, Stylianos; Hauer, Volker; Day, Christian

    2014-01-01

    Highlights: • Subdivertor gas flows calculations in tokamaks by coupling the B2-EIRENE and DSMC method. • The results include pressure, temperature, bulk velocity and particle fluxes in the subdivertor. • Gas recirculation effect towards the plasma chamber through the vertical targets is found. • Comparison between DSMC and the ITERVAC code reveals a very good agreement. - Abstract: This paper presents a new innovative scientific and engineering approach for describing sub-divertor gas flows of fusion devices by coupling the B2-EIRENE (SOLPS) code and the Direct Simulation Monte Carlo (DSMC) method. The present study exemplifies this with a computational investigation of neutral gas flow in the ITER's sub-divertor region. The numerical results include the flow fields and contours of the overall quantities of practical interest such as the pressure, the temperature and the bulk velocity assuming helium as model gas. Moreover, the study unravels the gas recirculation effect located behind the vertical targets, viz. neutral particles flowing towards the plasma chamber. Comparison between calculations performed by the DSMC method and the ITERVAC code reveals a very good agreement along the main sub-divertor ducts

  12. Structural analysis of the ITER Divertor toroidal rails

    Energy Technology Data Exchange (ETDEWEB)

    Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

    2013-10-15

    The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

  13. Simulation of the ASDEX divertor performance after hardening

    International Nuclear Information System (INIS)

    Schneider, W.; Lackner, K.; Neuhauser, J.; Wunderlich, R.

    1985-05-01

    Two combined computer models - a fluid description of the plasma scrape-off layer (SOLID) and a Monte-Carlo code for the neutral gas dynamics (DEGAS) - are used to assess changes in the divertor performance expected from the modifications in geometry needed for hardening the ASDEX divertor chamber for long-pulse, high-power heating. Stand-alone DEGAS calculations with assumed fixed scrape-off plasma parameters predict a doubling of the neutral escape probability, which, however, still remains so low, that achievement of the high divertor recycling regime can be expected over roughly the same operational regime as before modifications. This conclusion is also supported by fully self-consistent calculations with the combined model. Due to the reduced divertor, a significant reduction is predicted in the divertor time constant, which is expected to affect transient phenomena. (orig.)

  14. Divertor heat flux control and plasma-material interaction

    International Nuclear Information System (INIS)

    Kikuchi, Yusuke; Nagata, Masayoshi; Sawada, Keiji; Takamura, Shuichi; Ueda, Yoshio

    2014-01-01

    Development of reliable radiative-cooling divertors is essential in DEMO reactor because it uses low-activation materials with low heat removal and the plasma heat flux exhausted from the confined region is 5 times as large as in ITER. It is important to predict precisely the heat and particle flux toward the divertor plate by simulation. In this present article, theoretical and experimental data of the reflection, secondary emission and surface recombination coefficients of the divertor plate by ion bombardment are given and their effects on the power transmission coefficient are discussed. In addition, some topics such as the erosion process of the divertor plate by ELM and the plasma disruption, the thermal shielding due to the vapor layer on the divertor plate and the formation of fuzz structure on W by helium plasma irradiation, are described. (author)

  15. Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

    Energy Technology Data Exchange (ETDEWEB)

    Laugier, F

    2000-10-19

    We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

  16. Measurements of poloidal and toroidal energy deposition asymmetries in the ASDEX divertors

    International Nuclear Information System (INIS)

    Evans, T.E.

    1991-03-01

    Energy deposition characteristics in the ASDEX divertors have been analyzed over a wide range of discharges and wall conditions during ohmically heated, additionally heated, or lower hybrid current drive experiments. Changes in discharge operating parameters with high power additional heating produce a diversity of effects on the magnitudes and distributions of the energy absorbed in the divertors. Poloidally and toroidally resolved energy deposition patterns are particularly sensitive to changes in the edge safety factor, the type and power level of additional heating used, and the vertical position of the plasma. In most additionally heated discharges, a large fraction of the incremental divertor loading is found on only one or two target rings. Poloidal in-out asymmetries, which typically favor the low-field side by a factor of 2.5 in ohmic discharges, commonly range between a factor of 2.5 and 4.5 in additionally heated experiments and in extreme cases can be as large as a factor of 5.6. At the same time, toroidal asymmetries on individual target rings are found to range between a factor of 1.4 and 3.8 in typical ICRH and NBI cases with extreme LHCD cases of 4.3. A model, proposed to explain the cause of discharge asymmetries, is compared with the experimental observations. Under some conditions, for example during LHCD experiments, the model is in good agreement with the data. A method is proposed for supressing discharge asymmetries which may generally improve the divertor performance as well. (orig./AH)

  17. Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

    Science.gov (United States)

    Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

    2017-10-01

    In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

  18. Design analysis of the ITER divertor

    International Nuclear Information System (INIS)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

    2007-01-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  19. Design analysis of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

    2007-07-01

    The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

  20. On the refuelling of large divertor experiments

    International Nuclear Information System (INIS)

    Staebler, A.; Haas, G.; Ott, W.; Speth, E.

    1976-01-01

    The use of fast hydrogen atoms, molecules and clusters for refuelling large divertor-experiments like ASDEX is investigated. Three criteria for the choice among the various methods are discussed. It is shown that clusters suffer from lack of penetration. Molecules, created by fragmentation of clusters, offer the advantage of plasma-like energy combined with appreciable penetration. Large penetration and high ionization efficiency can only be achieved at energies for above the plasma temperature with H 0 -atoms of several tens of keV

  1. H-mode WEST tungsten divertor operation: deuterium and nitrogen seeded simulations with SOLEDGE2D-EIRENE

    Directory of Open Access Journals (Sweden)

    G. Ciraolo

    2017-08-01

    Full Text Available Simulations of WEST H-mode divertor scenarios have been performed with SOLEDGE2D-EIRENE edge plasma transport code, both for pure deuterium and nitrogen seeded discharge. In the pure deuterium case, a target heat flux of 8 MW/m2 is reached, but misalignment between heat and the particle outflux yields 50 eV plasma temperature at the target plates. With nitrogen seeding, the heat and particle outflux are observed to be aligned so that lower plasma temperatures at the target plates are achieved together with the required high heat fluxes. This change in heat and particle outflux alignment is analysed with respect to the role of divertor geometry and the impact of vertical vs horizontal target plates on neutrals spreading.

  2. Operation method for thermonuclear device and divertor for it

    International Nuclear Information System (INIS)

    Kotake, Michiko; Yoshioka, Ken; Fukumoto, Hideshi; Okazaki, Takashi; Kinoshita, Shigemi; Takeuchi, Kazuhiro.

    1992-01-01

    Divertor plates are disposed subsequently along with circumferential direction of a vacuum vessel in a region where magnetic fluxed generated from the divertor coils are injected toward a container wall. Each of the divertor plates is moved in a state that the injection position of the magnetic fluxes enter to the vacuum vessel is kept constant. Alternatively, each of the divertor plates is inclined at an angle facing the injection direction of plasma particle fluxes, or it is inclined so that the angle between the injection surface and the magnetic fluxes makes an acute angle. Since each of the divertor coils is moved in the state of keeping the injection position of the magnetic fluxes during firing of plasmas, in other words, with on change of the current of the divertor coils, the position of the magnetic fluxed is kept at a predetermined condition. Accordingly, charged particles are prevented from concentrating locally without causing eddy current in the coils and the vacuum vessel, which can contribute to the reduction of the wear of the divertor plates. (N.H.)

  3. 2D imaging of helium ion velocity in the DIII-D divertor

    Science.gov (United States)

    Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

    2018-05-01

    Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

  4. Impact of the impurity seeding for divertor protection on the performance of fusion reactors

    Science.gov (United States)

    Siccinio, Mattia; Fable, Emiliano; Angioni, Clemente; Saarelma, Samuli; Scarabosio, Andrea; Zohm, Hartmut

    2017-10-01

    A 0D divertor and scrape-off layer (SOL) model has been coupled to the 1.5D core transport code ASTRA. The resulting numerical tool has been employed for various parameter scans in order to identify the most convenient choices for the operation of electricity producing fusion devices with seeded impurities for the divertor protection. In particular, the repercussions of such radiative species on the main plasma through the fuel dilution have been taken into account. The main result we found is that, when the limits on the maximum tolerable divertor heat flux are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, i.e. no improvement would descend from a further increase of R or BT once the maximum has been reached. This occurrence appears as an intrinsic physical limit for all devices where a radiative SOL is needed to deal with the power exhaust. Furthermore, the relative importance of the different power loss channels (e.g. hydrogen radiation, charge exchange, perpendicular transport and impurity radiation), through which the power entering the SOL is dissipated before reaching the target plate, is investigated with our model.

