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

DOE PAGES

Si, Hang; Guo, Houyang Y.; Covele, Brent; ...

2018-04-04

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 frommore » $$1.18\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$ to $$0.88\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$. Furthermore, 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 $$0.67\\times {{10}^{19}}\\,{{{\\rm m}}^{-3}}$$, thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.« less

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

NASA Astrophysics Data System (ADS)

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.

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

DOE PAGES

Soukhanovskii, V. A.

2017-04-28

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

A review of radiative detachment studies in tokamak advanced magnetic divertor configurations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

The present vision for a plasma–material interface in the tokamak is an axisymmetric poloidal magnetic X-point divertor. Four tasks are accomplished by the standard poloidal X-point divertor: plasma power exhaust; particle control (D/T and He pumping); reduction of impurity production (source); and impurity screening by the divertor scrape-off layer. A low-temperature, low heat flux divertor operating regime called radiative detachment is viewed as the main option that addresses these tasks for present and future tokamaks. Advanced magnetic divertor configuration has the capability to modify divertor parallel and cross-field transport, radiative and dissipative losses, and detachment front stability. Advanced magnetic divertormore » configurations are divided into four categories based on their salient qualitative features: (1) multiple standard X-point divertors; (2) divertors with higher order nulls; (3) divertors with multiple X-points; and (4) long poloidal leg divertors (and also with multiple X-points). As a result, this paper reviews experiments and modeling in the area of radiative detachment in the advanced magnetic divertor configurations.« less

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

NASA Astrophysics Data System (ADS)

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.

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

NASA Astrophysics Data System (ADS)

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

2013-10-01

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.

Plasma detachment in divertor tokamaks

NASA Astrophysics Data System (ADS)

Leonard, A. W.

2018-04-01

Observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasma E× B drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.

SOLPS modeling of the effect on plasma detachment of closing the lower divertor in DIII-D

DOE PAGES

Sang, C. F.; Stangeby, P. C.; Guo, H. Y.; ...

2016-12-15

SOLPS modeling has been carried out to assess the effect of tightly closing the lower divertor in DIII-D, which at present is almost fully open, on the achievement of cold dissipative/detached divertor conditions. To isolate the impact of other factors on the divertor plasma solution and to make direct comparisons, most of the parameters including the meshes were kept as similar as possible. Only the neutral baffling was modified to compare a fully open divertor with a tightly closed one. The modeling shows that the tightly closed divertor greatly improves trapping of recycling neutrals, thereby increasing radiative and charge exchangemore » losses in the divertor and reducing the electron temperature T et and deposited power density q dep at the target plate. Furthermore, the closed structure enables the divertor plasma to enter into highly dissipative and detached divertor conditions at a significantly lower upstream density. The effects of divertor closure on the neutral density and pressure, and their correlation with the divertor plasma conditions are also demonstrated. The effect of molecular D 2- ion D + elastic collisions and neutral-neutral collisions on the divertor plasma solution are assessed.« less

Plasma detachment in divertor tokamaks

DOE PAGES

Leonard, A. W.

2018-02-07

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Plasma detachment in divertor tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, A. W.

In this study, observations of divertor plasma detachment in tokamaks are reviewed. Plasma detachment is characterized in terms of transport and dissipation of power, momentum and particle flux along the open field lines from the midplane to the divertor. Asymmetries in detachment onset and other characteristics between the inboard and outboard divertor plasmas is found to be primarily driven by plasmamore » $$\\vec{E}$$ x $$\\vec{B}$$ drifts. The effect of divertor plate geometry and magnetic configuration on divertor detachment is summarized. Control of divertor detachment has progressed with a development of a number of diagnostics to characterize the detached state in real-time. Finally the compatibility of detached divertor operation with high performance core plasmas is examined.« less

Divertor scenario development for NSTX Upgrade

NASA Astrophysics Data System (ADS)

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.

Small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

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.

Detachment experiments in new DIII-D upper divertor

NASA Astrophysics Data System (ADS)

Moser, A. L.; Leonard, A. W.; Groebner, R. J.; Guo, H.; Wang, H.; Watkins, J. G.; McLean, A. G.; Fenstermacher, M. E.; Shafer, M. W.; Briesemeister, A. R.; Hinson, E. T.

2017-10-01

Installation of the Small Angle Slot (SAS) in the upper divertor of DIII-D enables new studies of the effect of target and baffle geometry on divertor detachment. This structure provides a more-closed upper divertor as well as the SAS divertor itself. Initial SAS experiment results indicate that divertor detachment occurs at a lower line-averaged density than in the more-open, lower single null divertor configurations on DIII-D. In contrast, the increased divertor closure of the new installation did not reduce the upstream density required for detachment beyond that achieved with the previous upper divertor structure. Particle pumping in the upper divertor structure is found to produce a 10 % reduction in the pedestal density required for detachment compared to the case with no pumping. Comparisons focus on both the onset of detachment (measured by in-target Langmuir probes) as a function of upstream density, as well as the effect of the new divertor configurations on pedestal density profiles. Work supported by US DOE under DE-FC02-04ER54698, DE-AC05-00OR22725, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-SC00013911.

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

DOE PAGES

Frerichs, H.; Schmitz, O.; Covele, B.; ...

2018-02-28

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. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter 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 diverter 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 three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frerichs, H.; Schmitz, O.; Covele, B.

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. Therefore, small changes in the strikemore » point location can be expected to have a large impact on diverter 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 diverter 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 three dimensional edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.« less

Magnetic configuration flexibility of snowflake divertor for HL-2M [Analysis of snowflake divertor configurations for HL-2M

DOE PAGES

Zheng, G. Y.; Xu, X. Q.; Ryutov, D. D.; ...

2014-07-09

HL-2M (Li, 2013 [1]) is a tokamak device that is under construction. Based on the magnetic coils design of HL-2M, four kinds of divertor configurations are calculated by CORSICA code (Pearlstein et al., 2001 [2]) with the same main plasma parameters, which are standard divertor, exact snowflake divertor, snowflake-plus divertor and snowflake-minus divertor configurations. The potential properties of these divertors are analyzed and presented in this paper: low poloidal field area around X-point, connection length from outside mid-plane to the primary X-point, target plate design and magnetic field shear. The results show that the snowflake configurations not only can reducemore » the heat load at divertor target plates, but also may improve the magneto-hydrodynamic stability by stronger magnetic shear at the edge. Furthermore, a new divertor configuration, named “tripod divertor”, is designed by adjusting the positions of the two X-points according to plasma parameters and magnetic coils current of HL-2M.« less

Advanced Divertor Design and Application under Modern Superconducting Tokamak Constraints

NASA Astrophysics Data System (ADS)

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

2013-10-01

With current ITER projections already predicting divertor exhaust heat loads in the 5-10 MW/m2 range, i.e. at the maximum tolerance, it is clear that the divertor heat load problem will only be exacerbated for future superconducting tokamaks, as well as perhaps some modern tokamaks today. Thus, an advanced divertor, such as the X-Divertor (XD), Super-X Divertor (SXD), or Snowflake (SF) will become a virtual necessity to reduce incident heat flux at the target plates. Using the 2D magnetic equilibrium code CORSICA, we explore the possibilities of creating an advanced divertor for a next-generation superconducting tokamak (Ip = 15 MA, BT = 5.3 T, R = 6.2 m) under nominal engineering constraints. Advanced divertors were achieved with no in-vessel PF coils, PF current densities below 30 MA/m2, and vertical maintenance access, all of which are favorable conditions for tokamaks today. Both the XD and SF divertors are readily achievable while maintaining core plasma performance, and the advantages and disadvantages of each are discussed in turn. Some thought is given as to how the divertor cassette will need to be modified to accommodate advanced divertors. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Investigations on the heat flux and impurity for the HL-2M divertor

NASA Astrophysics Data System (ADS)

Zheng, G. Y.; Cai, L. Z.; Duan, X. R.; Xu, X. Q.; Ryutov, D. D.; Cai, L. J.; Liu, X.; Li, J. X.; Pan, Y. D.

2016-12-01

The controllability of the heat load and impurity in the divertor is very important, which could be one of the critical problems to be solved in order to ensure the success for a steady state tokamak. HL-2M has the advantage of the poloidal field (PF) coils placed inside the demountable toroidal field (TF) coils and close to the main plasma. As a result, it is possible to make highly accurate configuration control of the advanced divertor for HL-2M. The divertor target geometry of HL-2M has been designed to be compatible with different divertor configurations to study the divertor physics and support the high performance plasma operations. In this paper, the heat loads and impurities with different divertor configurations, including the standard X-point divertor, the snowflake-minus divertor and two tripod divertor configurations for HL-2M, are investigated by numerical simulations with the SOLPS5.0 code under the current design of the HL-2M divertor geometry. The plasmas with different conditions, such as the low discharge parameters with {{I}\\text{p}}   =  0.5 MA at the first stage of HL-2M and the high parameters with {{I}\\text{p}}   =  2.0 MA during the normal operations, are simulated. The heat load profiles and the impurity distributions are obtained, and the control of the peak heat load and the effect of impurity on the core plasma are discussed. The compatibility of different divertor configurations for HL-2M is also evaluated. It is seen that the excellent compatibility of different divertor configurations with the current divertor geometry has been verified. The results show that the snowflake-minus divertor and the tripod divertor with {{d}x}=30 \\text{cm} present good performance in terms of the heat load profiles and the impurity distributions under different conditions, which may not have a big effect on the core plasma. In addition, it is possible to optimize the distance between the two X-points, {{d}x} , to achieve a better performance in terms of the parameters of discharges.

"Snowflake" divertor configuration in NSTX

NASA Astrophysics Data System (ADS)

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.

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

NASA Astrophysics Data System (ADS)

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.

Comparative divertor-transport study for helical devices

NASA Astrophysics Data System (ADS)

Feng, Y.; Kobayashi, M.; Sardei, F.; Masuzaki, S.; Kisslinger, J.; Morisaki, T.; Grigull, P.; Yamada, H.; McCormick, K.; Ohyabu, N.; König, R.; Yamada, I.; Giannone, L.; Narihara, K.; Wenzel, U.; Morita, S.; Thomsen, H.; Miyazawa, J.; Hildebrandt, D.; Watanabe, T.; Wagner, F.; Ashikawa, N.; Ida, K.; Komori, A.; Motojima, O.; Nakamura, Y.; Peterson, B. J.; Sato, K.; Shoji, M.; Tamura, N.; Tokitani, M.; LHD experimental Group

2009-09-01

Using the island divertors (IDs) 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 sizes 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. The fundamental role of low-order magnetic islands in both divertor concepts is emphasized. 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 predicted to exist for W7-X. The paper focuses on identifying and understanding the role of divertors for high density plasma operations in helical devices. In this regard, special attention is paid to investigating the divertor function for controlling intrinsic impurities. Impurity transport behaviour and wall-sputtering processes of CX-neutrals are studied under different divertor plasma conditions. A divertor retention effect on intrinsic impurities at high SOL collisonalities is predicted for all the three devices. The required SOL plasma conditions and the underlying mechanisms are analysed in detail. Numerical results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. Different SOL transport regimes are numerically identified for the standard divertor configuration of W7-X and the possible consequences on high density plasmas are assessed. All the EMC3-EIRENE simulations presented in this paper are based on vacuum fields and comparisons with local diagnostics are made for low-ß plasmas.

Divertor heat flux mitigation in the National Spherical Torus Experimenta)

NASA Astrophysics Data System (ADS)

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.

Divertor detachment

NASA Astrophysics Data System (ADS)

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.

A review of direct experimental measurements of detachment

DOE PAGES

Boedo, J.; McLean, A. G.; Rudakov, D. L.; ...

2018-02-22

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A review of direct experimental measurements of detachment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boedo, J.; McLean, A. G.; Rudakov, D. L.

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. Here, we review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson Scattering (TS) in the divertor regionmore » and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.« less

A super-cusp divertor configuration for tokamaks

NASA Astrophysics Data System (ADS)

Ryutov, D. D.

2015-10-01

> This study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase's cusp divertor. It turns out that the set of remote coils can indeed produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough control that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called `a super-cusp'. General geometrical features of the three-null configurations produced by remote coils are described. Issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.

A review of direct experimental measurements of detachment

NASA Astrophysics Data System (ADS)

Boedo, J.; McLean, A. G.; Rudakov, D. L.; Watkins, J. G.

2018-04-01

Detached divertor plasmas feature strong radial and parallel gradients of density, temperature, electric fields and flow over the divertor volume and therefore, sampling the divertor plasma directly provides crucial knowledge to the interpretation and modeling efforts. We review the contribution of diagnostics that directly sample the plasma to the advancement of knowledge of the physics of detachment and detached divertors, such as the characteristics of the various regimes, discovery and quantification of drifts and identification of convection of heat and particles. We focus on wall probes, scanning probes, retarding field analyzers and Thomson scattering in the divertor region and also include the contribution of measurements away from the divertor that provide insight on how divertor detachment affects core, edge or pedestal conditions. Wall probes are critical as they can be installed in closed volumes of difficult access to other diagnostics and measure plasma parameters at the divertor structures, which define the plasma boundary conditions and where detachment effects are more likely to be strongest.

Thermal Analysis of the Divertor Primary Heat Transfer System Piping During the Gas Baking Process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yoder Jr, Graydon L; Harvey, Karen; Ferrada, Juan J

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.

Advantages and Challenges of Radiative Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

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.

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

NASA Astrophysics Data System (ADS)

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.

OEDGE modeling of plasma contamination efficiency of Ar puffing from different divertor locations in EAST

NASA Astrophysics Data System (ADS)

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.

Characteristics of the Secondary Divertor on DIII-D

NASA Astrophysics Data System (ADS)

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.

Divertor-leg instability for finite beta and radially-tilted divertor plate

NASA Astrophysics Data System (ADS)

Cohen, R. H.; Ryutov, D. D.

2004-11-01

Plasma in the divertor leg may experience a fast instability caused by sheath boundary conditions (BC). Perturbations cannot penetrate beyond the X point because of very strong shearing in its vicinity. Accordingly, this instability could increase cross-field transport in the divertor leg, and thereby reduce the heat load on the divertor plate, without having any appreciable negative effect on core plasma confinement. A way of describing the role of shearing in terms of the surface resistivity attributed to a ``control plane'' below the X point has recently been suggested (Contr. Plasma Phys., v. 44, p. 168, 2004). We use this BC, plus sheath BC at the divertor plate. We include effects of finite beta and of the radial tilt of the divertor plate. We optimize the radial tilt in order to maximize radial transport in divertor legs. We discuss experimental signatures of the instability: i) phase velocity and wave-numbers of the most unstable modes; ii) correlations between fluctuations of various parameters; and iii) the differences between fluctuations in the common and private flux regions.

A super-cusp divertor configuration for tokamaks

DOE PAGES

Ryutov, D. D.

2015-08-26

Our study demonstrates a remarkable flexibility of advanced divertor configurations created with the remote poloidal field coils. The emphasis here is on the configurations with three poloidal field nulls in the divertor area. We are seeking the structures where all three nulls lie on the same separatrix, thereby creating two zones of a very strong flux expansion, as envisaged in the concept of Takase’s cusp divertor. It turns out that the set of remote coils can produce a cusp divertor, with additional advantages of: (i) a large stand-off distance between the divertor and the coils and (ii) a thorough controlmore » that these coils exert over the fine features of the configuration. In reference to these additional favourable properties acquired by the cusp divertor, the resulting configuration could be called ‘a super-cusp’. General geometrical features of the three-null configurations produced by remote coils are described. Furthermore, issues on the way to practical applications include the need for a more sophisticated control system and possible constraints related to excessively high currents in the divertor coils.« less

Implementation of a long leg X-point target divertor in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

Kuang, A. Q.; Cao, N. M.; Creely, A. J.; Dennett, C. A.; Hecla, J.; Hoffman, H.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

2017-10-01

A long leg X-point target divertor geometry in a double null geometry has been implemented in the ARC pilot plant design, exploiting ARC's demountable toroidal field (TF) coils and FLiBe immersion blanket, which allow superconducting poloidal field coils to be located inside the TF coils, adequately shielded from neutrons. This new design maintains the original TF coil size, core plasma shape, and attains a tritium breedin ratio 1.08. The long leg divertor geometry provides significant advantages. Neutron transport computations indicate a factor of 10 reduction in divertor material neutron damage rate compared to the first wall, easing requirements for high heat flux components. Simulations have shown that long legged divertors are able to maintain a passively stable detachment front that stays in the divertor leg over a wide power window, in principle, responding immediately to fast changes in power exhaust. The ARC design exploits this new paradigm for divertor heat flux control: fewer concerns about coping with fast transients and a focus on neutron-tolerant diagnostics to measure and adjust detachment front locations in the outer divertor legs over long timescales.

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

NASA Astrophysics Data System (ADS)

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 λ<450 nm) to prevent the 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.

Response to "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)

NASA Astrophysics Data System (ADS)

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

2014-05-01

Relying on coil positions relative to the plasma, the "Comment on `Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake' " [Phys. Plasmas 21, 054701 (2014)], emphasizes a criterion for divertor characterization that was critiqued to be ill posed [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)]. We find that no substantive physical differences flow from this criteria. However, using these criteria, the successful NSTX experiment by Ryutov et al. [Phys. Plasmas 21, 054701 (2014)] has the coil configuration of an X-divertor (XD), rather than a snowflake (SF). On completing the divertor index (DI) versus distance graph for this NSTX shot (which had an inexplicably missing region), we find that the DI is like an XD for most of the outboard wetted divertor plate. Further, the "proximity condition," used to define an SF [M. Kotschenreuther et al., Phys. Plasmas 20, 102507 (2013)], does not have a substantive physics basis to override metrics based on flux expansion and line length. Finally, if the criteria of the comment are important, then the results of NSTX-like experiments could have questionable applicability to reactors.

Theory of Advanced Magnetic Divertors

NASA Astrophysics Data System (ADS)

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 < 1) fall on opposite sides of the standard divertor SD (DI = 1). Amongst other things, DI signifies the rate of convergence (divergence) of the flux surfaces near the divertor plate; the flux surfaces of SFD are more convergent contracting) than the SD while the 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.

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

DOE PAGES

Rognlien, Thomas D.; McLean, Adam G.; Fenstermacher, Max E.; ...

2017-01-27

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 H-mode regime. The data set, which spans a range of plasmas densities for both forward and reverse toroidal magnetic field (B t) over a range of plasma densities, is provided by divertor Thomson scattering (DTS). Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (T e) and density (n e) across both divertor legs for individual discharges. The calculations show the same features of in/out plasma asymmetries as measured inmore » the experiment, with the normal B t direction (ion ∇B drift toward the X-point) having higher n e and lower T e 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. Furthermore, these 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.« less

Alternative divertor target concepts for next step fusion devices

NASA Astrophysics Data System (ADS)

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.

DOE FES FY2017 Joint Research Target Fourth Quarter Milestone Report for theNational Spherical Torus Experiment Upgrade.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A.

2017-09-13

A successful high-performance plasma operation with a radiative divertor has been demonstrated on many tokamak devices, however, significant uncertainty remains in accurately modeling detachment thresholds, and in how detachment depends on divertor geometry. Whereas it was originally planned to perform dedicated divertor experiments on the National Spherical Tokamak Upgrade to address critical detachment and divertor geometry questions for this milestone, the experiments were deferred due to technical difficulties. Instead, existing NSTX divertor data was summarized and re-analyzed where applicable, and additional simulations were performed.

Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

NASA Astrophysics Data System (ADS)

Soukhanovskii, Vsevolod

2007-11-01

Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required for detachment.

Divertor target shape optimization in realistic edge plasma geometry

NASA Astrophysics Data System (ADS)

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

2014-07-01

Tokamak divertor design for next-step fusion reactors heavily relies on numerical simulations of the plasma edge. Currently, the design process is mainly done in a forward approach, where the designer is strongly guided by his experience and physical intuition in proposing divertor shapes, which are then thoroughly assessed by numerical computations. On the other hand, automated design methods based on optimization have proven very successful in the related field of aerodynamic design. By recasting design objectives and constraints into the framework of a mathematical optimization problem, efficient forward-adjoint based algorithms can be used to automatically compute the divertor shape which performs the best with respect to the selected edge plasma model and design criteria. In the past years, we have extended these methods to automated divertor target shape design, using somewhat simplified edge plasma models and geometries. In this paper, we build on and extend previous work to apply these shape optimization methods for the first time in more realistic, single null edge plasma and divertor geometry, as commonly used in current divertor design studies. In a case study with JET-like parameters, we show that the so-called one-shot method is very effective is solving divertor target design problems. Furthermore, by detailed shape sensitivity analysis we demonstrate that the development of the method already at the present state provides physically plausible trends, allowing to achieve a divertor design with an almost perfectly uniform power load for our particular choice of edge plasma model and design criteria.

Tokamak power exhaust with the snowflake divertor: Present results and outstanding issues

DOE PAGES

Soukhanovskii, V. A.; Xu, X.

2015-09-15

Here, a snowflake divertor magnetic configuration (Ryutov in Phys Plasmas 14(6):064502, 2007) with the second-order poloidal field null offers a number of possible advantages for tokamak plasma heat and particle exhaust in comparison with the standard poloidal divertor with the first-order null. Results from snowflake divertor experiments are briefly reviewed and future directions for research in this area are outlined.

Interpretations of the impact of cross-field drifts on divertor flows in DIII-D with UEDGE

DOE PAGES

Jaervinen, Aaro E.; Allen, Steve L.; Groth, Mathias; ...

2017-01-27

Simulations using the multi-fluid code UEDGE indicates that, in low confinement (Lmode) plasmas in DIII-D, recycling driven flows dominate poloidal particle flows in the divertor, whereas E×B drift flows dominate the radial particle flows. In contrast, in high confinement (H-mode) conditions E×B drift flows dominate both poloidal and radial particle flows in the divertor. UEDGE indicates that the toroidal C 2+ flow velocities in the divertor plasma are entrained within 30% to the background deuterium flow in both Land H-mode plasmas in the plasma region where the CIII 465 nm emission is measured. Therefore, UEDGE indicates that the Carbon Dopplermore » Coherence Imaging System (CIS), measuring the toroidal velocity of the C 2+ ions, can provide insight to the deuterium flows in the divertor. Parallel-to-B velocity dominates the toroidal divertor flow; direct drift impact being less than 1%. Toroidal divertor flow is predicted to reverse when the magnetic field is reversed. This is explained by the parallel-B flow towards the nearest divertor plate corresponding to opposite toroidal directions in opposite toroidal field configurations. Due to strong poloidal E×B flows in H-mode, net poloidal particle transport can be in opposite direction than the poloidal component of the parallel-B plasma flow.« less

Testing the role of molecular physics in dissipative divertor operations through helium plasmas at DIII-D

DOE PAGES

Canik, John M.; Briesemeister, Alexis R.; McLean, Adam G.; ...

2017-05-10

Recent experiments in DIII-D helium plasmas are examined to resolve the role of atomic and molecular physics in major discrepancies between experiment and modeling of dissipative divertor operation. Helium operation removes the complicated molecular processes of deuterium plasmas that are a prime candidate for the inability of standard fluid models to reproduce dissipative divertor operation, primarily the consistent under-prediction of radiated power. Modeling of these experiments shows that the full divertor radiation can be accounted for, but only if measures are taken to ensure that the model reproduces the measured divertor density. Relying on upstream measurements instead results in amore » lower divertor density and radiation than is measured, indicating a need for improved modeling of the connection between the diverter and the upstream scrape-off layer. Furthermore, these results show that fluid models are able to quantitatively describe the divertor-region plasma, including radiative losses, and indicate that efforts to improve the fidelity of the molecular deuterium models are likely to help resolve the discrepancy in radiation for deuterium plasmas.« less

Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

DOE Data Explorer

Frerichs, H. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Waters, I. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Schmitz, O. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Canal, G. P. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Evans, T. E. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Feng, Y. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States); Soukhanovskii, V. A. [Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)

2016-06-01

The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads are so called 'advanced divertors' which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their interaction based on the NSTX-U setup. An overview of different divertor configurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which is related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.

Evaluation of heat and particle controllability on the JT-60SA divertor

NASA Astrophysics Data System (ADS)

Kawashima, H.; Hoshino, K.; Shimizu, K.; Takizuka, T.; Ide, S.; Sakurai, S.; Asakura, N.

2011-08-01

The JT-60SA divertor design has been established on the basis of engineering requirements and physics analysis. Heat and particle fluxes under the full input power of 41 MW can give severe heat loads on the divertor targets, while the allowable heat load is limited below 15 MW/m2. Dependence of the heat flux mitigation on a D2 gas-puff is evaluated by SONIC simulations for high density (ne_ave ˜ 1 × 1020 m-3) high current plasmas. It is found that the peak heat load 10 MW/m2 with dense (ned > 4 × 1020 m-3) and cold (Ted, Tid ⩽ 1 eV) divertor plasmas are obtained at a moderate gas-puff of Γpuff = 15 × 1021 s-1. Divertor plasmas are controlled from attached to detached condition using the divertor pump with pumping-speed below 100 m3/s. In full non-inductive current drive plasmas with low density (ne_ave ˜ 5 × 1019 m-3), the reduction of divertor heat load is achieved with the Ar injection.

Optimized tokamak power exhaust with double radiative feedback in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Eich, T.; Fuchs, J. C.; Giannone, L.; Herrmann, A.; Schweinzer, J.; Treutterer, W.; the ASDEX Upgrade Team

2012-12-01

A double radiative feedback technique has been developed on the ASDEX Upgrade tokamak for optimization of power exhaust with a standard vertical target divertor. The main chamber radiation is measured in real time by a subset of three foil bolometer channels and controlled by argon injection in the outer midplane. The target heat flux is in addition controlled by nitrogen injection in the divertor private flux region using either a thermoelectric sensor or the scaled divertor radiation obtained by a bolometer channel in the outer divertor. No negative interference of the two radiation controllers has been observed so far. The combination of main chamber and divertor radiative cooling extends the operational space of a standard divertor configuration towards high values of P/R. Pheat/R = 14 MW m-1 has been achieved so far with nitrogen seeding alone as well as with combined N + Ar injection, with the time-averaged divertor peak heat flux below 5 MW m-2. Good plasma performance can be maintained under these conditions, namely H98(y,2) = 1 and βN = 3.

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

NASA Astrophysics Data System (ADS)

Covele, B.; Kotschenreuther, M.; Mahajan, S.; Valanju, P.; Leonard, A.; Watkins, J.; Makowski, M.; Fenstermacher, M.; Si, H.

2017-08-01

The X-divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at 10-20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. However, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. The model also points to carbon radiation as the primary driver of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency for core operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.

Evaluating Stellarator Divertor Designs with EMC3

NASA Astrophysics Data System (ADS)

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.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak.

PubMed

Xu, J C; Wang, L; Xu, G S; Luo, G N; Yao, D M; Li, Q; Cao, L; Chen, L; Zhang, W; Liu, S C; Wang, H Q; Jia, M N; Feng, W; Deng, G Z; Hu, L Q; Wan, B N; Li, J; Sun, Y W; Guo, H Y

2016-08-01

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triple probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.

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

NASA Astrophysics Data System (ADS)

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.

Upgrade of Langmuir probe diagnostic in ITER-like tungsten mono-block divertor on experimental advanced superconducting tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, J. C.; Jia, M. N.; Feng, W.

2016-08-15

In order to withstand rapid increase in particle and power impact onto the divertor and demonstrate the feasibility of the ITER design under long pulse operation, the upper divertor of the EAST tokamak has been upgraded to actively water-cooled, ITER-like tungsten mono-block structure since the 2014 campaign, which is the first attempt for ITER on the tokamak devices. Therefore, a new divertor Langmuir probe diagnostic system (DivLP) was designed and successfully upgraded on the tungsten divertor to obtain the plasma parameters in the divertor region such as electron temperature, electron density, particle and heat fluxes. More specifically, two identical triplemore » probe arrays have been installed at two ports of different toroidal positions (112.5-deg separated toroidally), which can provide fundamental data to study the toroidal asymmetry of divertor power deposition and related 3-dimension (3D) physics, as induced by resonant magnetic perturbations, lower hybrid wave, and so on. The shape of graphite tip and fixed structure of the probe are designed according to the structure of the upper tungsten divertor. The ceramic support, small graphite tip, and proper connector installed make it possible to be successfully installed in the very narrow interval between the cassette body and tungsten mono-block, i.e., 13.5 mm. It was demonstrated during the 2014 and 2015 commissioning campaigns that the newly upgraded divertor Langmuir probe diagnostic system is successful. Representative experimental data are given and discussed for the DivLP measurements, then proving its availability and reliability.« less

Real-time control of divertor detachment in H-mode with impurity seeding using Langmuir probe feedback in JET-ITER-like wall

NASA Astrophysics Data System (ADS)

Guillemaut, C.; Lennholm, M.; Harrison, J.; Carvalho, I.; Valcarcel, D.; Felton, R.; Griph, S.; Hogben, C.; Lucock, R.; Matthews, G. F.; Perez Von Thun, C.; Pitts, R. A.; Wiesen, S.; contributors, JET

2017-04-01

Burning plasmas with 500 MW of fusion power on ITER will rely on partially detached divertor operation to keep target heat loads at manageable levels. Such divertor regimes will be maintained by a real-time control system using the seeding of radiative impurities like nitrogen (N), neon or argon as actuator and one or more diagnostic signals as sensors. Recently, real-time control of divertor detachment has been successfully achieved in Type I ELMy H-mode JET-ITER-like wall discharges by using saturation current (I sat) measurements from divertor Langmuir probes as feedback signals to control the level of N seeding. The degree of divertor detachment is calculated in real-time by comparing the outer target peak I sat measurements to the peak I sat value at the roll-over in order to control the opening of the N injection valve. Real-time control of detachment has been achieved in both fixed and swept strike point experiments. The system has been progressively improved and can now automatically drive the divertor conditions from attached through high recycling and roll-over down to a user-defined level of detachment. Such a demonstration is a successful proof of principle in the context of future operation on ITER which will be extensively equipped with divertor target probes.

Impurity seeding for tokamak power exhaust: from present devices via ITER to DEMO

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Dux, R.; Casali, L.; Eich, T.; Giannone, L.; Herrmann, A.; McDermott, R.; Mlynek, A.; Müller, H. W.; Reimold, F.; Schweinzer, J.; Sertoli, M.; Tardini, G.; Treutterer, W.; Viezzer, E.; Wenninger, R.; Wischmeier, M.; the ASDEX Upgrade Team

2013-12-01

A future fusion reactor is expected to have all-metal plasma facing materials (PFMs) to ensure low erosion rates, low tritium retention and stability against high neutron fluences. As a consequence, intrinsic radiation losses in the plasma edge and divertor are low in comparison to devices with carbon PFMs. To avoid localized overheating in the divertor, intrinsic low-Z and medium-Z impurities have to be inserted into the plasma to convert a major part of the power flux into radiation and to facilitate partial divertor detachment. For burning plasma conditions in ITER, which operates not far above the L-H threshold power, a high divertor radiation level will be mandatory to avoid thermal overload of divertor components. Moreover, in a prototype reactor, DEMO, a high main plasma radiation level will be required in addition for dissipation of the much higher alpha heating power. For divertor plasma conditions in present day tokamaks and in ITER, nitrogen appears most suitable regarding its radiative characteristics. If elevated main chamber radiation is desired as well, argon is the best candidate for the simultaneous enhancement of core and divertor radiation, provided sufficient divertor compression can be obtained. The parameter Psep/R, the power flux through the separatrix normalized by the major radius, is suggested as a suitable scaling (for a given electron density) for the extrapolation of present day divertor conditions to larger devices. The scaling for main chamber radiation from small to large devices has a higher, more favourable dependence of about Prad,main/R2. Krypton provides the smallest fuel dilution for DEMO conditions, but has a more centrally peaked radiation profile compared to argon. For investigation of the different effects of main chamber and divertor radiation and for optimization of their distribution, a double radiative feedback system has been implemented in ASDEX Upgrade (AUG). About half the ITER/DEMO values of Psep/R have been achieved so far, and close to DEMO values of Prad,main/R2, albeit at lower Psep/R. Further increase of this parameter may be achieved by increasing the neutral pressure or improving the divertor geometry.

The influence of Filaments in the Private Flux Region on Divertor Power and Particle Deposition

NASA Astrophysics Data System (ADS)

Harrison, James

2014-10-01

Recent advances in imaging of the MAST divertor have revealed, for the first time, evidence for filaments in the private flux region (PFR). Detailed analysis of the image data shows 3 distinct types of fluctuations occurring within the divertor volume: highly sheared filaments in the SOL originating from the outer midplane, high frequency (>50 kHz) filaments near the separatrix of the outer divertor leg and filaments in the private flux region originating from inner divertor leg. With the need to extrapolate divertor performance from existing machines to future devices, these observations can contribute to our quantitative understanding of transport in the PFR. In particular, they suggest that transport in the PFR is, at least in part, driven by turbulence, which may not be well captured by the Eich/Wagner description of the divertor footprint, expressed in terms of exponential decay in space above the X-point and Gaussian spreading below the X-point. The PFR filaments are observed to move largely parallel with the flux surfaces in a way equivalent to a toroidal angular velocity of order 2 ×104 rad/s in H-mode, and slower by a factor of order 2 in L-mode. During their transit parallel to the flux surfaces across the PFR, the filaments eject plasma in bursts, away from the separatrix, deeper into the private flux region. Correlation analysis suggests that they are generated by processes local to the inner divertor leg, as there is a weak correlation between fluctuations in the SOL and PFR above what is expected from line integration effects. Scaling of filament properties with machine operating parameters, such as plasma current, density and auxiliary heating power will be presented, together with a comparison with data from divertor Langmuir probes and IR thermography to estimate the role PFR filaments play in determining the width of the divertor footprint.

An innovative small angle slot divertor concept for long pulse advanced tokamaks

NASA Astrophysics Data System (ADS)

Guo, Houyang

2017-10-01

A new Small Angle Slot (SAS) divertor is being developed in DIII-D to address the challenge of efficient divertor heat dispersal at the relatively low plasma density required for non-inductive current drive in future advanced tokamaks. SAS features a small incident angle near the plasma strike point on the divertor target plate with a progressively opening slot. SOLPS (B2-Eirene) edge code analysis finds that SAS can achieve strong plasma cooling when the strike point is placed near the small angle target plate in the slot, leading to low electron temperature Te across the entire divertor target. This is enabled by strong coupling between a gas tight slot and directed neutral recycling by the small angle target to enhance neutral buildup near the target. SOLPS analysis reveals a strong correlation between Te and D2 density at the target for various divertor configurations including the flat target, slanted target, and lower single null divertor. The strong correlation suggests that achievement of low Te may reduce essentially to identifying the divertor baffle geometry that achieves the highest target gas density at a given upstream condition. The SAS divertor concept has recently been tested in DIII-D for a range of plasma configurations and conditions with precise control of slot strike point location. In confirmation of SOLPS predictions, a sharp transition is observed when the strike point is moved to the critical outer corner of SAS. A set of Langmuir probes imbedded in SAS show that the Te radial profile, which is peaked at the strike point when it is located away from the SAS corner, becomes low across the target when the strike point is located near the corner. With further increase in density, deep-slot detachment occurs with Te 1 eV, measured by the unique DIII-D divertor Thomson Scattering diagnostic. Work supported by US DOE under DE-FC02-04ER54698.

Exploration of magnetic perturbation effects on advanced divertor configurations in NSTX-U

DOE PAGES

Frerichs, H.; Schmitz, O.; Waters, I.; ...

2016-06-01

The control of divertor heat loads - both steady state and transient - remains a key challenge for the successful operation of ITER and FNSF. Magnetic perturbations provide a promising technique to control ELMs (transients), but understanding their detailed impact is difficult due to their symmetry breaking nature. One approach for reducing steady state heat loads are so called 'advanced divertors' which aim at optimizing the magnetic field configuration: the snowflake and the (super-)X-divertor. It is likely that both concepts - magnetic perturbations and advanced divertors - will have to work together, and we explore their inter- action based onmore » the NSTX-U setup. An overview of different divertor con gurations under the impact of magnetic perturbations is presented, and the resulting impact on plasma edge transport is investigated with the EMC3-EIRENE code. Variations in size of the magnetic footprint of the perturbed separatrix are found, which is related to the level of flux expansion on the divertor target. Non-axisymmetric peaking of the heat flux related to the perturbed separatrix is found at the outer strike point, but only in locations where flux expansion is not too large.« less

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

NASA Astrophysics Data System (ADS)

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.

Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

NASA Astrophysics Data System (ADS)

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.

Favorable effects of turbulent plasma mixing on the performance of innovative tokamak divertors

NASA Astrophysics Data System (ADS)

Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.

2013-10-01

The problem of reducing the heat load on plasma-facing components is one of the most demanding issues for MFE devices. The general approach to the solution of this problem is the use of a specially configured poloidal magnetic field, so called magnetic divertors. In recent years, novel divertors possessing the 2-nd and 3-rd order nulls of the poloidal field (PF) have been proposed. They are called a ``snowflake'' (SF) and a ``cloverleaf'' (CL) divertor, respectively, due to characteristic shape of the magnetic separatrix. Among several beneficial features of such divertors is an effect of strong turbulent plasma mixing that is intrinsic to the zone of weak PF near the null-point. The turbulence spreads the heat flux between multiple divertor exhaust channels and increases the heat flux width within each channel. Among physical processes affecting the onset of convection the curvature-driven mode of axisymmetric rolls is most prominent. The effect is quite significant for the SF and is even stronger for the CL divertor. Projections to future ITER-scale facilities are discussed. Work performed for U.S. DoE by LLNL under Contract DE-AC52-07NA27344.

Design of ITER divertor VUV spectrometer and prototype test at KSTAR tokamak

NASA Astrophysics Data System (ADS)

Seon, Changrae; Hong, Joohwan; Song, Inwoo; Jang, Juhyeok; Lee, Hyeonyong; An, Younghwa; Kim, Bosung; Jeon, Taemin; Park, Jaesun; Choe, Wonho; Lee, Hyeongon; Pak, Sunil; Cheon, MunSeong; Choi, Jihyeon; Kim, Hyeonseok; Biel, Wolfgang; Bernascolle, Philippe; Barnsley, Robin; O'Mullane, Martin

2017-12-01

Design and development of the ITER divertor VUV spectrometer have been performed from the year 1998, and it is planned to be installed in the year 2027. Currently, the design of the ITER divertor VUV spectrometer is in the phase of detail design. It is optimized for monitoring of chord-integrated VUV signals from divertor plasmas, chosen to contain representative lines emission from the tungsten as the divertor material, and other impurities. Impurity emission from overall divertor plasmas is collimated through the relay optics onto the entrance slit of a VUV spectrometer with working wavelength range of 14.6-32 nm. To validate the design of the ITER divertor VUV spectrometer, two sets of VUV spectrometers have been developed and tested at KSTAR tokamak. One set of spectrometer without the field mirror employs a survey spectrometer with the wavelength ranging from 14.6 nm to 32 nm, and it provides the same optical specification as the spectrometer part of the ITER divertor VUV spectrometer system. The other spectrometer with the wavelength range of 5-25 nm consists of a commercial spectrometer with a concave grating, and the relay mirrors with the same geometry as the relay mirrors of the ITER divertor VUV spectrometer. From test of these prototypes, alignment method using backward laser illumination could be verified. To validate the feasibility of tungsten emission measurement, furthermore, the tungsten powder was injected in KSTAR plasmas, and the preliminary result could be obtained successfully with regard to the evaluation of photon throughput. Contribution to the Topical Issue "Atomic and Molecular Data and their Applications", edited by Gordon W.F. Drake, Jung-Sik Yoon, Daiji Kato, Grzegorz Karwasz.

Development of a Method for Local Electron Temperature and Density Measurements in the Divertor of the JET Tokamak

NASA Technical Reports Server (NTRS)

Jupen, C.; Meigs, A.; Bhatia, A. K.; Brezinsek, S.; OMullane, M.

2004-01-01

Plasma volume recombination in the divertor, a process in which charged particles recombine to neutral atoms, contributes to plasma detachment and hence cooling at the divertor target region. Detachment has been observed at JET and other tokamaks and is known to occur at low electron temperatures (T(sub e)<1 eV) and at high electron density (n(sub e)>10(exp 20)/m(exp 3)). The ability to measure such low temperatures is therefore of interest for modelling the divertor. In present work we report development of a new spectroscopic technique for investigation of local electron density (n(sub e)) and temperature (T,) in the outer divertor at JET.

Divertor electron temperature and impurity diffusion measurements with a spectrally resolved imaging radiometer.

PubMed

Clayton, D J; Jaworski, M A; Kumar, D; Stutman, D; Finkenthal, M; Tritz, K

2012-10-01

A divertor imaging radiometer (DIR) diagnostic is being studied to measure spatially and spectrally resolved radiated power P(rad)(λ) in the tokamak divertor. A dual transmission grating design, with extreme ultraviolet (~20-200 Å) and vacuum ultraviolet (~200-2000 Å) gratings placed side-by-side, can produce coarse spectral resolution over a broad wavelength range covering emission from impurities over a wide temperature range. The DIR can thus be used to evaluate the separate P(rad) contributions from different ion species and charge states. Additionally, synthetic spectra from divertor simulations can be fit to P(rad)(λ) measurements, providing a powerful code validation tool that can also be used to estimate electron divertor temperature and impurity transport.

Electron pressure balance in the SOL through the transition to detachment

DOE PAGES

McLean, A. G.; Leonard, A. W.; Makowski, M. A.; ...

2015-02-07

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of P inj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure T e down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurementsmore » of n e, T e and p e in the divertor versus Lparallel demonstrate a rapid transition from Te ≥ 15 eV to ≤3 eV occurring both at the outer strike point and upstream of the X-point. Furthermore, these observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.« less

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE PAGES

Covele, Brent; Kotschenreuther, M.; Mahajan, S.; ...

2017-06-23

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Electron pressure balance in the SOL through the transition to detachment

NASA Astrophysics Data System (ADS)

McLean, A. G.; Leonard, A. W.; Makowski, M. A.; Groth, M.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Briesemeister, A. R.; Carlstrom, T. N.; Eldon, D.; Fenstermacher, M. E.; Hill, D. N.; Lasnier, C. J.; Liu, C.; Osborne, T. H.; Petrie, T. W.; Soukhanovskii, V. A.; Stangeby, P. C.; Tsui, C.; Unterberg, E. A.; Watkins, J. G.

2015-08-01

Upgrades to core and divertor Thomson scattering (DTS) diagnostics at DIII-D have provided measurements of electron pressure profiles in the lower divertor from attached- to fully-detached divertor plasma conditions. Detailed, multistep sequences of discharges with increasing line-averaged density were run at several levels of Pinj. Strike point sweeping allowed 2D divertor characterization using DTS optimized to measure Te down to 0.5 eV. The ionization front at the onset of detachment is found to move upwards in a controlled manner consistent with the indication that scrape-off layer parallel power flux is converted from conducted to convective heat transport. Measurements of ne, Te and pe in the divertor versus Lparallel demonstrate a rapid transition from Te ⩾ 15 eV to ⩽3 eV occurring both at the outer strike point and upstream of the X-point. These observations provide a strong benchmark for ongoing modeling of divertor detachment for existing and future tokamak devices.

Increased heat dissipation with the X-divertor geometry facilitating detachment onset at lower density in DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Impurity ion flow and temperature measured in a detached divertor with externally applied non-axisymmetric fields on DIII-D

DOE PAGES

Briesemeister, A. R.; Isler, R. C.; Allen, S. L.; ...

2014-11-15

In this study, externally applied non-axisymmetric magnetic fields are shown to have little effect on the impurity ion flow velocity and temperature as measured by the multichord divertor spectrometer in the DIII-D divertor for both attached and detached conditions. These experiments were performed in H-mode plasmas with the grad-B drift toward the target plates, with and without n = 3 resonant magnetic perturbations (RMPs). The flow velocity in the divertor is shown to change by as much as 30% when deuterium gas puffing is used to create detachment of the divertor plasma. No measurable changes in the C III flowmore » were observed in response to the RMP fields for the conditions used in this work. Images of the C III emission are used along with divertor Thomson scattering to show that the local electron and C III temperatures are equilibrated for the conditions shown.« less

Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

NASA Astrophysics Data System (ADS)

Masuzaki, S.; Tokitani, M.; Otsuka, T.; Oya, Y.; Hatano, Y.; Miyamoto, M.; Sakamoto, R.; Ashikawa, N.; Sakurada, S.; Uemura, Y.; Azuma, K.; Yumizuru, K.; Oyaizu, M.; Suzuki, T.; Kurotaki, H.; Hamaguchi, D.; Isobe, K.; Asakura, N.; Widdowson, A.; Heinola, K.; Jachmich, S.; Rubel, M.; contributors, JET

2017-12-01

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

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

NASA Astrophysics Data System (ADS)

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.

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

NASA Astrophysics Data System (ADS)

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.

The dependence of divertor power sharing on magnetic flux balance in near double-null configurations on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Brunner, D.; Kuang, A. Q.; LaBombard, B.; Terry, J. L.

2018-07-01

Management of power exhaust will be a crucial task for tokamak fusion reactors. Reactor concepts are often proposed with double-null divertors, i.e. having two magnetic separatrices in an up-down symmetric configuration. This arrangement is potentially advantageous since the majority of the tokamak exhaust power tends to flow to the outer pair of divertor legs at large major radius, where the geometry is favorable for spreading the heat over a large surface area and there is more room for advanced divertor configurations. Despite the importance, there have been relatively few studies of divertor power sharing in near double null configurations and no studies at the poloidal magnetic fields and scrape-off layer power widths anticipated for a reactor. Motivated by this need we have undertaken a systematic study on Alcator C-Mod, examining the effect of magnetic flux balance on the power sharing among the four divertor legs in near double-null plasmas. Ohmic L-modes at three values of plasma current and ICRF-heated enhanced D-alpha (EDA) H-modes and I-modes at a single value of plasma current are explored, producing poloidal magnetic fields of 0.42, 0.62 and 0.85 Tesla. For Ohmic L-modes and ICRF-heated EDA H-modes, we find that the point of equal power sharing between upper and lower divertors occurs remarkably close to a balanced double null. Power sharing amongst the outer (upper versus lower) and inner (upper versus lower) pairs of divertors can be described in terms of a logistic function of magnetic flux balance, consistent with heat flux mapping along magnetic field lines to the outer midplane. Power sharing between inner and outer legs is found to follow a Gaussian-like function of magnetic flux balance with non-zero power to the inner divertors at double null. The overall behavior of H-modes operated near double null and for I-modes operating to within one heat flux e-folding of double null are found similar to Ohmic L-modes, with a significant reduction of power on the inner divertor legs. The results are encapsulated in terms of empirically-informed analytic functions of magnetic flux balance. When combined with magnetic equilibrium control system specifications, these relationships can be used to specify the power flux handling requirements for each of the four divertor target plates.

Basic physical processes and reduced models for plasma detachment

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2018-04-01

The divertor of a tokamak reactor will have to satisfy a number of critical constraints, the first of which is that the divertor targets not fail due to excessive heating or sputter-erosion. This paramount constraint of target survival defines the operating window for the principal plasma properties at the divertor target, the density n t and temperature, T t. In particular T et < 10 eV is shown to be required. Code and experimental studies show that the pressure–momentum loss by the plasma that occurs along flux tubes in the edge, between the divertor entrance and target, (i) correlates strongly with T et, and (ii) begins to increase as T et falls below 10 eV, becoming very strong by 1 eV. The transition between the high-recycling regime and the detached divertor regime has therefore been defined here to occur when T et < 10 eV. Simple analytic models are developed (i) to relate (T t, n t) to the controlling conditions ‘upstream’ e.g. at the divertor entrance, and (ii) in turn to relate (T t, n t) to other important divertor quantities including (a) the required level of radiative cooling in the divertor, and (b) the ion flux to the target in the presence of volumetric loss of particles, momentum and power in the divertor. The 2 Point Model, 2PM, is a widely used analytic model for relating (T t, n t) to the controlling upstream conditions. The 2PM is derived here for various levels of complexity regarding the effects included. Analytic models of divertor detachment provide valuable insight and useful approximations, but more complete modeling requires the use of edge codes such as EDGE2D, SOLPS, SONIC, UEDGE, etc. Edge codes have grown to become quite sophisticated and now constitute, in effect, ‘code-experiments’ that—just as for actual experiments—can benefit from interpretation in terms of simple conceptual frameworks. 2 Point Model Formatting, 2PMF, of edge code output can provide such a conceptual framework. Methods of applying 2PMF are illustrated here with some examples.

Design, R&D and commissioning of EAST tungsten divertor

NASA Astrophysics Data System (ADS)

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.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

NASA Astrophysics Data System (ADS)

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; Allen, S. L.; Stangeby, P. C.; Thomas, D.; Unterberg, E. A.; Abrams, T.; Boedo, J.; Briesemeister, A. R.; Buchenauer, D.; Bykov, I.; Canik, J. M.; Chrobak, C.; Covele, B.; Ding, R.; Doerner, R.; Donovan, D.; Du, H.; Elder, D.; Eldon, D.; Lasa, A.; Groth, M.; Guterl, J.; Jarvinen, A.; Hinson, E.; Kolemen, E.; Lasnier, C. J.; Lore, J.; Makowski, M. A.; McLean, A.; Meyer, B.; Moser, A. L.; Nygren, R.; Owen, L.; Petrie, T. W.; Porter, G. D.; Rognlien, T. D.; Rudakov, D.; Sang, C. F.; Samuell, C.; Si, H.; Schmitz, O.; Sontag, A.; Soukhanovskii, V.; Wampler, W.; Wang, H.; Watkins, J. G.

2016-12-01

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Developing and validating advanced divertor solutions on DIII-D for next-step fusion devices

DOE PAGES

Guo, H. Y.; Hill, D. N.; Leonard, A. W.; ...

2016-09-14

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, whichmore » we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). In conclusion, this paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.« less

Impurity-induced divertor plasma oscillations

DOE PAGES

Smirnov, R. D.; Kukushkin, A. S.; Krasheninnikov, S. I.; ...

2016-01-07

Two different oscillatory plasma regimes induced by seeding the plasma with high- and low-Z impurities are found for ITER-like divertor plasmas, using computer modeling with the DUSTT/UEDGE and SOLPS4.3 plasma-impurity transport codes. The oscillations are characterized by significant variations of the impurity-radiated power and of the peak heat load on the divertor targets. Qualitative analysis of the divertor plasma oscillations reveals different mechanisms driving the oscillations in the cases of high- and low-Z impurity seeding. The oscillations caused by the high-Z impurities are excited near the X-point by an impurity-related instability of the radiation-condensation type, accompanied by parallel impurity ionmore » transport affected by the thermal and plasma friction forces. The driving mechanism of the oscillations induced by the low-Z impurities is related to the cross-field transport of the impurity atoms, causing alteration between the high and low plasma temperature regimes in the plasma recycling region near the divertor targets. As a result, the implications of the impurity-induced plasma oscillations for divertor operation in the next generation tokamaks are also discussed.« less

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

NASA Astrophysics Data System (ADS)

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.

Modelling of Divertor Detachment in MAST Upgrade

NASA Astrophysics Data System (ADS)

Moulton, David; Carr, Matthew; Harrison, James; Meakins, Alex

2017-10-01

MAST Upgrade will have extensive capabilities to explore the benefits of alternative divertor configurations such as the conventional, Super-X, x divertor, snowflake and variants in a single device with closed divertors. Initial experiments will concentrate on exploring the Super-X and conventional configurations, in terms of power and particle loads to divertor surfaces, access to detachment and its control. Simulations have been carried out with the SOLPS5.0 code validated against MAST experiments. The simulations predict that the Super-X configuration has significant advantages over the conventional, such as lower detachment threshold (2-3x lower in terms of upstream density and 4x higher in terms of PSOL). Synthetic spectroscopy diagnostics from these simulations have been created using the Raysect ray tracing code to produce synthetic filtered camera images, spectra and foil bolometer data. Forward modelling of the current set of divertor diagnostics will be presented, together with a discussion of future diagnostics and analysis to improve estimates of the plasma conditions. Work supported by the RCUK Energy Programme [Grant Number EP/P012450/1] and EURATOM.

Thermal Fatigue Study on the Divertor Plate Materials

NASA Astrophysics Data System (ADS)

Wu, Ji-hong; Zhang, Fu; Xu, Zeng-yu; Yan, Jian-cheng

2002-10-01

Thermal fatigue property of the divertor plate is one of the key issues that governs the lifetime of the divertor plate. Taking tungsten as surface material, a small-mock-up divertor plate was made by hot isostatic press welding (HIP). A thermal cycling experiment for divertor mock-up was carried out in the vacuum, where a high-heat-flux electronic gun was used as the thermal source. A cyclic heat flux of 9 MW/m2 was loaded onto the mock-up, a heating duration of 20 s was selected, the cooling water flow rate was 80 ml/s. After 1000 cycles, the surface and the W/Cu joint of the mock-up did not show any damage. The SEM was used to analyze the microstructure of the welding joint, where no cracks were found also.

Modelling of 13CH4 injection and local carbon deposition at the outer divertor of ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Aho-Mantila, L.; Airila, M. I.; Wischmeier, M.; Krieger, K.; Pugno, R.; Coster, D. P.; Chankin, A. V.; Neu, R.; Rohde, V.

2009-12-01

Numerical modelling of 13CH4 injection into the outer divertor plasma of the full tungsten, vertical target of ASDEX Upgrade is presented. The SOLPS5.0 code package is used to calculate a realistic scrape-off layer plasma background corresponding to L-mode discharges in the attached divertor plasma regime. The ERO code is then used for detailed modelling of the hydrocarbon break-up, re-deposition and re-erosion processes. The deposition patterns observed at two different poloidal locations are shown to strongly reflect the cross-field gradients in divertor plasma density and temperature, as well as the local plasma collisionality. Experimental results with forward and reversed BT, accompanied by numerical modelling, also point towards a significant poloidal hydrocarbon E×B drift in the divertor region.

Conceptual design of a divertor Thomson scattering diagnostic for NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L.

2014-11-15

A conceptual design for a divertor Thomson scattering (DTS) diagnostic has been developed for the NSTX-U device to operate in parallel with the existing multipoint Thomson scattering system. Higher projected peak heat flux in NSTX-U will necessitate application of advanced magnetics geometries and divertor detachment. Interpretation and modeling of these divertor scenarios will depend heavily on local measurement of electron temperature, T{sub e}, and density, n{sub e}, which DTS provides in a passive manner. The DTS design for NSTX-U adopts major elements from the successful DIII-D DTS system including 7-channel polychromators measuring T{sub e} to 0.5 eV. If implemented onmore » NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.« less

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

NASA Astrophysics Data System (ADS)

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.

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE PAGES

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.; ...

2017-06-22

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

Effect of 3D magnetic perturbations on divertor conditions and detachment in tokamak and stellarator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

DTT: a divertor tokamak test facility for the study of the power exhaust issues in view of DEMO

NASA Astrophysics Data System (ADS)

Albanese, R.; WPDTT2 Team; DTT Project Proposal Contributors, the

2017-01-01

In parallel with the programme to optimize the operation with a conventional divertor based on detached conditions to be tested on the ITER device, a project has been launched to investigate alternative power exhaust solutions for DEMO, aimed at the definition and the design of a divertor tokamak test facility (DTT). The DTT project proposal refers to a set of parameters selected so as to have edge conditions as close as possible to DEMO, while remaining compatible with DEMO bulk plasma performance in terms of dimensionless parameters and given constraints. The paper illustrates the DTT project proposal, referring to a 6 MA plasma with a major radius of 2.15 m, an aspect ratio of about 3, an elongation of 1.6-1.8, and a toroidal field of 6 T. This selection will guarantee sufficient flexibility to test a wide set of divertor concepts and techniques to cope with large heat loads, including conventional tungsten divertors; liquid metal divertors; both conventional and advanced magnetic configurations (including single null, snow flake, quasi snow flake, X divertor, double null); internal coils for strike point sweeping and control of the width of the scrape-off layer in the divertor region; and radiation control. The Poloidal Field system is planned to provide a total flux swing of more than 35 Vs, compatible with a pulse length of more than 100 s. This is compatible with the mission of studying the power exhaust problem and is obtained using superconducting coils. Particular attention is dedicated to diagnostics and control issues, especially those relevant for plasma control in the divertor region, designed to be as compatible as possible with a DEMO-like environment. The construction is expected to last about seven years, and the selection of an Italian site would be compatible with a budget of 500 M€.

X-Divertors on ITER - with no hardware changes

NASA Astrophysics Data System (ADS)

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

2014-10-01

Using CORSICA, we have discovered that X-Divertor (XD) equilibria are possible on ITER - without any extra PF coils inside the TF coils, and with no changes to ITER's poloidal field (PF) coil set, divertor cassette, strike points, or first wall. Starting from the Standard Divertor (SD), a sequence of XD configurations (with increasing flux expansions at the divertor plate) can be made by reprogramming ITER PF coil currents while keeping them all under their design limits (Lackner and Zohm have shown this to be impossible for Snowflakes). The strike point is held fixed, so no changes in the divertor or pumping hardware will be needed. The main plasma shape is kept very close to the SD case, so no hardware changes to the main chamber will be needed. Time-dependent ITER-XD operational scenarios are being checked using TSC. This opens the possibility that many XDs could be tested and used to assist in high-power operation on ITER. Because of the toroidally segmented ITER divertor plates, strongly detached operation may be critical for making use of the largest XD flux expansion possible. The flux flaring in XDs is expected to increase the stability of detachment, so that H-mode confinement is not affected. Detachment stability is being examined with SOLPS. This work supported by US DOE Grants DE-FG02-04ER54742 and DE-FG02-04ER54754 and by TACC at UT Austin.

Impact of the plasma geometry on divertor power exhaust: experimental evidence from TCV and simulations with SolEdge2D and TOKAM3X

NASA Astrophysics Data System (ADS)

Gallo, A.; Fedorczak, N.; Elmore, S.; Maurizio, R.; Reimerdes, H.; Theiler, C.; Tsui, C. K.; Boedo, J. A.; Faitsch, M.; Bufferand, H.; Ciraolo, G.; Galassi, D.; Ghendrih, P.; Valentinuzzi, M.; Tamain, P.; the EUROfusion MST1 Team; the TCV Team

2018-01-01

A deep understanding of plasma transport at the edge of magnetically confined fusion plasmas is needed for the handling and control of heat loads on the machine first wall. Experimental observations collected on a number of tokamaks over the last three decades taught us that heat flux profiles at the divertor targets of X-point configurations can be parametrized by using two scale lengths for the scrape-off layer (SOL) transport, separately characterizing the main SOL ({λ }q) and the divertor SOL (S q ). In this work we challenge the current interpretation of these two scale lengths as well as their dependence on plasma parameters by studying the effect of divertor geometry modifications on heat exhaust in the Tokamak à Configuration Variable. In particular, a significant broadening of the heat flux profiles at the outer divertor target is diagnosed while increasing the length of the outer divertor leg in lower single null, Ohmic, L-mode discharges. Efforts to reproduce this experimental finding with both diffusive (SolEdge2D-EIRENE) and turbulent (TOKAM3X) modelling tools confirm the validity of a diffusive approach for simulating heat flux profiles in more traditional, short leg, configurations while highlighting the need of a turbulent description for modified, long leg, ones in which strongly asymmetric divertor perpendicular transport develops.

