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

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Dissipative divertor operation in the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

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

1995-01-01

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

The DIII-D Radiative Divertor Project: Status and plans

International Nuclear Information System (INIS)

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

1996-10-01

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

Plasma shape control calculations for BPX divertor design

International Nuclear Information System (INIS)

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

1991-01-01

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

Divertor heat flux mitigation in the National Spherical Torus Experimenta)

Science.gov (United States)

Soukhanovskii, V. A.; Maingi, R.; Gates, D. A.; Menard, J. E.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Bell, M. G.; Bell, R. E.; Boedo, J. A.; Bush, C. E.; Kaita, R.; Kugel, H. W.; Leblanc, B. P.; Mueller, D.; NSTX Team

2009-02-01

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Analysis of divertor asymmetry using a simple five-point model

International Nuclear Information System (INIS)

Hayashi, Nobuhiko; Takizuka, Tomonori; Hatayama, Akiyoshi; Ogasawara, Masatada.

1997-03-01

A simple five-point model of the scrape-off layer (SOL) plasma outside the separatrix of a diverted tokamak has been developed to study the inside/outside divertor asymmetry. The SOL current, gas pumping/puffing in the divertor region, and divertor plate biasing are included in this model. Gas pumping/puffing and biasing are shown to control divertor asymmetry. In addition, the SOL current is found to form asymmetric solutions without external controls of gas pumping/puffing and biasing. (author)

Structural design of the DIII-D radiative divertor

International Nuclear Information System (INIS)

Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

1996-10-01

The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria

OEDGE modeling of the DIII-D double null (CH4)-C-13 puffing experiment

International Nuclear Information System (INIS)

Elder, J.D.; Wampler, W.R.; McLean, A.G.; Stangeby, P.C.; Allen, S.L.; Bray, B.D.; Brooks, N.H.; Leonard, A.W.; Unterberg, Ezekial A.; Watkins, J.G.

2011-01-01

Unbalanced double null ELMy H-mode configurations in DIII-D are used to simulate the situation in ITER high triangularity, burning plasma magnetic equilibria, where the second X-point lies close to the top of the vacuum vessel, creating a secondary divertor region at the upper blanket modules. The measured plasma conditions in the outer secondary divertor closely duplicated those projected for ITER. (CH4)-C-13 was injected into the secondary outer divertor to simulate sputtering there. The majority of the C-13 found was in the secondary outer divertor. This material migration pattern is radically different than that observed for main wall (CH4)-C-13 injections into single null configurations where the deposition is primarily at the inner divertor. The implications for tritium codeposition resulting from sputtering at the secondary divertor in ITER are significant since release of tritium from Be co-deposits at the main wall bake temperature for ITER, 240 degrees C, is incomplete. The principal features of the measured C-13 deposition pattern have been replicated by the OEDGE interpretive code.

Divertor design for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Hill, D.N.; Braams, B.

1994-05-01

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

Engineering design of a radiative divertor for DIII-D

International Nuclear Information System (INIS)

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

1995-10-01

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

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

Directory of Open Access Journals (Sweden)

M.V. Umansky

2017-08-01

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

Experimental studies of the snowflake divertor in TCV

NARCIS (Netherlands)

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

2017-01-01

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

Exposures of tungsten nanostructures to divertor plasmas in DIII-D

International Nuclear Information System (INIS)

Rudakov, D L; Doerner, R P; Baldwin, M J; Boedo, J A; Hollmann, E M; Moyer, R A; Wong, C P C; Chrobak, C P; Guo, H Y; Leonard, A W; Pace, D C; Thomas, D M; Wright, G M; Abrams, T; Briesemeister, A R; McLean, A G; Fenstermacher, M E; Lasnier, C J; Watkins, J G

2016-01-01

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 little obvious damage except in the areas where unipolar arcing occurred. 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. (paper)

The comparison of heat flux pattern on lower divertor in KSTAR

International Nuclear Information System (INIS)

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

2015-01-01

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

ENERGY DISSIPATION IN MAGNETIC NULL POINTS AT KINETIC SCALES

International Nuclear Information System (INIS)

Olshevsky, Vyacheslav; Lapenta, Giovanni; Divin, Andrey; Eriksson, Elin; Markidis, Stefano

2015-01-01

We use kinetic particle-in-cell and MHD simulations supported by an observational data set to investigate magnetic reconnection in clusters of null points in space plasma. The magnetic configuration under investigation is driven by fast adiabatic flux rope compression that dissipates almost half of the initial magnetic field energy. In this phase powerful currents are excited producing secondary instabilities, and the system is brought into a state of “intermittent turbulence” within a few ion gyro-periods. Reconnection events are distributed all over the simulation domain and energy dissipation is rather volume-filling. Numerous spiral null points interconnected via their spines form null lines embedded into magnetic flux ropes; null point pairs demonstrate the signatures of torsional spine reconnection. However, energy dissipation mainly happens in the shear layers formed by adjacent flux ropes with oppositely directed currents. In these regions radial null pairs are spontaneously emerging and vanishing, associated with electron streams and small-scale current sheets. The number of spiral nulls in the simulation outweighs the number of radial nulls by a factor of 5–10, in accordance with Cluster observations in the Earth's magnetosheath. Twisted magnetic fields with embedded spiral null points might indicate the regions of major energy dissipation for future space missions such as the Magnetospheric Multiscale Mission

Evaluation of null-point detection methods on simulation data

Science.gov (United States)

Olshevsky, Vyacheslav; Fu, Huishan; Vaivads, Andris; Khotyaintsev, Yuri; Lapenta, Giovanni; Markidis, Stefano

2014-05-01

We model the measurements of artificial spacecraft that resemble the configuration of CLUSTER propagating in the particle-in-cell simulation of turbulent magnetic reconnection. The simulation domain contains multiple isolated X-type null-points, but the majority are O-type null-points. Simulations show that current pinches surrounded by twisted fields, analogous to laboratory pinches, are formed along the sequences of O-type nulls. In the simulation, the magnetic reconnection is mainly driven by the kinking of the pinches, at spatial scales of several ion inertial lentghs. We compute the locations of magnetic null-points and detect their type. When the satellites are separated by the fractions of ion inertial length, as it is for CLUSTER, they are able to locate both the isolated null-points, and the pinches. We apply the method to the real CLUSTER data and speculate how common are pinches in the magnetosphere, and whether they play a dominant role in the dissipation of magnetic energy.

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

International Nuclear Information System (INIS)

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

1997-06-01

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

Snowflake divertor experiments on TCV

International Nuclear Information System (INIS)

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

2010-01-01

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

Two-point model for divertor transport

International Nuclear Information System (INIS)

Galambos, J.D.; Peng, Y.K.M.

1984-04-01

Plasma transport along divertor field lines was investigated using a two-point model. This treatment requires considerably less effort to find solutions to the transport equations than previously used one-dimensional (1-D) models and is useful for studying general trends. It also can be a valuable tool for benchmarking more sophisticated models. The model was used to investigate the possibility of operating in the so-called high density, low temperature regime

Operating conditions of the BPX divertor

International Nuclear Information System (INIS)

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

1991-01-01

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

The ‘churning mode’ of plasma convection in the tokamak divertor region

International Nuclear Information System (INIS)

Ryutov, D D; Cohen, R H; Farmer, W A; Rognlien, T D; Umansky, M V

2014-01-01

The churning mode can arise in a toroidally-symmetric plasma where it causes convection in the vicinity of the poloidal magnetic field null. The mode is driven by the toroidal curvature of magnetic field lines coupled with a pressure gradient. The toroidal equilibrium conditions cannot be satisfied easily in the virtual absence of the poloidal field (PF)—hence the onset of this mode, which ‘churns’ the plasma around the PF null without perturbing the strong toroidal field. We find the conditions under which this mode can be excited in magnetic configurations with first-, second-, and third-order PF nulls (i.e., in the geometry of standard, snowflake and cloverleaf divertors). The size of the affected zone in second- and third-order-null divertors is much larger than in a standard first-order-null divertor. The proposed phenomenological theory allows one to evaluate observable characteristics of the mode, in particular the frequency and amplitude of the PF perturbations. The mode spreads the tokamak heat exhaust between multiple divertor legs and may lead to a broadening of the plasma width in each leg. The mode causes much more intense plasma convection in the poloidal plane than the classical plasma drifts. (invited comment)

Analysis of particle transport in a gas target divertor

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

Current singularities at finitely compressible three-dimensional magnetic null points

International Nuclear Information System (INIS)

Pontin, D.I.; Craig, I.J.D.

2005-01-01

The formation of current singularities at line-tied two- and three-dimensional (2D and 3D, respectively) magnetic null points in a nonresistive magnetohydrodynamic environment is explored. It is shown that, despite the different separatrix structures of 2D and 3D null points, current singularities may be initiated in a formally equivalent manner. This is true no matter whether the collapse is triggered by flux imbalance within closed, line-tied null points or driven by externally imposed velocity fields in open, incompressible geometries. A Lagrangian numerical code is used to investigate the finite amplitude perturbations that lead to singular current sheets in collapsing 2D and 3D null points. The form of the singular current distribution is analyzed as a function of the spatial anisotropy of the null point, and the effects of finite gas pressure are quantified. It is pointed out that the pressure force, while never stopping the formation of the singularity, significantly alters the morphology of the current distribution as well as dramatically weakening its strength. The impact of these findings on 2D and 3D magnetic reconnection models is discussed

MAGNETIC NULL POINTS IN KINETIC SIMULATIONS OF SPACE PLASMAS

International Nuclear Information System (INIS)

Olshevsky, Vyacheslav; Innocenti, Maria Elena; Cazzola, Emanuele; Lapenta, Giovanni; Deca, Jan; Divin, Andrey; Peng, Ivy Bo; Markidis, Stefano

2016-01-01

We present a systematic attempt to study magnetic null points and the associated magnetic energy conversion in kinetic particle-in-cell simulations of various plasma configurations. We address three-dimensional simulations performed with the semi-implicit kinetic electromagnetic code iPic3D in different setups: variations of a Harris current sheet, dipolar and quadrupolar magnetospheres interacting with the solar wind, and a relaxing turbulent configuration with multiple null points. Spiral nulls are more likely created in space plasmas: in all our simulations except lunar magnetic anomaly (LMA) and quadrupolar mini-magnetosphere the number of spiral nulls prevails over the number of radial nulls by a factor of 3–9. We show that often magnetic nulls do not indicate the regions of intensive energy dissipation. Energy dissipation events caused by topological bifurcations at radial nulls are rather rare and short-lived. The so-called X-lines formed by the radial nulls in the Harris current sheet and LMA simulations are rather stable and do not exhibit any energy dissipation. Energy dissipation is more powerful in the vicinity of spiral nulls enclosed by magnetic flux ropes with strong currents at their axes (their cross sections resemble 2D magnetic islands). These null lines reminiscent of Z-pinches efficiently dissipate magnetic energy due to secondary instabilities such as the two-stream or kinking instability, accompanied by changes in magnetic topology. Current enhancements accompanied by spiral nulls may signal magnetic energy conversion sites in the observational data

ARIES-III divertor engineering design

International Nuclear Information System (INIS)

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

1992-01-01

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

ARIES-III divertor engineering design

Energy Technology Data Exchange (ETDEWEB)

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

1992-01-01

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

Innovative Divertor Development to Solve the Plasma Heat-Flux Problem

International Nuclear Information System (INIS)

Rognlien, T.; Ryutov, D.; Makowski, M.; Soukhanovskii, V.; Umansky, M.; Cohen, R.; Hill, D.; Joseph, I.

2009-01-01

) and generating Resonant Magnetic Perturbations by the SOL currents (3). However, the specific approaches discussed here are part of a wider class of innovative divertor ideas that have come from the community in the last several years, and we certainly advocate the need to consider a range of options. Indeed, the most effective solution to the heat-flux problem may well contain features of various ideas. For example, there are the X-divertor (Kotschenreuther et al. (4)) that expands the magnetic flux surface in the vicinity of the near-X-point divertor plate, and the super X-divertor (Valanju et al. (5)) that guides the near-separatrix SOL flux tubes to a larger major radius to increase the surface area available for power deposition. These approaches have the common feature of manipulation of the edge magnetic geometry. Another approach is the use of liquid divertor surfaces that can increase the heat-flux capability by flowing the heated material to a cooling region and eventually out of the machine, and/or by being able to withstand a higher peak heat flux (6). All of these areas are only emerging concepts that require substantially more analysis and definitive experimental tests, and given the need for a large improvement in this area, we advocate a substantial program to systematically assess the approaches. Because of space limitation here, we present some details of one of the concepts, namely the Snowflake divertor configuration. The Snowflake (SF) divertor (2) exploits a tokamak geometry in which the poloidal magnetic field varies quadratically with distance from the X-point null, Δr. The name stems from the characteristic hexagonal, snowflake-like, shape of the multi-branched separatrix for this exact second-order null. In contrast, the standard X-point configuration has a poloidal field varying linearly with ?r. The different variations mean that a flux expansion is much larger in the vicinity of a null of a snowflake divertor, and one can try to exploit

Experimental studies of the snowflake divertor in TCV

Directory of Open Access Journals (Sweden)

B. Labit

2017-08-01

Full Text Available To address the risk that, in a fusion reactor, the conventional single-null divertor (SND configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD, are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized.

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

International Nuclear Information System (INIS)

Bertolini, E.

1995-01-01

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

A snowflake divertor: a possible solution to the power exhaust problem for tokamaks

Energy Technology Data Exchange (ETDEWEB)

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

2012-11-21

This paper summarizes recent progress in the theory of a snowflake divertor, a possible path to reduce both steady-state and intermittent heat loads on the divertor plates to an acceptable level. The most important feature of a SF divertor is the presence of a large zone of a very weak poloidal magnetic field around the poloidal field (PF) null. Qualitative explanation of a variety of new features characteristic of a SF divertor is provided based on simple scaling relations. The main part of the paper is focused on the concept of spreading of the heat flux by curvature-driven convection near the PF null. References to experimental results from the NSTX and TCV tokamaks are provided.

Null point of discrimination in crustacean polarisation vision.

Science.gov (United States)

How, Martin J; Christy, John; Roberts, Nicholas W; Marshall, N Justin

2014-07-15

The polarisation of light is used by many species of cephalopods and crustaceans to discriminate objects or to communicate. Most visual systems with this ability, such as that of the fiddler crab, include receptors with photopigments that are oriented horizontally and vertically relative to the outside world. Photoreceptors in such an orthogonal array are maximally sensitive to polarised light with the same fixed e-vector orientation. Using opponent neural connections, this two-channel system may produce a single value of polarisation contrast and, consequently, it may suffer from null points of discrimination. Stomatopod crustaceans use a different system for polarisation vision, comprising at least four types of polarisation-sensitive photoreceptor arranged at 0, 45, 90 and 135 deg relative to each other, in conjunction with extensive rotational eye movements. This anatomical arrangement should not suffer from equivalent null points of discrimination. To test whether these two systems were vulnerable to null points, we presented the fiddler crab Uca heteropleura and the stomatopod Haptosquilla trispinosa with polarised looming stimuli on a modified LCD monitor. The fiddler crab was less sensitive to differences in the degree of polarised light when the e-vector was at -45 deg than when the e-vector was horizontal. In comparison, stomatopods showed no difference in sensitivity between the two stimulus types. The results suggest that fiddler crabs suffer from a null point of sensitivity, while stomatopods do not. © 2014. Published by The Company of Biologists Ltd.

Engineering design of a Radiative Divertor for DIII-D

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Science.gov (United States)

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

Modelling of profile control with LH wave injection in the HL-2A single-null divertor plasma

International Nuclear Information System (INIS)

Gao Qingdi; Yuan Baoshan; Li Fangzhu; Wang, Aike; Budny, R.V.

2005-01-01

In the HL-2A tokamak a single-null divertor configuration has been established. The separatrix of the single-null diverted plasma was identified with a filament model, and the determined striking area on the target plate is in agreement with the measurements of electric probe array. Higher power LH wave (1.5MW) is injected to the diverted plasma with a nearly symmetric spectrum. Dominant electron heating and current profile control are investigated with numerical simulation. Plasma heating by electron Landau interaction results in operation scenarios of preferentially dominant electron heating. Due to the off-axis driven current, an optimized q-profile is formed, and an enhanced confinement regime with steep electron temperature gradient is produced. The clear decrease of the electron thermal conductivity in the LH power deposition region shows that an electron-ITB is developed. When higher LH power injects into the target plasma that is heated by NBI (0.5MW), the ion temperature has a large increment in addition to the high increase of electron temperature. The temperature profiles indicate that an enhanced core confinement is established with both ion-ITB and electron-ITB developed. (author)

Modeling of thermal effects on TIBER II divertor during plasma disruptions

International Nuclear Information System (INIS)

Bruhn, M.L.; Perkins, L.J.

1987-01-01

Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs

Theory and Simulations of ELM Control with a Snowflake Divertor

Energy Technology Data Exchange (ETDEWEB)

Ryutov, D.; Cohen, B.; Cohen, R.; Makowski, M. A.; Menard, J.; Rognlien, T.; Soukhanovskii, V.; Umansky, M.; Xu, X., E-mail: ryutov1@llnl.gov [Lawrence Livermore National Laboratory, Livermore (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, Princeton (United States)

2012-09-15

Full text: This paper is concerned with the use of a snowflake (SF) divertor for the control and mitigation of edge localized modes (ELMs). Our research is focused on the following three issues: 1. Effect of the SF geometry on neoclassical ion orbits near the separatrix, including prompt ion losses and the related control mechanism for the electric field and plasma flow in the pedestal; 2. Influence of the thereby modified flow and of high poloidal plasma beta in the divertor region on plasma turbulence and transport in the snowflake-plus geometry; 3. Reaction of the SF divertor to type-1 ELM events. Neoclassical ion orbits in the vicinity of the SF separatrix are changed due to a much weaker poloidal field near the null and much longer particle dwell-time in this area. This leads to an increase of the prompt ion loss, which then affects the radial electric field profile near the separatrix. The resulting E x B flow shear in the pedestal region affects the onset of ELMs. The electric field and velocity shear are then used as a background for two-fluid simulations of the edge plasma turbulence in a realistic geometry with the 3D BOUT code. A SF-plus geometry is chosen, so that the separatrix topology remains the same as for the standard X-point divertor, whereas the magnetic shear both inside and outside the separatrix increases dramatically. It is found that mesoscale instabilities are suppressed when the geometry is close to a perfect SF. In situations where complete suppression of ELMs is impossible, the SF divertor offers a path to reducing heat loads during ELM events to an acceptable level. Two effects, both related to the weakness of the poloidal field near the SF null, act synergistically in the same favorable direction. The first is the onset of strong, curvature-driven convection in the divertor, triggered by the increase of the poloidal pressure during the ELM and leading to the splitting of the heat flux between all four (as is the case in a SF geometry

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Engineering design of cryocondensation pumps for the DIII-D Radiative Divertor Program

International Nuclear Information System (INIS)

Bozek, A.S.; Baxi, C.B.; Del Bene, J.V.; Laughon, G.J.; Reis, E.E.; Shatoff, H.D.; Smith, J.P.

1995-01-01

A new double-null, slotted divertor configuration will be installed for the DIII-D Radiative Divertor Program at General Atomics in late 1996. Four cryocondensation pumps, three new and one existing, will be part of this new divertor. The purpose of the pumps is to provide plasma density control and to limit the impurities entering the plasma core by providing pumping at each divertor strike point. The three new pumps are based on the design of the existing pump, installed in 1992 as part of the Advanced Divertor Program. The pump continues to operate successfully. The new pumps require geometry modifications to the original design. Therefore, extensive modal and dynamic analyses were performed to determine the behavior of these pumps and their helium and nitrogen feed lines during disruption events. Thermal and fluid analyses were also performed to characterize the helium two-phase flow regime in the pumps and their feedlines. A flow testing program was completed to test the change in geometry of the pump feed lines with respect to helium flow stability. The results were compared to the helium thermal and fluid analyses to verify predicted flow regimes and flow stability

THE NATURE OF FLARE RIBBONS IN CORONAL NULL-POINT TOPOLOGY

International Nuclear Information System (INIS)

Masson, S.; Aulanier, G.; Pariat, E.; Schrijver, C. J.

2009-01-01

Flare ribbons are commonly attributed to the low-altitude impact, along the footprints of separatrices or quasi-separatrix layers (QSLs), of particle beams accelerated through magnetic reconnection. If reconnection occurs at a three-dimensional coronal magnetic null point, the footprint of the dome-shaped fan surface would map a closed circular ribbon. This paper addresses the following issues: does the entire circular ribbon brighten simultaneously, as expected because all fan field lines pass through the null point? And since the spine separatrices are singular field lines, do spine-related ribbons look like compact kernels? What can we learn from these observations about current sheet formation and magnetic reconnection in a null-point topology? The present study addresses these questions by analyzing Transition Region and Coronal Explorer and Solar and Heliospheric Observatory/Michelson Doppler Imager observations of a confined flare presenting a circular ribbon. Using a potential field extrapolation, we linked the circular shape of the ribbon with the photospheric mapping of the fan field lines originating from a coronal null point. Observations show that the flare ribbon outlining the fan lines brightens sequentially along the counterclockwise direction and that the spine-related ribbons are elongated. Using the potential field extrapolation as initial condition, we conduct a low-β resistive magnetohydrodynamics simulation of this observed event. We drive the coronal evolution by line-tied diverging boundary motions, so as to emulate the observed photospheric flow pattern associated with some magnetic flux emergence. The numerical analysis allows us to explain several observed features of the confined flare. The vorticity induced in the fan by the prescribed motions causes the spines to tear apart along the fan. This leads to formation of a thin current sheet and induces null-point reconnection. We also find that the null point and its associated topological

Engineering design of the Aries-IV gaseous divertor

International Nuclear Information System (INIS)

Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

1994-01-01

ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

Directory of Open Access Journals (Sweden)

T. Lunt

2017-08-01

Full Text Available We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF and the X-divertor (XD, that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN and double-null (DN would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

Multi-Fluid Modeling of Low-Recycling Divertor Regimes

International Nuclear Information System (INIS)

Smirnov, R.D.; Pigarov, A.Y.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

2010-01-01

The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Radiative and SOL experiments in open and baffled divertors on DIII-D

International Nuclear Information System (INIS)

Allen, S.L.; Brooks, N.H.; Bastasz, R.

1998-11-01

The authors present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. They have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D 2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with puff and pump techniques using SOL D 2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to n e /n gw ∼ 0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN

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

NARCIS (Netherlands)

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

2018-01-01

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

Liquid metal cooled divertor for ARIES

International Nuclear Information System (INIS)

Muraviev, E.

1995-01-01

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

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

2016-01-01

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. (paper)

Multi-fluid modeling of low-recycling divertor regimes

International Nuclear Information System (INIS)

Smirnov, R.D.; Pigarov, A.Yu.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

2010-01-01

The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

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

Science.gov (United States)

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

2018-05-01

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

Null-point titration measurements of free magnesium in stored erythrocytes

International Nuclear Information System (INIS)

Bock, J.L.; Yusuf, Y.; Puntillo, E.

1987-01-01

Free intracellular magnesium concentration (Mg/sub i/) was measured in stored human erythrocytes, using null-point titration with ionophore A23187. For cells stored 31 P NMR spectroscopy, which showed a decrease in Mg/sub i/ with storage. However, the NMR measurements are performed with no pretreatment of the cells, while the null-point method requires an initial washing step, which alters pH/sub i/ and may also alter Mg/sub i/. The titration-measured Mg/sub i/ values are still surprisingly low for long-stored cells, considering that depletion of ATP and 2,3-DPG should release bound Mg. Using the titration-measured Mg/sub i/ values along with measurements of total Mg, ATP, and 2,3-DPG, they estimate that an additional buffer contains about 47% of total Mg in cells stored 21 days. Mg/sub i/ determinations by both 31 P NMR and null-point titration thus indicate that erythrocyte Mg is largely bound to a high-capacity, low-affinity buffer whose relative importance increases during cell storage. Discrepancies between the methods require further investigation

NSTX Tangential Divertor Camera

International Nuclear Information System (INIS)

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

2004-01-01

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

L-H power threshold studies with tungsten/carbon divertor on the EAST tokamak

DEFF Research Database (Denmark)

Chen, L.; Xu, G. S.; Gao, W.