  5. ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

    International Nuclear Information System (INIS)

    1982-03-01

    ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

  6. Erosion products of ITER divertor materials under plasma disruption simulation

    Energy Technology Data Exchange (ETDEWEB)

    Guseva, M.I.; Gureev, V.M.; Kolbasov, B.N.; Korshunov, S.N.; Martynenko, Yu.V. E-mail: martyn@nfi.kiae.ru; Stolyarova, V.G.; Strunnikov, V.M.; Vasiliev, V.I

    2003-09-01

    Candidate ITER divertor armor materials: carbon-fiber-composite and four tungsten grades/alloys as well as mixed re-deposited W+Be and W+C layers were exposed in electrodynamic plasma accelerator MKT which provided a pulsed deuterium plasma flux simulating plasma disruptions with maximum ion energy of 1-2 keV, an energy density of 300 kJ/m{sup 2} per shot and a pulse duration of {approx}60 {mu}s. The number of pulses was from 2 to 10. The resultant erosion products were collected on a basalt filter and Si-collectors and studied in terms of morphology and size distribution using both scanning and transmission electron microscopy. Metal erosion products usually occurred in the form of spherical droplets, sometimes flakes. Their size distribution depended on the positioning of the collector. Simultaneously irradiated W, CFC and mixed W+Be targets appeared to have undergone a greater erosion than the same targets irradiated individually. Particles sized from 0.01 to 30 {mu}m were found on collectors and on a molten W-surface. A model of droplet emission and behavior in shielding plasma is provided.

  7. Experimental test campaign on an ITER divertor mock-up

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D.

    2002-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests

  8. Experimental test campaign on an ITER divertor mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: giovanni.dellorco@brasimone.enea.it; Malavasi, A.; Merola, M.; Polazzi, G.; Simoncini, M.; Zito, D

    2002-11-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body (CB), designed with some mechanical and hydraulic simplifications with respect to the reference body and its actively cooled Dummy Armour Prototype (DAP). This DAP consists of a Vertical Target (VT), a Wing (WI) and a Dump Target (DT), manufactured by European industries, which are integrated to the Gas Box Liner (GBL) supplied by the Russian Federation ITER Home Team. In 1999, in parallel with the manufacturing activity, the ITER European Home Team decided to assign to ENEA a Task for checking the component integration and performing the thermal-hydraulic and thermal mechanical testing of the DAP and CB. In 1999-2000, ENEA performed the experimental campaign at Brasimone Labs. The present work presents the experimental results of the component integration and the thermal-hydraulic and thermo-mechanical fatigue tests.

  9. Disruption characteristics in PDX with limiter and divertor discharges

    International Nuclear Information System (INIS)

    Couture, P.; McGuire, K.

    1986-09-01

    A comparison has been made between the characteristics of disruptions with limiter and divertor configurations in PDX. A large data base on disruptions has been collected over four years of machine operation, and a total of 15,000 discharges are contained in the data file. It was found that divertor discharges have less disruptions during ramp up and flattop of the plasma current. However, for divertor discharges a large number of fast, low current disruptions take place during the current ramp down. These disruptions are probably caused by the deformation of the plasma shape

  10. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    International Nuclear Information System (INIS)

    Dell'Orco, G.; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D.

    2001-01-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'

  11. Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

    Energy Technology Data Exchange (ETDEWEB)

    Dell' Orco, G. E-mail: dellorco@brasimone.enea.it; Canneta, A.; Cattadori, G.; Gaspari, G.P.; Merola, M.; Polazzi, G.; Vieider, G.; Zito, D

    2001-10-01

    In 1998, in the frame of the European R and D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the 'Component Integration and Thermal-Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body'.

  12. Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

    Directory of Open Access Journals (Sweden)

    T.D. Rognlien

    2017-08-01

    Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

  13. Heat and particle transport of sol/divertor plasma in the W-shaped divertor on JT-60U

    International Nuclear Information System (INIS)

    Asakura, N.; Sakurai, S.; Hosogane, N.

    1999-01-01

    The plasma profile and parallel flow in the scrape-off layer (SOL) were systematically measured using Mach probes installed at the midplane and the divertor x-point. Quantitative evaluation of a parallel flow: naturally produced in a torus to keep the pressure constant along the field line, was consistent with the measurement. Geometry effects of the W-shaped divertor on the divertor plasma and particle recycling at the newly installed baffle plates were evaluated quantitatively using the edge plasma data. (author)

  14. Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

    Science.gov (United States)

    Subba, Fabio; Aho-Mantila, Leena; Coster, David; Maddaluno, Giorgio; Nallo, Giuseppe F.; Sieglin, Bernard; Wenninger, Ronald; Zanino, Roberto

    2018-03-01

    In this paper we present a computational study on the divertor heat load mitigation through impurity injection for the EU DEMO. The study is performed by means of the SOLPS5.1 code. The power crossing the separatrix is considered fixed and corresponding to H-mode operation, whereas the machine operating condition is defined by the outboard mid-plane upstream electron density and the impurity level. The selected impurity for this study is Ar, based on its high radiation efficiency at SOL characteristic temperatures. We consider a conventional vertical target geometry for the EU DEMO and monitor target conditions for different operational points, considering as acceptability criteria the target electron temperature (≤5 eV to provide sufficiently low W sputtering rate) and the peak heat flux (below 5-10 MW m-2 to guarantee safe steady-state cooling conditions). Our simulations suggest that, neglecting the radiated power deposition on the plate, it is possible to satisfy the desired constraints. However, this requires an upstream density of the order of at least 50% of the Greenwald limit and a sufficiently high argon fraction. Furthermore, if the radiated power deposition is taken into account, the peak heat flux on the outer plate could not be reduced below 15 MW m-2 in these simulations. As these simulations do not take into account neutron loading, they strongly indicate that the vertical target divertor solution with a radiative front distributed along the divertor leg has a very marginal operational space in an EU DEMO sized reactor.

  15. Resonant island divertor experiments on text

    International Nuclear Information System (INIS)

    deGrassie, J.S.; Evans, T.E.; Jackson, G.L.

    1988-09-01

    The first experimental tests of the resonant island divertor (RID) concept have been carried out on the Texas Experimental Tokamak (TEXT). Modular perturbation coils produce static resonant magnetic fields at the tokamak boundary. The resulting magnetic islands are used to guide heat and particle fluxes around a small scoop limiter head. An enhancement in the limiter collection efficiency over the nonisland operation, as evidenced by enhanced neutral density within the limiter head, of up to a factor of 4 is obtained. This enhancement is larger than one would expect given the measured magnitude of the cross-field particle transport in TEXT. It is proposed that electrostatic perturbations occur which enhance the ion convection rate around the islands. Preliminary experiments utilizing electron cyclotron heating (ECH) in conjunction with RID operation have also have been performed. 6 refs., 3 figs

  16. Evaluation of helium cooling for fusion divertors

    International Nuclear Information System (INIS)

    Baxi, C.B.

    1993-09-01

    The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

  17. Transport and divertor studies in the FM-1 spherator

    International Nuclear Information System (INIS)

    Ando, K.; Ejima, S.; Davis, S.; Hawryluk, R.; Hsuan, H.; Meade, D.; Okabayaski, M.; Sauthoff, N.; Schmidt, J.; Sinnis, J.

    1974-10-01

    Fundamental problems of toroidal fusion devices have been investigated in the FM-1 Spherator. These subjects include the transport due to drift wave turbulence in the trapped electron regime, poloidal divertor and impurities, and lower hybrid heating. (auth)

  18. Status of National Spherical Torus Experiment Liquid Lithium Divertor

    Science.gov (United States)

    Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

    2009-11-01

    Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

  19. Two-dimensional divertor modeling and scaling laws

    International Nuclear Information System (INIS)

    Catto, P.J.; Connor, J.W.; Knoll, D.A.

    1996-01-01

    Two-dimensional numerical models of divertors contain large numbers of dimensionless parameters that must be varied to investigate all operating regimes of interest. To simplify the task and gain insight into divertor operation, we employ similarity techniques to investigate whether model systems of equations plus boundary conditions in the steady state admit scaling transformations that lead to useful divertor similarity scaling laws. A short mean free path neutral-plasma model of the divertor region below the x-point is adopted in which all perpendicular transport is due to the neutrals. We illustrate how the results can be used to benchmark large computer simulations by employing a modified version of UEDGE which contains a neutral fluid model. (orig.)

  20. Electron beam irradiation experiments of monoblock divertor mock-up

    International Nuclear Information System (INIS)

    Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

    1993-03-01

    It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

  1. Divertor pumping system with NBI cryopump for JT-60

    International Nuclear Information System (INIS)

    Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

    1998-08-01

    The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

  2. Development of liquid lithium divertor for fusion reactor

    International Nuclear Information System (INIS)

    Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

    2000-01-01

    Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

  3. Investigation of electron parallel pressure balance in the scrape-off layer of deuterium-based radiative divertor discharges IN DIII-D

    International Nuclear Information System (INIS)

    Petrie, T.W.; Carlstrom, T.N.; Allen, S.L.

    1996-10-01

    Electron density, temperature, and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2, and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of ∼ 2 between the midplane separatrix and the X-point, and then an additional ∼3--5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power maintains good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate, and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7--30 eV

  4. Thermomechanical simulation of WEST actively cooled upper divertor

    International Nuclear Information System (INIS)

    Batal, T.; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

    2016-01-01

    The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m 2 . This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m 2 , and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m 2 . The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

  5. Thermomechanical simulation of WEST actively cooled upper divertor

    Energy Technology Data Exchange (ETDEWEB)

    Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

    2016-11-15

    The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

  6. Non-ambipolar divertor flows in heliotron E

    International Nuclear Information System (INIS)

    Chechkin, V.V.; Voitsenya, V.S.; Smirnova, M.S.; Sorokovoj, E.L.; Mizuuchi, T.; Nagasaki, K.; Okada, H.; Funaba, H.; Hamada, T.; Sano, F.; Zushi, H.; Nakasuga, M.; Kondo, K.; Masuzaki, S.; Motojima, O.

