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The power exhaust concept and an appropriate divertor design are common critical issues for tokamak DEMO design activities which have been carried out in Europe, Japan, China, Korea and the USA. Conventional divertor concepts and power exhaust studies for recent DEMO designs (Pfusion = 1 – 2 GW, Rp = 7 – 9 m) are reviewed from the viewpoints of the plasma physics issues and the divertor engineering design. Radiative cooling is a common approach for the power fusion scenario. Requirements on the main plasma radiation fraction (fradmain = Pradmain/Pheat) and the plasma performance constrain the divertor design concept. Different challenges contribute to optimizing the future DEMO designs: for example, (i) increasing the main plasma radiation fraction for ITER-level Psep/Rp designs and simplifying the divertor geometry, and (ii) extending ITER divertor geometry with increasing divertor radiation (Praddiv) for larger Psep/Rp ≥ 25MWm−1 designs. Power exhaust simulations with large Psep = 150 – 300 MW have been performed using integrated divertor codes considering an ITER-based divertor geometry with longer leg length (1.6 – 1.7 m), as in a common baseline design. Geometry effects (ITER like geometry or more open one without baffle) on the plasma detachment profile and the required divertor radiation fraction (fraddiv = Praddiv/Psep) were key aspects of these studies. All simulations showed that the divertor plasma detachment were extended widely across the target plate with a reduction in the peak heat load of qtarget ≤ 10 MWm−2 for the large fraddiv = 0.7 – 0.8, while the peak qtarget location and value were noticeably different in the partially detached divertor. Simulation results also demonstrated that radial diffusion coefficients of the heat and particle fluxes were critical parameters for DEMO divertor design, and that effects of plasma drifts on outboard-enhanced asymmetry of the heat flux, suggested the need for longer divertor leg to ensure the existence of a detached divertor operation with qtarget ≤ 10 MWm−2.Integrated design of the water cooled divertor target, cassette body (CB) and cooling pipe routing has been developed for each DEMO concept, based on the ITER-like tungsten monoblock (W-MB) with Cu-alloy cooling pipes. Engineering design adequate under higher neutron irradiation condition was required. Therefore, inlet coolant temperature (Tcool) was increased. In current designs, it still shows a large potential variation between 70 °C and 200 °C. The influence of thermal softening on the Cu-alloy (CuCrZr) pipe was fostered near the strike-point when the high qtarget of ∼10 MWm−2 was studied. Improved technologies for high heat flux components based on the ITER W-MB unit have been developed for EU-DEMO. Different coolant conditions (low- and high-Tcool) were provided for Cu-alloy and reduced activation ferritic martensitic (RAFM) steel heat sink units, respectively. The high-Tcool coolant was also considered for the CB and supporting structures. Appropriate conditions for the high-Tcool coolant, i.e. 180 °C/ 5 MPa (EU-DEMO) and 290 °C/ 15 MPa (JA-DEMO, CFETR and K-DEMO), will be determined in the future optimizations of the divertor and DEMO design.

For the on-orbit high resolution satellites, attitude jitter is a common and complex phenomenon which may deteriorate the image quality and mapping accuracy. In this paper, an image-based jitter inversion and estimation model is derived considering time delay integration (TDI) effect, and the attitude is then updated by taking into account the estimated jitter information. The imagery is finally resampled through virtual re-imaging using the updated accurate attitude to improve the image geometric quality. The proposed method is validated by experiments using the Chinese Ziyuan-3 (ZY-3) satellite imagery.

The mission of WEST (tungsten-W Environment in Steady-state Tokamak) is to explore long pulse operation in a full tungsten (W) environment for preparing next-step fusion devices (ITER and DEMO) with a focus on testing the ITER actively cooled W divertor in tokamak conditions. Following the successful completion of phase 1 (2016-2021), phase 2 started in December 2022 with the lower divertor made entirely of actively cooled ITER-grade tungsten mono-blocks. A boronization prior the first plasma attempt allowed for a smooth startup with the new divertor. Despite the reduced operating window due to tungsten, rapid progress has been made in long pulse operation, resulting in discharges with a pulse length of 100 s and an injected energy of around 300 MJ per discharge. Plasma startup studies were carried out with equatorial boron nitride limiters to compare them with tungsten limiters, while Ion Cyclotron Resonance Heating assisted startup was attempted. High fluence operation in attached regime, which was the main thrust of the first campaigns, already showed the progressive build up of deposits and appearance of dust, impacting the plasma operation as the plasma fluence increased. In total, the cumulated injected energy during the first campaigns reached 43 GJ and the cumulated plasma time exceeded 5 h. Demonstration of controlled X-Point Radiator regime is also reported, opening a promising route for investigating plasma exhaust and plasma-wall interaction issues in more detached regime. This paper summarises the lessons learned from the manufacturing and the first operation of the ITER-grade divertor, describing the progress achieved in optimising operation in a full W environment with a focus on long pulse operation and plasma wall interaction.