2016-01-01

The power threshold for low (L) to high (H) confinement mode transition achieved by radio-frequency heating and molybdenum first wall with lithium coating has been experimentally investigated on the EAST tokamak for two sets of divertor geometries and materials: tungsten/carbon divertor and full...... carbon divertor. For both sets of divertors, the power threshold was found to decrease with gradual accumulation of the lithium wall coating, suggesting the important role played by the low Z impurities and/or the edge neutral density on the L-H power threshold. When operating in the upper single null...

Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

International Nuclear Information System (INIS)

Bruhn, M.L.; Perkins, L.J.

1987-01-01

Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

OBSERVATION OF MAGNETIC RECONNECTION AT A 3D NULL POINT ASSOCIATED WITH A SOLAR ERUPTION

International Nuclear Information System (INIS)

Sun, J. Q.; Yang, K.; Cheng, X.; Ding, M. D.; Zhang, J.

2016-01-01

Magnetic null has long been recognized as a special structure serving as a preferential site for magnetic reconnection (MR). However, the direct observational study of MR at null-points is largely lacking. Here, we show the observations of MR around a magnetic null associated with an eruption that resulted in an M1.7 flare and a coronal mass ejection. The Geostationary Operational Environmental Satellites X-ray profile of the flare exhibited two peaks at ∼02:23 UT and ∼02:40 UT on 2012 November 8, respectively. Based on the imaging observations, we find that the first and also primary X-ray peak was originated from MR in the current sheet (CS) underneath the erupting magnetic flux rope (MFR). On the other hand, the second and also weaker X-ray peak was caused by MR around a null point located above the pre-eruption MFR. The interaction of the null point and the erupting MFR can be described as a two-step process. During the first step, the erupting and fast expanding MFR passed through the null point, resulting in a significant displacement of the magnetic field surrounding the null. During the second step, the displaced magnetic field started to move back, resulting in a converging inflow and subsequently the MR around the null. The null-point reconnection is a different process from the current sheet reconnection in this flare; the latter is the cause of the main peak of the flare, while the former is the cause of the secondary peak of the flare and the conspicuous high-lying cusp structure.

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Advanced divertor configurations with large flux expansion

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-15

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

ON THE NATURE OF RECONNECTION AT A SOLAR CORONAL NULL POINT ABOVE A SEPARATRIX DOME

International Nuclear Information System (INIS)

Pontin, D. I.; Priest, E. R.; Galsgaard, K.

2013-01-01

Three-dimensional magnetic null points are ubiquitous in the solar corona and in any generic mixed-polarity magnetic field. We consider magnetic reconnection at an isolated coronal null point whose fan field lines form a dome structure. Using analytical and computational models, we demonstrate several features of spine-fan reconnection at such a null, including the fact that substantial magnetic flux transfer from one region of field line connectivity to another can occur. The flux transfer occurs across the current sheet that forms around the null point during spine-fan reconnection, and there is no separator present. Also, flipping of magnetic field lines takes place in a manner similar to that observed in the quasi-separatrix layer or slip-running reconnection

Magnetic Reconnection at a Three-dimensional Solar Null Point

DEFF Research Database (Denmark)

Frederiksen, Jacob Trier; Baumann, Gisela; Galsgaard, Klaus

2012-01-01

Using a specific solar null point reconnection case studied by Masson et al (2009; ApJ 700, 559) we investigate the dependence of the reconnection rate on boundary driving speed, numerical resolution, type of resistivity (constant or numerical), and assumed stratification (constant density or sol...

Shocks and currents in stratified atmospheres with a magnetic null point

Science.gov (United States)

Tarr, Lucas A.; Linton, Mark

2017-08-01

We use the resistive MHD code LARE (Arber et al 2001) to inject a compressive MHD wavepacket into a stratified atmosphere that has a single magnetic null point, as recently described in Tarr et al 2017. The 2.5D simulation represents a slice through a small ephemeral region or area of plage. The strong gradients in field strength and connectivity related to the presence of the null produce substantially different dynamics compared to the more slowly varying fields typically used in simple sunspot models. The wave-null interaction produces a fast mode shock that collapses the null into a current sheet and generates a set of outward propagating (from the null) slow mode shocks confined to field lines near each separatrix. A combination of oscillatory reconnection and shock dissipation ultimately raise the plasma's internal energy at the null and along each separatrix by 25-50% above the background. The resulting pressure gradients must be balanced by Lorentz forces, so that the final state has contact discontinuities along each separatrix and a persistent current at the null. The simulation demonstrates that fast and slow mode waves localize currents to the topologically important locations of the field, just as their Alfvenic counterparts do, and also illustrates the necessity of treating waves and reconnection as coupled phenomena.

Alfvén wave dynamics at the neighborhood of a 2.5D magnetic null-point

Science.gov (United States)

Sabri, S.; Vasheghani Farahani, S.; Ebadi, H.; Hosseinpour, M.; Fazel, Z.

2018-05-01

The aim of the present study is to highlight the energy transfer via the interaction of magnetohydrodynamic waves with a 2.5D magnetic null-point in a finite plasma-β regime of the solar corona. An initially symmetric Alfvén pulse at a specific distance from a magnetic null-point is kicked towards the isothermal null-point. A shock-capturing Godunov-type PLUTO code is used to solve the ideal magnetohydrodynamic set equations in the context of wave-plasma energy transfer. As the Alfvén wave propagates towards the magnetic null-point it experiences speed lowering which ends up in releasing energy along the separatrices. In this line owing to the Alfvén wave, a series of events take place that contribute towards coronal heating. Nonlinear induced waves are by products of the torsional Alfvén interaction with magnetic null-points. The energy of these induced waves which are fast magnetoacoustic (transverse) and slow magnetoacoustic (longitudinal) waves are supplied by the Alfvén wave. The nonlinearly induced density perturbations are proportional to the Alfvén wave energy loss. This supplies energy for the propagation of fast and slow magnetoacoustic waves, where in contrast to the fast wave the slow wave experiences a continuous energy increase. As such, the slow wave may transfer its energy to the medium at later times, maintaining a continuous heating mechanism at the neighborhood of a magnetic null-point.

T-12 divertor experiment

Energy Technology Data Exchange (ETDEWEB)

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

1981-01-01

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

Magnetoacoustic Waves in a Stratified Atmosphere with a Magnetic Null Point

Energy Technology Data Exchange (ETDEWEB)

Tarr, Lucas A.; Linton, Mark; Leake, James, E-mail: lucas.tarr.ctr@nrl.navy.mil [U.S. Naval Research Laboratory, 4555 Overlook Ave. SW, Washington, DC 20375 (United States)

2017-03-01

We perform nonlinear MHD simulations to study the propagation of magnetoacoustic waves from the photosphere to the low corona. We focus on a 2D system with a gravitationally stratified atmosphere and three photospheric concentrations of magnetic flux that produce a magnetic null point with a magnetic dome topology. We find that a single wavepacket introduced at the lower boundary splits into multiple secondary wavepackets. A portion of the packet refracts toward the null owing to the varying Alfvén speed. Waves incident on the equipartition contour surrounding the null, where the sound and Alfvén speeds coincide, partially transmit, reflect, and mode-convert between branches of the local dispersion relation. Approximately 15.5% of the wavepacket’s initial energy ( E {sub input}) converges on the null, mostly as a fast magnetoacoustic wave. Conversion is very efficient: 70% of the energy incident on the null is converted to slow modes propagating away from the null, 7% leaves as a fast wave, and the remaining 23% (0.036 E {sub input}) is locally dissipated. The acoustic energy leaving the null is strongly concentrated along field lines near each of the null’s four separatrices. The portion of the wavepacket that refracts toward the null, and the amount of current accumulation, depends on the vertical and horizontal wavenumbers and the centroid position of the wavepacket as it crosses the photosphere. Regions that refract toward or away from the null do not simply coincide with regions of open versus closed magnetic field or regions of particular field orientation. We also model wavepacket propagation using a WKB method and find that it agrees qualitatively, though not quantitatively, with the results of the numerical simulation.

The effect of density on divertor conditions in ASDEX-Upgrade

International Nuclear Information System (INIS)

Pitcher, C.S.; Bosch, H.-S.; Buechl, K.; Field, A.; Fuchs, C.; Haas, G.; Junker, W.; Neu, R.; Neuhauser, J.; Wenzel, U.

1995-01-01

Detailed experimental divertor data are presented on the profiles of density and temperature in the inner and outer divertor fans, the radiated power distribution, the gas pressure and the spectroscopically derived particle fluxes, all as a function of the discharge density. At low and medium density, the inner divertor is cold and dense compared to the outer divertor. At high density, strong X-point MARFE and separatrix radiation partially detaches the inner divertor. Probe measurements which penetrate into the X-point MARFE at the outer divertor are presented. ((orig.))

Comparison between a pumped-limiter and a divertor for the next step machines

International Nuclear Information System (INIS)

Harrison, F.F.A.

1985-01-01

The paper presents a simple description of the physics issues which influence the conceptual design of a pumped-limiter and single-null poloidal divertor in a next step, long burn tokamak of NET/INTOR scale. Predicted performance of the limiter and divertor are compared in regard to localised recycling, sputtering of the plasma collection surfaces, penetration of sputtered impurities into the fusion plasma, surface power loading and exhaust of helium ash. It is concluded that the performance of the divertor is superior and that it can be predicted with a reasonable degree of confidence. The viability of the limiter remains in doubt but the concept cannot be rejected at the present time

Design of the divertor Thomson scattering system on DIII-D

International Nuclear Information System (INIS)

Carlstrom, T.N.; Foote, J.H.; Nilson, D.G.; Rice, B.W.

1994-05-01

Local measurements of n e and T e in the divertor region are necessary for a more complete understanding of divertor physics. We have designed an extension to the existing multipulse Thomson scattering system to measure n e in the range 5 x 10 18 to 5 x 10 20 m -3 and T e 5--500 eV, with 1 cm resolution from 1--21 cm above the floor of the DIII-D vessel, in the region of the X-point for lower single-null diverted plasmas. One of the existing 8, 20 Hz, ND:YAG lasers will be redirected to a separate vertical port, and viewed radially with a specially designed, f/6.8 lens. Fiber optics carry the light to additional polychromators whose interference filters have been optimized for low T e measurements. Other aspect of the system, including the beam path to the vessel, polychromator design, real time data acquisition, laser control, calibration facility, and DIII-D timing and data acquisition interface will be shared with the existing multipulse Thomson system. An in-situ laser alignment monitor will provide alignment information for each laser pulse

Small angle slot divertor concept for long pulse advanced tokamaks

Science.gov (United States)

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

2017-04-01

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

Actively convected liquid metal divertor

International Nuclear Information System (INIS)

Shimada, Michiya; Hirooka, Yoshi

2014-01-01

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

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

Science.gov (United States)

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

2018-04-01

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

Kinetic modeling of particle acceleration in a solar null point reconnection region

DEFF Research Database (Denmark)

Baumann, Gisela; Haugbølle, Troels; Nordlund, Åke

2013-01-01

The primary focus of this paper is on the particle acceleration mechanism in solar coronal 3D reconnection null-point regions. Starting from a potential field extrapolation of a SOHO magnetogram taken on 2002 November 16, we first performed MHD simulations with horizontal motions observed by SOHO...... particles and 3.5 billion grid cells of size 17.5\\,km --- these simulations offer a new opportunity to study particle acceleration in solar-like settings....... applied to the photospheric boundary of the computational box. After a build-up of electric current in the fan-plane of the null-point, a sub-section of the evolved MHD data was used as initial and boundary conditions for a kinetic particle-in-cell model of the plasma. We find that sub...

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

Directory of Open Access Journals (Sweden)

T.D. Rognlien

2017-08-01

Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

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

Science.gov (United States)

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

2018-05-01

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

Design of DIII-D advanced divertor

International Nuclear Information System (INIS)

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

1989-01-01

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

Design of DIII-D Advanced Divertor

International Nuclear Information System (INIS)

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

1989-11-01

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

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

International Nuclear Information System (INIS)

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

1979-01-01

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm 2 ) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm 2 ) and for ISX-B 2 (11 kA/cm 2 ). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure

Physical study of experimental fusion breeder FEB divertor

International Nuclear Information System (INIS)

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

1999-10-01

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

Modelling of radial electric field profile for different divertor configurations

International Nuclear Information System (INIS)

Rozhansky, V; Kaveeva, E; Voskoboynikov, S; Counsell, G; Kirk, A; Meyer, H; Coster, D; Conway, G; Schirmer, J; Schneider, R

2006-01-01

The impact of divertor configuration on the structure of the radial electric field has been simulated by the B2SOLPS5.0 transport fluid code. It is shown that the change in the parallel flows in the scrape-off layer, which are transported through the separatrix due to turbulent viscosity and diffusivity, should result in variation of the radial electric field and toroidal rotation in the separatrix vicinity. The modelling predictions are compared with the measurements of the radial electric field for the low field side equatorial mid-plane of ASDEX Upgrade in lower, upper and double-null (DN) divertor configurations. The parallel (toroidal) flows in the scrape-off layer and mechanisms for their formation are analysed for different geometries. It is demonstrated that a spike in the electric field exists at the high field side equatorial mid-plane in the connected DN divertor configuration. Its origin is connected with different potential drops between the separatrix vicinity and divertor plates in the two disconnected scrape-off layers, while the separatrix should be at almost the same potential. The spike might be important for additional turbulent suppression

Observation of a 3D Magnetic Null Point

Energy Technology Data Exchange (ETDEWEB)

Romano, P.; Falco, M. [INAF—Osservatorio Astrofisico di Catania, Via S. Sofia 78, I-95123 Catania (Italy); Guglielmino, S. L.; Murabito, M., E-mail: prom@oact.inaf.it [Dipartimento di Fisica e Astronomia—Sezione Astrofisica, Università di Catania, Via S. Sofia 78, I-95123 Catania (Italy)

2017-03-10

We describe high-resolution observations of a GOES B-class flare characterized by a circular ribbon at the chromospheric level, corresponding to the network at the photospheric level. We interpret the flare as a consequence of a magnetic reconnection event that occurred at a three-dimensional (3D) coronal null point located above the supergranular cell. The potential field extrapolation of the photospheric magnetic field indicates that the circular chromospheric ribbon is cospatial with the fan footpoints, while the ribbons of the inner and outer spines look like compact kernels. We found new interesting observational aspects that need to be explained by models: (1) a loop corresponding to the outer spine became brighter a few minutes before the onset of the flare; (2) the circular ribbon was formed by several adjacent compact kernels characterized by a size of 1″–2″; (3) the kernels with a stronger intensity emission were located at the outer footpoint of the darker filaments, departing radially from the center of the supergranular cell; (4) these kernels started to brighten sequentially in clockwise direction; and (5) the site of the 3D null point and the shape of the outer spine were detected by RHESSI in the low-energy channel between 6.0 and 12.0 keV. Taking into account all these features and the length scales of the magnetic systems involved in the event, we argue that the low intensity of the flare may be ascribed to the low amount of magnetic flux and to its symmetric configuration.

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

Science.gov (United States)

Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

2017-10-01

In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

Heat and particle transport of sol/divertor plasma in the W-shaped divertor on JT-60U

International Nuclear Information System (INIS)

Asakura, N.; Sakurai, S.; Hosogane, N.

1999-01-01

The plasma profile and parallel flow in the scrape-off layer (SOL) were systematically measured using Mach probes installed at the midplane and the divertor x-point. Quantitative evaluation of a parallel flow: naturally produced in a torus to keep the pressure constant along the field line, was consistent with the measurement. Geometry effects of the W-shaped divertor on the divertor plasma and particle recycling at the newly installed baffle plates were evaluated quantitatively using the edge plasma data. (author)

The MAST improved divertor

International Nuclear Information System (INIS)

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

2005-01-01

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

3D Solar Null Point Reconnection MHD Simulations

Science.gov (United States)

Baumann, G.; Galsgaard, K.; Nordlund, Å.

2013-06-01

Numerical MHD simulations of 3D reconnection events in the solar corona have improved enormously over the last few years, not only in resolution, but also in their complexity, enabling more and more realistic modeling. Various ways to obtain the initial magnetic field, different forms of solar atmospheric models as well as diverse driving speeds and patterns have been employed. This study considers differences between simulations with stratified and non-stratified solar atmospheres, addresses the influence of the driving speed on the plasma flow and energetics, and provides quantitative formulas for mapping electric fields and dissipation levels obtained in numerical simulations to the corresponding solar quantities. The simulations start out from a potential magnetic field containing a null-point, obtained from a Solar and Heliospheric Observatory (SOHO) Michelson Doppler Imager (MDI) magnetogram magnetogram extrapolation approximately 8 hours before a C-class flare was observed. The magnetic field is stressed with a boundary motion pattern similar to - although simpler than - horizontal motions observed by SOHO during the period preceding the flare. The general behavior is nearly independent of the driving speed, and is also very similar in stratified and non-stratified models, provided only that the boundary motions are slow enough. The boundary motions cause a build-up of current sheets, mainly in the fan-plane of the magnetic null-point, but do not result in a flare-like energy release. The additional free energy required for the flare could have been partly present in non-potential form at the initial state, with subsequent additions from magnetic flux emergence or from components of the boundary motion that were not represented by the idealized driving pattern.

A new scaling for divertor detachment

Science.gov (United States)

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

2017-05-01

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

Divertor erosion in DIII-D

International Nuclear Information System (INIS)

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

1998-05-01

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

Developing physics basis for the snowflake divertor in the DIII-D tokamak

Science.gov (United States)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

Science.gov (United States)

Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

1999-01-01

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

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

Directory of Open Access Journals (Sweden)

J. Uljanovs

2017-08-01

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

Studies of high-δ (baffled) and low-δ (open) pumped divertor operation on DIII-D

International Nuclear Information System (INIS)

Allen, S.L.; Fenstermacher, M.E.; Greenfield, C.M.

1998-08-01

The authors report new experimental results with the RDP-OB (Radiative Divertor Project-outer baffle) and cryopump in both upper single-null (USN) and double-null (DN) ELMing H-mode discharges. The baffled divertor reduced the core ionization (∼2--2.5x), in reasonable agreement with predictions from UEDGE/DEGAS modeling (∼3.75x). The upper cryopump achieved density control of n e /n gw ∼ 0.22 (line density/Greenwald density) with Z eff ∼ 2 in high-δ plasmas. The measured exhaust is comparable to the lower pump, except at lower core electron densities (n e 19 m -3 ). Efficient impurity exhaust was obtained with deuterium SOL flow. Preliminary experiments with DN operation has shown that the particle exhaust to the upper pump depends on the up/down magnetic balance. Preliminary experiments indicate that the DN exhaust is roughly 40--50% of the USN exhaust at n e ∼ 4 x 10 19 m -3

Two-dimensional divertor modeling and scaling laws

International Nuclear Information System (INIS)

Catto, P.J.; Connor, J.W.; Knoll, D.A.

1996-01-01

Two-dimensional numerical models of divertors contain large numbers of dimensionless parameters that must be varied to investigate all operating regimes of interest. To simplify the task and gain insight into divertor operation, we employ similarity techniques to investigate whether model systems of equations plus boundary conditions in the steady state admit scaling transformations that lead to useful divertor similarity scaling laws. A short mean free path neutral-plasma model of the divertor region below the x-point is adopted in which all perpendicular transport is due to the neutrals. We illustrate how the results can be used to benchmark large computer simulations by employing a modified version of UEDGE which contains a neutral fluid model. (orig.)

Divertor radiation in the ASDEX upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

International Nuclear Information System (INIS)

Bruhn, M.L.

1988-04-01

Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

Divertor erosion in DIII-D

International Nuclear Information System (INIS)

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

1998-08-01

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

Snowflake divertor configuration studies in National Spherical Torus Experiment

Energy Technology Data Exchange (ETDEWEB)

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

2012-08-15

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

Evaluating Stellarator Divertor Designs with EMC3

Science.gov (United States)

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

2013-10-01

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

Divertor detachment

Science.gov (United States)

Krasheninnikov, Sergei

2015-11-01

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

Advanced divertor concepts

International Nuclear Information System (INIS)

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

1996-01-01

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

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

Directory of Open Access Journals (Sweden)

F. Subba

2017-08-01

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

Module of lithium divertor for KTM tokamak

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

Disruption Neutral Point Experiment on Alcator C-Mod

Science.gov (United States)

Granetz, R. S.; Nakamura, Y.

2000-10-01

Disruptions of single-null elongated plasmas generally result in loss of vertical position control, leading to a current quench occurring at the top or bottom of the machine, with all the attendant problems of halo and eddy currents flowing in divertor structures. On JT-60U, it has been found that if the plasma is operated with its magnetic axis at a particular height, called the neutral point, the initial vertical drift after a thermal quench is significantly slower than usual, and sometimes can even be arrested, thereby avoiding a current quench in the divertor region entirely. In an ongoing collaboration between MIT and JAERI, the neutral point concept is being tested in Alcator C-Mod, which has a significantly higher plasma elongation than JT-60U (1.65 vs 1.3). Calculations using TSC predict a neutral point at z~=+1 cm above the midplane (a=22 cm). The existence of a neutral point has now been experimentally confirmed, albeit at a height of z=+2.7 cm. The plasma has remained vertically stable for up to 9 ms after the disruption thermal quench, which in principle, is long enough for the PF control system to respond, if programmed appropriately. In addition, the physics of the neutral point stability on C-Mod appears to be somewhat different than that on JT-60U.

Utilization of vanadium alloys in the DIII-D Radiative Divertor Program

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1995-10-01

Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics (GA), in conjunction with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan for the utilization of vanadium alloys as part of the Radiative Divertor (RD) upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy (DOE). This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components, and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming Radiative Divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development (R and D) efforts to support fabrication development and to resolve key issues related to environmental effects

Utilization of vanadium alloys in the DIII-D radiative divertor program

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1996-01-01

Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics in conjunction with Argonne National Laboratory and Oak Ridge National Laboratory has developed a plan for the utilization of vanadium alloys as part of the radiative divertor upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy. This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming radiative divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development efforts to support fabrication development and to resolve key issues related to environmental effects. (orig.)

LHD helical divertor

International Nuclear Information System (INIS)

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

1993-07-01

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

The ITER divertor concept

International Nuclear Information System (INIS)

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

1995-01-01

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

Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

2009-12-15

Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

International Nuclear Information System (INIS)

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

1990-09-01

The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs

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

International Nuclear Information System (INIS)

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

2010-01-01

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.

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

Science.gov (United States)

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

2013-10-01

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

Toroidal asymmetries in divertor impurity influxes in NSTX

Directory of Open Access Journals (Sweden)

F. Scotti

2017-08-01

Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Standard Test Method for Measuring Extreme Heat-Transfer Rates from High-Energy Environments Using a Transient, Null-Point Calorimeter

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers the measurement of the heat-transfer rate or the heat flux to the surface of a solid body (test sample) using the measured transient temperature rise of a thermocouple located at the null point of a calorimeter that is installed in the body and is configured to simulate a semi-infinite solid. By definition the null point is a unique position on the axial centerline of a disturbed body which experiences the same transient temperature history as that on the surface of a solid body in the absence of the physical disturbance (hole) for the same heat-flux input. 1.2 Null-point calorimeters have been used to measure high convective or radiant heat-transfer rates to bodies immersed in both flowing and static environments of air, nitrogen, carbon dioxide, helium, hydrogen, and mixtures of these and other gases. Flow velocities have ranged from zero (static) through subsonic to hypersonic, total flow enthalpies from 1.16 to greater than 4.65 × 101 MJ/kg (5 × 102 to greater than 2 × 104 ...