    1999-01-01

    The object of the work is to find out (1) the poloidal distributions of PEC in different poloidal cross-sections of the torus within one field period; (2) the link between PEC in the divertor flows (DF) and the characteristics of the divertor field lines; (3) the effect of different methods and regimes of heating on PEC. The data having been obtained enable us to understand at least partially the nature of PEC in the diverted plasma of H-E

  7. Dynamic behavior of detached recombining plasmas during ELM-like plasma heat pulses in the divertor plasma simulator NAGDIS-II

    International Nuclear Information System (INIS)

    Uesugi, Y.; Hattori, N.; Nishijima, D.; Ohno, N.; Takamura, S.

    2001-01-01

    It has been recognized that the ELMs associated with a good confinement at the edge, such as H-mode, must bring an enormous energy to the divertor target plate through SOL and detached plasmas. The understanding of the ELM energy transport through SOL to the divertor target is rather poor at the moment, which leads to an ambiguous estimation of the deposited heat load on the divertor target in ITER. In the present work the ELM-like plasma heat pulse is generated by rf heating in a linear divertor plasma simulator. Energetic electrons with an energy range 10-40 eV are effectively generated by rf heating in low temperature plasmas with (T e )< ∼1 eV. It is observed experimentally that the energetic electrons ionize the highly excited Rydberg atoms quickly, bringing a rapid increase of the ion particle flux to the target, and make the detached plasmas attached to the target. Detailed physical processes about the interaction between the heat pulse with conduction and convection, and detached recombining plasmas are discussed

  8. Multi-fluid modeling of low-recycling divertor regimes

    International Nuclear Information System (INIS)

    Smirnov, R.D.; Pigarov, A.Yu.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

    2010-01-01

    The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

  9. Multi-Fluid Modeling of Low-Recycling Divertor Regimes

    International Nuclear Information System (INIS)

    Smirnov, R.D.; Pigarov, A.Y.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

    2010-01-01

    The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate.

  10. Development of a radiative divertor for DIII-D

    Energy Technology Data Exchange (ETDEWEB)

    Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

    1995-04-01

    We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

  11. Structural design of the DIII-D radiative divertor

    International Nuclear Information System (INIS)

    Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

    1996-10-01

    The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria

  12. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhitlukhin, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)]. E-mail: zhitlukh@triniti.ru; Klimov, N. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Landman, I. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany)]. E-mail: j.linke@fz-juelich.de; Loarte, A. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Merola, M. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Podkovyrov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Federici, G. [ITER JWS Garching, Boltzmannstr. 2, 85748 Garching (Germany); Bazylev, B. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Safronov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Hirai, T. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany); Maynashev, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Levashov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Muzichenko, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-06-15

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m{sup 2} and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 {mu}m per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m{sup 2} and was increased to the average value 0.3 {mu}m per pulse at 1.5 MJ/m{sup 2}. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m{sup 2}. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m{sup 2}.

  13. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

    Zhitlukhin, A.; Klimov, N.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Federici, G.; Bazylev, B.; Pestchanyi, S.; Safronov, V.; Hirai, T.; Maynashev, V.; Levashov, V.; Muzichenko, A.

    2007-06-01

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

  14. ADX: a high field, high power density, advanced divertor and RF tokamak

    Science.gov (United States)

    LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

    2015-05-01

    The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

  15. An operational non destructive examination for ITER divertor plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Durocher, A.; Escourbiac, F.; Farjon, J.L.; Vignal, N.; Cismondi, F. [Association Euratom-CEA Cadarache, 13 - Saint-Paul-lez-Durance (France). Dept. de Recherches sur la Fusion Controlee; Merola, M. [ITER International Team, Cadarache, 13 - St Paul Lez Durance (France); Riccardi, B. [CEFDA CSU-Garching, Garching bei Munchen (Germany)

    2007-07-01

    Full text of publication follows: To meet the power exhaust - heat flux of 20 MW/m{sup 2} - requirements of Plasma Facing Components (PFCs) during plasma operation requires control of their thermal and mechanical integrity. As heat exhaust capability and lifetime of PFCs during in-situ operation are linked to the manufacturing quality, it is an absolute requirement to develop reliable nondestructive examination methods, in particular of the CFC-CuCrZr joint, throughout the manufacturing process. Within the framework of Tokamak Tore Supra upgrade, a pioneering activity has been developed to evaluate the capability of the PFC to be efficiently cooled. In 1998 a test bed - so called SATIR - based on the heat transient method was developed by the CEA and is used today as an inspection tool in order to guarantee the PFCs performances. The technical procurement plan of ITER Divertor targets stated that all Cu cast layers on CFC armour should be subjected to 100% thermographic examination. Each ITER Party should demonstrate its technical capability to carry out the PFC with the required cooling efficiently. The ITER Divertor PFCs pose new challenges especially for the mono-block CFC thickness, and the number of full scale units to be tested which is higher than on any existing or under construction fusion machine. The SATIR method as functional inspection has been identified as the basis test to decide upon the final acceptance of the Divertor PFCs. In order to increase the detection sensitivity of SATIR test bed, several possibilities have been assessed i) the increase of the convective heat transfer coefficient, which improved in a significant way the sensitivity of SATIR diagnostic on ITER components. ii) the installation of a digital infrared camera and the improvement of the thermal signal processing, has led to a considerable increase of performances iii) an innovative process based on spatial image autocorrelation will allow to localize the interlayer defect

  16. EMC3-EIRENE modeling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

    Energy Technology Data Exchange (ETDEWEB)

    Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

    2015-08-15

    Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

  17. First measurements of the ion energy distribution at the divertor strike point during DIII-D disruptions

    International Nuclear Information System (INIS)

    Parks, P.B.; Brooks, N.H.; West, W.P.; Wong, C.P.C.; Bastasz, R.; Wampler, W.R.; Whyte, D.

    1995-12-01

    Plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV

  18. The effect of density on divertor conditions in ASDEX-Upgrade

    International Nuclear Information System (INIS)

    Pitcher, C.S.; Bosch, H.-S.; Buechl, K.; Field, A.; Fuchs, C.; Haas, G.; Junker, W.; Neu, R.; Neuhauser, J.; Wenzel, U.

    1995-01-01

    Detailed experimental divertor data are presented on the profiles of density and temperature in the inner and outer divertor fans, the radiated power distribution, the gas pressure and the spectroscopically derived particle fluxes, all as a function of the discharge density. At low and medium density, the inner divertor is cold and dense compared to the outer divertor. At high density, strong X-point MARFE and separatrix radiation partially detaches the inner divertor. Probe measurements which penetrate into the X-point MARFE at the outer divertor are presented. ((orig.))

  19. R(and)D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project

    International Nuclear Information System (INIS)

    Hirai, T.; Maier, H.; Rubel, M.

    2006-01-01

    The ITER-like Wall Project was initiated at JET, with the goal of testing the reference material combination chosen for ITER: beryllium (Be) in the main chamber (wall and limiters) and tungsten (W) in the divertor. The major aims are to study the tritium retention, material mixing, melt layer behavior and to optimize plasma operation scenarios with a full metal wall. The project requires major design and engineering efforts in R(and)D: (i) bulk W tile, (ii) W coatings on carbon fibre composites (CFC) (iii) Be coatings on Inconel, (iv) Be marker tiles. For the W divertor, two R(and)D tasks were initiated: (1) development of a conceptual design for a bulk W tile as the main outer divertor target plate, and (2) W coating selection from 14 different samples produced by various techniques for the other divertor plates and neutral beam shine. The bulk W tile must withstand power loads of 7 MW/m 2 for 10 s. JET divertor plates are not actively cooled, therefore, heat capacity of the tiles is an important design parameter. In addition to power handling, mechanical structural stability under electromagnetic forces and compatibility with remote handling are the key requirements in the design. The design has been completed. The test-tile survived 100 pulses at 7 MW/m 2 for 10 s in the electron beam facility, JUDITH. The W coatings with different thickness, thin ( 2 and 200 pulses at 10 MW/m 2 for 5 s. In all tested samples cracks developed perpendicularly to the fiber bundles in CFC because of contraction of the coating in the cooling phase. Coatings were also exposed to 1000 ELM-like loading pulses. The thin coatings showed fatigue leading to delamination, whereas for thick coatings better resistance in ELM-like loading. As a result of R(and)D a full W divertor was decided: bulk metal at the outer divertor and W coating at other areas. Be-related R(and)D activities are in two areas. Production of 8-9 μm layers on inner wall cladding Inconel tiles ensures the full coating of

  20. Plasma performance of Wendelstein 7-AS with the new boundary-island divertor modules

    International Nuclear Information System (INIS)

    McCormick, K.; Grigull, P.; Burhenn, R.; Brakel, R.; Ehmler, H.; Feng, Y.; Gadelmeier, F.; Giannone, L.; Hildebrandt, D.; Hirsch, M.; Jaenicke, R.; Kisslinger, J.; Klinger, T.; Klose, S.; Knauer, J.P.; Konig, R.; Kuhner, G.; Laqua, H.P.; Naujoks, D.; Niedermeyer, H.; Pasch, E.; Ramasubramanian, N.; Rust, N.; Sardei, F.; Wagner, F.; Weller, A.; Wenzel, U.; Werner, A.