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent ‘pedestal-to-divertor’ integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T _e and radiation which need control. High- β _P scenarios with q _min > 2, q _95 –7.9, β _N > 4, β _T –3.3% and H _98y2 > 1.5 are sustained with high density ( $\bar n$ = 7E19 m ^−3 , f _G –1) for 6 τ _E , improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

To address the needs for a fusion pilot plan design, DIII-D/EAST joint experiments on DIII-D have demonstrated high normalized beta β _N ∼ 4.2, toroidal beta β _T ∼ 3.3% with q _min > 2, q _95 ⩽ 8 sustained for more than six energy confinement times in high poloidal beta regime. The excellent energy confinement quality ( H _98y2 ∼ 1.8) is achieved with an internal transport barrier at high line-averaged Greenwald density fraction f _Gr > 0.9. The trapped gyro-Landau fluid (TGLF) modeling of the transport characteristics shows that the beam-driven rotation does not play an important role in the high confinement quality. The modeling also captures very well several transport features, giving us confidence in using integrated modeling to project these experimental results to future machines. The high-performance phase is terminated by fast-growing modes triggered near the n = 1 ideal-wall kink stability limit. New radio frequency (RF) capabilities for off-axis current drive could remove the residual ohmic current to achieve a fully non-inductive state, and improve the mode–wall coupling to increase the ideal-wall β _N limit, enabling sustainment of the fully non-inductive high performance plasma in stationary conditions.

The neutral-induced material erosion is still an unsolved issue for ITER and future reactors, which influences the lifetime of the first wall material and tritium retention rate. A low energy neutral particle analyzer (LENPA) and a quartz crystal microbalance (QMB) were recently installed together on the equatorial port of sector H in EAST to study neutral-induced material erosion. The QMB can provide real-time and in-situ measurements on the erosion process of the wall material, while the neutral energy spectrum given by LENPA can help to calculate the material erosion rate for comparison with QMB results. In the 2021 EAST summer campaign, neutral-induced Al erosion was successfully studied with the two available diagnostics for the first time. In a series of long-pulse discharges, the average Al erosion rates measured by the QMB are consistent with the theoretical calculations using the LENPA data. Furthermore, the changes in plasma density, heating power and the real-time lithium powder injection result in different neutral flux and energy, which lead to changes in the neutral-induced material erosion rates.

Tungsten (W) erosion and edge transport are investigated for EAST L-mode discharges with different gas injection. It is found that W erosion can be suppressed or mitigated by Ne or D2 seeding when divertor detachment is achieved. Compared to edge D2 fueling, Ne seeding from the divertor target is favorable for full detachment condition and thus W erosion suppression. Increasing the upstream plasma density by edge D2 fueling can affect the divertor condition, which may lead to a W erosion mitigation. D2 puffing at divertor target is less effective on increasing the upstream plasma density than OMP D2 puffing, and thus it is less effective on reduction of W erosion rates. W gross erosion profiles with different amount of injected D2 at the divertor are reproduced by a mixed material W erosion model, which indicates that there exist a Li-C overlayer on the W surface of EAST divertor. Ne and D2 injection are also found to have different impact on the normalized core W density. For the attached divertor condition, divertor Ne seeding will increase W leakage, but a suitable D2 fueling from divertor target can strengthen the edge W screening. After the divertor detachment, although the W source is dramatically reduced, the W core density is kept in a high level for the Ne seeding discharges, and even increased for the D2 fueling discharges.

The neutral energy spectrums with an energy range of 20–3000 eV was measured by the low energy neutral particle analyzer (LENPA) in the 2021 EAST summer campaign. The proportion of the neutral particles below 1000 eV measured by LENPA is more than 85 %. In the ohmic discharges, the integrated neutral flux increases with increasing the line-averaged density. However, the neural particles with higher energy (1000–3000 eV) are insensitive to the line-averaged density. In addition, the neutral flux in the energy range of 20–1000 eV of the discharges with supersonic molecular beam injection (SMBI) was lower than those without SMBI, which was different from neutrals in the range of 1000–3000 eV. The neutral flux increases with higher auxiliary heating power in all energy range. Compared to the ohmic discharges, the mean energy of neutral particles in auxiliary heated discharges is increased significantly.