Double null X-point operation in JET with NBI and ICRH heating

International Nuclear Information System (INIS)

Tubbing, B.; Bhatnagar, V.

1989-01-01

In this paper we report on a selection of experiments on H-modes, in 3 and 3.5MA discharges, in the double null X-point configuration. The first experiment, section 2, is an attempt to couple ICRH power to H-modes. Here we also report on a rather unique H-mode with a smaller than usual distance between plasma and limiter. The second experiment, section 3, is on H-modes in the low density, hot ion regime. (author) 5 refs., 4 figs

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

Science.gov (United States)

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.

Hydrogen recycling and transport in the helical divertor of TEXTOR

Energy Technology Data Exchange (ETDEWEB)

Clever, Meike

2010-07-01

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

Hydrogen recycling and transport in the helical divertor of TEXTOR

International Nuclear Information System (INIS)

Clever, Meike

2010-01-01

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

NSTX plasma operation with a Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

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

2012-10-15

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

Plans of LHD divertor experiment

International Nuclear Information System (INIS)

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

1996-01-01

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

Divertor characterization experiments

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

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

1991-02-01

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

Calculation of null points in SPAIR, FLAIR, and STIR. Part 2. Application to pelvic diffusion-weighted imaging with SPAIR at 3T

International Nuclear Information System (INIS)

Kita, Miho; Kawano, Kazuhiro; Kometani, Katsuya; Kondo, Toshihiko; Shimamoto, Kazuhiro; Tanaka, Humihiro; Oda, Hideyuki; Kojima, Akihiro; Sato, Morio

2013-01-01

Diffusion-weighted imaging (DWI) requires adequate fat suppression because of its sensitivity to chemical shift artifacts, especially at 3 Tesla (T). We investigated the utility of calculating the inversion time of the null point (TI null ) in pelvic DWI with spectral attenuated inversion recovery (SPAIR) at 3 T for obtaining adequate fat suppression. Thirteen volunteers underwent pelvic SPAIR-DWI using various SPAIR inversion delay times (TI) at 5-ms intervals in the range of the calculated TI null ±25 ms. The degree of fat suppression was evaluated into 3 grades and was compared among the various SPAIR-TIs with the calculated TI null . In 65 cases of prostatic disease, we evaluated the ratio of adequate fat suppression obtained using the calculated TI null . We obtained adequate fat suppression in all 13 volunteers and in 61 (94%) of the 65 patients using the calculated TI null . Fat suppression was best when the calculated TI null was used (P null (P null increased. In conclusion, this method of calculating the TI null may be useful for obtaining adequate fat suppression for pelvic SPAIR-DWI at 3T. (author)

Compatibility of detached divertor operation with robust edge pedestal performance

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Thermomechanical simulation of WEST actively cooled upper divertor

International Nuclear Information System (INIS)

Batal, T.; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-01-01

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m 2 . This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m 2 , and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m 2 . The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Thermomechanical simulation of WEST actively cooled upper divertor

Energy Technology Data Exchange (ETDEWEB)

Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-11-15

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Optimization and limitations of known DEMO divertor concepts

Energy Technology Data Exchange (ETDEWEB)

Reiser, Jens, E-mail: Jens.Reiser@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany); Rieth, Michael [Karlsruhe Institute of Technology, Institute for Applied Materials, P.O. Box 3640, 76021 Karlsruhe (Germany)

2012-08-15

Highlights: Black-Right-Pointing-Pointer Limitations of the materials. Black-Right-Pointing-Pointer Improved H{sub 2}O cooled divertor. Black-Right-Pointing-Pointer Improved He cooled divertor. - Abstract: In this work we will introduce and discuss improvements for two types of DEMO divertors based on known designs: (i) gas cooled designs and (ii) liquid coolant concepts. In a first step, the advantages and disadvantages of gas cooling as well as the necessity of a jet impingement to increase the heat transfer coefficients will be discussed. Further discussion deals with the pros and cons of liquid coolant concepts, like for example, liquid metal or water cooling. Thereafter, we will present two rather contrary DEMO divertor concepts which are based on today's knowledge on refractory materials science, fabrication and joining technology. The first improved concept uses water flowing through steel pipes, typically made of Eurofer steel. It is well known that using Eurofer at low temperatures is critical due to its severe embrittlement under neutron irradiation. Here we make a proposal how it could be possible to use the Eurofer steel anyway: the solution could consist in a limited operation period followed by an annealing cycle at 550 Degree-Sign C for a few hours during any maintenance shut down phases. The second design is based on the known helium cooling concept using jet impingement. Drawbacks of the actual He-cooled divertor design are small scale parts as well as the necessary high helium inlet temperature of about 600-800 Degree-Sign C which leads to the question: How can we deal with such high helium temperatures? This paper shows a solution for large scale components as well as a new thermal management for the helium outlet gas that we call 'cooling of the coolant'. Both concepts are discussed in terms of materials selection due to material limits and joining technology with a special focus on the material issue using already existing and

The importance of topographically corrected null models for analyzing ecological point processes.

Science.gov (United States)

McDowall, Philip; Lynch, Heather J

2017-07-01

Analyses of point process patterns and related techniques (e.g., MaxEnt) make use of the expected number of occurrences per unit area and second-order statistics based on the distance between occurrences. Ecologists working with point process data often assume that points exist on a two-dimensional x-y plane or within a three-dimensional volume, when in fact many observed point patterns are generated on a two-dimensional surface existing within three-dimensional space. For many surfaces, however, such as the topography of landscapes, the projection from the surface to the x-y plane preserves neither area nor distance. As such, when these point patterns are implicitly projected to and analyzed in the x-y plane, our expectations of the point pattern's statistical properties may not be met. When used in hypothesis testing, we find that the failure to account for the topography of the generating surface may bias statistical tests that incorrectly identify clustering and, furthermore, may bias coefficients in inhomogeneous point process models that incorporate slope as a covariate. We demonstrate the circumstances under which this bias is significant, and present simple methods that allow point processes to be simulated with corrections for topography. These point patterns can then be used to generate "topographically corrected" null models against which observed point processes can be compared. © 2017 by the Ecological Society of America.

Visible spectroscopy in the DIII-D divertor

International Nuclear Information System (INIS)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

Suzuki, Satoshi

2011-01-01

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

Novel Remarks on Point Mass Sources, Firewalls, Null Singularities and Gravitational Entropy

Science.gov (United States)

Perelman, Carlos Castro

2016-01-01

A continuous family of static spherically symmetric solutions of Einstein's vacuum field equations with a spatial singularity at the origin r = 0 is found. These solutions are parametrized by a real valued parameter λ (ranging from 0 to 1) and such that the radial horizon's location is displaced continuously towards the singularity ( r = 0 ) as λ increases. In the extreme limit λ = 1, the location of the singularity and horizon merges leading to a null singularity. In this extreme case, any infalling observer hits the null singularity at the very moment he/she crosses the horizon. This fact may have important consequences for the resolution of the fire wall problem and the complementarity controversy in black holes. An heuristic argument is provided how one might avoid the Hawking particle emission process in this extreme case when the singularity and horizon merges. The field equations due to a delta-function point-mass source at r = 0 are solved and the Euclidean gravitational action corresponding to those solutions is evaluated explicitly. It is found that the Euclidean action is precisely equal to the black hole entropy (in Planck area units). This result holds in any dimensions D ≥ 3.

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

International Nuclear Information System (INIS)

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

1994-07-01

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

Safety characteristics of the monolithic CFC divertor

International Nuclear Information System (INIS)

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

1994-01-01

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

Safety characteristics of the monolithic CFC divertor

Science.gov (United States)

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

1994-09-01

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

The WEST project: Current status of the ITER-like tungsten divertor

International Nuclear Information System (INIS)

Missirlian, M.; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

2014-01-01

Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues

The WEST project: Current status of the ITER-like tungsten divertor

Energy Technology Data Exchange (ETDEWEB)

Missirlian, M., E-mail: marc.missirlian@cea.fr; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

2014-10-15

Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

Investigation of electron parallel pressure balance in the scrape-off layer of deuterium-based radiative divertor discharges IN DIII-D

International Nuclear Information System (INIS)

Petrie, T.W.; Carlstrom, T.N.; Allen, S.L.

1996-10-01

Electron density, temperature, and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2, and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of ∼ 2 between the midplane separatrix and the X-point, and then an additional ∼3--5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power maintains good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate, and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7--30 eV

Design, R&D and commissioning of EAST tungsten divertor

Science.gov (United States)

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

2016-02-01

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

Plasma parameters in the COMPASS divertor during Ohmic plasmas

Energy Technology Data Exchange (ETDEWEB)

Dimitrova, M. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Dejarnac, R.; Stoeckel, J.; Havlicek, J.; Janky, F.; Panek, R. [Institute of Plasma Physics, Academy of Sciences of the Czech Republic v.v.i., Prague (Czech Republic); Popov, Ts.K. [Faculty of Physics, St. Kl. Ohridski University of Sofia (Bulgaria); Ivanova, P.; Vasileva, E. [Emil Djakov Institute of Electronics, Bulgarian Academy of Sciences, Sofia (Bulgaria); Kovacic, J. [Jozef Stefan Institute, Ljubljana (Slovenia)

2014-04-15

This paper reports on probe measurements of the electron energy distribution function and plasma potential in the divertor region of the COMPASS tokamak during D-shaped plasmas. The probe data have been processed using the novel first-derivative technique. A comparison with the results obtained by processing the same data with the classical probe technique, which assumes Maxwellian electron energy distribution functions is presented and discussed. In the vicinity of the inner and outer strike points of the divertor the electron energy distribution function can be approximated by a bi-Maxwellian, with a dominating low-energy electron population (4-7 eV) and a minority of higher energy electrons (12-25 eV). In the private flux region between the two strike points the electron energy distribution function is found to be Maxwellian with temperatures in the range of 7-10 eV. The comparative analysis using both techniques has allowed a better insight into the underlying physical processes at the divertor region of the COMPASS tokamak. (copyright 2014 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

VUV Spectroscopy in DIII-D Divertor

International Nuclear Information System (INIS)

Alkesh Punjabi; Nelson Jalufka

2004-01-01

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

A tangentially viewing visible TV system for the DIII-D divertor

International Nuclear Information System (INIS)

Fenstermacher, M.E.; Meyer, W.H.; Wood, R.D.; Nilson, D.G.; Ellis, R.; Brooks, N.H.

1997-01-01

A video camera system has been installed on the DIII-D tokamak for two-dimensional spatial studies of line emission in the lower divertor region. The system views the divertor tangentially at approximately the height of the X point through an outer port. At the tangency plane, the entire divertor from the inner wall to outside the DIII-D bias ring is viewed with spatial resolution of ∼1 cm. The image contains information from ∼90 deg of toroidal angle. In a recent upgrade, remotely controllable filter changers were added which have produced images from nominally identical discharges using different spectral lines. Software was developed to calculate the response function matrix of the optical system using distributed computing techniques and assuming toroidal symmetry. Standard sparse matrix algorithms are then used to invert the three-dimensional images onto a poloidal plane. Spatial resolution of the inverted images is 2 cm; higher resolution simply increases the size of the response function matrix. Initial results from a series of experiments with multiple identical discharges show that the emission from CII and CIII, which appears along the inner scrape-off layer above and below the X point during ELMing H mode, moves outward and becomes localized near the X point in radiative divertor operation induced by deuterium injection. copyright 1997 American Institute of Physics

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-15

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

Controlling marginally detached divertor plasmas

Science.gov (United States)

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

2017-06-01

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

EFFECT OF ION ∇ B DRIFT DIRECTION ON TURBULENCE FLOW AND FLOW SHEAR

International Nuclear Information System (INIS)

FENZI, C; McKEE, G.R; BURRELL, K.H; CARLSTROM, T.N; FONCK, R.J; GROEBNER, R.J

2003-01-01

The divertor magnetic geometry has a significant effect on the poloidal flow and resulting flow shear of turbulence in the outer region of L-mode tokamak plasmas, as determined via two-dimensional measurements of density fluctuations with Beam Emission Spectroscopy on DIII-D. Plasmas with similar parameters, except that in one case the ion (del)B drift points towards the divertor X-point (lower single-null, LSN), and in the other case, the ion (del)B drift points away from the divertor X-point (upper single-null, USN), are compared. Inside of r/a=0.9, the turbulence characteristics (amplitude, flow direction, correlation lengths) are similar in both cases, while near r/a=0.92, a dramatic reversal of the poloidal flow of turbulence relative to the core flow direction is observed in plasmas with the ion (del)B drift pointing towards the divertor X-point. No such flow reversal is observed in plasmas with the ion (del)B drift pointing away from the divertor X-point. This poloidal flow reversal results in a significantly larger local shear in the poloidal turbulence flow velocity in plasmas with the ion (del)B drift pointing towards the divertor X-point. Additionally, these plasmas locally exhibit significant dispersion, with two distinct and counter-propagating turbulence modes. Likewise, the radial correlation length of the turbulence is reduced in these plasmas, consistent with biorthogonal decomposition measurements of dominant turbulence structures. The naturally occurring turbulence flow shear in these LSN plasmas may facilitate the LH transition that occurs at an input power of roughly one-half to one-third that of corresponding plasmas with the ion (del)B drift pointing away from the X-point

Models for poloidal divertors

International Nuclear Information System (INIS)

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

1982-07-01

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

Models for poloidal divertors

Energy Technology Data Exchange (ETDEWEB)

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

1982-07-01

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

Vanadium alloys for the radiative divertor program of DIII-D

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1995-10-01

Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects

Innovative divertor concepts for LHD

International Nuclear Information System (INIS)

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

1994-07-01

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

On the dependence of energy confinement on elongation in Single Null divertor plasmas

International Nuclear Information System (INIS)

Weisen, H.; Martin, Y.; Moret, J. M.; Others

2002-03-01

An analysis of the effect of magnetic geometry on heat flux in a wide range of Single Null diverted discharge configurations with 1.5 κ ∼ 74, based on the configurations in the study. This is very close to dependences from the ITER database for discharge conditions, such as ELMy H-modes, obtained from a large number of experiments in various tokamak devices. (author)

Operating windows of pebble divertor

International Nuclear Information System (INIS)

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

2001-01-01

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

Divertor ‘death-ray’ explained: An artifact of a Langmuir probe operating at negative bias in a high-recycling divertor

International Nuclear Information System (INIS)

Brunner, D.; Umansky, M.V.; LaBombard, B.; Rognlien, T.D.

2013-01-01

The divertor ‘death-ray’, enhanced plasma pressure near the outer strike-point relative to ‘upstream’ values, was thought to correspond to axisymmetric increased divertor heat flux. Recent measurements on Alcator C-Mod show that the ‘death-ray’ is localized to biased Langmuir probes. Heat fluxes deduced from plasma-sheath theory and surface thermocouples agree in sheath-limited and moderate-recycling regimes. They diverge in high-recycling and detached regimes; surface thermocouples measure reduced heat flux while a ‘death-ray’ appears on Langmuir probes. The ‘death-ray’ is caused by the probe’s negative bias affecting the local flux tube. With the bias, electron heat flux to the probe surface is reduced. Thus, the local electron temperature is raised, enhancing neutral ionization and increasing the ion flux to the probe. The plasma fluid code UEDGE is used to simulate and reproduce many of the features of this integrated biased probe/divertor system

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

2017-01-01

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. (paper)

Versator divertor experiment: preliminary designs

International Nuclear Information System (INIS)

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

1984-08-01

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

Snowflake divertor configuration studies for NSTX-Upgrade

International Nuclear Information System (INIS)

Soukhanovskii, V.A.

2011-01-01

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

Towards fully authentic modelling of ITER divertor plasmas

International Nuclear Information System (INIS)

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

1991-01-01

Ignited next step tokamaks such as NET or ITER are expected to use a poloidal magnetic divertor to facilitate exhaust of plasma particles and energy. We report a development coupling together detailed computational models for both plasma and recycled neutral particle transport processes, to produce highly detailed and consistent design solutions. A particular aspect is involvement of an accurate specification of edge magnetic geometries, determined by an original equilibrium discretisation code, named LINDA. Initial results for a prototypical 22MA ITER double-null configuration are presented. Uncertainties in such modelling are considered, especially with regard to intrinsic physical scale lengths. Similar results produced with a simple, analytical treatment of recycling are also compared. Finally, a further extension allowing true oblique target sections is anticipated. (author) 8 refs., 5 figs

Model-based radiation scalings for the ITER-like divertors of JET and ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Brezinsek, S. [Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich (Germany); Federici, G. [EFDA PPP& T Department, D-85748 Garching (Germany)

2015-08-15

Effects of N-seeding in L-mode experiments in ASDEX Upgrade and JET are analysed numerically with the SOLPS5.0 code package. The modelling yields 3 qualitatively different radiative regimes with increasing N concentration, when initially attached outer divertor conditions are studied. The radiation pattern is observed to evolve asymmetrically, with radiation increasing first in the inner divertor, then in the outer divertor, and finally on closed field lines above the X-point. The properties of these radiative regimes are observed to be sensitive to cross-field drifts and they differ between the two devices. The modelled scaling of the divertor radiated power with the divertor neutral pressure is similar to an experimental scaling law for H-mode radiation. The same parametric dependencies are not observed in simulations without drifts.

Optimization for steady-state and hybrid operations of ITER by using scaling models of divertor heat load

International Nuclear Information System (INIS)

Murakami, Yoshiki; Itami, Kiyoshi; Sugihara, Masayoshi; Fujieda, Hirobumi.

1992-09-01

Steady-state and hybrid mode operations of ITER are investigated by 0-D power balance calculations assuming no radiation and charge-exchange cooling in divertor region. Operation points are optimized with respect to divertor heat load which must be reduced to the level of ignition mode (∼5 MW/m 2 ). Dependence of the divertor heat load on the variety of the models, i.e., constant-χ model, Bohm-type-χ model and JT-60U empirical scaling model, is also discussed. The divertor heat load increases linearly with the fusion power (P FUS ) in all models. The possible highest fusion power much differs for each model with an allowable divertor heat load. The heat load evaluated by constant-χ model is, for example, about 1.8 times larger than that by Bohm-type-χ model at P FUS = 750 MW. Effect of reduction of the helium accumulation, improvements of the confinement capability and the current-drive efficiency are also investigated aiming at lowering the divertor heat load. It is found that NBI power should be larger than about 60 MW to obtain a burn time longer than 2000 s. The optimized operation point, where the minimum divertor heat load is achieved, does not depend on the model and is the point with the minimum-P FUS and the maximum-P NBI . When P FUS = 690 MW and P NBI = 110 MW, the divertor heat load can be reduced to the level of ignition mode without impurity seeding if H = 2.2 is achieved. Controllability of the current-profile is also discussed. (J.P.N.)

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

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

1991-01-01

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

Textor bundle divertor

International Nuclear Information System (INIS)

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

1982-01-01

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

TEXTOR bundle divertor

International Nuclear Information System (INIS)

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

1982-01-01

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

Divertor cooling device

International Nuclear Information System (INIS)

Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

1993-01-01

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

Innovations in the LHD divertor program

International Nuclear Information System (INIS)

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

1995-01-01

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

The WEST programme: Minimizing technology and operational risks of a full actively cooled tungsten divertor on ITER

Energy Technology Data Exchange (ETDEWEB)

Grosman, André, E-mail: andre.grosman@cea.fr [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Bucalossi, Jérôme; Doceul, Louis [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Escourbiac, Frédéric [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Lipa, Manfred [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Merola, Mario [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Missirlian, Marc [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France); Pitts, Richard A. [ITER Organization, Cadarache, 13115 St. Paul-lez-Durance (France); Samaille, Franck; Tsitrone, Emmanuelle [CEA, IRFM, F-13108 Saint-Paul-Lez-Durance (France)

2013-10-15

Highlights: ► The WEST programme is a unique opportunity to experience the industrial scale manufacture of tungsten plasma-facing components similar to the ITER divertor ones. ► In Tore Supra, it will bring important know how for actively cooled W divertor operation. ► This can be done by a reasonable modification of the Tore Supra tokamak. ► A fast implementation of the project would make this information available in due time. ► This allows a significant contribution to the W ITER divertor risk minimization in its manufacturing and operation phase. -- Abstract: The WEST programme consists in transforming the Tore Supra tokamak into an X point divertor device, while taking advantage of its long discharge capability. This is obtained by inserting in vessel coils to create the X point while adapting the in-vessel elements to this new geometry. This will allow the full tungsten divertor technology to be used on ITER to be tested in anticipation of its use on ITER under relevant heat loading conditions and pulse duration. The early manufacturing of a significant industrial series of ITER-similar W plasma-facing units will contribute to the ITER divertor manufacturing risk mitigation and to that associated with early W divertor plasma operation on ITER.

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

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

1991-04-01

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

A survey of problems in divertor and edge plasma theory

International Nuclear Information System (INIS)

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

1992-01-01

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

Bursty fluctuation characteristics in SOL/divertor plasmas of Large Helical Device

International Nuclear Information System (INIS)

Ohno, N.; Masuzaki, S.; Morisaki, T.; Ohyabu, N.; Komori, A.; Budaev, V.P.; Miyoshi, H.; Takamura, S.

2006-10-01

Bursty electrostatic fluctuation in the scrape off layer (SOL) and the divertor region of the Large Helical Device (LHD) have been investigated by using a Langmuir probe array on a divertor plate and a reciprocating Langmuir probe. Large positive bursty events were often observed in the ion saturation current measured with a divertor probe near the divertor leg at which the magnetic line of force connected to the area of a low-field side with a short connection length. Condition averaging result of the positive bursty events indicates the intermittent feature with a rapid increase and a slow decay is similar to that of plasma blobs observed in tokamaks. On the other hand, at a striking point with a long connection length, negative spikes were observed. Statistical analysis based on probability distribution function (PDF) was employed to investigate the bursty fluctuation property. The observed scaling exponents disagree with the predictions for the self-organized criticality (SOC) paradigm. (author)

Scrape-off layer and divertor theory meeting: Proceedings

International Nuclear Information System (INIS)

1994-03-01

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

Design, fabrication, and testing of a helium-cooled module for the ITER divertor

International Nuclear Information System (INIS)

Baxi, C.B.; Smith, J.P.; Youchison, D.

1994-08-01

The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

Thermal effects of divertor sweeping in ITER

International Nuclear Information System (INIS)

Wesley, J.C.

1992-01-01

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

Engineering conceptual design of CFETR divertor

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Plasma structure change and intermittent fluctuation near magnetic island X-point under detached plasma condition in LHD

International Nuclear Information System (INIS)

Ohno, N.; Tsuji, Y.; Tanaka, H.; Masuzaki, S.; Kobayashi, M.; Akiyama, T.; Morisaki, T.; Motojima, G.; Narushima, Y.