    2002-01-01

    A promising new plasma operational regime on the Wendelstein stellarator W7-AS has been discovered. It is extant above a threshold density and characterized by flat density profiles, high energy- and low impurity-confinement times and edge-localized radiation. Impurity accumulation is avoided. Quasi-stationary discharges with line-averaged densities n e to 4x10 20 m -3 , radiation levels to 90%, and partial plasma detachment at the divertor target plates can be simultaneously realized. Energy confinement is up to twice that predicted by a conventional scaling. Copyright (2002) Australian National University- Research School of Physical Sciences and Engineering

  1. The role of ''momentum removal'' in divertor detachment

    Energy Technology Data Exchange (ETDEWEB)

    Kukushkin, A.S. [Kurchatov Institute, Moscow (Russian Federation); NRNU MEPhI, Moscow (Russian Federation); Pacher, H.D. [INRS-EMT, Varennes, Quebec (Canada)

    2016-08-15

    The role of ''momentum removal'' (the drag force on the plasma ion flow) in divertor detachment is considered and analysed in detail. This analysis of the 2D modelling results shows that the drag force cannot reduce the power and particle flux to the target directly. However, it is essential for creating the conditions for efficient radiation and volumetric plasma recombination, which in turn do the job. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  2. Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

    International Nuclear Information System (INIS)

    Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Soukhanovskii, V. A.; Umansky, M. V.

    2014-01-01

    In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription

  3. Divertor characteristics and control on the W-shaped divertor with pump of JT-60U

    International Nuclear Information System (INIS)

    Hosogane, N.; Kubo, H.; Higashijima, S.

    1999-01-01

    Roles of the inner leg pumping and the private dome, which are special features of the W-shaped divertor of JT-60U, have been investigated. The following observations were made: The inner leg pumping functions well in attached states or partially detached states with weak X-point MARFE where the inner particle recycling is enhanced. A combination of main gas puff and inner leg pump is effective in reduction of intrinsic carbon impurity. Geometrical effects of the private dome on transport of hydrocarbons in the private flux region was confirmed by spectroscopic measurements of CD-band intensity profile and impurity transport simulation code using experimental data. (author)

  4. Experimental study on highly collisional edge plasmas in W7-AS island divertor configurations

    International Nuclear Information System (INIS)

    Grigull, P.; Hildebrandt, D.; Sardei, F.; Feng, Y.; Herre, G.; Herrmann, A.; Hofmann, J.V.; Kisslinger, J.; Kuehner, G.; Niedermeyer, H.; Schneider, R.; Verbeek, H.; Wagner, F.; Wolf, R.; Zhang, X.D.

    1997-01-01

    Edge plasma scenarios in island divertor configurations ('natural' magnetic islands intersected by targets) are studied by comparing data from moderate to high density NBI discharges with 3D code (EMC3/EIRENE) results. The data strongly indicate that high recycling with significant particle flux enhancement was achieved in this geometry. But, plasma pressure losses towards the targets are relatively strong, and high recycling sets in only at n e >10 20 m -3 . The respective density enhancement in front of the targets is moderate (up to a factor of about three relative to the upstream density). These scenarios are also in basic agreement with B2/EIRENE code predictions. At n e >1.5 x 10 20 m -3 detachment seems to develop. Improvements are expected from additional coils controlling the field line pitch inside the islands, and from optimized targets which will better focus recycling neutrals into the islands. Both are in preparation. (orig.)

  5. A program to evaluate the erosion on the CFC tiles of the ITER divertor

    Energy Technology Data Exchange (ETDEWEB)

    D' Agata, E. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany)], E-mail: elio.dagata@iter.org; Ogorodnikova, O.V. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France); Tivey, R. [ITER International Team, ITER Joint Work Site, Boltzmannstr 2, 85748 Garching (Germany); Lowry, C.; Schlosser, J. [Association Euratom-CEA, CEA/DSM/DRFC, CEA/Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2007-10-15

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets, and carbon fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime due to a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, from start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations very well validated against experimental data. The code has been developed in the APDL language to operate inside a commercial and certificated finite element program such as ANSYS.

  6. A program to evaluate the erosion on the CFC tiles of the ITER divertor

    International Nuclear Information System (INIS)

    D'Agata, E.; Ogorodnikova, O.V.; Tivey, R.; Lowry, C.; Schlosser, J.

    2007-01-01

    The plasma-facing surfaces of the ITER divertor are armoured with tungsten in the upper part of the inner and outer vertical targets, and carbon fibre composite (CFC) in the lower part, the region where the scrape-off layer intercepts the divertor. The CFC in the form of a monoblock in the vertical target is the most loaded part of the plasma-facing surfaces, and hence it is subjected to high erosion and has a significant risk of failure. A program has been developed with the aim of understanding the impact on the erosion lifetime due to a combination of two main effects: the material property variations (particularly pronounced in CFC) and the presence of joining defects. The software allows the evolution of the surface profile of the armour to be predicted and the margin on critical heat flux at the heat-sink-to-coolant interface to be estimated for a range of postulated defects, from start-of-life through to end-of-life of the component. In assessing erosion, the code takes account of geometry and sublimation, and physical and chemical erosion of the CFC armour. The incident angle (a glancing angle of a few degrees) of the particle and heat flux onto the target is taken into account. The program has been validated by comparison with analytical approximations very well validated against experimental data. The code has been developed in the APDL language to operate inside a commercial and certificated finite element program such as ANSYS

  7. A computational study of operating regimes for poloidal divertors

    International Nuclear Information System (INIS)

    Petravic, M.; Heifetz, D.; Post, D.

    1982-01-01

    We have identified three theoretical operating regimes for poloidal divertors. These regimes are determined by the geometry of the divertor and the input energy and particle fluxes, and are characterized by the divertor plasma density and temperature. A fully self-consistent two-dimensional model for the plasma and neutral atom and molecule transport was used to study poloidal divertor operation. Extensions of our previous calculations important to this study were the inclusion of parallel electron and ion thermal conduction. We find that the key physics in divertor operation is the neutral recycling near the neutralizer plate. This can be parametrized by R = GAMMAsub(P)/GAMMAsub(O), the ratio of particle flux striking the neutralizer plate to the particle flux entering the divertor. Values of R approx. equal to 1 can be produced by large pumping rates near the neutralizer plates resulting in low neutral recycling and a high temperature, low density divertor plasma. By decreasing the pumping near the neutralizer plate, R can be raised to an intermediate value of 5-10, the plasma temperature lowered by the same factor, and the density raised by a factor of 10-30. In this regime, escape of the neutrals back to the main plasma is virtually blocked. By further restricting the pumping, R can be raised to twenty or more, thereby lowering the temperature by a factor of twenty or more and raising the density by a factor of ninety or more. Such high density regimes have been observed on D-III and appear to offer the most promise for impurity control and particle control on large reactor experiments such as INTOR or FED. In this paper, we explore the range 3 < R < 16. (orig.)

  8. Development of divertor simulation research in the GAMMA 10/PDX tandem mirror

    International Nuclear Information System (INIS)

    Nakashima, Y.; Sakamoto, M.; Yoshikawa, M.; Oki, K.; Takeda, H.; Ichimura, K.; Hosoi, K.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Kariya, T.; Katanuma, I.; Kohagura, J.; Minami, R.; Numakura, T.; Wang, X.; Iwamoto, M.; Hosoda, Y.; Asakura, Nobuyuki; Fukumoto, Masakatsu; Kubo, Hirotaka; Hatayama, A.; Hirooka, Y.; Masuzaki, S.; Sagara, A.; Shoji, M.; Kado, S.; Matsuura, H.; Nagata, S.; Shikama, T.; Nishino, N.; Ohno, N.; Tonegawa, A.; Ueda, Y.

    2014-10-01

    This paper describes the recent development of divertor simulation research towards the characterization and control of the detached plasma. In the end-mirror of large tandem mirror device GAMMA 10/PDX, additional ICRF heating experiments in the anchor-cells significantly increases the density in both the anchor and the central cells, which attained the highest particle flux up to 1.7×10 23 particles/s·m 2 at the end-mirror exit. Massive gas injection (H 2 and noble gases) to enhance the radiation cooling in divertor simulation experimental module (D-module) was performed and we have succeeded for the first time in achieving detachment of high temperature plasma equivalent to the SOL plasma of tokamaks by using linear device. A remarkable reduction of the electron temperature (from few tens eV to < 3 eV) on the target plate was successfully achieved associated with the strong reduction of particle and heat fluxes. Two-dimensional image of Hα emission in D-module observed with high-speed camera showed the bright emission in upstream region and strong reduction near the target plate. These results indicate radiation cooling and formation of detached plasma due to gas injection. It is also found that Xe gas is much effective on achieving detached plasma than Ar gas. Simultaneous injection of noble gas and hydrogen gas showed the most effective results on detached plasma generation, which indicates the effect of molecular activated recombination (MAR) processes. The above results will contribute to establishment of detached plasma control and clarification of radiation cooling mechanism towards the development of future divertor systems. (author)

  9. Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

    International Nuclear Information System (INIS)

    Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

    1985-01-01

    Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented

  10. Two-dimensional impurity transport calculations for a high recycling divertor

    International Nuclear Information System (INIS)

    Brooks, J.N.