Flow reversal, convection, and modeling in the DIII-D divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

1998-12-01

Measurements of the parallel Mach number of background plasma in the DIII-D tokamak divertor [M. A. Mahdavi {ital et al.} in {ital Proceedings, 16th International Conference}, Montreal, 1996 (International Atomic Energy Agency, Vienna, 1997) Vol. I, p. 397] were performed using a fast scanning Mach probe. The parallel particle flow shows evidence of complex behavior such as reverse flow, i.e., flow away from the target plate, stagnant flow, and large scale convection. For detached discharges, measurements confirm predictions of convective flow towards the divertor target plate at near sound speed over large regions in the divertor. The resulting convected heatmore » flux is a dominant heat transport mechanism in the divertor. For attached discharges with high recycling, particle flow reversal in a thin region at or near the outer separatrix, thereby confirming the existence of a mechanism by which impurities can be transported away from the divertor target plates. Modeling results from the two-dimensional fluid code UEDGE [G. D. Porter and the DIII-D Team, {open_quotes}Divertor characterization experiments and modelling in DIII-D,{close_quotes} in {ital Proceedings of the 23rd European Conference on Controlled Fusion and Plasma Physics}, 24{endash}28 June 1996, Kiev, Ukraine (European Physical Society, Petit-Lancy, Switzerland, 1996), Vol. 20C, Part II, p. 699] can reproduce the main features of the experimental observations. {copyright} {ital 1998 American Institute of Physics.}« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, M.; Jaworski, M. A.; Kaita, R.

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. 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-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Covele, Brent; Kotschenreuther, M.; Mahajan, S.

The X-Divertor geometry on DIII-D has demonstrated reduced particle and heat fluxes to the target, facilitating detachment onset at ~20% lower upstream density and higher H-mode pedestal pressure than a standard divertor. SOLPS modeling suggests that this effect cannot be explained by an increase in total connection length alone, but rather by the addition of connection length specifically in the power-dissipating volume near the target, via poloidal flux expansion and flaring. But, poloidal flaring must work synergistically with divertor closure to most effectively reduce the detachment density threshold. Furthermore, the model also points to carbon radiation as the primary drivermore » of power dissipation in divertors on the DIII-D floor, which is consistent with experimental observations. Sustainable divertor detachment at lower density has beneficial consequences for energy confinement and current drive efficiency in the core for advanced tokamak (AT) operation, while simultaneously satisfying the exhaust requirements of the plasma-facing components.« less

Status of National Spherical Torus Experiment Liquid Lithium Divertor

NASA Astrophysics Data System (ADS)

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.

Gyrokinetic projection of the divertor heat-flux width from present tokamaks to ITER

DOE PAGES

Chang, Choong Seock; Ku, Seung -Hoe; Loarte, Alberto; ...

2017-07-11

Here, the XGC1 edge gyrokinetic code is used to study the width of the heat-flux to divertor plates in attached plasma condition. The flux-driven simulation is performed until an approximate power balance is achieved between the heat-flux across the steep pedestal pressure gradient and the heat-flux on the divertor plates.

Innovative divertor concept development on DIII-D and EAST

DOE PAGES

Guo, H. Y.; Allen, S.; Canik, J.; ...

2016-06-02

A critical issue facing the design and operation of next-step high-power steady-state fusion devices is the control of heat fluxes and erosion at the plasma-facing components, in particular, the divertor target plates. A new initiative has been launched on DIII-D to develop and demonstrate innovative boundary plasma-materials interface solutions. The central purposes of this new initiative are to advance scientific understanding in this critical area and develop an advanced divertor concept for application to next-step fusion devices. Finally, DIII-D will leverage strong collaborative efforts on the EAST superconducting tokamak for extending integrated high performance advanced divertor solutions to true steady-state.

Enhanced visible and near-infrared capabilities of the JET mirror-linked divertor spectroscopy system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lomanowski, B. A., E-mail: b.a.lomanowski@durham.ac.uk; Sharples, R. M.; Meigs, A. G.

2014-11-15

The mirror-linked divertor spectroscopy diagnostic on JET has been upgraded with a new visible and near-infrared grating and filtered spectroscopy system. New capabilities include extended near-infrared coverage up to 1875 nm, capturing the hydrogen Paschen series, as well as a 2 kHz frame rate filtered imaging camera system for fast measurements of impurity (Be II) and deuterium Dα, Dβ, Dγ line emission in the outer divertor. The expanded system provides unique capabilities for studying spatially resolved divertor plasma dynamics at near-ELM resolved timescales as well as a test bed for feasibility assessment of near-infrared spectroscopy.

Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soukhanovskii, V. A., E-mail: vlad@llnl.gov; Kaita, R.; Stratton, B.

2016-11-15

A radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T{sub e} estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPherson Modelmore » 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T{sub e}-dependent signal within a characteristic divertor detachment equilibration time of ∼10–15 ms is expected.« less

Real-time radiative divertor feedback control development for the NSTX-U tokamak using a vacuum ultraviolet spectrometer

DOE PAGES

Soukhanovskii, V. A.; Kaita, R.; Stratton, B.

2016-08-04

Here, a radiative divertor technique is planned for the NSTX-U tokamak to prevent excessive erosion and thermal damage of divertor plasma-facing components in H-mode plasma discharges with auxiliary heating up to 12 MW. In the radiative (partially detached) divertor, extrinsically seeded deuterium or impurity gases are used to increase plasma volumetric power and momentum losses. A real-time feedback control of the gas seeding rate is planned for discharges of up to 5 s duration. The outer divertor leg plasma electron temperature T e estimated spectroscopically in real time will be used as a control parameter. A vacuum ultraviolet spectrometer McPhersonmore » Model 251 with a fast charged-coupled device detector is developed for temperature monitoring between 5 and 30 eV, based on the Δn = 0, 1 line intensity ratios of carbon, nitrogen, or neon ion lines in the spectral range 300–1600 Å. A collisional-radiative model-based line intensity ratio will be used for relative calibration. A real-time T e-dependent signal within a characteristic divertor detachment equilibration time of ~10–15 ms is expected.« less

Tungsten migration in Alcator C-Mod: sputtering and melting

NASA Astrophysics Data System (ADS)

Wright, G. M.; Barnard, H.; Lipschultz, B.; Whyte, D. G.

2010-11-01

A row of bulk tungsten (W) tiles were installed near the typical outer strike-point location in the Alcator C-Mod divertor in 2007. In the 2009/2010 campaign, one of the W tiles mechanically failed resulting in significant W melting at that location. Post-campaign PIXE surface analysis has been used to observe tungsten (W) deposition and migration patterns in the divertor for the typical operations (sputtering only) and operation with melted components. For sputtering conditions, W deposition of up to 20 nm equivalent thickness is observed at various divertor surfaces indicating prompt re-deposition at the outer divertor, neutral and ion transport through the private-flux region and ion transport in the scrape off layer. For melting conditions, W deposition of up to 400 nm equivalent thickness is observed at some locations at the outer divertor. However, the toroidal distribution of W on the outer divertor is strongly non-uniform. There is no W deposition measured on the inner wall limiter. These results indicate that impurity migration is affected by the erosion mechanism and source, with the migration from melting being less predictable and uniform than from the sputtering case. Supported by USDoE award DE-SC00-02060.

Scrape off layer modelling studies for SST-I

NASA Astrophysics Data System (ADS)

Warrier, M.; Jaishankar, S.; Deshpande, S.; Coster, D.; Schneider, R.; Chaturvedi, S.; Srinivasan, R.; Braams, B. J.; SST Team

SOL modelling results for SST-1 (SST Team, Proceedings of the 16th IEEE/NPSS Symposium on Fusion Engineering, Champaign, IL, vol. II, 1995, p. 481) show a sheath limited flow regime. This is due to the low edge densities required by lower hybrid current drive (LHCD), coupled with high power input per unit volume. Coupled plasma-neutral transport studies using B2-Eirene [R. Schneider et al., J. Nucl. Mater. 196-198 (1992) 810] show significantly high charge exchange losses and radiated power from the core. It also shows that the heat flux to the inner divertor is higher than that to the outer divertor due to thinner inner SOL widths. The Monte-Carlo neutral transport code DEGAS [D. Heifitz et al., J. Comput. Phys. 46 (1982) 309] was used to optimise the baffle plate geometry and it was seen that a configuration where the baffle plate shields the main plasma from the divertor strike point results in reduced backflow of neutrals. The divertor erosion code DIVER (M. Warrier et al., SST Divertor Modelling Report, 1996-1997) was used to predict a steady state operating temperature for the SST divertor plate lying in the range 750-1000°C for which the erosion will be minimum.

Results from core-edge experiments in high Power, high performance plasmas on DIII-D

DOE PAGES

Petrie, T. W.; Fenstermacher, M. E.; Holcomb, C. T.; ...

2016-12-24

Here, significant challenges to reducing divertor heat flux in highly powered near-double null divertor (DND) hybrid plasmas, while still maintaining both high performance metrics and low enough density for application of RF heating, are identified. For these DNDs on DIII-D, the scaling of the peak heat flux at the outer target (q ⊥ P) ∝ [P SOL x I P] 0.92 for P SOL = 8-19 MW and I P = 1.0–1.4 MA, and is consistent with standard ITPA scaling for single-null H-mode plasmas. Two divertor heat flux reduction methods were tested. First, applying the puff-and-pump radiating divertor to DIII-Dmore » plasmas may be problematical at high power and H98 (≥ 1.5) due to improvement in confinement time with deuterium gas puffing which can lead to unacceptably high core density under certain conditions. Second, q ⊥ P for these high performance DNDs was reduced by ≈35% when an open divertor is closed on the common flux side of the outer divertor target (“semi-slot”) but also that heating near the slot opening is a significant source for impurity contamination of the core.« less

Electric field divertor plasma pump

DOEpatents

Schaffer, Michael J.

1994-01-01

An electric field plasma pump includes a toroidal ring bias electrode (56) positioned near the divertor strike point of a poloidal divertor of a tokamak (20), or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix (40) of the poloidal divertor contacts the ring electrode (56), which then also acts as a divertor plate. A plenum (54) or other duct near the electrode (56) includes an entrance aperture open to receive electrically-driven plasma. The electrode (56) is insulated laterally with insulators (63,64), one of which (64) is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode (56) and a vacuum vessel wall (22), with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E.times.B/B.sup.2 drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable.

Electric field divertor plasma pump

DOEpatents

Schaffer, M.J.

1994-10-04

An electric field plasma pump includes a toroidal ring bias electrode positioned near the divertor strike point of a poloidal divertor of a tokamak, or similar plasma-confining apparatus. For optimum plasma pumping, the separatrix of the poloidal divertor contacts the ring electrode, which then also acts as a divertor plate. A plenum or other duct near the electrode includes an entrance aperture open to receive electrically-driven plasma. The electrode is insulated laterally with insulators, one of which is positioned opposite the electrode at the entrance aperture. An electric field E is established between the ring electrode and a vacuum vessel wall, with the polarity of the bias applied to the electrode being relative to the vessel wall selected such that the resultant electric field E interacts with the magnetic field B already existing in the tokamak to create an E [times] B/B[sup 2] drift velocity that drives plasma into the entrance aperture. The pumped plasma flow into the entrance aperture is insensitive to variations, intentional or otherwise, of the pump and divertor geometry. Pressure buildups in the plenum or duct connected to the entrance aperture in excess of 10 mtorr are achievable. 11 figs.

ELM elimination with lithium aerosol injection in upper-single null discharges using the tungsten divertor in EAST

NASA Astrophysics Data System (ADS)

Sun, Z.; Maingi, R.; Hu, J.; Lunsford, R.; Diallo, A.; Tritz, K.; Osborne, T.; Canik, J.; Zuo, G.; Wang, L.; Xu, G.; Gong, X.; EAST Team Team

2017-10-01

A reproducible, fully non-inductive H-mode regime devoid of large ELMs has been achieved by continuous Li injection in EAST into the upper `ITER-like' tungsten divertor, extending previous results on the graphite divertor. These discharges did not suffer from density or impurity accumulation, and maintained constant core radiated power. The new results extend the energy confinement multiplier H98(y,2) 1.2, as compared to H98(y,2) 0.75 previously on the graphite divertor. The observed ELM elimination is correlated with a decrease in particle recycling, as expected from the strong Li coating before the experiment, and real-time Li aerosol injection. In addition, core W concentration was reduced during the Li injection. ELM elimination is likely related to the reduced recycling and density /temperature profile changes. A low-n electromagnetic coherent mode (MCM) at 40kHz became stronger in amplitude and also more coherent. The MCM shows strong magnetic fluctuations as measured by fast Mirnov coils, but weak density fluctuations. As compared to the graphite divertor, Li injection into the tungsten divertor eliminated ELMs at twice the previous auxiliary heating power, and reduced pedestal collisionality.

Plasma-wall interactions in ITER

NASA Astrophysics Data System (ADS)

Parker, R.; Janeschitz, G.; Pacher, H. D.; Post, D.; Chiocchio, S.; Federici, G.; Ladd, P.; Iter Joint Central Team; Home Teams

1997-02-01

This paper reviews the status of the design of the divertor and first-wall/shield, the main in-vessel components for ITER. Under nominal ignited conditions, 300 MW of alpha power will be produced and must be removed from the divertor and first-wall. Additional power from auxiliary sources up to the level of 100 MW must also be removed in the case of driven burns. In the ignited case, about 100 MW will be radiated to the first wall as bremsstrahlung. Allowing the remaining power to be conducted to the divertor target plates would result in excessive heat fluxes. The power handling strategy is to radiate an additional 100-150 MW in the SOL and the divertor channel via a combination of radiation from hydrogen, and intrinsic and seeded impurities. Vertical targets have been adopted for the baseline divertor configuration. This geometry promotes partial detachment, as found in present experiments and in the results of modelling runs for ITER conditions, and power densities on the target plates can be ≤ 5 MW/ m2. Such regimes promote relatively high pressure (> 1 Pa) in the divertor and even with a low helium enrichment factor of 0.2, the required pumping speed to pump helium is ≤ 50 m3/ s. An important physics question is the quality of core confinement in these attractive divertor regimes. In addition to power and particle handling issues, the effects of disruptions play a major role in the design and performance of in-vessel components. Both centered disruptions and VDE's produce stresses in the first-wall/shield modules, backplate and the divertor wings and cassettes that are near or even somewhat in excess of allowables for normal operation. Also plasma-wall contact from disruptions, including at the divertor target, together with material properties are major factors determining component lifetime. Considering the potential for impurity contamination and minimizing tritium inventory as well as thermomechanical performance, the present material selection calls for carbon divertor targets near the strike point, tungsten on the rest of the target and on the baffle where the charge-exchange flux could be high, and beryllium elsewhere. All three materials and relevant joining techniques are being developed in the R&D program and the final selection for the first assembly will be made at the end of the EDA.

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

NASA Astrophysics Data System (ADS)

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 of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

A Fast Visible Camera Divertor-Imaging Diagnostic on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roquemore, A; Maingi, R; Lasnier, C

2007-06-19

In recent campaigns, the Photron Ultima SE fast framing camera has proven to be a powerful diagnostic when applied to imaging divertor phenomena on the National Spherical Torus Experiment (NSTX). Active areas of NSTX divertor research addressed with the fast camera include identification of types of EDGE Localized Modes (ELMs)[1], dust migration, impurity behavior and a number of phenomena related to turbulence. To compare such edge and divertor phenomena in low and high aspect ratio plasmas, a multi-institutional collaboration was developed for fast visible imaging on NSTX and DIII-D. More specifically, the collaboration was proposed to compare the NSTX smallmore » type V ELM regime [2] and the residual ELMs observed during Type I ELM suppression with external magnetic perturbations on DIII-D[3]. As part of the collaboration effort, the Photron camera was installed recently on DIII-D with a tangential view similar to the view implemented on NSTX, enabling a direct comparison between the two machines. The rapid implementation was facilitated by utilization of the existing optics that coupled the visible spectral output from the divertor vacuum ultraviolet UVTV system, which has a view similar to the view developed for the divertor tangential TV camera [4]. A remote controlled filter wheel was implemented, as was the radiation shield required for the DIII-D installation. The installation and initial operation of the camera are described in this paper, and the first images from the DIII-D divertor are presented.« less

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

DOE PAGES

Ono, M.; Jaworski, M. A.; Kaita, R.; ...

2016-08-05

Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition to those issues, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety issues. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues while potentially improving the reactor plasma performance. 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-freemore » core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 500°C than the first wall ~ 600 – 700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover in tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory issue. With an expected T fraction of ≤ 0.7 %, an acceptable level of T inventory can be achieved. In NSTX-U, preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issue on the ARLLD concept. A LL-loop is also being prepared off line for prototyping future use on NSTX-U.« less

The snowflake divertor

DOE PAGES

Ryutov, D. D.; Soukhanovskii, V. A.

2015-11-17

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. One of the most interesting effects of the snowflake geometry is the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation tomore » the existing theoretical models is described. Divertor concepts utilizing the properties of a snowflake configuration are briefly discussed.« less

Divertor target for magnetic containment device

DOEpatents

Luzzi, Jr., Theodore E.

1982-01-01

In a plasma containment device of a type having superconducting field coils for magnetically shaping the plasma into approximately the form of a torus, an improved divertor target for removing impurities from a "scrape off" region of the plasma comprises an array of water cooled swirl tubes onto which the scrape off flux is impinged. Impurities reflected from the divertor target are removed from the target region by a conventional vacuum getter system. The swirl tubes are oriented and spaced apart within the divertor region relative to the incident angle of the scrape off flux to cause only one side of each tube to be exposed to the flux to increase the burnout rating of the target. The divertor target plane is oriented relative to the plane of the path of the scrape off flux such that the maximum heat flux onto a swirl tube is less than the tube design flux. The containment device is used to contain the plasma of a tokamak fusion reactor and is applicable to other long pulse plasma containment systems.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE PAGES

Maingi, R.; Hu, J. S.; Sun, Z.; ...

2018-01-05

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

NASA Astrophysics Data System (ADS)

Bader, A.; Hegna, C. C.; Cianciosa, M.; Hartwell, G. J.

2018-05-01

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated that vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. These results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.

Development of heat sink concept for near-term fusion power plant divertor

NASA Astrophysics Data System (ADS)

Rimza, Sandeep; Khirwadkar, Samir; Velusamy, Karupanna

2017-04-01

Development of an efficient divertor concept is an important task to meet in the scenario of the future fusion power plant. The divertor, which is a vital part of the reactor has to discharge the considerable fraction of the total fusion thermal power (∼15%). Therefore, it has to survive very high thermal fluxes (∼10 MW/m2). In the present paper, an efficient divertor heat exchanger cooled by helium is proposed for the fusion tokamak. The Plasma facing surface of divertor made-up of several modules to overcome the stresses caused by high heat flux. The thermal hydraulic performance of one such module is numerically investigated in the present work. The result shows that the proposed design is capable of handling target heat flux values of 10 MW/m2. The computational model has been validated against high-heat flux experiments and a satisfactory agreement is noticed between the present simulation and the reported results.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

NASA Astrophysics Data System (ADS)

Maingi, R.; Hu, J. S.; Sun, Z.; Tritz, K.; Zuo, G. Z.; Xu, W.; Huang, M.; Meng, X. C.; Canik, J. M.; Diallo, A.; Lunsford, R.; Mansfield, D. K.; Osborne, T. H.; Gong, X. Z.; Wang, Y. F.; Li, Y. Y.; EAST Team

2018-02-01

We report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3-5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previous ELM elimination results via Li injection into the lower carbon divertor in EAST (Hu et al 2015 Phys. Rev. Lett. 114 055001). These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs (Lang et al 2017 Nucl. Fusion 57 016030), highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.

ELM elimination with Li powder injection in EAST discharges using the tungsten upper divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Maingi, R.; Hu, J. S.; Sun, Z.

Here, we report the first successful use of lithium (Li) to eliminate edge-localized modes (ELMs) with tungsten divertor plasma-facing components in the EAST device. Li powder injected into the scrape-off layer of the tungsten upper divertor successfully eliminated ELMs for 3–5 s in EAST. The ELM elimination became progressively more effective in consecutive discharges at constant lithium delivery rates, and the divertor D α baseline emission was reduced, both signatures of improved wall conditioning. A modest decrease in stored energy and normalized energy confinement was also observed, but the confinement relative to H98 remained well above 1, extending the previousmore » ELM elimination results via Li injection into the lower carbon divertor in EAST. These results can be compared with recent observations with lithium pellets in ASDEX-Upgrade that failed to mitigate ELMs, highlighting one comparative advantage of continuous powder injection for real-time ELM elimination.« less

Double-null divertor configuration discharge and disruptive heat flux simulation using TSC on EAST

NASA Astrophysics Data System (ADS)

Bo, SHI; Jinhong, YANG; Cheng, YANG; Desheng, CHENG; Hui, WANG; Hui, ZHANG; Haifei, DENG; Junli, QI; Xianzu, GONG; Weihua, WANG

2018-07-01

The tokamak simulation code (TSC) is employed to simulate the complete evolution of a disruptive discharge in the experimental advanced superconducting tokamak. The multiplication factor of the anomalous transport coefficient was adjusted to model the major disruptive discharge with double-null divertor configuration based on shot 61 916. The real-time feed-back control system for the plasma displacement was employed. Modeling results of the evolution of the poloidal field coil currents, the plasma current, the major radius, the plasma configuration all show agreement with experimental measurements. Results from the simulation show that during disruption, heat flux about 8 MW m‑2 flows to the upper divertor target plate and about 6 MW m‑2 flows to the lower divertor target plate. Computations predict that different amounts of heat fluxes on the divertor target plate could result by adjusting the multiplication factor of the anomalous transport coefficient. This shows that TSC has high flexibility and predictability.

Experience on divertor fuel retention after two ITER-Like Wall campaigns

NASA Astrophysics Data System (ADS)

Heinola, K.; Widdowson, A.; Likonen, J.; Ahlgren, T.; Alves, E.; Ayres, C. F.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Guillemaut, C.; Jepu, I.; Krat, S.; Lahtinen, A.; Matthews, G. F.; Mayer, M.; Contributors, JET

2017-12-01

The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during the operation of a future fusion reactor. Present work summarizes and reports the plasma fuel retention in the divertor resulting from the two first experimental campaigns with the ITER-Like Wall. The deposition pattern in the divertor after the second campaign shows same trend as was observed after the first campaign: highest deposition of 10-15 μm was found on the top part of the inner divertor. Due to the change in plasma magnetic configurations from the first to the second campaign, and the resulted strike point locations, an increase of deposition was observed on the base of the divertor. The deuterium retention was found to be affected by the hydrogen plasma experiments done at the end of second experimental campaign.

Simple Map with Low MN Perturbation for a Single-Null Divertor Tokamak with Constant Width of Stochastic Layer

NASA Astrophysics Data System (ADS)

Verma, Arun; Smith, Terry; Punjabi, Alkesh; Boozer, Allen

1996-11-01

In this work, we investigate the effects of low MN perturbations in a single-null divertor tokamak with stochastic scrape-off layer. The unperturbed magnetic topology of a single-null divertor tokamak is represented by Simple Map (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). We choose the combinations of the map parameter k, and the strength of the low MN perturbation such that the width of stochastic layer remains unchanged. We give detailed results on the effects of low MN perturbation on the magnetic topology of the stochastic layer and on the footprint of field lines on the divertor plate given the constraint of constant width of the stochastic layer. The low MN perturbations occur naturally and therefore their effects are of considerable importance in tokamak divertor physics. This work is supported by US DOE OFES. Use of CRAY at HU and at NERSC is gratefully acknowledged.

On Heat Loading, Novel Divertors, and Fusion Reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, Mike

2006-10-01

A new magnetic divertor geometry has been proposed to solve reactor heat exhaust problems, which are far more severe for a reactor than for ITER. Using reactor-compatible coils to generate an extra X-point downstream from the main X-point, the new X-divertor (XD) is shown to greatly expand magnetic flux at the divertor plates. As a result, the heat is distributed over a larger area and the line length is greatly increased. The heat-flux limitations of a standard divertor (SD) force a high core radiation fraction (fRad) in most reactor designs that necessarily have a several times higher ratio of heating power to radius (P/R) than ITER. It is argued that such high values of fRad will probably have serious deleterious consequences on the core confinement and stability of a burning plasma. Operation with internal transport barriers (ITBs) does not appear to overcome this problem. By reducing the core fRad within an acceptable range, the X-divertor is shown to substantially lower the core confinement requirement for a fusion reactor. As a bonus, the XD also enables the use of liquid metals by reducing the MHD drag. A possible series of experiments for an efficient and attractive path to practical fusion power is suggested.

Plasma power recycling at the divertor surface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xian -Zhu; Guo, Zehua

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Plasma power recycling at the divertor surface

DOE PAGES

Tang, Xian -Zhu; Guo, Zehua

2016-12-03

With a divertor made of solid materials like carbon and tungsten, plasma ions are expected to be recycled at the divertor surface with a time-averaged particle recycling coefficient very close to unity in steady-state operation. This means that almost every plasma ion (hydrogen and helium) will be returned to the plasma, mostly as neutrals. The power flux deposited by the plasma on the divertor surface, on the other hand, can have varying recycling characteristics depending on the material choice of the divertor; the run-time atomic composition of the surface, which can be modified by material mix due to impurity migrationmore » in the chamber; and the surface morphology change over time. In general, a high-Z–material (such as tungsten) surface tends to reflect light ions and produce stronger power recycling, while a low-Z–material (such as carbon) surface tends to have a larger sticking coefficient for light ions and hence lower power recycling. Here, an explicit constraint on target plasma density and temperature is derived from the truncated bi-Maxwellian sheath model, in relation to the absorbed power load and power recycling coefficient at the divertor surface. Lastly, it is shown that because of the surface recombination energy flux, the attached plasma has a sharper response to power recycling in comparison to a detached plasma.« less

Conceptual design of divertor and first wall for DEMO-FNS

NASA Astrophysics Data System (ADS)

Sergeev, V. Yu.; Kuteev, B. V.; Bykov, A. S.; Gervash, A. A.; Glazunov, D. A.; Goncharov, P. R.; Dnestrovskij, A. Yu.; Khayrutdinov, R. R.; Klishchenko, A. V.; Lukash, V. E.; Mazul, I. V.; Molchanov, P. A.; Petrov, V. S.; Rozhansky, V. A.; Shpanskiy, Yu. S.; Sivak, A. B.; Skokov, V. G.; Spitsyn, A. V.

2015-11-01

Key issues of design of the divertor and the first wall of DEMO-FNS are presented. A double null closed magnetic configuration was chosen with long external legs and V-shaped corners. The divertor employs a cassette design similar to that of ITER. Water-cooled first wall of the tokamak is made of Be tiles and CuCrZr-stainless steel shells. Lithium injection and circulation technologies are foreseen for protection of plasma facing components. Simulations of thermal loads onto the first wall and divertor plates suggest a possibility to distribute heat loads making them less than 10 MW m-2. Evaluations of sputtering and evaporation of plasma-facing materials suggest that lithium may protect the first wall. To prevent Be erosion at the outer divertor plates either the full detached divertor operation or arrangement of the renewal lithium flow on targets should be implemented. Test bed experiments on the Tsefey-M facility with the first wall mockup coated by Ве tiles and cooled by water are presented. The temperature of the surface of tiles reached 280-300 °С at 5 MW m-2 and 600-650 °С at 10.5 MW m-2. The mockup successfully withstood 1000 cycles with the lower thermal loading and 100 cycles with higher thermal loading.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono, M. Jaworski, R. Kaita, C. N. Skinner, J.P. Allain, R. Maingi, F. Scotti, V.A. Soukhanovskii, and the NSTX-U Team

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTXU, the PMI research has received a strong emphasis. With ~ 15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m2 . To support the PMI research, a comprehensive set of PMI diagnostic tools are being implemented. The snow-flake configuration can produce exceptionally high divertor flux expansion of up to ~ 50.more » Combined with the radiative divertor concept, the snow-flake configuration has reduced the divertor heat flux by an order of magnitude in NSTX. Another area of active PMI investigation is the effect of divertor lithium coating (both in solid and liquid phases). The overall NSTX lithium PFC coating results suggest exciting opportunities for future magnetic confinement research including significant electron energy confinement improvements, Hmode power threshold reduction, the control of Edge Localized Modes (ELMs), and high heat flux handling. To support the NSTX-U/PPPL PMI research, there are also a number of associated PMI facilities implemented at PPPL/Princeton University including the Liquid Lithium R&D facility, Lithium Tokamak Experiment, and Laboratories for Materials Characterization and Surface Chemistry.« less

Conceptual design study for heat exhaust management in the ARC fusion pilot plant

NASA Astrophysics Data System (ADS)

Dennett, C. A.; Cao, N. M.; Creely, A. J.; Hecla, J.; Hoffman, H.; Kuang, A. Q.; Major, M.; Ruiz Ruiz, J.; Tinguely, R. A.; Tolman, E. A.; Brunner, D.; Labombard, B.; Sorbom, B. N.; Whyte, D. G.; Grover, P.; Laughman, C.

2017-10-01

The ARC pilot plant conceptual design study has been extended to explore solutions for managing heat exhaust resulting from 525 MW of fusion power in a compact (R 3.3 m) tokamak. Superconducting poloidal field coils are configured to produce double-null equilibria that support X-point target divertors while maintaining the original core plasma shape and toroidal field coil size. Long outer divertor legs are appended to the original vacuum vessel, providing both large surface areas for surface dissipation of radiative heat and significantly reduced neutron damage for divertor components. A molten salt FLiBe blanket adequately shields all superconductors and functions as a tritium breeder, with advanced neutronics calculations indicating a tritium breeding ratio of 1.08. In addition, FLiBe is used as the active coolant for the entire vessel. A tungsten swirl-tube cooling channel is implemented in the divertor, capable of exhausting 12 MW/m2, heat flux while keeping total FliBe pumping power below 1% of fusion power. Finally, three novel diagnostics are explored: Cherenkov radiation emitted in FLiBe to measure fusion reaction rate, microwave interferometry to measure divertor detachment front location, and IR imaging through the FLiBe blanket to monitor selected divertor ``hotspots.''

Actively convected liquid metal divertor

NASA Astrophysics Data System (ADS)

Shimada, Michiya; Hirooka, Yoshi

2014-12-01

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.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahn, J. -W.; Briesemester, A. R.; Kobayashi, M.

Enhanced perpendicular heat and momentum transport induces parallel pressure loss leading to divertor detachment, which can be produced by the increase of density in 2D tokamaks. However, in the 3D configurations such as tokamaks with 3D fields and stellarators, the fraction of perpendicular transport can be higher even in a lower density regime, which could lead to the early transition to detachment without passing through the high-recycling regime. 3D fields applied to the limiter tokamak plasmas produce edge stochastic layers close to the last closed flux surface (LCFS), which can allow for enhanced perpendicular transport and indeed the absence ofmore » high recycling regime and early detachment have been observed in TEXTOR and Tore Supra. However, in the X-point divertor tokamaks with the applied 3D fields, the parallel transport is still dominant and the detachment facilitation has not been observed yet. Rather, 3D fields affected detachment adversely under certain conditions, either by preventing detachment onset as seen in DIII-D or by re-attaching the existing detached plasma as shown in NSTX. The possible way for strong 3D effects to induce access to the early detachment in divertor tokamaks appears to be via significant perpendicular loss of parallel momentum by frictional force for the counter-streaming flows between neighboring flow channels in the divertor. In principle, the adjacent lobes in the 3D divertor tokamak may generate the counter-streaming flow channels. However, an EMC3-EIRENE simulation for ITER H-mode plasmas demonstrated that screened RMP leads to significantly reduced counter-flows near the divertor target, therefore the momentum loss effect leading to detachment facilitation is expected to be small. This is consistent with the observation in LHD, which showed screening (amplification) of RMP fields in the attachment (stable detachment) case. In conclusion, work for optimal parameter window for best divertor operation scenario is needed particularly for the 3D divertor tokamak configuration.« less

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

NASA Astrophysics Data System (ADS)

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 obtained from highly resolved density profiles. As a consequence, the discrepancy between probe and IR measurements is attributed to the ion power channel via a high mean impact energy of the ions at the inner target. The dominant contributing mechanism is proposed to be the directed loss of ions from the pedestal region into the inner divertor.

A multichannel visible spectroscopy system for the ITER-like W divertor on EAST.

PubMed

Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

2017-04-01

To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

The appearance and propagation of filaments in the private flux region in Mega Amp Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J.

2015-09-15

The transport of particles via intermittent filamentary structures in the private flux region (PFR) of plasmas in the MAST tokamak has been investigated using a fast framing camera recording visible light emission from the volume of the lower divertor, as well as Langmuir probes and IR thermography monitoring particle and power fluxes to plasma-facing surfaces in the divertor. The visible camera data suggest that, in the divertor volume, fluctuations in light emission above the X-point are strongest in the scrape-off layer (SOL). Conversely, in the region below the X-point, it is found that these fluctuations are strongest in the PFRmore » of the inner divertor leg. Detailed analysis of the appearance of these filaments in the camera data suggests that they are approximately circular, around 1–2 cm in diameter, but appear more elongated near the divertor target. The most probable toroidal quasi-mode number is between 2 and 3. These filaments eject plasma deeper into the private flux region, sometimes by the production of secondary filaments, moving at a speed of 0.5–1.0 km/s. Probe measurements at the inner divertor target suggest that the fluctuations in the particle flux to the inner target are strongest in the private flux region, and that the amplitude and distribution of these fluctuations are insensitive to the electron density of the core plasma, auxiliary heating and whether the plasma is single-null or double-null. It is found that the e-folding width of the time-average particle flux in the PFR decreases with increasing plasma current, but the fluctuations appear to be unaffected. At the outer divertor target, the fluctuations in particle and power fluxes are strongest in the SOL.« less

A multichannel visible spectroscopy system for the ITER-like W divertor on EAST

NASA Astrophysics Data System (ADS)

Mao, Hongmin; Ding, Fang; Luo, Guang-Nan; Hu, Zhenhua; Chen, Xiahua; Xu, Feng; Yang, Zhongshi; Chen, Jingbo; Wang, Liang; Ding, Rui; Zhang, Ling; Gao, Wei; Xu, Jichan; Wu, Chengrui

2017-04-01

To facilitate long-pulse high power operation, an ITER-like actively cooled tungsten (W) divertor was installed in Experimental Advanced Superconducting Tokamak (EAST) to replace the original upper graphite divertor in 2014. A dedicated multichannel visible spectroscopic diagnostic system has been accordingly developed for the characterization of the plasma and impurities in the W divertor. An array of 22 lines-of-sight (LOSs) provides a profile measurement of the light emitted from the plasma along upper outer divertor, and the other 17 vertical LOSs view the upper inner divertor, achieving a 13 mm poloidal resolution in both regions. The light emitted from the plasma is collected by a specially designed optical lens assembly and then transferred to a Czerny-Turner spectrometer via 40 m quartz fibers. At the end, the spectra dispersed by the spectrometer are recorded with an Electron-Multiplying Charge Coupled Device (EMCCD). The optical throughput and quantum efficiency of the system are optimized in the wavelength range 350-700 nm. The spectral resolution/coverage can be adjusted from 0.01 nm/3 nm to 0.41 nm/140 nm by switching the grating with suitable groove density. The frame rate depends on the setting of LOS number in EMCCD and can reach nearly 2 kHz for single LOS detection. The light collected by the front optical lens can also be divided and partly transferred to a photomultiplier tube array with specified bandpass filter, which can provide faster sampling rates by up to 200 kHz. The spectroscopic diagnostic is routinely operated in EAST discharges with absolute optical calibrations applied before and after each campaign, monitoring photon fluxes from impurities and H recycling in the upper divertor. This paper presents the technical details of the diagnostic and typical measurements during EAST discharges.

Various divertor biasing configurations and improved divertor performance with biasing on Tokamak de Varennes (TdeV)*

DOE Office of Scientific and Technical Information (OSTI.GOV)

Decoste, R.; Lachambre, J.; Abel, G.

1994-05-01

Electrically insulated divertor plates are used on TdeV (Tokamak de Varennes) [18[ital th] [ital EPS] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Physics] Berlin (European Physical Society, Petit-Lancy, 1991), Vol. 15C, Part I, pp. 1--141] to produce various biasing configurations, which can be decomposed into two basic modes. Plasma biasing, with a radial electric field [ital E][sub [ital r

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

NASA Astrophysics Data System (ADS)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Active Control of Power Exhaust in Strongly Heated ASDEX Upgrade Plasmas

NASA Astrophysics Data System (ADS)

Dux, Ralph; Kallenbach, Arne; Bernert, Matthias; Eich, Thomas; Fuchs, Christoph; Giannone, Louis; Herrmann, Albrecht; Schweinzer, Josef; Treutterer, Wolfgang

2012-10-01

Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.

Erosion and deposition in the JET divertor during the second ITER-like wall campaign

NASA Astrophysics Data System (ADS)

Mayer, M.; Krat, S.; Baron-Wiechec, A.; Gasparyan, Yu; Heinola, K.; Koivuranta, S.; Likonen, J.; Ruset, C.; de Saint-Aubin, G.; Widdowson, A.; Contributors, JET

2017-12-01

Erosion of plasma-facing materials and successive transport and redeposition of eroded material are crucial processes determining the lifetime of plasma-facing components and the trapped tritium inventory in redeposited material layers. Erosion and deposition in the JET divertor were studied during the second JET ITER-like wall campaign ILW-2 in 2013-2014 by using a poloidal row of specially prepared divertor marker tiles including the tungsten bulk tile 5. The marker tiles were analyzed using elastic backscattering with 3-4.5 MeV incident protons and nuclear reaction analysis using 0.8-4.5 MeV 3He ions before and after the campaign. The erosion/deposition pattern observed during ILW-2 is qualitatively comparable to the first campaign ILW-1 in 2011-2012: deposits consist mainly of beryllium with 5-20 at.% of carbon and oxygen and small amounts of Ni and W. The highest deposition with deposited layer thicknesses up to 30 μm per campaign is still observed on the upper and horizontal parts of the inner divertor. Outer divertor tiles 5, 6, 7 and 8 are net W erosion areas. The observed D inventory is roughly comparable to the inventory observed during ILW-1. The results obtained during ILW-2 therefore confirm the positive results observed in ILW-1 with respect to reduced material deposition and hydrogen isotopes retention in the divertor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dr. Ricardo Maqueda; Dr. Fred M. Levinton

Nova Photonics, Inc. has a collaborative effort at the National Spherical Torus Experiment (NSTX). This collaboration, based on fast imaging of visible phenomena, has provided key insights on edge turbulence, intermittency, and edge phenomena such as edge localized modes (ELMs) and multi-faceted axisymmetric radiation from the edge (MARFE). Studies have been performed in all these areas. The edge turbulence/intermittency studies make use of the Gas Puff Imaging diagnostic developed by the Principal Investigator (Ricardo Maqueda) together with colleagues from PPPL. This effort is part of the International Tokamak Physics Activity (ITPA) edge, scrape-off layer and divertor group joint activity (DSOL-15:more » Inter-machine comparison of blob characteristics). The edge turbulence/blob study has been extended from the current location near the midplane of the device to the lower divertor region of NSTX. The goal of this effort was to study turbulence born blobs in the vicinity of the X-point region and their circuit closure on divertor sheaths or high density regions in the divertor. In the area of ELMs and MARFEs we have studied and characterized the mode structure and evolution of the ELM types observed in NSTX, as well as the study of the observed interaction between MARFEs and ELMs. This interaction could have substantial implications for future devices where radiative divertor regions are required to maintain detachment from the divertor plasma facing components.« less

Dynamic divertor control using resonant mixed toroidal harmonic magnetic fields during ELM suppression in DIII-D

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Paz-Soldan, C.; Nazikian, R.; Gu, S.; Liu, Y. Q.; Abrams, T.; Bykov, I.; Cui, L.; Evans, T.; Garofalo, A.; Guo, W.; Gong, X.; Lasnier, C.; Logan, N. C.; Makowski, M.; Orlov, D.; Wang, H. H.

2018-05-01

Experiments using Resonant Magnetic Perturbations (RMPs), with a rotating n = 2 toroidal harmonic combined with a stationary n = 3 toroidal harmonic, have validated predictions that divertor heat and particle flux can be dynamically controlled while maintaining Edge Localized Mode (ELM) suppression in the DIII-D tokamak. Here, n is the toroidal mode number. ELM suppression over one full cycle of a rotating n = 2 RMP that was mixed with a static n = 3 RMP field has been achieved. Prominent heat flux splitting on the outer divertor has been observed during ELM suppression by RMPs in low collisionality regime in DIII-D. Strong changes in the three dimensional heat and particle flux footprint in the divertor were observed during the application of the mixed toroidal harmonic magnetic perturbations. These results agree well with modeling of the edge magnetic field structure using the TOP2D code, which takes into account the plasma response from the MARS-F code. These results expand the potential effectiveness of the RMP ELM suppression technique for the simultaneous control of divertor heat and particle load required in ITER.

Divertor Coil Design and Implementation on Pegasus

NASA Astrophysics Data System (ADS)

Shriwise, P. C.; Bongard, M. W.; Cole, J. A.; Fonck, R. J.; Kujak-Ford, B. A.; Lewicki, B. T.; Winz, G. R.

2012-10-01

An upgraded divertor coil system is being commissioned on the Pegasus Toroidal Experiment in conjunction with power system upgrades in order to achieve higher β plasmas, reduce impurities, and possibly achieve H-mode operation. Design points for the divertor coil locations and estimates of their necessary current ratings were found using predictive equilibrium modeling based upon a 300 kA target plasma. This modeling represented existing Pegasus coil locations and current drive limits. The resultant design calls for 125 kA-turns from the divertor system to support the creation of a double null magnetic topology in plasmas with Ip<=300 kA. Initial experiments using this system will employ 900 V IGBT power supply modules to provide IDIV<=4 kA. The resulting 20 kA-turn capability of the existing divertor coil will be augmented by a new coil providing additional A-turns in series. Induced vessel wall current modeling indicates the time response of a 28 turn augmentation coil remains fast compared to the poloidal field penetration rate through the vessel. First results operating the augmented system are shown.

Parallel Energy Transport in Detached DIII-D Divertor Plasmas

NASA Astrophysics Data System (ADS)

Leonard, A. W.; Lore, J. D.; Canik, J. M.; McLean, A. G.; Makowski, M. A.

2017-10-01

A comparison of experiment and modeling of detached divertor plasmas is examined in the context of parallel energy transport. Experimental estimates of power carried by electron thermal conduction versus plasma convection are experimentally inferred from power balance measurements of radiated power and target plate heat flux combined with Thomson scattering measurements of the Te profile along the divertor leg. Experimental profiles of Te exhibit relatively low gradients with Te < 15 eV from the X-point to the target implying transport dominated by convection. In contrast, fluid modeling with SOLPS produces sharp Te gradients for Te > 3 eV, characteristic of transport dominated by electron conduction through the bulk of the divertor. This discrepancy with experimental transport dominated by convection and modeling by conduction has significant implications for the radiative capacity of divertor plasmas and may explain at least part of the difficulty for fluid modeling to obtain the experimentally observed radiative losses. Comparisons are also made for helium plasmas where the match between experiment and modeling is much better. Work supported by the US DOE under DE-FC02-04ER54698.

Minimum magnetic curvature for resilient divertors using Compact Toroidal Hybrid geometry

DOE PAGES

Bader, Aaron; Hegna, C. C.; Cianciosa, Mark R.; ...

2018-03-16

The properties of resilient divertors are explored using equilibria derived from Compact Toroidal Hybrid (CTH) geometries. Resilience is defined here as the robustness of the strike point patterns as the plasma geometry and/or plasma profiles are changed. The addition of plasma current in the CTH configurations significantly alters the shape of the last closed flux surface and the rotational transform profile, however, it does not alter the strike point pattern on the target plates, and hence has resilient divertor features. The limits of when a configuration transforms to a resilient configuration is then explored. New CTH-like configurations are generated thatmore » vary from a perfectly circular cross section to configurations with increasing amounts of toroidal shaping. It is found that even small amounts of toroidal shaping lead to strike point localization that is similar to the standard CTH configuration. Lastly, these results show that only a small degree of three-dimensional shaping is necessary to produce a resilient divertor, implying that any highly shaped optimized stellarator will possess the resilient divertor property.« less

Designing divertor targets for uniform power load

NASA Astrophysics Data System (ADS)

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

2015-08-01

Divertor design for next step fusion reactors heavily relies on 2D edge plasma modeling with codes as e.g. B2-EIRENE. While these codes are typically used in a design-by-analysis approach, in previous work we have shown that divertor design can alternatively be posed as a mathematical optimization problem, and solved very efficiently using adjoint methods adapted from computational aerodynamics. This approach has been applied successfully to divertor target shape design for more uniform power load. In this paper, the concept is further extended to include all contributions to the target power load, with particular focus on radiation. In a simplified test problem, we show the potential benefits of fully including the radiation load in the design cycle as compared to only assessing this load in a post-processing step.

Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

NASA Astrophysics Data System (ADS)

Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

2018-05-01

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on W transport assumptions during the ELM, a maximum ELM frequency is also identified above which core tungsten accumulation takes place.

The Thomson scattering diagnostic at Wendelstein 7-X and its performance in the first operation phase

NASA Astrophysics Data System (ADS)

Bozhenkov, S. A.; Beurskens, M.; Dal Molin, A.; Fuchert, G.; Pasch, E.; Stoneking, M. R.; Hirsch, M.; Höfel, U.; Knauer, J.; Svensson, J.; Trimino Mora, H.; Wolf, R. C.

2017-10-01

The optimized stellarator Wendelstein 7-X started operation in December 2015 with a 10 week limiter campaign. Divertor experiments will begin in the second half of 2017. The W7-X Thomson scattering system is an essential diagnostic for electron density and temperature profiles. In this paper the Thomson scattering diagnostic is described in detail, including its design, calibration, data evaluation and first experimental results. Plans for further development are also presented. The W7-X Thomson system is a Nd:YAG setup with up to five lasers, two sets of light collection lenses viewing the entire plasma cross-section, fiber bundles and filter based polychromators. To reduce hardware costs, two or three scattering volumes are measured with a single polychromator. The relative spectral calibration is carried out with the aid of a broadband supercontinuum light source. The absolute calibration is performed by observing Raman scattering in nitrogen. The electron temperatures and densities are recovered by Bayesian modelling. In the first campaign, the diagnostic was equipped for 10 scattering volumes. It provided temperature profiles comparable to those measured using an electron cyclotron emission diagnostic and line integrated densities within 10% of those from a dispersion interferometer.

Simulation of turbulence in the divertor region of tokamak edge plasma

NASA Astrophysics Data System (ADS)

Umansky, M. V.; Rognlien, T. D.; Xu, X. Q.

2005-03-01

Results are presented for turbulence simulations with the fluid edge turbulence code BOUT [X.Q. Xu, R.H. Cohen, Contr. Plas. Phys. 36 (1998) 158]. The present study is focussed on turbulence in the divertor leg region and on the role of the X-point in the structure of turbulence. Results of the present calculations indicate that the ballooning effects are important for the divertor fluctuations. The X-point shear leads to weak correlation of turbulence across the X-point regions, in particular for large toroidal wavenumber. For the saturated amplitudes of the divertor region turbulence it is found that amplitudes of density fluctuations are roughly proportional to the local density of the background plasma. The amplitudes of electron temperature and electric potential fluctuations are roughly proportional to the local electron temperature of the background plasma.

Divertor with a third-order null of the poloidal field

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ryutov, D. D.; Umansky, M. V.

2013-09-15

A concept and preliminary feasibility analysis of a divertor with the third-order poloidal field null is presented. The third-order null is the point where not only the field itself but also its first and second spatial derivatives are zero. In this case, the separatrix near the null-point has eight branches, and the number of strike-points increases from 2 (as in the standard divertor) to six. It is shown that this magnetic configuration can be created by a proper adjustment of the currents in a set of three divertor coils. If the currents are somewhat different from the required values, themore » configuration becomes that of three closely spaced first-order nulls. Analytic approach, suitable for a quick orientation in the problem, is used. Potential advantages and disadvantages of this configuration are briefly discussed.« less

Initial development of the DIII–D snowflake divertor control

NASA Astrophysics Data System (ADS)

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.

Nuclear analysis of structural damage and nuclear heating on enhanced K-DEMO divertor model

NASA Astrophysics Data System (ADS)

Park, J.; Im, K.; Kwon, S.; Kim, J.; Kim, D.; Woo, M.; Shin, C.

2017-12-01

This paper addresses nuclear analysis on the Korean fusion demonstration reactor (K-DEMO) divertor to estimate the overall trend of nuclear heating values and displacement damages. The K-DEMO divertor model was created and converted by the CAD (Pro-Engineer™) and Monte Carlo automatic modeling programs as a 22.5° sector of the tokamak. The Monte Carlo neutron photon transport and ADVANTG codes were used in this calculation with the FENDL-2.1 nuclear data library. The calculation results indicate that the highest values appeared on the upper outboard target (OT) area, which means the OT is exposed to the highest radiation conditions among the three plasma-facing parts (inboard, central and outboard) in the divertor. Especially, much lower nuclear heating values and displacement damages are indicated on the lower part of the OT area than others. These are important results contributing to thermal-hydraulic and thermo-mechanical analyses on the divertor and also it is expected that the copper alloy materials may be partially used as a heat sink only at the lower part of the OT instead of the reduced activation ferritic-martensitic steel due to copper alloy’s high thermal conductivity.

Power exhaust scenarios and control for projected high-power NSTX-U operation

NASA Astrophysics Data System (ADS)

Menard, Jonathan; Gerhardt, S. P.; Myers, C. E.; Reinke, M. L.; Brooks, A.; Mardenfeld, M.; NSTX Upgrade Team

2017-10-01

An important goal of the NSTX Upgrade (NSTX-U) research program is to characterize energy confinement in the low-aspect-ratio spherical tokamak configuration over a significantly expanded range of plasma current, toroidal field, and heating power, while increasing flattop durations up to 5 seconds. However, the narrowing of the scrape-off layer at higher current combined with an improved understanding of expected halo-current loads has motivated a significant re-design of NSTX-U plasma facing components in the high-heat-flux regions of the divertor. In order to reduce the expected divertor heat flux to acceptable levels, a combination of mitigation techniques will be used: increased divertor poloidal flux expansion, increased divertor radiation, and controlled strike-point sweeping. The machine requirements for these various mitigation techniques are studied here using a newly implemented reduced heat-flux model. Systematic equilibrium scans are used to quantify the required divertor coil currents and to verify vertical stability for a range of plasma shapes. Free-boundary control schemes to constrain the strike-point location and field-line angle-of-incidence will also be discussed. Work supported by DOE contract DE-AC02- 09CH11466.

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

Exposures of tungsten nanostructures to divertor plasmas in DIII-D

DOE PAGES

Rudakov, D. L.; Wong, C. P. C.; Doerner, R. P.; ...

2016-01-22

Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with littlemore » obvious damage except in the areas where unipolar arcing occurred. In conclusion, arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.« less

Plasma transport in a simulated magnetic-divertor configuration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strawitch, C. M.

1981-03-01

The transport properties of plasma on magnetic field lines that intersect a conducting plate are studied experimentally in the Wisconsin internal ring D.C. machine. The magnetic geometry is intended to simulate certain aspects of plasma phenomena that may take place in a tokamak divertor. It is found by a variety of measurements that the cross field transport is non-ambipolar; this may have important implications in heat loading considerations in tokamak divertors. The undesirable effects of nonambipolar flow make it preferable to be able to eliminate it. However, we find that though the non-ambipolarity may be reduced, it is difficult tomore » eliminate entirely. The plasma flow velocity parallel to the magnetic field is found to be near the ion acoustic velocity in all cases. The experimental density and electron temperature profiles are compared to the solutions to a one dimensional transport model that is commonly used in divertor theory.« less

Structural investigation of re-deposited layers in JET

NASA Astrophysics Data System (ADS)

Likonen, J.; Vainonen-Ahlgren, E.; Khriachtchev, L.; Coad, J. P.; Rubel, M.; Renvall, T.; Arstila, K.; Hole, D. E.; Contributors to the EFDA-JET Work-programme

2008-07-01

JET Mk-II Gas Box divertor tiles exposed in 1998-2001 have been analysed with various ion beam techniques, secondary ion mass spectrometry (SIMS) and Raman spectroscopy. Inner divertor wall tiles removed in 2001 were covered with a duplex film. The inner layer was very rich in metallic impurities, with Be/C ˜ 1 and H-isotopes only present at low concentrations. The outer layer contained higher concentrations of D than normal for plasma-facing surfaces in JET (D/C ˜ 0.4), and Be/C ˜ 0.14. Raman and SIMS analyses show that the deposited films on inner divertor tiles are hydrogenated amorphous carbon with low sp 3 fractions. The deposits have polymeric structure and low density. Both Raman scattering and SIMS indicate that films on inner divertor wall Tiles 1 and 3, and on floor Tile 4 have some differences in the chemical structure of the deposited films

Initial development of the DIII–D snowflake divertor control

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.

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 plasmamore » 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. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

Initial development of the DIII–D snowflake divertor control

DOE PAGES

Kolemen, Egemen; Vail, P. J.; Makowski, M. A.; ...

2018-04-11

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 plasmamore » 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. In conclusion, the SFD resulted in a 2.5×reduction in the peak heat flux for many energy confinement times (2–3s) without any adverse effects on core plasma performance.« less

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murphy, Christopher; Nygren, R. E.; Chrobak, C P.

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 ofmore » 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.« less

Three-dimensional modeling of plasma edge transport and divertor fluxes during application of resonant magnetic perturbations on ITER

NASA Astrophysics Data System (ADS)

Schmitz, O.; Becoulet, M.; Cahyna, P.; Evans, T. E.; Feng, Y.; Frerichs, H.; Loarte, A.; Pitts, R. A.; Reiser, D.; Fenstermacher, M. E.; Harting, D.; Kirschner, A.; Kukushkin, A.; Lunt, T.; Saibene, G.; Reiter, D.; Samm, U.; Wiesen, S.

2016-06-01

Results from three-dimensional modeling of plasma edge transport and plasma-wall interactions during application of resonant magnetic perturbation (RMP) fields for control of edge-localized modes in the ITER standard 15 MA Q  =  10 H-mode are presented. The full 3D plasma fluid and kinetic neutral transport code EMC3-EIRENE is used for the modeling. Four characteristic perturbed magnetic topologies are considered and discussed with reference to the axisymmetric case without RMP fields. Two perturbation field amplitudes at full and half of the ITER ELM control coil current capability using the vacuum approximation are compared to a case including a strongly screening plasma response. In addition, a vacuum field case at high q 95  =  4.2 featuring increased magnetic shear has been modeled. Formation of a three-dimensional plasma boundary is seen for all four perturbed magnetic topologies. The resonant field amplitudes and the effective radial magnetic field at the separatrix define the shape and extension of the 3D plasma boundary. Opening of the magnetic field lines from inside the separatrix establishes scrape-off layer-like channels of direct parallel particle and heat flux towards the divertor yielding a reduction of the main plasma thermal and particle confinement. This impact on confinement is most accentuated at full RMP current and is strongly reduced when screened RMP fields are considered, as well as for the reduced coil current cases. The divertor fluxes are redirected into a three-dimensional pattern of helical magnetic footprints on the divertor target tiles. At maximum perturbation strength, these fingers stretch out as far as 60 cm across the divertor targets, yielding heat flux spreading and the reduction of peak heat fluxes by 30%. However, at the same time substantial and highly localized heat fluxes reach divertor areas well outside of the axisymmetric heat flux decay profile. Reduced RMP amplitudes due to screening or reduced RMP coil current yield a reduction of the width of the divertor flux spreading to about 20-25 cm and cause increased peak heat fluxes back to values similar to those in the axisymmetric case. The dependencies of these features on the divertor recycling regime and the perpendicular transport assumptions, as well as toroidal averaged effects mimicking rotation of the RMP field, are discussed in the paper.

Safety characteristics of the monolithic CFC divertor

NASA Astrophysics Data System (ADS)

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.

Upstream Density for Plasma Detachment with Conventional and Lithium Vapor-Box Divertors

NASA Astrophysics Data System (ADS)

Goldston, Rj; Schwartz, Ja

2016-10-01

Fusion power plants are likely to require detachment of the divertor plasma from material targets. The lithium vapor box divertor is designed to achieve this, while limiting the flux of lithium vapor to the main plasma. We develop a simple model of near-detachment to evaluate the required upstream plasma density, for both conventional and lithium vapor-box divertors, based on particle and dynamic pressure balance between up- and down-stream, at near-detachment conditions. A remarkable general result is found, not just for lithium-induced detachment, that the upstream density divided by the Greenwald-limit density scales as (P 5 / 8 /B 3 / 8) Tdet1 / 2 / (ɛcool + γTdet) , with no explicit size scaling. Tdet is the temperature just before strong pressure loss, 1/2 of the ionization potential of the dominant recycling species, ɛcool is the average plasma energy lost per injected hydrogenic and impurity atom, and γ is the sheath heat transmission factor. A recent 1-D calculation agrees well with this scaling. The implication is that the plasma exhaust problem cannot be solved by increasing R. Instead significant innovation, such as the lithium vapor box divertor, will be required. This work supported by DOE Contract No. DE-AC02-09CH11466.

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Brunner, D.; Burke, W.; Kuang, A. Q.; LaBombard, B.; Lipschultz, B.; Wolfe, S.

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Feedback system for divertor impurity seeding based on real-time measurements of surface heat flux in the Alcator C-Mod tokamak.