2014-10-01

Plasma profiles and intermittent fluctuations near the helical divertor X-point and on a divertor plate were investigated using a fast scanning Langmuir probe and a probe array embedded on a divertor plate in detached divertor condition that was sustained by applying a resonant magnetic perturbation (RMP) field in LHD. When the RMP induced magnetic island X-point (n/m = 1/1) is located near the helical divertor X-point, the reduction of particle flux accompanied by the plasma detachment occurred near the helical divertor X-point (n/m = 2/10), which leads to the reduction of the particle flux at the strike point on the divertor plate. We also found that when the divertor plasma turned to be the detached condition, the enhanced plasma fluctuations were confirmed between the helical divertor X-point and ergodic region, which exhibited a dynamic behavior having a large amount of positive-spike components with highly intermittent property. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

To predict the particle and heat controllability in the divertor of tokamak reactors such as ITER and to optimize the divertor design, comprehensive simulations by integrated modelling with taking in various physical processes are indispensable. For the design study of ITER divertor, the modelling codes such as B2, UEDGE and EDGE2D have been developed, and their results have contributed to the evolution of the divertor concept. In Japan Atomic Energy Agency (JAEA), SOL/divertor codes have also been developed for the interpretation and the prediction on behaviours of plasmas, neutrals and impurities in the SOL/divertor regions. The code development is originally carried out since physics models can be verified quickly and flexibly under the circumstance of close collaboration with JT-60 team. Figure 1 shows our code system, which consists of the 2 dimensional fluid code SOLDOR, the neutral Monte Carlo (MC) code NEUT2D, and the impurity MC code IMPMC. The particle simulation code PARASOL has also been developed in order to establish the physics modelling used in fluid simulations. Integration of SOLDOR, NEUT2D and IMPMC, called the '' SONIC '' code, is being carried out to simulate self-consistently the SOL/divertor plasmas in present tokamaks and in future devices. Combination of the SOLDOR and NEUT2D was completed, which has the features such as 1) high-resolution oscillation-free scheme in solving fluid equations, 2) neutral transport calculation under the fine meshes, 3) success in reduction of MC noise, 4) optimization on the massive parallel computer, etc. The simulation reproduces the X-point MARFE in the JT-60U experiment. It is found that the chemically sputtered carbon at the dome causes the radiation peaking near the X-point. The performance of divertor pumping in JT-60U is evaluated from the particle balances. We also present the divertor designing of JT-60SA, which is the modification program of JT-60U to establish high beta steady-state operation. To

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

International Nuclear Information System (INIS)

Ribeiro, T.; Scott, B.

2007-01-01

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

TCV divertor upgrade for alternative magnetic configurations

Directory of Open Access Journals (Sweden)

H. Reimerdes

2017-08-01

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

Particle recirculation in the ergodic divertor of Tore Supra

International Nuclear Information System (INIS)

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

1999-01-01

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

Numerical studies on divertor experiments

International Nuclear Information System (INIS)

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

1988-04-01

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

Comparative divertor-transport study for helical devices

International Nuclear Information System (INIS)

Feng, Y.; Sardei, F.; Kobayashi, M.

2008-10-01

Using the island divertors (ID) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine-size following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. Revealed is the fundamental role of the low-order magnetic islands in both divertor concepts. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime which is absent from W7-AS and LHD is expected for W7-X. Topics addressed are restricted to the basic function elements of a divertor such as particle flux enhancement and impurity retention. In particular, the divertor function on reducing the influx of intrinsic impurities is examined for all the three devices under different divertor plasma conditions. Special attention is paid to examining the island screening potential of intrinsic impurities which has been predicted for all the three devices under high divertor collisionality conditions. The results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. (author)

Manufacturing and joining technologies for helium cooled divertors

International Nuclear Information System (INIS)

Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

2014-01-01

Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

Development of divertor remote maintenance system

Energy Technology Data Exchange (ETDEWEB)

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

1998-04-01

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

Development of divertor remote maintenance system

International Nuclear Information System (INIS)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

1999-06-01

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

Theory of Advanced Magnetic Divertors

Science.gov (United States)

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

2013-10-01

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

Feasibility study for an engineering concept of a stainless steel/copper divertor plate protected by W-5 Re alloy or graphite armor

International Nuclear Information System (INIS)

Renda, V.; Federici, G.; Papa, L.

1988-01-01

The latest Joint Research Centre (JRC)-Ispra proposal is presented to support the design of a divertor concept that has long been considered the most crucial component of the plasma impurity control system for the Next Europen Torus (NET) tokamak fusion reactor. Because of the harsh tokamak environment, the divertor panel is the plasma facing component that suffers the most severe loading conditions, such as high thermal stresses, thermal fatigue, severe erosion rates and neutron damage. An analysis of a new divertor panel concept has evolved from the previous studies carried out at JRC-Ispra. The materials considered in this study are AISI 316 stainless steel for the cooling tubes, pure copper for the heat sink, and W-5 Re alloy or graphite for the protective armor. The panel is cooled by pressurized water circulation in U-tubes. A preliminary thermo-hydraulic analysis has been carried out to evaluate a set of reference parameters, such as optimum coolant velocity, maximum outlet water temperature, convective heat exchange coefficient, and the expected pressure drops in the channels. Thermal and mechanical calculations, performed by using the finite element technique, showed encouraging results about the engineering feasibility of the pressure boundary of the divertor for loading conditions similar to those of NET double null, assumed as the reference mainframe

Fabrication of divertor cassette for ITER

International Nuclear Information System (INIS)

Sanguinetti, G.P.

2008-01-01

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

Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Porter, G.D.; Rognlien, T.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

2001-01-01

Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the ExB drift speed, ion diamagnetism and nite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

Turbulence studies in tokamak boundary plasmas with realistic divertor geometry

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Por, G.D. ter; Rognlien, T.D.; Ryutov, D.D.; Myra, J.R.; D'Ippolito, D.A.; Moyer, R.; Groebner, R.J.

1999-01-01

Results are presented from the 3D nonlocal electromagnetic turbulence code BOUT and the linearized shooting code BAL for studies of turbulence in tokamak boundary plasmas and its relationship to the L-H transition, in a realistic divertor plasma geometry. The key results include: (1) the identification of the dominant resistive X-point mode in divertor geometry and (2) turbulence suppression in the L-H transition by shear in the E x B drift speed, ion diamagnetism and finite polarization. Based on the simulation results, a parameterization of the transport is given that includes the dependence on the relevant physical parameters. (author)

A large divertor manipulator for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

A review of progress towards radiative divertor

International Nuclear Information System (INIS)

Zagorski, Roman

1997-07-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

International Nuclear Information System (INIS)

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

2014-01-01

In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription

Plasma flow in the DIII-D divertor

International Nuclear Information System (INIS)

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

1998-07-01

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

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Study on poloidal field coil optimization and equilibrium control of ITER

International Nuclear Information System (INIS)

Shinya, Kichiro; Sugihara, Masayoshi; Nishio, Satoshi

1989-03-01

The purpose of this report is to present general features of the poloidal field coil optimization for the ITER plasma, flexibility analysis for various plasma options and some other aspect of the equilibrium control which is required for understanding plasma operation in more detail. Double null divertor plasma was selected as a main object of the optimization. Single null divertor plasma was assumed to be an alternative, because single null divertor plasma can be operational within the amounts of the total stored energy and ampere-turns of the double null divertor plasma, if it is shaped appropriately. Plasma parameters used in the present analysis are mainly those employed in the preliminary study by the Basic Device Engineering group of the ITER design team. The most part of the optimization study, however, utilizes the parameters proposed for discussion by the Japan team before starting joint design work at Garching. Plasma shape, and solenoid coil shape and size, which maximize available flux swing with reasonable amounts of the stored energy and ampere-turns, are discussed. Location and minimum number of the poloidal field coils with adequate shaping controllability were also discussed for various plasma options. Some other aspect of the equilibrium control, such as separatrix swing, moving null point operation during plasma heating and possible range of li, were evaluated and the guideline for the engineering design was proposed. Finally, fusion power output was estimated for the different pressure profiles and combinations of the average density and temperature, and the magnetic quantities of the scrape-off region was calculated to be available for the future divertor analysis. (author)

Thermal transients due to sweeping of the separatrix on the monoblock divertor concept for ITER

International Nuclear Information System (INIS)

Renda, V.; Papa, L.; Soria, A.

1991-01-01

The ITER divertor plate considered in the present study is the monoblock design option, consisting of an armour of CFC-SEP-Carb graphite tiles, crossed by the tubes of the water cooling system made in Mo-Re alloy. Preliminary steady-state calculations for a peak flux of 15 MW/m 2 showed that the allowable thickness to limit the maximum temperature to 1273 K (1000degC) is about 5 mm. This small value reduces the lifetime of the plate, due to the expected erosion rate, to an unacceptable value from the engineering standpoint. A sweeping of the separatrix has been proposed to reduce the erosion of the protective armour and to lessen the thermomechanical effects of the localized peak surface heat flux. A rotation of the null points of the separatrix of 30 mm radius with a frequency of 0.3 Hz for a surface heat flux of 15 MW/m 2 was assumed as nominal working condition. Several scenarios were considered as off-normal conditions: the loss of sweeping accident, the change in frequency from 0.3 to 0.1 Hz and the change of the peak of the surface heat flux from 15 to 30 MW/m 2 . The results related to the nominal condition show that a 16 mm thick armour could be allowed; this value should ensure an acceptable lifetime for the divertor plate. The loss of sweeping accident leads the surface temperature to reach about 2273 K in few seconds; the change in frequency raises the maximum temperature of 423 K, but its range doubles; the change in peak flux leads to a maximum temperature of about 2373 K. (author)

The ITER divertor cassette project meeting

International Nuclear Information System (INIS)

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

1999-01-01

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

Detached divertor plasmas in Alcator C-Mod: A study of the role of atomic physics

International Nuclear Information System (INIS)

Lipschultz, B.; Boswell, C.; Goetz, J.A.

1999-01-01

Detailed profiles of the volumetric recombination occurring in Alcator C-Mod plasmas are presented. During detachment the recombination sink is compared to the divertor plate sink as well as the divertor ion source. Depending on plasma conditions, volume recombination removes between 10 and 75% of the ions before they reach the plates. A second, equally important process that leads to a drop in plate ion current is inferred to be a reduction in divertor ion source, which is correlated with a drop in power flowing into the ionization region and the pressure loss of detachment. For high n e the divertor recombination can cross the separatrix near the x-point, cool the core and lead to a disruption. Experimental measurements show a difference in ion and neutral velocities for H-mode detached plasmas. The resulting ion-neutral collisions are found to be more efficacious than recombination in removing momentum from the ions. The neutral component of volumetric power emission from the divertor has been measured by means of a novel filtering technique to be substantial (∼ 20% of the total divertor volumetric emission). (author)

Plasma facing components integration studies for the WEST divertor

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-15

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

Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

International Nuclear Information System (INIS)

Mai, L.P.; Malick, F.S.

1981-01-01

The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented

A Lithium Vapor Box Divertor Similarity Experiment

Science.gov (United States)

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

2017-10-01

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

Single Null Negative Triangularity Tokamak for Power Handling

Science.gov (United States)

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.

Commissioning and preliminary operation of HL-2A tokamak

International Nuclear Information System (INIS)

Liu Dequan; Liu Yong; Yan Jianchen; Cao Zeng; Yang Qingwei; Zhou Caipin; Li Xiaodong

2005-01-01

HL-2A is a divertor tokamak located at new site of Southwestern Institute of Physics (SWIP), Chengdu, China. It can be operated in double-null and single-null divertor with closed configurations. The effects of the divertor on impurity behaviors, MHD instabilities, transport, wall conditioning and divertor physics are key issues to be studied during the first step operation on HL-2A. Preliminary experiments with limiter and single null divertor configurations were carried out in 2003. Main parameters achieved are the plasma current 168 KA, duration time of 920 ms and plasma linear-averaged density of 1.7 x 10 19 m -3 . The impurities, especially of low Z, are clearly decreased during the divertor experiments

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Towards a physics-integrated view on divertor pumping

International Nuclear Information System (INIS)

Day, Chr.; Gleason-González, C.; Hauer, V.; Igitkhanov, Y.; Kalupin, D.; Varoutis, S.

2014-01-01

Highlights: • Physics-integrated design approaches are to be preferred over approaches based on simple requirement lists. • A physics-integrated assessment is presented for the divertor vacuum pumping system based on detachment onset conditions for the divertor. • This approach considers density dependent pump albedo to reflect the effects of gas recycling at the divertor and the changes in flow regime with density. • A comparison with DEMO indicates that the divertor pumping system for a pulsed DEMO scales less than linearly with fusion power. - Abstract: One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset

Conceptual design of CFETR divertor remote handling compatible structure

International Nuclear Information System (INIS)

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

2016-01-01

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

Conceptual design of CFETR divertor remote handling compatible structure

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Rapidly Moving Divertor Plates In A Tokamak

International Nuclear Information System (INIS)

Zweben, S.

2011-01-01

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

Role of molecular effects in divertor plasma recombination

Directory of Open Access Journals (Sweden)

A.S. Kukushkin

2017-08-01

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

Island divertor studies on W7-AS

International Nuclear Information System (INIS)

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

1997-01-01

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

Atomic and molecular processes in JT-60U divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

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

1997-01-01

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

Driving mechanism of SOL plasma flow and effects on the divertor performance in JT-60U

International Nuclear Information System (INIS)

Asakura, N.

2002-01-01

SOL plasma flow plays an important role in the plasma transport along the field lines, and influences control of the divertor plasma and impurity ions. Recently, mechanisms producing the SOL flow such as drifts produced by electric field and pressure gradient are pointed out. In JT-60U, three reciprocating Mach probes were installed at the high-field-side (HFS) baffle, low-field-side (LFS) midplane and just below the X-point. The measurements of the SOL flow and plasma profiles both at the HFS and LFS, for the first time, found out the SOL flow pattern and its driving mechanism. 'Flow reversal' was found near the separatrix of the HFS and LFS. Radial profiles of the SOL flow were similar to those calculated numerically using the UEDGE code with the plasma drifts included. SOL particle fluxes towards the HFS and LFS divertors were, for the first time, evaluated. Important physics issues for the divertor design and operation, such as in-out asymmetries of the heat and particle fluxes, and control of impurity ions with intense gas puff and divertor pump (puff and pump), were investigated. (author)

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

International Nuclear Information System (INIS)

Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

2006-01-01

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

Tungsten nano-tendril growth in the Alcator C-Mod divertor

International Nuclear Information System (INIS)

Wright, G.M.; Brunner, D.; Labombard, B.; Lipschultz, B.; Terry, J.L.; Whyte, D.G.; Baldwin, M.J.; Doerner, R.P.

2012-01-01

Growth of tungsten nano-tendrils (‘fuzz’) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50–100 nm) and the depth of the layer (600 ± 150 nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices. (letter)

Comparison between stellarator and tokamak divertor transport

International Nuclear Information System (INIS)

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

2010-11-01

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

The divertor remote maintenance project

International Nuclear Information System (INIS)

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

2001-01-01

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

The divertor remote maintenance project

International Nuclear Information System (INIS)

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

1999-01-01

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

FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

International Nuclear Information System (INIS)

O'NEIL, RC; STAMBAUGH, RD

2002-01-01

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

'EU divertor celebration day'

International Nuclear Information System (INIS)

Merola, M.

2002-01-01

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

Advantages and Challenges of Radiative Liquid Lithium Divertor

Science.gov (United States)

Ono, Masayuki

2017-10-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

ITER tungsten divertor design development and qualification program

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper.

High temperature divertor plasma operation

International Nuclear Information System (INIS)

Ohyabu, Nobuyoshi.

1991-02-01

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

Method for comparison of tokamak divertor strike point data with magnetic perturbation models

Czech Academy of Sciences Publication Activity Database

Cahyna, Pavel; Peterka, Matěj; Nardon, E.; Frerichs, H.; Pánek, Radomír

2014-01-01

Roč. 54, č. 6 (2014), 064002-064002 ISSN 0029-5515. [International Workshop on Stochasticity in Fusion Plasmas /6./. Jülich, 18.03.2013-20.03.2013] R&D Projects: GA ČR GAP205/11/2341 Institutional support: RVO:61389021 Keywords : divertor * resonant magnetic perturbation Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 3.062, year: 2014 http://iopscience.iop.org/0029-5515/54/6/064002/pdf/0029-5515_54_6_064002.pdf

Observation of Dust in DIII-D Divertor and SOL by Visible Imaging

International Nuclear Information System (INIS)

Rudakov, D L; West, W P; Groth, M; Yu, J H; Wong, C C; Boedo, J A; Brooks, N H; Evans, T E; Fenstermacher, M E; Hollmann, E M; Hyatt, A W; Lasnier, C J; McLean, A G; Moyer, R A; Pigarov, A; Smirnov, R; Solomon, W M; Watkins, J G

2007-01-01

Dust is commonly found in fusion devices. Though generally of no concern in the present day machines, dust may pose serious safety and operational concerns for ITER. Micron-size dust usually dominates the samples collected from tokamaks. During a plasma discharge micron-size dust particles can become highly mobile and travel over distances of a few meters. Once inside the plasma, dust particles heat up to over 3000 K and emit thermal radiation that can be detected by visible imaging techniques. Observations of naturally occurring and artificially introduced dusts have been performed in DIII-D divertor and scrape-off layer (SOL) using standard frame rate CMOS cameras, a gated-intensified CID camera, and a fast-framing CMOS camera. In the first 2-3 plasma discharges after a vent with personnel entry inside the vacuum vessel ('dirty vent') dust levels were quite high with thousands of particles observed in each discharge. Individual particles moving at velocities of up to a few hundred m/s and breakup of larger particles into pieces were observed. After about 15 discharges dust was virtually gone during the stationary portion of a discharge, and appeared at much reduced levels during the plasma initiation and termination phases. After a few days of plasma operations (about 70 discharges) dust levels were further reduced to just a few observed events per discharge except in discharges with current disruptions that produced significant amounts of dust. An injection of a few milligram of micron-size (6 micron median diameter) carbon dust into a high-power lower single-null ELMing H-mode discharge with strike points swept across the lower divertor floor was performed. A significant increase of the core carbon radiation was observed for about 250 ms after the injection, as the total radiated power increased twofold. Dust particles from the injection were observed by the fast framing camera in the outboard SOL near the midplane. The amount of dust observed by the fast

Divertor plasma studies on DIII-D: Experiment and modeling

International Nuclear Information System (INIS)

West, W.P.; Brooks, N.H.; Allen, S.L.

1996-09-01

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

On the Robinson theorem and shearfree geodesic null congruences

International Nuclear Information System (INIS)

Tafel, J.

1985-01-01

Null electromagnetic fields and shearfree geodesic null congruences in curved and flat spacetimes are studied. We point out some mathematical problems connected with the validity of the Robinson theorem. The problem of finding nonanalytic twisting congruences in the Minkowski space is reduced to the construction of holomorphic functions with specific boundary conditions. (orig.)

Characteristics of the Secondary Divertor on DIII-D

Science.gov (United States)

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

2009-11-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Reactor application of an improved bundle divertor

International Nuclear Information System (INIS)

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

1978-11-01

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

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

Science.gov (United States)

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

2017-12-01

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

First measurements of the ion energy distribution at the divertor strike point during DIII-D disruptions

International Nuclear Information System (INIS)

Parks, P.B.; Brooks, N.H.; West, W.P.; Wong, C.P.C.; Bastasz, R.; Wampler, W.R.; Whyte, D.

1995-12-01

Plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV

Modeling detachment physics in the NSTX snowflake divertor

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

Development of a radiative divertor for DIII-D

International Nuclear Information System (INIS)

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

1995-01-01

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

Development of a radiative divertor for DIII-D

Energy Technology Data Exchange (ETDEWEB)

Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

1995-04-01

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

An Asdex-type divertor for ITER

International Nuclear Information System (INIS)

Fowler, T.K.

1989-01-01

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

Divertor plasma physics experiments on the DIII-D tokamak

International Nuclear Information System (INIS)

Mahdavi, M.A.; Allen, S.L.; Evans, T.E.

1996-10-01

In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model

Physics design and experimental study of tokamak divertor

International Nuclear Information System (INIS)

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

2007-06-01

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

Measurements of flows in the DIII-D divertor by Mach probes

International Nuclear Information System (INIS)

Boedo, J.A.; Lehmer, R.; Moyer, R.A.; Watkins, J.G.; Porter, G.D.; Evans, T.E.; Leonard, A.W.; Schaffer, M.J.

1998-06-01

First measurements of Mach number of background plasma in the DIII-D divertor are presented in conjunction with temperature T e and density n e using a fast scanning probe array. To validate the probe measurements, the authors compared the T e , n e and J sat data to Thomson scattering data and find good overall agreement in attached discharges and some discrepancy for T e and n e in detached discharges. The discrepancy is mostly due to the effect of large fluctuations present during detached plasmas on the probe characteristic; the particle flux is accurately measured in every case. A composite 2-D map of measured flows is presented for an ELMing H-mode discharge and they focus on some of the details. They have also documented the temperature, density and Mach number in the private flux region of the divertor and the vicinity of the X-point, which are important transition regions that have been little studied or modeled. Background parallel plasma flows and electric fields in the divertor region show a complex structure

3D modelling of the island divertor for W7-AS

International Nuclear Information System (INIS)

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

1996-01-01

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

Thermal properties of redeposition layers in the JT-60U divertor region

International Nuclear Information System (INIS)

Ishimoto, Y.; Gotoh, Y.; Arai, T.; Masaki, K.; Miya, N.; Oyama, N.; Asakura, N.

2006-01-01

Thermal properties of the redeposition layer on the inner plate of the W-shaped divertor of JT-60U have been measured with laser flash method so as to estimate transient heat loads onto the divertor. Morphology analysis of the redeposition layer was conducted with a scanning electron microscope. Measurement of a redeposition layer sample of more than 200 μm thick, which had been produced near the most frequent striking point, showed following results: (1) the bulk density of the redeposition layer is about half of that of carbon fiber composite material; (2) the specific heat of the layer is roughly equal to that of the isotropic graphite; (3) the thermal conductivity of the redeposition layer is two orders of magnitude smaller than that of the carbon fiber composite. This low thermal conductivity of the redeposition layer is considered to be caused by a low graphitization degree of the redeposition layer. The difference between the divertor heat loads and the loss of the plasma stored energy becomes smaller taking account of thermal properties of the redeposition layer on the inner divertor, whereas estimated heat loads due to the ELMs is still larger than the loss. This is probably caused by the poloidal distribution of the thermal properties

Divertor cassette movers prototypes for ITER

International Nuclear Information System (INIS)

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

1998-01-01

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

Divertor Heat Flux Reduction by Resonant Magnetic Perturbations in the LHD-Type Helical DEMO Reactor

Energy Technology Data Exchange (ETDEWEB)

Yanagi, N.; Sagara, A.; Goto, T.; Masuzaki, S.; Miyazawa, J., E-mail: yanagi@lhd.nifs.ac.jp [National Institute for Fusion Science, Toki (Japan)

2012-09-15

Full text: The conceptual design studies of the LHD-type helical fusion DEMO reactor, FFHR-d1, are progressing steadfastly. The LHD-type heliotron magnetic configuration equipped with the built- in helical divertors has a potential to realize low divertor heat flux in spatial average. However, the toroidal asymmetry may give more than a couple of times higher peak heat flux at some locations, as has been experimentally observed in LHD and confirmed by magnetic field-line tracing. By providing radiation dispersion accompanied with a plasma detachment, the heat flux may decrease significantly though the compatibility with a good core plasma confinement is an important issue to be explored. Whereas the engineering difficulties for developing materials to be used under the neutron environment require even further decrease of the heat flux (even though the heliotron is a unique configuration that divertor plates be largely shielded from the direct irradiation of neutrons by breeder blankets). In this respect, we proposed, in the last IAEA FEC, a new strike point sweeping scheme using a set of auxiliary helical coils, termed helical divertor (HD) coils. The HD coils carrying a few percent of the current amplitude of the main helical coils sweep the divertor strike points without altering the core plasma. Though this scheme is effective in dispersing the heat flux in the poloidal direction, the toroidal asymmetry still remains. The AC operation may also give unforeseen engineering difficulties. We here propose that the peak heat flux be mitigated using RMP fields in steady-state. The magnetic field-lines are numerically traced in the vacuum configuration and their footprints coming to the divertor regions are counted. Their fraction plotted as a function of the toroidal angle indicates that the peak heat flux be mitigated to {approx} 20 MW per square meters at 3 GW fusion power generation without having radiation dispersion when an RMP field is applied. We note that the

Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

International Nuclear Information System (INIS)

Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

1991-01-01

The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

Design integration of liquid surface divertors

International Nuclear Information System (INIS)

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

2004-01-01

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

Simulation of the ASDEX divertor performance after hardening

International Nuclear Information System (INIS)

Schneider, W.; Lackner, K.; Neuhauser, J.; Wunderlich, R.