    1986-04-01

    Two dimensional analysis of impurity transport in a high recycling divertor shows asymmetric particle fluxes to the divertor plate, low helium pumping efficiency, and high scrapeoff zone shielding for sputtered impurities

  11. The ITER divertor cassette. Steady state characterisation and draining and drying transient hydraulic analyses

    International Nuclear Information System (INIS)

    Pietro Alessandro Di Maio; Valerio Tomarchio; Giuseppe Vella; Irene Zammuto; Giovanni Dell'Orco

    2005-01-01

    Full text of publication follows: The divertor is one of the most challenging components of the next step ITER nuclear fusion reactor. It is aimed at controlling the characteristics of boundary plasma, reducing the impurities in the plasma and sustaining the heat and particle fluxes arising from it, during normal and transient operations as well as during disruption events. The ITER divertor consists of 54 cassettes, each one mainly composed of three Plasma-Facing Components (PFCs), namely the inner vertical target, the outer vertical target and the dome-liner, actively cooled by subcooled pressurized water. Each PFC consists in a number of plasma facing units, cooled in parallel and assembled onto a supporting structure. The water maximum total flow rate, for the whole divertor, should be 1000 kg/s, with 100-150 deg. C inlet/outlet temperatures, 4.2 MPa inlet pressure and a maximum pressure drop of 1.4 MPa. The PFCs are cooled in series, with a maximum water velocity in the channel of 11 m/s, whilst the water coolant is routed via the cassette body. Due to the extremely high heat loads expected onto the PFCs (up to 20 MW/m 2 over 20 s), the hydraulic design of the divertor is particularly demanding. It shall ensure that the foreseen flow rate actually reaches each plasma-facing unit to ensure an adequate cooling and to prevent any risk of Critical Heat Flux (CHF). Sufficient margin ( > 40 %) to avoid the reaching of a CHR limit on the PFCs could be obtained by using hypervapotron design inside the flat channels and swirl flow turbulence tape promoters inside the vertical target cooling tubes. Furthermore the overall pressure drop and flow rate shall be within the specified design limit to avoid an unduly high pumping power. Another important issue is the definition of a proper procedure to drain the coolant and dry the divertor components prior to the maintenance operations as well as to refill them with water after maintenance, ensuring a complete elimination of

  12. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  13. The edge plasma and divertor in TIBER

    Energy Technology Data Exchange (ETDEWEB)

    Barr, W.L.

    1987-10-16

    An open divertor configuration has been adopted for TIBER. Most recent designs, including DIII-D, NET and CIT use open configurations and rely on a dense edge plasma to shield the plasma from the gas produced at the neutralizer plate. Experiments on ASDEX, PDX, D-III, and recently on DIII-D have shown that a dense edge plasma can be produced by re-ionizing most of the gas produced at the plate. This high recycling mode allows a large flux of particles to carry the heat to the plate, so that the mean energy per particle can be low. Erosion of the plate can be greatly reduced if the average impact energy of the ions at the plate can be reduced to near or below the threshold for sputtering of the plate material. The present configuration allows part of the flux of edge plasma ions to be neutralized at the entrance to the pumping duct so that helium is pumped as well as hydrogen. 7 refs., 3 figs.

  14. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  15. The edge plasma and divertor in TIBER

    International Nuclear Information System (INIS)

    Barr, W.L.

    1987-01-01

    An open divertor configuration has been adopted for TIBER. Most recent designs, including DIII-D, NET and CIT use open configurations and rely on a dense edge plasma to shield the plasma from the gas produced at the neutralizer plate. Experiments on ASDEX, PDX, D-III, and recently on DIII-D have shown that a dense edge plasma can be produced by re-ionizing most of the gas produced at the plate. This high recycling mode allows a large flux of particles to carry the heat to the plate, so that the mean energy per particle can be low. Erosion of the plate can be greatly reduced if the average impact energy of the ions at the plate can be reduced to near or below the threshold for sputtering of the plate material. The present configuration allows part of the flux of edge plasma ions to be neutralized at the entrance to the pumping duct so that helium is pumped as well as hydrogen. 7 refs., 3 figs

  16. Divertor conceptual designs for a fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

    2007-01-01

    The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it

  17. Theory and Simulations of ELM Control with a Snowflake Divertor

    Energy Technology Data Exchange (ETDEWEB)

    Ryutov, D.; Cohen, B.; Cohen, R.; Makowski, M. A.; Menard, J.; Rognlien, T.; Soukhanovskii, V.; Umansky, M.; Xu, X., E-mail: ryutov1@llnl.gov [Lawrence Livermore National Laboratory, Livermore (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton (United States)

    2012-09-15

    Full text: This paper is concerned with the use of a snowflake (SF) divertor for the control and mitigation of edge localized modes (ELMs). Our research is focused on the following three issues: 1. Effect of the SF geometry on neoclassical ion orbits near the separatrix, including prompt ion losses and the related control mechanism for the electric field and plasma flow in the pedestal; 2. Influence of the thereby modified flow and of high poloidal plasma beta in the divertor region on plasma turbulence and transport in the snowflake-plus geometry; 3. Reaction of the SF divertor to type-1 ELM events. Neoclassical ion orbits in the vicinity of the SF separatrix are changed due to a much weaker poloidal field near the null and much longer particle dwell-time in this area. This leads to an increase of the prompt ion loss, which then affects the radial electric field profile near the separatrix. The resulting E x B flow shear in the pedestal region affects the onset of ELMs. The electric field and velocity shear are then used as a background for two-fluid simulations of the edge plasma turbulence in a realistic geometry with the 3D BOUT code. A SF-plus geometry is chosen, so that the separatrix topology remains the same as for the standard X-point divertor, whereas the magnetic shear both inside and outside the separatrix increases dramatically. It is found that mesoscale instabilities are suppressed when the geometry is close to a perfect SF. In situations where complete suppression of ELMs is impossible, the SF divertor offers a path to reducing heat loads during ELM events to an acceptable level. Two effects, both related to the weakness of the poloidal field near the SF null, act synergistically in the same favorable direction. The first is the onset of strong, curvature-driven convection in the divertor, triggered by the increase of the poloidal pressure during the ELM and leading to the splitting of the heat flux between all four (as is the case in a SF geometry

  18. Probabilistic analysis of divertor plate lifetime in tokamak reactors

    International Nuclear Information System (INIS)

    Golinescu, R.P.; Kazimi, M.S.

    1994-01-01

    Defining a methodology for a reliability estimate of the International Tokamak Experimental Reactor (ITER) divertor is the objective of the study summarized in this paper. If ITER could be designed such that no transients of any type occurred, the divertor reliability would be controlled by erosion of material during normal operation. The occurrence of several transient events results in important contribution to the expected divertor failure rate. Some transients cause the temperature in the divertor plate (DP) to rise; if these temperatures get too high, the structural elements in the DP will weaken and subsequently suffer structural failure and possibly reach the melting temperature. Using the limited data available leads to the result that there is a high probability that the DP will reliably withstand a peak heat flux of 11 MW/m 2 . However, transient events will lead to a much shorter lifetime than desirable for DP's, mainly due to the expected severe effects of plasma disruptions. If transients occurred, but the shutdown mechanism succeeded to perform without inducing a disruption, divertor reliability could be significantly improved. Improved characterization of the disruption conditions, and enlarged scope of failure modes should be pursued to gain confidence in the present conclusions

  19. Optimal thermal-hydraulic performance for helium-cooled divertors

    International Nuclear Information System (INIS)

    Izenson, M.G.; Martin, J.L.

    1996-01-01

    Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

  20. Analysis of sweeping heat loads on divertor plate materials

    International Nuclear Information System (INIS)

    Hassanein, A.

    1991-01-01

    The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m 2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

  1. Toroidal asymmetries in divertor impurity influxes in NSTX

    Directory of Open Access Journals (Sweden)

    F. Scotti

    2017-08-01

    Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

  2. Non-inductive current start-up and plasma equilibrium with an inboard poloidal field null by means of electron cyclotron waves in QUEST

    International Nuclear Information System (INIS)

    Zushi, H.; Hasegawa, M.; Hanada, K.; Idei, H.; Nakamura, K.; Fujisawa, A.; Nagashima, Y.; Matsuoka, K.; Tashima, S.; Ishiguro, M.; Banerjee, S.; Sharma, S.K.; Liu, H.; Nishino, N.; Isobe, M.; Toi, K.; Okamura, S.; Maekawa, T.; Fukuyama, A.; Ejiri, A.; Yamaguchi, T.; Hiratsuka, J.; Takase, Y.; Kikuchi, Mitsuru; Ueda, Y.; Mitarai, O.