Physical sputtering caused by particle bombardment is believed to be the main erosion mechanism of W materials in fusion devices, in which W atoms are the sputtering products. However, the tungsten deuteride molecule (WD) spectra have been observed in both TEXTOR and ASDEX Upgrade, which was believed to be the product of chemically assisted physical sputtering (CAPS), a new sputtering mechanism that has been proposed in recent years. In this paper, we report the spectroscopic observation of WD molecules in the EAST W divertor. The behaviors of WD molecules sputtering are compared with W atoms sputtering via the spectral measurements of the ro-vibrational band emission of WD 6Π → 6Σ+ in the spectral range between 673 nm and 678 nm and the WI line emission at 400.9 nm. The physical sputtering characters and chemical sputtering characters of WD molecule sputtering were confirmed in EAST. The measurements in EAST showed that there is an energy threshold for WD molecules sputtering and that the sputtering energy threshold of WD molecules is smaller than that of W atoms. Furthermore, the dependence of WD molecule sputtering efficiency (the absolute WD photon flux (PhWD) normalized to the particle flux (Γion) reaching the target with a constant Te) on impact energy and heat flux were studied, presenting significantly different behaviors compared with W atom sputtering. In addition, it is found that the decrease of WD sputtering efficiency with the heat flux is accompanied by the rise of the photon flux of Dδ (410.06 nm) normalized to the particle flux (Γion) at the divertor target, which may imply the enhanced deuterium desorption at the W surface. The further increase of heat flux hitting the target surface could elevate the surface temperature and benefit the deuterium release from the surface, which may influence the formation of WD molecules at the surface layer.

The scrape-off layer (SOL) power width ( λ _q ) is an important parameter for predicting the heat load on divertor targets for future magnetically confined devices. Currently, the underlying physics for λ _q scaling is not fully understood. This paper extends the previous SOL particle flux width ( λ _js , λ _q ≈ λ _js is assumed, j _s is the measured ion current density used as a proxy for particle flux) scaling measured by the inboard divertor Langmuir probes (Div-LPs) in deuterium plasmas (Liu et al 2019 Plasma Phys. Control. Fusion 61 045001) to the λ _js scalings measured by the outboard Div-LPs in both deuterium and helium plasmas on EAST. A systematic method has been used to correct the upper-outer Div-LP measurements to reduce the measurement uncertainty of λ _js . About 520 discharges are selected for the construction of six databases (H-mode, L-mode, and Ohmic for deuterium and helium plasmas) to scale λ _js . Since the published λ _q scalings do not fit EAST λ _js well in the databases, four separate λ _js scalings are proposed using statistical method to minimize the scaling parameters. It is found that λ _js has a robust scaling dependence on the plasma line-averaged density ( ${\bar n_{\text{e}}}$ ) for both deuterium and helium plasmas and a strong positive scaling dependence on the divertor leg length ( L _div,leg ). The scaling exponent of the stored energy agrees well with the previous λ _q / λ _js scalings for deuterium plasmas. The near unity scaling exponents of edge safety factor and L _div,leg reveal the importance of parallel connection length in determining λ _js . The positive scaling dependence of λ _js on ${\bar n_{\text{e}}}$ and the 2 ∼ 4 times larger scaling amplitude compared with the published λ _q scalings probably indicate the turbulence broadening of λ _js on EAST. The comparison of λ _js for deuterium and helium plasmas using the constructed databases shows no significant difference. Statistical comparison of the normalized λ _js with different ELM types suggests that the grassy-ELM λ _js is slightly larger than the inter large-ELM (type-I ELM) λ _js for both deuterium and helium plasmas.

Tungsten (W) is one of the most promising plasma-facing materials for future fusion devices. Although its melting point is the highest among all metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for ITER and future reactors. Actively cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, and provide a good opportunity for direct comparison of the damage mechanism for the two types of PFCs. Various in situ melting phenomena on the lower divertor have been observed by CCD cameras, which have been further verified by post-mortem inspections. Severe melting and even exfoliation of the edge-beveled W plates were observed on some W/Cu flat-type components at horizontal outer targets. Many droplets were ejected during long-pulse operations, which induced a significant increase of W impurities and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even caused disruptions. Two different shaping structures of flat-type PFCs show different positions of melting and the corresponding mechanisms. Slight melting was found on the sharp leading edges of W/Cu monoblocks between cassette modules (inter-CM) for horizontal targets with small droplet ejection, which was much improved compared to that observed on the upper W divertor, illustrating that the application of a large-sized bevel chamfer inter-CM was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for the divertor provides important experiences and lessons for the engineering design and optimization of divertor PFCs in future fusion devices.

The edge turbulence characteristics and the induced radial transport have been investigated in edge localized mode (ELM) suppression by using the $n = $ 4 resonant magnetic perturbation coils on EAST, with ${q_{95}} = $ 3.6 and the electron collisionality $\nu _{\text{e}}^* \approx $ 0.5. During ELM suppression, the edge turbulence is enhanced dramatically, as measured by the reciprocating probe and the poloidal correlation reflectometry. In the near SOL, the low frequency turbulence (250 kHz), which could be beneficial to form the observed turbulent transport. The estimated upstream cross-field particle flux is consistent with the total particle flux deposited on divertor targets, demonstrating that the enhanced radial turbulent particle transport is an important mechanism for particle exhaust in ELM suppression.