1985-05-01

Two combined computer models - a fluid description of the plasma scrape-off layer (SOLID) and a Monte-Carlo code for the neutral gas dynamics (DEGAS) - are used to assess changes in the divertor performance expected from the modifications in geometry needed for hardening the ASDEX divertor chamber for long-pulse, high-power heating. Stand-alone DEGAS calculations with assumed fixed scrape-off plasma parameters predict a doubling of the neutral escape probability, which, however, still remains so low, that achievement of the high divertor recycling regime can be expected over roughly the same operational regime as before modifications. This conclusion is also supported by fully self-consistent calculations with the combined model. Due to the reduced divertor, a significant reduction is predicted in the divertor time constant, which is expected to affect transient phenomena. (orig.)

CHANGES IN EDGE AND SCRAPE-OFF LAYER PLASMA BEHAVIOE DUE TO VAARIATION IN MAGNETIC BALANCE IN DIII-D

International Nuclear Information System (INIS)

PETRIE, T.W.; WATKINS, J.G.; BAYLOR, L.R.; BROOKS, N.H.; FENSTERMACHER, M.E.; HYATT, A.W.; JACKSON, G.L.; LASNIER, C.J.; LEONARD, A.W.; PIGAROV, A.YU.; RENSINK, M.E.; ROGNLIEN, T.D.; SCHAFFER, M.J.; WOLF, N.S.

2002-01-01

Changes in the divertor magnetic balance in DIII-D H-mode plasmas affects core, edge, and divertor plasma behavior. Both the pedestal density n e,PED and plasma stored energy W T were sensitive to changes in magnetic balance near the double-null (DN) configuration, e.g., both decreased 20%-30% when the DN shifted to a slightly unbalanced DN, where the B x (del)B drift direction pointed away from the main X-point. Recycling at each of the four divertor targets was sensitive to changes in magnetic balance and the B x (del)B drift direction. The poloidal distribution of the recycling in DN is in qualitative agreement with the predictions of UEDGE modeling with particle drifts included. The particle flux at the inner divertor target is shown to be much more sensitive to magnetic balance than the particle flux at the outer divertor target near the DN shape. These results suggest possible advantages and drawbacks for balanced DN operation

A solid tungsten divertor for ASDEX Upgrade

International Nuclear Information System (INIS)

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

2011-01-01

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

Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

International Nuclear Information System (INIS)

Kirschner, A.; Matveev, D.; Borodin, D.; Airila, M.; Brezinsek, S.; Groth, M.; Wiesen, S.; Widdowson, A.; Beal, J.; Esser, H.G.; Likonen, J.; Bekris, N.; Ding, R.

2015-01-01

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented

Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

Energy Technology Data Exchange (ETDEWEB)

Kirschner, A., E-mail: a.kirschner@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Matveev, D.; Borodin, D. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Airila, M. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Brezinsek, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Groth, M. [Aalto University, Otakaari 4, 02015 Espoo (Finland); Wiesen, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Widdowson, A. [Culham Centre for Fusion Energy, Abingdon OX14 3DB (United Kingdom); Beal, J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Esser, H.G. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Likonen, J. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Bekris, N. [Karlsruhe Institute of Technology, Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, Bau 451, 76344 Eggenstein-Leopoldshafen (Germany); Ding, R. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei, Anhui 230031 (China)

2015-08-15

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

Operation method for thermonuclear device and divertor for it

International Nuclear Information System (INIS)

Kotake, Michiko; Yoshioka, Ken; Fukumoto, Hideshi; Okazaki, Takashi; Kinoshita, Shigemi; Takeuchi, Kazuhiro.

1992-01-01

Divertor plates are disposed subsequently along with circumferential direction of a vacuum vessel in a region where magnetic fluxed generated from the divertor coils are injected toward a container wall. Each of the divertor plates is moved in a state that the injection position of the magnetic fluxes enter to the vacuum vessel is kept constant. Alternatively, each of the divertor plates is inclined at an angle facing the injection direction of plasma particle fluxes, or it is inclined so that the angle between the injection surface and the magnetic fluxes makes an acute angle. Since each of the divertor coils is moved in the state of keeping the injection position of the magnetic fluxes during firing of plasmas, in other words, with on change of the current of the divertor coils, the position of the magnetic fluxed is kept at a predetermined condition. Accordingly, charged particles are prevented from concentrating locally without causing eddy current in the coils and the vacuum vessel, which can contribute to the reduction of the wear of the divertor plates. (N.H.)

Divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

Norajitra, P.; Ihli, T.; Janeschitz, G.; Abdel-Khalik, S.; Mazul, I.; Malang, S.

2007-01-01

The development of a divertor concept for post-ITER fusion power plants is deemed to be an urgent task to meet the EU Fast Track scenario. Developing a divertor is particularly challenging due to the wide range of requirements to be met including the high incident peak heat flux, the blanket design with which the divertor has to be integrated, sputtering erosion of the plasma-facing material caused by the incident a particles, radiation effects on the properties of structural materials, and efficient recovery and conversion of the divertor thermal power (∝15% of the total fusion thermal power) by maximizing the coolant operating temperature while minimizing the pumping power. In the course of the EU PPCS, three near-term (A, B and AB) and two advanced power plant models (C, D) were investigated. Model A utilizes a water-cooled lead-lithium (WCLL) blanket and a water-cooled divertor with a peak heat flux of 15 MW/m 2 . Model B uses a He-cooled ceramics/beryllium pebble bed (HCPB) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model AB uses a He-cooled lithium-lead (HCLL) blanket and a He-cooled divertor concept (10 MW/m 2 ). Model C is based on a dual-coolant (DC) blanket (lead/lithium self-cooled bulk and He-cooled structures) and a He-cooled divertor (10 MW/m 2 ). Model D employs a self-cooled lead/lithium (SCLL) blanket and lead-lithiumcooled divertor (5 MW/m 2 ). The values in parenthesis correspond to the maximum peak heat fluxes required. It can be noted that the helium-cooled divertor is used in most of the EU plant models; it has also been proposed for the US ARIES-CS reactor study. Since 2002, it has been investigated extensively in Europe under the PPCS with the goal of reaching a maximum heat flux of at least 10 MW/m2. Work has covered many areas including conceptual design, analysis, material and fabrication issues, and experiments. Generally, the helium-cooled divertor is considered to be a suitable solution for fusion power plants, as it

Divertor design for the TITAN reversed-field-pinch reactor

International Nuclear Information System (INIS)

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

1987-01-01

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

In situ measurement of erosion/deposition in the DIII-D divertor by colorimetry

International Nuclear Information System (INIS)

Weschenfelder, F.; Wienhold, P.; Winter, J.

1996-01-01

Colorimetry was introduced into the DIII-D tokamak to measure in situ the growth and erosion of transparent wall coatings (a-C:H) on the divertor. The colorimetric measurement system consisting of a halogen light source, a set of three filters and a black/white camera is described together with a first erosion measurement. An insertable graphite sample with a diameter of 4.7 cm was precoated with a 300 nm thick amorphous carbon film and was exposed in the divertor for several discharges with its surface coplanar to the surrounding graphite tiles. For each of the discharges the plasma strike point was moved onto the sample for 1 s to erode the coating. Between the discharges a camera signal with each filter was recorded and the film thickness was evaluated along a radial line across the DIMES sample. Thus it has been possible for the first time to measure erosion and deposition of divertor material in situ and shot-by-shot. The average peak heat flux with the strike point on DIMES was about 110 W cm -2 . The measurement shows a strong decrease in the film thickness almost over the entire sample with an average erosion rate of ∼ 9 nm s -1 . (Author)

EU R and D on divertor components

International Nuclear Information System (INIS)

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

2005-01-01

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

Integrated core-edge-divertor modeling studies

International Nuclear Information System (INIS)

Stacey, W.M.

2001-01-01

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

A carbon-metal brazing for divertor plates in fusion devices

International Nuclear Information System (INIS)

Matsuda, T.; Matsumoto, T.; Miki, S.; Sogabe, T.; Okada, M.; Kubota, Y.; Sagara, A.; Noda, N.; Motojima, O.; Hino, T.; Yamashina, T.

1993-01-01

A divertor unit, which consists of carbon armors brazed to a copper cooling channel, is under development for fusion devices. Isotropic graphite (IG-430U) and CFC (CX-2002U) are used for the armor, and a copper for the cooling tube. A technique named as dissolution and deposit of base metal was employed for brazing. The reliability of the brazed components was evaluated both by 4-point bending test and thermal shock test. According to the results of a 4-point bending test under the temperature ranged from RT to 800 C in a vacuum, it was found that the strength of the brazed surface at RT was maintained up to the higher temperature, 600 C. High heat load test has been also performed on the brazed sample in order to find whether the samples meet the requirement of the divertor plates of LHD (Large Helical Device). Active Cooling Teststand (ACT:NIFS) with electron beam power of 100kW was used. In LHD, it is presumed that the maximum heat flux is 10MW/m 2 . In addition, the surface temperature of divertor has to be kept below 1,200 C to avoid RES, by active cooling. The heat load test showed that the brazing components of CX-2002U (flat plate type CFC-Cu brazed) was stable at 1,300 C under a heat flux of 10MW/m 2 , when the flow velocity of cooling water was 6m/s. No damage nor deterioration was found at the brazed zone after the heat load test

Particle control in the DIII-D advanced divertor

International Nuclear Information System (INIS)

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

1991-11-01

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

Divertor heat flux control and plasma-material interaction

International Nuclear Information System (INIS)

Kikuchi, Yusuke; Nagata, Masayoshi; Sawada, Keiji; Takamura, Shuichi; Ueda, Yoshio

2014-01-01

Development of reliable radiative-cooling divertors is essential in DEMO reactor because it uses low-activation materials with low heat removal and the plasma heat flux exhausted from the confined region is 5 times as large as in ITER. It is important to predict precisely the heat and particle flux toward the divertor plate by simulation. In this present article, theoretical and experimental data of the reflection, secondary emission and surface recombination coefficients of the divertor plate by ion bombardment are given and their effects on the power transmission coefficient are discussed. In addition, some topics such as the erosion process of the divertor plate by ELM and the plasma disruption, the thermal shielding due to the vapor layer on the divertor plate and the formation of fuzz structure on W by helium plasma irradiation, are described. (author)

Detached divertor plasmas in JET

Energy Technology Data Exchange (ETDEWEB)

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

1994-07-01

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

Divertor design through shape optimization

International Nuclear Information System (INIS)

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

2012-01-01

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

DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

Energy Technology Data Exchange (ETDEWEB)

Rudakov, D L; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W P; Wong, C C; Allen, S L; Bastasz, R J; Boedo, J A; Brooks, N H; Boivin, R L; De Temmerman, G; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Wampler, W R; Watkins, J G; Wienhold, P; Whaley, J

2007-03-15

A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited practically no carbon deposition.

Density effects on tokamak edge turbulence and transport with magnetic X-points

International Nuclear Information System (INIS)

Xu, X.Q.; Cohen, R.H.; Nevins, W.M.; Rognlien, T.D.; Ryutov, D.D.; Umansky, M.V.; Pearlstein, L.D.; Bulmer, R.H.; Russell, D.A.; Myra, J.R.; D'Ippolito, D.A.; Greenwald, M.; Snyder, P.B.; Mahdavi, M.A.

2005-01-01

Results are presented from the 3D electromagnetic turbulence code BOUT, the 2D transport code UEDGE, and theoretical analysis of boundary turbulence and transport in a real divertor-plasma geometry and its relationship to the density limit. Key results include: (1) a transition of the boundary turbulence from resistive X-point to resistive-ballooning as a critical plasma density is exceeded; (2) formation of an X-point MARFE in 2D UEDGE transport simulations for increasing outboard radial transport as found by BOUT for increasing density; (3) identification of convective transport by localized plasma 'blobs' in the SOL at high density during neutral fueling, and decorrelation of turbulence between the midplane and the divertor leg due to strong X-point magnetic shear; (4) a new divertor-leg instability driven at high plasma beta by a radial tilt of the divertor plate. (author)

Divertor characteristics and control on the W-shaped divertor with pump of JT-60U

International Nuclear Information System (INIS)

Hosogane, N.; Kubo, H.; Higashijima, S.

1999-01-01

Roles of the inner leg pumping and the private dome, which are special features of the W-shaped divertor of JT-60U, have been investigated. The following observations were made: The inner leg pumping functions well in attached states or partially detached states with weak X-point MARFE where the inner particle recycling is enhanced. A combination of main gas puff and inner leg pump is effective in reduction of intrinsic carbon impurity. Geometrical effects of the private dome on transport of hydrocarbons in the private flux region was confirmed by spectroscopic measurements of CD-band intensity profile and impurity transport simulation code using experimental data. (author)

Yang-Mills theory in null path space

International Nuclear Information System (INIS)

Kent, S.L.

1982-01-01

A reformulation of classical GL(n,c) Yang-Mills theory is presented. The reformulation is in terms of a single matrix-valued function G on a six-dimensional subspace of the space of paths in Minkowski space, M. This subspace is defined as the null paths beginning at each point, (X/sup a/), of M and ending at future null infinity. A convenient parametrization of these paths is to give the Minkowski coordinates x/sup a/ of the starting point and the (complex) stereographic coordinates (xi, antixi) on S 2 which label the light cone generators of x/sup a/. A path is thus labeled by (x/sup a/,xi, antixi). The function G(x/sup a/,xi, antixi) is defined by the parallel propagation (with a given connection) of n linearly independent fiber vectors from x/sup a/ to null infinity along the (xi, antixi) generator. From knowledge of G(x/sup a/,xi, antixi) the connection one-form γ/sub a/ at the point x/sup a/ can be obtained is shown. Furthermore how the vacuum Yang-Mills equations can be imposed on the G is shown. This results in a rather complicated integro-differential equation for G which involves the characteristic initial data (essentially the radiation field) acting as the driving term. Two simple special cases are immediately obtainable; in the case of self-dual (or anti-self dual) fields the author obtains a simple derivation of the Sparling equation, namely delta G = -GA, while for Abelian (Maxwell) theories obtained the equation delta anti delta log G = -anti delta A-anti delta A, where A and its conjugate anti A are the characteristic free data given on null infinity. The latter equation is equivalent to the vacuum Maxwell equations

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

Science.gov (United States)

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.

Finite element modelling of transport and drift effects in tokamak divertor and SOL

International Nuclear Information System (INIS)

Simard, M.; Marchand, R.; Boucher, C.; Gunn, J.P.

1996-01-01

A finite element code is used to simulate transport of a single-species plasma in the edge and divertor of a tokamak. The physical model is based on Braginskii's fluid equations for the conservation of particles, parallel momentum, ion and electron energy. In modelling recycling, transport of neutral density and energy is treated in the diffusion approximation. The electrostatic potential is obtained from the generalized Ohm's law. It is used to compute the electric field and the associated E x B drift. In a first approximation, transport is assumed to be ambipolar. The system of equations is discretized on an unstructured triangular mesh, thus permitting good spatial resolution near the X-point and an accurate description of divertor plates of arbitrary shape. Special care must be taken to prevent numerical corruption of the highly anisotropic thermal diffusion. Comparisons will be made between simulations and experimental results from TdeV. This will focus, in particular, on density and temperature profiles at the divertor plates, and on the plasma parallel velocity in the SOL. The asymmetry in the power deposited to the inner and outer divertors and the effect of magnetic field reversal will be considered. Comparisons with B2-Eirene simulation results will also be presented

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

Science.gov (United States)

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.

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

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

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

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

Divertor IR thermography on Alcator C-Moda)

Science.gov (United States)

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

2010-10-01

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

Structural analysis of the ITER Divertor toroidal rails

Energy Technology Data Exchange (ETDEWEB)

Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

2013-10-15

The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

Deposition of deuterium and metals on divertor tiles in the DIII-D tokamak

International Nuclear Information System (INIS)

Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

1991-01-01

Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the D3-D tokamak. To reduce metallic impurities in D3-D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However erosion, redeposition and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the side of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium and metals were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as fast a 1 cm from the plasma-facing and containing up to forty percent of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

Towards the procurement of the ITER divertor

International Nuclear Information System (INIS)

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

2006-01-01

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

Alternative divertor target concepts for next step fusion devices

Science.gov (United States)

Mazul, I. V.

2016-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Harrison, J. R.; Fishpool, G. M.; Thornton, A. J. [CCFE, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Walkden, N. R. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom)

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 PFR 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.

Neutral particle retention in the JET MK I divertor

International Nuclear Information System (INIS)

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

1997-01-01

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

Narrow power deposition profiles on the JET divertor target

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Science.gov (United States)

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

2018-03-01

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

Evaluation of divertor conceptual designs for a fusion power plant

International Nuclear Information System (INIS)

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

2001-01-01

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

Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

Energy Technology Data Exchange (ETDEWEB)

Laugier, F

2000-10-19

We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

Optimal thermal-hydraulic performance for helium-cooled divertors

International Nuclear Information System (INIS)

Izenson, M.G.; Martin, J.L.

1996-01-01

Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

The control of divertor carbon erosion/redeposition in the DIII-D tokamak

International Nuclear Information System (INIS)

Whyte, D.G.; West, W.P.; Wong, C.P.C.

2001-01-01

The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m 2 /burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

Divertor, thermonuclear device and method of neutralizing high temperature plasma

International Nuclear Information System (INIS)

Ikegami, Hideo.

1995-01-01

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

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2015-10-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

The ITER divertor cassette project

International Nuclear Information System (INIS)

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

2001-01-01

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

The ITER divertor cassette project

International Nuclear Information System (INIS)

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

1999-01-01

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

Thermal effects of runaway electrons in an armoured divertor

International Nuclear Information System (INIS)

Stad, R.C.L. van der.

1993-12-01

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

Review of JT-60 experimental results from June to October, 1988

International Nuclear Information System (INIS)

1989-03-01

After the achievement of AEC (Atomic Energy Commission) milestone in October, 1987, a new divertor coil has been installed to produce divertor plasmas with a single null at the buttom. A pellet injector was also installed during the same shut down period. This report presents the results from the experiment held from June to October in 1988. Major experimental issues during this phase are: 1) confinement and 2) divertor characteristics of lower X-point discharges, 3) enhancement of confinement with pellet injection, 4) ion heating with ICRF of LHRF, 5) current drive with LHRF or ICRF, and evaluation of bootstrap current. (J.P.N.)

Recent advances towards a lithium vapor box divertor

Directory of Open Access Journals (Sweden)

R.J. Goldston

2017-08-01

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

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

International Nuclear Information System (INIS)

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

1993-01-01

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

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

Energy Technology Data Exchange (ETDEWEB)

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

1994-12-31

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

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.S.; Herrmann, A.

1999-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (author)

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.-S.; Herrmann, A.

2001-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Measurements presented in this paper reveal a reduction of the maximum heat flux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

Low energy neutral particle fluxes in the JET divertor

International Nuclear Information System (INIS)

Reichle, R.; Horton, L.D.; Ingesson, L.C.; Jaeckel, H.J.; McCormick, G.K.; Loarte, A.; Simonini, R.; Stamp, M.F.

1997-01-01

First measurements are presented of the total power loss through neutral particles and their average energy in the JET divertor. The method used distinguishes between the heat flux and the electromagnetic radiation on bolometers. This is done by comparing measurements from inside the divertor either with opposite lines of sight or with a tomographic reconstruction of the radiation. The typical value of the total power loss in the divertor through neutrals is about 1 MW. The average energy of the neutral particles at the inner divertor leg is 1.5-3 eV when detachment is in progress, which agrees with EDGE2D/NIMBUS modelling. (orig.)

Influence of stray light for divertor spectroscopy in ITER

International Nuclear Information System (INIS)

Kajita, Shin; Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael; Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi

2015-01-01

The influence of stray light in the divertor spectroscopy system in ITER is quantitatively investigated using a ray tracing simulation. Simulation results show that the stray light is negligible at positions in the divertor where the plasma emission is strong. However, it is also shown that the stray light can be significantly greater than the real signal if the plasma intensity is low. Deuterium and beryllium emissions are used for the assessment; for beryllium cases in particular, since the emission profile may be non-uniform in the divertor region, the influence of stray light can be non-negligible at some positions, e.g., above the divertor dome

Continuous development of current sheets near and away from magnetic nulls

International Nuclear Information System (INIS)

Kumar, Sanjay; Bhattacharyya, R.

2016-01-01

The presented computations compare the strength of current sheets which develop near and away from the magnetic nulls. To ensure the spontaneous generation of current sheets, the computations are performed congruently with Parker's magnetostatic theorem. The simulations evince current sheets near two dimensional and three dimensional magnetic nulls as well as away from them. An important finding of this work is in the demonstration of comparative scaling of peak current density with numerical resolution, for these different types of current sheets. The results document current sheets near two dimensional magnetic nulls to have larger strength while exhibiting a stronger scaling than the current sheets close to three dimensional magnetic nulls or away from any magnetic null. The comparative scaling points to a scenario where the magnetic topology near a developing current sheet is important for energetics of the subsequent reconnection.

Deposition of deuterium and metals on divertor tiles in the DIII--D tokamak

International Nuclear Information System (INIS)

Walsh, D.S.; Doyle, B.L.; Jackson, G.L.

1992-01-01

Hydrogen recycling and impurity influx are important issues in obtaining high confinement discharges in the DIII--D tokamak. To reduce metallic impurities in DIII--D, 40% of the wall area, including the highest heat flux zones, have been covered with graphite tiles. However, erosion, redeposition, and hydrogen retention in the tiles, as well as metal transport from the remaining Inconel walls, can lead to enhanced recycling and impurity influx. Hydrogen and metal retention in divertor floor tiles have been measured using external ion beam analysis techniques following four campaigns where tiles were exposed to several thousand tokamak discharges. The areal density of deuterium retained following exposure to tokamak plasmas was measured with external nuclear reaction analysis. External proton-induced x-ray emission analysis was used to measure the areal densities of metallic impurities deposited upon the divertor tiles either by sputtering of metallic components during discharges or as contamination during tile fabrication. Measurements for both deuterium and metallic impurities were taken on both the tile surfaces which face the operating plasma and the surfaces on the sides of the tiles which form the small gaps separating each of the tiles in the divertor. The highest areal densities of both deuterium (from 2 to 8 x 10 18 atoms/cm 2 ) and metals (from 0.2 to 1 x 10 18 atoms/cm 2 ) were found on the plasma-facing surface near the inner strike point region of each set of divertor tiles. Significant deposits, extending as far as 1 cm from the plasma-facing surface and containing up to 40% of the total divertor deposition, were also observed on the gap-forming surfaces of the tiles

Strike Point Control on EAST Using an Isoflux Control Method

International Nuclear Information System (INIS)

Xing Zhe; Xiao Bingjia; Luo Zhengping; Walker, M. L.; Humphreys, D. A.

2015-01-01

For the advanced tokamak, the particle deposition and thermal load on the divertor is a big challenge. By moving the strike points on divertor target plates, the position of particle deposition and thermal load can be shifted. We could adjust the Poloidal Field (PF) coil current to achieve the strike point position feedback control. Using isoflux control method, the strike point position can be controlled by controlling the X point position. On the basis of experimental data, we establish relational expressions between X point position and strike point position. Benchmark experiments are carried out to validate the correctness and robustness of the control methods. The strike point position is successfully controlled following our command in the EAST operation. (paper)

Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

1985-01-01

Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented

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

International Nuclear Information System (INIS)

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

1994-05-01

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

Analysis of sweeping heat loads on divertor plate materials

International Nuclear Information System (INIS)

Hassanein, A.