    2012-11-01

    Non-inductive current start-up via relativistic electron cyclotron resonance interaction is investigated for the high ratio (∼10%) of vertical B v to toroidal B t fields and the concave field lines in the QUEST spherical tokamak. In the start-up scenario with an internal poloidal field null (IPN), the fast current start-up rate of 0.3-0.5 MA/sec and correlation with mildly relativistic electrons accelerated due to multiple ECR interaction are observed. In steady state high β p equilibrium characterized by the inboard null (R s ∼ 0.7×R 0 ) and εβ p of 1.5 is achieved, where ε, β p are the inverse aspect ratio and poloidal beta, respectively. Relaxation oscillations in this equilibrium and confinement of the energetic electrons are discussed. (author)

  3. A study on the fusion reactor - A study on the design feature of fusion reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Jin [Chosun University, Kwangju (Korea, Republic of); Paek, Won Pil; Jang, Soon Hong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Sim, Young Jae [Kyungsang University, Jinju (Korea, Republic of)

    1996-09-01

    The contents and scope of the project can be summarized as, - study on the trend of divertor design - study on characteristics of coolant materials - study on characteristics of divertor materials - study on the thermal analysis method of divertor design. 36 refs., 12 tabs., 16 figs. (author)

  4. Characterization of the island divertor plasma of W7-AS stellarator in the deeply detached state with volume recombination

    International Nuclear Information System (INIS)

    Ramasubramanian, N.; Koenig, R.; Feng, Y.; Giannone, L.; Grigull, P.; Klinger, T.; McCormick, K.; Thomsen, H.; Wenzel, U.

    2004-01-01

    In the high-density H-mode of the Stellarator Wendelstein 7-AS, the plasma detaches from the island divertor targets when the line-averaged density exceeds a critical value. This quasi-stationary detachment is found to be partial and shows edge-localized, poloidally asymmetric radiation. The spectroscopic characteristics of the deeply detached plasma are reported, including evidence for volume recombination. The detached plasma radiates up to 90% of the absorbed power with larger contributions from the locations close to magnetic X-points outside the divertor region. The spectral analysis of the Balmer series indicate very high densities and low temperatures at the detached regions. The results of the spectral analysis underline the importance of three-dimensional modelling. An initial comparison is made with the latest results from EMC3-EIRENE modelling. (author)

  5. Critical heat flux analysis and R and D for the design of the ITER divertor

    International Nuclear Information System (INIS)

    Raffray, A.R.; Chiocchio, S.; Merola, M.; Tivey, R.; Vieider, G.; Schlosser, J.; Driemeyer, D.; Escourbiac, F.; Grigoriev, S.; Youchison, D.

    1999-01-01

    The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW/m 2 over ∼10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R and D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R and D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of ∼1.4 or more for all divertor components. (orig.)

  6. Developing physics basis for the snowflake divertor in the DIII-D tokamak

    Science.gov (United States)

    Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

    2018-03-01

    Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

  7. Aberrations in preliminary design of ITER divertor impurity influx monitor

    Energy Technology Data Exchange (ETDEWEB)

    Kitazawa, Sin-iti, E-mail: kitazawa.siniti@jaea.go.jp [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Ogawa, Hiroaki [Naka Fusion Institute, Japan Atomic Energy Agency, JAEA, Naka 311-0193 (Japan); Katsunuma, Atsushi; Kitazawa, Daisuke [Core Technology Center, Nikon Corporation, Yokohama 244-8533 (Japan); Ohmori, Keisuke [Customized Products Business Unit, Nikon Corporation, Mito 310-0843 (Japan)

    2015-12-15

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  8. Neutron activation behavior of NET/ITER divertor structural materials

    International Nuclear Information System (INIS)

    Smid, I.; Weimann, G.; Kny, E.; Kneringer, G.; Reheis, N.

    1995-01-01

    The post-activation behavior of the materials carbon, TZM (99.3 % Mo) and Mo.41Re, as well as of high temperature brazes suitable for their joining after irradiation with 14 MeV neutrons has been evaluated. The activity, dose rate and energy generation after exposure to an ignited fusion plasma is presented for various time steps after shutdown. The impact of the activity and the afterheat production on the handling and storage conditions of retired divertor components is simulated, the required protection for maintenance is discussed. Further the temperature of stored divertor elements after a full time operation in NET was calculated. No major afterheat production will occur and thus no special cooling is to be provided after approximately one month. Taking into account convection and radiation the equilibrium temperature of vertically stored environment/aircooled divertor elements is predicted to be approximately 100 degree C. (author)

  9. The simple map for a single-null divertor tokamak

    International Nuclear Information System (INIS)

    Punjabi, A.; Verma, A.; Boozer, A.

    1996-01-01

    We present the simple map for a single-null divertor tokamak. The simple map is an area-preserving map based on the idea that magnetic field lines are a single-degree-of-freedom time-dependent Hamiltonian system, and that the basic features of such systems near the X-point are generic. We obtain the properties of this map and the resulting footprints of field lines on the divertor plate. These include the width of the stochastic layer, the edge safety factor, the area of the footprint and the amount of magnetic flux diverted. We give the safety factor profile, the average and median values of strike angles, lengths and the Liapunov exponents. We describe how the effects of magnetic perturbations can be included in the simple map. We show how the map can be applied to the problem of the determination of heat flux on the divertor plate in tokamaks. (Author)

  10. Charge exchange in a divertor plasma with excited particles

    International Nuclear Information System (INIS)

    Krasheninnikov, S.I.; Lisitsa, V.S.; Pigarov, A.Y.

    1988-01-01

    A model is constructed for the dynamics of neutral atoms and multicharged ions in a tokamak plasma. The influence of cascade excitation on charge exchange and ionization is taken into account. The effective rates of the resonant charge exchange of a proton with a hydrogen atom, the nonresonant charge exchange of a helium atom with a proton, and that of an α particle with atomic hydrogen are calculated as functions of the parameters of the divertor plasma in a tokamak. The charge exchange H + +He→H+He + can represent a significant fraction (∼30%) of the total helium ionization rate. Incorporating the charge exchange of He 2+ with atomic hydrogen under the conditions prevailing in the divertor plasma of the INTOR reactor can lead to substantial He 2+ →He + conversion and thereby reduce the sputtering of the divertor plates by helium ions

  11. Radiative divertor plasmas with convection in DIII-D

    International Nuclear Information System (INIS)

    Leornard, A.W.; Porter, G.D.; Wood, R.D.

    1998-01-01

    The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

  12. Development of divertor pumping system with superpermeable membrane

    International Nuclear Information System (INIS)

    Nakamura, Y.; Ohyabu, N.; Suzuki, H.; Nakahara, Y.; Livshits, A.; Notkin, M.; Alimov, V.; Busnyuk, A.

    2000-01-01

    A new divertor pumping system with superpermeable membranes of group Va-metals (Nb, V) is now under research and development. Properties of membrane pumping were investigated with the use of a plasma device simulating divertor plasma conditions. The deposition of metal (Fe) and non-metal (C) impurities on the membrane upstream surface results in a degradation of plasma driven superpermeation at the membrane temperature T m m ≥800 deg. C. The same temperature effect on superpermeation is observed at sputtering of membrane surface by energetic plasma ions. In addition, the first application of the membrane pumping to fusion devices has been carried out and a deuterium pumping through the membrane was demonstrated under the conditions of divertor plasma in the JFT-2M tokamak

  13. Aberrations in preliminary design of ITER divertor impurity influx monitor

    International Nuclear Information System (INIS)

    Kitazawa, Sin-iti; Ogawa, Hiroaki; Katsunuma, Atsushi; Kitazawa, Daisuke; Ohmori, Keisuke

    2015-01-01

    Highlights: • Divertor impurity influx monitor for ITER (DIM) is procured by JADA. • DIM is designed to observe light from nuclear fusion plasma directly. • DIM is under preliminary design phase. • The spot diagrams were suppressed within the core of receiving fiber. • The aberration of DIM is suppressed in the preliminary design. - Abstract: Divertor impurity influx monitor for ITER (DIM) is a diagnostic system that observes light from nuclear fusion plasma directly. This system is affected by various aberrations because it observes light from the fan-array chord near the divertor in the ultraviolet–near infrared wavelength range. The aberrations should be suppressed to the extent possible to observe the light with very high spatial resolution. In the preliminary design of DIM, spot diagrams were suppressed within the core of the receiving fiber's cross section, and the resulting spatial resolutions satisfied the design requirements.

  14. Experimental demonstration of vector E x vector B plasma divertor

    International Nuclear Information System (INIS)

    Strait, E.J.; Kerst, D.W.; Sprott, J.C.

    1977-01-01

    The vector E x vector B drift due to an applied radial electric field in a tokamak with poloidal divertor can speed the flow of plasma out of the scrape-off region, and provide a means of externally controlling the flow rate and thus the width of the density fall-off. An experiment in the Wisconsin levitated toroidal octupole, using vector E x vector B drifts alone, demonstrates divertor-like behavior, including 70% reduction of plasma density near the wall and 40% reduction of plasma flux to the wall, with no adverse effects on confinement of the main plasma

  15. Divertor development for a future fusion power plant

    International Nuclear Information System (INIS)

    Norajitra, Prachai

    2011-01-01

    Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m 2 . In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading

  16. Divertor plate concept with carbon based armour for NET

    International Nuclear Information System (INIS)

    Moons, F.; Howard, R.; Kneringer, G.; Stickler, R.