PubMed

Brunner, D; Burke, W; Kuang, A Q; LaBombard, B; Lipschultz, B; Wolfe, S

2016-02-01

Mitigation of the intense heat flux to the divertor is one of the outstanding problems in fusion energy. One technique that has shown promise is impurity seeding, i.e., the injection of low-Z gaseous impurities (typically N2 or Ne) to radiate and dissipate the power before it arrives to the divertor target plate. To this end, the Alcator C-Mod team has created a first-of-its-kind feedback system to control the injection of seed gas based on real-time surface heat flux measurements. Surface thermocouples provide real-time measurements of the surface temperature response to the plasma heat flux. The surface temperature measurements are inputted into an analog computer that "solves" the 1-D heat transport equation to deliver accurate, real-time signals of the surface heat flux. The surface heat flux signals are sent to the C-Mod digital plasma control system, which uses a proportional-integral-derivative (PID) algorithm to control the duty cycle demand to a pulse width modulated piezo valve, which in turn controls the injection of gas into the private flux region of the C-Mod divertor. This paper presents the design and implementation of this new feedback system as well as initial results using it to control divertor heat flux.

Fast imaging of filaments in the X-point region of Alcator C-Mod

DOE PAGES

Terry, J. L.; Ballinger, S.; Brunner, D.; ...

2017-01-27

A rich variety of field-aligned fluctuations has been revealed using fast imaging of D α emission from Alcator C-Mod's lower X-point region. Field-aligned filamentary fluctuations are observed along the inner divertor leg, within the Private-Flux-Zone (PFZ), in the Scrape-Off Layer (SOL) outside the outer divertor leg, and, under some conditions, at or above the X-point. The locations and dynamics of the filaments in these regions are strikingly complex in C-Mod. Changes in the filaments’ generation appear to be ordered by plasma density and magnetic configuration. Filaments are not observed for plasmas with n/nGreenwald ≲ 0.12 nor are they observed inmore » Upper Single Null configurations. In a Lower Single Null with 0.12 ≲ n/nGreenwald ≲ 0.45 and Bx∇B directed down, filaments typically move up the inner divertor leg toward the X-point. Reversing the field direction results in the appearance of filaments outside of the outer divertor leg. With the divertor targets “detached”, filaments inside the LCFS are seen. Lastly, these studies were motivated by observations of filaments in the X-point and PFZ regions in MAST, and comparisons with those observations are made.« less

Advanced simulation of mixed-material erosion/evolution and application to low and high-Z containing plasma facing components

NASA Astrophysics Data System (ADS)

Brooks, J. N.; Hassanein, A.; Sizyuk, T.

2013-07-01

Plasma interactions with mixed-material surfaces are being analyzed using advanced modeling of time-dependent surface evolution/erosion. Simulations use the REDEP/WBC erosion/redeposition code package coupled to the HEIGHTS package ITMC-DYN mixed-material formation/response code, with plasma parameter input from codes and data. We report here on analysis for a DIII-D Mo/C containing tokamak divertor. A DIII-D/DiMES probe experiment simulation predicts that sputtered molybdenum from a 1 cm diameter central spot quickly saturates (˜4 s) in the 5 cm diameter surrounding carbon probe surface, with subsequent re-sputtering and transport to off-probe divertor regions, and with high (˜50%) redeposition on the Mo spot. Predicted Mo content in the carbon agrees well with post-exposure probe data. We discuss implications and mixed-material analysis issues for Be/W mixing at the ITER outer divertor, and Li, C, Mo mixing at an NSTX divertor.

Initial results from divertor heat-flux instrumentation on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Labombard, B.; Brunner, D.; Payne, J.; Reinke, M.; Terry, J. L.; Hughes, J. W.; Lipschultz, B.; Whyte, D.

2009-11-01

Physics-based plasma transport models that can accurately simulate the heat-flux power widths observed in the tokamak boundary are lacking at the present time. Yet this quantity is of fundamental importance for ITER and most critically important for DEMO, a reactor similar to ITER but with ˜4 times the power exhaust. In order to improve our understanding, C-Mod, DIII-D and NSTX will aim experiments in FY10 towards characterizing the divertor ``footprint'' and its connection to conditions ``upstream'' in the boundary and core plasmas [2]. Standard IR-based heat-flux measurements are particularly difficult in C-Mod, due to its vertical-oriented divertor targets. To overcome this, a suite of embedded heat-flux sensor probes (tile thermocouples, calorimeters, surface thermocouples) combined with IR thermography was installed during the FY09 opening, along with a new divertor bolometer system. This paper will report on initial experiments aimed at unfolding the heat-flux dependencies on plasma operating conditions. [2] a proposed US DoE Joint Facilities Milestone.

HSX as an example of a resilient non-resonant divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bader, A.; Boozer, A. H.; Hegna, C. C.

This study describes an initial description of the resilient divertor properties of quasi-symmetric (QS) stellarators using the HSX (Helically Symmetric eXperiment) configuration as a test-case. Divertors in high-performance QS stellarators will need to be resilient to changes in plasma configuration that arise due to evolution of plasma pressure profiles and bootstrap currents for divertor design. Resiliency is tested by examining the changes in strike point patterns from the field line following, which arise due to configurational changes. A low strike point variation with high configuration changes corresponds to high resiliency. The HSX edge displays resilient properties with configuration changes arisingmore » from the (1) wall position, (2) plasma current, and (3) external coils. The resilient behavior is lost if large edge islands intersect the wall structure. The resilient edge properties are corroborated by heat flux calculations from the fully 3-D plasma simulations using EMC3-EIRENE. Additionally, the strike point patterns are found to correspond to high curvature regions of magnetic flux surfaces.« less

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

NASA Astrophysics Data System (ADS)

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.

Experiences with tungsten coatings in high heat flux tests and under plasma load in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Herrmann, A.; Greuner, H.; Fuchs, J. C.; de Marné, P.; Neu, R.; ASDEX Upgrade Team

2009-12-01

ASDEX Upgrade was operated with about 6400 s plasma discharge during the scientific program in 2007/2008 exploring tungsten as a first wall material in tokamaks. In the first phase, the heating power was restricted to 10 MW. It was increased to 15 MW in the second phase. During this operational period, a delamination of the 200 μm W-VPS coating happened at 2 out of 128 tiles of the outer divertor and an unscheduled opening was required. In the third phase, ASDEX Upgrade was operated with partly predamaged tiles and up to 15 MW heating power. The target load was actively controlled by N2-seeding. This paper presents the screening test of target tiles in the high heat flux test facility GLADIS, experiences with operation and detected damages of the outer divertor as well as the heat load to the outer divertor and the reasons for the toroidal asymmetry of the divertor load.

Design of snowflake-diverted equilibria of CFETR

NASA Astrophysics Data System (ADS)

Hang, LI; Xiang, GAO; Guoqiang, LI; Zhengping, LUO; Damao, YAO; Yong, GUO

2018-03-01

The Chinese Fusion Engineering Test Reactor (CFETR) represents the next generation of full superconducting fusion reactors in China. Recently, CFETR was redesigned with a larger size and will be operated in two phases. To reduce the heat flux on the target plate, a snowflake (SF) divertor configuration is proposed. In this paper we show that by adding two dedicated poloidal field (PF) coils, the SF configuration can be achieved in both phases. The equilibria were calculated by TEQ code for a range of self-inductances l i3. The coil currents were calculated at some fiducial points in the flattop phase. The results indicate that the PF coil system has the ability to maintain a long flattop phase in 7.5 and 10 MA inductive scenarios for the single null divertor (SND) and SF divertor configurations. The properties of the SF configuration were also analyzed. The connection length and flux expansion of the SF divertor were both increased significantly over the SND.

HSX as an example of a resilient non-resonant divertor

DOE PAGES

Bader, A.; Boozer, A. H.; Hegna, C. C.; ...

2017-03-16

This study describes an initial description of the resilient divertor properties of quasi-symmetric (QS) stellarators using the HSX (Helically Symmetric eXperiment) configuration as a test-case. Divertors in high-performance QS stellarators will need to be resilient to changes in plasma configuration that arise due to evolution of plasma pressure profiles and bootstrap currents for divertor design. Resiliency is tested by examining the changes in strike point patterns from the field line following, which arise due to configurational changes. A low strike point variation with high configuration changes corresponds to high resiliency. The HSX edge displays resilient properties with configuration changes arisingmore » from the (1) wall position, (2) plasma current, and (3) external coils. The resilient behavior is lost if large edge islands intersect the wall structure. The resilient edge properties are corroborated by heat flux calculations from the fully 3-D plasma simulations using EMC3-EIRENE. Additionally, the strike point patterns are found to correspond to high curvature regions of magnetic flux surfaces.« less

Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

NASA Astrophysics Data System (ADS)

McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

2017-10-01

An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

OEDGE Modeling of Collector Probe measurements in L-mode from the DIII-D tungsten ring campaign

NASA Astrophysics Data System (ADS)

Elder, J. D.; Stangeby, P. C.; Unterberg, Z.; Donovan, D.; Wampler, W. R.; Watkins, J.; Abrams, T.; McLean, A. G.

2017-10-01

During the tungsten ring campaign on DIII-D, a collector probe system with multiple diameter, dual-facing collector rods was inserted into the far scrape off layer (SOL) near the outer midplane to measure the plasma tungsten content. For most probes more tungsten was observed on the side connected along field lines to the inner divertor, with the larger probes showing largest divertor-facing asymmetries The OEDGE code is used to model the tungsten erosion, transport and deposition. It has been enhanced with (i) a peripheral particle transport and deposition model to record the impurity content in the peripheral region outside the regular mesh, and (ii) a collector probe model. The OEDGE results approximately reproduce both the divertor-facing asymmetries and the radial decay of each collector probe profile. The effect of changing impurity transport assumptions and wall location are examined. The measured divertor-facing asymmetries imply a higher tungsten density in the plasma upstream of the probe; this was expected theoretically from the effect of the parallel ion temperature gradient force driving upstream transport of tungsten from the outer divertor and was also found in the code analysis. Work supported by the US Department of Energy under DE-FC02-04ER54698, DE-NA0003525, DE-AC05-00OR22725, and DE-AC52-07NA27344.

Characterization of chemical sputtering using the Mark II DiMES porous plug injector in attached and semi-detached divertor plasmas of DIII-D

NASA Astrophysics Data System (ADS)

McLean, A. G.; Davis, J. W.; Stangeby, P. C.; Allen, S. L.; Boedo, J. A.; Bray, B. D.; Brezinsek, S.; Brooks, N. H.; Fenstermacher, M. E.; Groth, M.; Haasz, A. A.; Hollmann, E. M.; Isler, R. C.; Lasnier, C. J.; Mu, Y.; Petrie, T. W.; Rudakov, D. L.; Watkins, J. G.; West, W. P.; Whyte, D. G.; Wong, C. P. C.

2009-06-01

An improved, self-contained gas injection system for the divertor material evaluation system (DiMES) on DIII-D has been employed for in situ study of chemical erosion in the tokamak divertor environment. To minimize perturbation to local plasma, the Mark II porous plug injector (PPI) releases methane through a porous graphite surface at the outer strike point at a rate precisely controlled by a micro-orifice flow restrictor to be approximately equal as that predicted for intrinsic chemical sputtering. Effective photon efficiencies resulting from CH 4 are found to be 58 ± 12 in an attached divertor ( ne ˜ 1.5 × 10 13/cm 3, Te ˜ 25 eV, Tsurf ˜ 450 K), and 94 ± 20 in a semi-detached cold divertor ( ne ˜ 6.0 × 10 13/cm 3, Te ˜ 2-3 eV, Tsurf ˜ 350 K). These values are significantly more than previous measurements in similar plasma conditions, indicating the importance of the injection rate and local re-erosion for the integrity of this analysis. The contribution of chemical versus physical sputtering to the source of C + at the target is assessed through simultaneous measurement of CII line, and CD plus CH-band emissions during release of CH 4 from the PPI, then compared with that seen in intrinsic sputtering.

Characterizing Tungsten Sourcing and SOL Transport during the Metal Rings Campaign

NASA Astrophysics Data System (ADS)

Thomas, D. M.; Abrams, T.; Unterberg, E. A.; Donovan, D.; Elder, J. D.; Wampler, W. R.; DIII-D Team

2017-10-01

The Metal Rings Campaign on DIII-D utilized two isotopically and poloidally distinct toroidal arrays of tungsten coated inserts in the lower divertor to study W divertor erosion near the outer strike point (OSP) and divertor entrance and subsequent migration in a mixed-material (C-W) environment. In AT hybrid discharges (PAUX = 14 MW, H98 = 1.6, βN = 3.7) with rapid ELMs (fELM 200 Hz, δW/W 0.7%) W impurities are seen to reach the midplane predominantly from the OSP region rather than the divertor entrance (far-SOL). Conversely, in scenarios with less frequent larger ELMs (fELM 60 Hz, δW/W 3.6%), the W impurities are found to transport equally from the OSP and entrance region. ELM-resolved spectroscopic measurements of W sourcing indicate that large ELMs can source W at many times the inter ELM rate. The peak W erosion rate can shift radially outwards consistent with the ELM energy flux, thereby shifting the balance between strikepoint and far-SOL sources. Changes in the peak erosion locations between forward and reversed Bt discharges are consistent with ExB ion drift effects. Evidence for a near-SOL impurity buildup between the divertors driven by the parallel grad-Ti force is also seen. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698.

Controlling marginally detached divertor plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eldon, David; Kolemen, Egemen; Barton, Joseph L.

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 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 (proportionalintegral-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. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

Progress in extrapolating divertor heat fluxes towards large fusion devices

NASA Astrophysics Data System (ADS)

Sieglin, B.; Faitsch, M.; Eich, T.; Herrmann, A.; Suttrop, W.; Collaborators, JET; the MST1 Team; the ASDEX Upgrade Team

2017-12-01

Heat load to the plasma facing components is one of the major challenges for the development and design of large fusion devices such as ITER. Nowadays fusion experiments can operate with heat load mitigation techniques, e.g. sweeping, impurity seeding, but do not generally require it. For large fusion devices however, heat load mitigation will be essential. This paper presents the current progress of the extrapolation of steady state and transient heat loads towards large fusion devices. For transient heat loads, so-called edge localized modes are considered a serious issue for the lifetime of divertor components. In this paper, the ITER operation at half field (2.65 T) and half current (7.5 MA) will be discussed considering the current material limit for the divertor peak energy fluence of 0.5 {MJ}/{{{m}}}2. Recent studies were successful in describing the observed energy fluence in the JET, MAST and ASDEX Upgrade using the pedestal pressure prior to the ELM crash. Extrapolating this towards ITER results in a more benign heat load compared to previous scalings. In the presence of magnetic perturbation, the axisymmetry is broken and a 2D heat flux pattern is induced on the divertor target, leading to local increase of the heat flux which is a concern for ITER. It is shown that for a moderate divertor broadening S/{λ }{{q}}> 0.5 the toroidal peaking of the heat flux disappears.

The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

DOE PAGES

Perkins, R. J.; Hosea, J. C.; Jaworski, M. A.; ...

2015-04-13

The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. We demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heatmore » flux transmission coefficient in the presence of the RF field. Though the precise comparison between computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. Our work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.« less

Numerical analyses of baseline JT-60SA design concepts with the COREDIV code

NASA Astrophysics Data System (ADS)

Zagórski, R.; Gałązka, K.; Ivanova-Stanik, I.; Stępniewski, W.; Garzotti, L.; Giruzzi, G.; Neu, R.; Romanelli, M.

2017-06-01

JT-60SA reference design scenarios at high (#3) and low (#2) density have been analyzed with the help of the self-consistent COREDIV code. Simulations results for a standard C wall and full W wall have been compared in terms of the influence of impurities, both intrinsic (C, W) and seeded (N, Ar, Ne, Kr), on the radiation losses and plasma parameters. For scenario #3 in a C environment, the regime of detachment on divertor plates can be achieved with N or Ne seeding, whereas for the low density and high power scenario (#2), the C and seeding impurity radiation does not effectively reduce power to the targets. In this case, only an increase of either average density or edge density together with Kr seeding might help to develop conditions with strong radiation losses and semi-detached conditions in the divertor. The calculations show that, in the case of a W divertor, the power load to the plate is mitigated by seeding and the central plasma dilution is smaller compared to the C divertor. For the high density case (#3) with Ne seeding, operation in full detachment mode is predicted. Ar seems to be an optimal choice for the low-density high-power scenario #2, showing a wide operating window, whereas Ne leads to high plasma dilution at high seeding levels albeit not achieving semi-detached conditions in the divertor.

Quiescence near the X-point of MAST measured by high speed visible imaging

NASA Astrophysics Data System (ADS)

Walkden, N. R.; Harrison, J.; Silburn, S. A.; Farley, T.; Henderson, S. S.; Kirk, A.; Militello, F.; Thornton, A.; The MAST Team

2017-12-01

Using high speed imaging of the divertor volume, the region close to the X-point in MAST is shown to be quiescent. This is confirmed by three different analysis techniques and the quiescent X-point region (QXR) spans from the separatrix to the \\psiN = 1.02 flux surface. Local reductions to the atomic density and effects associated with the camera viewing geometry are ruled out as causes of the QXR, leaving quiescence in the local plasma conditions as being the most likely cause. The QXR is found to be ubiquitous across a significant operational space in MAST including L-mode and H-mode discharges across maximal ranges of 9.8×1019~m-2 in line integrated density, 0.36 MA in plasma current, 0.11 T in toroidal magnetic field and 3.2 MW in NBI power. When mapped to the divertor target the QXR occupies approximately an e-folding length of the heat-flux profile, containing  ∼60% of the total heat flux to the target, and also shows a tendency towards higher frequency shorter lived fluctuations in the ion-saturation current. This is consistent with short-lived divertor localised filamentary structures observed further down the outer divertor leg in the camera images, and suggests a complex multi-region picture of filamentary transport in the divertor.

Controlling marginally detached divertor plasmas

DOE PAGES

Eldon, David; Kolemen, Egemen; Barton, Joseph L.; ...

2017-05-04

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 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 (proportionalintegral-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. However, the observed bifurcation in plasma conditions at the outer strike point with the ion B ×more » $$\

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

The snowflake divertor

NASA Astrophysics Data System (ADS)

Ryutov, D. D.; Soukhanovskii, V. A.

2015-11-01

The snowflake magnetic configuration is characterized by the presence of two closely spaced poloidal field nulls that create a characteristic hexagonal (reminiscent of a snowflake) separatrix structure. The magnetic field properties and the plasma behaviour in the snowflake are determined by the simultaneous action of both nulls, this generating a lot of interesting physics, as well as providing a chance for improving divertor performance. Among potential beneficial effects of this geometry are: increased volume of a low poloidal field around the null, increased connection length, and the heat flux sharing between multiple divertor channels. The authors summarise experimental results obtained with the snowflake configuration on several tokamaks. Wherever possible, relation to the existing theoretical models is described.

Suppression of tritium retention in remote areas of ITER by nonperturbative reactive gas injection.

PubMed

Tabarés, F L; Ferreira, J A; Ramos, A; van Rooij, G; Westerhout, J; Al, R; Rapp, J; Drenik, A; Mozetic, M

2010-10-22

A technique based on reactive gas injection in the afterglow region of the divertor plasma is proposed for the suppression of tritium-carbon codeposits in remote areas of ITER when operated with carbon-based divertor targets. Experiments in a divertor simulator plasma device indicate that a 4  nm/min deposition can be suppressed by addition of 1  Pa·m³ s⁻¹ ammonia flow at 10 cm from the plasma. These results bolster the concept of nonperturbative scavenger injection for tritium inventory control in carbon-based fusion plasma devices, thus paving the way for ITER operation in the active phase under a carbon-dominated, plasma facing component background.

Toroidally symmetric plasma vortex at tokamak divertor null point

DOE PAGES

Umansky, M. V.; Ryutov, D. D.

2016-03-09

Reduced MHD equations are used for studying toroidally symmetric plasma dynamics near the divertor null point. Numerical solution of these equations exhibits a plasma vortex localized at the null point with the time-evolution defined by interplay of the curvature drive, magnetic restoring force, and dissipation. Convective motion is easier to achieve for a second-order null (snowflake) divertor than for a regular x-point configuration, and the size of the convection zone in a snowflake configuration grows with plasma pressure at the null point. In conclusion, the trends in simulations are consistent with tokamak experiments which indicate the presence of enhanced transportmore » at the null point.« less

Divertor sheath power studies in DIII-D using fixed Langmuir probes and three-dimensional modeling of tile heat flows

NASA Astrophysics Data System (ADS)

Donovan, D.; Nygren, R.; Buchenauer, D.; Watkins, J.; Rudakov, D.; Leonard, A.; Wong, C. P. C.; Makowski, M.

2014-04-01

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (Jsat) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

Fast plasma shutdown by killer pellet injection in JT-60U with reduced heat flux on the divertor plate and avoiding runaway electron generation

NASA Astrophysics Data System (ADS)

Yoshino, R.; Kondoh, T.; Neyatani, Y.; Itami, K.; Kawano, Y.; Isei, N.

1997-02-01

A killer pellet is an impurity pellet that is injected into a tokamak plasma in order to terminate a discharge without causing serious damage to the tokamak machine. In JT-60U neon ice pellets have been injected into OH and NB heated plasmas and fast plasma shutdowns have been demonstrated without large vertical displacement. The heat pulse on the divertor plate has been greatly reduced by killer pellet injection (KPI), but a low-power heat flux tail with a long time duration is observed. The total energy on the divertor plate increases with longer heat flux tail, so it has been reduced by shortening the tail. Runaway electron (RE) generation has been observed just after KPI and/or in the later phase of the plasma current quench. However, RE generation has been avoided when large magnetic perturbations are excited. These experimental results clearly show that KPI is a credible fast shutdown method avoiding large vertical displacement, reducing heat flux on the divertor plate, and avoiding (or minimizing) RE generation.

ADX - Advanced Divertor and RF Tokamak Experiment

NASA Astrophysics Data System (ADS)

Greenwald, Martin; Labombard, Brian; Bonoli, Paul; Irby, Jim; Terry, Jim; Wallace, Greg; Vieira, Rui; Whyte, Dennis; Wolfe, Steve; Wukitch, Steve; Marmar, Earl

2015-11-01

The Advanced Divertor and RF Tokamak Experiment (ADX) is a design concept for a compact high-field tokamak that would address boundary plasma and plasma-material interaction physics challenges whose solution is critical for the viability of magnetic fusion energy. This device would have two crucial missions. First, it would serve as a Divertor Test Tokamak, developing divertor geometries, materials and operational scenarios that could meet the stringent requirements imposed in a fusion power plant. By operating at high field, ADX would address this problem at a level of power loading and other plasma conditions that are essentially identical to those expected in a future reactor. Secondly, ADX would investigate the physics and engineering of high-field-side launch of RF waves for current drive and heating. Efficient current drive is an essential element for achieving steady-state in a practical, power producing fusion device and high-field launch offers the prospect of higher efficiency, better control of the current profile and survivability of the launching structures. ADX would carry out this research in integrated scenarios that simultaneously demonstrate the required boundary regimes consistent with efficient current drive and core performance.

FLIT: Flowing LIquid metal Torus

NASA Astrophysics Data System (ADS)

Kolemen, Egemen; Majeski, Richard; Maingi, Rajesh; Hvasta, Michael

2017-10-01

The design and construction of FLIT, Flowing LIquid Torus, at PPPL is presented. FLIT focuses on a liquid metal divertor system suitable for implementation and testing in present-day fusion systems, such as NSTX-U. It is designed as a proof-of-concept fast-flowing liquid metal divertor that can handle heat flux of 10 MW/m2 without an additional cooling system. The 72 cm wide by 107 cm tall torus system consisting of 12 rectangular coils that give 1 Tesla magnetic field in the center and it can operate for greater than 10 seconds at this field. Initially, 30 gallons Galinstan (Ga-In-Sn) will be recirculated using 6 jxB pumps and flow velocities of up to 10 m/s will be achieved on the fully annular divertor plate. FLIT is designed as a flexible machine that will allow experimental testing of various liquid metal injection techniques, study of flow instabilities, and their control in order to prove the feasibility of liquid metal divertor concept for fusion reactors. FLIT: Flowing LIquid metal Torus. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

Coupled Kinetic-MHD Simulations of Divertor Heat Load with ELM Perturbations

NASA Astrophysics Data System (ADS)

Cummings, Julian; Chang, C. S.; Park, Gunyoung; Sugiyama, Linda; Pankin, Alexei; Klasky, Scott; Podhorszki, Norbert; Docan, Ciprian; Parashar, Manish

2010-11-01

The effect of Type-I ELM activity on divertor plate heat load is a key component of the DOE OFES Joint Research Target milestones for this year. In this talk, we present simulations of kinetic edge physics, ELM activity, and the associated divertor heat loads in which we couple the discrete guiding-center neoclassical transport code XGC0 with the nonlinear extended MHD code M3D using the End-to-end Framework for Fusion Integrated Simulations, or EFFIS. In these coupled simulations, the kinetic code and the MHD code run concurrently on the same massively parallel platform and periodic data exchanges are performed using a memory-to-memory coupling technology provided by EFFIS. The M3D code models the fast ELM event and sends frequent updates of the magnetic field perturbations and electrostatic potential to XGC0, which in turn tracks particle dynamics under the influence of these perturbations and collects divertor particle and energy flux statistics. We describe here how EFFIS technologies facilitate these coupled simulations and discuss results for DIII-D, NSTX and Alcator C-Mod tokamak discharges.

Using Divertor Strike Point Splitting to Study Plasma Response and Its Sensitivity to Equilibrium Uncertainties

NASA Astrophysics Data System (ADS)

Teklu, Abraham; Orlov, D. M.; Moyer, R. A.; Bykov, I.; Evans, T. E.; Wu, W.; Trevisan, G. L.; Lyons, B. C.; Abrams, T.; Makowski, M. A.; Lasnier, C. S.; Fenstermacher, M. E.

2017-10-01

Resonant magnetic perturbations (RMPs) from 3D coils have been varied to modify the splitting of the divertor strike points in DIII-D. This splitting is imaged in filtered visible and infrared emission from the divertor to determine the particle and heat flux patterns on the target plates. The observed splitting is compared to vacuum and plasma response modeling in discharges where a subset of the RMP coils were ramped to shift the divertor footprints from dominantly n = 3 to n = 2 pattern. These results will be used to determine if the plasma response model can be validated with the measured splitting. We will also study the sensitivity of the modeled splitting to details of the 2D equilibrium. This RMP ramp technique could be used in ITER to spread out the heat flux while avoiding excessive forces on the RMP coils. Work supported by U.S. DOE under the Science Undergraduate Laboratory Internship (SULI) program and DE-FC02-04ER54698, DE-FG02-07ER54917, DE-FG02-05ER54809 and DE-AC52-07NA27344.

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

NASA Astrophysics Data System (ADS)

Deng, Guozhong; Liu, Xiaoju; Wang, Liang; Liu, Shaocheng; Xu, Jichan; Feng, Wei; Liu, Jianbin; Liu, Huan; Gao, Xiang

2017-04-01

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λ q and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I p. Strong inverse scaling of the SOL width with I p has been achieved for both L-mode and H-mode plasmas in the forms of {λ }q,{{L}\\text-\\text{mode}}=4.98× {I}{{p}}-0.68 and {λ }q,{{H}\\text-\\text{mode}}=1.86× {I}{{p}}-1.08. Similar trends have also been demonstrated in the study of heat spreading with {S}{{L}\\text-\\text{mode}}=1.95× {I}{{p}}-0.542 and {S}{{H}\\text-\\text{mode}}=0.756× {I}{{p}}-0.872. In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR).

Overview of decade-long development of plasma-facing components at ASIPP

NASA Astrophysics Data System (ADS)

Luo, G.-N.; Liu, G. H.; Li, Q.; Qin, S. G.; Wang, W. J.; Shi, Y. L.; Xie, C. Y.; Chen, Z. M.; Missirlian, M.; Guilhem, D.; Richou, M.; Hirai, T.; Escourbiac, F.; Yao, D. M.; Chen, J. L.; Wang, T. J.; Bucalossi, J.; Merola, M.; Li, J. G.; EAST Team

2017-06-01

The first EAST (Experimental Advanced Superconducting Tokamak) plasma ignited in 2006 with non-actively cooled steel plates as the plasma-facing materials and components (PFMCs) which were then upgraded into full graphite tiles bolted onto water-cooled copper heat sinks in 2008. The first wall was changed further into molybdenum alloy in 2012, while keeping the graphite for both the upper and lower divertors. With the rapid increase in heating and current driving power in EAST, the W/Cu divertor project was launched around the end of 2012, aiming at achieving actively cooled full W/Cu-PFCs for the upper divertor, with heat removal capability up to 10 MW m-2. The W/Cu upper divertor was finished in the spring of 2014, consisting of 80 cassette bodies toroidally assembled. Commissioning of the EAST upper W/Cu divertor in 2014 was unsatisfactory and then several practical measures were implemented to improve the design, welding quality and reliability, which helped us achieve successful commissioning in the 2015 Spring Campaign. In collaboration with the IO and CEA teams, we have demonstrated our technological capability to remove heat loads of 5000 cycles at 10 MW m-2 and 1000 cycles at 20 MW m-2 for the small scale monoblock mockups, and surprisingly over 300 cycles at 20 MW m-2 for the flat-tile ones. The experience and lessons we learned from batch production and commissioning are undoubtedly valuable for ITER (International Thermonuclear Experimental Reactor) engineering validation and tungsten-related plasma physics.

Divertor-localized fluctuations in NSTX-U L-mode discharges

NASA Astrophysics Data System (ADS)

Scotti, Filippo; Soukhanovskii, V. A.; Zweben, S.; Myra, J.; Baver, D.; Sabbagh, S. A.

2017-10-01

The 3-D structure of divertor turbulence is characterized in NSTX-U by means of fast camera imaging. Edge and divertor turbulence can be important in determining the heat flux width in fusion devices. Field-aligned filaments are found on the divertor legs via imaging of C III and D- α emission in NBI-heated diverted L-mode discharges, similar to observations in Alcator C-Mod and MAST. These flute-like fluctuations of up to 10-20% in RMS/mean are radially localized around the separatrix and limited to the region below the X-point. Poloidal and parallel correlation lengths are a few cm (10-50ρi) and several meters, respectively. For the outer leg filaments, poloidal correlation lengths decrease along the leg away from the strike point and typical effective toroidal mode numbers are in the range of 10-20. Opposite toroidal rotation is observed for inner (co-current rotation) and outer leg (counter-current rotation) filaments with apparent poloidal propagation of 1 km/s. The poloidal motion of outer leg filaments is opposite to the one typically observed for NSTX upstream blobs in the scrape-off layer. The shape, dynamics and absence of correlation with upstream turbulence suggest that these fluctuations are generated and localized in the divertor region. Supported by US DOE DE-AC52-07NA27344, DE-AC02-09CH11466, DE-FG02- 02ER54678, DE-FG02-99ER54524.

Controlling marginally detached divertor plasmas

NASA Astrophysics Data System (ADS)

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   ×  \

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lumsdaine, A.; Bjorholm, T.; Harris, J.

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Lesson from Tungsten Leading Edge Heat Load Analysis in KSTAR Divertor

NASA Astrophysics Data System (ADS)

Hong, Suk-Ho; Pitts, Richard Anthony; Lee, Hyeong-Ho; Bang, Eunnam; Kang, Chan-Soo; Kim, Kyung-Min; Kim, Hong-Tack; ITER Organization Collaboration; Kstar Team Team

2016-10-01

An important design issue for the ITER tungsten (W) divertor and in fact for all such components using metallic plasma-facing elements and which are exposed to high parallel power fluxes, is the question of surface shaping to avoid melting of leading edges. We have fabricated a series of tungsten blocks with a variety of leading edge heights (0.3, 0.6, 1.0, and 2.0 mm), from the ITER worst case to heights even beyond the extreme value tested on JET. They are mounted into adjacent, inertially cooled graphite tile installed in the central divertor region of KSTAR, within the field of view of an infra-red (IR) thermography system with a spatial resolution to 0.4 mm/pixel. Adjustment of the outer divertor strike point position is used to deposit power on the different blocks in different discharges. The measured power flux density on flat regions of the surrounding graphite tiles is used to obtain the parallel power flux, q|| impinging on the various W blocks. Experiments have been performed in Type I ELMing H-mode with Ip = 600 kA, BT = 2 T, PNBI = 3.5 MW, leading to a hot attached divertor with typical pulse lengths of 10 s. Three dimensional ANSYS simulations using q|| and assuming geometric projection of the heat flux are found to be consistent with the observed edge loading. This research was partially supported by Ministry of Science, ICT, and Future Planning under KSTAR project.

Overview of design and analysis activities for the W7-X scraper element

DOE PAGES

Lumsdaine, A.; Bjorholm, T.; Harris, J.; ...

2016-08-18

The Wendelstein 7-X stellarator is in final stages of commissioning, and will begin operation in late 2015. In the first phase, the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current will evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A highmore » heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a test divertor unit scraper element (TDU-SE) in 2017 during an earlier operational phase. Several U.S. fusion energy science laboratories have been involved in the design, analysis (structural and thermal finite element, as well as computational fluid dynamics), plasma simulation, planning, prototyping, and diagnostic development around the scraper element program (both TDU-SE and HHF-SE). As a result, this paper presents an overview of all of these activities and their current status.« less

Changes in divertor conditions in response to changing core density with RMPs

DOE PAGES

Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.; ...

2017-06-07

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicated non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have components which has at least one but typically many resonances with the rotational transform of the plasma. RMPs are found to alter inter-ELM heat flux to the divertormore » by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that, local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. Furthermore, these trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied.« less

Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

DOE PAGES

Wang, H. Q.; Guo, Houyang Y.; Petrie, Thomas W.; ...

2017-03-18

The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N 2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N 2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss powermore » to the L-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection leads to an increase in core radiation, up to 50% of the injected power, and a reduction in pedestal temperature over 60%, thus significantly degrading the confinement, i.e., with H 98 reducing from 1.1 to below 0.7. As for Ne seeding, H 98 near 0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N 2 seeded plasmas, radiation is predominately confined in the boundary plasma, with up to 50% of heating power being radiated in the divertor region and less than 25% in the core at the onset of detachment. In the case of strong N 2 gas puffing, the confinement recovers during the detachment, from ~20% reduction at the onset of the detachment to greater than that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.« less

Fabrication of divertor mock-up with ODS-Cu and W by the improved brazing technique

NASA Astrophysics Data System (ADS)

Tokitani, M.; Hamaji, Y.; Hiraoka, Y.; Masuzaki, S.; Tamura, H.; Noto, H.; Tanaka, T.; Muroga, T.; Sagara, A.; FFHR Design Group

2017-07-01

Copper alloy has been considered as a divertor cooling tube or heat sink not only in the helical reactor FFHR-d1 but also in the tokamak DEMO reactor, because it has a high thermal conductivity. This work focused on applying an oxide dispersion strengthened copper alloy (ODS-Cu), GlidCop® (Cu-0.3 wt%Al2O3) as the divertor heat sink material of FFHR-d1. This alloy has superior high temperature yield strength exceeding 300 MPa at room temperature even after annealing up to ~1000 °C. The change in material properties of Pure-Cu, GlidCop® and CuCrZr by neutron irradiation are summarized in this paper. A primary dose limit is the radiation-induced hardening/softening (~0.2 dpa/1-2 dpa) which has a temperature dependence. According to such an evaluation, the GlidCop® can be selected as the current best candidate material in the commercial base of the divertor heat sink, and its temperature should be maintained as close as possible to 300 °C during operation. Bonding between the W armour and the GlidCop® heat sink was successfully performed by using an improved brazing technique with BNi-6 (Ni-11%P) filler material. The bonding strength was measured by a three-point bending test and reached up to approximately 200 MPa. Surprisingly, several specimens showed an obvious yield point. This means that the BNi-6 brazing (bonding) layer caused relaxation of the applied stress. The small-scale divertor mock-up of the W/BNi-6/GlidCop® was successfully fabricated by using the improved brazing technique. The heat loading test was carried out by the electron beam device ACT2 in NIFS. The mock-up showed an excellent heat removal capability for use in the FFHR-d1 divertor.

Changes in divertor conditions in response to changing core density with RMPs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Briesemeister, Alexis R.; Ahn, Joon -Wook; Canik, John M.

The effects of changes in core density on divertor electron temperature, density and heat flux when resonant magnetic perturbations (RMPs) are applied are presented, notably a reduction in RMP induced secondary radial peaks in the electron temperature profile at the target plate is observed when the core density is increased, which is consistent with modeling. RMPs is used here to indicated non-axisymmetric magnetic field perturbations, created using in-vessel control coils, which have components which has at least one but typically many resonances with the rotational transform of the plasma. RMPs are found to alter inter-ELM heat flux to the divertormore » by modifying the core plasma density. It is shown that applying RMPs reduces the core density and increases the inter-ELM heat flux to both the inner and outer targets. Using gas puffing to return the core density to the pre-RMP levels more than eliminates the increase in inter-ELM heat flux, but a broadening of the heat flux to the outer target remains. These measurements were made at a single toroidal location, but the peak in the heat flux profile was found near the outer strike point where simulations indicate little toroidal variation should exist and tangentially viewing diagnostics showed no evidence of strong asymmetries. In experiments where divertor Thomson scattering measurements were available it is shown that, local secondary peaks in the divertor electron temperature profile near the target plate are reduced as the core density is increased, while peaks in the divertor electron density profile near the target are increased. Furthermore, these trends observed in the divertor electron temperature and density are qualitatively reproduced by scanning the upstream density in EMC3-Eirene modeling. Measurements are presented showing that higher densities are needed to induce detachment of the outer strike point in a case where an increase in electron temperature, likely due to a change in MHD activity, is seen after RMPs are applied.« less

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

NASA Astrophysics Data System (ADS)

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 with a fatigue life of approximately 0.55 years based on the expected operation scenario of a prototype or test reactor. This study extends the state of knowledge of the technological limit of a divertor based on a RAFM steel pipe structure.

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE PAGES

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.; ...

2017-07-12

Here, 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. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » 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 lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki; Majeski, Richard P.; Jaworski, Michael A.

Here, 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. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peakmore » 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 lithium divertor (RLLD) concept and its variant, the active liquid lithium divertor concept (ARLLD), taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned. We examined two key technology issues: 1) dust or solid particle removal and 2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust / impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~ 1 l/sec LL flow, even a small 0.1% dust content by weight (or 0.5 g per sec) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW fusion power) fusion power plant, about 0.5 g / sec of tritium is needed to maintain the fusion fuel cycle assuming ~ 1 % fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap (SCT) concept.« less

Liquid lithium loop system to solve challenging technology issues for fusion power plant

NASA Astrophysics Data System (ADS)

Ono, M.; Majeski, R.; Jaworski, M. A.; Hirooka, Y.; Kaita, R.; Gray, T. K.; Maingi, R.; Skinner, C. H.; Christenson, M.; Ruzic, D. N.

2017-11-01

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. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance. 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 lithium divertor concept and its variant, the active liquid lithium divertor concept, taking advantage of the enhanced or non-coronal Li radiation in relatively poorly confined divertor plasmas. To maintain the LL purity in a 1 GW-electric class fusion power plant, a closed LL loop system with a modest circulating capacity of ~1 l s-1 is envisioned. We examined two key technology issues: (1) dust or solid particle removal and (2) real time recovery of tritium from LL while keeping the tritium inventory level to an acceptable level. By running the LL-loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to the outside where the dust/impurities can be removed by relatively simple dust filter, cold trap and/or centrifugal separation systems. With ~1 l s-1 LL flow, even a small 0.1% dust content by weight (or 0.5 g s-1) suggests that the LL-loop could carry away nearly 16 tons of dust per year. In a 1 GW-electric (or ~3 GW fusion power) fusion power plant, about 0.5 g s-1 of tritium is needed to maintain the fusion fuel cycle assuming ~1% fusion burn efficiency. It appears feasible to recover tritium (T) in real time from LL while maintaining an acceptable T inventory level. Laboratory tests are being conducted to investigate T recovery feasibility with the surface cold trap concept.

Lithium As Plasma Facing Component for Magnetic Fusion Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Masayuki Ono

The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor ofmore » two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m2 . By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective flexible means to improve the fusion reactor performance, while providing a practical solution to the highly challenging divertor heat handling issue confronting the steadystate magnetic fusion reactors.« less

Numerical study of the current-convective instability driven by asymmetry of detachment in inner and outer divertors

NASA Astrophysics Data System (ADS)

Stepanenko, A. A.; Krasheninnikov, S. I.

2018-01-01

One of the possible mechanisms responsible for strong radiation fluctuations observed in recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nucl. Fusion 54, 013001 (2014)] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer divertors of the tokamak [S. Krasheninnikov and A. Smolyakov, Phys. Plasmas 23, 092505 (2016)]. In this study, we present the physical model, used to simulate the CCI, and the first numerical results of modeling of the CCI dynamics in ASDEX Upgrade-like conditions. The simulation results provide frequency spectra of turbulent divertor plasma oscillations showing reasonably good agreement with the available experimental data.

Surface heat flux feedback controlled impurity seeding experiments with Alcator C-Mod’s high-Z vertical target plate divertor: performance, limitations and implications for fusion power reactors

NASA Astrophysics Data System (ADS)

Brunner, D.; Wolfe, S. M.; LaBombard, B.; Kuang, A. Q.; Lipschultz, B.; Reinke, M. L.; Hubbard, A.; Hughes, J.; Mumgaard, R. T.; Terry, J. L.; Umansky, M. V.; The Alcator C-Mod Team

2017-08-01

The Alcator C-Mod team has recently developed a feedback system to measure and control surface heat flux in real-time. The system uses real-time measurements of surface heat flux from surface thermocouples and a pulse-width modulated piezo valve to inject low-Z impurities (typically N2) into the private flux region. It has been used in C-Mod to mitigate peak surface heat fluxes  >40 MW m-2 down to  <10 MW m-2 while maintaining excellent core confinement, H 98  >  1. While the system works quite well under relatively steady conditions, use of it during transients has revealed important limitations on feedback control of impurity seeding in conventional vertical target plate divertors. In some cases, the system is unable to avoid plasma reattachment to the divertor plate or the formation of a confinement-damaging x-point MARFE. This is due to the small operational window for mitigated heat flux in the parameters of incident plasma heat flux, plasma density, and impurity density as well as the relatively slow response of the impurity gas injection system compared to plasma transients. Given the severe consequences for failure of such a system to operate reliably in a reactor, there is substantial risk that the conventional vertical target plate divertor will not provide an adequately controllable system in reactor-class devices. These considerations motivate the need to develop passively stable, highly compliant divertor configurations and experimental facilities that can test such possible solutions.

Achievement of radiative feedback control for long-pulse operation on EAST

NASA Astrophysics Data System (ADS)

Wu, K.; Yuan, Q. P.; Xiao, B. J.; Wang, L.; Duan, Y. M.; Chen, J. B.; Zheng, X. W.; Liu, X. J.; Zhang, B.; Xu, J. C.; Luo, Z. P.; Zang, Q.; Li, Y. Y.; Feng, W.; Wu, J. H.; Yang, Z. S.; Zhang, L.; Luo, G.-N.; Gong, X. Z.; Hu, L. Q.; Hu, J. S.; Li, J.

2018-05-01

The active feedback control of radiated power to prevent divertor target plates overheating during long-pulse operation has been developed and implemented on EAST. The radiation control algorithm, with impurity seeding via a supersonic molecular beam injection (SMBI) system, has shown great success in both reliability and stability. By seeding a sequence of short neon (Ne) impurity pulses with the SMBI from the outer mid-plane, the radiated power of the bulk plasma can be well controlled, and the duration of radiative control (feedforward and feedback) is 4.5 s during a discharge of 10 s. Reliable control of the total radiated power of bulk plasma has been successfully achieved in long-pulse upper single null (USN) discharges with a tungsten divertor. The achieved control range of {{f}rad} is 20%–30% in L-mode regimes and 18%–36% in H-mode regimes. The temperature of the divertor target plates was maintained at a low level during the radiative control phase. The peak particle flux on the divertor target was decreased by feedforward Ne injection in the L-mode discharges, while the Ne pulses from the SMBI had no influence on the peak particle flux because of the very small injecting volume. It is shown that although the radiated power increased, no serious reduction of plasma-stored energy or confinement was observed during the control phase. The success of the radiation control algorithm and current experiments in radiated power control represents a significant advance for steady-state divertor radiation and heat flux control on EAST for near-future long-pulse operation.

High-resolution disruption halo current measurements using Langmuir probes in Alcator C-Mod

NASA Astrophysics Data System (ADS)

Tinguely, R. A.; Granetz, R. S.; Berg, A.; Kuang, A. Q.; Brunner, D.; LaBombard, B.

2018-01-01

Halo currents generated during disruptions on Alcator C-Mod have been measured with Langmuir ‘rail’ probes. These rail probes are embedded in a lower outboard divertor module in a closely-spaced vertical (poloidal) array. The dense array provides detailed resolution of the spatial dependence (~1 cm spacing) of the halo current distribution in the plasma scrape-off region with high time resolution (400 kHz digitization rate). As the plasma limits on the outboard divertor plate, the contact point is clearly discernible in the halo current data (as an inversion of current) and moves vertically down the divertor plate on many disruptions. These data are consistent with filament reconstructions of the plasma boundary, from which the edge safety factor of the disrupting plasma can be calculated. Additionally, the halo current ‘footprint’ on the divertor plate is obtained and related to the halo flux width. The voltage driving halo current and the effective resistance of the plasma region through which the halo current flows to reach the probes are also investigated. Estimations of the sheath resistance and halo region resistivity and temperature are given. This information could prove useful for modeling halo current dynamics.

Low temperature tungsten spectroscopy on a Penning Ionization Discharge

NASA Astrophysics Data System (ADS)

Kumar, Deepak; Englesbe, Alexander; Stutman, Dan; Finkenthal, Michael

2011-10-01

Complete Tungsten divertor operation is being planned on many tokamaks including Tore Supra and ITER. Thus, low temperature tungsten spectroscopy is important for aiding the divertor diagnostics on larger machines. A Penning Ionization Discharge (PID) at the Johns Hopkins University produces steady state plasmas with Te ~ 2 eV, ne ~1013 cm-3 and a fast electron fraction at ~ 10 s eV. Similar bi-Maxwellian distributions, but with slightly higher electron temperatures, are found in the divertor plasmas of tokamaks. The two significant populating mechanisms for higher charge states in the PID are: (a) collisional excitation from bulk electrons, and (b) inner shell ionization from the fast electrons. The PID is diagnosed in a wide wavelength range - XUV, VUV and visible, to differentiate the two populating mechanisms. W is introduced in the PID by the sputtering of cathodes made of CuW alloy. Spectral emission from significantly higher charge states of W (up to W IV) has been observed in the experiment. This poster will describe results indicating the populating mechanism of W ions and also describe plans on upgrading the experiment to achieve higher temperatures which are closer to the divertor conditions. Supported by USDOE.

Carbon Deposition in the Inner JET Divertor Measured by Means of Quartz Microbalance

NASA Astrophysics Data System (ADS)

Esser, H. G.; Philipps, V.; Freisinger, M.; Coad, P.; Matthews, G. F.; Neill, G.; JET EFDA Contributors

A Quartz Microbalance (QMB) system was implemented in the inner divertor region of JET in order to measure in situ and time resolved (minimum exposure time ≥0.1 s) material fluxes (mainly carbon) and layer deposition. The system has been developed to operate at temperatures up to 200°C. The aim is to investigate carbon transport to the remote areas, and hence the tritium retention in dependence on plasma conditions. This question is still a major concern for the ITER operation. The mass sensitivity of the system is Sm = 1.5 A~— 10-8 [g/Hz cm2]. First reliable measurements were made during the C5 campaign (March–May 2002; â‰e1000 plasma discharges). The results presented are based on 74 selected exposures (694 s) under various conditions (strike point position, input power, neutral pressure, ELM frequency). Most influencing on the carbon deposition in the remote area seems to be the geometry i.e. the strike point position on the divertor tiles. In average 1.9 A~— 10-4 C-atom are deposited per deuterium ion flowing into the inner divertor.

Characterization of edge turbulence in different states of divertor detachment using reflectometry in the ASDEX Upgrade tokamak

NASA Astrophysics Data System (ADS)

Nikolaeva, V.; Guimarais, L.; Manz, P.; Carralero, D.; Manso, M. E.; Stroth, U.; Silva, C.; Conway, G. D.; Seliunin, E.; Vicente, J.; Brida, D.; Aguiam, D.; Santos, J.; Silva, A.; ASDEX Upgrade team; MST1 team

2018-05-01

Transport in the scrape-off layer (SOL) depends on the state of divertor detachment. L-mode discharges were analyzed where the state of divertor detachment is varied through a density ramp-up. By means of reflectometry measurements at the low (LFS) and the high field side (HFS), midplane density fluctuations are studied for the first time in ASDEX Upgrade simultaneously at both sides of the tokamak. Radial density fluctuation profiles (δ {n}e/{n}e) increase with radius in both the HFS and the LFS. It is found that in the SOL density fluctuations at the LFS have about a factor of two larger amplitude than at the HFS in agreement with ballooned transport. Density fluctuations at the LFS show a modest variation with increasing background density resulting mainly from a rise of low frequency components. Experimental results are in good agreement with an enhanced convection of filaments at the LFS at the beginning of outer divertor detachment leading to a flatter SOL density profile. In this phase of the discharge, density fluctuations measured at the HFS far-SOL display a strong increase, which may be associated with the presence of faster filaments originated at the LFS.

Heat removal capability of divertor coaxial tube assembly

NASA Astrophysics Data System (ADS)

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.

Overview of the JET results

NASA Astrophysics Data System (ADS)

Romanelli, F.; JET Contributors,

2015-10-01

Since the installation of an ITER-like wall, the JET programme has focused on the consolidation of ITER design choices and the preparation for ITER operation, with a specific emphasis given to the bulk tungsten melt experiment, which has been crucial for the final decision on the material choice for the day-one tungsten divertor in ITER. Integrated scenarios have been progressed with the re-establishment of long-pulse, high-confinement H-modes by optimizing the magnetic configuration and the use of ICRH to avoid tungsten impurity accumulation. Stationary discharges with detached divertor conditions and small edge localized modes have been demonstrated by nitrogen seeding. The differences in confinement and pedestal behaviour before and after the ITER-like wall installation have been better characterized towards the development of high fusion yield scenarios in DT. Post-mortem analyses of the plasma-facing components have confirmed the previously reported low fuel retention obtained by gas balance and shown that the pattern of deposition within the divertor has changed significantly with respect to the JET carbon wall campaigns due to the absence of thermally activated chemical erosion of beryllium in contrast to carbon. Transport to remote areas is almost absent and two orders of magnitude less material is found in the divertor.

Influence of impurity seeding on the plasma radiation in the EAST tokamak

NASA Astrophysics Data System (ADS)

Liping, DONG; Yanmin, DUAN; Kaiyun, CHEN; Xiuda, YANG; Ling, ZHANG; Feng, XU; Jingbo, CHEN; Songtao, MAO; Zhenwei, WU; Liqun, HU

2018-04-01

Plasma radiation characteristics in EAST argon (Ar) gas and neon (Ne) gas seeding experiments are studied. The radiation profiles reconstructed from the fast bolometer measurement data by tomography method are compared with the ones got from the simulation program based on corona model. And the simulation results coincide roughly with the experimental data. For Ar seeding discharges, the substantial enhanced radiations can be generally observed in the edge areas at normalized radius ρ pol∼0.7–0.9, while the enhanced regions are more outer for Ne seeding discharges. The influence of seeded Ar gas on the core radiation is related to the injected position. In discharges with LSN divertor configuration, the Ar ions can permeate into the core region more easily when being injected from the opposite upper divertor ports. In USN divertor configuration, the W impurity sputtered from the upper divertor target plates are observed to be an important contributor to the increase of the core radiation no matter impurity seeding from any ports. The maximum radiated power fractions f rad (P rad/P heat) about 60%–70% have been achieved in the recent EAST experimental campaign in 2015–2016.

Rapid change of blob structure in the outer scrape-off layer (SOL)

NASA Astrophysics Data System (ADS)

Cohen, R. H.

2005-10-01

Nonlinear structures (``blobs'') driven by the magnetic field curvature and highly elongated along the field lines may exist in the tokamak SOL.footnotetextS.I. Krasheninnikov. Phys. Lett. A 283, 368 (2001) The contact of the blob end with the divertor plate significantly affects the blob structure and velocity. However, the strong shearing of the flux-tube near the X-point makes impossible direct electrical contact of the blob in the upper SOL and the divertor, so that the sheath boundary condition (BC) has to be replaced by a BC imposed near the X point.footnotetextD. Ryutov, R.H. Cohen. Contr. Pl. Phys 44, 168 (2004) We show that, at larger distances from the separatrix, in the far SOL, the connection between the upper SOL and the divertor plate is re-established, and the sheath BC becomes again relevant. During the blob's outward radial motion, this event is reflected in a sudden change of its length, from the blob extending only to the X point to the blob extending down to the plate. Likewise, a blob initially existing only in the divertor leg becomes suddenly longer, and extends to the whole SOL.

Automated divertor target design by adjoint shape sensitivity analysis and a one-shot method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dekeyser, W., E-mail: Wouter.Dekeyser@kuleuven.be; Reiter, D.; Baelmans, M.

As magnetic confinement fusion progresses towards the development of first reactor-scale devices, computational tokamak divertor design is a topic of high priority. Presently, edge plasma codes are used in a forward approach, where magnetic field and divertor geometry are manually adjusted to meet design requirements. Due to the complex edge plasma flows and large number of design variables, this method is computationally very demanding. On the other hand, efficient optimization-based design strategies have been developed in computational aerodynamics and fluid mechanics. Such an optimization approach to divertor target shape design is elaborated in the present paper. A general formulation ofmore » the design problems is given, and conditions characterizing the optimal designs are formulated. Using a continuous adjoint framework, design sensitivities can be computed at a cost of only two edge plasma simulations, independent of the number of design variables. Furthermore, by using a one-shot method the entire optimization problem can be solved at an equivalent cost of only a few forward simulations. The methodology is applied to target shape design for uniform power load, in simplified edge plasma geometry.« less

Development of high poloidal beta, steady-state scenario with ITER-like tungsten divertor on EAST

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Gong, X. Z.; Qian, J.; Chen, J.; Li, G.; Li, K.; Li, M. H.; Zhai, X.; Bonoli, P.; Brower, D.; Cao, L.; Cui, L.; Ding, S.; Ding, W. X.; Guo, W.; Holcomb, C.; Huang, J.; Hyatt, A.; Lanctot, M.; Lao, L. L.; Liu, H.; Lyu, B.; McClenaghan, J.; Peysson, Y.; Ren, Q.; Shiraiwa, S.; Solomon, W.; Zang, Q.; Wan, B.

2017-07-01

Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H 98y2 ~ 1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drive (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.

Impact of the impurity seeding for divertor protection on the performance of fusion reactors

NASA Astrophysics Data System (ADS)

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.

Analysis of a multi-machine database on divertor heat fluxesa)

NASA Astrophysics Data System (ADS)

Makowski, M. A.; Elder, D.; Gray, T. K.; LaBombard, B.; Lasnier, C. J.; Leonard, A. W.; Maingi, R.; Osborne, T. H.; Stangeby, P. C.; Terry, J. L.; Watkins, J.

2012-05-01

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D, and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip, which all three tokamaks independently demonstrate. An improved Thomson scattering system on DIII-D has yielded very accurate scrape off layer (SOL) profile measurements from which tests of parallel transport models have been made. It is found that a flux-limited model agrees best with the data at all collisionalities, while a Spitzer resistivity model agrees at higher collisionality where it is more valid. The SOL profile measurements and divertor heat flux scaling are consistent with a heuristic drift based model as well as a critical gradient model.

Development of high poloidal beta, steady-state scenario with ITER-like tungsten divertor on EAST

DOE PAGES

Garofalo, Andrea M.; Gong, X. Z.; Qian, J.; ...

2017-06-07

Recent experiments on EAST have achieved the first long pulse H-mode (61 s) with zero loop voltage and an ITER-like tungsten divertor, and have demonstrated access to broad plasma current profiles by increasing the density in fully-noninductive lower hybrid current-driven discharges. These long pulse discharges reach wall thermal and particle balance, exhibit stationary good confinement (H 98y2~1.1) with low core electron transport, and are only possible with optimal active cooling of the tungsten armors. In separate experiments, the electron density was systematically varied in order to study its effect on the deposition profile of the external lower hybrid current drivemore » (LHCD), while keeping the plasma in fully-noninductive conditions and with divertor strike points on the tungsten divertor. A broadening of the current profile is found, as indicated by lower values of the internal inductance at higher density. A broad current profile is attractive because, among other reasons, it enables internal transport barriers at large minor radius, leading to improved confinement as shown in companion DIII-D experiments. These experiments strengthen the physics basis for achieving high performance, steady state discharges in future burning plasmas.« less

Effects of low and high mode number tearing modes in divertor tokamaks

NASA Astrophysics Data System (ADS)

Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

2007-08-01

The topological effects of magnetic perturbations on a divertor tokamak, such as DIII-D, are studied using field-line maps that were developed by Punjabi et al. [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)]. The studies consider both long-wavelength perturbations, such as those of m =1, n =1 tearing modes, and localized perturbations, which are represented as a magnetic dipole. The parameters of the dipole map are set using DIII-D data from shot 115467 in which the C-coils were activated [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The long-wavelength perturbations alter the structure of the interception of magnetic field lines with the divertor plates, but the interception is in sharp lines. The dipole perturbations cause a spreading of the interception of the field lines with the divertor plates, which alleviates problems associated with heat deposition. Magnetic field lines are the trajectories of a one-and-a-half degree of freedom Hamiltonian, which strongly constrains the topological features of the lines. Although the field line maps that we use do not accurately represent the trajectories through ordinary space of individual field lines, they do represent their topological structure.

Preliminary Results on the Heat Deposition on Divertor Plate using Low MN Map

NASA Astrophysics Data System (ADS)

Ali, Halima; Punjabi, Alkesh; Boozer, Allen

2003-10-01

The study of magnetic field line behavior and the closely related plasma behavior are important not only for their tokamak application but also for their application to other Hamiltonian, or near-Hamiltonian, systems. The behavior of field lines near a tokamak separatrix has been studied extensively using various approaches. Our approach is called method of maps. In this paper, we introduce an area-preserving map called Low MN map. We first derive the map from the general theory of maps /1/, and then use it to calculate the effects of m = 1, n = 1 perturbations on the stochastic layer and magnetic footprint in single-null divertor tokamaks. We show that there are self-similarities, singularities, and topological equivalences in the pattern of physical parameters that characterize the stochastic layer and the magnetic footprint. Preliminary results in the investigation on the heat distribution on the divertor plate indicate multiple peaked in heat flux profile distributed radially across the divertor target when the amplitude is 10-3. This, and other features, are in good agreement with experimental observations. This work is done under the DOE grant number DE-FG02-01ER54624. 1. A. Punjabi et al, J. Plasma Phys. 52, 91 (1994).