1991-01-01

The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m 2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

FLP: a field line plotting code for bundle divertor design

International Nuclear Information System (INIS)

Ruchti, C.

1981-01-01

A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

Divertor pumping system with NBI cryopump for JT-60

International Nuclear Information System (INIS)

Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

1998-08-01

The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

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

Science.gov (United States)

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

2008-08-01

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

Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

Science.gov (United States)

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

2015-11-01

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

CIT divertor conceptual design

International Nuclear Information System (INIS)

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

1988-06-01

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

Plasma diagnostics for the DIII-D divertor upgrade (abstract)

International Nuclear Information System (INIS)

Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

1990-01-01

The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

Optimization of a bundle divertor for FED

International Nuclear Information System (INIS)

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

1982-01-01

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

The WEST project mechanical analysis of the divertor structure according to the nuclear construction code

Energy Technology Data Exchange (ETDEWEB)

Larroque, S., E-mail: sebastien.larroque@cea.fr [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France); Portafaix, C. [ITER Organization, 13108 Saint-Paul-lez-Durance (France); Saille, A.; Doceul, L.; Bucalossi, J.; Samaille, F.; Freslon, S. de [CEA Cadarache, IRFM, F-13108 Saint-Paul-lez-Durance (France)

2014-10-15

Highlights: • Divertor structure is mainly loaded by electromagnetical forces. • A simplified FEM analysis give the stresses in the structure. • RCCM criteria are required for the sizing. • Refined finite element models are used for local overstresses. - Abstract: The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST project, launched in support to the ITER tungsten divertor strategy. The installation of coils inside the vacuum vessel led to the design of a divertor supporting platform able to meet the project requirements and the associated electromagnetic loads. This paper illustrates the design, the method and the results of the thermomechanical elastic stress analyses performed in 2012. The validation of the integrity of the structure is based on the compliance with RCCMR design criteria (even though these Design and Construction rules for Mechanical Components of nuclear installations are not required for such experimental fusion device). Several 3D analyses are performed with the ANSYS code. The major one is a global analysis of half structure which determinates the stresses in the main part of the components. It gives an idea of the areas which needs local analyses. It also provides the interface loads for junction studies or simplified local model.

Standard and Null Weak Values

OpenAIRE

Zilberberg, Oded; Romito, Alessandro; Gefen, Yuval

2013-01-01

Weak value (WV) is a quantum mechanical measurement protocol, proposed by Aharonov, Albert, and Vaidman. It consists of a weak measurement, which is weighed in, conditional on the outcome of a later, strong measurement. Here we define another two-step measurement protocol, null weak value (NVW), and point out its advantages as compared to WV. We present two alternative derivations of NWVs and compare them to the corresponding derivations of WVs.

Development of liquid lithium divertor for fusion reactor

International Nuclear Information System (INIS)

Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

2000-01-01

Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

International Nuclear Information System (INIS)

Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

2015-01-01

Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

Modelling of neutral particle transport in divertor plasma

International Nuclear Information System (INIS)

Kakizuka, Tomonori; Shimizu, Katsuhiro

1995-01-01

An outline of the modelling of neutral particle transport in the diverter plasma was described in the paper. The characteristic properties of divertor plasma were largely affected by interaction between neutral particles and divertor plasma. Accordingly, the behavior of neutral particle should be investigated quantitatively. Moreover, plasma and neutral gas should be traced consistently in the plasma simulation. There are Monte Carlo modelling and the neutral gas fluid modelling as the transport modelling. The former need long calculation time, but it is able to make the physical process modelling. A ultra-large parallel computer is good for the former. In spite of proposing some kinds of models, the latter has not been established. At the view point of reducing calculation time, a work station is good for the simulation of the latter, although some physical problems have not been solved. On the Monte Carlo method particle modelling, reducing the calculation time and introducing the interaction of particles are important subjects to develop 'the evolutional Monte Carlo Method'. To reduce the calculation time, two new methods: 'Implicit Monte Carlo method' and 'Free-and Diffusive-Motion Hybrid Monte-Carlo method' have been developing. (S.Y.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-07-01

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

Divertor design and its integration into the ITER-FEAT machine

International Nuclear Information System (INIS)

Janeschitz, G.; Antipenkov, A.; Federici, G.; Ibbott, C.; Kukushkin, A.; Ladd, P.; Martin, E.; Tivey, R.

2001-01-01

The physics of the edge and divertor plasma is strongly coupled with the divertor and the fuel cycle design. Due to the limited space available the design as well as the remote maintenance approach for the ITER divertor are highly optimized to allow maximum space for the divertor plasma. Several auxiliary systems (e.g. in vessel viewing, glow discharge electrodes...) as well as a part of the pumping and fuelling system have to be integrated together with the divertor into the lower level of the ITER machine. Two main options exist for the choice of the plasma-facing material in the divertor, i.e. W and CFC. Based on already existing R and D results one can be optimistic that the material choice will be mainly based on physics considerations and material issues (e.g. C-T co-deposition). The requirements for the ITER fuel cycle arise from plasma physics as well as from the envisaged operation scenarios. Due to the complex dynamic relationship of the fuel cycle subsystems among themselves and with the plasma, codes are employed for their optimization. This paper elaborates these interacting issues and gives the latest design status. (author)

ERATO-code analysis of vacuum magnetic field oscillations in JT-60 divertor configuration

International Nuclear Information System (INIS)

Ozeki, Takahisa; Tokuda, Shinji; Tsunematsu, Toshihide; Ishida, Shinichi; Neyatani, Yuzuru; Itami, Kiyoshi; Azumi, Masafumi

1989-07-01

Magnetic field oscillations caused by external kink instabilities are numerically studied by using the ideal MHD stability code ERATO-J. Dependence of a spatial distribution of their amplitude and phase on aspect-ratio, beta-poloidal, shaping of conducting shell and divertor/limiter configurations is examined in detail. In the low aspect ratio plasma, the amplitude of magnetic oscillations in the inner side of the torus is larger than that in the outer. On the contrary, as the poloidal beta increases, the amplitude in the outer side of the torus becomes larger than that in the inner. In the divertor configuration, the amplitude of oscillations reduces near the X-point and the phase is locally modulated. The coherent magnetic oscillations observed in JT-60 agree well with the theoretical results, where the vacuum vessel is assumed to be an ideal conducting shell. The non-uniformity of the poloidal distribution observed in JT-60 can be explained by the combined effects of the finite beta, the X-point and the shape of shell. (author)

Induced tungsten melting events in the divertor of ASDEX Upgrade and their influence on plasma performance

International Nuclear Information System (INIS)

Krieger, K.; Lunt, T.; Dux, R.; Janzer, A.; Kallenbach, A.; Mueller, H.W.; Neu, R.; Puetterich, T.; Rohde, V.

2011-01-01

Tungsten rods of 1 x 1 x 3 mm were exposed at the outer divertor plate of ASDEX Upgrade using a manipulator system. Controlled melting of the W-rod in H-mode discharges was induced by moving the outer strike point towards the W-rod position. Visible light emission of ejected W droplets was recorded by fast camera systems. The resulting increase of tungsten concentration in the confined plasma was compared to that induced by W laser ablation into the outer main chamber boundary plasma. The resulting divertor retention expressed as ratio of tungsten core penetration probability from a divertor source to that of a main chamber source is ∼100. Ejected droplets are found to move always in general direction of the plasma flow. The measured magnitude of droplet acceleration shows that droplets are mainly subject to rocket forces and friction forces. Typical initial droplet size can be inferred from the time evolution of the droplet light emission to be ≥100μm.

Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

International Nuclear Information System (INIS)

R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

1999-01-01

The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

Modeling of impurity spectroscopy in the divertor and SOL of DIII-D using the 1D multifluid model NEWT1D

International Nuclear Information System (INIS)

West, W.P.; Evans, T.E.; Brooks, N.H.

1996-10-01

NEWT1D, a one dimensional multifluid model of the scrape-off layer and divertor plasma, has been used to model the plasma including the distribution of carbon ionization states in the SOL and divertor of ELMing H-mode at two injected power levels in DIII-D. Comparison of the code predictions to the measured divertor and scrape-off layer (SOL) plasma density and temperature shows good agreement. Comparison of the predicted line emissions to the spectroscopic data suggests that physically sputtered carbon from the strike point is not transported up the flux tube; a distributed source of carbon a few centimeters up the flux tube is required to achieve reasonable agreement

First results from the dynamic ergodic divertor at TEXTOR

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Science.gov (United States)

Samuell, C. M.; Porter, G. D.; Meyer, W. H.; Rognlien, T. D.; Allen, S. L.; Briesemeister, A.; Mclean, A. G.; Zeng, L.; Jaervinen, A. E.; Howard, J.

2018-05-01

Two-dimensional imaging of parallel ion velocities is compared to fluid modeling simulations to understand the role of ions in determining divertor conditions and benchmark the UEDGE fluid modeling code. Pure helium discharges are used so that spectroscopic He+ measurements represent the main-ion population at small electron temperatures. Electron temperatures and densities in the divertor match simulated values to within about 20%-30%, establishing the experiment/model match as being at least as good as those normally obtained in the more regularly simulated deuterium plasmas. He+ brightness (HeII) comparison indicates that the degree of detachment is captured well by UEDGE, principally due to the inclusion of E ×B drifts. Tomographically inverted Coherence Imaging Spectroscopy measurements are used to determine the He+ parallel velocities which display excellent agreement between the model and the experiment near the divertor target where He+ is predicted to be the main-ion species and where electron-dominated physics dictates the parallel momentum balance. Upstream near the X-point where He+ is a minority species and ion-dominated physics plays a more important role, there is an underestimation of the flow velocity magnitude by a factor of 2-3. These results indicate that more effort is required to be able to correctly predict ion momentum in these challenging regimes.

New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

International Nuclear Information System (INIS)

Damiani, C.; Baldi, L.; Galbiati, L.; Irving, M.; Lorenzelli, L.; Micciche, G.; Muro, L.; Nucci, S.; Varocchi, G.; Poggianti, A.; Fermani, G.; Maisonnier, D.; Palmer, J.; Martin, E.; Friconneau, J.P.; Gravez, P.; Takeda, N.

2001-01-01

The divertor assembly for the ITER fusion reactor consists of a number of rail mounted cassettes (54 now in ITER FEAT) located in the bottom region of the vacuum vessel. These cassettes shall be removed/installed remotely during the life of the reactor by means of specific devices. To demonstrate and optimise the feasibility of the in-vessel maintenance process the Divertor Test Platform (DTP) has been established at the ENEA Research Centre in Brasimone, Italy, as a major part of the large ITER R and D project L7. A first set of tests has been already carried out and reported during 1998, when the basic feasibility of the divertor replacement was demonstrated. In the present period (January 1999-July 2000), new activities, including both site tests and other 'external' R and D works, have been carried out in order to refine and improve the ITER divertor maintenance scenario. These include the study of abnormal maintenance operations and of possible handling equipment failure and its consequences; the procurement and testing of new sub-systems (e.g. a force reflection manipulator arm), and the development of remote handling techniques including a virtual reality system. Following a short description of the DTP, this paper reports on the new results and achievements, draws the relevant conclusions, and finally discusses future activities

Feasibility study of fast swept divertor strike point suppressing transient heat fluxes in big tokamaks.

Czech Academy of Sciences Publication Activity Database

Horáček, Jan; Cunningham, G.; Entler, Slavomír; Dobias, P.; Duban, R.; Imríšek, Martin; Markovič, Tomáš; Havlíček, Josef; Enikeev, R.

2017-01-01

Roč. 123, November (2017), s. 646-649 ISSN 0920-3796. [SOFT 2016: Symposium on Fusion Technology /29./. Prague, 05.09.2016-09.09.2016] R&D Projects: GA ČR(CZ) GA16-14228S; GA MŠk(CZ) LM2015045; GA MŠk(CZ) 8D15001; GA MŠk LG14002 Institutional support: RVO:61389021 Keywords : DEMO * ELM * Divertor * Heat flux * Tokamak Subject RIV: JF - Nuclear Energetics OBOR OECD: Nuclear related engineering Impact factor: 1.319, year: 2016 http://www.sciencedirect.com/science/article/pii/S0920379617300376

The quantum null energy condition in curved space

Science.gov (United States)

Fu, Zicao; Koeller, Jason; Marolf, Donald

2017-11-01

The quantum null energy condition (QNEC) is a conjectured bound on components (Tkk = Tab ka k^b) of the stress tensor along a null vector k a at a point p in terms of a second k-derivative of the von Neumann entropy S on one side of a null congruence N through p generated by k a . The conjecture has been established for super-renormalizeable field theories at points p that lie on a bifurcate Killing horizon with null tangent k a and for large-N holographic theories on flat space. While the Koeller-Leichenauer holographic argument clearly yields an inequality for general ( p, k^a) , more conditions are generally required for this inequality to be a useful QNEC. For d≤slant 3 , for arbitrary backgroud metric we show that the QNEC is naturally finite and independent of renormalization scheme when the expansion θ of N at the point p vanishes. This is consistent with the original QNEC conjecture which required θ and the shear σab to satisfy θ \\vert _p= \\dotθ\\vert p =0 , σab\\vert _p=0 . But for d=4, 5 more conditions than even these are required. In particular, we also require the vanishing of additional derivatives and a dominant energy condition. In the above cases the holographic argument does indeed yield a finite QNEC, though for d≥slant6 we argue these properties to fail even for weakly isolated horizons (where all derivatives of θ, σab vanish) that also satisfy a dominant energy condition. On the positive side, a corrollary to our work is that, when coupled to Einstein-Hilbert gravity, d ≤slant 3 holographic theories at large N satisfy the generalized second law (GSL) of thermodynamics at leading order in Newton’s constant G. This is the first GSL proof which does not require the quantum fields to be perturbations to a Killing horizon.

Overall feature of EAST operation space by using simple Core-SOL-Divertor model

International Nuclear Information System (INIS)

Hiwatari, R.; Hatayama, A.; Zhu, S.; Takizuka, T.; Tomita, Y.

2005-01-01

We have developed a simple Core-SOL-Divertor (C-S-D) model to investigate qualitatively the overall features of the operational space for the integrated core and edge plasma. To construct the simple C-S-D model, a simple core plasma model of ITER physics guidelines and a two-point SOL-divertor model are used. The simple C-S-D model is applied to the study of the EAST operational space with lower hybrid current drive experiments under various kinds of trade-off for the basic plasma parameters. Effective methods for extending the operation space are also presented. As shown by this study for the EAST operation space, it is evident that the C-S-D model is a useful tool to understand qualitatively the overall features of the plasma operation space. (author)

Divertor plate for thermonuclear reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Constrained ripple optimization of Tokamak bundle divertors

International Nuclear Information System (INIS)

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

1983-02-01

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

Kinetic modeling of divertor heat load fluxes in the Alcator C-Mod and DIII-D tokamaks

Energy Technology Data Exchange (ETDEWEB)

Pankin, A. Y. [Tech-X Corporation, Boulder, Colorado 80303 (United States); Rafiq, T.; Kritz, A. H. [Department of Physics, Lehigh University, Bethlehem, Pennsylvania 18015 (United States); Park, G. Y. [National Fusion Research Institute, Daejeon, 305-333 (Korea, Republic of); Chang, C. S.; Ku, S. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08540 (United States); Brunner, D.; Hughes, J. W.; LaBombard, B.; Terry, J. L. [MIT Plasma Science and Fusion Center, Cambridge, Massachusetts 02139 (United States); Groebner, R. J. [General Atomics, San Diego, California 92121 (United States)

2015-09-15

The guiding-center kinetic neoclassical transport code, XGC0 [Chang et al., Phys. Plasmas 11, 2649 (2004)], is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions, and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that the width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current I{sub p.} The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the 1/I{sub p} scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the 1/I{sub p} scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical 1/I{sub p} scaling. The Bohm or gyro-Bohm scalings of anomalous transport do not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER.

Upgraded divertor Thomson scattering system on DIII-D

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

A computational study of operating regimes for poloidal divertors

International Nuclear Information System (INIS)

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

1982-01-01

We have identified three theoretical operating regimes for poloidal divertors. These regimes are determined by the geometry of the divertor and the input energy and particle fluxes, and are characterized by the divertor plasma density and temperature. A fully self-consistent two-dimensional model for the plasma and neutral atom and molecule transport was used to study poloidal divertor operation. Extensions of our previous calculations important to this study were the inclusion of parallel electron and ion thermal conduction. We find that the key physics in divertor operation is the neutral recycling near the neutralizer plate. This can be parametrized by R = GAMMAsub(P)/GAMMAsub(O), the ratio of particle flux striking the neutralizer plate to the particle flux entering the divertor. Values of R approx. equal to 1 can be produced by large pumping rates near the neutralizer plates resulting in low neutral recycling and a high temperature, low density divertor plasma. By decreasing the pumping near the neutralizer plate, R can be raised to an intermediate value of 5-10, the plasma temperature lowered by the same factor, and the density raised by a factor of 10-30. In this regime, escape of the neutrals back to the main plasma is virtually blocked. By further restricting the pumping, R can be raised to twenty or more, thereby lowering the temperature by a factor of twenty or more and raising the density by a factor of ninety or more. Such high density regimes have been observed on D-III and appear to offer the most promise for impurity control and particle control on large reactor experiments such as INTOR or FED. In this paper, we explore the range 3 < R < 16. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

Lore, J.D., E-mail: lorejd@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Reinke, M.L. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); LaBombard, B. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Lipschultz, B. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Churchill, R.M. [Plasma Science and Fusion Center, MIT, Cambridge, MA 02139 (United States); Pitts, R.A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Feng, Y. [Max Planck Institute for Plasma Physics, Greifswald (Germany)

2015-08-15

Experiments on Alcator C-Mod with toroidally and poloidally localized divertor nitrogen injection have been modeled using the three-dimensional edge transport code EMC3-EIRENE to elucidate the mechanisms driving measured toroidal asymmetries. In these experiments five toroidally distributed gas injectors in the private flux region were sequentially activated in separate discharges resulting in clear evidence of toroidal asymmetries in radiated power and nitrogen line emission as well as a ∼50% toroidal modulation in electron pressure at the divertor target. The pressure modulation is qualitatively reproduced by the modeling, with the simulation yielding a toroidal asymmetry in the heat flow to the outer strike point. Toroidal variation in impurity line emission is qualitatively matched in the scrape-off layer above the strike point, however kinetic corrections and cross-field drifts are likely required to quantitatively reproduce impurity behavior in the private flux region and electron temperatures and densities directly in front of the target.

Disruption characteristics in PDX with limiter and divertor discharges

International Nuclear Information System (INIS)

Couture, P.; McGuire, K.

1986-09-01

A comparison has been made between the characteristics of disruptions with limiter and divertor configurations in PDX. A large data base on disruptions has been collected over four years of machine operation, and a total of 15,000 discharges are contained in the data file. It was found that divertor discharges have less disruptions during ramp up and flattop of the plasma current. However, for divertor discharges a large number of fast, low current disruptions take place during the current ramp down. These disruptions are probably caused by the deformation of the plasma shape

Installation and initial operation of the DIII-D advanced divertor cryocondensation pump

International Nuclear Information System (INIS)

Smith, J.P.; Schaubel, K.M.; Baxi, C.B.; Campbell, G.L.; Hyatt, A.W.; Laughon, G.J.; Mahdavi, M.A.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.; Menon, M.M.

1993-10-01

Phase two of a divertor cryocondensation pump, the Advanced Divertor Program, is now installed in the DIII-D tokamak at General Atomics and complements the phase one biasable ring electrode. The installation consists of a 10 m long cryocondensation pump located in the divertor baffle chamber to study plasma density control by pumping of the divertor. The design is a toroidally electrically continuous liquid helium-cooled panel with 1 m 2 of pumping surface. The helium panel is single point grounded to the nitrogen shield to minimize eddy currents. The nitrogen shield is toroidally continuous and grounded to the vacuum vessel in 24 locations to prevent voltage potentials from building up between the pump and vacuum vessel wall. A radiation/particle shield surrounds the nitrogen-cooled surface to minimize the heat load and prevent water molecules condensed on the nitrogen surface from being released by impact of energetic particles. Large currents (>5000 A) are driven in the helium and nitrogen panels during ohmic coil ramp up and during disruptions. The pump is designed to accommodate both the thermal and mechanical loads due to these currents. A feedthrough for the cryogens allows for both radial and vertical motion of the pump with respect to the vacuum vessel. Thermal performance measured on a prototype verified the analytical model and thermal design of the pump. Characterization tests of the installed pump show the pumping speed in deuterium is 42,000 ell/sec for a pressure of 5 mTorr. Induction heating of the pump (at 300 W) resulted in no degradation of pumping speed. Plasma operations with the cryopump show a 60% lower density in H-mode

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

International Nuclear Information System (INIS)

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

2011-01-01

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.

Magnetic field models and their application in optimal magnetic divertor design

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Study of particle pumping characteristics for different pumping geometries in JT-60U and DIII-D divertors

International Nuclear Information System (INIS)

Takenaga, H.; Sakasai, A.; Kubo, H.

2001-01-01

Particle pumping characteristics were compared between pumping from the inner side private flux region (IPP) and pumping from both sides of the private flux region (BPP) in the JT-60U W shaped divertor, and between JT-60U IPP and pumping in the DIII-D lower baffled divertor. The pumping flux for BPP is smaller than that for IPP by about a factor of 2 with weak in-out asymmetry of recycling neutral flux and by a factor of 3.5-6.5 with strong in-out asymmetry. The reduction of the pumping flux for BPP is consistent with Monte Carlo simulations, where backflow at the outer pumping slot is observed due to in-out recycling asymmetry. The pumping flux in DIII-D at I p =0.8 MA and B T =1.6 T is comparable to or smaller than that for JT-60U IPP at I p =1.0 MA, B T =3.8 T and I p =1.5 MA, B T =3.5 T in the same density regime. In the DIII-D divertor with pumping from the private flux region, the pumping flux decreases with increasing in-out asymmetry. The pumping flux normalized by the integrated D α emission over the whole plasma exhibits a similar dependence on the distance between the pumping slot and the strike point in JT-60U IPP and the DIII-D lower divertor with pumping through the outer divertor plasma region. (author)

Preliminary concept design of the divertor remote handling system for DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2014-11-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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.

Compact poloidal divertor reference design for TNS

International Nuclear Information System (INIS)

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

1977-01-01

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

Estimation of Neutral Density in Edge Plasma with Double Null Configuration in EAST

International Nuclear Information System (INIS)

Zhang Ling; Xu Guosheng; Ding Siye; Gao Wei; Wu Zhenwei; Chen Yingjie; Huang Juan; Liu Xiaoju; Zang Qing; Chang Jiafeng; Zhang Wei; Li Yingying; Qian Jinping

2011-01-01

In this work, population coefficients of hydrogen's n = 3 excited state from the hydrogen collisional-radiative (CR) model, from the data file of DEGAS 2, are used to calculate the photon emissivity coefficients (PECs) of hydrogen Balmer-α (n = 3 → n = 2) (H α ). The results are compared with the PECs from Atomic Data and Analysis Structure (ADAS) database, and a good agreement is found. A magnetic surface-averaged neutral density profile of typical double-null (DN) plasma in EAST is obtained by using FRANTIC, the 1.5-D fluid transport code. It is found that the sum of integral D α and H α emission intensity calculated via the neutral density agrees with the measured results obtained by using the absolutely calibrated multi-channel poloidal photodiode array systems viewing the lower divertor at the last closed flux surface (LCFS). It is revealed that the typical magnetic surface-averaged neutral density at LCFS is about 3.5 x 10 16 m -3 . (magnetically confined plasma)

Electron beam irradiation experiments of monoblock divertor mock-up

International Nuclear Information System (INIS)

Satoh, Kazuyoshi; Akiba, Masato; Araki, Masanori; Suzuki, Satoshi; Yokoyama, Kenji; Smid, I.; Cardella, A.; Duwe, R.; Di Pietro, E.