    1989-01-01

    A series of tests has been performed on simulated divertor elements for NET at the JET neutral beam injector test bed. The test section consisted of a water cooled main structure, the surface of which was protected with a carbon based armour in the form of tiles. The scope of these was to study the thermal behaviour of mechanically attached tiles with the use of an intermediate soft carbon layer to improve the thermal contact under divertor relevant conditions. (author). 4 refs.; 4 figs.; 1 tab

  17. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    2001-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  18. Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

    International Nuclear Information System (INIS)

    Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

    1999-01-01

    Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

  19. Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

    International Nuclear Information System (INIS)

    Mai, L.P.; Malick, F.S.

    1981-01-01

    The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented

  20. Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

    International Nuclear Information System (INIS)

    Yoder, Graydon L. Jr.; Harvey, Karen; Ferrada, Juan J.

    2011-01-01

    A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

  1. OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

    Science.gov (United States)

    Pengfei, ZHANG; Ling, ZHANG; Zhenwei, WU; Zong, XU; Wei, GAO; Liang, WANG; Qingquan, YANG; Jichan, XU; Jianbin, LIU; Hao, QU; Yong, LIU; Juan, HUANG; Chengrui, WU; Yumei, HOU; Zhao, JIN; J, D. ELDER; Houyang, GUO

    2018-04-01

    Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.

  2. He-cooled divertor for DEMO. Fabrication technology for tungsten cooling fingers

    Energy Technology Data Exchange (ETDEWEB)

    Reiser, J.; Norajitra, P.; Widak, V.; Krauss, W. [Forschungszentrum Karlsruhe GmbH (Germany)

    2008-07-01

    A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) has been developed for the ''post-ITER'' demonstration reactor (DEMO) at the Forschungszentrum Karlsruhe [1, 2]. The main function of the divertor is to keep the plasma free from impurities by catching particles, such as fusion ash and eroded particles from the first wall. From the divertor surface, a maximum heat load of 10 MW/m{sup 2} at least has to be removed. The whole divertor is split up into a number of cassettes (48 according to the latest design studies [3]). Each cassette is cooled separately. The target plates are provided with several cooling fingers to keep the thermal stresses low. Each cooling finger consists of a tungsten tile which is brazed to a thimble-like cap made of a tungsten alloy W-1%La2O3 (WL10) underneath. The thimble has to be connected to the ODS EUROFER steel structure, which is accomplished by brazing again. The tungsten/tungsten brazing is exposed to 1200 C operation temperature while the tungsten/steel brazing joint must withstand 700 C operating temperature. Cooling of the finger is achieved by multi-jet impingement with helium. The inlet temperature of helium is 600 C and rises up to 700 C at the outlet. With this kind of cooling, a mean heat transfer coefficient of 35.000 W/(m{sup 2*}K) can be reached. This compact report will focus on the manufacturing of such a cooling finger unit at FZK. It will cover the machining of the tungsten tile as well as of the thimble and, the brazing of the parts. The major aim of this activity is, on the one hand, to obtain functioning mock-ups with high quality and high reliability, in particular in terms of minimising the surface roughness, cracks, and micro-cracks. On the other hand, effort should also be laid on realising the mass production from economic point of view. (orig.)

  3. Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

    International Nuclear Information System (INIS)

    Kaufmann, M.; Bosch, H.S.; Herrmann, A.

    1999-01-01

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (author)

  4. Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

    International Nuclear Information System (INIS)

    Kaufmann, M.; Bosch, H.-S.; Herrmann, A.

    2001-01-01

    In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

  5. Discharge power dependence of carbon dust flux in a divertor simulator

    International Nuclear Information System (INIS)

    Nishiyama, Katsushi; Morita, Yasuhiko; Uchida, Giichiro; Yamashita, Daisuke; Kamataki, Kunihiro; Seo, Hyunwoong; Itagaki, Naho; Koga, Kazunori; Shiratani, Masaharu; Ashikawa, Naoko; Masuzaki, Suguru; Nishimura, Kiyohiko; Sagara, Akio; Bornholdt, Sven; Kersten, Holger

    2013-01-01

    In fusion devices, dust particles are generated due to plasma–wall interactions and may cause safety or operational problems. Therefore it is necessary to clarify the generation and transport mechanisms of dust particles. Here we have measured energy influx from H 2 plasmas toward a graphite target using a calorimetric probe and compared the results with the dust flux toward a dust collecting substrate set on the reactor wall. The dust flux decreases with increasing the energy influx. For the higher discharge power, the more number of dust particles tend to redeposit onto the graphite target due to the higher ion drag force and hence the dust flux toward the reactor wall becomes smaller. The results show that dust inventory depends strongly on energy influx to graphite divertor plates in fusion devices

  6. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    International Nuclear Information System (INIS)

    Missirlian, M.; Escourbiac, F.; Schmidt, A.; Riccardi, B.; Bobin-Vastra, I.

    2011-01-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  7. Examination of high heat flux components for the ITER divertor after thermal fatigue testing

    Energy Technology Data Exchange (ETDEWEB)

    Missirlian, M., E-mail: marc.missirlian@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Escourbiac, F., E-mail: frederic.escourbiac@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Schmidt, A., E-mail: a.schmidt@fz-juelich.de [Forschungszentrum Juelich, IFE-2 (Germany); Riccardi, B., E-mail: Bruno.Riccardi@f4e.europa.eu [Fusion For Energy, E-08019 Barcelona (Spain); Bobin-Vastra, I., E-mail: isabelle.bobinvastra@areva.com [AREVA-NP, 71200 Le Creusot (France)

    2011-10-01

    An extensive development programme has been carried out in the EU on high heat flux components within the ITER project. In this framework, a full-scale vertical target (VTFS) prototype was manufactured with all the main features of the corresponding ITER divertor design. The fatigue cycling campaign on CFC and W armoured regions, proved the capability of such a component to meet the ITER requirements in terms of heat flux performances for the vertical target. This paper discusses metallographic observations performed on both CFC and W part after this intensive thermal fatigue testing campaign for a better understanding of thermally induced mechanical stress within the component, especially close to the armour-heat sink interface.

  8. Manufacturing and testing of a copper/CFC divertor mock-up for JET

    International Nuclear Information System (INIS)

    Brossa, M.; Ciric, D.; Deksnis, E.; Falter, H.; Guerreschi, U.; Peacock, A.; Pick, M.; Rossi, M.; Shen, Y.; Zacchia, F.

    1995-01-01

    An actively cooled divertor is a possible option for future developments at The Joint European Torus (JET). A proof of principle actively cooled tile has been produced in order to qualify the relevant manufacturing technologies and the non destructive control processes. In this frame Ansaldo Ricerche (ARI) has been involved in the construction of a mock-up comprising 6 OFHC copper tubes for water cooling that are brazed to a plate made out of carbon fibre composite (CFC). The final objective was the high heat flux testing of the mock-up at JET in order to evaluate the general behaviour of the component under relevant operating conditions. The key point of the work was the realisation of a sound joint by adapting the expertise gained in ARI in previous R and D activities on brazing heterogeneous materials. Reliable methods for ultrasonic examinations of the pieces were also set up. For successful application to the JET pumped divertor a water-cooled CFC target plate must show surface temperatures of 2 . Furthermore, global hydraulic considerations specific to JET limit the system pressure to 0.7 MPa. In such a design, critical heat flux is not the key limit, rather the reliability of the CFC-copper joint in terms of extent of wetting. First tests in the neutral beam test bed at JET show an adequate response for fluxes up to 15 MW/m 2 . (orig.)

  9. Divertor load footprint of ELMs in pellet triggering and pacing experiments at JET

    Energy Technology Data Exchange (ETDEWEB)

    Frigione, D., E-mail: domenico.frigione@frascati.enea.it [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Garzotti, L. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Lennholm, M. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Alper, B. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Artaserse, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bennett, P. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Giovannozzi, E. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Eich, T. [Max Planck Institute for Plasma Physics, Garching (Germany); Kocsis, G. [WIGNER RCP RMI, POB 49, 1525 Budapest (Hungary); Lang, P.T. [Max Planck Institute for Plasma Physics, Garching (Germany); Maddaluno, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Mooney, R. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Rack, M. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich, 52425 Jülich (Germany); Sips, G. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Tvalashvili, G. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Viola, B. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Wilkes, D. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom)

    2015-08-15

    An investigation of pellet pacing and triggering of Edge Localized Modes (ELMs) was carried out in the frame of ELM mitigation studies aimed at reducing their damaging effects on the plasma-facing components (PFCs). The divertor power load footprint of triggered ELMs was compared with gas puffing controlled ELMs. Small pellets, corresponding to a few per cent of the target plasma particle inventory, were used to minimize the fueling effect and the total particle throughput. There is no evidence that pellets can reduce the divertor power load with respect to gas fueling when operating at the same ELM frequency. The line average density and the energy confinement time remained constant when the gas was progressively substituted by pellets. The launch from the Vertical High Field Side (VHFS) confirmed to be more efficient in ELM triggering than from the Low Field Side (LFS) while the power load footprint remained the same both in time evolution and in spatial distribution when changing the injection geometry.