Two-point modeling of SOL losses of HHFW power in NSTX

NASA Astrophysics Data System (ADS)

Kish, Ayden; Perkins, Rory; Ahn, Joon-Wook; Diallo, Ahmed; Gray, Travis; Hosea, Joel; Jaworski, Michael; Kramer, Gerrit; Leblanc, Benoit; Sabbagh, Steve

2017-10-01

High-harmonic fast-wave (HHFW) heating is a heating and current-drive scheme on the National Spherical Torus eXperiment (NSTX) complimentary to neutral beam injection. Previous experiments suggest that a significant fraction, up to 50%, of the HHFW power is promptly lost to the scrape-off layer (SOL). Research indicates that the lost power reaches the divertor via wave propagation and is converted to a heat flux at the divertor through RF rectification rather than heating the SOL plasma at the midplane. This counter-intuitive hypothesis is investigated using a simplified two-point model, relating plasma parameters at the divertor to those at the midplane. Taking measurements at the divertor region of NSTX as input, this two-point model is used to predict midplane parameters, using the predicted heat flux as an indicator of power input to the SOL. These predictions are compared to measurements at the midplane to evaluate the extent to which they are consistent with experiment. This work was made possible by funding from the Department of Energy for the Summer Undergraduate Laboratory Internship (SULI) program. This work is supported by the US DOE Contract No. DE-AC02-09CH11466.

Measurements of tungsten migration in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Wampler, W. R.; Rudakov, D. L.; Watkins, J. G.; McLean, A. G.; Unterberg, E. A.; Stangeby, P. C.

2017-12-01

An experimental study of migration of tungsten in the DIII-D divertor is described, in which the outer strike point of L-mode plasmas was positioned on a toroidal ring of tungsten-coated metal inserts. Net deposition of tungsten on the divertor just outside the strike point was measured on graphite samples exposed to various plasma durations using the divertor materials evaluation system. Tungsten coverage, measured by Rutherford backscattering spectroscopy (RBS), was found to be low and nearly independent of both radius and exposure time closer to the strike point, whereas farther from the strike point the W coverage was much larger and increased with exposure time. Depth profiles from RBS show this was due to accumulation of thicker mixed-material deposits farther from the strike point where the plasma temperature is lower. These results are consistent with a low near-surface steady-state coverage on graphite undergoing net erosion, and continuing accumulation in regions of net deposition. This experiment provides data needed to validate, and further improve computational simulations of erosion and deposition of material on plasma-facing components and transport of impurities in magnetic fusion devices. Such simulations are underway and will be reported later.

Predictive modelling of JT-60SA high-beta steady-state plasma with impurity accumulation

NASA Astrophysics Data System (ADS)

Hayashi, N.; Hoshino, K.; Honda, M.; Ide, S.

2018-06-01

The integrated modelling code TOPICS has been extended to include core impurity transport, and applied to predictive modelling of JT-60SA high-beta steady-state plasma with the accumulation of impurity seeded to reduce the divertor heat load. In the modelling, models and conditions are selected for a conservative prediction, which considers a lower bound of plasma performance with the maximum accumulation of impurity. The conservative prediction shows the compatibility of impurity seeding with core plasma with high-beta (β N  >  3.5) and full current drive conditions, i.e. when Ar seeding reduces the divertor heat load below 10 MW m‑2, its accumulation in the core is so moderate that the core plasma performance can be recovered by additional heating within the machine capability to compensate for Ar radiation. Due to the strong dependence of accumulation on the pedestal density gradient, high separatrix density is important for the low accumulation as well as the low divertor heat load. The conservative prediction also shows that JT-60SA has enough capability to explore the divertor heat load control by impurity seeding in high-beta steady-state plasmas.

Extreme Ultraviolet Spectra of Few-Times Ionized Tungsten for Divertor Plasma Diagnostics

DOE PAGES

Clementson, Joel; Lennartsson, Thomas; Beiersdorfer, Peter

2015-09-09

The extreme ultraviolet (EUV) emission from few-times ionized tungsten atoms has been experimentally studied at the Livermore electron beam ion trap facility. The ions were produced and confined during low-energy operations of the EBIT-I electron beam ion trap. By varying the electron-beam energy from around 30–300 eV, tungsten ions in charge states expected to be abundant in tokamak divertor plasmas were excited, and the resulting EUV emission was studied using a survey spectrometer covering 120–320 Å. It is found that the emission strongly depends on the excitation energy; below 150 eV, it is relatively simple, consisting of strong isolated linesmore » from a few charge states, whereas at higher energies, it becomes very complex. For divertor plasmas with tungsten impurity ions, this emission should prove useful for diagnostics of tungsten flux rates and charge balance, as well as for radiative cooling of the divertor volume. Several lines in the 194–223 Å interval belonging to the spectra of five- and seven-times ionized tungsten (Tm-like W VI and Ho-like W VIII) were also measured using a high-resolution spectrometer.« less

The near infrared imaging system for the real-time protection of the JET ITER-like wall

NASA Astrophysics Data System (ADS)

Huber, A.; Kinna, D.; Huber, V.; Arnoux, G.; Balboa, I.; Balorin, C.; Carman, P.; Carvalho, P.; Collins, S.; Conway, N.; McCullen, P.; Jachmich, S.; Jouve, M.; Linsmeier, Ch; Lomanowski, B.; Lomas, P. J.; Lowry, C. G.; Maggi, C. F.; Matthews, G. F.; May-Smith, T.; Meigs, A.; Mertens, Ph; Nunes, I.; Price, M.; Puglia, P.; Riccardo, V.; Rimini, F. G.; Sergienko, G.; Tsalas, M.; Zastrow, K.-D.; contributors, JET

2017-12-01

This paper describes the design, implementation and operation of the near infrared (NIR) imaging diagnostic system of the JET ITER-like wall (JET-ILW) plasma experiment and its integration into the existing JET protection architecture. The imaging system comprises four wide-angle views, four tangential divertor views, and two top views of the divertor covering 66% of the first wall and up to 43% of the divertor. The operation temperature ranges which must be observed by the NIR protection cameras are, for the materials used on JET: Be 700 °C-1400 °C W coating 700 °C-1370 °C W bulk 700 °C-1400 °C. The Real-Time Protection system operates routinely since 2011 and successfully demonstrated its capability to avoid the overheating of the main chamber beryllium wall as well as of the divertor W and W-coated carbon fibre composite (CFC) tiles. During this period, less than 0.5% of the terminated discharges were aborted by a malfunction of the system. About 2%-3% of the discharges were terminated due to the detection of actual hot spots.

Predictions of VRF on a Langmuir Probe under the RF Heating Spiral on the Divertor Floor on NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosea, J C; Perkins, R J; Jaworski, M A

RF heating deposition spirals are observed on the divertor plates on NSTX as shown in for a NB plus RF heating case. It has been shown that the RF spiral is tracked quite well by the spiral mapping of the strike points on the divertor plate of magnetic field lines passing in front of the high harmonic fast wave (HHFW) antenna on NSTX. Indeed, both current instrumented tiles and Langmuir probes respond to the spiral when it is positioned over them. In particular, a positive increment in tile current (collection of electrons) is obtained when the spiral is over themore » tile. This current can be due to RF rectification and/or RF heating of the scrape off layer (SOL) plasma along the magnetic field lines passing in front of the the HHFW antenna. It is important to determine quantitatively the relative contributions of these processes. Here we explore the properties of the characteristics of probes on the lower divertor plate to determine the likelyhood that the primary cause of the RF heat deposition is RF rectification.« less

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

NASA Astrophysics Data System (ADS)

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.

Parameter dependences of the separatrix density in nitrogen seeded ASDEX Upgrade H-mode discharges

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Sun, H. J.; Eich, T.; Carralero, D.; Hobirk, J.; Scarabosio, A.; Siccinio, M.; ASDEX Upgrade Team; EUROfusion MST1 Team

2018-04-01

The upstream separatrix electron density is an important interface parameter for core performance and divertor power exhaust. It has been measured in ASDEX Upgrade H-mode discharges by means of Thomson scattering using a self-consistent estimate of the upstream electron temperature under the assumption of Spitzer-Härm electron conduction. Its dependence on various plasma parameters has been tested for different plasma conditions in H-mode. The leading parameter determining n e,sep was found to be the neutral divertor pressure, which can be considered as an engineering parameter since it is determined mainly by the gas puff rate and the pumping speed. The experimentally found parameter dependence of n e,sep, which is dominated by the divertor neutral pressure, could be approximately reconciled by 2-point modelling.

Thermal strain measurement of EAST tungsten divertor component with bare fiber Bragg grating sensors

NASA Astrophysics Data System (ADS)

Wang, Xingli; Wang, Wanjing; Wang, Jichao; Wei, Ran; Sun, Zhaoxuan; Li, Qiang; Xie, Chunyi; Luo, Guang-Nan

2017-12-01

Fiber Bragg Gratings (FBGs) have been widely used in the sensor field to monitor temperature and strain. However, the weak mechanical property of optical fibers and insufficient heat-resistant property of general optic-fiber sensors have prevented it from being widely used, such as in some extreme engineering situations. In this work, a bare FBG sensor system had been introduced to measure thermal strain of an Experimental Advanced Superconducting Tokamak tungsten divertor component under baking condition. This strain measurement system had withstood as high temperature as 210 °C and finished the measurement experiment successfully. Meaningful measurement results had been obtained and analyzed, which showed the applicability of such a bare fiber grating sensor system and as well contributed to studying on tungsten divertor's thermal strain conditions.

Failure study of helium-cooled tungsten divertor plasma-facing units tested at DEMO relevant steady-state heat loads

NASA Astrophysics Data System (ADS)

Ritz, G.; Hirai, T.; Norajitra, P.; Reiser, J.; Giniyatulin, R.; Makhankov, A.; Mazul, I.; Pintsuk, G.; Linke, J.

2009-12-01

Tungsten was selected as armor material for the helium-cooled divertor in future DEMO-type fusion reactors and fusion power plants. After realizing the design and testing of them under cyclic thermal loads of up to ~14 MW m-2, the tungsten divertor plasma-facing units were examined by metallography; they revealed failures such as cracks at the thermal loaded and as-machined surfaces, as well as degradation of the brazing layers. Furthermore, in order to optimize the machining processes, the quality of tungsten surfaces prepared by turning, milling and using a diamond cutting wheel were examined. This paper presents a metallographic examination of the tungsten plasma-facing units as well as technical studies and the characterization on machining of tungsten and alternative brazing joints.

Overview of ASDEX Upgrade results

NASA Astrophysics Data System (ADS)

Stroth, U.; Adamek, J.; Aho-Mantila, L.; Äkäslompolo, S.; Amdor, C.; Angioni, C.; Balden, M.; Bardin, S.; Barrera Orte, L.; Behler, K.; Belonohy, E.; Bergmann, A.; Bernert, M.; Bilato, R.; Birkenmeier, G.; Bobkov, V.; Boom, J.; Bottereau, C.; Bottino, A.; Braun, F.; Brezinsek, S.; Brochard, T.; Brüdgam, M.; Buhler, A.; Burckhart, A.; Casson, F. J.; Chankin, A.; Chapman, I.; Clairet, F.; Classen, I. G. J.; Coenen, J. W.; Conway, G. D.; Coster, D. P.; Curran, D.; da Silva, F.; de Marné, P.; D'Inca, R.; Douai, D.; Drube, R.; Dunne, M.; Dux, R.; Eich, T.; Eixenberger, H.; Endstrasser, N.; Engelhardt, K.; Esposito, B.; Fable, E.; Fischer, R.; Fünfgelder, H.; Fuchs, J. C.; Gál, K.; García Muñoz, M.; Geiger, B.; Giannone, L.; Görler, T.; da Graca, S.; Greuner, H.; Gruber, O.; Gude, A.; Guimarais, L.; Günter, S.; Haas, G.; Hakola, A. H.; Hangan, D.; Happel, T.; Härtl, T.; Hauff, T.; Heinemann, B.; Herrmann, A.; Hobirk, J.; Höhnle, H.; Hölzl, M.; Hopf, C.; Houben, A.; Igochine, V.; Ionita, C.; Janzer, A.; Jenko, F.; Kantor, M.; Käsemann, C.-P.; Kallenbach, A.; Kálvin, S.; Kantor, M.; Kappatou, A.; Kardaun, O.; Kasparek, W.; Kaufmann, M.; Kirk, A.; Klingshirn, H.-J.; Kocan, M.; Kocsis, G.; Konz, C.; Koslowski, R.; Krieger, K.; Kubic, M.; Kurki-Suonio, T.; Kurzan, B.; Lackner, K.; Lang, P. T.; Lauber, P.; Laux, M.; Lazaros, A.; Leipold, F.; Leuterer, F.; Lindig, S.; Lisgo, S.; Lohs, A.; Lunt, T.; Maier, H.; Makkonen, T.; Mank, K.; Manso, M.-E.; Maraschek, M.; Mayer, M.; McCarthy, P. J.; McDermott, R.; Mehlmann, F.; Meister, H.; Menchero, L.; Meo, F.; Merkel, P.; Merkel, R.; Mertens, V.; Merz, F.; Mlynek, A.; Monaco, F.; Müller, S.; Müller, H. W.; Münich, M.; Neu, G.; Neu, R.; Neuwirth, D.; Nocente, M.; Nold, B.; Noterdaeme, J.-M.; Pautasso, G.; Pereverzev, G.; Plöckl, B.; Podoba, Y.; Pompon, F.; Poli, E.; Polozhiy, K.; Potzel, S.; Püschel, M. J.; Pütterich, T.; Rathgeber, S. K.; Raupp, G.; Reich, M.; Reimold, F.; Ribeiro, T.; Riedl, R.; Rohde, V.; Rooij, G. v.; Roth, J.; Rott, M.; Ryter, F.; Salewski, M.; Santos, J.; Sauter, P.; Scarabosio, A.; Schall, G.; Schmid, K.; Schneider, P. A.; Schneider, W.; Schrittwieser, R.; Schubert, M.; Schweinzer, J.; Scott, B.; Sempf, M.; Sertoli, M.; Siccinio, M.; Sieglin, B.; Sigalov, A.; Silva, A.; Sommer, F.; Stäbler, A.; Stober, J.; Streibl, B.; Strumberger, E.; Sugiyama, K.; Suttrop, W.; Tala, T.; Tardini, G.; Teschke, M.; Tichmann, C.; Told, D.; Treutterer, W.; Tsalas, M.; Van Zeeland, M. A.; Varela, P.; Veres, G.; Vicente, J.; Vianello, N.; Vierle, T.; Viezzer, E.; Viola, B.; Vorpahl, C.; Wachowski, M.; Wagner, D.; Wauters, T.; Weller, A.; Wenninger, R.; Wieland, B.; Willensdorfer, M.; Wischmeier, M.; Wolfrum, E.; Würsching, E.; Yu, Q.; Zammuto, I.; Zasche, D.; Zehetbauer, T.; Zhang, Y.; Zilker, M.; Zohm, H.

2013-10-01

The medium size divertor tokamak ASDEX Upgrade (major and minor radii 1.65 m and 0.5 m, respectively, magnetic-field strength 2.5 T) possesses flexible shaping and versatile heating and current drive systems. Recently the technical capabilities were extended by increasing the electron cyclotron resonance heating (ECRH) power, by installing 2 × 8 internal magnetic perturbation coils, and by improving the ion cyclotron range of frequency compatibility with the tungsten wall. With the perturbation coils, reliable suppression of large type-I edge localized modes (ELMs) could be demonstrated in a wide operational window, which opens up above a critical plasma pedestal density. The pellet fuelling efficiency was observed to increase which gives access to H-mode discharges with peaked density profiles at line densities clearly exceeding the empirical Greenwald limit. Owing to the increased ECRH power of 4 MW, H-mode discharges could be studied in regimes with dominant electron heating and low plasma rotation velocities, i.e. under conditions particularly relevant for ITER. The ion-pressure gradient and the neoclassical radial electric field emerge as key parameters for the transition. Using the total simultaneously available heating power of 23 MW, high performance discharges have been carried out where feed-back controlled radiative cooling in the core and the divertor allowed the divertor peak power loads to be maintained below 5 MW m-2. Under attached divertor conditions, a multi-device scaling expression for the power-decay length was obtained which is independent of major radius and decreases with magnetic field resulting in a decay length of 1 mm for ITER. At higher densities and under partially detached conditions, however, a broadening of the decay length is observed. In discharges with density ramps up to the density limit, the divertor plasma shows a complex behaviour with a localized high-density region in the inner divertor before the outer divertor detaches. Turbulent transport is studied in the core and the scrape-off layer (SOL). Discharges over a wide parameter range exhibit a close link between core momentum and density transport. Consistent with gyro-kinetic calculations, the density gradient at half plasma radius determines the momentum transport through residual stress and thus the central toroidal rotation. In the SOL a close comparison of probe data with a gyro-fluid code showed excellent agreement and points to the dominance of drift waves. Intermittent structures from ELMs and from turbulence are shown to have high ion temperatures even at large distances outside the separatrix.

Particle simulations on transport control in divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kashiwagi, Mieko; Ido, Shunji

1995-04-01

Particle orbit simulations are carried out to study the reflection of He ions recycled from a tokamak divertor by RF electric fields, which have the frequency close to ion cyclotron resonance frequency (ICRF). The performance of particle reflection and the requirement to the intensity of RF fields are studied. The control of He recycling by ICRF fields is found to be available. 4 refs., 4 figs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

M. Ono; Jaworski, M.; Kaita, R.

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

ADX: a high field, high power density, advanced divertor and RF tokamak

NASA Astrophysics Data System (ADS)

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 (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Experiments and numerical modeling of fast flowing liquid metal thin films under spatially varying magnetic field conditions

NASA Astrophysics Data System (ADS)

Narula, Manmeet Singh

Innovative concepts using fast flowing thin films of liquid metals (like lithium) have been proposed for the protection of the divertor surface in magnetic fusion devices. However, concerns exist about the possibility of establishing the required flow of liquid metal thin films because of the presence of strong magnetic fields which can cause flow disrupting MHD effects. A plan is underway to design liquid lithium based divertor protection concepts for NSTX, a small spherical torus experiment at Princeton. Of these, a promising concept is the use of modularized fast flowing liquid lithium film zones, as the divertor (called the NSTX liquid surface module concept or NSTX LSM). The dynamic response of the liquid metal film flow in a spatially varying magnetic field configuration is still unknown and it is suspected that some unpredicted effects might be lurking. The primary goal of the research work being reported in this dissertation is to provide qualitative and quantitative information on the liquid metal film flow dynamics under spatially varying magnetic field conditions, typical of the divertor region of a magnetic fusion device. The liquid metal film flow dynamics have been studied through a synergic experimental and numerical modeling effort. The Magneto Thermofluid Omnibus Research (MTOR) facility at UCLA has been used to design several experiments to study the MHD interaction of liquid gallium films under a scaled NSTX outboard divertor magnetic field environment. A 3D multi-material, free surface MHD modeling capability is under development in collaboration with HyPerComp Inc., an SBIR vendor. This numerical code called HIMAG provides a unique capability to model the equations of incompressible MHD with a free surface. Some parts of this modeling capability have been developed in this research work, in the form of subroutines for HIMAG. Extensive code debugging and benchmarking exercise has also been carried out. Finally, HIMAG has been used to study the MHD interaction of fast flowing liquid metal films under various divertor relevant magnetic field configurations through numerical modeling exercises.

Experimental and analytical studies of high heat flux components for fusion experimental reactor

NASA Astrophysics Data System (ADS)

Araki, Masanori

1993-03-01

In this report, the experimental and analytical results concerning the development of plasma facing components of ITER are described. With respect to developing high heat removal structures for the divertor plates, an externally-finned swirl tube was developed based on the results of critical heat flux (CHF) experiments on various tube structures. As the result, the burnout heat flux, which also indicates incident CHF, of 41 (+/-) 1 MW/sq m was achieved in the externally-finned swirl tube. The applicability of existing CHF correlations based on uniform heating conditions was evaluated by comparing the CHF experimental data with the smooth and the externally-finned tubes under one-sided heating condition. As the results, experimentally determined CHF data for straight tube show good agreement, for the externally-finned tube, no existing correlations are available for prediction of the CHF. With respect to the evaluation of the bonds between carbon-based material and heat sink metal, results of brazing tests were compared with the analytical results by three dimensional model with temperature-dependent thermal and mechanical properties. Analytical results showed that residual stresses from brazing can be estimated by the analytical three directional stress values instead of the equivalent stress value applied. In the analytical study on the separatrix sweeping for effectively reducing surface heat fluxes on the divertor plate, thermal response of the divertor plate was analyzed under ITER relevant heat flux conditions and has been tested. As the result, it has been demonstrated that application of the sweeping technique is very effective for improvement in the power handling capability of the divertor plate and that the divertor mock-up has withstood a large number of additional cyclic heat loads.

Development and application of W/Cu flat-type plasma facing components at ASIPP

NASA Astrophysics Data System (ADS)

Li, Q.; Zhao, S. X.; Sun, Z. X.; Xu, Y.; Li, B.; Wei, R.; Wang, W. J.; Qin, S. G.; Shi, Y. L.; Xie, C. Y.; Wang, J. C.; Wang, X. L.; Missirlian, M.; Guilhem, D.; Liu, G. H.; Yang, Z. S.; Luo, G.-N.

2017-12-01

W/Cu flat-type plasma facing components (PFCs) were widely used in divertor of fusion device because of its advantages, such as low cost, light in weight and good machinability. However, it is very difficult to manufacture them due to the large mismatch between the thermo-mechanical properties of W and Cu. Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) has successfully developed W/Cu flat-type PFCs for EAST W/Cu divertor project by hot isostatic pressing (HIP) technology. This paper presents the development and application of W/Cu flat-type PFCs at ASIPP. The optimized manufacturing process is to cast pure copper onto the rear side of W tiles at temperature of 1200 °C firstly, and then to HIP the W/Cu tiles onto CuCrZr heat sink at temperature of 600 °C, pressure of 150 MPa and duration of 3 h. W/Cu flat-type testing mock-up for EAST survived 1000 cycles at heat load of 5 MW m-2 in high heat flux tests. And then ASIPP prepared two mock-ups for CEA’s tungsten environment in steady-state tokamak (WEST) project. One mock-up withstood successfully 302 cycles of 20 MW m-2, which are far beyond the design requirement. Since 2014, W/Cu flat-type PFCs were wildly used in EAST upper divertor as baffle and dome components which showed excellent performance in 2015 and 2016 campaigns. Given the success in EAST upper divertor, W/Cu flat-type concept is as well applied in the design of actively cooled Langmuir probes which will be mounted onto EAST divertor targets soon.

Integration of uncooled scraper elements and its diagnostics into Wendelstein 7-X

DOE PAGES

Fellinger, Joris; Loesser, Doug; Neilson, Hutch; ...

2017-08-08

The modular stellarator Wendelstein 7-X in Greifswald (Germany) successfully started operation in 2015 with short pulse limiter plasmas. In 2017, the next operation phase (OP) OP1.2 will start once 10 uncooled test divertor units (TDU) with graphite armor will be installed. The TDUs allow for plasma pulses of 10 s with 8 MW heating. OP2, allowing for steady state operation, is planned for 2020 after the TDUs will be replaced by 10 water cooled CFC armored divertors. Due to the development of plasma currents like bootstrap currents in long pulse plasmas in OP2, the plasma could hit the edge ofmore » the divertor targets which has a reduced cooling capacity compared to the central part of the target tiles. To prevent overloading of these edges, a so-called scraper element can be positioned in front of the divertor, intersecting those strike lines that would otherwise hit the divertor edges. As a result, these edges are protected but as a drawback the pumping efficiency of neutrals is also reduced. As a test an uncooled scraper element with graphite tiles will be placed in two out of ten half modules in OP1.2. A decision to install ten water cooled scraper elements for OP2 is pending on the results of this test in OP1.2. To monitor the impact of the scraper element on the plasma, Langmuir probes are integrated in the plasma facing surface, and a neutral gas manometer measures the neutral density directly behind the plasma facing surface. Moreover, IR and VIS cameras observe the plasma facing surface and thermocouples monitor the temperatures of the graphite tiles and underlying support structure. This paper describes the integration of the scraper element and its diagnostics in Wendelstein 7-X.« less

Quantification of Chemical Erosion in the DIII-D Divertor

NASA Astrophysics Data System (ADS)

McLean, Adam

2009-11-01

Chemical erosion (CE) yield at the graphite divertor target in DIII-D was measured to be substantially lower in cold near-detached plasma conditions compared to well-attached ones, with major implications for ITER. Current estimates of tritium retention by co-deposition with hydrocarbons (HCs) in ITER place potentially severe restrictions on operation. However, calculations done to date have been based on excessively conservative assumptions, due to limited understanding of cold divertor plasmas (1-5eV) which bridge energy thresholds for complex atomic and molecular processes not present in attached conditions. Hydrocarbon injection through a unique porous graphite plate which realistically simulates secondary reactions of HCs with a graphite surface has been used to measure CE in-situ. For the first time in a divertor, measurements were made at extrinsic CH4 injection rates comparable to the expected intrinsic CE rate of C, with the resulting spectroscopic emissions separated from those of the intrinsic sources. Under cold plasma conditions the contribution of CE-produced C relative to total C sources in the divertor declined dramatically from ˜50% to <15%. Photon efficiencies for products from the breakup of injected CH4 were greater than previous measurements at higher puff rates, indicating the importance of minimizing perturbation to the local plasma. At 350K, the measured CE yield near the outer strike point was ˜2.6% in attachment dropping to only ˜0.5% in cold plasma; results are consistent with some theoretical predications and lab studies. Under full detachment, near total extinction of the CD band occurred, consistent with suppression of net C erosion. These findings have potentially major impact on projected target lifetime and tritium retention in future reactors, and for the PFC choice in ITER.

The Simple Map for a Single-null Divertor Tokamak: How to Find the Footprint of Field lines

NASA Astrophysics Data System (ADS)

Figgins, Montoya; Ali, Halima; Punjabi, Alkesh

2000-10-01

We are working with the Simple Map^1 to find the footprint of field lines on the diverter plate in a single-null tokamak. Footprint of a field line is the position of the line when it escapes across the divertor plate. The Simple Map represents the magnetic field in a single-null divertor tokamak. The path of a field line is given by the equations: X_n+1=X_n-kY_n(1-Y_n) and Y_n+1=Y_n+kX_n+1. In order to find the footprint, we must first find the last good surface which is Y=0.997135768 and X=0. The value of k is fixed at 0.6. The starting values X0 are fixed at X_0=0. We use 10,000 points between the last good surface and the X-point. The X-point is located at (0,1). We also use the Continuous Analog of the Simple Map given by the equations: X(φ)=X_0-kY0 (1-Y_0)φ and Y(φ)=Y_0+kX(φ)φ. This will tell us what the (φ,X) is which represents the field lines crossing the divertor plate. The divertor plate is located at Y=1. When graphed, the footprint of field lines looks like the rings of Saturn. This work is supported by US DOES OFES. Ms. Montoya Figgins is HU CFRT Summer Fusion High School Scholar from E. E. Smith High School in North Carolina. She is supported by NASA under its NASA SHARP Plus Program. 1. Punjabi A, Verma A, and Boozer A, Phys Rev Lett, 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994)

SOLPS simulations of X-divertor in NSTX-U

NASA Astrophysics Data System (ADS)

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.

Recent sheath physics studies on DIII-D

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Labombard, B.; Stangeby, P. C.; Lasnier, C. J.; McLean, A. G.; Nygren, R. E.; Boedo, J. A.; Leonard, A. W.; Rudakov, D. L.

2015-08-01

A study to examine some current issues in the physics of the plasma sheath has been recently carried out in DIII-D low power Ohmic plasmas using both flush and domed Langmuir probes, divertor Thomson scattering (DTS), an infrared camera (IRTV), and a new calorimeter triple probe assembly mounted on the Divertor Materials Evaluation System (DIMES). The sheath power transmission factor was found to be consistent with the theoretically predicted value of 7 (±2) for low power plasmas. Using this factor, the three heat flux profiles derived from the LP, DTS, and calorimeter diagnostic measurements agree. Comparison of flush and domed Langmuir probes and divertor Thomson scattering indicates that proper interpretation of flush probe data to get target plate density and temperature is feasible and could potentially yield accurate measurements of target plate conditions where the probes are located.

Upgrade of the infrared camera diagnostics for the JET ITER-like wall divertor.

PubMed

Balboa, I; Arnoux, G; Eich, T; Sieglin, B; Devaux, S; Zeidner, W; Morlock, C; Kruezi, U; Sergienko, G; Kinna, D; Thomas, P D; Rack, M

2012-10-01

For the new ITER-like wall at JET, two new infrared diagnostics (KL9B, KL3B) have been installed. These diagnostics can operate between 3.5 and 5 μm and up to sampling frequencies of ∼20 kHz. KL9B and KL3B image the horizontal and vertical tiles of the divertor. The divertor tiles are tungsten coated carbon fiber composite except the central tile which is bulk tungsten and consists of lamella segments. The thermal emission between lamellae affects the surface temperature measurement and therefore KL9A has been upgraded to achieve a higher spatial resolution (by a factor of 2). A technical description of KL9A, KL9B, and KL3B and cross correlation with a near infrared camera and a two-color pyrometer is presented.

Analysis of the plasma-wall interaction in the Heliotron E device

NASA Astrophysics Data System (ADS)

Motojima, O.; Mizuuchi, T.; Besshou, S.; Iiyoshi, A.; Uo, K.; Yamashina, T.; Mohri, M.; Satake, T.; Hashiba, M.; Amemiya, S.; Miwa, H.

1984-12-01

The plasma-wall interaction (PWI) of the currentless plasmas with temperature To, Tio ≤ 1.1 keV, density N¯e = (2-10)× 1013/cm3, and volume-averaged beta value of β$¯≤ 2% was investigated. We have observed that PWI took place mainly where the divertor field line intersected the chamber wall (called divertor traces). Boundary plasmas were measured with electrostatic probes, which showed the presence of the divertor region with the parameters in the range of Ned = 1010-1011/cm3 and Ted = 10-50 eV. Surface analysis techniques (ESCA, AES, and RBS) were applied to analyze the surface probes (Si, graphite and stainless steel) and the test pieces (SiC, TiC, and stainless steel), which were irradiated by plasmas for short and long times respectively.

Studies of Be migration in the JET tokamak using AMS with 10Be marker

NASA Astrophysics Data System (ADS)

Bykov, I.; Bergsåker, H.; Possnert, G.; Zhou, Y.; Heinola, K.; Pettersson, J.; Conroy, S.; Likonen, J.; Petersson, P.; Widdowson, A.

2016-03-01

The JET tokamak is operated with beryllium limiter tiles in the main chamber and tungsten coated carbon fiber composite tiles and solid W tiles in the divertor. One important issue is how wall materials are migrating during plasma operation. To study beryllium redistribution in the main chamber and in the divertor, a 10Be enriched limiter tile was installed prior to plasma operations in 2011-2012. Methods to take surface samples have been developed, an abrasive method for bulk Be tiles in the main chamber, which permits reuse of the tiles, and leaching with hot HCl to remove all Be deposited at W coated surfaces in the divertor. Quantitative analysis of the total amount of Be in cm2 sized samples was made with inductively coupled plasma atomic emission spectroscopy (ICP-AES). The 10Be/9Be ratio in the samples was measured with accelerator mass spectrometry (AMS). The experimental setup and methods are described in detail, including sample preparation, measures to eliminate contributions in AMS from the 10B isobar, possible activation due to plasma generated neutrons and effects of diffusive isotope mixing. For the first time marker concentrations are measured in the divertor deposits. They are in the range 0.4-1.2% of the source concentration, with moderate poloidal variation.

In-Vessel Tritium Retention and Removal in ITER-FEAT

NASA Astrophysics Data System (ADS)

Federici, G.; Brooks, J. N.; Iseli, M.; Wu, C. H.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ˜350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.

Particle and Power Exhaust in EAST

NASA Astrophysics Data System (ADS)

Wang, Liang; Ding, Fang; Yu, Yaowei; Gan, Kaifu; Liang, Yunfeng; Xu, Guosheng; Xiao, Bingjia; Sun, Youwen; Luo, Guangnan; Gong, Xianzu; Hu, Jiansheng; Li, Jiangang; Wan, Baonian; Maingi, Rajesh; Guo, Houyang; Garofalo, Andrea; EAST Team

2017-10-01

A total power injection up to 0.3GJ has been achieved in EAST long pulse USN operation with ITER-like water-cooling W-monoblock divertor, which has steady-state power exhaust capability of 10 MWm-2. The peak temperature of W target saturated at t = 12 s to the value T 500oC and a heat flux 3MWm-2was maintained. Great efforts to reduce heat flux and accommodate particle exhaust simultaneously have been made towards long pulse of 102s time scale. By exploiting the observation of Pfirsch-Schlüter flow direction in the SOL, the Bt direction with Bx ∇B away from the W divertor (more particles favor outer target in USN) was adopted along with optimizing the strike point location near the pumping slot, to facilitate particle and impurity exhaust with the top cryo-pump. By tailoring the 3D divertor footprint through edge magnetic topology change, the heat load was dispersed widely and thus peak heat flux and W sputtering was well controlled. Active feedback control of total radiative power with neon seeding was achieved within frad = 17-35%, exhibiting further potential for heat flux reduction with divertor and edge radiation. Other heat flux handling techniques, including quasi snowflake configuration, will also be presented.

High heat flux Langmuir probe array for the DIII-D divertor platesa)

NASA Astrophysics Data System (ADS)

Watkins, J. G.; Taussig, D.; Boivin, R. L.; Mahdavi, M. A.; Nygren, R. E.

2008-10-01

Two modular arrays of Langmuir probes designed to handle a heat flux of up to 25 MW/m2 for 10 s exposures have been installed in the lower divertor target plates of the DIII-D tokamak. The 20 pyrolytic graphite probe tips have more than three times higher thermal conductivity and 16 times larger mass than the original DIII-D isotropic graphite probes. The probe tips have a fixed 12.5° surface angle to distribute the heat flux more uniformly than the previous 6 mm diameter domed collectors and a symmetric "rooftop" design to allow operation with reversed toroidal magnetic field. A large spring-loaded contact area improves heat conduction from each probe tip through a ceramic insulator into a cooled graphite divertor floor tile. The probe tips, brazed to molybdenum foil to ensure good electrical contact, are mounted in a ceramic tray for electrical isolation and reliable cable connections. The new probes are located 1.5 cm radially apart in a staggered arrangement near the entrance to the lower divertor pumping baffle and are linearly spaced 3 cm apart on the shelf above the in-vessel cryopump. Typical target plate profiles of Jsat, Te, and Vf with 4 mm spatial resolution are shown.

Overview of Compact Toroidal Hybrid research program progress and plans

NASA Astrophysics Data System (ADS)

Maurer, David; Ennis, David; Hanson, James; Hartwell, Gregory; Herfindal, Jeffrey; Knowlton, Stephen; Ma, Xingxing; Pandya, Mihir; Roberds, Nicholas; Ross, Kevin; Traverso, Peter

2016-10-01

disruptive behavior on the level of applied 3D magnetic shaping; (2) test and advance the V3FIT reconstruction code and NIMROD modeling of CTH; and (3) study the implementation of an island divertor. Progress towards these goals and other developments are summarized. The disruptive density limit exceeds the Greenwald limit as the vacuum transform is increased, but a threshold for avoidance is not observed. Low- q disruptions, with 1.1 < q (a) <2.0, cease to occur if the vacuum transform is raised above 0.07. Application of vacuum transform can reduce and eliminate the vertical drift of elongated discharges that would otherwise be vertically unstable. Reconstructions using external magnetics give accurate estimates for quantities near the plasma boundary, and internal diagnostics have been implemented to extend the range of accuracy into the plasma core. Sawtooth behavior has been reproducibly modified with external transform and NIMROD is used to model these observations and reproduces experimental trends. An island divertor design has begun with connection length studies to model energy deposition on divertor plates located in an edge 1/3 island as well as the study of a non-resonant divertor configuration. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

2D imaging of helium ion velocity in the DIII-D divertor

NASA Astrophysics Data System (ADS)

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.

Symmetric Simple Map with Dipole Map for a Single-Null Divertor Tokamak

NASA Astrophysics Data System (ADS)

Ali, Halima; Watson, Michael; Punjabi, Alkesh; Boozer, Allen

1996-11-01

This investigation focuses on the effects of an externally placed dipole coil on the magnetic topology of a single-null divertor tokamak with a stochastic scrape-off layer using the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994). The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effect of dipole perturbation is repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The strength of the dipole perturbation and the distance of the coil from the last good surface are varied. We observe that the dipole perturbation causes spatially intermittent chaos. This has significant implications for radiative divertor concepts as well for impurity control. We also present the detailed results on the effects of the dipole coil on the properties of the stochastic layer and the footprint of the field lines on the divertor plate. This work is supported by the US DOE OFES.

Upgrades toward high-heat flux, liquid lithium plasma-facing components in the NSTX-U

DOE PAGES

Jaworski, M. A.; Brooks, A.; Kaita, R.; ...

2016-08-08

Liquid metal plasma-facing components (PFCs) provide numerous potential advantages over solid-material components. One critique of the approach is the relatively less developed technologies associated with deploying these components in a fusion plasma-experiment. Exploration of the temperature limits of liquid lithium PFCs in a tokamak divertor and the corresponding consequences on core operation are a high priority informing the possibilities for future liquid lithium PFCs. An all-metal NSTX-U is envisioned to make direct comparison between all high-Z wall operation and liquid lithium PFCs in a single device. By executing the all-metal upgrades incrementally, scientific productivity will be maintained while enabling physicsmore » and engineering-science studies to further develop the solid- and liquid-metal components. Six major elements of a flowing liquid-metal divertor system are described and a three-step program for implementing this system is laid out. The upgrade steps involve the first high-Z divertor target upgrade in NSTX-U, pre-filled liquid metal targets and finally, an integrated, flowing liquid metal divertor target. As a result, two example issues are described where the engineering and physics experiments are shown to be closely related in examining the prospects for future liquid metal PFCs.« less

Evaluation of cooling concepts and specimen geometries for high heat flux tests on neutron irradiated divertor elements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Bolt. H.; Breitbach, G.

1994-12-31

To assess the lifetime and the long term heat removal capabilities of plasma facing components in future thermonuclear fusion reactors such as ITER, neutron irradiation and subsequent high heat flux tests will be most essential. The effect of neutron damage will be simulated in material test reactors (such as the HFR-Petten) in a fission neutron environment. To investigate the heat loads during normal and off-normal operation scenarios a 60 kW electron beam test stand (Juelich Divertor Test Facility in Hot Cells, JUDITH) has been installed in a hot cell which can be operated by remote handling techniques. In this facilitymore » inertially cooled test coupons can be handled as well as small actively cooled divertor mock-ups. A special clamping mechanism for small test coupons (25 mm x 25 mm x 35 mm) with an integrated coolant channel within a copper or TZM heat sink has been developed and tested in an electron beam test bed. This method is an attractive alternative to costly large scale tests on complete divertor modules. The temperature and stress fields in individual CFC or beryllium tiles brazed to metallic heat sink (e.g. copper or TZM) can be investigated before and after neutron irradiation with moderate efforts.« less

Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

NASA Astrophysics Data System (ADS)

Silburn, S. A.; Matthews, G. F.; Challis, C. D.; Frigione, D.; Graves, J. P.; Mantsinen, M. J.; Belonohy, E.; Hobirk, J.; Iglesias, D.; Keeling, D. L.; King, D.; Kirov, K.; Lennholm, M.; Lomas, P. J.; Moradi, S.; Sips, A. C. C.; Tsalas, M.; Contributors, JET

2017-12-01

Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data.

Progress on the application of ELM control schemes to ITER scenarios from the non-active phase to DT operation

NASA Astrophysics Data System (ADS)

Loarte, A.; Huijsmans, G.; Futatani, S.; Baylor, L. R.; Evans, T. E.; Orlov, D. M.; Schmitz, O.; Becoulet, M.; Cahyna, P.; Gribov, Y.; Kavin, A.; Sashala Naik, A.; Campbell, D. J.; Casper, T.; Daly, E.; Frerichs, H.; Kischner, A.; Laengner, R.; Lisgo, S.; Pitts, R. A.; Saibene, G.; Wingen, A.

2014-03-01

Progress in the definition of the requirements for edge localized mode (ELM) control and the application of ELM control methods both for high fusion performance DT operation and non-active low-current operation in ITER is described. Evaluation of the power fluxes for low plasma current H-modes in ITER shows that uncontrolled ELMs will not lead to damage to the tungsten (W) divertor target, unlike for high-current H-modes in which divertor damage by uncontrolled ELMs is expected. Despite the lack of divertor damage at lower currents, ELM control is found to be required in ITER under these conditions to prevent an excessive contamination of the plasma by W, which could eventually lead to an increased disruptivity. Modelling with the non-linear MHD code JOREK of the physics processes determining the flow of energy from the confined plasma onto the plasma-facing components during ELMs at the ITER scale shows that the relative contribution of conductive and convective losses is intrinsically linked to the magnitude of the ELM energy loss. Modelling of the triggering of ELMs by pellet injection for DIII-D and ITER has identified the minimum pellet size required to trigger ELMs and, from this, the required fuel throughput for the application of this technique to ITER is evaluated and shown to be compatible with the installed fuelling and tritium re-processing capabilities in ITER. The evaluation of the capabilities of the ELM control coil system in ITER for ELM suppression is carried out (in the vacuum approximation) and found to have a factor of ˜2 margin in terms of coil current to achieve its design criterion, although such a margin could be substantially reduced when plasma shielding effects are taken into account. The consequences for the spatial distribution of the power fluxes at the divertor of ELM control by three-dimensional (3D) fields are evaluated and found to lead to substantial toroidal asymmetries in zones of the divertor target away from the separatrix. Therefore, specifications for the rotation of the 3D perturbation applied for ELM control in order to avoid excessive localized erosion of the ITER divertor target are derived. It is shown that a rotation frequency in excess of 1 Hz for the whole toroidally asymmetric divertor power flux pattern is required (corresponding to n Hz frequency in the variation of currents in the coils, where n is the toroidal symmetry of the perturbation applied) in order to avoid unacceptable thermal cycling of the divertor target for the highest power fluxes and worst toroidal power flux asymmetries expected. The possible use of the in-vessel vertical stability coils for ELM control as a back-up to the main ELM control systems in ITER is described and the feasibility of its application to control ELMs in low plasma current H-modes, foreseen for initial ITER operation, is evaluated and found to be viable for plasma currents up to 5-10 MA depending on modelling assumptions.

Effect of ELMs on deuterium-loaded-tungsten plasma facing components

NASA Astrophysics Data System (ADS)

Umstadter, K. R.; Rudakov, D. L.; Wampler, W.; Watkins, J. G.; Wong, C. P. C.

2011-08-01

Prior heat pulse testing of plasma facing components (PFCs) has been completed in vacuum environments without the presence of background plasma. Edge localized modes (ELMs) will not be this kind of isolated event and one should know the effect of a plasma background during these transients. Heat-pulse experiments have been conducted in the PISCES-A device utilizing laser heating in a divertor-like plasma background. Initial results indicate that the erosion of PFCs is enhanced as compared to heat pulse or plasma only tests. To determine if the enhanced erosion effect is a phenomena only witnessed in the laboratory PISCES device, tungsten and graphite samples were exposed to plasmas in the lower divertor of the DIII-D tokamak using the Divertor Material Evaluation System (DiMES). Mass loss analysis indicates that materials that contain significant deuterium prior to experiencing a transient heating event will erode faster than those that have no or little retained deuterium.

Neutral pressure behavior for diverted discharges in the Wendelstein 7-AS Stellarator

NASA Astrophysics Data System (ADS)

McCormick, K.; Grigull, P.; Burhenn, R.; Ehmler, H.; Feng, Y.; Giannone, L.; Haas, G.; Sardei, F.; NBI-, ECRH-; W7-AS Teams

2005-03-01

On the W7-AS stellarator, the subdivertor neutral pressure in an up-down divertor pair as well as at two points in the vicinity of a lower divertor module in the main chamber are measured. Results are presented for ι=5/9 island divertor discharges under conditions of normal confinement (NC) and the HDH-mode for: n˜0.1-4×1020 m-3, Pecrh = 0.5-1.5 MW, Pnbi = 2 MW, and H + and D + plasmas, with both normal- and reversed- Bt for H +. Subdivertor pressures are in the range 1-2 × 10 -3 mbar for HDH conditions. For plasma detachment at the target plates a strong up-down pressure asymmetry arises, with pup/ pdown ⩽ 5. The asymmetry reverses with reversed Bt. Main vessel pressures are a factor of 5-10 lower than the average subdivertor pressure for H +, with D + plasmas exhibiting still lower values.

Performance of V-4Cr-4Ti material exposed to DIII-D tokamak environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tsai, H.; Chung, H.M.; Smith, D.L.

1997-04-01

Test specimens made with the 832665 heat of V-4Cr-4Ti alloy were exposed in the DIII-D tokamak environment to support the installation of components made of a V-4Cr-4Ti alloy in the radiative divertor of the DIII-D. Some of the tests were conducted with the Divertor Materials Evaluation System (DiMES) to study the short-term effects of postvent bakeout, when concentrations of gaseous impurities in the DIII-D chamber are the highest. Other specimens were mounted next to the chamber wall behind the divertor baffle plate, to study the effects of longer-term exposures. By design, none of the specimens directly interacted with the plasma.more » Preliminary results from testing the exposed specimens indicate only minor degradation of mechanical properties. Additional testing and microstructural characterization are in progress.« less

Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

NASA Astrophysics Data System (ADS)

You, Jeong-Ha

2005-01-01

Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.

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

NASA Astrophysics Data System (ADS)

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.

An automated approach to magnetic divertor configuration design

NASA Astrophysics Data System (ADS)

Blommaert, M.; Dekeyser, W.; Baelmans, M.; Gauger, N. R.; Reiter, D.

2015-01-01

Automated methods based on optimization can greatly assist computational engineering design in many areas. In this paper an optimization approach to the magnetic design of a nuclear fusion reactor divertor is proposed and applied to a tokamak edge magnetic configuration in a first feasibility study. The approach is based on reduced models for magnetic field and plasma edge, which are integrated with a grid generator into one sensitivity code. The design objective chosen here for demonstrative purposes is to spread the divertor target heat load as much as possible over the entire target area. Constraints on the separatrix position are introduced to eliminate physically irrelevant magnetic field configurations during the optimization cycle. A gradient projection method is used to ensure stable cost function evaluations during optimization. The concept is applied to a configuration with typical Joint European Torus (JET) parameters and it automatically provides plausible configurations with reduced heat load.

Long-term fuel retention in JET ITER-like wall

NASA Astrophysics Data System (ADS)

Heinola, K.; Widdowson, A.; Likonen, J.; Alves, E.; Baron-Wiechec, A.; Barradas, N.; Brezinsek, S.; Catarino, N.; Coad, P.; Koivuranta, S.; Krat, S.; Matthews, G. F.; Mayer, M.; Petersson, P.; Contributors, JET

2016-02-01

Post-mortem studies with ion beam analysis, thermal desorption, and secondary ion mass spectrometry have been applied for investigating the long-term fuel retention in the JET ITER-like wall components. The retention takes place via implantation and co-deposition, and the highest retention values were found to correlate with the thickness of the deposited impurity layers. From the total amount of retained D fuel over half was detected in the divertor region. The majority of the retained D is on the top surface of the inner divertor, whereas the least retention was measured in the main chamber on the mid-plane of the inner wall limiter. The recessed areas of the inner wall showed significant contribution to the main chamber total retention. Thermal desorption spectroscopy analysis revealed the energetic T from DD reactions being implanted in the divertor. The total T inventory was assessed to be \\gt 0.3 {{mg}}.

High density operation for reactor-relevant power exhaust

NASA Astrophysics Data System (ADS)

Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

2015-08-01

With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

We conducted physics experiments: record normalized {Beta} = 4.9 achieved in VH-mode, {Beta} limits of ITER-like configurations evaluated, FWCD commissioning. The tokamak vessel was opened to atmosphere for six weeks and a number of key diagnostics for understanding the divertor were installed. The DIII-D Advisory Committee met in January to review the DIII-D program and plan. They commended us for recent progress and supported the vanadium divertor design. The U.S./Japan DIII-D steering committee met and recommended extending the agreement to the year 2000. The field work proposal for FY 96/97 was presented in Washington on March 29, 1995. A reviewmore » of the DIII-D plan to install vanadium structural components as part of the new radiative divertor modification was held in Washington 31, 1995 and the panel endorsed the plans. Preliminary plans were developed with PPPL for collaborations in FY96,« less

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Carbon Radiation Studies in the DIII-D Divertor with the Monte Carlo Impurity (MCI) Code

NASA Astrophysics Data System (ADS)

Evans, T. E.; Leonard, A. W.; West, W. P.; Finkenthal, D. F.; Fenstermacher, M. E.; Porter, G. D.; Chu, Y.

1998-11-01

Carbon sputtering and transport are modeled in the DIII--D divertor with the MCI code. Calculated 2-D radiation patterns are compared with measured radiation distributions. The results are particularly sensitive to Ti near the divertor target plates. For example, increasing the ion temperature from 8 eV to 20 eV in MCI raises P_rad^div from 1626 to 2862 kW. Although this presents difficulties in assessing which sputtering model best describes the plasma-surface interaction physics (because of experimental uncertainties in T_i), processes which either produce too much or too little radiated power compared to the measured value of 1718 kW can be eliminated. Based on this, the number of viable sputtering options has been reduced from 12 to 4. For the conditions studied, three of these options involve both physical and chemical sputtering, and one requires only physical sputtering.

Applications of Collisional Radiative Modeling of Helium and Deuterium for Image Tomography Diagnostic of Te, Ne, and ND in the DIII-D Tokamak

NASA Astrophysics Data System (ADS)

Munoz Burgos, J. M.; Brooks, N. H.; Fenstermacher, M. E.; Meyer, W. H.; Unterberg, E. A.; Schmitz, O.; Loch, S. D.; Balance, C. P.

2011-10-01

We apply new atomic modeling techniques to helium and deuterium for diagnostics in the divertor and scrape-off layer regions. Analysis of tomographically inverted images is useful for validating detachment prediction models and power balances in the divertor. We apply tomographic image inversion from fast tangential cameras of helium and Dα emission at the divertor in order to obtain 2D profiles of Te, Ne, and ND (neutral ion density profiles). The accuracy of the atomic models for He I will be cross-checked against Thomson scattering measurements of Te and Ne. This work summarizes several current developments and applications of atomic modeling into diagnostic at the DIII-D tokamak. Supported in part by the US DOE under DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, and DE-AC05-00OR22725.

Design and Test of Wendelstein 7-X Water-Cooled Divertor Scraper

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boscary, J.; Greuner, Henri; Ehrke, Gunnar

Heat load calculations have indicated the possible overloading of the ends of the water-cooled divertor facing the pumping gap beyond their technological limit. The intention of the scraper is the interception of some of the plasma fluxes both upstream and downstream before they reach the divertor surface. The scraper is divided into six modules of four plasma facing components (PFCs); each module has four PFCs hydraulically connected in series by two water boxes (inlet and outlet). A full-scale prototype of one module has been manufactured. Development activities have been carried out to connect the water boxes to the cooling pipesmore » of the PFCs by tungsten inert gas internal orbital welding. This prototype was successfully tested in the GLADIS facility with 17 MW/m2 for 500 cycles. The results of these activities have confirmed the possible technological basis for a fabrication of the water-cooled scraper.« less

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

Upgrades of edge, divertor and scrape-off layer diagnostics of W7-X for OP1.2

DOE PAGES

Hathiramani, D.; Ali, A.; Anda, G.; ...

2018-02-07

In this work, Wendelstein 7-X (W7-X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7-X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) tomore » an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently.The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. In conclusion, the focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.« less

A new scaling for divertor detachment

NASA Astrophysics Data System (ADS)

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.

Effects of two-dimensional magnetic uncertainties and three-dimensional error and perturbation fields on the Small Angle Slot divertor geometry and topology [Effects of two- and three-dimensional magnetic fields on the Small Angle Slot divertor magnetic topology

DOE PAGES

Trevisan, Gregorio L.; Lao, Lang L.; Evans, Todd E.; ...

2018-01-04

The Small Angle Slot (SAS) was recently installed on DIII-D as an advanced divertor, promising easier plasma detachment and lower temperatures across the whole target. A twofold study of the SAS magnetic topology is presented in this paper. On one hand, a twodimensional uncertainty quantification analysis is carried out through a Monte Carlo approach in order to understand the level of accuracy of two-dimensional equilibrium computations in reconstructing the strike point and angle onto the divertor. Under typical experimental conditions, the uncertainties are found to be roughly 6.8 mm and 0.56 deg, respectively. On the other hand, a three-dimensional “vacuum”more » analysis is carried out to understand the effects of typical external perturbation fields on the scrape-off layer topology. When the threedimensional I-coils are switched on, poloidally-localized lobes are found to appear, grow, and hit the SAS target, although barely, even for 5 kA; at the same time, the strike point modulation is found to be roughly 1.8 mm and thus negligible for most purposes. Furthermore, such results complement previous two-dimensional analyses in characterizing typical SAS equilibria and provide useful background information for planning and interpreting SAS experiments.« less

A new scaling for divertor detachment

DOE PAGES

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

2017-03-29

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 PsepB/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-likemore » 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.« less

Fast wave power flow along SOL field lines in NSTX

NASA Astrophysics Data System (ADS)

Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

2012-10-01

On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

Three-dimensional simulation of H-mode plasmas with localized divertor impurity injection on Alcator C-Mod using the edge transport code EMC3-EIRENE

DOE PAGES

Lore, Jeremy D.; Reinke, M. L.; Brunner, D.; ...

2015-04-28

We study experiments in Alcator C-Mod to assess the level of toroidal asymmetry in divertor conditions resulting from poloidally and toroidally localized extrinsic impurity gas seeding show a weak toroidal peaking (~1.1) in divertor electron temperatures for high-power enhanced D-alpha H-modeplasmas. This is in contrast to similar experiments in Ohmically heated L-modeplasmas, which showed a clear toroidal modulation in the divertor electron temperature. Modeling of these experiments using the 3D edge transport code EMC3-EIRENE [Y. Feng et al., J. Nucl. Mater. 241, 930 (1997)] qualitatively reproduces these trends, and indicates that the different response in the simulations is due tomore » the ionization location of the injected nitrogen. Low electron temperatures in the private flux region (PFR) in L-mode result in a PFR plasma that is nearly transparent to neutral nitrogen, while in H-mode the impurities are ionized in close proximity to the injection location, with this latter case yielding a largely axisymmetric radiation pattern in the scrape-off-layer. In conclusion, the consequences for the ITER gas injection system are discussed. Quantitative agreement with the experiment is lacking in some areas, suggesting potential areas for improving the physics model in EMC3-EIRENE.« less

TEMPEST simulations of the plasma transport in a single-null tokamak geometry

NASA Astrophysics Data System (ADS)

Xu, X. Q.; Bodi, K.; Cohen, R. H.; Krasheninnikov, S.; Rognlien, T. D.

2010-06-01

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. To study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. A series of TEMPEST simulations were conducted to investigate the transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. We also show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.

Plasma-surface interaction in the Be/W environment: Conclusions drawn from the JET-ILW for ITER

NASA Astrophysics Data System (ADS)

Brezinsek, S.; JET-EFDA contributors

2015-08-01

The JET ITER-Like Wall experiment (JET-ILW) provides an ideal test bed to investigate plasma-surface interaction (PSI) and plasma operation with the ITER plasma-facing material selection employing beryllium in the main chamber and tungsten in the divertor. The main PSI processes: material erosion and migration, (b) fuel recycling and retention, (c) impurity concentration and radiation have be1en studied and compared between JET-C and JET-ILW. The current physics understanding of these key processes in the JET-ILW revealed that both interpretation of previously obtained carbon results (JET-C) and predictions to ITER need to be revisited. The impact of the first-wall material on the plasma was underestimated. Main observations are: (a) low primary erosion source in H-mode plasmas and reduction of the material migration from the main chamber to the divertor (factor 7) as well as within the divertor from plasma-facing to remote areas (factor 30 - 50). The energetic threshold for beryllium sputtering minimises the primary erosion source and inhibits multi-step re-erosion in the divertor. The physical sputtering yield of tungsten is low as 10-5 and determined by beryllium ions. (b) Reduction of the long-term fuel retention (factor 10 - 20) in JET-ILW with respect to JET-C. The remaining retention is caused by implantation and co-deposition with beryllium and residual impurities. Outgassing has gained importance and impacts on the recycling properties of beryllium and tungsten. (c) The low effective plasma charge (Zeff = 1.2) and low radiation capability of beryllium reveal the bare deuterium plasma physics. Moderate nitrogen seeding, reaching Zeff = 1.6 , restores in particular the confinement and the L-H threshold behaviour. ITER-compatible divertor conditions with stable semi-detachment were obtained owing to a higher density limit with ILW. Overall JET demonstrated successful plasma operation in the Be/W material combination and confirms its advantageous PSI behaviour and gives strong support to the ITER material selection.

Dynamic ELM and divertor control using resonant toroidal multi-mode magnetic fields in DIII-D and EAST

NASA Astrophysics Data System (ADS)

Sun, Youwen

2017-10-01

A rotating n = 2 Resonant Magnetic Perturbation (RMP) field combined with a stationary n = 3 RMP field has validated predictions that access to ELM suppression can be improved, while divertor heat and particle flux can also be dynamically controlled in DIII-D. Recent observations in the EAST tokamak indicate that edge magnetic topology changes, due to nonlinear plasma response to magnetic perturbations, play a critical role in accessing ELM suppression. MARS-F code MHD simulations, which include the plasma response to the RMP, indicate the nonlinear transition to ELM suppression is optimized by configuring the RMP coils to drive maximal edge stochasticity. Consequently, mixed toroidal multi-mode RMP fields, which produce more densely packed islands over a range of additional rational surfaces, improve access to ELM suppression, and further spread heat loading on the divertor. Beneficial effects of this multi-harmonic spectrum on ELM suppression have been validated in DIII-D. Here, the threshold current required for ELM suppression with a mixed n spectrum, where part of the n = 3 RMP field is replaced by an n = 2 field, is smaller than the case with pure n = 3 field. An important further benefit of this multi-mode approach is that significant changes of 3D particle flux footprint profiles on the divertor are found in the experiment during the application of a rotating n = 2 RMP field superimposed on a static n = 3 RMP field. This result was predicted by modeling studies of the edge magnetic field structure using the TOP2D code which takes into account plasma response from MARS-F code. These results expand physics understanding and potential effectiveness of the technique for reliably controlling ELMs and divertor power/particle loading distributions in future burning plasma devices such as ITER. Work supported by USDOE under DE-FC02-04ER54698 and NNSF of China under 11475224.