1993-03-01

It is one of the key issues for ITER to develop the divertor plate. Electron beam irradiation tests were carried out on a NET divertor mock-up using JEBIS at JAERI under a collaboration between The NET team, JAERI and KFA Juelich. Screening tests (maximum heat flux of 23 MW/m 2 ) and thermal cycling tests (18 MW/m 2 , 30s, 1000cycle) were carried out. As a result of the screening tests, the erosion caused by sublimation of C/C was observed on the surface of armor tile. No serious damage such as cracks or detachments, however, were found. As a result of the thermal cycling tests, no major damage was detected on the C/C surface. However cooling time constant of the divertor mock-up increased over 600cycle. Therefore it implies that some defects would occur at the brazing interface of the divertor mock-up. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Divertor impurity injection using high voltage arcs for impurity transport studies on the Mega Amp Spherical Tokamak

International Nuclear Information System (INIS)

Leggate, H. J.; Turner, M. M.; Lisgo, S. W.; Harrison, J. R.; Elmore, S.; Allan, S. Y.; Gaffka, R. C.; Stephen, R. C.

2014-01-01

The operation of next-generation fusion reactors will be significantly affected by impurity transport in the scrape-off layer (SOL). Current modelling efforts are restricted by a lack of detailed data on impurity transport in the SOL. In order to address this, a carbon injector has been designed and installed on the Mega Amp Spherical Tokamak (MAST). The injector creates short lived carbon plumes originating at the MAST divertor lasting less than 50 μs. High voltage capacitor banks are used to create a discharge across concentric carbon electrodes located in a probe mounted on the Divertor Science Facility in the MAST lower divertor. This results in a very short plume duration allowing observation of the evolution of the plume and precise localisation of the plume relative to the X-point on MAST. The emission from the carbon plume was imaged using fast visible cameras filtered in order to isolate the carbon II and carbon III emission lines centered around 514 nm and 465 nm

Two-dimensional impurity transport calculations for a high recycling divertor

International Nuclear Information System (INIS)

Brooks, J.N.

1986-04-01

Two dimensional analysis of impurity transport in a high recycling divertor shows asymmetric particle fluxes to the divertor plate, low helium pumping efficiency, and high scrapeoff zone shielding for sputtered impurities

Divertor development for a future fusion power plant

International Nuclear Information System (INIS)

Norajitra, Prachai

2011-01-01

Nuclear fusion is considered as a future source of sustainable energy supply. In the first chapter, the physical principle of magnetic plasma confinement, and the function of a tokamak are described. Since the discovery of the H-mode in ASDEX experiment ''Divertor I'' in 1982, the divertor has been an integral part of all modern tokamaks and stellarators, not least the ITER machine. The goal of this work is to develop a feasible divertor design for a fusion power plant to be built after ITER. This task is particularly challenging because a fusion power plant formulates much greater demands on the structural material and the design than ITER in terms of neutron wall load and radiation. First several divertor concepts proposed in the literature e.g. the Power Plant Conceptual Study (PPCS) using different coolants are reviewed and analyzed with respect to their performance. As a result helium cooled divertor concept exhibited the best potential to come up to the highest safety requirements and therefore has been chosen for the design process. From the third chapter the necessary steps towards this goal are described. First, the boundary conditions for the arrangement of a divertor with respect to the fusion plasma are discussed, as this determines the main thermal and neutronic load parameters. Based on the loads material selection criteria are inherently formulated. In the next step, the reference design is defined in accordance with the established functional design specifications. The developed concept is of modular nature and consists of cooling fingers of tungsten using an impingement cooling in order to achieve a heat dissipation of 10 MW/m 2 . In the next step, the design was subjected to the thermal-hydraulic and thermo-mechanical calculations in order to analyze and improve the performance and the manufacturing technologies. Based on these results, a prototype was produced and experimentally tested on their cooling capacity, their thermo-cyclic loading

Probabilistic analysis of divertor plate lifetime in tokamak reactors

International Nuclear Information System (INIS)

Golinescu, R.P.; Kazimi, M.S.

1994-01-01

Defining a methodology for a reliability estimate of the International Tokamak Experimental Reactor (ITER) divertor is the objective of the study summarized in this paper. If ITER could be designed such that no transients of any type occurred, the divertor reliability would be controlled by erosion of material during normal operation. The occurrence of several transient events results in important contribution to the expected divertor failure rate. Some transients cause the temperature in the divertor plate (DP) to rise; if these temperatures get too high, the structural elements in the DP will weaken and subsequently suffer structural failure and possibly reach the melting temperature. Using the limited data available leads to the result that there is a high probability that the DP will reliably withstand a peak heat flux of 11 MW/m 2 . However, transient events will lead to a much shorter lifetime than desirable for DP's, mainly due to the expected severe effects of plasma disruptions. If transients occurred, but the shutdown mechanism succeeded to perform without inducing a disruption, divertor reliability could be significantly improved. Improved characterization of the disruption conditions, and enlarged scope of failure modes should be pursued to gain confidence in the present conclusions

Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

International Nuclear Information System (INIS)

Di Gironimo, G.; Carfora, D.; Esposito, G.; Lanzotti, A.; Marzullo, D.; Siuko, M.

2015-01-01

Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed

Concept design of the DEMO divertor cassette-to-vacuum vessel locking system adopting a systems engineering approach

Energy Technology Data Exchange (ETDEWEB)

Di Gironimo, G., E-mail: giuseppe.digironimo@unina.it [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Carfora, D. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland); Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Esposito, G.; Lanzotti, A.; Marzullo, D. [Università degli Studi di Napoli “Federico II”, Dipartimento di Ingegneria Industriale, Piazzale Tecchio 80, 80135 Napoli (Italy); Siuko, M. [VTT Technical Research Centre of Finland, Tekniikankatu 1, PO Box 1300, FI-33101 Tampere (Finland)

2015-05-15

Highlights: • An iterative and incremental design process for cassette-to-VV locking system of DEMO divertor is presented. • Three different concepts have been developed with a systematic design approach. • The final concept has been selected with Fuzzy-Analytic Hierarchy Process in virtual reality. - Abstract: This paper deals with pre-concept studies of DEMO divertor cassette-to-vacuum vessel locking system under the work program WP13-DAS-07-T06: Divertor Remote Maintenance System pre-concept study. An iterative design process, consistent with Systems Engineering guidelines and named Iterative and Participative Axiomatic Design Process (IPADeP), is used in this paper to propose new innovative solutions for divertor locking system, which can overcome the difficulties in applying the ITER principles to DEMO. The solutions conceived have been analysed from the structural point of view using the software Ansys and, eventually, evaluated using the methodology known as Fuzzy-Analytic Hierarchy Process. Due to the lack and the uncertainty of the requirements in this early conceptual design stage, the aim is to cover a first iteration of an iterative and incremental process to propose an innovative design concept to be developed in more details as the information will be completed.

The control of convection by fuelling and pumping in the JET pumped divertor

Energy Technology Data Exchange (ETDEWEB)

Harbour, P J; Andrew, P; Campbell, D; Clement, S; Davies, S; Ehrenberg, J; Erents, S K; Gondhalekar, A; Gadeberg, M; Gottardi, N; Von Hellermann, M; Horton, L; Loarte, A; Lowry, C; Maggi, C; McCormick, K; O` Brien, D; Reichle, R; Saibene, G; Simonini, R; Spence, J; Stamp, M; Stork, D; Taroni, A; Vlases, G [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

1994-07-01

Convection from the scrape-off layer (SOL) to the divertor will control core impurities, if it retains them in a cold, dense, divertor plasma. This implies a high impurity concentration in the divertor, low at its entrance. Particle flux into the divertor entrance can be varied systematically in JET, using the new fuelling and pumping systems. The convection ratio has been estimated for various conditions of operation. Particle convection into the divertor should increase thermal convection, decreasing thermal conduction, and temperature and density gradients along the magnetic field, hence increasing the frictional force and decreasing the thermal force on impurities. Changes in convection in the SOL, caused by gaseous fuelling, have been studied, both experimentally in the JET Mk I divertor and with EDGE2/NIMBUS. 1 ref., 4 figs., 1 tab.

Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

International Nuclear Information System (INIS)

Peterson, R.R.; MacFarlane, J.J.; Wang, P.

1994-01-01

Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

Experimental studies on an axisymmetric divertor in DIVA(JFT-2a)

International Nuclear Information System (INIS)

Yamamoto, Shin

1979-03-01

DIVA(JFT-2a) is the first tokamak with an axisymmetric divertor in the world. Objectives of the experiments were i) Plasma production and confinement in a tokamak with a separatrix magnetic surface, and ii) divertor effects on radiation loss and plasma confinement. The results so far are as follows: i) The equilibrium with a separatrix magnetic surface is stable during the discharge. ii) There is an ergodic region near the separatrix magnetic surface due to non-axisymmetric magnetic perturbations. iii) The divertor reduces radiation loss and increases energy confinement time. iv) The divertor does not affect the transport process in the main plasma. (author)

Magnetic divertor design for the compact reversed-field pinch reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Miller, R.L.; Krakowski, R.A.

1984-01-01

A recently completed design of a pumped-limiter-based Compact Reversed-Field Pinch Reactor is used to estimate for the first time the impact of magnetic divertors. A range of divertor options for the low-toroidal-field RFP is examined, and a design selection is made constrained by consideration of field ripple (magnetic island), blanket displacement, recirculating power, cost, heat flux, and access. Design choices based on diversion of minority (toroidal) field lead to a preference for (poloidally) symmetric or bundle divertor geometries

Radiative divertor plasmas with convection in DIII-D

International Nuclear Information System (INIS)

Leornard, A.W.; Porter, G.D.; Wood, R.D.

1998-01-01

The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

Development of divertor pumping system with superpermeable membrane

International Nuclear Information System (INIS)

Nakamura, Y.; Ohyabu, N.; Suzuki, H.; Nakahara, Y.; Livshits, A.; Notkin, M.; Alimov, V.; Busnyuk, A.

2000-01-01

A new divertor pumping system with superpermeable membranes of group Va-metals (Nb, V) is now under research and development. Properties of membrane pumping were investigated with the use of a plasma device simulating divertor plasma conditions. The deposition of metal (Fe) and non-metal (C) impurities on the membrane upstream surface results in a degradation of plasma driven superpermeation at the membrane temperature T m m ≥800 deg. C. The same temperature effect on superpermeation is observed at sputtering of membrane surface by energetic plasma ions. In addition, the first application of the membrane pumping to fusion devices has been carried out and a deuterium pumping through the membrane was demonstrated under the conditions of divertor plasma in the JFT-2M tokamak

Status of National Spherical Torus Experiment Liquid Lithium Divertor

Science.gov (United States)

Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

2009-11-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Conceptual Design for a Bulk Tungsten Divertor Tile in JET

International Nuclear Information System (INIS)

Mertens, P.; Neubauer, O.; Philipps, V.; Schweer, B.; Samm, U.; Hirai, T.; Sadakov, S.

2006-01-01

With ITER on the verge of being build, the ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant to the support of decisions to the first wall construction and, from the point of view of plasma physics, to the corresponding investigations of possible plasma configuration and plasma-wall interaction. In both respects, tungsten plays a key role in the divertor cladding whereas beryllium will be used for the vessel's first wall. For the central tile, also called LB-SRP for '' Load-Bearing Septum Replacement Plate '', resort to bulk tungsten is envisaged in order to cope with the high loads expected (up to 10 MW/m 2 for about 10 s). This is indeed the preferred plasma-facing component for positioning the outer strike-point in the divertor. Forschungszentrum Juelich has developed a conceptual design for this tile, based on an assembly of tungsten blades or lamellae. It was selected in the frame of an extensive R-and-D study in search of a suitable, inertially cooled component(T. Hirai et al., R-and-D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project: this conference). As reported elsewhere, the design is actually driven by electromagnetic considerations in the first place(S. Sadakov et al., Detailed electromagnetic analysis for optimisation of a tungsten divertor plate for JET: this conference). The lamellae are grouped in four stacks per tile which are independently attached to an equally re-designed supporting structure. A so-called adapter plate, also a new design, takes care of an appropriate interface to the base carrier of JET, onto which modules of two tiles are positioned and screwed by remote handling (RH) procedures. The compatibility of the design on the whole with RH requirements is another essential ingredient which was duly taken into account throughout. The concept and the underlying philosophy will be presented along with important

Modular He-cooled divertor for power plant application

International Nuclear Information System (INIS)

Diegele, Eberhard; Kruessmann, R.; Malang, S.; Norajitra, P.; Rizzi, G.

2003-01-01

Gas cooled divertor concepts are regarded as a suitable option for fusion power plants because of an increased thermal efficiency for power conversion systems and the use of a coolant compatible with all blanket systems. A modular helium cooled divertor concept is proposed with an improved heat transfer. The concept employs small tiles made of tungsten and brazed to a finger-like structure made of Mo-alloy (TZM). Design goal was a heat flux of at least 15 MW/m 2 and a minimum temperature of the structure of 600 deg.C. The divertor has to survive a number of cycles (100-1000) between operating temperature and room temperature even for the steady state operation assumed. Thermo-hydraulic design requirements for the concepts include to keep the pumping power below 10% of the thermal power to the divertor plates, and simultaneously achieving a heat transfer coefficient in excess of 60 kW/m 2 K. Inelastic stress analysis indicates that design allowable stress limits on primary and secondary (thermal) stresses as required by the ITER structural design criteria are met even under conservative assumptions. Finally, critical issues for future development are addressed

Plasma/neutral gas transport in divertors and limiters

International Nuclear Information System (INIS)

Gierszewski, P.J.

1983-09-01

The engineering design of the divertor and first wall region of fusion reactors requires accurate knowledge of the energies and particle fluxes striking these surfaces. Simple calculations indicate that approx. 10 MW/m 2 heat fluxes and approx. 1 cm/yr erosion rates are possible, but there remain fundamental physics questions that bear directly on the engineering design. The purpose of this study was to treat hydrogen plasma and neutral gas transport in divertors and pumped limiters in sufficient detail to answer some of the questions as to the actual conditions that will be expected in fusion reactors. This was accomplished in four parts: (1) a review of relevant atomic processes to establish the dominant interactions and their data base; (2) a steady-state coupled O-D model of the plasma core, scrape-off layer and divertor exhaust to determine gross modes of operation and edge conditions; (3) a 1-D kinetic transport model to investigate the case of collisionless divertor exhaust, including non-Maxwellian ions and neutral atoms, highly collisional electrons, and a self-consistent electric field; and (4) a 3-D Monte Carlo treatment of neutral transport to correctly account for geometric effects

Results of the H-mode experiments with JT-60 outer and lower divertors

International Nuclear Information System (INIS)

Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

1989-08-01

In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

Non-ambipolar divertor flows in heliotron E

International Nuclear Information System (INIS)

Chechkin, V.V.; Voitsenya, V.S.; Smirnova, M.S.; Sorokovoj, E.L.; Mizuuchi, T.; Nagasaki, K.; Okada, H.; Funaba, H.; Hamada, T.; Sano, F.; Zushi, H.; Nakasuga, M.; Kondo, K.; Masuzaki, S.; Motojima, O.

1999-01-01

The object of the work is to find out (1) the poloidal distributions of PEC in different poloidal cross-sections of the torus within one field period; (2) the link between PEC in the divertor flows (DF) and the characteristics of the divertor field lines; (3) the effect of different methods and regimes of heating on PEC. The data having been obtained enable us to understand at least partially the nature of PEC in the diverted plasma of H-E

Investigation of the influence of divertor recycling on global plasma confinement in JET ITER-like wall

NARCIS (Netherlands)

Tamain, P.; Joffrin, E.; Bufferand, H.; Jarvinen, A.; Brezinsek, S.; Ciraolo, G.; Delabie, E.; Frassinetti, L.; Giroud, C.; Groth, M.; Lipschultz, B.; Lomas, P.; Marsen, S.; Menmuir, S.; Oberkofler, M.; Stamp, M.; Wiesen, S.; JET-EFDA Contributors,

2015-01-01

Abstract The impact of the divertor geometry on global plasma confinement in type I ELMy H-mode has been investigated in the JET tokamak equipped with ITER-Like Wall. Discharges have been performed in which the position of the strike-points was changed while keeping the bulk plasma equilibrium

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

Science.gov (United States)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

Design and analysis of the W7-X divertor scraper element

International Nuclear Information System (INIS)

Lumsdaine, A.; Tipton, J.; Lore, J.; McGinnis, D.; Canik, J.; Harris, J.; Peacock, A.; Boscary, J.; Tretter, J.; Andreeva, T.

2013-01-01

Highlights: • A high heat flux actively cooled divertor component is thermally modeled with CFD. • CFC monoblocks are analyzed to verify peak steady-state temperatures do not exceed 1200 °C. • A field line diffusion code is developed to determine the heat flux on the divertor components. • Iteration is required to develop a surface that meets the criteria and fits into the limited space. -- Abstract: Thehigh heat-flux divertor of the Wendelstein 7-X large stellarator experiment consists of 10 divertor units which are designed to carry a steady-state heat flux of 10 MW/m 2 . However, the edge elements of this divertor are limited to only 5 MW/m 2 , and may be overloaded in certain plasma scenarios. It is proposed to reduce this heat by placing an additional “scraper element” in each of the ten divertor locations. It will be constructed using carbon fiber composite (CFC) monoblock technology. The design of the monoblocks and the path of the cooling tubes must be optimized in order to survive the significant steady-state heat loads, provide adequate coverage for the existing divertor, be located within sub-millimeter accuracy, and take into account the boundaries to other in vessel components, all at a minimum cost. Computational fluid dynamics modeling has been performed to examine the thermal transfer through the monoblock swirl tube channels for the design of the monoblock orientation. An iterative physics modeling and computer aided design process is being performed to optimize the placement of the scraper element within the severe spatial restrictions

Effect of variation in equilibrium shape on ELMing H-mode performance in DIII-D diverted plasmas

International Nuclear Information System (INIS)

Fenstermacher, M.E.; Osborne, T.H.; Petrie, T.W.

2001-01-01

The changes in the performance of the core, pedestal, scrape-off-layer (SOL), and divertor plasmas as a result of changes in triangularity, δ, up/down magnetic balance, and secondary divertor volume were examined in shape variation experiments using ELMing H mode plasmas on DIII-D. In moderate density, unpumped plasmas, high δ∼0.7 increased the energy in the H mode pedestal and the global energy confinement of the core, primarily due to an increase in the margin by which the edge pressure gradient exceeded the value which would have been expected had it been limited by infinite-n ideal ballooning modes. In addition, a nearly balanced double-null (DN) shape was effective for sharing the peak heat flux in the divertor in these attached plasmas. For detached plasmas good heat flux sharing was obtained for a substantial range of unbalanced DN shapes. Finally, the presence of a second X-point in unbalanced DN shapes did not degrade the plasma performance if it was sufficiently far inside the vacuum vessel. These results indicate that a high δ unbalanced DN shape has some advantages over a single null shape for future high power tokamak operation. (author)

He-cooled divertor for DEMO. Fabrication technology for tungsten cooling fingers

Energy Technology Data Exchange (ETDEWEB)

Reiser, J.; Norajitra, P.; Widak, V.; Krauss, W. [Forschungszentrum Karlsruhe GmbH (Germany)

2008-07-01

A modular helium-cooled divertor design based on the multi-jet impingement concept (HEMJ) has been developed for the ''post-ITER'' demonstration reactor (DEMO) at the Forschungszentrum Karlsruhe [1, 2]. The main function of the divertor is to keep the plasma free from impurities by catching particles, such as fusion ash and eroded particles from the first wall. From the divertor surface, a maximum heat load of 10 MW/m{sup 2} at least has to be removed. The whole divertor is split up into a number of cassettes (48 according to the latest design studies [3]). Each cassette is cooled separately. The target plates are provided with several cooling fingers to keep the thermal stresses low. Each cooling finger consists of a tungsten tile which is brazed to a thimble-like cap made of a tungsten alloy W-1%La2O3 (WL10) underneath. The thimble has to be connected to the ODS EUROFER steel structure, which is accomplished by brazing again. The tungsten/tungsten brazing is exposed to 1200 C operation temperature while the tungsten/steel brazing joint must withstand 700 C operating temperature. Cooling of the finger is achieved by multi-jet impingement with helium. The inlet temperature of helium is 600 C and rises up to 700 C at the outlet. With this kind of cooling, a mean heat transfer coefficient of 35.000 W/(m{sup 2*}K) can be reached. This compact report will focus on the manufacturing of such a cooling finger unit at FZK. It will cover the machining of the tungsten tile as well as of the thimble and, the brazing of the parts. The major aim of this activity is, on the one hand, to obtain functioning mock-ups with high quality and high reliability, in particular in terms of minimising the surface roughness, cracks, and micro-cracks. On the other hand, effort should also be laid on realising the mass production from economic point of view. (orig.)

Local island divertor experiments on LHD

International Nuclear Information System (INIS)

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

2005-01-01

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

Conceptual design of divertor cassette handling by remote handling system for JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2007-01-01

The JT-60SA aims to contribute and supplement ITER toward DEMO reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is prohibited. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor modules. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor module, which is 10 degrees wide in toroidal direction and weighs 500kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor module to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the module by a pallet installed from outside the VV. (author)

Conceptual design of divertor cassette handling by remote handling system of JT-60SA

International Nuclear Information System (INIS)

Hayashi, Takao; Sakurai, Shinji; Masaki, Kei; Tamai, Hiroshi; Yoshida, Kiyoshi; Matsukawa, Makoto

2008-01-01

The JT-60SA aims to contribute and supplement ITER toward demonstration fusion reactor based on tokamak concept. One of the features of JT-60SA is its high power long pulse heating, causing the large annual neutron fluence. Because the expected dose rate at the vacuum vessel (VV) may exceed 1 mSv/hr after 10 years operation and three month cooling, the human access inside the VV is restricted. Therefore a remote handling (RH) system is necessary for the maintenance and repair of in-vessel components. This paper described the RH system of JT-60SA, especially the expansion of the RH rail and exchange of the divertor cassettes. The RH rail is divided into nine and three-point mounting. The nine sections can cover 225 degrees in toroidal direction. A divertor cassette, which is 10 degrees wide in toroidal direction and weighs 500 kg itself due to the limitations of port width and handling weight, can be exchanged by heavy weight manipulator (HWM). The HWM brings the divertor cassette to the front of the other RH port, which is used for supporting the rail and/or carrying in and out equipments. Then another RH device receives and brings out the cassette by a pallet installed from outside the VV. (author)

Progress in ergodic divertor operation on Tore Supra

International Nuclear Information System (INIS)

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

1999-09-01

Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q ∼ 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

Mechanical Design of the NSTX Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

2009-02-19

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Enhancing the DEMO divertor target by interlayer engineering

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Enhancing the DEMO divertor target by interlayer engineering

International Nuclear Information System (INIS)

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

2015-01-01

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

Mechanical Design of the NSTX Liquid Lithium Divertor

International Nuclear Information System (INIS)

Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

2009-01-01

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Fabrication and installation of the DIII-D radiative divertor structures

International Nuclear Information System (INIS)

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

1997-11-01

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

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

Science.gov (United States)

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.

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

International Nuclear Information System (INIS)

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

1995-10-01

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

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-09-01

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

Conceptual design for a bulk tungsten divertor tile in JET

International Nuclear Information System (INIS)

Mertens, Ph.; Hirai, T.; Linke, J.; Neubauer, O.; Pintsuk, G.; Philipps, V.; Sadakov, S.; Samm, U.; Schweer, B.