  10. Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

    International Nuclear Information System (INIS)

    Hayashi, T.; Ochiai, K.; Masaki, K.; Gotoh, Y.; Kutsukake, C.; Arai, T.; Nishitani, T.; Miya, N.

    2006-01-01

    Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12 C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12 C data were lower than 0.006. Additionally, the maximum (H + D)/ 12 C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk

  11. Modeling results for a linear simulator of a divertor

    International Nuclear Information System (INIS)

    Hooper, E.B.; Brown, M.D.; Byers, J.A.; Casper, T.A.; Cohen, B.I.; Cohen, R.H.; Jackson, M.C.; Kaiser, T.B.; Molvik, A.W.; Nevins, W.M.; Nilson, D.G.; Pearlstein, L.D.; Rognlien, T.D.

    1993-01-01

    A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach ∼ 1 Gw/m 2 along the magnetic fieldlines and > 10 MW/m 2 on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report

  12. 2-D fluid transport simulations of gaseous/radiative divertors

    International Nuclear Information System (INIS)

    Rognlien, T.D.; Brown, P.N.; Campbell, R.B.; Kaiser, T.B.; Knoll, D.A.; McHugh, P.R.; Porter, G.D.; Rensink, M.E.; Smith, G.R.

    1994-01-01

    The features of the fully implicit 2-D fluid code UEDGE are described. The utility of the code is demonstrated by showing bifurcations or multiple solutions of the tokamak edge plasma for both deuterium and impurity injection in the divertor. (orig.)

  13. Modelling of radial electric field profile for different divertor configurations

    International Nuclear Information System (INIS)

    Rozhansky, V; Kaveeva, E; Voskoboynikov, S; Counsell, G; Kirk, A; Meyer, H; Coster, D; Conway, G; Schirmer, J; Schneider, R

    2006-01-01

    The impact of divertor configuration on the structure of the radial electric field has been simulated by the B2SOLPS5.0 transport fluid code. It is shown that the change in the parallel flows in the scrape-off layer, which are transported through the separatrix due to turbulent viscosity and diffusivity, should result in variation of the radial electric field and toroidal rotation in the separatrix vicinity. The modelling predictions are compared with the measurements of the radial electric field for the low field side equatorial mid-plane of ASDEX Upgrade in lower, upper and double-null (DN) divertor configurations. The parallel (toroidal) flows in the scrape-off layer and mechanisms for their formation are analysed for different geometries. It is demonstrated that a spike in the electric field exists at the high field side equatorial mid-plane in the connected DN divertor configuration. Its origin is connected with different potential drops between the separatrix vicinity and divertor plates in the two disconnected scrape-off layers, while the separatrix should be at almost the same potential. The spike might be important for additional turbulent suppression

  14. Plasma/neutral gas transport in divertors and limiters

    International Nuclear Information System (INIS)

    Gierszewski, P.J.

    1983-09-01

    The engineering design of the divertor and first wall region of fusion reactors requires accurate knowledge of the energies and particle fluxes striking these surfaces. Simple calculations indicate that approx. 10 MW/m 2 heat fluxes and approx. 1 cm/yr erosion rates are possible, but there remain fundamental physics questions that bear directly on the engineering design. The purpose of this study was to treat hydrogen plasma and neutral gas transport in divertors and pumped limiters in sufficient detail to answer some of the questions as to the actual conditions that will be expected in fusion reactors. This was accomplished in four parts: (1) a review of relevant atomic processes to establish the dominant interactions and their data base; (2) a steady-state coupled O-D model of the plasma core, scrape-off layer and divertor exhaust to determine gross modes of operation and edge conditions; (3) a 1-D kinetic transport model to investigate the case of collisionless divertor exhaust, including non-Maxwellian ions and neutral atoms, highly collisional electrons, and a self-consistent electric field; and (4) a 3-D Monte Carlo treatment of neutral transport to correctly account for geometric effects

  15. Electron and molecular ion collisions relevant to divertor plasma

    International Nuclear Information System (INIS)

    Takagi, H.

    2005-01-01

    We introduce the concept of the multi-channel quantum defect theory (MQDT) and show the outline of the MQDT newly extended to include the dissociative states. We investigate some molecular processes relevant to the divertor plasma by using the MQDT: the dissociative recombination, dissociative excitation, and rotation-vibrational transition in the hydrogen molecular ion and electron collisions. (author)

  16. Modular He-cooled divertor for power plant application

    International Nuclear Information System (INIS)

    Diegele, Eberhard; Kruessmann, R.; Malang, S.; Norajitra, P.; Rizzi, G.

    2003-01-01

    Gas cooled divertor concepts are regarded as a suitable option for fusion power plants because of an increased thermal efficiency for power conversion systems and the use of a coolant compatible with all blanket systems. A modular helium cooled divertor concept is proposed with an improved heat transfer. The concept employs small tiles made of tungsten and brazed to a finger-like structure made of Mo-alloy (TZM). Design goal was a heat flux of at least 15 MW/m 2 and a minimum temperature of the structure of 600 deg.C. The divertor has to survive a number of cycles (100-1000) between operating temperature and room temperature even for the steady state operation assumed. Thermo-hydraulic design requirements for the concepts include to keep the pumping power below 10% of the thermal power to the divertor plates, and simultaneously achieving a heat transfer coefficient in excess of 60 kW/m 2 K. Inelastic stress analysis indicates that design allowable stress limits on primary and secondary (thermal) stresses as required by the ITER structural design criteria are met even under conservative assumptions. Finally, critical issues for future development are addressed

  17. Lifetime analysis for fusion reactor first walls and divertor plates

    International Nuclear Information System (INIS)

    Horie, T.; Tsujimura, S.; Minato, A.; Tone, T.

    1987-01-01

    Lifetime analysis of fusion reactor first walls and divertor plates is performed by (1) a one-dimensional analytical plate model, and (2) a two-dimensional elastic-plastic finite element method. Life-limiting mechanisms and the limits of applicability for these analysis methods are examined. Structural design criteria are also discussed. (orig.)

  18. Mechanical Design of the NSTX Liquid Lithium Divertor

    Energy Technology Data Exchange (ETDEWEB)

    R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

    2009-02-19

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  19. Taming the plasma-material interface with the snowflake divertor.

    Energy Technology Data Exchange (ETDEWEB)

    Soukhanovskii, V A

    2015-04-24

    Experiments in several tokamaks have provided increasing support for the snowflake configuration as a viable tokamak heat exhaust concept. This white paper summarizes the snowflake properties predicted theoretically and studied experimentally, and identifies outstanding issues to be resolved in existing and future facilities before the snowflake divertor can qualify for the reactor interface.

  20. Mechanical Design of the NSTX Liquid Lithium Divertor

    International Nuclear Information System (INIS)

    Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

    2009-01-01

    The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

  1. Manufacturing and joining technologies for helium cooled divertors

    International Nuclear Information System (INIS)

    Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

    2014-01-01

    Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

  2. Optimization and limitations of known DEMO divertor concepts

    International Nuclear Information System (INIS)

    Reiser, Jens; Rieth, Michael

    2012-01-01

    Highlights: ► Limitations of the materials. ► Improved H 2 O cooled divertor. ► Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 °C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600–800 °C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call ‘cooling of the coolant’. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and available materials.

  3. Initial development of the DIII–D snowflake divertor control

    Science.gov (United States)

    Kolemen, E.; Vail, P. J.; Makowski, M. A.; Allen, S. L.; Bray, B. D.; Fenstermacher, M. E.; Humphreys, D. A.; Hyatt, A. W.; Lasnier, C. J.; Leonard, A. W.; McLean, A. G.; Maingi, R.; Nazikian, R.; Petrie, T. W.; Soukhanovskii, V. A.; Unterberg, E. A.

    2018-06-01

    Simultaneous control of two proximate magnetic field nulls in the divertor region is demonstrated on DIII–D to enable plasma operations in an advanced magnetic configuration known as the snowflake divertor (SFD). The SFD is characterized by a second-order poloidal field null, created by merging two first-order nulls of the standard divertor configuration. The snowflake configuration has many magnetic properties, such as high poloidal flux expansion, large plasma-wetted area, and additional strike points, that are advantageous for divertor heat flux management in future fusion reactors. However, the magnetic configuration of the SFD is highly-sensitive to changes in currents within the plasma and external coils and therefore requires complex magnetic control. The first real-time snowflake detection and control system on DIII–D has been implemented in order to stabilize the configuration. The control algorithm calculates the position of the two nulls in real-time by locally-expanding the Grad–Shafranov equation in the divertor region. A linear relation between variations in the poloidal field coil currents and changes in the null locations is then analytically derived. This formulation allows for simultaneous control of multiple coils to achieve a desired SFD configuration. It is shown that the control enabled various snowflake configurations on DIII–D in scenarios such as the double-null advanced tokamak. The SFD resulted in a 2.5×  reduction in the peak heat flux for many energy confinement times (2–3 s) without any adverse effects on core plasma performance.

  4. New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

    International Nuclear Information System (INIS)

    Damiani, C.; Baldi, L.; Galbiati, L.; Irving, M.; Lorenzelli, L.; Micciche, G.; Muro, L.; Nucci, S.; Varocchi, G.; Poggianti, A.; Fermani, G.; Maisonnier, D.; Palm