The ARIES Advanced and Conservative Tokamak Power Plant Study

DOE PAGES

Kessel, C. E; Tillak, M. S; Najmabadi, F.; ...

2015-12-22

Tokamak power plants are studied with advanced and conservative design philosophies to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared with older studies. The advanced configuration assumes a self-cooled lead lithium blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q₉₅ of 4.5, aᵦ total N of 5.75, an H98 of 1.65,more » an n/n Gr of 1.0, and a peak divertor heat flux of 13.7 MW/m² . The conservative configuration assumes a dual-coolant lead lithium blanket concept with reduced activation ferritic martensitic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma has a major radius of 9.75 m, a toroidal field of 8.75 T, a q₉₅ of 8.0, aᵦ total N of 2.5, an H₉₈ of 1.25, an n/n Gr of 1.3, and a peak divertor heat flux of 10 MW/m² . The divertor heat flux treatment with a narrow power scrape off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range 10 to 15 MW/m² . Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Other papers in this issue provide more detailed discussion of the work summarized here.« less

The ARIES Advanced And Conservative Tokamak (ACT) Power Plant Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kessel, C. E.; Poli, F. M.; Ghantous, K.

2014-03-05

Tokamak power plants are studied with advanced and conservative design philosophies in order to identify the impacts on the resulting designs and to provide guidance to critical research needs. Incorporating updated physics understanding, and using more sophisticated engineering and physics analysis, the tokamak configurations have developed a more credible basis compared to older studies. The advanced configuration assumes a self-cooled lead lithium (SCLL) blanket concept with SiC composite structural material with 58% thermal conversion efficiency. This plasma has a major radius of 6.25 m, a toroidal field of 6.0 T, a q95 of 4.5, a βN total of 5.75, Hmore » 98 of 1.65, n/nGr of 1.0, and peak divertor heat flux of 13.7 MW/m 2. The conservative configuration assumes a dual coolant lead lithium (DCLL) blanket concept with ferritic steel structural material and helium coolant, achieving a thermal conversion efficiency of 45%. The plasma major radius is 9.75 m, a toroidal field of 8.75 T, a q95 of 8.0, a βN total of 2.5, H 98 of 1.25, n/n Gr of 1.3, and peak divertor heat flux of 10 MW/m 2. The divertor heat flux treatment with a narrow power scrape-off width has driven the plasmas to larger major radius. Edge and divertor plasma simulations are targeting a basis for high radiated power fraction in the divertor, which is necessary for solutions to keep the peak heat flux in the range of 10-15 MW/m 2. Combinations of the advanced and conservative approaches show intermediate sizes. A new systems code using a database approach has been used and shows that the operating point is really an operating zone with some range of plasma and engineering parameters and very similar costs of electricity. Papers in this issue provide more detailed discussion of the work summarized here.« less

Divertor, scrape-off layer and pedestal particle dynamics in the ELM cycle on ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Laggner, F. M.; Keerl, S.; Gnilsen, J.; Wolfrum, E.; Bernert, M.; Carralero, D.; Guimarais, L.; Nikolaeva, V.; Potzel, S.; Cavedon, M.; Mink, F.; Dunne, M. G.; Birkenmeier, G.; Fischer, R.; Viezzer, E.; Willensdorfer, M.; Wischmeier, M.; Aumayr, F.; the EUROfusion MST1 Team; the ASDEX Upgrade Team

2018-02-01

In addition to the relaxation of the pedestal, edge localised modes (ELMs) introduce changes to the divertor and scrape-off layer (SOL) conditions. Their impact on the inter-ELM pedestal recovery is investigated, with emphasis on the electron density (n e) evolution. The typical ELM cycle occurring in an exemplary ASDEX Upgrade discharge interval at moderate applied gas puff and heating power is characterised, utilising several divertor, SOL and pedestal diagnostics. In the studied discharge interval the inner divertor target is detached before the ELM crash, while the outer target is attached. The particles and power expelled by the ELM crash lead to a re-attachment of the inner target plasma. After the ELM crash, the outer divertor target moves into a high recycling regime with large n e in front of the plate, which is accompanied by high main chamber neutral fluxes. On similar timescales, the inner target fully detaches and the high field side high density region (HFSHD) is formed reaching up to the high field side midplane. This state evolves again to the pre-ELM state, when the main chamber neutral fluxes are reduced later in the ELM cycle. Neither the timescale of the appearance of the HFSHD nor the increase of the main chamber neutral fluxes fit the timescale of the n e pedestal, which is faster. It is found that during the n e pedestal recovery, the magnetic activity at the low field side midplane is strongly reduced indicating a lower level of fluctuations. A rough estimation of the particle flux across the pedestal suggests that the particle flux is reduced in this period. In conclusion, the evolution of the n e pedestal is determined by a combination of neutral fluxes, HFSHD and reduced particle flux across the pedestal. A reduced particle flux explains the fast, experimentally observed re-establishment of the n e pedestal best, whereas neutrals and HFSHD impact on the evolution of the SOL and separatrix conditions.

Recognition of white matter bundles using local and global streamline-based registration and clustering.

PubMed

Garyfallidis, Eleftherios; Côté, Marc-Alexandre; Rheault, Francois; Sidhu, Jasmeen; Hau, Janice; Petit, Laurent; Fortin, David; Cunanne, Stephen; Descoteaux, Maxime

2018-04-15

Virtual dissection of diffusion MRI tractograms is cumbersome and needs extensive knowledge of white matter anatomy. This virtual dissection often requires several inclusion and exclusion regions-of-interest that make it a process that is very hard to reproduce across experts. Having automated tools that can extract white matter bundles for tract-based studies of large numbers of people is of great interest for neuroscience and neurosurgical planning. The purpose of our proposed method, named RecoBundles, is to segment white matter bundles and make virtual dissection easier to perform. This can help explore large tractograms from multiple persons directly in their native space. RecoBundles leverages latest state-of-the-art streamline-based registration and clustering to recognize and extract bundles using prior bundle models. RecoBundles uses bundle models as shape priors for detecting similar streamlines and bundles in tractograms. RecoBundles is 100% streamline-based, is efficient to work with millions of streamlines and, most importantly, is robust and adaptive to incomplete data and bundles with missing components. It is also robust to pathological brains with tumors and deformations. We evaluated our results using multiple bundles and showed that RecoBundles is in good agreement with the neuroanatomical experts and generally produced more dense bundles. Across all the different experiments reported in this paper, RecoBundles was able to identify the core parts of the bundles, independently from tractography type (deterministic or probabilistic) or size. Thus, RecoBundles can be a valuable method for exploring tractograms and facilitating tractometry studies. Copyright © 2017 Elsevier Inc. All rights reserved.

On the concept of a filtered bundle

NASA Astrophysics Data System (ADS)

Bruce, Andrew James; Grabowska, Katarzyna; Grabowski, Janusz

We present the notion of a filtered bundle as a generalization of a graded bundle. In particular, we weaken the necessity of the transformation laws for local coordinates to exactly respect the weight of the coordinates by allowing more general polynomial transformation laws. The key examples of such bundles include affine bundles and various jet bundles, both of which play fundamental roles in geometric mechanics and classical field theory. We also present the notion of double filtered bundles which provide natural generalizations of double vector bundles and double affine bundles. Furthermore, we show that the linearization of a filtered bundle — which can be seen as a partial polarization of the admissible changes of local coordinates — is well defined.

Anatomy of the anterior cruciate ligament with regard to its two bundles.

PubMed

Petersen, Wolf; Zantop, Thore

2007-01-01

The anterior cruciate ligament (ACL) consists of two major fiber bundles, namely the anteromedial and posterolateral bundle. When the knee is extended, the posterolateral bundle (PL) is tight and the anteromedial (AM) bundle is moderately lax. As the knee is flexed, the femoral attachment of the ACL becomes a more horizontal orientation; causing the AM bundle to tighten and the PL bundle to relax. There is some degree of variability for the femoral origin of the anterome-dial and posterolateral bundle. The anteromedial bundle is located proximal and anterior in the femoral ACL origin (high and deep in the notch when the knee is flexed at 90 degrees ); the posterolateral bundle starts in the distal and posterior aspect of the femoral ACL origin (shallow and low when the knee is flexed at 90 degrees ). In the frontal plane the anteromedial bundle origin is in the 10:30 clock position and the postero-lateral bundle origin in the 9:30 clock position. At the tibial insertion the ACL fans out to form the foot region. The anteromedial bundle insertion is in the anterior part of the tibial ACL footprint, the posterolateral bundle in the posterior part. While the anteromedial bundle is the primary restraint against anterior tibial translation, the posterolateral bundle tends to stabilize the knee near full extension, particularly against rotatory loads.

Overview of innovative PMI research on NSTX-U and associated PMI facilities at PPPL

DOE PAGES

M. Ono; Jaworski, M.; Kaita, R.; ...

2013-05-01

Developing a reactor compatible divertor and managing the associated plasma material interaction (PMI) has been identified as a high priority research area for magnetic confinement fusion. Accordingly on NSTX-U, the PMI research has received a strong emphasis. Moreover, with ˜15 MW of auxiliary heating power, NSTX-U will be able to test the PMI physics with the peak divertor plasma facing component (PFC) heat loads of up to 40-60 MW/m 2.

Ideal plasma response to vacuum magnetic fields with resonant magnetic perturbations in non-axisymmetric tokamaks

DOE PAGES

Kim, Kimin; Ahn, J. -W.; Scotti, F.; ...

2015-09-03

Ideal plasma shielding and amplification of resonant magnetic perturbations in non-axisymmetric tokamak is presented by field line tracing simulation with full ideal plasma response, compared to measurements of divertor lobe structures. Magnetic field line tracing simulations in NSTX with toroidal non-axisymmetry indicate the ideal plasma response can significantly shield/amplify and phase shift the vacuum resonant magnetic perturbations. Ideal plasma shielding for n = 3 mode is found to prevent magnetic islands from opening as consistently shown in the field line connection length profile and magnetic footprints on the divertor target. It is also found that the ideal plasma shielding modifiesmore » the degree of stochasticity but does not change the overall helical lobe structures of the vacuum field for n = 3. Furthermore, amplification of vacuum fields by the ideal plasma response is predicted for low toroidal mode n = 1, better reproducing measurements of strong striation of the field lines on the divertor plate in NSTX.« less

Manufacturing and testing of a prototypical divertor vertical target for ITER

NASA Astrophysics Data System (ADS)

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.

A practical globalization of one-shot optimization for optimal design of tokamak divertors

NASA Astrophysics Data System (ADS)

Blommaert, Maarten; Dekeyser, Wouter; Baelmans, Martine; Gauger, Nicolas R.; Reiter, Detlev

2017-01-01

In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes only state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.

Characterized the pattern of the material deposition in the HL-2A tokamak

NASA Astrophysics Data System (ADS)

Cai, Laizhong; Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao; Hai, Ran; Ding, Hongbin

2017-03-01

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Global transport of light elements boron and carbon in the full-W ASDEX Upgrade

NASA Astrophysics Data System (ADS)

ASDEX Upgrade Team; Hakola, A.; Likonen, J.; Koivuranta, S.; Krieger, K.; Mayer, M.; Neu, R.; Rohde, V.; Sugiyama, K.

2011-08-01

Transport of carbon and boron has been investigated in the full-W ASDEX Upgrade after experimental campaigns with (2008) and without (2007) boronizations. For this purpose, poloidal deposition profiles of the two elements on tungsten and graphite regions of lower-divertor tiles have been determined. Carbon is mainly deposited in the inner divertor - 80-90% of the determined 12C and 13C inventories on W - while boron shows a much more symmetric deposition profile. In the unboronized machine, the boron inventories are a factor of 10 smaller than in the boronized case and result from residual boron atoms left in the torus prior to the 2007 campaign. Both carbon and boron are deposited more efficiently and/or show less erosion on graphite than on tungsten, particularly in the outer divertor. For 13C, the difference is 10-100 in favor of graphite. This is most probably caused by a higher re-erosion from tungsten surfaces.

EMC3-EIRENE modelling of toroidally-localized divertor gas injection experiments on Alcator C-Mod

DOE PAGES

Lore, Jeremy D.; Reinke, M. L.; LaBombard, Brian; ...

2014-09-30

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 modelling, with the simulation yielding a toroidal asymmetry in the heat flow to the outermore » strike point. Finally, 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.« less

Powder Injection Molding for mass production of He-cooled divertor parts

NASA Astrophysics Data System (ADS)

Antusch, S.; Norajitra, P.; Piotter, V.; Ritzhaupt-Kleissl, H.-J.

2011-10-01

A He-cooled divertor for future fusion power plants has been developed at KIT. Tungsten and tungsten alloys are presently considered the most promising materials for functional and structural divertor components. The advantages of tungsten materials lie, e.g. in the high melting point, and low activation, the disadvantages are high hardness and brittleness. The machinig of tungsten, e.g. milling, is very complex and cost-intensive. Powder Injection Molding (PIM) is a method for cost effective mass production of near-net-shape parts with high precision. The complete W-PIM process route is outlined and, results of product examination discussed. A binary tungsten powder feedstock with a grain size distribution in the range 0.7-1.7 μm FSSS, and a solid load of 50 vol.% was developed. After heat treatment, the successfully finished samples showed promising results, i.e. 97.6% theoretical density, a grain size of approximately 5 μm, and a hardness of 457 HV0.1.

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tang, Xianzhu; Guo, Zehua

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Bohm criterion and plasma particle/power exhaust to and recycling at the wall

DOE PAGES

Tang, Xianzhu; Guo, Zehua

2017-06-07

The plasma particle and power exhaust to the divertor surface drives both particle and power recycling at the surface, which in return constrains the plasma density and temperature at the target and their profile further upstream. Both particle and power exhaust fluxes are mediated by the plasma sheath next to the divertor surface. In particular, the Bohm criterion constrains the ion exit flow speed, which enters directly into the particle flux and the kinetic flow energy component of the ion power flux, and indirectly into the electron power flux through the sheath potential drop. Here we give an overview onmore » how the Bohm speed is set in a general plasma and how it enters power exhaust and power recycling at the divertor surface, and the implication on the correct implementation of sheath boundary conditions in numerical codes. The cases of ideal and non-ideal Bohm speed are distinguished as a result of the physics discussion.« less

Operational limits on WEST inertial divertor sector during the early phase experiment

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Corre, Y.; Languille, P.; Greuner, H.; Autissier, E.; Desgranges, C.; Guilhem, D.; Gunn, J. P.; Lipa, M.; Missirlian, M.; Pascal, J.-Y.; Pocheau, C.; Richou, M.; Tsitrone, E.

2016-02-01

The primary goal of the WEST project is to be a test bed to characterize the fatigue and lifetime of ITER-like W divertor components subjected to relevant thermal loads. During the first phase of exploitation (S2 2016), these components (W monoblock plasma facing unit—W-PFU) will be installed in conjunction with graphite components (G-PFU). Since the G-PFU will not be actively cooled, it is necessary to ensure the expected pulse duration allows the W-PFU to reach its steady state without overheating the G-PFU assembly structure or the embedded stainless-steel diagnostics. High heat flux tests were performed at the GLADIS facility to assess the thermal behavior of the G-PFU. Some operational limits based on plasma parameters were determined. It was found that it is possible to operate at an injected power such that the maximal incident heat flux on the lower divertor is 10 MW m-2 for the required pulse length.

Time-to-burnout data for a prototypical ITER divertor tube during a simulated loss of flow accident

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, T.D.; Watson, R.D.; McDonald, J.M.

The Loss of Flow Accident (LOFA) is a serious safety concern for the International Thermonuclear Experimental Reactor (ITER) as it has been suggested that greater than 100 seconds are necessary to safely shutdown the plasma when ITER is operating at full power. In this experiment, the thermal response of a prototypical ITER divertor tube during a simulated LOFA was studied. The divertor tube was fabricated from oxygen-free high-conductivity copper to have a square geometry with a circular coolant channel. The coolant channel inner diameter was 0.77 cm, the heated length was 4.0 cm, and the heated width was 1.6 cm.more » The mockup did not feature any flow enhancement techniques, i.e., swirl tape, helical coils, or internal fins. One-sided surface heating of the mockup was accomplished through the use of the 30 kW Sandia Electron Beam Test System. After reaching steady state temperatures in the mockup, as determined by two Type-K thermocouples installed 0.5 mm beneath the heated surface, the coolant pump was manually tripped off and the coolant flow allowed to naturally coast down. Electron beam heating continued after the pump trip until the divertor tube`s heated surface exhibited the high temperature transient normally indicative of rapidly approaching burnout. Experimental data showed that time-to-burnout increases proportionally with increasing inlet velocity and decreases proportionally with increasing incident heat flux.« less

The lithium vapor box divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

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

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. Our 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 almore » 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. Furthermore, 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 in order 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.« less

The lithium vapor box divertor

NASA Astrophysics Data System (ADS)

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

2016-02-01

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.

Non-solenoidal Startup with High-Field-Side Local Helicity Injection on the Pegasus ST

NASA Astrophysics Data System (ADS)

Perry, J. M.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Pierren, C.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Reusch, J. A.; Weberski, J. D.

2017-10-01

Local Helicity Injection (LHI) is a non-solenoidal startup technique utilizing electron current injectors at the plasma edge to initiate a tokamak-like plasma at high Ip . Recent experiments on Pegasus explore the inherent tradeoffs between high-field-side (HFS) injection in the lower divertor region and low-field-side (LFS) injection at the outboard midplane. Trade-offs include the relative current drive contributions of HI and poloidal induction, and the magnetic geometry required for relaxation to a tokamak-like state. HFS injection using a set of two increased-area injectors (Ainj = 4 cm2, Vinj 1.5 kV, and Iinj 8 kA) in the lower divertor is demonstrated over the full range of toroidal field available on Pegasus (BT 0 <= 0.15 T). Increased PMI on both the injectors and the lower divertor plates was observed during HFS injection, and was substantively mitigated through optimization of injector geometry and placement of local limiters to reduce scrape-off density in the divertor region. Ip up to 200 kA is achieved with LHI as the dominant current drive, consistent with expectations from helicity balance. To date, experiments support Ip increasing linearly with helicity injection rate. The high normalized current (IN >= 10) attainable with LHI and the favorable stability of the ultra-low aspect ratio, low-li LHI-driven plasmas allow access to high βt-up to 100 % , as indicated by kinetically-constrained equilibrium reconstructions. Work supported by US DOE Grant DE-FG02-96ER54375.

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE PAGES

X. Q. Xu; Bodi, K.; Cohen, R. H.; ...

2010-05-28

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

TEMPEST Simulations of the Plasma Transport in a Single-Null Tokamak Geometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

X. Q. Xu; Bodi, K.; Cohen, R. H.

We present edge kinetic ion transport simulations of tokamak plasmas in magnetic divertor geometry using the fully nonlinear (full-f) continuum code TEMPEST. Besides neoclassical transport, a term for divergence of anomalous kinetic radial flux is added to mock up the effect of turbulent transport. In order to study the relative roles of neoclassical and anomalous transport, TEMPEST simulations were carried out for plasma transport and flow dynamics in a single-null tokamak geometry, including the pedestal region that extends across the separatrix into the scrape-off layer and private flux region. In a series of TEMPEST simulations were conducted to investigate themore » transition of midplane pedestal heat flux and flow from the neoclassical to the turbulent limit and the transition of divertor heat flux and flow from the kinetic to the fluid regime via an anomalous transport scan and a density scan. The TEMPEST simulation results demonstrate that turbulent transport (as modelled by large diffusion) plays a similar role to collisional decorrelation of particle orbits and that the large turbulent transport (large diffusion) leads to an apparent Maxwellianization of the particle distribution. Moreover, we show the transition of parallel heat flux and flow at the entrance to the divertor plates from the fluid to the kinetic regime. For an absorbing divertor plate boundary condition, a non-half-Maxwellian is found due to the balance between upstream radial anomalous transport and energetic ion endloss.« less

The lithium vapor box divertor

DOE PAGES

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

2016-01-13

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. Our 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 almore » 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. Furthermore, 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 in order 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.« less

Time-dependent modeling of dust injection in semi-detached ITER divertor plasma

NASA Astrophysics Data System (ADS)

Smirnov, Roman; Krasheninnikov, Sergei

2017-10-01

At present, it is generally understood that dust related issues will play important role in operation of the next step fusion devices, i.e. ITER, and in the development of future fusion reactors. Recent progress in research on dust in magnetic fusion devises has outlined several topics of particular concern: a) degradation of fusion plasma performance; b) impairment of in-vessel diagnostic instruments; and c) safety issues related to dust reactivity and tritium retention. In addition, observed dust events in fusion edge plasmas are highly irregular and require consideration of temporal evolution of both the dust and the fusion plasma. In order to address the dust-related fusion performance issues, we have coupled the dust transport code DUSTT and the edge plasma transport code UEDGE in time-dependent manner, allowing modeling of transient dust-induced phenomena in fusion edge plasmas. Using the coupled codes we simulate burst-like injection of tungsten dust into ITER divertor plasma in semi-detached regime, which is considered as preferable ITER divertor operational mode based on the plasma and heat load control restrictions. Analysis of transport of the dust and the dust-produced impurities, and of dynamics of the ITER divertor and edge plasma in response to the dust injection will be presented. This material is based upon work supported by the U.S. Department of Energy, Office of Science, Office of Fusion Energy Sciences, under Award Number DE-FG02-06ER54852.

Verification of the FBR fuel bundle-duct interaction analysis code BAMBOO by the out-of-pile bundle compression test with large diameter pins

NASA Astrophysics Data System (ADS)

Uwaba, Tomoyuki; Ito, Masahiro; Nemoto, Junichi; Ichikawa, Shoichi; Katsuyama, Kozo

2014-09-01

The BAMBOO computer code was verified by results for the out-of-pile bundle compression test with large diameter pin bundle deformation under the bundle-duct interaction (BDI) condition. The pin diameters of the examined test bundles were 8.5 mm and 10.4 mm, which are targeted as preliminary fuel pin diameters for the upgraded core of the prototype fast breeder reactor (FBR) and for demonstration and commercial FBRs studied in the FaCT project. In the bundle compression test, bundle cross-sectional views were obtained from X-ray computer tomography (CT) images and local parameters of bundle deformation such as pin-to-duct and pin-to-pin clearances were measured by CT image analyses. In the verification, calculation results of bundle deformation obtained by the BAMBOO code analyses were compared with the experimental results from the CT image analyses. The comparison showed that the BAMBOO code reasonably predicts deformation of large diameter pin bundles under the BDI condition by assuming that pin bowing and cladding oval distortion are the major deformation mechanisms, the same as in the case of small diameter pin bundles. In addition, the BAMBOO analysis results confirmed that cladding oval distortion effectively suppresses BDI in large diameter pin bundles as well as in small diameter pin bundles.

Halo current diagnostic system of experimental advanced superconducting tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, D. L.; Shen, B.; Sun, Y.

2015-10-15

The design, calibration, and installation of disruption halo current sensors for the Experimental Advanced Superconducting Tokamak are described in this article. All the sensors are Rogowski coils that surround conducting structures, and all the signals are analog integrated. Coils with two different cross-section sizes have been fabricated, and their mutual inductances are calibrated. Sensors have been installed to measure halo currents in several different parts of both the upper divertor (tungsten) and lower divertor (graphite) at several toroidal locations. Initial measurements from disruptions show that the halo current diagnostics are working well.

Full toroidal imaging of non-axisymmetric plasma material interaction in the National Spherical Torus Experiment divertor.

PubMed

Scotti, Filippo; Roquemore, A L; Soukhanovskii, V A

2012-10-01

A pair of two dimensional fast cameras with a wide angle view (allowing a full radial and toroidal coverage of the lower divertor) was installed in the National Spherical Torus Experiment in order to monitor non-axisymmetric effects. A custom polar remapping procedure and an absolute photometric calibration enabled the easier visualization and quantitative analysis of non-axisymmetric plasma material interaction (e.g., strike point splitting due to application of 3D fields and effects of toroidally asymmetric plasma facing components).

Signal detection by active, noisy hair bundles

NASA Astrophysics Data System (ADS)

O'Maoiléidigh, Dáibhid; Salvi, Joshua D.; Hudspeth, A. J.

2018-05-01

Vertebrate ears employ hair bundles to transduce mechanical movements into electrical signals, but their performance is limited by noise. Hair bundles are substantially more sensitive to periodic stimulation when they are mechanically active, however, than when they are passive. We developed a model of active hair-bundle mechanics that predicts the conditions under which a bundle is most sensitive to periodic stimulation. The model relies only on the existence of mechanotransduction channels and an active adaptation mechanism that recloses the channels. For a frequency-detuned stimulus, a noisy hair bundle's phase-locked response and degree of entrainment as well as its detection bandwidth are maximized when the bundle exhibits low-amplitude spontaneous oscillations. The phase-locked response and entrainment of a bundle are predicted to peak as functions of the noise level. We confirmed several of these predictions experimentally by periodically forcing hair bundles held near the onset of self-oscillation. A hair bundle's active process amplifies the stimulus preferentially over the noise, allowing the bundle to detect periodic forces less than 1 pN in amplitude. Moreover, the addition of noise can improve a bundle's ability to detect the stimulus. Although, mechanical activity has not yet been observed in mammalian hair bundles, a related model predicts that active but quiescent bundles can oscillate spontaneously when they are loaded by a sufficiently massive object such as the tectorial membrane. Overall, this work indicates that auditory systems rely on active elements, composed of hair cells and their mechanical environment, that operate on the brink of self-oscillation.

Crossed Module Bundle Gerbes; Classification, String Group and Differential Geometry

NASA Astrophysics Data System (ADS)

Jurčo, Branislav

We discuss nonabelian bundle gerbes and their differential geometry using simplicial methods. Associated to any crossed module there is a simplicial group NC, the nerve of the 1-category defined by the crossed module and its geometric realization |NC|. Equivalence classes of principal bundles with structure group |NC| are shown to be one-to-one with stable equivalence classes of what we call crossed module gerbes bundle gerbes. We can also associate to a crossed module a 2-category C'. Then there are two equivalent ways how to view classifying spaces of NC-bundles and hence of |NC|-bundles and crossed module bundle gerbes. We can either apply the W-construction to NC or take the nerve of the 2-category C'. We discuss the string group and string structures from this point of view. Also a simplicial principal bundle can be equipped with a simplicial connection and a B-field. It is shown how in the case of a simplicial principal NC-bundle these simplicial objects give the bundle gerbe connection and the bundle gerbe B-field.

Anatomical approach to permanent His bundle pacing: Optimizing His bundle capture.

PubMed

Vijayaraman, Pugazhendhi; Dandamudi, Gopi

2016-01-01

Permanent His bundle pacing is a physiological alternative to right ventricular pacing. In this article we describe our approach to His bundle pacing in patients with AV nodal and intra-Hisian conduction disease. It is essential for the implanters to understand the anatomic variations of the His bundle course and its effect on the type of His bundle pacing achieved. We describe several case examples to illustrate our anatomical approach to permanent His bundle pacing in this article. Copyright © 2016 Elsevier Inc. All rights reserved.

Warps, grids and curvature in triple vector bundles

NASA Astrophysics Data System (ADS)

Flari, Magdalini K.; Mackenzie, Kirill

2018-06-01

A triple vector bundle is a cube of vector bundle structures which commute in the (strict) categorical sense. A grid in a triple vector bundle is a collection of sections of each bundle structure with certain linearity properties. A grid provides two routes around each face of the triple vector bundle, and six routes from the base manifold to the total manifold; the warps measure the lack of commutativity of these routes. In this paper we first prove that the sum of the warps in a triple vector bundle is zero. The proof we give is intrinsic and, we believe, clearer than the proof using decompositions given earlier by one of us. We apply this result to the triple tangent bundle T^3M of a manifold and deduce (as earlier) the Jacobi identity. We further apply the result to the triple vector bundle T^2A for a vector bundle A using a connection in A to define a grid in T^2A . In this case the curvature emerges from the warp theorem.

A New Scaling for Divertor Detachment

NASA Astrophysics Data System (ADS)

Goldston, Robert

2017-10-01

The ITER design and future fusion power plant designs depend on divertor detachment, whether partial, pronounced or complete, both to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. Generally the parallel heat flux, estimated as proportional to Psep / R or Psep B / R , is used as a proxy for the difficulty of achieving detachment. Here we argue that the impurity cooling required for detachment is strongly dependent on the upstream separatrix density, which is limited by Greenwald scaling. Taking this into account self-consistently, along with the Heuristic Drift (HD) model for the SOL width, and using a Lengyel radiation model that includes non-coronal effects, we find that the relative impurity concentration, cz ≡nz /ne , required for detachment scales dominantly as cz Psep /Bp(nsep /nGW) 2 . The absence of any explicit favorable size scaling is concerning, as Psep must increase by an order of magnitude from present experiments to an economic fusion power system, while increases in the poloidal magnetic field strength are limited by magnet technology and MHD stability. This result should not be surprising, as it follows from the simplest scaling, Psep czne2VSOL , taking into account the Greenwald density limit and the HD SOL volume scaling. Reinke has combined a similar approach with the requirement to maintain H-mode, which sets a lower limit on Psep, and also arrives at an incentive for high field and disincentive for large size. These results should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. In particular measurements are required for extrinsic divertor impurity concentration over a range of power and density conditions far from the regime where detachment can be achieved with deuterium puffing and intrinsic impurities alone. Nonetheless, these results suggest that higher magnetic field, stronger shaping, double-null operation, `advanced' divertor magnetic and baffle configurations, as well as lithium vapor targets merit greater attention. This work supported by the US DOE under contract DE-AC02-09CH11466.

Investigation into the formation of the scrape-off layer density shoulder in JET ITER-like wall L-mode and H-mode plasmas

NASA Astrophysics Data System (ADS)

Wynn, A.; Lipschultz, B.; Cziegler, I.; Harrison, J.; Jaervinen, A.; Matthews, G. F.; Schmitz, J.; Tal, B.; Brix, M.; Guillemaut, C.; Frigione, D.; Huber, A.; Joffrin, E.; Kruzei, U.; Militello, F.; Nielsen, A.; Walkden, N. R.; Wiesen, S.; Contributors, JET

2018-05-01

The low temperature boundary layer plasma (scrape-off layer or SOL) between the hot core and the surrounding vessel determines the level of power loading, erosion and implantation of material surfaces, and thus the viability of tokamak-based fusion as an energy source. This study explores mechanisms affecting the formation of flattened density profiles, so-called ‘density shoulders’, in the low-field side (LFS) SOL, which modify ion and neutral fluxes to surfaces—and subsequent erosion. We find that increases in SOL parallel resistivity, Λdiv (=[L || ν eiΩi]/c sΩe), postulated to lead to shoulder growth through changes in SOL turbulence characteristics, correlates with increases in SOL shoulder amplitude, A s, only under a subset of conditions (D2-fuelled L-mode density scans with outer strike point on the horizontal target). Λdiv fails to correlate with A s for cases of N2 seeding or during sweeping of the strike point across the horizontal target. The limited correlation of Λdiv and A s is also found for H-mode discharges. Thus, while it may be necessary for Λdiv to be above a threshold of ~1 for shoulder formation and/or growth, another mechanism is required. More significantly, we find that in contrast to parallel resistivity, outer divertor recycling, as quantified by the total outer divertor Balmer D α emission, I-D α , does scale with A s where Λdiv does and even where Λdiv does not. Divertor recycling could lead to SOL density shoulder formation through: (a) reducing the parallel to the field flow (loss) of ions out of the SOL to the divertor; and (b) changes in radial electric fields which lead to E  ×  B poloidal flows as well as potentially affecting SOL turbulence birth characteristics. Thus, changes in divertor recycling may be the sole process involved in bringing about SOL density shoulders or it may be that it acts in tandem with parallel resistivity.

Evidence and modeling of 3D divertor footprint induced by lower hybrid waves on EAST with tungsten divertor operations

NASA Astrophysics Data System (ADS)

Feng, W.; Wang, L.; Rack, M.; Liang, Y.; Guo, H. Y.; Xu, G. S.; Xu, J. C.; Liu, J. B.; Sun, Y. W.; Jia, M. N.; Yang, Q. Q.; Zhang, B.; Zou, X. L.; Liu, H.; Zhang, T.; Ding, F.; Chen, J. B.; Duan, Y. M.; Zheng, X. W.; Dai, S. Y.; Deng, G. Z.; Chen, R.; Hu, G. H.; Yan, N.; Si, H.; Liu, S. C.; Xu, S.; Wang, M.; Li, M. H.; Ding, B. J.; Wingen, A.; Huang, J.; Gao, X.; Luo, G. N.; Gong, X. Z.; Garofalo, A. M.; Li, J.; Wan, B. N.; the EAST Team

2017-12-01

Three dimensional (3D) divertor particle flux footprints induced by the lower hybrid wave (LHW) have been systematically investigated in the EAST superconducting tokamak during the recent experimental campaign. We find that the striated particle flux (SPF) peaks away from the strike point (SP) closely fit the pitch of the edge magnetic field line for different safety factors q 95, as predicted by a field line tracing code taking into account the helical current filaments (HCFs) in the scrape-off-layer (SOL). As LHW power increases, it requires the fuelling to be increased e.g. by super molecular beam injection (SMBI), to maintain a similar plasma density, which may be attributed to the pump-out effect due to LHW, and may thus be beneficial for EAST steady state operations. The 3D SPF structure is observed with a LHW power threshold (P LHW ~ 0.9 MW). The ratio of the particle fluxes between SPF and outer strike point (OSP), i.e. {{Γ }ion,SPF}/{{Γ }ion,OSP} , increases with the LHW power. Upon transition to divertor detachment, the particle flux at the main OSP decreases, as expected, however, the particle flux at SPF continues increasing, in contrast to the RMP-induced striations that vanish with increasing divertor density. In addition, we also find that the in-out asymmetry of the 3D particle flux footprint pattern exhibits a clear dependence on the toroidal field direction (B  ×    ∇   B  ↓  and B  ×    ∇   B↑). Experiments using neon impurity seeding show a promising capability in 3D particle and heat flux control on EAST. LHW-induced particle and heat flux striations are also present in the H-mode plasmas, reducing the peak heat flux and erosion at the main strike point, thus facilitating long-pulse operation with a new steady-state H-mode over 60 s being recently achieved in EAST.

Steady-state stiffness of utricular hair cells depends on macular location and hair bundle structure.

PubMed

Spoon, Corrie; Moravec, W J; Rowe, M H; Grant, J W; Peterson, E H

2011-12-01

Spatial and temporal properties of head movement are encoded by vestibular hair cells in the inner ear. One of the most striking features of these receptors is the orderly structural variation in their mechanoreceptive hair bundles, but the functional significance of this diversity is poorly understood. We tested the hypothesis that hair bundle structure is a significant contributor to hair bundle mechanics by comparing structure and steady-state stiffness of 73 hair bundles at varying locations on the utricular macula. Our first major finding is that stiffness of utricular hair bundles varies systematically with macular locus. Stiffness values are highest in the striola, near the line of hair bundle polarity reversal, and decline exponentially toward the medial extrastriola. Striolar bundles are significantly more stiff than those in medial (median: 8.9 μN/m) and lateral (2.0 μN/m) extrastriolae. Within the striola, bundle stiffness is greatest in zone 2 (106.4 μN/m), a band of type II hair cells, and significantly less in zone 3 (30.6 μN/m), which contains the only type I hair cells in the macula. Bathing bundles in media that break interciliary links produced changes in bundle stiffness with predictable time course and magnitude, suggesting that links were intact in our standard media and contributed normally to bundle stiffness during measurements. Our second major finding is that bundle structure is a significant predictor of steady-state stiffness: the heights of kinocilia and the tallest stereocilia are the most important determinants of bundle stiffness. Our results suggest 1) a functional interpretation of bundle height variability in vertebrate vestibular organs, 2) a role for the striola in detecting onset of head movement, and 3) the hypothesis that differences in bundle stiffness contribute to diversity in afferent response dynamics.

Steady-state stiffness of utricular hair cells depends on macular location and hair bundle structure

PubMed Central

Spoon, Corrie; Moravec, W. J.; Rowe, M. H.; Grant, J. W.

2011-01-01

Spatial and temporal properties of head movement are encoded by vestibular hair cells in the inner ear. One of the most striking features of these receptors is the orderly structural variation in their mechanoreceptive hair bundles, but the functional significance of this diversity is poorly understood. We tested the hypothesis that hair bundle structure is a significant contributor to hair bundle mechanics by comparing structure and steady-state stiffness of 73 hair bundles at varying locations on the utricular macula. Our first major finding is that stiffness of utricular hair bundles varies systematically with macular locus. Stiffness values are highest in the striola, near the line of hair bundle polarity reversal, and decline exponentially toward the medial extrastriola. Striolar bundles are significantly more stiff than those in medial (median: 8.9 μN/m) and lateral (2.0 μN/m) extrastriolae. Within the striola, bundle stiffness is greatest in zone 2 (106.4 μN/m), a band of type II hair cells, and significantly less in zone 3 (30.6 μN/m), which contains the only type I hair cells in the macula. Bathing bundles in media that break interciliary links produced changes in bundle stiffness with predictable time course and magnitude, suggesting that links were intact in our standard media and contributed normally to bundle stiffness during measurements. Our second major finding is that bundle structure is a significant predictor of steady-state stiffness: the heights of kinocilia and the tallest stereocilia are the most important determinants of bundle stiffness. Our results suggest 1) a functional interpretation of bundle height variability in vertebrate vestibular organs, 2) a role for the striola in detecting onset of head movement, and 3) the hypothesis that differences in bundle stiffness contribute to diversity in afferent response dynamics. PMID:21918003

Double-bundle anterior cruciate ligament reconstruction is superior to single-bundle reconstruction in terms of revision frequency: a study of 22,460 patients from the Swedish National Knee Ligament Register.

PubMed

Svantesson, Eleonor; Sundemo, David; Hamrin Senorski, Eric; Alentorn-Geli, Eduard; Musahl, Volker; Fu, Freddie H; Desai, Neel; Stålman, Anders; Samuelsson, Kristian

2017-12-01

Studies comparing single- and double-bundle anterior cruciate ligament (ACL) reconstructions often include a combined analysis of anatomic and non-anatomic techniques. The purpose of this study was to compare the revision rates between single- and double-bundle ACL reconstructions in the Swedish National Knee Ligament Register with regard to surgical variables as determined by the anatomic ACL reconstruction scoring checklist (AARSC). Patients from the Swedish National Knee Ligament Register who underwent either single- or double-bundle ACL reconstruction with hamstring tendon autograft during the period 2007-2014 were included. The follow-up period started with primary ACL reconstruction, and the outcome measure was set as revision surgery. An online questionnaire based on the items of the AARSC was used to determine the surgical technique implemented in the single-bundle procedures. These were organized into subgroups based on surgical variables, and the revision rates were compared with the double-bundle ACL reconstruction. Hazard ratios (HR) with 95% confidence interval (CI) was calculated and adjusted for confounders by Cox regression. A total of 22,460 patients were included in the study, of which 21,846 were single-bundle and 614 were double-bundle ACL reconstruction. Double-bundle ACL reconstruction had a revision frequency of 2.0% (n = 12) and single-bundle 3.2% (n = 689). Single-bundle reconstruction had an increased risk of revision surgery compared with double-bundle [adjusted HR 1.98 (95% CI 1.12-3.51), p = 0.019]. The subgroup analysis showed a significantly increased risk of revision surgery in patients undergoing single-bundle with anatomic technique using transportal drilling [adjusted HR 2.51 (95% CI 1.39-4.54), p = 0.002] compared with double-bundle ACL reconstruction. Utilizing a more complete anatomic technique according to the AARSC lowered the hazard rate considerably when transportal drilling was performed but still resulted in significantly increased risk of revision surgery compared with double-bundle ACL reconstruction [adjusted HR 1.87 (95% CI 1.04-3.38), p = 0.037]. Double-bundle ACL reconstruction is associated with a lower risk of revision surgery than single-bundle ACL reconstruction. Single-bundle procedures performed using transportal femoral drilling technique had significantly higher risk of revision surgery compared with double-bundle. However, a reference reconstruction with transportal drilling defined as a more complete anatomic reconstruction reduces the risk of revision surgery considerably. III.

Modelling of mitigation of the power divertor loading for the EU DEMO through Ar injection

NASA Astrophysics Data System (ADS)

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.

Moisture separator reheater with round tube bundle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Byerley, W. M.

1984-11-27

A moisture separator reheater having a central chamber with cylindrical wall protions and a generally round tube bundle, the tube bundle having arcuate plates disposed on each side of the bundle which form a wrapper on each side of the bundle and having a tongue and groove juncture between the wrapper and cylindrical wall portions to provide a seal therebetween and a track for installing and removing the tube bundle from the central chamber.

["Habitual" left branch block alternating with 2 "disguised" bracnch block].

PubMed

Lévy, S; Jullien, G; Mathieu, P; Mostefa, S; Gérard, R

1976-10-01

Two cases of alternating left bundle branch block and "masquerading block" (with left bundle branch morphology in the stnadard leads and right bundle branch block morphology in the precordial leads) were studied by serial tracings and his bundle electrocardiography. In case 1 "the masquerading" block was associated with a first degree AV block related to a prolongation of HV interval. This case is to our knowledge the first cas of alternating bundle branch block in which his bundle activity was recorded in man. In case 2, the patient had atrial fibrilation and His bundle recordings were performed while differents degrees of left bundle branch block were present: The mechanism of the alternation and the concept of "masquerading" block are discussed. It is suggested that this type of block represents a right bundle branch block associated with severe lesions of the "left system".

Characterization of active hair-bundle motility by a mechanical-load clamp

NASA Astrophysics Data System (ADS)

Salvi, Joshua D.; Maoiléidigh, Dáibhid Ó.; Fabella, Brian A.; Tobin, Mélanie; Hudspeth, A. J.

2015-12-01

Active hair-bundle motility endows hair cells with several traits that augment auditory stimuli. The activity of a hair bundle might be controlled by adjusting its mechanical properties. Indeed, the mechanical properties of bundles vary between different organisms and along the tonotopic axis of a single auditory organ. Motivated by these biological differences and a dynamical model of hair-bundle motility, we explore how adjusting the mass, drag, stiffness, and offset force applied to a bundle control its dynamics and response to external perturbations. Utilizing a mechanical-load clamp, we systematically mapped the two-dimensional state diagram of a hair bundle. The clamp system used a real-time processor to tightly control each of the virtual mechanical elements. Increasing the stiffness of a hair bundle advances its operating point from a spontaneously oscillating regime into a quiescent regime. As predicted by a dynamical model of hair-bundle mechanics, this boundary constitutes a Hopf bifurcation.

Bundled payments in orthopedic surgery.

PubMed

Bushnell, Brandon D

2015-02-01

As a result of reading this article, physicians should be able to: 1. Describe the concept of bundled payments and the potential applications of bundled payments in orthopedic surgery. 2. For specific situations, outline a clinical episode of care, determine the participants in a bundling situation, and define care protocols and pathways. 3. Recognize the importance of resource utilization management, quality outcome measurement, and combined economic-clinical value in determining the value of bundled payment arrangements. 4. Identify the implications of bundled payments for practicing orthopedists, as well as the legal issues and potential future directions of this increasingly popular alternative payment method. Bundled payments, the idea of paying a single price for a bundle of goods and services, is a financial concept familiar to most American consumers because examples appear in many industries. The idea of bundled payments has recently gained significant momentum as a financial model with the potential to decrease the significant current costs of health care. Orthopedic surgery as a field of medicine is uniquely positioned for success in an environment of bundled payments. This article reviews the history, logistics, and implications of the bundled payment model relative to orthopedic surgery. Copyright 2015, SLACK Incorporated.

Compatibility of separatrix density scaling for divertor detachment with H-mode pedestal operation in DIII-D

NASA Astrophysics Data System (ADS)

Leonard, A. W.; McLean, A. G.; Makowski, M. A.; Stangeby, P. C.

2017-08-01

The midplane separatrix density is characterized in response to variations in upstream parallel heat flux density and central density through deuterium gas injection. The midplane density is determined from a high spatial resolution Thomson scattering diagnostic at the midplane with power balance analysis to determine the separatrix location. The heat flux density is varied by scans of three parameters, auxiliary heating, toroidal field with fixed plasma current, and plasma current with fixed safety factor, q 95. The separatrix density just before divertor detachment onset is found to scale consistent with the two-point model when radiative dissipation is taken into account. The ratio of separatrix to pedestal density, n e,sep/n e,ped varies from  ⩽30% to  ⩾60% over the dataset, helping to resolve the conflicting scaling of core plasma density limit and divertor detachment onset. The scaling of the separatrix density at detachment onset is combined with H-mode power threshold scaling to obtain a scaling ratio of minimum n e,sep/n e,ped expected in future devices.

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

DOE Office of Scientific and Technical Information (OSTI.GOV)

Budaev, V. P., E-mail: budaev@mail.ru

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 ofmore » 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.« less

Divertor tungsten tile melting and its effect on core plasma performance

NASA Astrophysics Data System (ADS)

Lipschultz, B.; Coenen, J. W.; Barnard, H. S.; Howard, N. T.; Reinke, M. L.; Whyte, D. G.; Wright, G. M.

2012-12-01

For the 2007 and 2008 run campaigns, Alcator C-Mod operated with a full toroidal row of tungsten tiles in the high heat flux region of the outer divertor; tungsten levels in the core plasma were below measurement limits. An accidental creation of a tungsten leading edge in the 2009 campaign led to this study of a melting tungsten source: H-mode operation with strike point in the region of the melting tile was immediately impossible due to some fraction of tungsten droplets reaching the main plasma. Approximately 15 g of tungsten was lost from the tile over ˜100 discharges. Less than 1% of the evaporated tungsten was found re-deposited on surfaces, the rest is assumed to have become dust. The strong discharge variability of the tungsten reaching the core implies that the melt layer topology is always varying. There is no evidence of healing of the surface with repeated melting. Forces on the melted tungsten tend to lead to prominences that extend further into the plasma. A discussion of the implications of melting a divertor tungsten monoblock on the ITER plasma is presented.

Overview of experimental preparation for the ITER-Like Wall at JET

NASA Astrophysics Data System (ADS)

Jet Efda Contributors Brezinsek, S.; Fundamenski, W.; Eich, T.; Coad, J. P.; Giroud, C.; Huber, A.; Jachmich, S.; Joffrin, E.; Krieger, K.; McCormick, K.; Lehnen, M.; Loarer, T.; de La Luna, E.; Maddison, G.; Matthews, G. F.; Mertens, Ph.; Nunes, I.; Philipps, V.; Riccardo, V.; Rubel, M.; Stamp, M. F.; Tsalas, M.

2011-08-01

Experiments in JET with carbon-based plasma-facing components have been carried out in preparation of the ITER-Like Wall with beryllium main chamber and full tungsten divertor. The preparatory work was twofold: (i) development of techniques, which ensure safe operation with the new wall and (ii) provision of reference plasmas, which allow a comparison of operation with carbon and metallic wall. (i) Compatibility with the W divertor with respect to energy loads could be achieved in N2 seeded plasmas at high densities and low temperatures, finally approaching partial detachment, with only moderate confinement reduction of 10%. Strike-point sweeping increases the operational space further by re-distributing the load over several components. (ii) Be and C migration to the divertor has been documented with spectroscopy and QMBs under different plasma conditions providing a database which will allow a comparison of the material transport to remote areas with metallic walls. Fuel retention rates of 1.0-2.0 × 1021 D s-1 were obtained as references in accompanied gas balance studies.

Modelling the detachment dependence on strike point location in the small angle slot divertor (SAS) with SOLPS

NASA Astrophysics Data System (ADS)

Casali, Livia; Covele, Brent; Guo, Houyang

2017-10-01

The new Small Angle Slot (SAS) divertor in DIII-D is characterized by a shallow-angle target enclosed by a slot structure about the strike point (SP). SOLPS modelling results of SAS have demonstrated divertor closure's utility in widening the range of acceptable densities for adequate heat handling. An extensive database of runs has been built to study the detachment dependence on SP location in SAS. Density scans show that lower Te at lower upstream density occur when the SP is at the critical location in the slot. The cooling front spreads across the entire target at higher densities, in agreement with experimental Langmuir probe measurements. A localized increase of the atomic and molecular density takes place near the SP, which reduces the target incident power density and facilitates detachment at lower upstream density. Systematic scans of variables such as power, transport, and viscosity have been carried out to assess the detachment sensitivity. Therein, a positive role of the viscosity is found. This work supported by DOE Contract Number DE-FC02-04ER54698.

A practical globalization of one-shot optimization for optimal design of tokamak divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blommaert, Maarten, E-mail: maarten.blommaert@kuleuven.be; Dekeyser, Wouter; Baelmans, Martine

In past studies, nested optimization methods were successfully applied to design of the magnetic divertor configuration in nuclear fusion reactors. In this paper, so-called one-shot optimization methods are pursued. Due to convergence issues, a globalization strategy for the one-shot solver is sought. Whereas Griewank introduced a globalization strategy using a doubly augmented Lagrangian function that includes primal and adjoint residuals, its practical usability is limited by the necessity of second order derivatives and expensive line search iterations. In this paper, a practical alternative is offered that avoids these drawbacks by using a regular augmented Lagrangian merit function that penalizes onlymore » state residuals. Additionally, robust rank-two Hessian estimation is achieved by adaptation of Powell's damped BFGS update rule. The application of the novel one-shot approach to magnetic divertor design is considered in detail. For this purpose, the approach is adapted to be complementary with practical in parts adjoint sensitivities. Using the globalization strategy, stable convergence of the one-shot approach is achieved.« less

Numerical investigation of disruption characteristics for the snowflake divertor configuration in HL-2M

NASA Astrophysics Data System (ADS)

Xue, L.; Duan, X. R.; Zheng, G. Y.; Liu, Y. Q.; Pan, Y. D.; Yan, S. L.; Dokuka, V. N.; Lukash, V. E.; Khayrutdinov, R. R.

2016-05-01

Cold and hot vertical displacement events (VDEs) are frequently related to the disruption of vertically-elongated tokamaks. The weak poloidal magnetic field around the null-points of a snowflake divertor configuration may influence the vertical displacement process. In this paper, the major disruption with a cold VDE and the vertical disruption in the HL-2M tokamak are investigated by the DINA code. In order to better illustrate the effect from the weak poloidal field, a double-null snowflake configuration is compared with the standard divertor (SD) configuration under the same plasma parameters. Computational results show that the weak poloidal magnetic field can be partly beneficial for mitigating the vertical instability of the plasma under small perturbations. For major disruption, the peak poloidal halo current fraction is almost the same between the snowflake and the SD configurations. However, this fraction becomes much larger for the snowflake in the event of a hot VDE. Furthermore, during the disruption for a snowflake configuration, the distribution of electromagnetic force on a vacuum vessel gets more non-uniform during the current quench.

Measurements of C V flows from thermal charge-exchange excitation in divertor plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zaniol, B.; Isler, R. C.; Brooks, N. H.

2001-10-01

Certain transitions of C IV (C{sup 3+}) from n=7 to n=6 ({approx}7226 {angstrom}) and from n=6 to n=5 ({approx}4660 {angstrom}) sometimes appear much brighter in tokamak divertors than expected for electron-impact excitation from the ground state. This situation occurs because of charge exchange between C V (C{sup 4+}) and recycling thermal deuterium atoms in the n=2 level. As a result, it is possible to extend parallel flow measurements of carbon, which have previously been performed on C II--C IV ions using Doppler shift spectroscopy, to include flows of the He-like C V ions. The work described here includes modeling ofmore » the spectral features, correlation of state populations with classical Monte Carlo trajectory (CTMC) predictions, and applications to flow measurements in the DIII-D divertor [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515].« less

Measurements of C V flows from thermal charge-exchange excitation in divertor plasmas

NASA Astrophysics Data System (ADS)

Zaniol, B.; Isler, R. C.; Brooks, N. H.; West, W. P.; Olson, R. E.

2001-10-01

Certain transitions of C IV (C3+) from n=7 to n=6 (≈7226 Å) and from n=6 to n=5 (≈4660 Å) sometimes appear much brighter in tokamak divertors than expected for electron-impact excitation from the ground state. This situation occurs because of charge exchange between C V (C4+) and recycling thermal deuterium atoms in the n=2 level. As a result, it is possible to extend parallel flow measurements of carbon, which have previously been performed on C II-C IV ions using Doppler shift spectroscopy, to include flows of the He-like C V ions. The work described here includes modeling of the spectral features, correlation of state populations with classical Monte Carlo trajectory (CTMC) predictions, and applications to flow measurements in the DIII-D divertor [Plasma Physics Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515].

Partial detachment of high power discharges in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Bernert, M.; Beurskens, M.; Casali, L.; Dunne, M.; Eich, T.; Giannone, L.; Herrmann, A.; Maraschek, M.; Potzel, S.; Reimold, F.; Rohde, V.; Schweinzer, J.; Viezzer, E.; Wischmeier, M.; the ASDEX Upgrade Team

2015-05-01

Detachment of high power discharges is obtained in ASDEX Upgrade by simultaneous feedback control of core radiation and divertor radiation or thermoelectric currents by the injection of radiating impurities. So far 2/3 of the ITER normalized heat flux Psep/R = 15 MW m-1 has been obtained in ASDEX Upgrade under partially detached conditions with a peak target heat flux well below 10 MW m-2. When the detachment is further pronounced towards lower peak heat flux at the target, substantial changes in edge localized mode (ELM) behaviour, density and radiation distribution occur. The time-averaged peak heat flux at both divertor targets can be reduced below 2 MW m-2, which offers an attractive DEMO divertor scenario with potential for simpler and cheaper technical solutions. Generally, pronounced detachment leads to a pedestal and core density rise by about 20-40%, moderate (<20%) confinement degradation and a reduction of ELM size. For AUG conditions, some operational challenges occur, like the density cut-off limit for X-2 electron cyclotron resonance heating, which is used for central tungsten control.

Calculation of ion distribution functions and neoclassical transport in the edge of single-null divertor tokamaks

NASA Astrophysics Data System (ADS)

Rognlien, T. D.; Cohen, R. H.; Xu, X. Q.

2007-11-01

The ion distribution function in the H-mode pedestal region and outward across the magnetic separatrix is expected to have a substantial non-Maxwellian character owing to the large banana orbits and steep gradients in temperature and density. The 4D (2r,2v) version of the TEMPEST continuum gyrokinetic code is used with a Coulomb collision model to calculate the ion distribution in a single-null tokamak geometry throughout the pedestal/scrape-off-layer regions. The mean density, parallel velocity, and energy radial profiles are shown at various poloidal locations. The collisions cause neoclassical energy transport through the pedestal that is then lost to the divertor plates along the open field lines outside the separatrix. The resulting heat flux profiles at the inner and outer divertor plates are presented and discussed, including asymmetries that depend on the B-field direction. Of particular focus is the effect on ion profiles and fluxes of a radial electric field exhibiting a deep well just inside the separatrix, which reduces the width of the banana orbits by the well-known squeezing effect.

Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Raffray, A.R.; Chiocchio, S.

1995-12-31

This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC`s) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m{sup 2}) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM`s) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects themore » target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC`s clad with different PFM`s are discussed.« less

Quantitative study of bundle size effect on thermal conductivity of single-walled carbon nanotubes

NASA Astrophysics Data System (ADS)

Feng, Ya; Inoue, Taiki; An, Hua; Xiang, Rong; Chiashi, Shohei; Maruyama, Shigeo

2018-05-01

Compared with isolated single-walled carbon nanotubes (SWNTs), thermal conductivity is greatly impeded in SWNT bundles; however, the measurement of the bundle size effect is difficult. In this study, the number of SWNTs in a bundle was determined based on the transferred horizontally aligned SWNTs on a suspended micro-thermometer to quantitatively study the effect of the bundle size on thermal conductivity. Increasing the bundle size significantly degraded the thermal conductivity. For isolated SWNTs, thermal conductivity was approximately 5000 ± 1000 W m-1 K-1 at room temperature, three times larger than that of the four-SWNT bundle. The logarithmical deterioration of thermal conductivity resulting from the increased bundle size can be attributed to the increased scattering rate with neighboring SWNTs based on the kinetic theory.

Cations Modulate Actin Bundle Mechanics, Assembly Dynamics, and Structure.

PubMed

Castaneda, Nicholas; Zheng, Tianyu; Rivera-Jacquez, Hector J; Lee, Hyun-Ju; Hyun, Jaekyung; Balaeff, Alexander; Huo, Qun; Kang, Hyeran

2018-04-12

Actin bundles are key factors in the mechanical support and dynamic reorganization of the cytoskeleton. High concentrations of multivalent counterions promote bundle formation through electrostatic attraction between actin filaments that are negatively charged polyelectrolytes. In this study, we evaluate how physiologically relevant divalent cations affect the mechanical, dynamic, and structural properties of actin bundles. Using a combination of total internal reflection fluorescence microscopy, transmission electron microscopy, and dynamic light scattering, we demonstrate that divalent cations modulate bundle stiffness, length distribution, and lateral growth. Molecular dynamics simulations of an all-atom model of the actin bundle reveal specific actin residues coordinate cation-binding sites that promote the bundle formation. Our work suggests that specific cation interactions may play a fundamental role in the assembly, structure, and mechanical properties of actin bundles.

Spinal surgery: variations in health care costs and implications for episode-based bundled payments.

PubMed

Ugiliweneza, Beatrice; Kong, Maiying; Nosova, Kristin; Huang, Kevin T; Babu, Ranjith; Lad, Shivanand P; Boakye, Maxwell

2014-07-01

Retrospective, observational. To simulate what episodes of care in spinal surgery might look like in a bundled payment system and to evaluate the associated costs and characteristics. Episode-based payment bundling has received considerable attention as a potential method to help curb the rise in health care spending and is being investigated as a new payment model as part of the Affordable Care Act. Although earlier studies investigated bundled payments in a number of surgical settings, very few focused on spine surgery, specifically. We analyzed data from MarketScan. Patients were included in the study if they underwent cervical or lumbar spinal surgery during 2000-2009, had at least 2-year preoperative and 90-day postoperative follow-up data. Patients were grouped on the basis of their diagnosis-related group (DRG) and then tracked in simulated episodes-of-care/payment bundles that lasted for the duration of 30, 60, and 90 days after the discharge from the index-surgical hospitalization. The total cost associated with each episode-of-care duration was measured and characterized. A total of 196,918 patients met our inclusion criteria. Significant variation existed between DRGs, ranging from $11,180 (30-day bundle, DRG 491) to $107,642 (30-day bundle, DRG 456). There were significant cost variations within each individual DRG. Postdischarge care accounted for a relatively small portion of overall bundle costs (range, 4%-8% in 90-day bundles). Total bundle costs remained relatively flat as bundle-length increased (total average cost of 30-day bundle: $33,522 vs. $35,165 for 90-day bundle). Payments to hospitals accounted for the largest portion of bundle costs (76%). There exists significant variation in total health care costs for patients who undergo spinal surgery, even within a given DRG. Better characterization of impacts of a bundled payment system in spine surgery is important for understanding the costs of index procedure hospital, physician services, and postoperative care on potential future health care policy decision making. N/A.