2007-01-01

The ITER-like Wall project (ILW) for JET aims at providing the plasma chamber of the tokamak with an environment of mixed materials which will be relevant for the actual first wall construction on ITER. Tungsten plays a key role in the divertor cladding. For the central tile, also called LB-SRP for 'load-bearing septum replacement plate', bulk tungsten is envisaged in order to cope with the high heat loads expected (up to 10 MW/m 2 for 10 s). The outer strike-point in the divertor will be positioned on this tile for the most relevant configurations. Forschungszentrum Juelich (FZJ) has developed a conceptual design based on an assembly of tungsten blades or lamellae. An appropriate interface with the base carrier of JET, on which modules of two tiles are positioned and fixed by remote handling procedures, is a substantial part of the integral design. Important issues are the electromagnetic forces and expected temperature distributions. Material choices combine tungsten, TZM TM , Inconel and ceramic parts. The completed design has been finalised in a proposal to the ILW project, with utmost ITER-relevance

ITER divertor, design issues and research and development

International Nuclear Information System (INIS)

Tivey, R.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Janeschitz, G.; Raffray, R.; Mazul, I.; Pacher, H.; Ulrickson, M.; Vieider, G.

1999-01-01

Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R and D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R and D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m -2 10 MW m -2 . Analysis and experiment show that a CfC armour thickness of ∝20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within ∝6 months. (orig.)

ITER divertor, design issues and research and development

Energy Technology Data Exchange (ETDEWEB)

Tivey, R.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Janeschitz, G.; Raffray, R. [ITER Joint Central Team, Garching (Germany). Joint Central Work Site; Akiba, M. [Japan Atomic Energy Research Institute, Naka-machi, Ibaraki-ken (Japan); Mazul, I. [Efremov Institute, St Petersburg (Russian Federation); Pacher, H. [NET Team, Boltzmannstr. 2, D-85748, Garching (Germany); Ulrickson, M. [Sandia National Laboratories, Albuquerque, NM (United States); Vieider, G. [NET Team, Boltzmannstr. 2, D-85748, Garching (Germany)

1999-11-01

Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R and D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R and D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m{sup -2}10 MW m{sup -2}. Analysis and experiment show that a CfC armour thickness of {proportional_to}20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within {proportional_to}6 months. (orig.)

Numerical simulations of resistive magnetohydrodynamic instabilities in a poloidal divertor tokamak

International Nuclear Information System (INIS)

Uchimoto, E.

1988-03-01

A new 3-D resistive MHD initial value code RPD has been successfully developed from scratch to study the linear and nonlinear evolution of long wavelength resistive MHD instabilities in a square cross-section tokamak with or without a poloidal divertor. The code numerically advances the full set of compressible resistive MHD equations in a toroidal geometry, with an important option of permitting the divertor separatrix and the region outside it to be in the computational domain. A severe temporal step size restriction for numerical stability imposed by the fast compressional waves was removed by developing and implementing a new, efficient semi-implicit scheme extending one first proposed by Harned and Kerner. As a result, the code typically runs faster than that with a mostly explicit scheme by a factor of about the aspect ratio. The equilibrium input for RPD is generated by a new 2-D code EQPD that is based on the Chodura-Schluter method. The RPD code, as well as the new semi-implicit scheme, has passed very extensive numerical tests in both divertor and divertorless geometries. Linear and nonlinear simulations in a divertorless geometry have reproduced the standard, previously known results. In a geometry with a four-node divertor the m = 2,n = 1 (2/1) tearing mode tends to be linearly stabilized as the q = 2 surface approaches the divertor separatrix. However, the m = 1,n = 1 (1/1) resistive kink mode remains relatively unaffected by the nearness of the q = 1 surface to the divertor separatrix. When plasma current is added to the region outside the divertor separatrix, the 2/1 tearing mode is linearly stabilized not by this current, but by the profile modifications induced near the q = 2 surface and the divertor separatrix. A similar stabilization effect is seen for the 1/1 resistive kink mode, but to a lesser extent. 77 refs., 91 figs

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-01

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

Development of database for the divertor recycling in JT-60U and its analysis

Energy Technology Data Exchange (ETDEWEB)

Takizuka, Tomonori; Shimizu, Katsuhiro; Hayashi, Nobuhiko; Asakura, Nobuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arakawa, Kazuya [Komatsu, Ltd., Tokyo (Japan)

2003-05-01

We have developed a database for the divertor recycling in JT-60U plasmas. This database makes it possible to investigate behaviors of the neutral-particle flux in plasmas and the ion flux to divertor plates under a condition for core-plasma parameters, such as electron density and heating power. The correlation between the electron density and the heating power is not strong in this database, and parameter scans for the density and the power in wide ranges are realized. On the basis of this database, we have analyzed the ion flux to divertor plates. The divertor-plate ion flux amplified by the recycling grows nonlinearly with the increase of the electron density n{sub e}. Its averaged dependence is a linear growth ({approx}n{sub e}{sup 1.0}) at the low density, and becomes a nonlinear growth ({approx}n{sub e}{sup 1.5}) at the high density. The spread of dependence from the averaged one is very large. This spread is caused mainly by complex physical characteristics of divertor plasmas, though it is little dependent on the heating power. The behavior of ion flux depends strongly on divertor configurations and divertor-plate/first-wall conditions. It is confirmed that the bifurcated transition takes place from the low-recycling divertor plasma at the low density to the high-recycling divertor plasma at the high density. The density at the transition is nearly proportional to the 1/4 power of the heating power. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-08-15

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

McLean, A. G., E-mail: mclean@fusion.gat.com; Soukhanovskii, V. A.; Allen, S. L. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States); Carlstrom, T. N. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); LeBlanc, B. P.; Ono, M.; Stratton, B. C. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States)

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 on NSTX-U, the divertor TS system would provide an invaluable diagnostic for the boundary program to characterize the edge plasma.

Accounting of the Power Balance for Neutral-beam heated H-Mode Plasmas in NSTX

International Nuclear Information System (INIS)

Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H.

2004-01-01

A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor

Technological development of the Monobloc Divertor Concept

International Nuclear Information System (INIS)

DiPietro, E.; Brossa, M.; Guerreschi, U.; Suresh, D.; Cardella, A.

1992-01-01

This paper reports on a technological program devoted to the assessment of the feasibility and the qualification of the Monobloc Divertor Concept for the divertor of the NET/ITER Machine which has been developed with the joint collaboration between ENEA, the NET Team, Ansaldo DNT and Metallwerk Plansee. The basic idea guiding the development of the monobloc divertor consists in obtaining a component suitable to sustain the operation thermal loads, attaining peak values in the range of 15 MW/2 in steady state conditions, by a proper arrangement of refractory tiles (acting as an armour) directly brazed to the cooling pipes. In the first phase the main activities have been devoted to find a reliable joint between the armour and the cooling pipes. A number of candidate armour materials have been investigated chosen among the most promising CFC currently available in combination with molybdenum alloys (T2M and Mo41Re) and dispersion strengthened copper. The most relevant results of the test activity including the comparison of different brazing alloys and techniques and the evaluation of suitable NDE techniques are reported

Spatial radiation profiles in the ASDEX Upgrade divertor for detached plasmas

International Nuclear Information System (INIS)

Wenzel, U.; Thoma, A.; Dux, R.; Fuchs, C.; Herrmann, A.; Hirsch, S.; Kallenbach, A.; Kastelewicz, H.; Laux, M.; Mast, F.; Napiontek, B.

1997-01-01

In this paper we describe impurity line emission measurements in the divertor of ASDEX Upgrade during high power neutral beam heated discharges. We focus on detached conditions where the dominating part of the radiation comes from the X-point region. Spatially resolved line emission in the VUV and visible spectral region of the intrinsic carbon and additionally puffed impurities (neon and nitrogen) is presented. A simple interpretation of the line emission profiles is given and they are also compared to the results of bolometry. (orig.)

Experimental study of the topological aspect of the ergodic divertor in Tore-supra tokamak; Etude experimentale des aspects topologiques du divertor ergodique de Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Costanzo, L

2001-10-01

The control of power deposition onto plasma facing components in tokamaks is a determining factor for future thermonuclear fusion reactors. Plasma surface interaction can be performed using limiters or divertors. The ergodic divertor installed on Tore Supra is an atypical example of a magnetic divertor. It consists in applying a magnetic perturbation which establishes a particular topology of the plasma in contact with the wall (edge plasma). We carried out dedicated experiments in order to study parallel heat flux which strike the divertor neutralizers. This quantitative and qualitative analysis of heat flux as a function of experimental conditions allows to determine the profiles of power deposition along the neutralizers. The influence of plasma electron density, additional heating, impurities and injected gas was established. An experimental study of the sheath heat transmission factor {gamma} was carried out by correlating measurements made with Langmuir probes and infrared imaging. This study gave rise to a major conclusion: for ohmic discharges with deuterium injection and most of the time with helium, it was experimentally confirmed that {gamma}=7 in agreement with classical sheath theory. However, an increase of this factor with additional power has been shown. Detached plasma, which is an attractive regime in order to reduce the power deposition, requires an optimized control. A new measurement of the detachment onset has been developed. It is based on the variation of heat flux onto the plates derived from infrared measurements. A detachment cartography with the determination of a new 2D 'IR' Degree of Detachment was carried out allowing to locate the zone where the detachment starts. We can apply this concept both to other tokamaks such as JET and ITER. A comparison between the axisymmetric divertor and the ergodic divertor is also presented concerning the power deposition in the two configurations. Low heat flux with the ergodic divertor is a

Biased divertor performance under auxiliary heating conditions on the TdeV tokamak

International Nuclear Information System (INIS)

Decoste, R.; Lachambre, J.L.; Demers, Y.

1994-01-01

Plasma biasing has been shown on TdeV in the ohmic regime to be very promising for divertor applications. Negative biasing, with shortened SOL density gradients, improves the divertor performance, whereas positive biasing, with longer gradients, does not do much for the divertor. The next objectives were to extrapolate those results to auxiliary heated plasmas and optimize/simplify the biasing geometry for future upgrades. New results are now available with an improved divertor geometry and auxiliary heating/current drive provided by a new lower hybrid (LH) system. The new geometry, optimized for positive biasing with predictably acceptable negative biasing performances, allows for a fair comparison between the two polarities. (author) 4 refs., 5 figs

Divertor and scoop limiter experiments on PDX

International Nuclear Information System (INIS)

McGuire, K.; Beirsdorfer, P.; Bell, M.

1985-01-01

Routine operation in the enhanced-energy-confinement (or H-mode) regime during neutral-beam injection was achieved by modifying the PDX divertor hardware to inhibit the influx of neutral gas from the divertor region to the main plasma chamber. A particle scoop limiter has been studied as a mechanical means of controlling particles at the plasma edge, and neutral-beam-heated discharges with this limiter show similar confinement times (normalized to tausub(E)/Isub(p)) to average H-mode plasma. Two new instabilities are observed near the plasma edge in PDX during H-mode operation. The first, a quasi-coherent fluctuation, occurred in bursts at well-defined frequencies (Δω/ω<=0.1) in the range 50 to 180 kHz, and had no obvious effects on confinement. The second instability, the edge relaxation phenomena (ERP), did cause deterioration in the global confinement time. The ERPs are characterized by sharp spikes in the divertor plasma density, Hsub(α) emission, and on the X-ray signals they appear as sawtooth-like relaxations at the plasma edge with an inversion radius near the separatrix. Attempts to obtain high βsub(T) in the H-mode discharges were hampered by a deterioration in the H-mode confinement and major disruptions which limited the achievable βsub(T). A study of the stability of both the limiter L-mode and divertor H-mode discharge close to the theoretical β boundary showed that the major disruptions observed there are sometimes caused by a fast growing m/n=1/1 mode with no observable external precursor oscillations. (author)

The remote exchange of the JET divertor

International Nuclear Information System (INIS)

Pick, M.

1999-01-01

In 1997 a series of experiments were performed in the JET machine using deuterium-tritium (D-T) mixtures and resulting in discharges with record breaking fusion power and fusion energy. The experiments demonstrated a key technology required for fusion, namely the on-line operation of a tritium fuel re-processing plant. These experiments left the inside of the JET vessel inaccessible to manned access for approximately one year. During this time, the complete Mark IIA divertor, a major system within the torus, was successfully removed and replaced with a new divertor design, the Mark II Gas Box divertor, using only remote handling techniques. This was the first application of the JET remote handling system and a demonstration of a further key ITER technology. The paper explains the methodology and operational approach taken to achieve the results using the remote handling system developed at JET. It describes the remote handling equipment including the force-reflecting servo-manipulator, the specialised tools designed, the facilities needed, and the trials, planning and training carried out to ensure the safe, reliable and rapid completion of the remote handling tasks. The planned tasks are outlined including the execution of the novel procedure for a remote, sub-millimetre precision, dimensional survey of the divertor support structure using digital photogrammetry. Furthermore the paper shows how the adaptability of the system was used to successfully undertake a large number of unplanned tasks including the removal of damaged tiles, a damaged diagnostic system and the vacuum cleaning of diagnostic windows. (author)

ELM-induced transient tungsten melting in the JET divertor

Science.gov (United States)

Coenen, J. W.; Arnoux, G.; Bazylev, B.; Matthews, G. F.; Autricque, A.; Balboa, I.; Clever, M.; Dejarnac, R.; Coffey, I.; Corre, Y.; Devaux, S.; Frassinetti, L.; Gauthier, E.; Horacek, J.; Jachmich, S.; Komm, M.; Knaup, M.; Krieger, K.; Marsen, S.; Meigs, A.; Mertens, Ph.; Pitts, R. A.; Puetterich, T.; Rack, M.; Stamp, M.; Sergienko, G.; Tamain, P.; Thompson, V.; Contributors, JET-EFDA

2015-02-01

The original goals of the JET ITER-like wall included the study of the impact of an all W divertor on plasma operation (Coenen et al 2013 Nucl. Fusion 53 073043) and fuel retention (Brezinsek et al 2013 Nucl. Fusion 53 083023). ITER has recently decided to install a full-tungsten (W) divertor from the start of operations. One of the key inputs required in support of this decision was the study of the possibility of W melting and melt splashing during transients. Damage of this type can lead to modifications of surface topology which could lead to higher disruption frequency or compromise subsequent plasma operation. Although every effort will be made to avoid leading edges, ITER plasma stored energies are sufficient that transients can drive shallow melting on the top surfaces of components. JET is able to produce ELMs large enough to allow access to transient melting in a regime of relevance to ITER. Transient W melt experiments were performed in JET using a dedicated divertor module and a sequence of IP = 3.0 MA/BT = 2.9 T H-mode pulses with an input power of PIN = 23 MW, a stored energy of ˜6 MJ and regular type I ELMs at ΔWELM = 0.3 MJ and fELM ˜ 30 Hz. By moving the outer strike point onto a dedicated leading edge in the W divertor the base temperature was raised within ˜1 s to a level allowing transient, ELM-driven melting during the subsequent 0.5 s. Such ELMs (δW ˜ 300 kJ per ELM) are comparable to mitigated ELMs expected in ITER (Pitts et al 2011 J. Nucl. Mater. 415 (Suppl.) S957-64). Although significant material losses in terms of ejections into the plasma were not observed, there is indirect evidence that some small droplets (˜80 µm) were released. Almost 1 mm (˜6 mm3) of W was moved by ˜150 ELMs within 7 subsequent discharges. The impact on the main plasma parameters was minor and no disruptions occurred. The W-melt gradually moved along the leading edge towards the high-field side, driven by j × B forces. The evaporation rate determined

Null geodesics and wave front singularities in the Gödel space-time

Science.gov (United States)

Kling, Thomas P.; Roebuck, Kevin; Grotzke, Eric

2018-01-01

We explore wave fronts of null geodesics in the Gödel metric emitted from point sources both at, and away from, the origin. For constant time wave fronts emitted by sources away from the origin, we find cusp ridges as well as blue sky metamorphoses where spatially disconnected portions of the wave front appear, connect to the main wave front, and then later break free and vanish. These blue sky metamorphoses in the constant time wave fronts highlight the non-causal features of the Gödel metric. We introduce a concept of physical distance along the null geodesics, and show that for wave fronts of constant physical distance, the reorganization of the points making up the wave front leads to the removal of cusp ridges.

Comparison of Ne and Ar seeded radiative divertor plasmas in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Nakano, T., E-mail: nakano.tomohide@jaea.go.jp

2015-08-15

In H-mode plasmas with Ne, Ar and a mixture of Ne and Ar injection, the divertor radiation power fractions amongst these impurities in addition to an intrinsic impurity, C, are investigated. In plasmas with the inner divertor plasma attached, carbon is the biggest radiator, whichever impurity, Ne, Ar or a mixture of Ar and Ne is injected. In contrast, in plasmas with the inner divertor plasma detached, Ne is the biggest radiator due to a significantly high recombination radiation from Ne VIII. Ar is always a minor contributor in plasmas with the inner divertor both attached and detached.

Development of a full-size divertor cassette prototype for ITER

Energy Technology Data Exchange (ETDEWEB)

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

1996-10-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

A preliminary analysis has been performed examining the temperature distribution in the Divertor Primary Heat Transfer System (PHTS) piping and the divertor itself during the gas baking process. During gas baking, it is required that the divertor reach a temperature of 350 C. Thermal losses in the piping and from the divertor itself require that the gas supply temperature be maintained above that temperature in order to ensure that all of the divertor components reach the required temperature. The analysis described in this report was conducted in order to estimate the required supply temperature from the gas heater.

Optical design for divertor Thomson scattering system for JT-60SA

International Nuclear Information System (INIS)

Kajita, Shin; Enokuchi, Akito; Hatae, Takaki; Itami, Kiyoshi; Hamano, Takashi; Kado, Shinichiro; Ohno, Noriyasu; Takeyama, Norihide

2014-01-01

Highlights: •A detailed designing for collection optical system of divertor Thomson scattering system in JT-60SA is conducted. •The assessment of the density and temperature errors of the measurement system is conducted. •It is shown that the measurement could be done with the temperature error of 50% when the density was 10 20 m −3 . •The availability of the laser transmission mirrors for the measurement system is discussed. •Several guidelines to improve the measurement system are discussed. -- Abstract: Optical design for divertor Thomson scattering system in JT-60SA has been conducted. The measurement system will use a Nd:YAG laser at 1064 nm, and scattered photons are collected by a collection optical system. The collection optics consists of primary mirror, secondary mirror, relay optics, and fiber collection optics. The laser transmission mirror and collection optics were designed to be installed in a slender lower port of JT-60SA. The assessment of the measurement errors in temperature was conducted for the designed collection optical system. Because of spatial limitation, the solid angle from the measurement points would be small especially for the measurement points in high field side, and consequently, the temperature errors in the high field side would be considerably large. The effects of several improvements on the error are discussed. Moreover, an assessment for the in-vessel laser transmission metallic mirrors is conducted for the present design

pyNSMC: A Python Module for Null-Space Monte Carlo Uncertainty Analysis

Science.gov (United States)

White, J.; Brakefield, L. K.

2015-12-01

The null-space monte carlo technique is a non-linear uncertainty analyses technique that is well-suited to high-dimensional inverse problems. While the technique is powerful, the existing workflow for completing null-space monte carlo is cumbersome, requiring the use of multiple commandline utilities, several sets of intermediate files and even a text editor. pyNSMC is an open-source python module that automates the workflow of null-space monte carlo uncertainty analyses. The module is fully compatible with the PEST and PEST++ software suites and leverages existing functionality of pyEMU, a python framework for linear-based uncertainty analyses. pyNSMC greatly simplifies the existing workflow for null-space monte carlo by taking advantage of object oriented design facilities in python. The core of pyNSMC is the ensemble class, which draws and stores realized random vectors and also provides functionality for exporting and visualizing results. By relieving users of the tedium associated with file handling and command line utility execution, pyNSMC instead focuses the user on the important steps and assumptions of null-space monte carlo analysis. Furthermore, pyNSMC facilitates learning through flow charts and results visualization, which are available at many points in the algorithm. The ease-of-use of the pyNSMC workflow is compared to the existing workflow for null-space monte carlo for a synthetic groundwater model with hundreds of estimable parameters.

Magnetic Fluctuations during plasma current rise of divertor discharge in JT-60

International Nuclear Information System (INIS)

Ushigusa, Kenkichi; Kikuchi, Mitsuru; Hosogane, Nobuyuki; Tsuji, Syunji; Hayashi, Kazuo.

1986-03-01

During a current rise phase in the JT-60 divertor discharge, a series of magnetic fluctuations which do not rotate poloidally (phase-locking) is observed. They cause a cooling of plasma periphery and an enhancement of H α emission in the divertor chamber. A significant increase in β P + 1 i /2 with minor disruptions during the phase-locked magnetic fluctuation suggests a relaxation of the current profile in the current rise phase of the divertor discharge. (author)

Stochasticity about a poloidal divertor separatrix

International Nuclear Information System (INIS)

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

1986-10-01

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

Molecular deuterium behaviour in tungsten divertor on JET

Energy Technology Data Exchange (ETDEWEB)

Sergienko, G., E-mail: g.sergienko@fz-juelich.de [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Arnoux, G. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Brezinsek, S.; Clever, M.; Huber, A.; Kruezi, U. [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Meigs, A.G. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Mertens, Ph.; Samm, U. [Institute of Energy and Climate Research –Plasma Physics, Forschungszentrum Jülich, EURATOM Association, Trilateral Euregio Cluster, D-52425 Jülich (Germany); Stamp, M. [Euratom-CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom)

2013-07-15

Molecular spectroscopy was used to observe molecular deuterium at the outer strike point of the new bulk tungsten JET divertor. The rotational and vibrational populations of the deuterium molecules in the ground state were determined from the deuterium Q-branches of Fulcher-α band emission (d{sup 3}Π{sub u}{sup -}→a{sup 3}Σ{sub g}{sup +}) in the 600–640 nm spectral range. For L-mode plasmas in the low recycling regime the molecular emission maximum is located in the vicinity of the strike point. The spatial profile of the emission was strongly modified during plasma detachment in both L- and H-mode plasmas. The rotational temperature of excited molecules reached 2760 K in L-mode. The vibrational population has a peculiarity: a remarkably high population of the d{sup 3}Π{sub u}{sup -}(v = 0) vibrational level indicating a non-Boltzmann vibrational distribution of D{sub 2} in tungsten environment.

He-cooled divertor development for DEMO

International Nuclear Information System (INIS)

Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

2007-01-01

Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

Automated magnetic divertor design for optimal power exhaust

Energy Technology Data Exchange (ETDEWEB)

Blommaert, Maarten

2017-07-01

The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation

Automated magnetic divertor design for optimal power exhaust

International Nuclear Information System (INIS)

Blommaert, Maarten

2017-01-01

The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation. These flaws

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

Science.gov (United States)

Pengfei, ZHANG; Ling, ZHANG; Zhenwei, WU; Zong, XU; Wei, GAO; Liang, WANG; Qingquan, YANG; Jichan, XU; Jianbin, LIU; Hao, QU; Yong, LIU; Juan, HUANG; Chengrui, WU; Yumei, HOU; Zhao, JIN; J, D. ELDER; Houyang, GUO

2018-04-01

Modeling with OEDGE was carried out to assess the initial and long-term plasma contamination efficiency of Ar puffing from different divertor locations, i.e. the inner divertor, the outer divertor and the dome, in the EAST superconducting tokamak for typical ohmic plasma conditions. It was found that the initial Ar contamination efficiency is dependent on the local plasma conditions at the different gas puff locations. However, it quickly approaches a similar steady state value for Ar recycling efficiency >0.9. OEDGE modeling shows that the final equilibrium Ar contamination efficiency is significantly lower for the more closed lower divertor than that for the upper divertor.