Nurses' perceptions of a pressure ulcer prevention care bundle: a qualitative descriptive study.

PubMed

Roberts, Shelley; McInnes, Elizabeth; Wallis, Marianne; Bucknall, Tracey; Banks, Merrilyn; Chaboyer, Wendy

2016-01-01

Pressure ulcer prevention is a critical patient safety indicator for acute care hospitals. An innovative pressure ulcer prevention care bundle targeting patient participation in their care was recently tested in a cluster randomised trial in eight Australian hospitals. Understanding nurses' perspectives of such an intervention is imperative when interpreting results and translating evidence into practice. As part of a process evaluation for the main trial, this study assessed nurses' perceptions of the usefulness and impact of a pressure ulcer prevention care bundle intervention on clinical practice. This qualitative descriptive study involved semi-structured interviews with nursing staff at four Australian hospitals that were intervention sites for a cluster randomised trial testing a pressure ulcer prevention care bundle. Four to five participants were purposively sampled at each site. A trained interviewer used a semi-structured interview guide to question participants about their perceptions of the care bundle. Interviews were digitally recorded, transcribed and analysed using thematic analysis. Eighteen nurses from four hospitals participated in the study. Nurses' perceptions of the intervention are described in five themes: 1) Awareness of the pressure ulcer prevention care bundle and its similarity to current practice; 2) Improving awareness, communication and participation with the pressure ulcer prevention care bundle; 3) Appreciating the positive aspects of patient participation in care; 4) Perceived barriers to engaging patients in the pressure ulcer prevention care bundle; and 5) Partnering with nursing staff to facilitate pressure ulcer prevention care bundle implementation. Overall, nurses found the care bundle feasible and acceptable. They identified a number of benefits from the bundle, including improved communication, awareness and participation in pressure ulcer prevention care among patients and staff. However, nurses thought the care bundle was not appropriate or effective for all patients, such as those who were cognitively impaired. Perceived enablers to implementation of the bundle included facilitation through effective communication and dissemination of evidence about the care bundle; strong leadership and ability to influence staff behaviour; and simplicity of the care bundle.

MOX fuel assembly design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reese, A.P.; Crowther, R.L. Jr.

1992-02-18

This patent describes improvement in a boiling water reactor core having a plurality of vertically upstanding fuel bundles; each fuel bundle containing longitudinally extending sealed rods with fissile material therein; the improvement comprises the fissile material including a mixture of uranium and recovered plutonium in rods of the fuel bundle at locations other than the corners of the fuel bundle; and, neutron absorbing material being located in rods of the fuel bundle at rod locations adjacent the corners of the fuel bundles whereby the neutron absorbing material has decreased shielding from the plutonium and maximum exposure to thermal neutrons formore » shaping the cold reactivity shutdown zone in the fuel bundle.« less

Some problems of brazing technology for the divertor plate manufacturing

NASA Astrophysics Data System (ADS)

Prokofiev, Yu. G.; Barabash, V. R.; Khorunov, V. F.; Maksimova, S. V.; Gervash, A. A.; Fabritsiev, S. A.; Vinokurov, V. F.

1992-09-01

Among the different design options of the ITER reactor divertor, the joints of the carbon-based materials and molybdenum alloys and joints of tungsten and copper alloys are considered. High-temperature brazing is one of the most promising joining methods for the plasma facing and heat sink materials. The use of brazing for creation of W-Cu and graphite-Mo joints are given here. In addition, the investigation results of microstructure, microhardness and mechanical properties of the joints are presented. For W-Cu samples an influence of the neutron irradiation on the joining strength was studied.

Divertor for use in fusion reactors

DOEpatents

Christensen, Uffe R.

1979-01-01

A poloidal divertor for a toroidal plasma column ring having a set of poloidal coils co-axial with the plasma ring for providing a space for a thick shielding blanket close to the plasma along the entire length of the plasma ring cross section and all the way around the axis of rotation of the plasma ring. The poloidal coils of this invention also provide a stagnation point on the inside of the toroidal plasma column ring, gently curving field lines for vertical stability, an initial plasma current, and the shaping of the field lines of a separatrix up and around the shielding blanket.

Polycation induced actin bundles.

PubMed

Muhlrad, Andras; Grintsevich, Elena E; Reisler, Emil

2011-04-01

Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon the addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations and the neutralization of repulsive interactions of negative charges on actin. The attractive forces between the filaments are strong, as shown by the low (in nanomolar range) critical concentration of their bundling at low ionic strength. These bundles are sensitive to ionic strength and disassemble partially in 100 mM NaCl, but both the dissociation and ionic strength sensitivity can be countered by higher polycation concentrations. Cys374 residues of actin monomers residing on neighboring filaments in the bundles can be cross-linked by the short span (5.4Å) MTS-1 (1,1-methanedyl bismethanethiosulfonate) cross-linker, which indicates a tight packing of filaments in the bundles. The interfilament cross-links, which connect monomers located on oppositely oriented filaments, prevent disassembly of bundles at high ionic strength. Cofilin and the polysaccharide polyanion heparin disassemble lysozyme induced actin bundles more effectively than the polylysine-induced bundles. The actin-lysozyme bundles are pathologically significant as both proteins are found in the pulmonary airways of cystic fibrosis patients. Their bundles contribute to the formation of viscous mucus, which is the main cause of breathing difficulties and eventual death in this disorder. Copyright © 2011 Elsevier B.V. All rights reserved.

Post-natal molecular adaptations in anteromedial and posterolateral bundles of the ovine anterior cruciate ligament: one structure with two parts or two distinct ligaments?

PubMed

Huebner, Kyla D; O'Brien, Etienne J O; Heard, Bryan J; Chung, May; Achari, Yamini; Shrive, Nigel G; Frank, Cyril B

2012-01-01

The human anterior cruciate ligament (ACL) is a composite structure of two anatomically distinct bundles: an anteromedial (AM) and posterolateral (PL) bundles. Tendons are often used as autografts for surgical reconstruction of ACL following severe injury. However, despite successful surgical reconstruction, some people experience re-rupture and later development of osteoarthritis. Understanding the structure and molecular makeup of normal ACL is essential for its optimal replacement. Reportedly the two bundles display different tensions throughout joint motion and may be fundamentally different. This study assessed the similarities and differences in ultrastructure and molecular composition of the AM and PL bundles to test the hypothesis that the two bundles of the ACL develop unique characteristics with maturation. ACLs from nine mature and six immature sheep were compared. The bundles were examined for mRNA and protein levels of collagen types I, III, V, and VI, and two proteoglycans. The fibril diameter composition of the two bundles was examined with transmission electron microscopy. Maturation does alter the molecular and structural composition of the two bundles of ACL. Although the PL band appears to mature slower than the AM band, no significant differences were detected between the bundles in the mature animals. We thus reject our hypothesis that the two ACL bundles are distinct. The two anatomically distinct bundles of the sheep ACL can be considered as two parts of one structure at maturity and material that would result in a structure of similar functionality can be used to replace each ACL bundle in the sheep.

An exploration of advanced X-divertor scenarios on ITER

NASA Astrophysics Data System (ADS)

Covele, B.; Valanju, P.; Kotschenreuther, M.; Mahajan, S.

2014-07-01

It is found that the X-divertor (XD) configuration (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA), CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502) can be made with the conventional poloidal field (PF) coil set on ITER (Tomabechi et al and Team 1991 Nucl. Fusion 31 1135), where all PF coils are outside the TF coils. Starting from the standard divertor, a sequence of desirable XD configurations are possible where the PF currents are below the present maximum design limits on ITER, and where the baseline divertor cassette is used. This opens the possibility that the XD could be tested and used to assist in high-power operation on ITER, but some further issues need examination. Note that the increased major radius of the super-X-divertor (Kotschenreuther et al 2007 Bull. Am. Phys. Soc. 53 11, Valanju et al 2009 Phys. Plasmas 16 5, Kotschenreuther et al 2010 Nucl. Fusion 50 035003, Valanju et al 2010 Fusion Eng. Des. 85 46) is not a feature of the XD geometry. In addition, we present an XD configuration for K-DEMO (Kim et al 2013 Fusion Eng. Des. 88 123) to demonstrate that it is also possible to attain the XD configuration in advanced tokamak reactors with all PF coils outside the TF coils. The results given here for the XD are far more encouraging than recent calculations by Lackner and Zohm (2012 Fusion Sci. Technol. 63 43) for the Snowflake (Ryutov 2007 Phys. Plasmas 14 064502, Ryutov et al 2008 Phys. Plasmas 15 092501), where the required high PF currents represent a major technological challenge. The magnetic field structure in the outboard divertor SOL (Kotschenreuther 2013 Phys. Plasmas 20 102507) in the recently created XD configurations reproduces what was presented in the earlier XD papers (Kotschenreuther et al 2004 Proc. 20th Int. Conf. on Fusion Energy (Vilamoura, Portugal, 2004) (Vienna: IAEA) CD-ROM file [IC/P6-43] www-naweb.iaea.org/napc/physics/fec/fec2004/datasets/index.html, Kotschenreuther et al 2006 Proc. 21st Int. Conf. on Fusion Energy 2006 (Chengdu, China, 2006) (Vienna: IAEA) CD-ROM file [IC/P7-12] www-naweb.iaea.org/napc/physics/FEC/FEC2006/html/index.htm, Kotschenreuther et al 2007 Phys. Plasmas 14 072502). Consequently, the same advantages accrue, but no close-in PF coils are employed.

Nonabelian Bundle Gerbes, Their Differential Geometry and Gauge Theory

NASA Astrophysics Data System (ADS)

Aschieri, Paolo; Cantini, Luigi; Jurčo, Branislav

2005-03-01

Bundle gerbes are a higher version of line bundles, we present nonabelian bundle gerbes as a higher version of principal bundles. Connection, curving, curvature and gauge transformations are studied both in a global coordinate independent formalism and in local coordinates. These are the gauge fields needed for the construction of Yang-Mills theories with 2-form gauge potential.

Bundled payment fails to gain a foothold In California: the experience of the IHA bundled payment demonstration.

PubMed

Ridgely, M Susan; de Vries, David; Bozic, Kevin J; Hussey, Peter S

2014-08-01

To determine whether bundled payment could be an effective payment model for California, the Integrated Healthcare Association convened a group of stakeholders (health plans, hospitals, ambulatory surgery centers, physician organizations, and vendors) to develop, through a consensus process, the methods and means of implementing bundled payment. In spite of a high level of enthusiasm and effort, the pilot did not succeed in its goal to implement bundled payment for orthopedic procedures across multiple payers and hospital-physician partners. An evaluation of the pilot documented a number of barriers, such as administrative burden, state regulatory uncertainty, and disagreements about bundle definition and assumption of risk. Ultimately, few contracts were signed, which resulted in insufficient volume to test hypotheses about the impact of bundled payment on quality and costs. Although bundled payment failed to gain a foothold in California, the evaluation provides lessons for future bundled payment initiatives. Project HOPE—The People-to-People Health Foundation, Inc.

Double-bundle ACL reconstruction can improve rotational stability.

PubMed

Yagi, Masayoshi; Kuroda, Ryosuke; Nagamune, Kouki; Yoshiya, Shinichi; Kurosaka, Masahiro

2007-01-01

Double-bundle anterior cruciate ligament (ACL) reconstruction reproduces anteromedial and posterolateral bundles, and thus has theoretical advantages over conventional single-bundle reconstruction in controlling rotational torque in vitro. However, its superiority in clinical practice has not been proven. We analyzed rotational stability with three reconstruction techniques in 60 consecutive patients who were randomly divided into three groups (double-bundle, anteromedial single-bundle, posterolateral single-bundle). In the reconstructive procedure, the hamstring tendon was harvested and used as a free tendon graft. Followup examinations were performed 1 year after surgery. Anteroposterior laxity of the knee was examined with a KT-1000 arthrometer, whereas rotatory instability, as elicited by the pivot shift test, was assessed using a new measurement system incorporating three-dimensional electromagnetic sensors. Routine clinical evaluations, including KT examination, demonstrated no differences among the three groups. However, using the new measurement system, patients with double-bundle ACL reconstruction showed better pivot shift control of complex instability than patients with anteromedial and posterolateral single-bundle reconstruction.

Risk of Revision Was Not Reduced by a Double-bundle ACL Reconstruction Technique: Results From the Scandinavian Registers.

PubMed

Aga, Cathrine; Kartus, Jüri-Tomas; Lind, Martin; Lygre, Stein Håkon Låstad; Granan, Lars-Petter; Engebretsen, Lars

2017-10-01

Double-bundle anterior cruciate ligament (ACL) reconstruction has demonstrated improved biomechanical properties and moderately better objective outcomes compared with single-bundle reconstructions. This could make an impact on the rerupture rate and reduce the risk of revisions in patients undergoing double-bundle ACL reconstruction compared with patients reconstructed with a traditional single-bundle technique. The National Knee Ligament Registers in Scandinavia provide information that can be used to evaluate the revision outcome after ACL reconstructions. The purposes of the study were (1) to compare the risk of revision between double-bundle and single-bundle reconstructions, reconstructed with autologous hamstring tendon grafts; (2) to compare the risk of revision between double-bundle hamstring tendon and single-bundle bone-patellar tendon-bone autografts; and (3) to compare the hazard ratios for the same two research questions after Cox regression analysis was performed. Data collection of primary ACL reconstructions from the National Knee Ligament Registers in Denmark, Norway, and Sweden from July 1, 2005, to December 31, 2014, was retrospectively analyzed. A total of 60,775 patients were included in the study; 994 patients were reconstructed with double-bundle hamstring tendon grafts, 51,991 with single-bundle hamstring tendon grafts, and 7790 with single-bundle bone-patellar tendon-bone grafts. The double-bundle ACL-reconstructed patients were compared with the two other groups. The risk of revision for each research question was detected by the risk ratio, hazard ratio, and the corresponding 95% confidence intervals. Kaplan-Meier analysis was used to estimate survival at 1, 2, and 5 years for the three different groups. Furthermore, a Cox proportional hazard regression model was applied and the hazard ratios were adjusted for country, age, sex, meniscal or chondral injury, and utilized fixation devices on the femoral and tibial sides. There were no differences in the crude risk of revision between the patients undergoing the double-bundle technique and the two other groups. A total of 3.7% patients were revised in the double-bundle group (37 of 994 patients) versus 3.8% in the single-bundle hamstring tendon group (1952 of 51,991; risk ratio, 1.01; 95% confidence interval (CI), 0.73-1.39; p = 0.96), and 2.8% of the patients were revised in the bone-patellar tendon-bone group (219 of the 7790 bone-patellar tendon-bone patients; risk ratio, 0.76; 95% CI, 0.54-1.06; p = 0.11). Cox regression analysis with adjustment for country, age, sex, menisci or cartilage injury, and utilized fixation device on the femoral and tibial sides, did not reveal any further difference in the risk of revision between the single-bundle hamstring tendon and double-bundle hamstring tendon groups (hazard ratio, 1.18; 95% CI, 0.85-1.62; p = 0.33), but the adjusted hazard ratio showed a lower risk of revision in the single-bundle bone-patellar tendon-bone group compared with the double-bundle group (hazard ratio, 0.62; 95% CI, 0.43-0.90; p = 0.01). Comparisons of the graft revision rates reported separately for each country revealed that double-bundle hamstring tendon reconstructions in Sweden had a lower hazard ratio compared with the single-bundle hamstring tendon reconstructions (hazard ratio, 1.00 versus 1.89; 95% CI, 1.09-3.29; p = 0.02). Survival at 5 years after index surgery was 96.0% for the double-bundle group, 95.4% for the single-bundle hamstring tendon group, and 97.0% for the single-bundle bone-patellar tendon-bone group. Based on the data from all three national registers, the risk of revision was not influenced by the reconstruction technique in terms of using single- or double-bundle hamstring tendons, although national differences in survival existed. Using bone-patellar tendon-bone grafts lowered the risk of revision compared with double-bundle hamstring tendon grafts. These findings should be considered when deciding what reconstruction technique to use in ACL-deficient knees. Future studies identifying the reasons for graft rerupture in single- and double-bundle reconstructions would be of interest to understand the findings of the present study. Level III, therapeutic study.

Heat Transfer Analysis in Wire Bundles for Aerospace Vehicles

NASA Technical Reports Server (NTRS)

Rickman, S. L.; Iamello, C. J.

2016-01-01

Design of wiring for aerospace vehicles relies on an understanding of "ampacity" which refers to the current carrying capacity of wires, either, individually or in wire bundles. Designers rely on standards to derate allowable current flow to prevent exceedance of wire temperature limits due to resistive heat dissipation within the wires or wire bundles. These standards often add considerable margin and are based on empirical data. Commercial providers are taking an aggressive approach to wire sizing which challenges the conventional wisdom of the established standards. Thermal modelling of wire bundles may offer significant mass reduction in a system if the technique can be generalized to produce reliable temperature predictions for arbitrary bundle configurations. Thermal analysis has been applied to the problem of wire bundles wherein any or all of the wires within the bundle may carry current. Wire bundles present analytical challenges because the heat transfer path from conductors internal to the bundle is tortuous, relying on internal radiation and thermal interface conductance to move the heat from within the bundle to the external jacket where it can be carried away by convective and radiative heat transfer. The problem is further complicated by the dependence of wire electrical resistivity on temperature. Reduced heat transfer out of the bundle leads to higher conductor temperatures and, hence, increased resistive heat dissipation. Development of a generalized wire bundle thermal model is presented and compared with test data. The steady state heat balance for a single wire is derived and extended to the bundle configuration. The generalized model includes the effects of temperature varying resistance, internal radiation and thermal interface conductance, external radiation and temperature varying convective relief from the free surface. The sensitivity of the response to uncertainties in key model parameters is explored using Monte Carlo analysis.

78 FR 76987 - Adjustment of Determination of Compulsory License Rates for Mechanical and Digital Phonorecords

Federal Register 2010, 2011, 2012, 2013, 2014

2013-12-20

... service bundles, music bundles, paid locker services, and purchased content locker services. The technical... downloads, limited offerings, mixed service bundles, music bundles, paid locker services, and purchased...

Interaction of plasmas with lithium and tungsten fusion plasma facing components

NASA Astrophysics Data System (ADS)

Fiflis, Peter Robert

One of the largest outstanding issues in magnetic confinement fusion is the interaction of the fusion plasma with the first wall of the device; an interaction which is strongest in the divertor region. Erosion, melting, sputtering, and deformation are all concerns which inform choices of divertor material. Of the many materials proposed for use in the divertor, only a few remain as promising choices. Tungsten has been chosen as the material for the ITER divertor, and liquid lithium stands poised as its replacement in higher heat flux devices. As a refractory metal, tungsten's large melting point and thermal conductivity as well as its low sputtering yield have led to its selection as the material of choice of the ITER divertor. Experiments have reinforced this choice demonstrating tungsten's ability to withstand large heat fluxes when adequately cooled. However, tungsten has shown a propensity to nanostructure under exposure within a certain temperature range to large fluxes of helium ions. These nanostructures if disrupted into the plasma as dust by an off-normal event would cause quenching of the plasma from the generated dust. Liquid lithium, meanwhile, has gathered growing interest within the fusion community in recent years as a divertor, limiter, and alternative first wall material. Liquid lithium is attractive as a low-Z material replacement for refractory metals due to its ability to getter impurities, while also being self-healing in nature. However, concerns exist about the stability of a liquid metal surface at the edge of a fusion device. Liquid metal pools, such as the Li-DiMes probe, have shown evidence of macroscopic lithium displacement as well as droplet formation and ejection into the plasma. These issues must be mitigated in future implementations of liquid lithium divertor concepts. Rayleigh-Taylor-like (RT) and Kelvin-Helmholtz-like (KH) instabilities have been claimed as the initiators of droplet ejection, yet not enough data exists to delineate a stability boundary. The influences of plasma pressure and current driven instabilities on lithium surfaces that lead to droplet ejection are investigated to determine which of the two effects is dominant for a given set of plasma conditions. This work studies the influence of large plasma fluxes on these two materials to better inform the selection and design of plasma facing components (PFCs). The nanostructuring of tungsten was investigated to determine the mechanisms by which tungsten nanostructures so that its formation may be mitigated. Experiments investigated the dependence of nanostructuring on temperature, looked at the morphological evolution, and grew nanostructures on a variety of metals to examine their similarity to tungsten. Additionally, a computational model is presented for the initial stages of fuzz formation showing good quantitative and qualitative agreement with experimental observations. The influences of RT and KH instabilities on the surface of liquid lithium were experimentally observed and quantified on the ThermoElectric-driven Liquid-metal plasma-facing Structures (TELS) chamber at the University of Illinois at Urbana-Champaign and the stabilizing effect of surface tension, an effect employed by the LiMIT concept as well as other liquid lithium concepts, was studied, and the stability boundary afforded by surface tension was compared between experiment, computational simulation, and theory.

Structures for handling high heat fluxes

NASA Astrophysics Data System (ADS)

Watson, R. D.

1990-12-01

The divertor is reconized as one of the main performance limiting components for ITER. This paper reviews the critical issues for structures that are designed to withstand heat fluxes > 5 MW/m 2. High velocity, sub-cooled water with twisted tape inserts for enhanced heat transfer provides a critical heat flux limit of 40-60 MW/m 2. Uncertainties in physics and engineering heat flux peaking factors require that the design heat flux not exceed 10 MW/m 2 to maintain an adequate burnout safety margin. Armor tiles and heat sink materials must have a well matched thermal expansion coefficient to minimize stresses. The divertor lifetime from sputtering erosion is highly uncertain. The number of disruptions specified for ITER must be reduced to achieve a credible design. In-situ plasma spray repair with thick metallic coatings may reduce the problems of erosion. Runaway electrons in ITER have the potential to melt actively cooled components in a single event. A water leak is a serious accident because of steam reactions with hot carbon, beryllium, or tungsten that can mobilize large amounts of tritium and radioactive elements. If the plasma does not shutdown immediately, the divertor can melt in 1-10 s after a loss of coolant accident. Very high reliability of carbon tile braze joints will be required to achieve adequate safety and performance goals. Most of these critical issues will be addressed in the near future by operation of the Tore Supra pump limiters and the JET pumped divertor. An accurate understanding of the power flow out of edge of a DT burning plasma is essential to successful design of high heat flux components.

Creating Hybrid Plasmas With Off-Axis ECCD for Radiating Divertor Studies in DIII-D

NASA Astrophysics Data System (ADS)

Petty, C. C.; Ferron, J. R.; Luce, T. C.; Osborne, T. H.; Petrie, T. W.; Turco, F.; Holcomb, C. T.; Thome, K. E.

2017-10-01

A long duration, high density, high power hybrid scenario has been developed for use in radiative divertor studies in DIII-D. Using 11.2 MW of co-NBI power and 3.4 MW of ECCD, with a total injected energy of up to 56 MJ, high performance hybrid plasmas with βN = 3.7 and H98y2 = 1.5 were created. The hybrid plasmas were fully non-inductive at densities of n 4.2 ×1019 m-3 with central ECCD, but the EC deposition needed to be moved to ρ = 0.45 to avoid the right-hand cutoff when the density was raised to n 5.8 ×1019 m-3 for radiative divertor studies. Although moving the EC deposition to ρ = 0.45 had the effect of dropping τE by 10%, the energy confinement time increased with higher density like τE n0.4, allowing high beta to be maintained. While the plasma current profile displays the usual self-organizing properties of hybrids - an anomalously broad profile with qmin >1 - local current drive can still have a measurable effect on stability, either positively or negatively. For example, hybrid discharges with radial ECH deposited at ρ = 0.45 proved to be more robustly stable to n = 1 modes (can be either a 1/1 or 2/1 mode) than similar discharges with co-ECCD at the same location. Interestingly, the large 1/1 mode had almost no effect on energy confinement but strongly degraded particle confinement; thus this mode needed to be suppressed to achieve the high pedestal densities required for radiative divertor studies. Work supported by USDOE under DE-FC02-04ER54698.

Tungsten: an option for divertor and main chamber plasma facing components in future fusion devices

NASA Astrophysics Data System (ADS)

Neu, R.; Dux, R.; Kallenbach, A.; Pütterich, T.; Balden, M.; Fuchs, J. C.; Herrmann, A.; Maggi, C. F.; O'Mullane, M.; Pugno, R.; Radivojevic, I.; Rohde, V.; Sips, A. C. C.; Suttrop, W.; Whiteford, A.; ASDEX Upgrade Team

2005-03-01

The tungsten programme in ASDEX Upgrade is pursued towards a full high-Z device. The spectroscopic diagnostic of W has been extended and refined and the cooling factor of W has been re-evaluated. The W coated surfaces now represent a fraction of 65% of all plasma facing components (24.8 m2). The only two major components that are not yet coated are the strikepoint region of the lower divertor as well as the limiters at the low field side. While extending the W surfaces, the W concentration and the discharge behaviour have changed gradually pointing to critical issues when operating with a W wall: anomalous transport in the plasma centre should not be too low, otherwise neoclassical accumulation can occur. One very successful remedy is the addition of central RF heating at the 20-30% level. Regimes with low ELM activity show increased impurity concentration over the whole plasma radius. These discharges can be cured by increasing the ELM frequency through pellet ELM pacemaking or by higher heating power. Moderate gas puffing also mitigates the impurity influx and penetration, however, at the expense of lower confinement. The erosion yield at the low field side guard limiter can be as high as 10-3 and fast particle losses from NBI were identified to contribute a significant part to the W sputtering. Discharges run in the upper W coated divertor do not show higher W concentrations than comparable discharges in the lower C based divertor. According to impurity transport calculations no strong high-Z accumulation is expected for the ITER standard scenario as long as the anomalous transport is at least as high as the neoclassical one.

Long-term evolution of the impurity composition and impurity events with the ITER-like wall at JET

NASA Astrophysics Data System (ADS)

Coenen, J. W.; Sertoli, M.; Brezinsek, S.; Coffey, I.; Dux, R.; Giroud, C.; Groth, M.; Huber, A.; Ivanova, D.; Krieger, K.; Lawson, K.; Marsen, S.; Meigs, A.; Neu, R.; Puetterich, T.; van Rooij, G. J.; Stamp, M. F.; Contributors, JET-EFDA

2013-07-01

This paper covers aspects of long-term evolution of intrinsic impurities in the JET tokamak with respect to the newly installed ITER-like wall (ILW). At first the changes related to the change over from the JET-C to the JET-ILW with beryllium (Be) as the main wall material and tungsten (W) in the divertor are discussed. The evolution of impurity fluxes in the newly installed W divertor with respect to studying material migration is described. In addition, a statistical analysis of transient impurity events causing significant plasma contamination and radiation losses is shown. The main findings comprise a drop in carbon content (×20) (see also Brezinsek et al (2013 J. Nucl. Mater. 438 S303)), low oxygen content (×10) due to the Be first wall (Douai et al 2013 J. Nucl. Mater. 438 S1172-6) as well as the evolution of the material mix in the divertor. Initially, a short period of repetitive ohmic plasmas was carried out to study material migration (Krieger et al 2013 J. Nucl. Mater. 438 S262). After the initial 1600 plasma seconds the material surface composition is, however, still evolving. With operational time, the levels of recycled C are increasing slightly by 20% while the Be levels in the deposition-dominated inner divertor are dropping, hinting at changes in the surface layer material mix made of Be, C and W. A steady number of transient impurity events, consisting of W and constituents of inconel, is observed despite the increase in variation in machine operation and changes in magnetic configuration as well as the auxiliary power increase.

Analysis of a Multi-Machine Database on Divertor Heat Fluxes

NASA Astrophysics Data System (ADS)

Makowski, M. A.

2011-10-01

A coordinated effort to measure divertor heat flux characteristics in fully attached, similarly shaped H-mode plasmas on C-Mod, DIII-D and NSTX was carried out in 2010 in order to construct a predictive scaling relation applicable to next step devices including ITER, FNSF, and DEMO. Few published scaling laws are available and those that have been published were obtained under widely varying conditions and divertor geometries, leading to conflicting predictions for this critically important quantity. This study was designed to overcome these deficiencies. Corresponding plasma parameters were systematically varied in each tokamak, resulting in a combined data set in which Ip varies by a factor 3, Bt varies by a factor of 14.5, and major radius varies by a factor of 2.6. The derived scaling relation consistently predicts narrower heat flux widths than relations currently in use. Analysis of the combined data set reveals that the primary dependence of the parallel heat flux width is robustly inverse with Ip. All three tokamaks independently demonstrate this dependence. The midplane SOL profiles in DIII-D are also found to steepen with higher Ip, similar to the divertor heat flux profiles. Weaker dependencies on the toroidal field and normalized Greenwald density, fGW, are also found, but vary across devices and with the measure of the heat flux width used, either FWHM or integral width. In the combined data set, the strongest size scaling is with minor radius resulting in an approximately linear dependence on a /Ip . This suggests a scaling correlated with the inverse of the poloidal field, as would be expected for critical gradient or drift-based transport. Supported by the US DOE under DE-AC52-07NA27344 and DE-FC02-04ER54698.

Access to edge scenarios for testing a scraper element in early operation phases of Wendelstein 7-X

DOE PAGES

Holbe, H.; Pedersen, T. Sunn; Geiger, J.; ...

2016-01-29

The edge topology of magnetic fusion devices is decisive for the control of the plasma exhaust. In Wendelstein 7-X, the island divertor concept will be used, for which the edge topology can change significantly as the internal currents in a plasma discharge evolve towards steady-state. Consequently, the device has been optimized to minimize such internal currents, in particular the bootstrap current [1]. Nonetheless, there are predicted pulse scenarios where effects of the remaining internal currents could potentially lead to overload of plasma-facing components. These internal currents are predicted to evolve on long time scales (tens of seconds) so their effectsmore » on the edge topology and the divertor heat loads may not be experimentally accessible in the first years of W7-X operation, where only relatively short pulses are possible. However, we show here that for at least one important long-pulse divertor operation issue, relevant physics experiments can be performed already in short-pulse operation, through judicious adjustment of the edge topology by the use of the existing coil sets. The specific issue studied here is a potential overload of the divertor element edges. This overload might be mitigated by the installation of an extra set of plasma-facing components, so-called scraper elements, as suggested in earlier publications. It is shown here that by a targeted control of edge topology, the effectiveness of such scraper elements can be tested already with uncooled test-scraper elements in short-pulse operation. Furthermore, this will allow an early and well-informed decision on whether long-pulse-capable (actively cooled) scraper elements should be built and installed.« less

Activation, decay heat, and waste classification studies of the European DEMO concept

NASA Astrophysics Data System (ADS)

Gilbert, M. R.; Eade, T.; Bachmann, C.; Fischer, U.; Taylor, N. P.

2017-04-01

Inventory calculations have a key role to play in designing future fusion power plants because, for a given irradiation field and material, they can predict the time evolution in chemical composition, activation, decay heat, gamma-dose, gas production, and even damage (dpa) dose. For conceptual designs of the European DEMO fusion reactor such calculations provide information about the neutron shielding requirements, maintenance schedules, and waste disposal prospects; thereby guiding future development. Extensive neutron-transport and inventory calculations have been performed for a reference DEMO reactor model with four different tritium-breeding blanket concepts. The results have been used to chart the post-operation variation in activity and decay heat from different vessel components, demonstrating that the shielding performance of the different blanket concepts—for a given blanket thickness—varies significantly. Detailed analyses of the simulated nuclide inventories for the vacuum vessel (VV) and divertor highlight the most dominant radionuclides, potentially suggesting how changes in material composition could help to reduce activity. Minor impurities in the raw composition of W used in divertor tiles, for example, are shown to produce undesirable long-lived radionuclides. Finally, waste classifications, based on UK regulations, and a recycling potential limit, have been applied to estimate the time-evolution in waste masses for both the entire vessel (including blanket modules, VV, divertor, and some ex-vessel components) and individual components, and also to suggest when a particular component might be suitable for recycling. The results indicate that the large mass of the VV will not be classifiable as low level waste on the 100 year timescale, but the majority of the divertor will be, and that both components will be potentially recyclable within that time.

Time-resolved deposition in the remote region of the JET-ILW divertor: measurements and modelling

NASA Astrophysics Data System (ADS)

Catarino, N.; Widdowson, A.; Baron-Wiechec, A.; Coad, J. P.; Heinola, K.; Rubel, M.; Alves, E.; Contributors, JET

2017-12-01

One crucial requirement for the development of fusion power is to know where, and how much, impurities collect in the machine, and how much of the fuelling isotope tritium will be trapped therein. The most relevant information on this issue comes from the operation of the Joint European Tokamak (JET), which is the world’s largest operating tokamak and has the same interior plasma-facing materials as the next step machine, ITER. Much of the information gained so far has been from post-mortem analysis of samples collected after whole campaigns involving varied types of operation. This paper describes time-resolved measurements of the deposition rate using rotating collectors (RC) placed in remote areas of the JET divertor during the 2013-2014 campaign with the ITER-like Wall (ILW). These techniques allow the effects of different types of operation to be distinguished. Rotating collectors made of silicon discs housed behind an aperture are exposed to the plasma. Each time the magnetic field coils are ramped up for a discharge the disc rotates, providing a linear relationship between the exposed region and the discharge number. Post-mortem ion beam analyses provide information on the deposit composition as a function of the discharge number. The results show that the Be deposition average for the RC in the corners of the inner and outer divertor are 4.9 × 1016 cm-2 and 1.8 × 1017 cm-2, respectively, accumulated over an average of ˜25 pulses. Data from the rotating collector below Tile 5 in the central region of divertor indicate a Be deposition rate of 9.3 × 1015 cm-2, per ˜25 pulses.

Cost-Effectiveness of a Central Venous Catheter Care Bundle

PubMed Central

Halton, Kate A.; Cook, David; Paterson, David L.; Safdar, Nasia; Graves, Nicholas

2010-01-01

Background A bundled approach to central venous catheter care is currently being promoted as an effective way of preventing catheter-related bloodstream infection (CR-BSI). Consumables used in the bundled approach are relatively inexpensive which may lead to the conclusion that the bundle is cost-effective. However, this fails to consider the nontrivial costs of the monitoring and education activities required to implement the bundle, or that alternative strategies are available to prevent CR-BSI. We evaluated the cost-effectiveness of a bundle to prevent CR-BSI in Australian intensive care patients. Methods and Findings A Markov decision model was used to evaluate the cost-effectiveness of the bundle relative to remaining with current practice (a non-bundled approach to catheter care and uncoated catheters), or use of antimicrobial catheters. We assumed the bundle reduced relative risk of CR-BSI to 0.34. Given uncertainty about the cost of the bundle, threshold analyses were used to determine the maximum cost at which the bundle remained cost-effective relative to the other approaches to infection control. Sensitivity analyses explored how this threshold alters under different assumptions about the economic value placed on bed-days and health benefits gained by preventing infection. If clinicians are prepared to use antimicrobial catheters, the bundle is cost-effective if national 18-month implementation costs are below $1.1 million. If antimicrobial catheters are not an option the bundle must cost less than $4.3 million. If decision makers are only interested in obtaining cash-savings for the unit, and place no economic value on either the bed-days or the health benefits gained through preventing infection, these cost thresholds are reduced by two-thirds. Conclusions A catheter care bundle has the potential to be cost-effective in the Australian intensive care setting. Rather than anticipating cash-savings from this intervention, decision makers must be prepared to invest resources in infection control to see efficiency improvements. PMID:20862246

Cost-effectiveness of a central venous catheter care bundle.

PubMed

Halton, Kate A; Cook, David; Paterson, David L; Safdar, Nasia; Graves, Nicholas

2010-09-17

A bundled approach to central venous catheter care is currently being promoted as an effective way of preventing catheter-related bloodstream infection (CR-BSI). Consumables used in the bundled approach are relatively inexpensive which may lead to the conclusion that the bundle is cost-effective. However, this fails to consider the nontrivial costs of the monitoring and education activities required to implement the bundle, or that alternative strategies are available to prevent CR-BSI. We evaluated the cost-effectiveness of a bundle to prevent CR-BSI in Australian intensive care patients. A Markov decision model was used to evaluate the cost-effectiveness of the bundle relative to remaining with current practice (a non-bundled approach to catheter care and uncoated catheters), or use of antimicrobial catheters. We assumed the bundle reduced relative risk of CR-BSI to 0.34. Given uncertainty about the cost of the bundle, threshold analyses were used to determine the maximum cost at which the bundle remained cost-effective relative to the other approaches to infection control. Sensitivity analyses explored how this threshold alters under different assumptions about the economic value placed on bed-days and health benefits gained by preventing infection. If clinicians are prepared to use antimicrobial catheters, the bundle is cost-effective if national 18-month implementation costs are below $1.1 million. If antimicrobial catheters are not an option the bundle must cost less than $4.3 million. If decision makers are only interested in obtaining cash-savings for the unit, and place no economic value on either the bed-days or the health benefits gained through preventing infection, these cost thresholds are reduced by two-thirds. A catheter care bundle has the potential to be cost-effective in the Australian intensive care setting. Rather than anticipating cash-savings from this intervention, decision makers must be prepared to invest resources in infection control to see efficiency improvements.

Adhering to a national surgical care bundle reduces the risk of surgical site infections

PubMed Central

Hopmans, Titia E. M.; Soetens, Loes C.; Wille, Jan C.; Geerlings, Suzanne E.; Vos, Margreet C.; van Benthem, Birgit H. B.; de Greeff, Sabine C.

2017-01-01

Background In 2008, a bundle of care to prevent Surgical Site Infections (SSIs) was introduced in the Netherlands. The bundle consisted of four elements: antibiotic prophylaxis according to local guidelines, no hair removal, normothermia and ‘hygiene discipline’ in the operating room (i.e. number of door movements). Dutch hospitals were advised to implement the bundle and to measure the outcome. This study’s goal was to assess how effective the bundle was in reducing SSI risk. Methods Hospitals assessed whether their staff complied with each of the bundle elements and voluntary reported compliance data to the national SSI surveillance network (PREZIES). From PREZIES data, we selected data from 2009 to 2014 relating to 13 types of surgical procedures. We excluded surgeries with missing (non)compliance data, and calculated for each remaining surgery with reported (non)compliance data the level of compliance with the bundle (that is, being compliant with 0, 1, 2, 3, or 4 of the elements). Subsequently, we used this level of compliance to assess the effect of bundle compliance on the SSI risk, using multilevel logistic regression techniques. Results 217 489 surgeries were included, of which 62 486 surgeries (29%) had complete bundle reporting. Within this group, the SSI risk was significantly lower for surgeries with complete bundle compliance compared to surgeries with lower compliance levels. Odds ratios ranged from 0.63 to 0.86 (risk reduction of 14% to 37%), while a 13% risk reduction was demonstrated for each point increase in compliance-level. Sensitivity analysis indicated that due to analysing reported bundles only, we probably underestimated the total effect of implementing the bundle. Conclusions This study demonstrated that adhering to a surgical care bundle significantly reduced the risk of SSIs. Reporting of and compliance with the bundle compliance can, however, still be improved. Therefore an even greater effect might be achieved. PMID:28877223

Ab initio density functional theory investigation of electronic properties of semiconducting single-walled carbon nanotube bundles

NASA Astrophysics Data System (ADS)

Moradian, Rostam; Behzad, Somayeh; Azadi, Sam

2008-09-01

By using ab initio density functional theory we investigated the structural and electronic properties of semiconducting (7, 0), (8, 0) and (10, 0) carbon nanotube bundles. The energetic and electronic evolutions of nanotubes in the bundling process are also studied. The effects of inter-tube coupling on the electronic dispersions of semiconducting carbon nanotube bundles are demonstrated. Our results show that the inter-tube coupling decreases the energy gap in semiconducting nanotubes. We found that bundles of (7, 0) and (8, 0) carbon nanotubes have metallic feature, while (10, 0) bundle is a semiconductor with an energy gap of 0.22 eV. To clarify our results the band structures of isolated and bundled nanotubes are compared.

Lexical bundles in an advanced INTOCSU writing class and engineering texts: A functional analysis

NASA Astrophysics Data System (ADS)

Alquraishi, Mohammed Abdulrahman

The purpose of this study is to investigate the functions of lexical bundles in two corpora: a corpus of engineering academic texts and a corpus of IEP advanced writing class texts. This study is concerned with the nature of formulaic language in Pathway IEPs and engineering texts, and whether those types of texts show similar or distinctive formulaic functions. Moreover, the study looked into lexical bundles found in an engineering 1.26 million-word corpus and an ESL 65000-word corpus using a concordancing program. The study then analyzed the functions of those lexical bundles and compared them statistically using chi-square tests. Additionally, the results of this investigation showed 236 unique frequent lexical bundles in the engineering corpus and 37 bundles in the pathway corpus. Also, the study identified several differences between the density and functions of lexical bundles in the two corpora. These differences were evident in the distribution of functions of lexical bundles and the minimal overlap of lexical bundles found in the two corpora. The results of this study call for more attention to formulaic language at ESP and EAP programs.

Biological natural retting for determining the hierarchical structuration of banana fibers.

PubMed

Gañán, Piedad; Zuluaga, Robin; Velez, Juan Manuel; Mondragon, Iñaki

2004-10-20

Extraction processes of natural fibers can be performed by different procedures that include mechanical, chemical and biological methods. Each method presents different advantages or drawbacks according to the amount of fiber produced or the quality and properties of fiber bundles obtained. In this study, biological natural retting was satisfactorily used for obtaining banana fibers from plant bunches. However, the most important contribution of this work refers to the description of the hierarchical microstructural ordering present in banana fiber bundles in both bundle surface and inner region. The chemical composition of banana fiber bundles has been evaluated by FTIR spectroscopy. Through exposure time, the fiber bundle configuration presents small variations in composition. The main changes are related to hemicellulose and pectins as they conform the outer walls of the bundle. Hierarchical helicoidal ordering in the bundle surface as well as orientation on the longitudinal axis of the bundle were observed by optical microscopy (OM) and scanning electron microscopy (SEM) for 3-4 microm surface fibers and 10-15 microm inner elementary fibers, respectively. With increasing exposure time, fiber bundle walls lose integrity, as reflected in their mechanical behavior.

Modeling of limiter heat loads and impurity transport in Wendelstein 7-X startup plasmas

NASA Astrophysics Data System (ADS)

Effenberg, Florian; Feng, Y.; Frerichs, H.; Schmitz, O.; Hoelbe, H.; Koenig, R.; Krychowiak, M.; Pedersen, T. S.; Bozhenkov, S.; Reiter, D.

2015-11-01

The quasi-isodynamic stellarator Wendelstein 7-X starts plasma operation in a limiter configuration. The field consists of closed magnetic flux surfaces avoiding magnetic islands in the plasma boundary. Because of the small size of the limiters and the absence of wall-protecting elements in this phase, limiter heat loads and impurity generation due to plasma surface interaction become a concern. These issues are studied with the 3D fluid plasma edge and kinetic neutral transport code EMC3-Eirene. It is shown that the 3D SOL consists of three separate helical magnetic flux bundles of different field line connection lengths. A density scan at input power of 4MW reveals a strong modulation of the plasma paramters with the connection length. The limiter peak heat fluxes drop from 14 MWm-2 down to 10 MWm-2 with raising the density from 1 ×1018m-3 to 1.9 ×1019m-3, accompanied by an increase of the heat flux channel widths λq. Radiative power losses can help to avoid thermal overloads of the limiters at the upper margin of the heating power. The power removal feasibility of the intrinsic carbon and other extrinsic light impurities via active gas injection is discussed as a preparation of this method for island divertor operation. Work supported in part by start up funds of the Department of Engineering Physics at the University of Wisconsin - Madison, USA and by the U.S. Department of Energy under grant DE-SC0013911.

Stimulation of hair cells with ultraviolet light

NASA Astrophysics Data System (ADS)

Azimzadeh, Julien B.; Fabella, Brian A.; Hudspeth, A. J.

2018-05-01

Hair bundles are specialized organelles that transduce mechanical inputs into electrical outputs. To activate hair cells, physiologists have resorted to mechanical methods of hair-bundle stimulation. Here we describe a new method of hair-bundle stimulation, irradiation with ultraviolet light. A hair bundle illuminated by ultraviolet light rapidly moves towards its tall edge, a motion typically associated with excitatory stimulation. The motion disappears upon tip-link rupture and is associated with the opening of mechanotransduction channels. Hair bundles can be induced to move sinusoidally with oscillatory modulation of the stimulation power. We discuss the implications of ultraviolet stimulation as a novel hair-bundle stimulus.

Formation and structural organization of the egg-sperm bundle of the scleractinian coral Montipora capitata

NASA Astrophysics Data System (ADS)

Padilla-Gamiño, J. L.; Weatherby, T. M.; Waller, R. G.; Gates, R. D.

2011-06-01

The majority of scleractinian corals are hermaphrodites that broadcast spawn their gametes separately or packaged as egg-sperm bundles during spawning events that are timed to the lunar cycle. The egg-sperm bundle is an efficient way of transporting gametes to the ocean surface where fertilization takes place, while minimizing sperm dilution and maximizing the opportunity for gamete encounters during a spawning event. To date, there are few studies that focus on the formation and structure of egg-sperm bundle. This study explores formation, ultrastructure, and longevity of the egg-sperm bundle in Montipora capitata, a major reef building coral in Hawai`i. Our results show that the egg-sperm bundle is formed by a mucus layer secreted by the oocytes. The sperm package is located at the center of each bundle, possibly reflecting the development of male and female gametes in different mesenteries. Once the egg-sperm bundle has reached the ocean surface, it breaks open within 10-35 min, depending on the environmental conditions (i.e., wind, water turbulence). Although the bundle has an ephemeral life span, the formation of an egg-sperm bundle is a fundamental part of the reproductive process that could be strongly influenced by climate change and deterioration of water quality (due to anthropogenic effects) and thus requires further investigation.

Robust Mapping of Incoherent Fiber-Optic Bundles

NASA Technical Reports Server (NTRS)

Roberts, Harry E.; Deason, Brent E.; DePlachett, Charles P.; Pilgrim, Robert A.; Sanford, Harold S.

2007-01-01

A method and apparatus for mapping between the positions of fibers at opposite ends of incoherent fiber-optic bundles have been invented to enable the use of such bundles to transmit images in visible or infrared light. The method is robust in the sense that it provides useful mapping even for a bundle that contains thousands of narrow, irregularly packed fibers, some of which may be defective. In a coherent fiber-optic bundle, the input and output ends of each fiber lie at identical positions in the input and output planes; therefore, the bundle can be used to transmit images without further modification. Unfortunately, the fabrication of coherent fiber-optic bundles is too labor-intensive and expensive for many applications. An incoherent fiber-optic bundle can be fabricated more easily and at lower cost, but it produces a scrambled image because the position of the end of each fiber in the input plane is generally different from the end of the same fiber in the output plane. However, the image transmitted by an incoherent fiber-optic bundle can be unscrambled (or, from a different perspective, decoded) by digital processing of the output image if the mapping between the input and output fiber-end positions is known. Thus, the present invention enables the use of relatively inexpensive fiber-optic bundles to transmit images.

A compressed sensing approach for resolution improvement in fiber-bundle based endomicroscopy

NASA Astrophysics Data System (ADS)

Dumas, John P.; Lodhi, Muhammad A.; Bajwa, Waheed U.; Pierce, Mark C.

2018-02-01

Endomicroscopy techniques such as confocal, multi-photon, and wide-field imaging have all been demonstrated using coherent fiber-optic imaging bundles. While the narrow diameter and flexibility of fiber bundles is clinically advantageous, the number of resolvable points in an image is conventionally limited to the number of individual fibers within the bundle. We are introducing concepts from the compressed sensing (CS) field to fiber bundle based endomicroscopy, to allow images to be recovered with more resolvable points than fibers in the bundle. The distal face of the fiber bundle is treated as a low-resolution sensor with circular pixels (fibers) arranged in a hexagonal lattice. A spatial light modulator is located conjugate to the object and distal face, applying multiple high resolution masks to the intermediate image prior to propagation through the bundle. We acquire images of the proximal end of the bundle for each (known) mask pattern and then apply CS inversion algorithms to recover a single high-resolution image. We first developed a theoretical forward model describing image formation through the mask and fiber bundle. We then imaged objects through a rigid fiber bundle and demonstrate that our CS endomicroscopy architecture can recover intra-fiber details while filling inter-fiber regions with interpolation. Finally, we examine the relationship between reconstruction quality and the ratio of the number of mask elements to the number of fiber cores, finding that images could be generated with approximately 28,900 resolvable points for a 1,000 fiber region in our platform.

Understanding nurses' views on a pressure ulcer prevention care bundle: a first step towards successful implementation.

PubMed

Chaboyer, Wendy; Gillespie, Brigid M

2014-12-01

To explore nurses' views of the barriers and facilitators to the use of a newly devised patient-centred pressure ulcer prevention care bundle. Given pressure ulcer prevention strategies are not implemented consistently, the use of a pressure ulcer care bundle may improve implementation given bundles generally assist in standardising care. A quality improvement project was undertaken after a pressure ulcer prevention care bundle was developed and pilot-tested. Short, conversational interviews with nurse explored their views of a patient-centred pressure ulcer care bundle. Interviews were audio-taped and transcribed. Inductive content analysis was used to analyse the transcripts. A total of 20 nurses were interviewed. Five categories with corresponding subcategories emerged from the analysis. They were increasing awareness of pressure ulcer prevention, prompting pressure ulcer prevention activities, promoting active patient participation, barriers to using a pressure ulcer prevention care bundle and enabling integration of the pressure ulcer prevention care bundle into routine practice. Benefits of using a patient-centred pressure ulcer prevention care bundle may include prompting patients and staff to implement prevention strategies and promote active patient participation in care. The success of the care bundle relied on both patients' willingness to participate and nurses' willingness to incorporate it into their routine work. A patient-centred pressure ulcer prevention care bundle may facilitate more consistent implementation of pressure ulcer prevention strategies and active patient participation in care. © 2014 John Wiley & Sons Ltd.

Concept development for the ITER equatorial port visible∕infrared wide angle viewing system.

PubMed

Reichle, R; Beaumont, B; Boilson, D; Bouhamou, R; Direz, M-F; Encheva, A; Henderson, M; Huxford, R; Kazarian, F; Lamalle, Ph; Lisgo, S; Mitteau, R; Patel, K M; Pitcher, C S; Pitts, R A; Prakash, A; Raffray, R; Schunke, B; Snipes, J; Diaz, A Suarez; Udintsev, V S; Walker, C; Walsh, M

2012-10-01

The ITER equatorial port visible∕infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R&D topics are outlined.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Umansky, M. V.; Ryutov, D. D.

Reduced MHD equations are used for studying toroidally symmetric plasma dynamics near the divertor null point. Numerical solution of these equations exhibits a plasma vortex localized at the null point with the time-evolution defined by interplay of the curvature drive, magnetic restoring force, and dissipation. Convective motion is easier to achieve for a second-order null (snowflake) divertor than for a regular x-point configuration, and the size of the convection zone in a snowflake configuration grows with plasma pressure at the null point. In conclusion, the trends in simulations are consistent with tokamak experiments which indicate the presence of enhanced transportmore » at the null point.« less

Crossed-field divertor for a plasma device

DOEpatents

Kerst, Donald W.; Strait, Edward J.

1981-01-01

A divertor for removal of unwanted materials from the interior of a magnetic plasma confinement device includes the division of the wall of the device into segments insulated from each other in order to apply an electric field having a component perpendicular to the confining magnetic field. The resulting crossed-field drift causes electrically charged particles to be removed from the outer part of the confinement chamber to a pumping chamber. This method moves the particles quickly past the saddle point in the poloidal magnetic field where they would otherwise tend to stall, and provides external control over the rate of removal by controlling the magnitude of the electric field.

Development and applications of 3D-DIVIMP(HC) Monte Carlo impurity modeling code

NASA Astrophysics Data System (ADS)

Mu, Yarong

A self-contained gas injection system for the Divertor Material Evaluation System (DiMES) on DIII-D, the Porous Plug Injector (PPI), has been employed by A. McLean for in-situ study of chemical erosion in the tokamak divertor environment by injection of CH4. The principal contribution of the present thesis is a new interpretive code, 3D-DIVIMP(HC), which has been developed and successfully applied to the interpretation of the CH, C I, and C II emissions measured during the PPI experiments. The two principal types of experimental data which are compared here with 3D-DIVIMP(HC) code modeling are (a) absolute emissivities measured with a high resolution spectrometer, and (b) 2D filtered camera (TV) pictures taken from a view essentially straight down on the PPI. Incorporating the Janev-Reiter database for the breakup reactions of methane molecules in a plasma, 3D-DIVIMP(HC) is able to replicate these measurements to within the combined experimental and database uncertainties. It is therefore concluded that the basic elements of the physics and chemistry controlling the breakup of methane entering an attached divertor plasma have been identified and are incorporated in 3D-DIVIMP(HC).

Spectroscopic measurements and modeling of tungsten erosion in the DIII-D divertor

NASA Astrophysics Data System (ADS)

Abrams, T. D.; Ding, R.; Guo, H. Y.; Leonard, A. W.; Thomas, D. M.; Allen, S. L.; McLean, A. G.; Briesemeister, A. R.; Unterberg, E. A.; Chrobak, C.; Doerner, R. P.; Rudakov, D. L.; Elder, J. D.; Stangeby, P. C.; Wampler, W. R.; Watkins, J. G.

2015-11-01

In situ time-resolved measurements of the gross W erosion rate have been performed in DIII-D by monitoring W/I (400.9 nm) emission in the divertor via a filtered camera and high-resolution spectrometer. The erosion rate of a thin W coating on DiMES, inferred via the S/XB method, was found to be ~ 0.7 nm/s during deuterim L-mode exposure, in fair agreement with post-mortem IBA analysis but lower than REDEP/WBC modeling. During H-mode He bombardment of W disks, average erosion rates of ~ 2.9 nm/s and ~ 9.0 nm/s were estimated during the inter-ELM and intra-ELM phases, using ne and Te from divertor Thomson scattering and Langmuir probes. Results will also be presented from additional W erosion experiments in preparation for the DIII-D mini-campaign to measure high-Z transport in the edge plasma. Comparisons will be made with ERO modeling Supported by US DOE DE-AC05-06OR23100, DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC05-00OR22725, DE-SC0001961, DE-AC04-94AL85000.

DiMES PMI research at DIII-D in support of ITER and beyond

DOE PAGES

Rudakov, Dimitry L.; Abrams, Tyler; Ding, Rui; ...

2017-03-27

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based onmore » experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. Furthermore, these techniques may find applications in the future devices.« less

Single Null Negative Triangularity Tokamak for Power Handling

NASA Astrophysics Data System (ADS)

Kikuchi, Mitsuru; Medvedev, S.; Takizuka, T.; Sauter, O.; Merle, A.; Coda, S.; Chen, D.; Li, J. X.

2017-10-01

Power and particle control in fusion reactor is challenge and we proposed the negative triangularity tokamak (NTT) to eliminate ELM by operating L-mode edge with improved core confinement. The SN configuration has more flexibility in shaping by adopting rectangular-shaped TF coils. The limiting normalized beta is 3.56 with wall stabilization and 3.14 without wall. The vertical stability is assured under a reasonable control system. The wetted area on the divertor plates becomes wider in proportion to the larger major radius at the divertor strike points due to the NT configuration. In addition to the major-radius effect, the ``Flux Tune Expansion (FTE)'' is adopted to further reduce the heat load on the divertor plate by factor of 2.6 with a coil current 3 MA. L-mode edge also allows further increase in wetted area. The fusion power of 3 GW is deliverable only at normalized beta 2.1. Therefore this reactor may be operable stably against the serious MHD activities. The CD power for SS operation is 175 MW at Q = 17. AC operation is also possible option. A required HH factor is relatively modest H = 1.12.

Design, Analysis and R&D of the EAST In-Vessel Components

NASA Astrophysics Data System (ADS)

Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

2008-06-01

In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

Long-term fuel retention and release in JET ITER-Like Wall at ITER-relevant baking temperatures

NASA Astrophysics Data System (ADS)

Heinola, K.; Likonen, J.; Ahlgren, T.; Brezinsek, S.; De Temmerman, G.; Jepu, I.; Matthews, G. F.; Pitts, R. A.; Widdowson, A.; Contributors, JET

2017-08-01

The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved from the W divertor and Be main chamber were annealed at 350 and 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 and 15 h to study the deuterium removal effectiveness at the nominal baking temperatures. The remained fraction was determined by emptying the samples fully of deuterium by heating W and Be samples up to 1000 and 775 °C,respectively. Results showed the deposits in the divertor having an increasing effect to the remaining retention at temperatures above baking. Highest remaining fractions 54 and 87 % were observed with deposit thicknesses of 10 and 40 μm, respectively. Substantially high fractions were obtained in the main chamber samples from the deposit-free erosion zone of the limiter midplane, in which the dominant fuel retention mechanism is via implantation: 15 h annealing resulted in retained deuterium higher than 90 % . TDS results from the divertor were simulated with TMAP7 calculations. The spectra were modelled with three deuterium activation energies resulting in good agreement with the experiments.

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

NASA Astrophysics Data System (ADS)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.; Evans, T. E.; West, W. P.; Whyte, D. G.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C2 under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discrimination of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C2 influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.

Spectroscopic determinations of carbon fluxes, sources, and shielding in the DIII-D divertors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Isler, R. C.; Colchin, R. J.; Brooks, N. H.

2001-10-01

The most important mechanisms for eroding plasma-facing components (PFCs) and introducing carbon into tokamak divertors are believed to be physical sputtering, chemical sputtering, sublimation, and radiation enhance sublimation (RES). The relative importance of these processes has been investigated by analyzing the spectral emission rates and the effective temperatures of CI, CD, and C{sub 2} under several operating conditions in the DIII-D tokamak [Plasma Physics Controlled Nuclear Fusion Research, 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159; Proceedings of the 18th IEEE/NPSS Symposium on Fusion Engineering, Albuquerque (Institute of Electrical and Electronic Engineers, Piscataway, 1999), p. 515]. Discriminationmore » of chemical sputtering from physical sputtering is accomplished by quantitatively relating the fraction of CI influxes expected from dissociation of hydrocarbons to the measured CD and C{sub 2} influxes. Characteristics of sublimation are studied from carbon test samples heated to surface temperatures exceeding 2000 K. The shielding efficiency of carbon produced at the divertor target is assessed from comparison of fluxes of neutral atoms and ions; approximately 95% of the primary influx appears to be redeposited before being transported far enough upstream to fuel the core plasma.« less

Realization of high heat flux tungsten monoblock type target with graded interlayer for application to DEMO divertor

NASA Astrophysics Data System (ADS)

Richou, M.; Gallay, F.; Böswirth, B.; Chu, I.; Lenci, M.; Loewenhoff, Th; Quet, A.; Greuner, H.; Kermouche, G.; Meillot, E.; Pintsuk, G.; Visca, E.; You, J. H.

2017-12-01

The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. In DEMO, divertor targets must survive an environment of high heat fluxes (˜up to 20 MW m-2 during slow transients) and neutron irradiation. One advanced concept for components in monoblock configuration concerns the insertion of a compositionally graded layer between tungsten and CuCrZr instead of the soft copper interlayer. As a first step, a thin graded layer (˜25 μm) was developed. As a second step, a thicker graded layer (˜500 μm), which is actually being developed, will also be inserted to study the compliant role of a macroscopic graded layer. This paper reports the results of cyclic high heat flux loading tests up to 20 MW m-2 and to heat flux higher than 25 MW m-2 that mock-ups equipped with thin graded layer survived without visible damage. First feedback on manufacturing steps is also presented. Moreover, the first results obtained on the development of the thick graded layer and its integration in a monoblock configuration are shown.

Pre-irradiation testing of actively cooled Be-Cu divertor modules

DOE Office of Scientific and Technical Information (OSTI.GOV)

Linke, J.; Duwe, R.; Kuehnlein, W.

1995-09-01

A set of neutron irradiation tests is prepared on different plasma facing materials (PFM) candidates and miniaturized components for ITER. Beside beryllium the irradiation program which will be performed in the High Flux Reactor (HFR) in Petten, includes different carbon fiber composites (CFQ) and tungsten alloys. The target values for the neutron irradiation will be 0.5 dpa at temperatures of 350{degrees}C and 700{degrees}C, resp.. The post irradiation examination (PIE) will cover a wide range of mechanical tests; in addition the degradation of thermal conductivity will be investigated. To determine the high heat flux (HHF) performance of actively cooled divertor modules,more » electron beam tests which simulate the expected heat loads during the operation of ITER, are scheduled in the hot cell electron beam facility JUDITH. These tests on a selection of different actively cooled beryllium-copper and CFC-copper divertor modules are performed before and after neutron irradiation; the pre-irradiation testing is an essential part of the program to quantify the zero-fluence high heat flux performance and to detect defects in the modules, in particular in the brazed joints.« less

Effect of n = 3 perturbation field amplitudes below the ELM triggering threshold on edge and SOL transport in NSTX

DOE PAGES

J. M. Canik; Lore, J. D.; Ahn, J. -W.; ...

2013-01-12

Here, the pulsed application of n = 3 magnetic perturbation fields with amplitudes below that which triggers ELMs results in distinct, transient responses observable on several edge and divertor diagnostics in NSTX. We refer to these responses as Sub-Threshold Edge Perturbations (STEPs). An analysis of edge measurements suggests that STEPs result in increased transport in the plasma edge and scrape-off layer, which leads to augmentation of the intrinsic strike point splitting due to error fields, i.e., an intensification of the helical divertor footprint flux pattern. These effects are much smaller in magnitude than those of triggered ELMs, and are observedmore » for the duration of the field perturbation measured internal to the vacuum vessel. In addition, STEPs are correlated with changes to the MHD activity, along with transient reductions in the neutron production rate. Ideally the STEPs could be used to provide density control and prevent impurity accumulation, in the same manner that on-demand ELM triggering is used on NSTX, without the impulsive divertor fluxes and potential for damage to plasma facing components associated with ELMs.« less

3D nonlinear numerical simulation of the current-convective instability in detached diverter plasma

NASA Astrophysics Data System (ADS)

Stepanenko, Alexander; Krasheninnikov, Sergei

2017-10-01

One of the possible mechanisms responsible for strong radiation fluctuations observed in the recent experiments with detached plasmas at ASDEX Upgrade [Potzel et al., Nuclear Fusion, 2014] can be related to the onset of the current-convective instability (CCI) driven by strong asymmetry of detachment in the inner and outer tokamak divertors [Krasheninnikov and Smolyakov, PoP, 2016]. In this study we present the first results of 3D nonlinear numerical simulations of the CCI in divertor plasma for the conditions relevant to the AUG experiment. The general physical model used to simulate the CCI, qualitative estimates for the instability characteristic growth rate and transverse wavelengths derived for plasma, which is spatially inhomogeneous both across and along the magnetic field lines, are presented. The simulation results, demonstrating nonlinear dynamics of the CCI, provide the frequency spectra of turbulent divertor plasma fluctuations showing good agreement with the available experimental data. This material is based upon the work supported by the U.S. Department of Energy under Award No. DE-FG02-04ER54739 at UCSD and by the Russian Ministry of Education and Science Grant No. 14.Y26.31.0008 at MEPhI.

Activation and Environmental Aspects of In-Vacuum Vessel Components of CFETR

NASA Astrophysics Data System (ADS)

Zhang, Xiaokang; Liu, Songlin; Zhu, Qingjun; Gao, Fangfang; Li, Jia

2016-11-01

The water-cooled ceramic breeder (WCCB) blanket is one of the three candidates of China's Fusion Engineering Test Reactor (CFETR). The evaluation of the radioactivity and decay heat produced by neutrons for the in-vacuum vessel components is essential for the assessment of radioactive wastes and the safety of CFETR. The activation calculation of CFETR in-vacuum vessel components was carried out by using the Monte Carlo N-Particle Transport Code MCNP, IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, and the nuclear inventory code FISPACT-2007 and corresponding EAF-2007 libraries. In these analyses, the three-dimensional (3-D) neutronics model was employed and the WCCB blanket, the divertor, and the shield were modeled in detail to provide the detailed spatial distribution of the neutron flux and energy spectra. Then the neutron flux, energy spectra and the materials specification were transferred to FISPACT for the activation calculation with an assumed irradiation scenario of CFETR. This paper presents the main results of the activation analysis to evaluate the radioactivity, the decay heat, the contact dose, and the waste classification of the radioactive materials. At the time of shutdown, the activity of the WCCB blanket is 1.88×1019 Bq and the specific activity, the decay heat and the contact dose rate are 1.7 × 1013 Bq/kg, 3.05 MW, and 2.0 × 103 Sv/h respectively. After cooling for 100 years, 79% (4166.4 tons) radioactive wastes produced from the blanket, divertor, high temperature shield (HTS) and low temperature shield (LTS) need near surface disposal, while 21% (1112.3 tons) need geological disposal. According to results of the contact dose rate, all the components of the blanket, divertor, HTS and LTS could potentially be recycled after shutdown by using advanced remote handling equipment. In addition, the selection of Eurofer97 or RAFM for the divertor is better than that of SS316 because SS316 makes the activity of the divertor-body keep at a relatively high level. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2013GB108004, 2015BG108002, 2014GB122000, 2014GB119000), National Natural Science Foundation of China (No. 11175207)

Mechanical Overstimulation of Hair Bundles: Suppression and Recovery of Active Motility

PubMed Central

Kao, Albert; Meenderink, Sebastiaan W. F.; Bozovic, Dolores

2013-01-01

We explore the effects of high-amplitude mechanical stimuli on hair bundles of the bullfrog sacculus. Under in vitro conditions, these bundles exhibit spontaneous limit cycle oscillations. Prolonged deflection exerted two effects. First, it induced an offset in the position of the bundle. Recovery to the original position displayed two distinct time scales, suggesting the existence of two adaptive mechanisms. Second, the stimulus suppressed spontaneous oscillations, indicating a change in the hair bundle’s dynamic state. After cessation of the stimulus, active bundle motility recovered with time. Both effects were dependent on the duration of the imposed stimulus. External calcium concentration also affected the recovery to the oscillatory state. Our results indicate that both offset in the bundle position and calcium concentration control the dynamic state of the bundle. PMID:23505461

Influence of knee flexion angle and transverse drill angle on creation of femoral tunnels in double-bundle anterior cruciate ligament reconstruction using the transportal technique: Three-dimensional computed tomography simulation analysis.

PubMed

Choi, Chong Hyuk; Kim, Sung-Jae; Chun, Yong-Min; Kim, Sung-Hwan; Lee, Su-Keon; Eom, Nam-Kyu; Jung, Min

2018-01-01

The purpose of this study was to find appropriate flexion angle and transverse drill angle for optimal femoral tunnels of anteromedial (AM) bundle and posterolateral (PL) bundle in double-bundle ACL reconstruction using transportal technique. Thirty three-dimensional knee models were reconstructed. Knee flexion angles were altered from 100° to 130° at intervals of 10°. Maximum transverse drill angle (MTA), MTA minus 10° and 20° were set up. Twelve different tunnels were determined by four flexion angles and three transverse drill angles for each bundle. Tunnel length, wall breakage, inter-tunnel communication and graft-bending angle were assessed. Mean tunnel length of AM bundle was >30mm at 120° and 130° of flexion in all transverse drill angles. Mean tunnel length of PL bundle was >30mm during every condition. There were ≥1 cases of wall breakage except at 120° and 130° of flexion with MTA for AM bundle. There was no case of wall breakage for PL bundle. Considering inter-tunnel gap of >2mm without communication and obtuse graft-bending angle, 120° of flexion and MTA could be recommended as optimal condition for femoral tunnels of AM and PL bundles. Flexion angle and transverse drill angle had combined effect on femoral tunnel in double-bundle ACL reconstruction using transportal technique. Achieving flexion angle of 120° and transverse drill angle close to the medial femoral condyle could be recommended as optimal condition for femoral tunnels of AM and PL bundles to avoid insufficient tunnel length, wall breakage, inter-tunnel communication and acute graft-bending angle. Copyright © 2017 Elsevier B.V. All rights reserved.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hughes, Michael John; McConnaughhay, Johnie Franklin

A combustor includes a tube bundle that extends radially across at least a portion of the combustor. The tube bundle includes an upstream surface axially separated from a downstream surface, and a plurality of tubes extend from the upstream surface through the downstream surface to provide fluid communication through the tube bundle. A barrier extends radially inside the tube bundle between the upstream and downstream surfaces, and a baffle extends axially inside the tube bundle between the upstream surface and the barrier.

Sampling system and method

DOEpatents

Decker, David L; Lyles, Brad F; Purcell, Richard G; Hershey, Ronald Lee

2014-05-20

An apparatus and method for supporting a tubing bundle during installation or removal. The apparatus includes a clamp for securing the tubing bundle to an external wireline. The method includes deploying the tubing bundle and wireline together, The tubing bundle is periodically secured to the wireline using a clamp.

Fiber-bundle-basis sparse reconstruction for high resolution wide-field microendoscopy.

PubMed

Mekhail, Simon Peter; Abudukeyoumu, Nilupaer; Ward, Jonathan; Arbuthnott, Gordon; Chormaic, Síle Nic

2018-04-01

In order to observe deep regions of the brain, we propose the use of a fiber bundle for microendoscopy. Fiber bundles allow for the excitation and collection of fluorescence as well as wide field imaging while remaining largely impervious to image distortions brought on by bending. Furthermore, their thin diameter, from 200-500 µ m, means their impact on living tissue, though not absent, is minimal. Although wide field imaging with a bundle allows for a high temporal resolution since no scanning is involved, the largest criticism of bundle imaging is the drastically lowered spatial resolution. In this paper, we make use of sparsity in the object being imaged to up sample the low resolution images from the fiber bundle with compressive sensing. We take each image in a single shot by using a measurement basis dictated by the quasi-crystalline arrangement of the bundle's cores. We find that this technique allows us to increase the resolution of a typical image taken through a fiber bundle.

Equilibrium polyelectrolyte bundles with different multivalent counterion concentrations

NASA Astrophysics Data System (ADS)

Sayar, Mehmet; Holm, Christian

2010-09-01

We present the results of molecular-dynamics simulations on the salt concentration dependence of the formation of polyelectrolyte bundles in thermodynamic equilibrium. Extending our results on salt-free systems we investigate here deficiency or excess of trivalent counterions in solution. Our results reveal that the trivalent counterion concentration significantly alters the bundle size and size distribution. The onset of bundle formation takes place at earlier Bjerrum length values with increasing trivalent counterion concentration. For the cases of 80%, 95%, and 100% charge compensation via trivalent counterions, the net charge of the bundles decreases with increasing size. We suggest that competition among two different mechanisms, counterion condensation and merger of bundles, leads to a nonmonotonic change in line-charge density with increasing Bjerrum length. The investigated case of having an abundance of trivalent counterions by 200% prohibits such a behavior. In this case, we find that the difference in effective line-charge density of different size bundles diminishes. In fact, the system displays an isoelectric point, where all bundles become charge neutral.

Overhead electric power transmission line jumpering system for bundles of five or more subconductors

DOEpatents

Winkelman, Paul F.

1982-01-01

Jumpering of electric power transmission lines at a dead end tower. Two transmission line conductor bundles each contain five or more spaced apart subconductors (5) arranged in the shape of a cylinder having a circular cross section. The ends of each bundle of subconductors are attached with insulators to a dead end tower (1). Jumpering allows the electric current to flow between the two bundles of subconductors using jumper buses, internal jumper conductors, and external jumper conductors. One or more current collecting jumper buses (37) are located inside each bundle of subconductors with each jumper bus being attached to the end of a subconductor. Small-diameter internal jumper conductors (33) are located in the inherently electrically shielded area inside each bundle of subconductors with each subconductor (except ones having an attached jumper bus) having one internal jumper conductor connected between that subconductor's end and a jumper bus. Large-diameter external jumper conductors (9) are located outside each bundle of subconductors with one or more external jumper conductors being connected between the jumper buses in one bundle of subconductors and the jumper buses in the other bundle.

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Characterization of midrib vascular bundles of selected medicinal species in Rubiaceae

NASA Astrophysics Data System (ADS)

Nurul-Syahirah, M.; Noraini, T.; Latiff, A.

2016-11-01

An anatomical study was carried out on mature leaves of five selected medicinal species of Rubiaceae from Peninsular Malaysia. The chosen medicinal species were Aidia densiflora, Aidia racemosa, Chasallia chartacea, Hedyotis auricularia and Ixora grandifolia. The objective of this study is to determine the taxonomic value of midrib anatomical characteristics. Leaves samples were collected from Taman Paku Pakis, Universiti Kebangsaan Malaysia, Bangi, Selangor and Kledang Saiong Forest Reserve, Perak, Malaysia. Leaves samples then were fixed in spirit and acetic acid (3:1), the midrib parts then were sectioned using sliding microtome, cleared using Clorox, stained in Safranin and Alcian blue, mounted in Euparal and were observed under light microscope. Findings in this study have shown all species have collateral bundles. The midrib vascular bundles characteristics that can be used as tool to differentiate between species or genus are vascular bundles system (opened or closed), shape and arrangement of main vascular bundles, presence of both additional and medullary vascular bundles, position of additional vascular bundles, shape of medullary vascular bundles, presence of sclerenchyma cells ensheathed the vascular bundles. As a conclusion, midrib anatomical characteristics can be used to identify and discriminate medicinal plants species studied in the Rubiaceae.

Evidence for two populations of hair bundles in the sea anemone, Nematostella vectensis.

PubMed

Menard, Shelcie S; Watson, Glen M

2017-06-01

Cytochalasin D (CD) was employed to disrupt F-actin within stereocilia of anemone hair bundles. CD treatment decreases the abundance of hair bundles (by 85%) while significantly impairing predation. The remaining hair bundles are 'CD-resistant.' Surprisingly, the morphology and F-actin content of resistant hair bundles are comparable to those of untreated controls. However, the resistant hair bundles fail to respond normally to the N-acetylated sugar, NANA, by elongating. Instead, they remain at resting length. Immediately after CD treatment, when only CD-resistant hair bundles are present, nematocyst discharge is normal into targets touched to tentacles in the absence of vibrations (i.e., baseline) but fails to increase normally in the presence of nearby vibrations at 56Hz, a key frequency. After CD treatment, the abundance of hair bundles recovers to control levels within three hours. At 2h after CD treatment, when CD-resistant and CD-sensitive hair bundles are both present, but a full-recovery is not yet complete, somewhat enhanced discharge of nematocysts occurs into targets touched to tentacles in the presence of nearby vibrations at 56Hz (at least as compared to the response of CD-treated animals to contact with test probes in the absence of vibrations). Additionally, at 2h after CD-treatment, prey capture recovers to normal. Thus, two populations of hair bundles may be present on tentacles of sea anemones: those that are CD-resistant and those that are CD-sensitive. The functions of these hair bundles may be distinct. Copyright © 2017 Elsevier Inc. All rights reserved.

Translating Pressure Ulcer Prevention Into Intensive Care Nursing Practice: Overlaying a Care Bundle Approach With a Model for Research Implementation.

PubMed

Tayyib, Nahla; Coyer, Fiona

This article reports on the development and implementation process used to integrate a care bundle approach (a pressure ulcer [PU] prevention bundle to improve patients' skin integrity in intensive care) and the Ottawa Model of Research Use (OMRU). The PU prevention care bundle demonstrated significant reduction in PU incidence, with the OMRU model providing a consolidated framework for the implementation of bundled evidence in an effective and consistent manner into daily clinical nursing practice.

Localized Statistics for DW-MRI Fiber Bundle Segmentation

PubMed Central

Lankton, Shawn; Melonakos, John; Malcolm, James; Dambreville, Samuel; Tannenbaum, Allen

2013-01-01

We describe a method for segmenting neural fiber bundles in diffusion-weighted magnetic resonance images (DWMRI). As these bundles traverse the brain to connect regions, their local orientation of diffusion changes drastically, hence a constant global model is inaccurate. We propose a method to compute localized statistics on orientation information and use it to drive a variational active contour segmentation that accurately models the non-homogeneous orientation information present along the bundle. Initialized from a single fiber path, the proposed method proceeds to capture the entire bundle. We demonstrate results using the technique to segment the cingulum bundle and describe several extensions making the technique applicable to a wide range of tissues. PMID:23652079

Field Emission Study of Carbon Nanotubes: High Current Density from Nanotube Bundle Arrays

NASA Technical Reports Server (NTRS)

Bronikowski, Micheal J.; Manohara, Harish M.; Siegel, Peter H.; Hunt, Brian D.

2004-01-01

We have investigated the field emission behavior of lithographically patterned bundles of multiwalled carbon nanotubes arranged in a variety of array geometries. Such arrays of nanotube bundles are found to perform significantly better in field emission than arrays of isolated nanotubes or dense, continuous mats of nanotubes, with the field emission performance depending on the bundle diameter and inter-bundle spacing. Arrays of 2-micrometers diameter nanotube bundles spaced 5 micrometers apart (edge-to-edge spacing) produced the largest emission densities, routinely giving 1.5 to 1.8 A/cm(sup 2) at approximately 4 V/micrometer electric field, and greater than 6 A/cm(sup 2) at 20 V/micrometers.

National Partnership for Maternal Safety: Consensus Bundle on Safe Reduction of Primary Cesarean Births-Supporting Intended Vaginal Births.

PubMed

Lagrew, David C; Low, Lisa Kane; Brennan, Rita; Corry, Maureen P; Edmonds, Joyce K; Gilpin, Brian G; Frost, Jennifer; Pinger, Whitney; Reisner, Dale P; Jaffer, Sara

2018-03-01

Cesarean births and associated morbidity and mortality have reached near epidemic proportions. The National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care responded by developing a patient safety bundle to reduce the number of primary cesarean births. Safety bundles outline critical practices to implement in every maternity unit. This National Partnership for Maternity Safety bundle, as with other bundles, is organized into four domains: Readiness, Recognition and Prevention, Response, and Reporting and Systems Learning. Bundle components may be adapted to individual facilities, but standardization within an institution is advised. Evidence-based resources and recommendations are provided to assist implementation.

Matrix remodeling between cells and cellular interactions with collagen bundle

NASA Astrophysics Data System (ADS)

Kim, Jihan; Sun, Bo

When cells are surrounded by complex environment, they continuously probe and interact with it by applying cellular traction forces. As cells apply traction forces, they can sense rigidity of their local environment and remodel the matrix microstructure simultaneously. Previous study shows that single human carcinoma cell (MDA-MB-231) remodeled its surrounding extracellular matrix (ECM) and the matrix remodeling was reversible. In this study we examined the matrix microstructure between cells and cellular interaction between them using quantitative confocal microscopy. The result shows that the matrix microstructure is the most significantly remodeled between cells consisting of aligned, and densified collagen fibers (collagen bundle)., the result shows that collagen bundle is irreversible and significantly change micromechanics of ECM around the bundle. We further examined cellular interaction with collagen bundle by analyzing dynamics of actin and talin formation along with the direction of bundle. Lastly, we analyzed dynamics of cellular protrusion and migrating direction of cells along the bundle.

Anatomic Double-Bundle Anterior Cruciate Ligament Reconstruction With a Free Quadriceps Tendon Autograft.

PubMed

Caterev, Sergiu; Nistor, Dan Viorel; Todor, Adrian

2016-10-01

Anatomic double-bundle anterior cruciate ligament (ACL) reconstruction aims to restore the 2 functional bundles of the ACL in an attempt to better reproduce the native biomechanics of the injured knee and promote long-term knee health. However, this concept is not fully accepted and is not performed on a standard basis. In addition, the superiority of this technique over the conventional single-bundle technique has been questioned, especially the long-term clinical results. One of the down sides of the double-bundle reconstruction is the complexity of the procedure, with increased risks, operative time, and costs compared with the single-bundle procedure. Also, the revision procedure, if necessary, is more challenging. We propose a technique that has some advantages over the traditional double-bundle procedure, using a single femoral tunnel, 2 tibial tunnels, and a free quadriceps tendon autograft.

Synchronization of Spontaneous Active Motility of Hair Cell Bundles

PubMed Central

Zhang, Tracy-Ying; Ji, Seung; Bozovic, Dolores

2015-01-01

Hair cells of the inner ear exhibit an active process, believed to be crucial for achieving the sensitivity of auditory and vestibular detection. One of the manifestations of the active process is the occurrence of spontaneous hair bundle oscillations in vitro. Hair bundles are coupled by overlying membranes in vivo; hence, explaining the potential role of innate bundle motility in the generation of otoacoustic emissions requires an understanding of the effects of coupling on the active bundle dynamics. We used microbeads to connect small groups of hair cell bundles, using in vitro preparations that maintain their innate oscillations. Our experiments demonstrate robust synchronization of spontaneous oscillations, with either 1:1 or multi-mode phase-locking. The frequency of synchronized oscillation was found to be near the mean of the innate frequencies of individual bundles. Coupling also led to an improved regularity of entrained oscillations, demonstrated by an increase in the quality factor. PMID:26540409

Exploring Differential Bundle Functioning in Mathematics by Gender: The Effect of Hierarchical Modelling

ERIC Educational Resources Information Center

Ong, Yoke Mooi; Williams, Julian; Lamprianou, Iasonas

2013-01-01

Researchers interested in exploring substantive group differences are increasingly attending to bundles of items (or testlets): the aim is to understand how gender differences, for instance, are explained by differential performances on different types or bundles of items, hence differential bundle functioning (DBF). Some previous work has…

Universal fiber-optic C.I.E. colorimeter

DOEpatents

Kronberg, James W.

1992-01-01

Apparatus for color measurements according to the C.I.E. system comprises a first fiber optic cable for receiving and linearizing light from a light source, a lens system for spectrally displaying the linearized light and focusing the light on one end of a trifurcated fiber optic assembly that integrates and separates the light according to the three C.I.E. tristimulus functions. The separated light is received by three photodiodes and electronically evaluated to determine the magnitude of the light corresponding to the tristimulus functions. The fiber optic assembly is made by forming, at one end, a bundle of optic fibers to match the contours of one of the tristimulus functions, encapsulating that bundle, adding a second bundle that, together with the first bundle, will match the contours of the first plus one other tristimulus function, encapsulating that second bundle, then adding a third bundle which together with the first and second bundles, has contours matching the sum of all three tristimulus functions. At the other end of the assembly the three bundles are separated and aligned with their respective photodiodes.

Implementing a Daily Maintenance Care Bundle to Prevent Central Line-Associated Bloodstream Infections in Pediatric Oncology Patients.

PubMed

Duffy, Elizabeth A; Rodgers, Cheryl C; Shever, Leah L; Hockenberry, Marilyn J

2015-01-01

Eliminating central line-associated bloodstream infection (CLABSI) is a national priority. Central venous catheter (CVC) care bundles are composed of a series of interventions that, when used together, are effective in preventing CLABSI. A CVC daily maintenance care bundle includes procedural guidelines for hygiene, dressing changes, and access as well as specific timeframes. Failure to complete one of the components of the care bundle predisposes the patient to a bloodstream infection. A nurse-led multidisciplinary team implemented and, for six months, sustained a daily maintenance care bundle for pediatric oncology patients. This quality improvement project focused on nursing staffs' implementation of the daily maintenance care bundle and the sustainment of the intervention. The project used a pre-post program design to evaluate outcomes of CVC daily maintenance care bundle compliancy and CLABSI. A statistically significant increase between the pre- and post-assessments of the compliance was noted with the CVC daily maintenance care bundle. CLABSI infection rates decreased during the intervention. Strategies to implement practice change and promote sustainability are discussed. © 2015 by Association of Pediatric Hematology/Oncology Nurses.

National Partnership for Maternal Safety: Consensus Bundle on Severe Hypertension During Pregnancy and the Postpartum Period.

PubMed

Bernstein, Peter S; Martin, James N; Barton, John R; Shields, Laurence E; Druzin, Maurice L; Scavone, Barbara M; Frost, Jennifer; Morton, Christine H; Ruhl, Catherine; Slager, Joan; Tsigas, Eleni Z; Jaffer, Sara; Menard, M Kathryn

2017-08-01

Complications arising from hypertensive disorders of pregnancy are among the leading causes of preventable severe maternal morbidity and mortality. Timely and appropriate treatment has the potential to significantly reduce hypertension-related complications. To assist health care providers in achieving this goal, this patient safety bundle provides guidance to coordinate and standardize the care provided to women with severe hypertension during pregnancy and the postpartum period. This is one of several patient safety bundles developed by multidisciplinary work groups of the National Partnership for Maternal Safety under the guidance of the Council on Patient Safety in Women's Health Care. These safety bundles outline critical clinical practices that should be implemented in every maternity care setting. Similar to other bundles that have been developed and promoted by the Partnership, the hypertension safety bundle is organized into four domains: Readiness, Recognition and Prevention, Response, and Reporting and Systems Learning. Although the bundle components may be adapted to meet the resources available in individual facilities, standardization within an institution is strongly encouraged. This commentary provides information to assist with bundle implementation.

Microtubule bundling plays a role in ethylene-mediated cortical microtubule reorientation in etiolated Arabidopsis hypocotyls.

PubMed

Ma, Qianqian; Sun, Jingbo; Mao, Tonglin

2016-05-15

The gaseous hormone ethylene is known to regulate plant growth under etiolated conditions (the 'triple response'). Although organization of cortical microtubules is essential for cell elongation, the underlying mechanisms that regulate microtubule organization by hormone signaling, including ethylene, are ambiguous. In the present study, we demonstrate that ethylene signaling participates in regulation of cortical microtubule reorientation. In particular, regulation of microtubule bundling is important for this process in etiolated hypocotyls. Time-lapse analysis indicated that selective stabilization of microtubule-bundling structures formed in various arrays is related to ethylene-mediated microtubule orientation. Bundling events and bundle growth lifetimes were significantly increased in oblique and longitudinal arrays, but decreased in transverse arrays in wild-type cells in response to ethylene. However, the effects of ethylene on microtubule bundling were partially suppressed in a microtubule-bundling protein WDL5 knockout mutant (wdl5-1). This study suggests that modulation of microtubule bundles that have formed in certain orientations plays a role in reorienting microtubule arrays in response to ethylene-mediated etiolated hypocotyl cell elongation. © 2016. Published by The Company of Biologists Ltd.

Abnormal Q waves in right sided chest leads provoked by onset of right bundle-branch block in patients with anteroseptal infarction.

PubMed Central

Rosenbaum, M B; Girotti, L A; Lázzari, J O; Halpern, M S; Elizari, M V

1982-01-01

In five cases of anteroseptal myocardial infarction complicated by intermittent right bundle-branch block, the onset of right bundle-branch block provoked the appearance of abnormal Q waves in leads V1 and V2, whereas a small initial R wave was present in the same leads during normal conduction. The intermittency of the conduction disturbance indicated that the Q waves were "right bundle-branch block dependent". It was also apparent that right bundle-branch block shifted the electrical location of the infarct towards the right, and made it look much larger. Right bundle-branch block dependent Q waves may arise during the acute stage of an anterior infarct suggesting, fallaciously, that an acute extension has occurred, or during the chronic stage, leading to the erroneous supposition that a new infarct had developed. The abnormal Q waves anteroseptal infarction complicated by fixed right bundle-branch block, though obviously related to the infarct, may be dependent on the right bundle-branch block. PMID:7059400

Actin-binding proteins sensitively mediate F-actin bundle stiffness

NASA Astrophysics Data System (ADS)

Claessens, Mireille M. A. E.; Bathe, Mark; Frey, Erwin; Bausch, Andreas R.

2006-09-01

Bundles of filamentous actin (F-actin) form primary structural components of a broad range of cytoskeletal processes including filopodia, sensory hair cell bristles and microvilli. Actin-binding proteins (ABPs) allow the cell to tailor the dimensions and mechanical properties of the bundles to suit specific biological functions. Therefore, it is important to obtain quantitative knowledge on the effect of ABPs on the mechanical properties of F-actin bundles. Here we measure the bending stiffness of F-actin bundles crosslinked by three ABPs that are ubiquitous in eukaryotes. We observe distinct regimes of bundle bending stiffness that differ by orders of magnitude depending on ABP type, concentration and bundle size. The behaviour observed experimentally is reproduced quantitatively by a molecular-based mechanical model in which ABP shearing competes with F-actin extension/compression. Our results shed new light on the biomechanical function of ABPs and demonstrate how single-molecule properties determine mesoscopic behaviour. The bending mechanics of F-actin fibre bundles are general and have implications for cytoskeletal mechanics and for the rational design of functional materials.

Negative hair-bundle stiffness betrays a mechanism for mechanical amplification by the hair cell.

PubMed

Martin, P; Mehta, A D; Hudspeth, A J

2000-10-24

Hearing and balance rely on the ability of hair cells in the inner ear to sense miniscule mechanical stimuli. In each cell, sound or acceleration deflects the mechanosensitive hair bundle, a tuft of rigid stereocilia protruding from the cell's apical surface. By altering the tension in gating springs linked to mechanically sensitive transduction channels, this deflection changes the channels' open probability and elicits an electrical response. To detect weak stimuli despite energy losses caused by viscous dissipation, a hair cell can use active hair-bundle movement to amplify its mechanical inputs. This amplificatory process also yields spontaneous bundle oscillations. Using a displacement-clamp system to measure the mechanical properties of individual hair bundles from the bullfrog's ear, we found that an oscillatory bundle displays negative slope stiffness at the heart of its region of mechanosensitivity. Offsetting the hair bundle's position activates an adaptation process that shifts the region of negative stiffness along the displacement axis. Modeling indicates that the interplay between negative bundle stiffness and the motor responsible for mechanical adaptation produces bundle oscillation similar to that observed. Just as the negative resistance of electrically excitable cells and of tunnel diodes can be embedded in a biasing circuit to amplify electrical signals, negative stiffness can be harnessed to amplify mechanical stimuli in the ear.

Influence of fiber packing structure on permeability

NASA Technical Reports Server (NTRS)

Cai, Zhong; Berdichevsky, Alexander L.

1993-01-01

The study on the permeability of an aligned fiber bundle is the key building block in modeling the permeability of advanced woven and braided preforms. Available results on the permeability of fiber bundles in the literature show that a substantial difference exists between numerical and analytical calculations on idealized fiber packing structures, such as square and hexagonal packing, and experimental measurements on practical fiber bundles. The present study focuses on the variation of the permeability of a fiber bundle under practical process conditions. Fiber bundles are considered as containing openings and fiber clusters within the bundle. Numerical simulations on the influence of various openings on the permeability were conducted. Idealized packing structures are used, but with introduced openings distributed in different patterns. Both longitudinal and transverse flow are considered. The results show that openings within the fiber bundle have substantial effect on the permeability. In the longitudinal flow case, the openings become the dominant flow path. In the transverse flow case, the fiber clusters reduce the gap sizes among fibers. Therefore the permeability is greatly influenced by these openings and clusters, respectively. In addition to the porosity or fiber volume fraction, which is commonly used in the permeability expression, another fiber bundle status parameter, the ultimate fiber volume fraction, is introduced to capture the disturbance within a fiber bundle.

Monoubiquitination Inhibits the Actin Bundling Activity of Fascin*

PubMed Central

Lin, Shengchen; Lu, Shuang; Mulaj, Mentor; Fang, Bin; Keeley, Tyler; Wan, Lixin; Hao, Jihui; Muschol, Martin; Sun, Jianwei; Yang, Shengyu

2016-01-01

Fascin is an actin bundling protein that cross-links individual actin filaments into straight, compact, and stiff bundles, which are crucial for the formation of filopodia, stereocillia, and other finger-like membrane protrusions. The dysregulation of fascin has been implicated in cancer metastasis, hearing loss, and blindness. Here we identified monoubiquitination as a novel mechanism that regulates fascin bundling activity and dynamics. The monoubiquitination sites were identified to be Lys247 and Lys250, two residues located in a positive charge patch at the actin binding site 2 of fascin. Using a chemical ubiquitination method, we synthesized chemically monoubiquitinated fascin and determined the effects of monoubiquitination on fascin bundling activity and dynamics. Our data demonstrated that monoubiquitination decreased the fascin bundling EC50, delayed the initiation of bundle assembly, and accelerated the disassembly of existing bundles. By analyzing the electrostatic properties on the solvent-accessible surface of fascin, we proposed that monoubiquitination introduced steric hindrance to interfere with the interaction between actin filaments and the positively charged patch at actin binding site 2. We also identified Smurf1 as a E3 ligase regulating the monoubiquitination of fascin. Our findings revealed a previously unidentified regulatory mechanism for fascin, which will have important implications for the understanding of actin bundle regulation under physiological and pathological conditions. PMID:27879315

Accuracy of Shack-Hartmann wavefront sensor using a coherent wound fibre image bundle

NASA Astrophysics Data System (ADS)

Zheng, Jessica R.; Goodwin, Michael; Lawrence, Jon

2018-03-01

Shack-Hartmannwavefront sensors using wound fibre image bundles are desired for multi-object adaptive optical systems to provide large multiplex positioned by Starbugs. The use of a large-sized wound fibre image bundle provides the flexibility to use more sub-apertures wavefront sensor for ELTs. These compact wavefront sensors take advantage of large focal surfaces such as the Giant Magellan Telescope. The focus of this paper is to study the wound fibre image bundle structure defects effect on the centroid measurement accuracy of a Shack-Hartmann wavefront sensor. We use the first moment centroid method to estimate the centroid of a focused Gaussian beam sampled by a simulated bundle. Spot estimation accuracy with wound fibre image bundle and its structure impact on wavefront measurement accuracy statistics are addressed. Our results show that when the measurement signal-to-noise ratio is high, the centroid measurement accuracy is dominated by the wound fibre image bundle structure, e.g. tile angle and gap spacing. For the measurement with low signal-to-noise ratio, its accuracy is influenced by the read noise of the detector instead of the wound fibre image bundle structure defects. We demonstrate this both with simulation and experimentally. We provide a statistical model of the centroid and wavefront error of a wound fibre image bundle found through experiment.

Spatial confinement of active microtubule networks induces large-scale rotational cytoplasmic flow

PubMed Central

Suzuki, Kazuya; Miyazaki, Makito; Takagi, Jun; Itabashi, Takeshi; Ishiwata, Shin’ichi

2017-01-01

Collective behaviors of motile units through hydrodynamic interactions induce directed fluid flow on a larger length scale than individual units. In cells, active cytoskeletal systems composed of polar filaments and molecular motors drive fluid flow, a process known as cytoplasmic streaming. The motor-driven elongation of microtubule bundles generates turbulent-like flow in purified systems; however, it remains unclear whether and how microtubule bundles induce large-scale directed flow like the cytoplasmic streaming observed in cells. Here, we adopted Xenopus egg extracts as a model system of the cytoplasm and found that microtubule bundle elongation induces directed flow for which the length scale and timescale depend on the existence of geometrical constraints. At the lower activity of dynein, kinesins bundle and slide microtubules, organizing extensile microtubule bundles. In bulk extracts, the extensile bundles connected with each other and formed a random network, and vortex flows with a length scale comparable to the bundle length continually emerged and persisted for 1 min at multiple places. When the extracts were encapsulated in droplets, the extensile bundles pushed the droplet boundary. This pushing force initiated symmetry breaking of the randomly oriented bundle network, leading to bundles aligning into a rotating vortex structure. This vortex induced rotational cytoplasmic flows on the length scale and timescale that were 10- to 100-fold longer than the vortex flows emerging in bulk extracts. Our results suggest that microtubule systems use not only hydrodynamic interactions but also mechanical interactions to induce large-scale temporally stable cytoplasmic flow. PMID:28265076

Reducing hospital-acquired pressure ulcers using bundle methodology in pediatric and neonatal patients receiving extracorporeal membrane oxygenation therapy: An integrative review and call to action.

PubMed

Courtwright, Suzanne E; Mastro, Kari A; Preuster, Christa; Dardashti, Navid; McGill, Sandra; Madelon, Myrlene; Johnson, Donna

2017-10-01

This review focuses on identifying (1) evidence of the effectiveness of care bundle methodology to reduce hospital-acquired pressure ulcers (HAPUs) in pediatric and neonatal patients receiving extracorporeal membrane oxygenation (ECMO) therapy and (2) barriers to implementing HAPU care bundles in this at-risk population. An integrative review was conducted and reported following the Preferred Reporting Items for Systematic Reviews and Meta-Analyses guidelines. A search of the scientific literature was performed. Studies included were published between January 2011 and February 2016. A total of seven articles met inclusion criteria. Data were extracted from each published article and analyzed to identify common themes, specifically bundle methodology and barriers to implementing HAPU bundles, in this population. There is limited research on effectiveness of care bundle methodology in reducing HAPUs in children, and no research specific to its effectiveness in pediatric or neonatal ECMO patients. No research was identified studying barriers to implementation of HAPU care bundles in this population. Nurses are well poised to test innovative strategies to prevent HAPUs. Nurses should consider implementing and testing bundle methodology to reduce HAPU in this at-risk population, and conduct research to identify any barriers to implementing this strategy. There is literature to support the use of nurses as unit-based skin care champions to facilitate teamwork and reliable use of the bundle, both critical components to the success of bundle methodology. © 2017 Wiley Periodicals, Inc.

Sustained Reduction of Ventilator-Associated Pneumonia Rates Using Real-Time Course Correction With a Ventilator Bundle Compliance Dashboard.

PubMed

Talbot, Thomas R; Carr, Devin; Parmley, C Lee; Martin, Barbara J; Gray, Barbara; Ambrose, Anna; Starmer, Jack

2015-11-01

The effectiveness of practice bundles on reducing ventilator-associated pneumonia (VAP) has been questioned. To implement a comprehensive program that included a real-time bundle compliance dashboard to improve compliance and reduce ventilator-associated complications. DESIGN Before-and-after quasi-experimental study with interrupted time-series analysis. SETTING Academic medical center. In 2007 a comprehensive institutional ventilator bundle program was developed. To assess bundle compliance and stimulate instant course correction of noncompliant parameters, a real-time computerized dashboard was developed. Program impact in 6 adult intensive care units (ICUs) was assessed. Bundle compliance was noted as an overall cumulative bundle adherence assessment, reflecting the percentage of time all elements were concurrently in compliance for all patients. The VAP rate in all ICUs combined decreased from 19.5 to 9.2 VAPs per 1,000 ventilator-days following program implementation (P<.001). Bundle compliance significantly increased (Z100 score of 23% in August 2007 to 83% in June 2011 [P<.001]). The implementation resulted in a significant monthly decrease in the overall ICU VAP rate of 3.28/1,000 ventilator-days (95% CI, 2.64-3.92/1,000 ventilator-days). Following the intervention, the VAP rate decreased significantly at a rate of 0.20/1,000 ventilator-days per month (95% CI, 0.14-0.30/1,000 ventilator-days per month). Among all adult ICUs combined, improved bundle compliance was moderately correlated with monthly VAP rate reductions (Pearson correlation coefficient, -0.32). A prevention program using a real-time bundle adherence dashboard was associated with significant sustained decreases in VAP rates and an increase in bundle compliance among adult ICU patients.

High-resolution imaging of the retinal nerve fiber layer in normal eyes using adaptive optics scanning laser ophthalmoscopy.

PubMed

Takayama, Kohei; Ooto, Sotaro; Hangai, Masanori; Arakawa, Naoko; Oshima, Susumu; Shibata, Naohisa; Hanebuchi, Masaaki; Inoue, Takashi; Yoshimura, Nagahisa

2012-01-01

To conduct high-resolution imaging of the retinal nerve fiber layer (RNFL) in normal eyes using adaptive optics scanning laser ophthalmoscopy (AO-SLO). AO-SLO images were obtained in 20 normal eyes at multiple locations in the posterior polar area and a circular path with a 3-4-mm diameter around the optic disc. For each eye, images focused on the RNFL were recorded and a montage of AO-SLO images was created. AO-SLO images for all eyes showed many hyperreflective bundles in the RNFL. Hyperreflective bundles above or below the fovea were seen in an arch from the temporal periphery on either side of a horizontal dividing line to the optic disc. The dark lines among the hyperreflective bundles were narrower around the optic disc compared with those in the temporal raphe. The hyperreflective bundles corresponded with the direction of the striations on SLO red-free images. The resolution and contrast of the bundles were much higher in AO-SLO images than in red-free fundus photography or SLO red-free images. The mean hyperreflective bundle width around the optic disc had a double-humped shape; the bundles at the temporal and nasal sides of the optic disc were narrower than those above and below the optic disc (P<0.001). RNFL thickness obtained by optical coherence tomography correlated with the hyperreflective bundle widths on AO-SLO (P<0.001) AO-SLO revealed hyperreflective bundles and dark lines in the RNFL, believed to be retinal nerve fiber bundles and Müller cell septa. The widths of the nerve fiber bundles appear to be proportional to the RNFL thickness at equivalent distances from the optic disc.

High-resolution imaging of retinal nerve fiber bundles in glaucoma using adaptive optics scanning laser ophthalmoscopy.

PubMed

Takayama, Kohei; Ooto, Sotaro; Hangai, Masanori; Ueda-Arakawa, Naoko; Yoshida, Sachiko; Akagi, Tadamichi; Ikeda, Hanako Ohashi; Nonaka, Atsushi; Hanebuchi, Masaaki; Inoue, Takashi; Yoshimura, Nagahisa

2013-05-01

To detect pathologic changes in retinal nerve fiber bundles in glaucomatous eyes seen on images obtained by adaptive optics (AO) scanning laser ophthalmoscopy (AO SLO). Prospective cross-sectional study. Twenty-eight eyes of 28 patients with open-angle glaucoma and 21 normal eyes of 21 volunteer subjects underwent a full ophthalmologic examination, visual field testing using a Humphrey Field Analyzer, fundus photography, red-free SLO imaging, spectral-domain optical coherence tomography, and imaging with an original prototype AO SLO system. The AO SLO images showed many hyperreflective bundles suggesting nerve fiber bundles. In glaucomatous eyes, the nerve fiber bundles were narrower than in normal eyes, and the nerve fiber layer thickness was correlated with the nerve fiber bundle widths on AO SLO (P < .001). In the nerve fiber layer defect area on fundus photography, the nerve fiber bundles on AO SLO were narrower compared with those in normal eyes (P < .001). At 60 degrees on the inferior temporal side of the optic disc, the nerve fiber bundle width was significantly lower, even in areas without nerve fiber layer defect, in eyes with glaucomatous eyes compared with normal eyes (P = .026). The mean deviations of each cluster in visual field testing were correlated with the corresponding nerve fiber bundle widths (P = .017). AO SLO images showed reduced nerve fiber bundle widths both in clinically normal and abnormal areas of glaucomatous eyes, and these abnormalities were associated with visual field defects, suggesting that AO SLO may be useful for detecting early nerve fiber bundle abnormalities associated with loss of visual function. Copyright © 2013 Elsevier Inc. All rights reserved.

Framework for shape analysis of white matter fiber bundles.

PubMed

Glozman, Tanya; Bruckert, Lisa; Pestilli, Franco; Yecies, Derek W; Guibas, Leonidas J; Yeom, Kristen W

2018-02-15

Diffusion imaging coupled with tractography algorithms allows researchers to image human white matter fiber bundles in-vivo. These bundles are three-dimensional structures with shapes that change over time during the course of development as well as in pathologic states. While most studies on white matter variability focus on analysis of tissue properties estimated from the diffusion data, e.g. fractional anisotropy, the shape variability of white matter fiber bundle is much less explored. In this paper, we present a set of tools for shape analysis of white matter fiber bundles, namely: (1) a concise geometric model of bundle shapes; (2) a method for bundle registration between subjects; (3) a method for deformation estimation. Our framework is useful for analysis of shape variability in white matter fiber bundles. We demonstrate our framework by applying our methods on two datasets: one consisting of data for 6 normal adults and another consisting of data for 38 normal children of age 11 days to 8.5 years. We suggest a robust and reproducible method to measure changes in the shape of white matter fiber bundles. We demonstrate how this method can be used to create a model to assess age-dependent changes in the shape of specific fiber bundles. We derive such models for an ensemble of white matter fiber bundles on our pediatric dataset and show that our results agree with normative human head and brain growth data. Creating these models for a large pediatric longitudinal dataset may improve understanding of both normal development and pathologic states and propose novel parameters for the examination of the pediatric brain. Copyright © 2017 Elsevier Inc. All rights reserved.

Implementation of a post-cardiac arrest care bundle including therapeutic hypothermia and hemodynamic optimization in comatose patients with return of spontaneous circulation after out-of-hospital cardiac arrest: a feasibility study.

PubMed

Walters, Elizabeth Lea; Morawski, Kyle; Dorotta, Ihab; Ramsingh, Davinder; Lumen, Kelly; Bland, David; Clem, Kathleen; Nguyen, H Bryant

2011-04-01

Patients who present to the emergency department (ED) with return of spontaneous circulation after cardiac arrest generally have poor outcomes. Guidelines for treatment can be complicated and difficult to implement. This study examined the feasibility of implementing a care bundle including therapeutic hypothermia (TH) and early hemodynamic optimization for comatose patients with return of spontaneous circulation after out-of-hospital cardiac arrest. The study included patients over a 2-year period in the ED and intensive care unit of an academic tertiary-care medical center. The first year (prebundle) provided a historical control, followed by a prospective observational period of bundle implementation during the second year. The bundle elements included (a) TH initiated; (b) central venous pressure/central venous oxygen saturation monitoring in 2 h; (c) target temperature in 4 h; (d) central venous pressure greater than 12 mmHg in 6 h; (e) MAP greater than 65 mmHg in 6 h; (f) central venous oxygen saturation greater than 70% in 6 h; (g) TH maintained for 24 h; and (h) decreasing lactate in 24 h. Fifty-five patients were enrolled, 26 patients in the prebundle phase and 29 patients in the bundle phase. Seventy-seven percent of bundle elements were completed during the bundle phase. In-hospital mortality in bundle compared with prebundle patients was 55.2% vs. 69.2% (P = 0.29). In the bundle patients, those patients who received all elements of the care bundle had mortality 33.3% compared with 60.9% in those receiving some of the bundle elements (P = 0.22). Bundle patients tended to achieve good neurologic outcome compared with prebundle patients, Cerebral Performance Category 1 or 2 in 31 vs. 12% patients, respectively (P = 0.08). Our study demonstrated that a post-cardiac arrest care bundle that incorporates TH and early hemodynamic optimization can be implemented in the ED and intensive care unit collaboratively and can achieve similar clinical benefits compared with those observed in previous clinical trials.

Concept development for the ITER equatorial port visible/infrared wide angle viewing system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reichle, R.; Beaumont, B.; Boilson, D.

2012-10-15

The ITER equatorial port visible/infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R and D topicsmore » are outlined.« less

Turkish and Native English Academic Writers' Use of Lexical Bundles

ERIC Educational Resources Information Center

Öztürk, Yusuf; Köse, Gül Durmusoglu

2016-01-01

Lexical bundles such as "on the other hand" and "as a result of" are extremely common and important in academic discourse. The appropriate use of lexical bundles typical of a specific academic discipline is important for writers and the absence of such bundles may not sound fluent and native-like. Recent studies (e.g. Adel…

Bundles over nearly-Kahler homogeneous spaces in heterotic string theory

NASA Astrophysics Data System (ADS)

Klaput, Michael; Lukas, Andre; Matti, Cyril

2011-09-01

We construct heterotic vacua based on six-dimensional nearly-Kahler homogeneous manifolds and non-trivial vector bundles thereon. Our examples are based on three specific group coset spaces. It is shown how to construct line bundles over these spaces, compute their properties and build up vector bundles consistent with supersymmetry and anomaly cancelation. It turns out that the most interesting coset is SU(3)/U(1)2. This space supports a large number of vector bundles which lead to consistent heterotic vacua, some of them with three chiral families.

Measurements of plasma sheath heat flux in the Alcator C-Mod divertor

NASA Astrophysics Data System (ADS)

Brunner, Dan; Labombard, Brian; Terry, Jim; Reinke, Matt

2010-11-01

Heat flux is one of the most important parameters controlling the lifetime of first-wall components in fusion experiments and reactors. The sheath heat flux coefficient (γ) is a parameter relating heat flux (from a plasma to a material surface) to the electron temperature and ion saturation current. Being such a simple expression for a kinetic process, it is of great interest to plasma edge fluid modelers. Under the assumptions of equal ion and electron temperatures, no secondary electron emission, and no net current to the surface the value of γ is approximately 7 [1]. Alcator C-Mod provides a unique opportunity among today's experiments to measure reactor-relevant heat fluxes (100's of MW/m^2 parallel to the magnetic field) in reactor-like divertor geometry. Motivated by the DoE 2010 joint milestone to measure heat flux footprints, the lower outer divertor of Alcator has been instrumented with a suite of Langmuir probes, novel surface thermocouples, and calorimeters in tiles purposefully ramped to eliminate shadowing; all within view of an IR camera. Initial results indicate that the experimentally inferred values of γ are found to agree with simple theory in the sheath limited regime and diverges to lower values as the density increases.

Modifications of W and Mo leading edges under plasma loads in DIII-D divertor

NASA Astrophysics Data System (ADS)

Rudakov, D. L.; Bykov, I.; Moyer, R. A.; Abrams, T.; Chrobak, C. P.; Guo, H. Y.; Stahl, B.; Thomas, D. M.; Barton, J. L.; Nygren, R. E.; Watkins, J. G.; Lasnier, C. J.; Litnovsky, Andrey; Stangeby, P. C.; Unterberg, E. A.

2017-10-01

Cracking and melting of W and Mo leading edges were observed in the lower divertor of DIII-D during experiments with intentionally misaligned W monoblocks (MBs) and in the course of the Metal Rings Campaign involving W-coated Mo tile inserts (TIs). MBs were exposed near the attached outer strike point during deuterium and helium L- and H-mode discharges using DiMES. Two of the MBs were misaligned by 0.3 mm and 1 mm, forming leading edges. Particulate ejection from a 1 mm leading edge was observed during the exposure, and evidence of melting and cracking was found post mortem. Two toroidal rings of TIs were installed in the lower outer divertor, the inner one at the floor and the outer one at the shelf. The floor TIs bowed during plasma exposure forming leading edges up to 1.2 mm high; about 40% of these edges experienced melting. Re-solidified melt layers up to 1 mm thick were observed, their shape being consistent with motion in the jx B direction with j driven by electron emission. Work supported by US DOE under DE-FC02-04ER54698, DE-FG02-07ER54917, DE-AC04-94AL85000, DE-AC52-07NA27344, and DE-AC05-00OR22725.

Formation of a three-dimensional plasma boundary after decay of the plasma response to resonant magnetic perturbation fields

NASA Astrophysics Data System (ADS)

Schmitz, O.; Evans, T. E.; Fenstermacher, M. E.; Lanctot, M. J.; Lasnier, C. L.; Mordijck, S.; Moyer, R. A.; Reimerdes, H.; the DIII-D Team

2014-01-01

First time experimental evidence is presented for a direct link between the decay of a n = 3 plasma response and the formation of a three-dimensional (3D) plasma boundary. We inspect a lower single-null L-mode plasma which first reacts at sufficiently high rotation with an ideal resonant screening response to an external toroidal mode number n = 3 resonant magnetic perturbation field. Decay of this response due to reduced bulk plasma rotation changes the plasma state considerably. Signatures such as density pump out and a spin up of the edge rotation—which are usually connected to formation of a stochastic boundary—are detected. Coincident, striation of the divertor single ionized carbon emission and a 3D emission structure in double ionized carbon at the separatrix is seen. The striated C II pattern follows in this stage the perturbed magnetic footprint modelled without a plasma response (vacuum approach). This provides for the first time substantial experimental evidence, that a 3D plasma boundary with direct impact on the divertor particle flux pattern is formed as soon as the internal plasma response decays. The resulting divertor structure follows the vacuum modelled magnetic field topology. However, the inward extension of the perturbed boundary layer can still not directly be determined from these measurements.

Control of particle and power exhaust in pellet fuelled ITER DT scenarios employing integrated models

NASA Astrophysics Data System (ADS)

Wiesen, S.; Köchl, F.; Belo, P.; Kotov, V.; Loarte, A.; Parail, V.; Corrigan, G.; Garzotti, L.; Harting, D.

2017-07-01

The integrated model JINTRAC is employed to assess the dynamic density evolution of the ITER baseline scenario when fuelled by discrete pellets. The consequences on the core confinement properties, α-particle heating due to fusion and the effect on the ITER divertor operation, taking into account the material limitations on the target heat loads, are discussed within the integrated model. Using the model one can observe that stable but cyclical operational regimes can be achieved for a pellet-fuelled ITER ELMy H-mode scenario with Q  =  10 maintaining partially detached conditions in the divertor. It is shown that the level of divertor detachment is inversely correlated with the core plasma density due to α-particle heating, and thus depends on the density evolution cycle imposed by pellet ablations. The power crossing the separatrix to be dissipated depends on the enhancement of the transport in the pedestal region being linked with the pressure gradient evolution after pellet injection. The fuelling efficacy of the deposited pellet material is strongly dependent on the E  ×  B plasmoid drift. It is concluded that integrated models like JINTRAC, if validated and supported by realistic physics constraints, may help to establish suitable control schemes of particle and power exhaust in burning ITER DT-plasma scenarios.

Surface thermocouples for measurement of pulsed heat flux in the divertor of the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Brunner, D.; LaBombard, B.

2012-03-01

A novel set of thermocouple sensors has been developed to measure heat fluxes arriving at divertor surfaces in the Alcator C-Mod tokamak, a magnetic confinement fusion experiment. These sensors operate in direct contact with the divertor plasma, which deposits heat fluxes in excess of ˜10 MW/m2 over an ˜1 s pulse. Thermoelectric EMF signals are produced across a non-standard bimetallic junction: a 50 μm thick 74% tungsten-26% rhenium ribbon embedded in a 6.35 mm diameter molybdenum cylinder. The unique coaxial geometry of the sensor combined with its single-point electrical ground contact minimizes interference from the plasma/magnetic environment. Incident heat fluxes are inferred from surface temperature evolution via a 1D thermal heat transport model. For an incident heat flux of 10 MW/m2, surface temperatures rise ˜1000 °C/s, corresponding to a heat flux flowing along the local magnetic field of ˜200 MW/m2. Separate calorimeter sensors are used to independently confirm the derived heat fluxes by comparing total energies deposited during a plasma pulse. Langmuir probes in close proximity to the surface thermocouples are used to test plasma-sheath heat transmission theory and to identify potential sources of discrepancies among physical models.

ADX: a high field, high power density, Advanced Divertor test eXperiment

NASA Astrophysics Data System (ADS)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

2014-10-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

Effects of 2D and 3D Error Fields on the SAS Divertor Magnetic Topology

NASA Astrophysics Data System (ADS)

Trevisan, G. L.; Lao, L. L.; Strait, E. J.; Guo, H. Y.; Wu, W.; Evans, T. E.

2016-10-01

The successful design of plasma-facing components in fusion experiments is of paramount importance in both the operation of future reactors and in the modification of operating machines. Indeed, the Small Angle Slot (SAS) divertor concept, proposed for application on the DIII-D experiment, combines a small incident angle at the plasma strike point with a progressively opening slot, so as to better control heat flux and erosion in high-performance tokamak plasmas. Uncertainty quantification of the error fields expected around the striking point provides additional useful information in both the design and the modeling phases of the new divertor, in part due to the particular geometric requirement of the striking flux surfaces. The presented work involves both 2D and 3D magnetic error field analysis on the SAS strike point carried out using the EFIT code for 2D equilibrium reconstruction, V3POST for vacuum 3D computations and the OMFIT integrated modeling framework for data analysis. An uncertainty in the magnetic probes' signals is found to propagate non-linearly as an uncertainty in the striking point and angle, which can be quantified through statistical analysis to yield robust estimates. Work supported by contracts DE-FG02-95ER54309 and DE-FC02-04ER54698.

Establishment of a low recycling state with full density control by active pumping of the closed helical divertor at LHD

NASA Astrophysics Data System (ADS)

Motojima, G.; Masuzaki, S.; Tanaka, H.; Morisaki, T.; Sakamoto, R.; Murase, T.; Tsuchibushi, Y.; Kobayashi, M.; Schmitz, O.; Shoji, M.; Tokitani, M.; Yamada, H.; Takeiri, Y.; The LHD Experiment Group

2018-01-01

Superior control of particle recycling and hence full governance of plasma density has been established in the Large Helical Device (LHD) using largely enhanced active pumping of the closed helical divertor (CHD). In-vessel cryo-sorption pumping systems inside the CHD in five out of ten inner toroidal divertor sections have been developed and installed step by step in the LHD. The total effective pumping speed obtained was 67  ±  5 m3 s-1 in hydrogen, which is approximately seven times larger than previously obtained. As a result, a low recycling state was observed with CHD pumping for the first time in LHD featuring excellent density control even under intense pellet fueling conditions. A global particle confinement time (τ p* ) is used for comparison of operation with and without the CHD pumping. The τ p* was evaluated from the density decay after the fueling of hydrogen pellet injection or gas puffing in NBI plasmas. A reliably low base density before the fueling and short τ p* after the fueling were obtained during the CHD pumping, demonstrating for the first time full control of the particle balance with active pumping in the CHD.

Electron temperature and heat load measurements in the COMPASS divertor using the new system of probes

NASA Astrophysics Data System (ADS)

Adamek, J.; Seidl, J.; Horacek, J.; Komm, M.; Eich, T.; Panek, R.; Cavalier, J.; Devitre, A.; Peterka, M.; Vondracek, P.; Stöckel, J.; Sestak, D.; Grover, O.; Bilkova, P.; Böhm, P.; Varju, J.; Havranek, A.; Weinzettl, V.; Lovell, J.; Dimitrova, M.; Mitosinkova, K.; Dejarnac, R.; Hron, M.; The COMPASS Team; The EUROfusion MST1 Team

2017-11-01

A new system of probes was recently installed in the divertor of tokamak COMPASS in order to investigate the ELM energy density with high spatial and temporal resolution. The new system consists of two arrays of rooftop-shaped Langmuir probes (LPs) used to measure the floating potential or the ion saturation current density and one array of Ball-pen probes (BPPs) used to measure the plasma potential with a spatial resolution of ~3.5 mm. The combination of floating BPPs and LPs yields the electron temperature with microsecond temporal resolution. We report on the design of the new divertor probe arrays and first results of electron temperature profile measurements in ELMy H-mode and L-mode. We also present comparative measurements of the parallel heat flux using the new probe arrays and fast infrared termography (IR) data during L-mode with excellent agreement between both techniques using a heat power transmission coefficient γ  =  7. The ELM energy density {{\\varepsilon }\\parallel } was measured during a set of NBI assisted ELMy H-mode discharges. The peak values of {{\\varepsilon }\\parallel } were compared with those predicted by model and with experimental data from JET, AUG and MAST with a good agreement.

On the Measurement of Electron Temperature by Single Langmuir Probes in High Recycling Divertors

NASA Astrophysics Data System (ADS)

Pitts, Richard; Horacek, Jan; Loarte, Alberto

2000-10-01

Under high recycling and detached conditions, divertor Langmuir probes often yield a significantly higher value of Te than expected. The influence of plasma turbulence and the effect of fast electrons/plasma collisionality are two reasons why this might occur. We concentrate on these two candidates, with particular reference to observations on the TCV tokamak. A systematic study of the effects of noise on simulated probe characteristics at low T_e, shows that the asymmetric, exponential nature of the characteristic favours electron collection such that fluctuations in Vf alone actually tend to reduce the derived Te from that which would otherwise be found. We have also studied the effects of correlated density and potential fluctuations, finding no effect on the fitted T_e. The sheath potential fall energetically filters electrons such that at high densities, the probe measured Te may be characteristic of hotter, more distant zones in the plasma. We use model parallel field profiles of Te and ne generated from B2-Eirene simulations of TCV discharges as input to the analytic theory of Wesson [1] to show how a divertor plate measurement of Te in TCV can exceed the expected value by factors of up to 6 as detachment is approached. [1] J. A. Wesson, Plasma Phys. and Contr. Fusion 37 (1995) 1459

Overview of recent results and future plans on the Compact Toroidal Hybrid experiment

NASA Astrophysics Data System (ADS)

Maurer, D. A.; Archmiller, M. C.; Cianciosa, M. R.; Ennis, D. A.; Hanson, J. D.; Hartwell, G. J.; Hebert, J. D.; Herfindal, J. L.; Knowlton, S. F.; Ma, X.; Massidda, S.; Pandya, M. D.; Roberds, N. A.; Traverso, P. J.

2015-11-01

Goals of the Compact Toroidal Hybrid (CTH) experiment are to: (1) investigate the dependence of plasma disruptive behavior on the level of applied 3D magnetic shaping, (2) test and advance 3D computational modeling tools in strongly shaped plasmas, and (3) study the implementation of a new island divertor. Progress towards these goals and other developments are summarized. The disruptive density limit is observed to exceed the Greenwald limit as the vacuum transform is increased, but a threshold for disruption avoidance is not observed. Low q operation is routine, with low q disruptions avoided when the vacuum transform is raised to the value of 0.07 or above. Application of vacuum transform has been demonstrated to reduce and eliminate the vertical drift of elongated discharges that would otherwise be vertically unstable. Current efforts at improved equilibrium reconstruction and diagnostic development will beoverviewed. NIMROD is used to model the current ramp phase of CTH and 3D shaped sawtooth behavior. An island divertor design has begun with connection length studies and initial EMC3-Eirene results to model energy deposition on divertor plates located in an edge 1/3 island. This work is supported by U.S. Department of Energy Grant No. DE- FG02-00ER54610.

Concept for a beryllium divertor with in-situ plasma spray surface regeneration

NASA Astrophysics Data System (ADS)

Smith, M. F.; Watson, R. D.; McGrath, R. T.; Croessmann, C. D.; Whitley, J. B.; Causey, R. A.

1990-04-01

Two serious problems with the use of graphite tiles on the ITER divertor are the limited lifetime due to erosion and the difficulty of replacing broken tiles inside the machine. Beryllium is proposed as an alternative low-Z armor material because the plasma spray process can be used to make in-situ repairs of eroded or damaged surfaces. Recent advances in plasma spray technology have produced beryllium coatings of 98% density with a 95% deposition efficiency and strong adhesion to the substrate. With existing technology, the entire active region of the ITER divertor surface could be coated with 2 mm of beryllium in less than 15 h using four small plasma spray guns. Beryllium also has other potential advantages over graphite, e.g., efficient gettering of oxygen, ten times less tritium inventory, reduced problems of transient fueling from D/T exchange and release, no runaway erosion cascades from self-sputtering, better adhesion of redeposited material, as well as higher strength, ductility, and fracture toughness than graphite. A 2-D finite element stress analysis was performed on a 3 mm thick Be tile brazed to an OFHC soft-copper saddle block, which was brazed to a high-strength copper tube. Peak stresses remained 50% below the ultimate strength for both brazing and in-service thermal stresses.

Electron energy distribution function in the divertor region of the COMPASS tokamak during neutral beam injection heating

NASA Astrophysics Data System (ADS)

Hasan, E.; Dimitrova, M.; Havlicek, J.; Mitošinková, K.; Stöckel, J.; Varju, J.; Popov, Tsv K.; Komm, M.; Dejarnac, R.; Hacek, P.; Panek, R.; the COMPASS Team

2018-02-01

This paper presents the results from swept probe measurements in the divertor region of the COMPASS tokamak in D-shaped, L-mode discharges, with toroidal magnetic field BT = 1.15 T, plasma current Ip = 180 kA and line-average electron densities varying from 2 to 8×1019 m-3. Using neutral beam injection heating, the electron energy distribution function is studied before and during the application of the beam. The current-voltage characteristics data are processed using the first-derivative probe technique. This technique allows one to evaluate the plasma potential and the real electron energy distribution function (respectively, the electron temperatures and densities). At the low average electron density of 2×1019 m-3, the electron energy distribution function is bi-Maxwellian with a low-energy electron population with temperatures 4-6 eV and a high-energy electron group 12-25 eV. As the line-average electron density is increased, the electron temperatures decrease. At line-average electron densities above 7×1019 m-3, the electron energy distribution function is found to be Maxwellian with a temperature of 6-8.5 eV. The effect of the neutral beam injection heating power in the divertor region is also studied.

Characteristics of the Footprints of Field Lines for Variable Location of Plate for a Single-Null Divertor Tokamak with the Effects of Dipole Coil Using Method of Maps

NASA Astrophysics Data System (ADS)

Burton, Joni; Ali, Halima; Punjabi, Alkesh

1996-11-01

We determine the properties of the footprint of the magnetic field lines from the stochastic scrape-off layer of a single-null divertor tokamak including the effects of an externally placed dipole coil as the location of the divertor plate is varied. We use the Method of Maps (Punjabi A, Verma A and Boozer A, Phys Rev Lett), 69, 3322 (1992) and J Plasma Phys, 52, 91 (1994) for this investigation. The unperturbed magnetic topology is represented by the Symmetric Simple Map (Ali H, Watson M, Mayer C, Punjabi A and Boozer A, Bull Am Phys Soc), 40, 1855 (1995). The effects of the dipole coil are repesented by the Dipole Map (Ali H, Watson M, Punjabi A and Boozer A, Sherwood Mtg), paper 1C20 (1996). A single dipole coil is placed across from the X-point below the last good surface. The area of the footprint is calculated using the method of fractal dimesion. This work is supported by US DOE OFES. Joni Burton is an undergraduate mathematics major at Hampton University. She is a Ronald E. McNair Scholar at HU supported by R. E. McNair Foundation.

Divertor power load feedback with nitrogen seeding in ASDEX Upgrade

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Dux, R.; Fuchs, J. C.; Fischer, R.; Geiger, B.; Giannone, L.; Herrmann, A.; Lunt, T.; Mertens, V.; McDermott, R.; Neu, R.; Pütterich, T.; Rathgeber, S.; Rohde, V.; Schmid, K.; Schweinzer, J.; Treutterer, W.; ASDEX Upgrade Team

2010-05-01

Feedback control of the divertor power load by means of nitrogen seeding has been developed into a routine operational tool in the all-tungsten clad ASDEX Upgrade tokamak. For heating powers above about 12 MW, its use has become inevitable to protect the divertor tungsten coating under boronized conditions. The use of nitrogen seeding is accompanied by improved energy confinement due to higher core plasma temperatures, which more than compensates the negative effect of plasma dilution by nitrogen on the neutron rate. This paper describes the technical details of the feedback controller. A simple model for its underlying physics allows the prediction of its behaviour and the optimization of the feedback gain coefficients used. Storage and release of nitrogen in tungsten surfaces were found to have substantial impact on the behaviour of the seeded plasma, resulting in increased nitrogen consumption with unloaded walls and a latency of nitrogen release over several discharges after its injection. Nitrogen is released from tungsten plasma facing components with moderate surface temperature in a sputtering-like process; therefore no uncontrolled excursions of the nitrogen wall release are observed. Overall, very stable operation of the high-Z tokamak is possible with nitrogen seeding, where core radiative losses are avoided due to its low atomic charge Z and a high ELM frequency is maintained.

Advanced Plasma Shape Control to Enable High-Performance Divertor Operation on NSTX-U

NASA Astrophysics Data System (ADS)

Vail, Patrick; Kolemen, Egemen; Boyer, Mark; Welander, Anders

2017-10-01

This work presents the development of an advanced framework for control of the global plasma shape and its application to a variety of shape control challenges on NSTX-U. Operations in high-performance plasma scenarios will require highly-accurate and robust control of the plasma poloidal shape to accomplish such tasks as obtaining the strong-shaping required for the avoidance of MHD instabilities and mitigating heat flux through regulation of the divertor magnetic geometry. The new control system employs a high-fidelity model of the toroidal current dynamics in NSTX-U poloidal field coils and conducting structures as well as a first-principles driven calculation of the axisymmetric plasma response. The model-based nature of the control system enables real-time optimization of controller parameters in response to time-varying plasma conditions and control objectives. The new control scheme is shown to enable stable and on-demand plasma operations in complicated magnetic geometries such as the snowflake divertor. A recently-developed code that simulates the nonlinear evolution of the plasma equilibrium is used to demonstrate the capabilities of the designed shape controllers. Plans for future real-time implementations on NSTX-U and elsewhere are also presented. Supported by the US DOE under DE-AC02-09CH11466.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vahala, G.; Tracy, E.

During the past year, the authors have concentrated on (1) divertor physics, (2) thermo-lattice Boltzmann (TLBE) approach to turbulence, and (3) phase space techniques in gyro-resonance problems in collaboration with Dieter Sigmar (MIT), Sergei Krasheninnikov (MIT), Linda Vahala (ODU), Joseph Morrison (AS and M/NASA-Langley), Pavol Pavlo and Josef Preinhaelter (institute of Plasma Physics, Czech Academy of Sciences) and Allan Kaufman (LBL/U.C.Berkeley). Using a 2-equation compressible closure model with a 2D mean flow, the authors are investigating the effects of 3D neutral turbulence on reducing the heat load to the divertor plate by various toroidal cavity geometries. These studies are beingmore » extended to examine 3D mean flows. Thermal Lattice Boltzmann (TLBE) methods are being investigated to handle 3D turbulent flows in nontrivial geometries. It is planned to couple the TLBE collisional regime to the weakly collisional regime and so be able to tackle divertor physics. In the application of phase space techniques to minority-ion RF heating, resonance heating is treated as a multi-stage process. A generalization of the Case-van Kampen analysis is presented for multi-dimensional non-uniform plasmas. Effects such as particle trapping and the ray propagation dynamics in tokamak geometry can now be handled using Weyl calculus.« less

Coupling and Elastic Loading Affect the Active Response by the Inner Ear Hair Cell Bundles

PubMed Central

Strimbu, Clark Elliott; Fredrickson-Hemsing, Lea; Bozovic, Dolores

2012-01-01

Active hair bundle motility has been proposed to underlie the amplification mechanism in the auditory endorgans of non-mammals and in the vestibular systems of all vertebrates, and to constitute a crucial component of cochlear amplification in mammals. We used semi-intact in vitro preparations of the bullfrog sacculus to study the effects of elastic mechanical loading on both natively coupled and freely oscillating hair bundles. For the latter, we attached glass fibers of different stiffness to the stereocilia and observed the induced changes in the spontaneous bundle movement. When driven with sinusoidal deflections, hair bundles displayed phase-locked response indicative of an Arnold Tongue, with the frequency selectivity highest at low amplitudes and decreasing under stronger stimulation. A striking broadening of the mode-locked response was seen with increasing stiffness of the load, until approximate impedance matching, where the phase-locked response remained flat over the physiological range of frequencies. When the otolithic membrane was left intact atop the preparation, the natural loading of the bundles likewise decreased their frequency selectivity with respect to that observed in freely oscillating bundles. To probe for signatures of the active process under natural loading and coupling conditions, we applied transient mechanical stimuli to the otolithic membrane. Following the pulses, the underlying bundles displayed active movement in the opposite direction, analogous to the twitches observed in individual cells. Tracking features in the otolithic membrane indicated that it moved in phase with the bundles. Hence, synchronous active motility evoked in the system of coupled hair bundles by external input is sufficient to displace large overlying structures. PMID:22479461

Prevalence of Split Nerve Fiber Layer Bundles in Healthy People Imaged with Spectral Domain Optical Coherence Tomography.

PubMed

Gür Güngör, Sirel; Akman, Ahmet; Sarıgül Sezenöz, Almila; Tanrıaşıkı, Gülşah

2016-12-01

The presence of retinal nerve fiber layer (RNFL) split bundles was recently described in normal eyes scanned using scanning laser polarimetry and by histologic studies. Split bundles may resemble RNFL loss in healthy eyes. The aim of our study was to determine the prevalence of nerve fiber layer split bundles in healthy people. We imaged 718 eyes of 359 healthy persons with the spectral domain optical coherence tomography in this cross-sectional study. All eyes had intraocular pressure of 21 mmHg or less, normal appearance of the optic nerve head, and normal visual fields (Humphrey Field Analyzer 24-2 full threshold program). In our study, a bundle was defined as 'split' when there is localized defect not resembling a wedge defect in the RNFL deviation map with a symmetrically divided RNFL appearance on the RNFL thickness map. The classification was performed by two independent observers who used an identical set of reference examples to standardize the classification. Inter-observer consensus was reached in all cases. Bilateral superior split bundles were seen in 19 cases (5.29%) and unilateral superior split was observed in 15 cases (4.16%). In 325 cases (90.52%) there was no split bundle. Split nerve fiber layer bundles, in contrast to single nerve fiber layer bundles, are not common findings in healthy eyes. In eyes with normal optic disc appearance, especially when a superior RNFL defect is observed in RNFL deviation map, the RNLF thickness map and graphs should also be examined for split nerve fiber layer bundles.

Computational imaging through a fiber-optic bundle

NASA Astrophysics Data System (ADS)

Lodhi, Muhammad A.; Dumas, John Paul; Pierce, Mark C.; Bajwa, Waheed U.

2017-05-01

Compressive sensing (CS) has proven to be a viable method for reconstructing high-resolution signals using low-resolution measurements. Integrating CS principles into an optical system allows for higher-resolution imaging using lower-resolution sensor arrays. In contrast to prior works on CS-based imaging, our focus in this paper is on imaging through fiber-optic bundles, in which manufacturing constraints limit individual fiber spacing to around 2 μm. This limitation essentially renders fiber-optic bundles as low-resolution sensors with relatively few resolvable points per unit area. These fiber bundles are often used in minimally invasive medical instruments for viewing tissue at macro and microscopic levels. While the compact nature and flexibility of fiber bundles allow for excellent tissue access in-vivo, imaging through fiber bundles does not provide the fine details of tissue features that is demanded in some medical situations. Our hypothesis is that adapting existing CS principles to fiber bundle-based optical systems will overcome the resolution limitation inherent in fiber-bundle imaging. In a previous paper we examined the practical challenges involved in implementing a highly parallel version of the single-pixel camera while focusing on synthetic objects. This paper extends the same architecture for fiber-bundle imaging under incoherent illumination and addresses some practical issues associated with imaging physical objects. Additionally, we model the optical non-idealities in the system to get lower modelling errors.

Monoubiquitination Inhibits the Actin Bundling Activity of Fascin.

PubMed

Lin, Shengchen; Lu, Shuang; Mulaj, Mentor; Fang, Bin; Keeley, Tyler; Wan, Lixin; Hao, Jihui; Muschol, Martin; Sun, Jianwei; Yang, Shengyu

2016-12-30

Fascin is an actin bundling protein that cross-links individual actin filaments into straight, compact, and stiff bundles, which are crucial for the formation of filopodia, stereocillia, and other finger-like membrane protrusions. The dysregulation of fascin has been implicated in cancer metastasis, hearing loss, and blindness. Here we identified monoubiquitination as a novel mechanism that regulates fascin bundling activity and dynamics. The monoubiquitination sites were identified to be Lys 247 and Lys 250 , two residues located in a positive charge patch at the actin binding site 2 of fascin. Using a chemical ubiquitination method, we synthesized chemically monoubiquitinated fascin and determined the effects of monoubiquitination on fascin bundling activity and dynamics. Our data demonstrated that monoubiquitination decreased the fascin bundling EC 50 , delayed the initiation of bundle assembly, and accelerated the disassembly of existing bundles. By analyzing the electrostatic properties on the solvent-accessible surface of fascin, we proposed that monoubiquitination introduced steric hindrance to interfere with the interaction between actin filaments and the positively charged patch at actin binding site 2. We also identified Smurf1 as a E3 ligase regulating the monoubiquitination of fascin. Our findings revealed a previously unidentified regulatory mechanism for fascin, which will have important implications for the understanding of actin bundle regulation under physiological and pathological conditions. © 2016 by The American Society for Biochemistry and Molecular Biology, Inc.

75 FR 51668 - Optional Mail Preparation Standards for Flat-Size Mailpieces in FSS Zones

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-23

... bundles of six or more addressed pieces each, subject to these standards: * * * * * [Revise item b of 13.2... combination. Mailers will then prepare bundles of uniform size from the pieces in the pool. Bundles must be... this option may be applied to the top piece of each bundle, unless otherwise required to be placed on...

Recognizing asymmetry in pseudo-symmetry; structural insights into the interaction between amphipathic α-helices and X-bundle proteins.

PubMed

Haddad, John Faissal; Yang, Yidai; Yeung, Sylvain; Couture, Jean-François

2017-11-01

An α-helix bundle is a small and compact protein fold always composed of more than 2 α-helices that typically run nearly parallel or antiparallel to each other. The repertoire of arrangements of α-helix bundle is such that these domains bind to a myriad of molecular entities including DNA, RNA, proteins and small molecules. A special instance of α-helical bundle is the X-type in which the arrangement of two α-helices interact at 45° to form an X. Among those, some X-helix bundle proteins bind to the hydrophobic section of an amphipathic α-helix in a seemingly orientation and sequence specific manner. In this review, we will compare the binding mode of amphipathic α-helices to X-helix bundle and α-helical bundle proteins. From these structures, we will highlight potential regulatory paradigms that may control the specific interactions of X-helix bundle proteins to amphipathic α-helices. This article is part of a Special Issue entitled: Biophysics in Canada, edited by Lewis Kay, John Baenziger, Albert Berghuis and Peter Tieleman. Copyright © 2017 Elsevier B.V. All rights reserved.

Muscle architecture of the elongated nose in the Asian elephant (Elephas maximus).

PubMed

Endo, H; Hayashi, Y; Komiya, T; Narushima, E; Sasaki, M

2001-05-01

The architecture of the M. caninus in the elongated nose was examined in the Asian elephant (Elephas maximus). The following complicated musculature of the M. caninus was observed in the proximal and distal regions of the nose: (1) Proximal region: In the superficial layer, the longitudinal bundles are confirmed in the dorsal part, and the obliquely-oriented ones in the ventral part. In the middle layer, some bundles run ventro-distally, while other ones represent longitudinally-oriented running. The deep layer consists of complicated architecture of many bundles. Some muscle bundles run medio-laterally, while the others extend proximo-distally in this space. (2) Distal region: In the dorsal part of the M. caninus, the bundles run at deep-superficial direction, while in the ventral part the bundles are longitudinally arranged. The bundles run at lateral direction near the septum of the nasal conduits. The N. facialis and N. infraorbitalis send many branches in the lateral area of the M. caninus in the trunk. This muscle architecture of multi-oriented bundles and well-developed innervation to them suggest that they enable the elongated nose to act as a refined manipulator in the Asian elephant.

The impact of a ventilator bundle on preventing ventilator-associated pneumonia: a multicenter study.

PubMed

Eom, Joong Sik; Lee, Mi-Suk; Chun, Hee-Kyung; Choi, Hee Jung; Jung, Sun-Young; Kim, Yeon-Sook; Yoon, Seon Jin; Kwak, Yee Gyung; Oh, Gang-Bok; Jeon, Min-Hyok; Park, Sun-Young; Koo, Hyun-Sook; Ju, Young-Su; Lee, Jin Seo

2014-01-01

For prevention of ventilator-associated pneumonia (VAP), a bundle approach was applied to patients receiving mechanical ventilation in intensive care units. The incidence of VAP and the preventive efficacy of the VAP bundle were investigated. A quasi-experimental study was conducted in adult intensive care units of 6 university hospitals with similar VAP rates. We implemented the VAP bundle between March 2011 and June 2011, then compared the rate of VAP after implementation of the VAP bundle with the rate in the previous 8 months. Our ventilator bundle included head of bed elevation, peptic ulcer disease prophylaxis, deep venous thrombosis prophylaxis, and oral decontamination with chlorhexidine 0.12%. Continuous aspiration of subglottic secretions was an option. Implementation of the VAP bundle reduced the VAP rate from a mean of 4.08 cases per 1,000 ventilator-days to 1.16 cases per 1,000 ventilator-days. The incidence density ratio (rate) was 0.28 (95% confidence interval, 0.275-0.292). Implementing the appropriate VAP bundle significantly decreased the incidence of VAP in patients with mechanical ventilation. Copyright © 2014 Association for Professionals in Infection Control and Epidemiology, Inc. Published by Mosby, Inc. All rights reserved.

Nanomechanics of Pectin-Linked β-Lactoglobulin Nanofibril Bundles.

PubMed

Loveday, Simon M; Gunning, A Patrick

2018-06-14

Nanofibrils of β-lactoglobulin can be assembled into bundles by site-specific noncovalent cross-linking with high-methoxyl pectin (Hettiarachchi et al. Soft Matter 2016, 12, 756). Here we characterized the nanomechanical properties of bundles using atomic force microscopy and force spectroscopy. Bundles had Gaussian cross sections and a mean height of 17.4 ± 1.4 nm. Persistence lengths were calculated using image analysis with the mean-squared end-to-end model. The relationship between the persistence length and the thickness had exponents of 1.69-2.30, which is consistent with previous reports for other fibril types. In force spectroscopy experiments, the bundles stretched in a qualitatively different manner to fibrils, and some of the force curves were consistent with peeling fibrils away from bundles. The flexibility of pectin-linked nanofibril bundles is likely to be tunable by modulating the stiffness and length of fibrils and the ratio of pectin to fibrils, giving rise to a wide range of structures and functionalities.

Tokyo Guidelines 2018: management bundles for acute cholangitis and cholecystitis.

PubMed

Mayumi, Toshihiko; Okamoto, Kohji; Takada, Tadahiro; Strasberg, Steven M; Solomkin, Joseph S; Schlossberg, David; Pitt, Henry A; Yoshida, Masahiro; Gomi, Harumi; Miura, Fumihiko; Garden, O James; Kiriyama, Seiki; Yokoe, Masamichi; Endo, Itaru; Asbun, Horacio J; Iwashita, Yukio; Hibi, Taizo; Umezawa, Akiko; Suzuki, Kenji; Itoi, Takao; Hata, Jiro; Han, Ho-Seong; Hwang, Tsann-Long; Dervenis, Christos; Asai, Koji; Mori, Yasuhisa; Huang, Wayne Shih-Wei; Belli, Giulio; Mukai, Shuntaro; Jagannath, Palepu; Cherqui, Daniel; Kozaka, Kazuto; Baron, Todd H; de Santibañes, Eduardo; Higuchi, Ryota; Wada, Keita; Gouma, Dirk J; Deziel, Daniel J; Liau, Kui-Hin; Wakabayashi, Go; Padbury, Robert; Jonas, Eduard; Supe, Avinash Nivritti; Singh, Harjit; Gabata, Toshifumi; Chan, Angus C W; Lau, Wan Yee; Fan, Sheung Tat; Chen, Miin-Fu; Ker, Chen-Guo; Yoon, Yoo-Seok; Choi, In-Seok; Kim, Myung-Hwan; Yoon, Dong-Sup; Kitano, Seigo; Inomata, Masafumi; Hirata, Koichi; Inui, Kazuo; Sumiyama, Yoshinobu; Yamamoto, Masakazu

2018-01-01

Management bundles that define items or procedures strongly recommended in clinical practice have been used in many guidelines in recent years. Application of these bundles facilitates the adaptation of guidelines and helps improve the prognosis of target diseases. In Tokyo Guidelines 2013 (TG13), we proposed management bundles for acute cholangitis and cholecystitis. Here, in Tokyo Guidelines 2018 (TG18), we redefine the management bundles for acute cholangitis and cholecystitis. Critical parts of the bundles in TG18 include the diagnostic process, severity assessment, transfer of patients if necessary, and therapeutic approach at each time point. Observance of these items and procedures should improve the prognosis of acute cholangitis and cholecystitis. Studies are now needed to evaluate the dissemination of these TG18 bundles and their effectiveness. Free full articles and mobile app of TG18 are available at: http://www.jshbps.jp/modules/en/index.php?content_id=47. Related clinical questions and references are also included. © 2017 Japanese Society of Hepato-Biliary-Pancreatic Surgery.

Hair-bundle proteomes of avian and mammalian inner-ear utricles

PubMed Central

Wilmarth, Phillip A.; Krey, Jocelyn F.; Shin, Jung-Bum; Choi, Dongseok; David, Larry L.; Barr-Gillespie, Peter G.

2015-01-01

Examination of multiple proteomics datasets within or between species increases the reliability of protein identification. We report here proteomes of inner-ear hair bundles from three species (chick, mouse, and rat), which were collected on LTQ or LTQ Velos ion-trap mass spectrometers; the constituent proteins were quantified using MS2 intensities, which are the summed intensities of all peptide fragmentation spectra matched to a protein. The data are available via ProteomeXchange with identifiers PXD002410 (chick LTQ), PXD002414 (chick Velos), PXD002415 (mouse Velos), and PXD002416 (rat LTQ). The two chick bundle datasets compared favourably to a third, already-described chick bundle dataset, which was quantified using MS1 peak intensities, the summed intensities of peptides identified by high-resolution mass spectrometry (PXD000104; updated analysis in PXD002445). The mouse bundle dataset described here was comparable to a different mouse bundle dataset quantified using MS1 intensities (PXD002167). These six datasets will be useful for identifying the core proteome of vestibular hair bundles. PMID:26645194

Reduction of Net Erosion of High-Z Divertor Surface by Local Redeposition in DIII-D

NASA Astrophysics Data System (ADS)

Stangeby, P. C.

2012-10-01

Utilizing the unique capability to expose material samples to well characterized diverted plasmas, recent DIII-D measurements have confirmed theoretical expectations of the relative net and gross erosion rates of molybdenum in the divertor region. Knowledge of these erosion rates is important for predicting first wall lifetime in future fusion devices. Theory suggests that the net erosion rate will be much less than gross erosion due to prompt local deposition of eroded ions by gyro-orbit motion, the strong E-field toward the target and friction with the fast plasma flow toward the target. However, experimental evidence to date has been contradictory. The results here, which are the most definitive to date, are consistent with the basic theoretical predictions. The net and gross erosion rates were measured utilizing 1-cm and 1-mm diameter Mo samples that are mounted on the DIII-D Divertor Material Evaluation System (DiMES) system and simultaneously exposed near the attached outer strike point of an L-mode plasma for 4 s. Due to the spatial extent of the re-deposition, the larger sample gives the net erosion while the smaller sample is indicative of the gross erosion. Post-mortem ion beam analysis (RBS) of the larger sample, indicates a 2.9 nm film thickness reduction (or 0.72 nm/s net erosion rate). Similar analysis of the smaller sample yields a 1.3 nm/s gross erosion rate, consistent with spectroscopic measurements of Mo I emission. The net to gross erosion ratio of 0.56 is consistent with calculations using a modeling package including REDEP/WBS and OEDGE codes. Using as input the measured plasma density and temperature profiles from divertor Langmuir probes, these codes estimate a net to gross erosion ratio of 0.46. Details of the modeling and implications for future devices will be discussed.

Improved confinement in highly powered high performance scenarios on DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Petrie, Thomas W.; Osborne, Thomas; Fenstermacher, Max E.

DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95 = 6, P IN up to 15 MW, H 98 = 1.4–1.8, and β N = 2.5–4.0. The ion B xmore » $$\\triangledown$$B direction is away from the primary X-point. While plasma conditions at lower to moderate power input (e.g., 11 MW) are shown to be favorable to successful puff-and-pump radiating divertor applications, particularly when using argon seeds, plasma behavior at higher powers (e.g., ≥14 MW) may make successful puff-and-pump operation more problematic. In contrast to lower powered high performance plasmas, both $$\\tau$$ E and β N in the high power cases (≥14 MW) increased and ELM frequency decreased, as density was raised by deuterium gas injection. Improved performance in the higher power plasmas was tied to higher pedestal pressure, which according to peeling-ballooning mode stability analysis using the ELITE code could increase with density along the kink/peeling stability threshold, while the pedestal pressure gradient in the lower power discharges were limited by the ballooning threshold. This resulted in improved fueling efficiency and ≈10% higher $$\\tau$$ E and β N than is normally observed in comparable high performance plasmas on DIII-D. Applying the puff-and-pump radiating divertor approach at moderate versus high power input is shown to result in a much different evolution in core and pedestal plasma behavior. In conclusion, we find that injecting deuterium gas into these highly powered DND plasmas may open up a new avenue for achieving elevated plasma performance, including better fueling, but the resulting higher density may also complicate application of a radiating divertor approach to heat flux reduction in present-day tokamaks, if scenarios involving second-harmonic electron cyclotron heating are used.« less

Improved confinement in highly powered high performance scenarios on DIII-D

DOE PAGES

Petrie, Thomas W.; Osborne, Thomas; Fenstermacher, Max E.; ...

2017-06-09

DIII-D has recently demonstrated improved energy confinement by injecting neutral deuterium gas into high performance near-double null divertor (DND) plasmas during high power operation. Representative parameters for these plasmas are: q 95 = 6, P IN up to 15 MW, H 98 = 1.4–1.8, and β N = 2.5–4.0. The ion B xmore » $$\\triangledown$$B direction is away from the primary X-point. While plasma conditions at lower to moderate power input (e.g., 11 MW) are shown to be favorable to successful puff-and-pump radiating divertor applications, particularly when using argon seeds, plasma behavior at higher powers (e.g., ≥14 MW) may make successful puff-and-pump operation more problematic. In contrast to lower powered high performance plasmas, both $$\\tau$$ E and β N in the high power cases (≥14 MW) increased and ELM frequency decreased, as density was raised by deuterium gas injection. Improved performance in the higher power plasmas was tied to higher pedestal pressure, which according to peeling-ballooning mode stability analysis using the ELITE code could increase with density along the kink/peeling stability threshold, while the pedestal pressure gradient in the lower power discharges were limited by the ballooning threshold. This resulted in improved fueling efficiency and ≈10% higher $$\\tau$$ E and β N than is normally observed in comparable high performance plasmas on DIII-D. Applying the puff-and-pump radiating divertor approach at moderate versus high power input is shown to result in a much different evolution in core and pedestal plasma behavior. In conclusion, we find that injecting deuterium gas into these highly powered DND plasmas may open up a new avenue for achieving elevated plasma performance, including better fueling, but the resulting higher density may also complicate application of a radiating divertor approach to heat flux reduction in present-day tokamaks, if scenarios involving second-harmonic electron cyclotron heating are used.« less

Upward-facing Lithium Flash Evaporator for NSTX-U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roquemore, A. L.

2013-07-09

NSTX plasma performance has been significantly enhanced by lithium conditioning [1]. To date, the lower divertor and passive plates have been conditioned by downward facing lithium evaporators (LITER) as appropriate for lower null plasmas. The higher power operation expected from NSTX-U requires double null plasma operation in order to distribute the heat flux between the upper and lower divertors making it desirable to coat the upper divertor region with Li as well. An upward aiming LITER (U-LITER) is presently under development and will be inserted into NSTX-U using a horizontal probe drive located in a 6" upper midplane port. Inmore » the retracted position the evaporator will be loaded with up to 300 mg of Li granules utilizing one of the calibrated NSTX Li powder droppers[2]. The evaporator will then be inserted into the vessel in a location within the shadow of the RF limiters and will remain in the vessel during the discharge. About 10 seconds before a discharge, it will be rapidly heated and the lithium completely evaporated onto the upper divertor, thus avoiding the complication of a shutter that prevents evaporation during the shot when the diagnostic shutters are open. The minimal time interval between the evaporation and the start of the discharge will avoid the passivation of the lithium by residual gases and enable the study of the conditioning effects of un-passivated Li surfaces [3]. Two methods are being investigated to accomplish the rapid (few second) heating of the lithium. A resistive method relies on passing a large current through a Li filled crucible. A second method requires using a 3 kW e-beam gun to heat the Li. In this paper the evaporator systems will be described and the pros and cons of each heating method will be discussed.« less

Non-boronized compared with boronized operation of ASDEX Upgrade with full-tungsten plasma facing components

NASA Astrophysics Data System (ADS)

Kallenbach, A.; Dux, R.; Mayer, M.; Neu, R.; Pütterich, T.; Bobkov, V.; Fuchs, J. C.; Eich, T.; Giannone, L.; Gruber, O.; Herrmann, A.; Horton, L. D.; Maggi, C. F.; Meister, H.; Müller, H. W.; Rohde, V.; Sips, A.; Stäbler, A.; Stober, J.; ASDEX Upgrade Team

2009-04-01

After completion of the tungsten coating of all plasma facing components, ASDEX Upgrade has been operated without boronization for 1 1/2 experimental campaigns. This has allowed the study of fuel retention under conditions of relatively low D co-deposition with low-Z impurities as well as the operational space of a full-tungsten device for the unfavourable condition of a relatively high intrinsic impurity level. Restrictions in operation were caused by the central accumulation of tungsten in combination with density peaking, resulting in H-L backtransitions induced by too low separatrix power flux. Most important control parameters have been found to be the central heating power, as delivered predominantly by ECRH, and the ELM frequency, most easily controlled by gas puffing. Generally, ELMs exhibit a positive impact, with the effect of impurity flushing out of the pedestal region overbalancing the ELM-induced W source. The restrictions of plasma operation in the unboronized W machine occurred predominantly under low or medium power conditions. Under medium-high power conditions, stable operation with virtually no difference between boronized and unboronized discharges was achieved. Due to the reduced intrinsic radiation with boronization and the limited power handling capability of VPS coated divertor tiles (≈10 MW m-2), boronized operation at high heating powers was possible only with radiative cooling. To enable this, a previously developed feedback system using (thermo-)electric current measurements as approximate sensor for the divertor power flux was introduced into the standard AUG operation. To avoid the problems with reduced ELM frequency due to core plasma radiation, nitrogen was selected as radiating species since its radiative characteristic peaks at lower electron temperatures in comparison with Ne and Ar, favouring SOL and divertor radiative losses. Nitrogen seeding resulted not only in the desired divertor power load reduction but also in improved energy confinement, as well as in smaller ELMs.

Experimental results of near real-time protection system for plasma facing components in Wendelstein 7-X at GLADIS

NASA Astrophysics Data System (ADS)

Ali, A.; Jakubowski, M.; Greuner, H.; Böswirth, B.; Moncada, V.; Sitjes, A. Puig; Neu, R.; Pedersen, T. S.; the W7-X Team

2017-12-01

One of the aims of stellarator Wendelstein 7-X (W7-X), is to investigate steady state operation, for which power exhaust is an important issue. The predominant fraction of the energy lost from the confined plasma region will be absorbed by an island divertors, which is designed for 10 {{MWm}}-2 steady state operation. In order to protect the divertor targets from overheating, 10 state-of-the-art infrared endoscopes will be installed at W7-X. In this work, we present the experimental results obtained at the high heat flux test facility GLADIS (Garching LArge DIvertor Sample test facility in IPP Garching) [1] during tests of a new plasma facing components (PFCs) protection algorithm designed for W7-X. The GLADIS device is equipped with two ion beams that can generate a heat load in the range from 3 MWm-2 to 55 MWm-2. The algorithms developed at W7-X to detect defects and hot spots are based on the analysis of surface temperature evolution and are adapted to work in near real-time. The aim of this work was to test the near real-time algorithms in conditions close to those expected in W7-X. The experiments were performed on W7-X pre-series tiles to detect CFC/Cu delaminations. For detection of surface layers, carbon fiber composite (CFC) blocks from the divertor of the Wendelstein 7-AS stellarator were used to observe temporal behavior of fully developed surface layers. These layers of re-deposited materials, like carbon, boron, oxygen and iron, were formed during the W7-AS operation. A detailed analysis of the composition and their thermal response to high heat fluxes (HHF) are described in [2]. The experiments indicate that the automatic detection of critical events works according to W7-X PFC protection requirements.

Ab initio density functional theory investigation of structural and electronic properties of silicon carbide nanotube bundles

NASA Astrophysics Data System (ADS)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-10-01

By using ab initio density functional theory the structural and electronic properties of isolated and bundled (8,0) and (6,6) silicon carbide nanotubes (SiCNTs) are investigated. Our results show that for such small diameter nanotubes the inter-tube interaction causes a very small radial deformation, while band splitting and reduction of the semiconducting energy band gap are significant. We compared the equilibrium interaction energy and inter-tube separation distance of (8,0) SiCNT bundle with (10,0) carbon nanotube (CNT) bundle where they have the same radius. We found that there is a larger inter-tube separation and weaker inter-tube interaction in the (8,0) SiCNT bundle with respect to (10,0) CNT bundle, although they have the same radius.

A collagen and elastic network in the wing of the bat.

PubMed

Holbrook, K A; Odland, G F

1978-05-01

Bundles of collagen fibrils, elastic fibres and fibroblasts are organized into a network that lies in the plane of a large portion of the bat wing. By ultrastructural (TEM and SEM) and biochemical analyses it was found that individual bundles of the net are similar to elastic ligaments. Although elastic fibres predominate, they are integrated and aligned in parallel with small bundles of collagen. A reticulum of fibroblasts, joined by focal junctions, forms a cellular framework throughout each bundle. Because of the unique features of the fibre bundles of the bat's wing, in particular their accessibility, and the parallel alignment of the collagen fibrils and elastic fibres in each easily isolatable fibre bundle, they should prove a most valuable model for connective tissue studies, particularly for the study of collagen-elastin interactions.

Analysis of the Proteome of Hair-Cell Stereocilia by Mass Spectrometry

PubMed Central

Krey, Jocelyn F.; Wilmarth, Philip A.; David, Larry L.; Barr-Gillespie, Peter G.

2017-01-01

Characterization of proteins that mediate mechanotransduction by hair cells, the sensory cells of the inner ear, is hampered by the scarcity of these cells and their sensory organelle, the hair bundle. Mass spectrometry, with its high sensitivity and identification precision, is the ideal method for determining which proteins are present in bundles and what proteins they interact with. We describe here the isolation of mouse hair bundles, as well as preparation of bundle-protein samples for mass spectrometry. We also describe protocols for data-dependent (shotgun) and parallel-reaction-monitoring (targeted) mass spectrometry that allow us to identify and quantify proteins of the hair bundle. These sensitive methods are particularly useful for comparing proteomes of wild-type and mice with deafness mutations affecting hair-bundle proteins. (120 words; maximum 250) PMID:28109437

PDS4 Bundle Creation Governance Using BPMN

NASA Astrophysics Data System (ADS)

Radulescu, C.; Levoe, S. R.; Algermissen, S. S.; Rye, E. D.; Hardman, S. H.

2015-06-01

The AMMOS-PDS Pipeline Service (APPS) provides a Bundle Builder tool, which governs the process of creating, and ultimately generates, PDS4 bundles incrementally, as science products are being generated.

Bundle Branch Block

MedlinePlus

... known cause. Causes can include: Left bundle branch block Heart attacks (myocardial infarction) Thickened, stiffened or weakened ... myocarditis) High blood pressure (hypertension) Right bundle branch block A heart abnormality that's present at birth (congenital) — ...

Deformation quantization with separation of variables of an endomorphism bundle

NASA Astrophysics Data System (ADS)

Karabegov, Alexander

2014-01-01

Given a holomorphic Hermitian vector bundle E and a star-product with separation of variables on a pseudo-Kähler manifold, we construct a star product on the sections of the endomorphism bundle of the dual bundle E∗ which also has the appropriately generalized property of separation of variables. For this star product we prove a generalization of Gammelgaard's graph-theoretic formula.

Effect of fabricated density and bamboo species on physical-mechanical properties of bamboo fiber bundle reinforced composites

Treesearch

Jiulong Xie; Jinqiu Qi; Tingxing Hu; Cornelis F. De Hoop; Chung Yun Hse; Todd F. Shupe

2016-01-01

Bamboo stems were subjected to a mechanical treatment process for the extraction of bamboo fiber bundles. The fiber bundles were used as reinforcement for the fabrication of high-performance composites with phenolic resins as matrix. The influence of fabricated density and bamboo species on physical–mechanical properties of bamboo fiber bundle reinforced composites (...

ACL double-bundle reconstruction with one tibial tunnel provides equal stability compared to two tibial tunnels.

PubMed

Drews, Björn Holger; Seitz, Andreas Martin; Huth, Jochen; Bauer, Gerhard; Ignatius, Anita; Dürselen, Lutz

2017-05-01

The purpose of this study was to investigate whether an anterior cruciate ligament (ACL) double-bundle reconstruction with one tibial tunnel displays the same in vitro stability as a conventional double-bundle reconstruction with two tibial tunnels when using the same tensioning protocol. In 11 fresh-frozen cadaveric knees, ACL double-bundle reconstruction with one and two tibial tunnels was performed. The two grafts were tightened using 80 N in different flexion angles (anteromedial-bundle at 60° and posterolateral-bundle at 15°). Anterior tibial translation (134 N) and translation with combined rotatory and valgus loads (10 Nm valgus stress and 4 Nm internal tibial torque) were determined at 0°, 30°, 60° and 90° flexion. Measurements were taken in intact ACL, resected ACL, three-tunnel reconstruction and four-tunnel reconstruction. Additionally, the tension on the grafts was determined. Student's t test was performed for statistical analysis of the related samples. Significance was set at p < 0.017 according to Bonferroni correction. The two reconstructive techniques displayed no significant differences in comparison with the intact ACL in anterior tibial translation at 0°, 60° and 90° of flexion. The same results were obtained for the anterior tibial translation with a combined rotatory load at 60° and 90°. When directly comparing both reconstructive techniques, there were no significant differences for the anterior tibial translation and combined rotatory load at all flexion angles. The measured tension on grafts displayed similar load sharing between both bundles. Except at full extension, both grafts displayed a significantly different tension increase under anterior tibial translation for both techniques (p = 0.0086). Tightening both bundles in ACL double-bundle reconstruction with one or two tibial tunnels in different flexion angles achieved comparable restoration of stability, although there was different load sharing on the bundles. With regard to individualized ACL reconstruction, the double-bundle technique with one tibial tunnel offers a possibility to address small tibial insertion sites without compromising the advantages of a double-bundle procedure.

Femoral tunnel placement in single-bundle anterior cruciate ligament reconstruction: a cadaveric study relating transtibial lateralized femoral tunnel position to the anteromedial and posterolateral bundle femoral origins of the anterior cruciate ligament.

PubMed

Rue, John-Paul H; Ghodadra, Neil; Bach, Bernard R

2008-01-01

There is controversy regarding the necessity of reconstructing both the posterolateral and anteromedial bundles of the anterior cruciate ligament. A laterally oriented transtibial drilled femoral tunnel replaces portions of the femoral footprints of the anteromedial and posterolateral bundles of the anterior cruciate ligament. Descriptive laboratory study. Footprints of the anteromedial and posterolateral bundles of the anterior cruciate ligament were preserved on 7 matched pairs (5 female, 2 male) of fresh-frozen human cadaveric femurs (14 femurs total). Each femur was anatomically oriented and secured in a custom size-appropriate, side-matched replica tibia model to simulate transtibial retrograde drilling of a 10-mm femoral tunnel in each specimen. The relationship of the tunnel relative to footprints of both bundles of the anterior cruciate ligament was recorded using a Microscribe MX digitizer. The angle of the femoral tunnel relative to the vertical 12-o'clock position was recorded for all 14 specimens; only 10 specimens were used for footprint measurements. On average, the 10-mm femoral tunnel overlapped 50% of the anteromedial bundle (range, 2%-83%) and 51% of the posterolateral bundle (range, 16%-97%). The footprint of the anteromedial bundle occupied 32% (range, 3%-49%) of the area of the tunnel; the footprint of the posterolateral bundle contributed 26% (range, 7%-41%). The remainder of the area of the 10-mm tunnel did not overlap with the anterior cruciate ligament footprint. The mean absolute angle of the femoral tunnel as measured directly on the specimen was 48 degrees (range, 42 degrees-53 degrees) from vertical, corresponding to approximately a 10:30 clock face position on a right knee. Anterior cruciate ligament reconstruction using a laterally oriented transtibial drilled femoral tunnel incorporates portions of the anteromedial and posterolateral bundle origins of the native anterior cruciate ligament. A laterally oriented transtibial drilled femoral tunnel placed at the 10:30 position (1:30 for left knees) reconstructs portions of the anteromedial and posterolateral bundles of the anterior cruciate ligament.

Bundling of elastic filaments induced by hydrodynamic interactions

NASA Astrophysics Data System (ADS)

Man, Yi; Page, William; Poole, Robert J.; Lauga, Eric

2017-12-01

Peritrichous bacteria swim in viscous fluids by rotating multiple helical flagellar filaments. As the bacterium swims forward, all its flagella rotate in synchrony behind the cell in a tight helical bundle. When the bacterium changes its direction, the flagellar filaments unbundle and randomly reorient the cell for a short period of time before returning to their bundled state and resuming swimming. This rapid bundling and unbundling is, at its heart, a mechanical process whereby hydrodynamic interactions balance with elasticity to determine the time-varying deformation of the filaments. Inspired by this biophysical problem, we present in this paper what is perhaps the simplest model of bundling whereby two or more straight elastic filaments immersed in a viscous fluid rotate about their centerline, inducing rotational flows which tend to bend the filaments around each other. We derive an integrodifferential equation governing the shape of the filaments resulting from mechanical balance in a viscous fluid at low Reynolds number. We show that such equation may be evaluated asymptotically analytically in the long-wavelength limit, leading to a local partial differential equation governed by a single dimensionless bundling number. A numerical study of the dynamics predicted by the model reveals the presence of two configuration instabilities with increasing bundling numbers: first to a crossing state where filaments touch at one point and then to a bundled state where filaments wrap along each other in a helical fashion. We also consider the case of multiple filaments and the unbundling dynamics. We next provide an intuitive physical model for the crossing instability and show that it may be used to predict analytically its threshold and adapted to address the transition to a bundling state. We then use a macroscale experimental implementation of the two-filament configuration in order to validate our theoretical predictions and obtain excellent agreement. This long-wavelength model of bundling will be applicable to other problems in biological physics and provides the groundwork for further, more realistic, models of flagellar bundling.

Bundling in Place: Translating the NGSS into Place-Based Earth-System Science Curricula

NASA Astrophysics Data System (ADS)

Semken, S. C.

2016-12-01

Bundling is the process of grouping Performance Expectations (PEs) from the Next Generation Science Standards (NGSS) into coherent units based on a defined topic, idea, question, or phenomenon. Bundling sorts the PEs for a given grade or grade band into a teachable narrative: a key stage in building curriculum, instruction, and assessment from the NGSS. To encourage and facilitate this, bundling guidelines have recently been released on the NGSS website (nextgenscience.org/glossary/bundlesbundling), and example bundles for different grade bands and disciplines are also being developed and posted there. According to these guidelines the iterative process of bundling begins with organization of PEs according to natural connections among them, and alignment of the three NGSS dimensions (Disciplinary Core Ideas, Cross-Cutting Concepts, and Science and Engineering Practices) that underpin each PE. Bundles are grouped by coherence and increasing complexity into courses, and courses into course sets that should encompass all PEs for a grade band. Bundling offers a natural way to translate the NGSS into highly contextualized curricula such as place-based (PB) teaching, which is situated in specific places or regions and focused on natural and cultural features, processes, phenomena, history, and challenges to sustainability therein. Attributes of place and our individual and collective connections to place (sense of place) directly inform PB curriculum, pedagogy, and assessment. PEs can be bundled by their relevance to these themes. Following the NGSS guidelines, I model the process for PB instruction by bundling PEs around the themes of Paleozoic geology and carbonate deposition and their relationships to mining and calcining of limestone in Anthropocene cement production for developing communities. The bundles integrate aspects of Earth history, the carbon cycle, mineral resources, climate change, and sustainability using specific local examples and narratives. They are designed for a hypothetical place-based high-school Earth-science course situated in the Greater American Southwest, but could be readily modified for another region with similar geology and resource use.

Effectiveness of a Model Bundle Payment Initiative for Femur Fracture Patients.

PubMed

Lott, Ariana; Belayneh, Rebekah; Haglin, Jack; Konda, Sanjit; Egol, Kenneth A

2018-05-28

Analyze the effectiveness of a BPCI (Bundle Payments for Care Improvement) initiative for patients who would be included in a future potential Surgical Hip and Femur Fracture Treatment (SHFFT) bundle. Retrospective cohort SETTING:: Single Academic Institution PATIENTS/PARTICIPANTS:: Patients discharged with operative fixation of a hip or femur fracture (DRG codes 480-482) between 1/2015-10/2016 were included. A BPCI initiative based upon an established program for BPCI Total Joint Arthroplasty (TJA) was initiated for patients with hip and femur fractures in January 2016. Patients were divided into non-bundle (care before initiative) and bundle (care with initiative) cohorts. Application of BPCI principles MAIN OUTCOME MEASURES:: Length of stay, location of discharge, readmissions RESULTS:: 116 patients participated in the "institutional bundle," and 126 received care prior to the initiative. There was a trend towards decreased mean length of stay, (7.3 ± 6.3 days vs. 6.8 ± 4.0 days, p=0.457) and decreased readmission within 90 days (22.2% vs. 18.1%, p=0.426). The number of patients discharged home doubled (30.2% vs. 14.3%, p=0.008). There was no difference in readmission rates in bundle vs. non-bundle patients based on discharged home status; however, bundle patients discharged to SNF trended towards less readmissions than non-bundle patients discharged to SNF (37.3% vs. 50.6%, p=0.402). Mean episode cost reduction due to initiative was estimated to be $6,450 using Medicare reimbursement data. This study demonstrates the potential success of a BPCI initiative at one institution in decreasing post-acute care facility utilization and cost of care when used for a hip and femur fracture population. Therapeutic Level IV. See Instructions for Authors for a complete description of levels of evidence.

* Hierarchically Structured Electrospun Scaffolds with Chemically Conjugated Growth Factor for Ligament Tissue Engineering.

PubMed

Pauly, Hannah M; Sathy, Binulal N; Olvera, Dinorath; McCarthy, Helen O; Kelly, Daniel J; Popat, Ketul C; Dunne, Nicholas J; Haut Donahue, Tammy Lynn

2017-08-01

The anterior cruciate ligament (ACL) of the knee is vital for proper joint function and is commonly ruptured during sports injuries or car accidents. Due to a lack of intrinsic healing capacity and drawbacks with allografts and autografts, there is a need for a tissue-engineered ACL replacement. Our group has previously used aligned sheets of electrospun polycaprolactone nanofibers to develop solid cylindrical bundles of longitudinally aligned nanofibers. We have shown that these nanofiber bundles support cell proliferation and elongation and the hierarchical structure and material properties are similar to the native human ACL. It is possible to combine multiple nanofiber bundles to create a scaffold that attempts to mimic the macroscale structure of the ACL. The goal of this work was to develop a hierarchical bioactive scaffold for ligament tissue engineering using connective tissue growth factor (CTGF)-conjugated nanofiber bundles and evaluate the behavior of mesenchymal stem cells (MSCs) on these scaffolds in vitro and in vivo. CTGF was immobilized onto the surface of individual nanofiber bundles or scaffolds consisting of multiple nanofiber bundles. The conjugation efficiency and the release of conjugated CTGF were assessed using X-ray photoelectron spectroscopy, assays, and immunofluorescence staining. Scaffolds were seeded with MSCs and maintained in vitro for 7 days (individual nanofiber bundles), in vitro for 21 days (scaled-up scaffolds of 20 nanofiber bundles), or in vivo for 6 weeks (small scaffolds of 4 nanofiber bundles), and ligament-specific tissue formation was assessed in comparison to non-CTGF-conjugated control scaffolds. Results showed that CTGF conjugation encouraged cell proliferation and ligament-specific tissue formation in vitro and in vivo. The results suggest that hierarchical electrospun nanofiber bundles conjugated with CTGF are a scalable and bioactive scaffold for ACL tissue engineering.

Hexagonally Ordered Arrays of α-Helical Bundles Formed from Peptide-Dendron Hybrids

DOE Office of Scientific and Technical Information (OSTI.GOV)

Barkley, Deborah A.; Rokhlenko, Yekaterina; Marine, Jeannette E.

Combining monodisperse building blocks that have distinct folding properties serves as a modular strategy for controlling structural complexity in hierarchically organized materials. We combine an α-helical bundle-forming peptide with self-assembling dendrons to better control the arrangement of functional groups within cylindrical nanostructures. Site-specific grafting of dendrons to amino acid residues on the exterior of the α-helical bundle yields monodisperse macromolecules with programmable folding and self-assembly properties. The resulting hybrid biomaterials form thermotropic columnar hexagonal mesophases in which the peptides adopt an α-helical conformation. Bundling of the α-helical peptides accompanies self-assembly of the peptide-dendron hybrids into cylindrical nanostructures. The bundle stoichiometrymore » in the mesophase agrees well with the size found in solution for α-helical bundles of peptides with a similar amino acid sequence.« less

Method and apparatus for extracting tritium and preparing radioactive waste for disposal

DOEpatents

Heung, Leung K.

1994-01-01

Apparatus for heating an object such as a nuclear target bundle to release and recover hydrogen and contain the disposable residue for disposal. The apparatus comprises an inverted furnace, a sleeve/crucible assembly for holding and enclosing the bundle, conveying equipment for placing the sleeve onto the crucible and loading the bundle into the sleeve/crucible, a lift for raising the enclosed bundle into the furnace, and hydrogen recovery equipment including a trap and strippers, all housed in a containment having negative internal pressure. The crucible/sleeve assembly has an internal volume that is sufficient to enclose and hold the bundle before heating; the crucible's internal volume is sufficient by itself to hold and enclose the bundle's volume after heating. The crucible can then be covered and disposed of; the sleeve, on the other hand, can be reused.

Spinor Geometry and Signal Transmission in Three-Space

NASA Astrophysics Data System (ADS)

Binz, Ernst; Pods, Sonja; Schempp, Walter

2002-09-01

For a singularity free gradient field in an open set of an oriented Euclidean space of dimension three we define a natural principal bundle out of an immanent complex line bundle. The elements of both bundles are called internal variables. Several other natural bundles are associated with the principal bundle and, in turn, determine the vector field. Two examples are given and it is shown that for a constant vector field circular polarized waves travelling along a field line can be considered as waves of internal variables. Einstein's equation epsilon = m [middle dot] c2 is derived from the geometry of the principal bundle. On SU(2) a relation between spin representations and Schrodinger representations is established. The link between the spin 1/2-model and the Schrodinger representations yields a connection between a microscopic and a macroscopic viewpoint.

System and method for reducing combustion dynamics in a combustor

DOEpatents

Uhm, Jong Ho; Johnson, Thomas Edward; Zuo, Baifang; York, William David

2015-09-01

A system for reducing combustion dynamics in a combustor includes an end cap having an upstream surface axially separated from a downstream surface, and tube bundles extend from the upstream surface through the downstream surface. A divider inside a tube bundle defines a diluent passage that extends axially through the downstream surface, and a diluent supply in fluid communication with the divider provides diluent flow to the diluent passage. A method for reducing combustion dynamics in a combustor includes flowing a fuel through tube bundles, flowing a diluent through a diluent passage inside a tube bundle, wherein the diluent passage extends axially through at least a portion of the end cap into a combustion chamber, and forming a diluent barrier in the combustion chamber between the tube bundle and at least one other adjacent tube bundle.

Fuel transfer system

DOEpatents

Townsend, Harold E.; Barbanti, Giancarlo

1994-01-01

A nuclear fuel bundle fuel transfer system includes a transfer pool containing water at a level above a reactor core. A fuel transfer machine therein includes a carriage disposed in the transfer pool and under the water for transporting fuel bundles. The carriage is selectively movable through the water in the transfer pool and individual fuel bundles are carried vertically in the carriage. In a preferred embodiment, a first movable bridge is disposed over an upper pool containing the reactor core, and a second movable bridge is disposed over a fuel storage pool, with the transfer pool being disposed therebetween. A fuel bundle may be moved by the first bridge from the reactor core and loaded into the carriage which transports the fuel bundle to the second bridge which picks up the fuel bundle and carries it to the fuel storage pool.