To evaluate the erosion effects of charge exchange (CX) neutrals on first wall materials, aluminum (Al) coated samples installed in a cylindrical sleeve were exposed to EAST plasmas with the Material and Plasma Evaluation System (MAPES). The 3D-GAPS code was firstly used to simulate the dedicated experiments with the neutral energy spectrums measured by the Low Energy Neutral Particle Analyzer located at the MAPES. The modeled Al erosion rates with isotropic angular distributions for incident D neutrals are consistent with the measured erosion thicknesses of Al layer by Rutherford backscattering spectrometry before and after the plasma exposure. The effects of neutral reflection coefficients and the heights of cylindrical sleeve on material erosion are also studied. The measured material erosion rate by CX neutrals in EAST verified by the modeling can help to us understand the effects of neutrals on first wall erosion, which is an important erosion process in future fusion reactors.

Abstract Background The aim of this study was to provide information on the semen quality pattern of infertile men and age thresholds for semen parameters in China. Methods This was a retrospective cross-sectional study investigating 71,623 infertile men from the Reproductive and Genetic Hospital of CITIC Xiangya in Hunan, China, from 2011 to 2017. The Kruskal-Wallis test, Mann-Kendall test, linear regression model and joinpoint regression were used. Results Although erratic changes were observed in the median semen parameters (sperm concentration 40.1–52.1 × 106/ml, total sperm count 117.8–153.1 × 106, sperm progressive motility 33.4–38.1%) during the 7 years of observation, no significant decrease in semen quality was found, and 47.88% of infertile men showed normal semen parameters according to the World Health Organization (WHO) criteria. According to the joinpoint regression analysis, sperm progressive motility appeared to decrease earlier than the sperm concentration and total sperm count (at 28, 58, and 42 years of age, respectively). Conclusions There is no evidence of a deterioration in semen quality among infertile men in Hunan, China. Semen parameters decreased with increasing age, with turning points noted at different ages. Semen parameters are not absolute evidence for the assessment of male fertility potential. Therefore, we believe that, among semen parameters, the sperm concentration is the best predictor of fertility for ART, followed by motility. Decreased sperm motility may affect natural pregnancy, but it is not necessary for successful IVF.

As net erosion of tungsten plasma-facing components in tokamak divertors is strongly reduced by tungsten prompt redeposition, predicting and monitoring tungsten prompt redeposition is crucial to estimate the lifetime of tungsten PFCs and the source of tungsten impurities in tokamak divertors. To that end, a theoretical and numerical analysis of tungsten prompt redeposition near divertor targets is presented. It is shown that the reduction of the Larmor radius of tungsten impurities due to multiple ionizations during the first gyro-orbit of those impurities does not significantly lower tungsten prompt redeposition, as the sheath electric field is much stronger than the Lorentz force in the sheath region because of the large inertia of tungsten particles. As a consequence, only tungsten impurities which are ionized near or beyond the sheath entrance avoid prompt redeposition. Tungsten prompt redeposition is therefore mainly governed by the ratio of the ionization mean-free path of sputtered neutral tungsten particles over the sheath width. This governing parameter has similar values for divertor plasma conditions in DIII-D experiments and in plasma conditions expected in the far-SOL of the ITER divertor, indicating that investigations of tungsten erosion in smaller devices like DIII-D are relevant for future fusion reactors like ITER. Furthermore, this governing parameter scales linearly with the magnetic field strength. Tungsten net erosion in the divertor region might thus significantly increase in fusion devices operating at high magnetic field. It is also observed that the fraction of tungsten impurities which promptly redeposit is only correlated to the fraction of impurities in the highest charge state reached during the first gyro-orbit. As a consequence, in-situ monitoring of tungsten prompt redeposition in divertors requires monitoring photon emissions associated with the ionizations of tungsten impurities in charge states Z>2+, typically W3+, W4+and W5+for plasma conditions expected in the far-SOL of the ITER divertor.

Hot spot is a serious challenge limiting long pulse operation with LHCD (lower hybrid current drive) in tokamak. In order to mitigate the hot spot in guard limiter and improve LHCD capability, effect of edge plasma density on inducing hot spot and current drive has been studied in EAST. The temperature in the guard limiter of the LH antenna, inducing hot spot directly, increases with edge density and LH power. Studies show that the hot spot is mainly ascribed to the heat flux in front of LH antenna. Further simulation indicates that such spot corresponds to the peak position of edge density due to local LH electric field. In addition, due to the stronger parametric instability (PI) behavior in the case of higher edge density, the current drive capability decreases with edge density. Strike-point splitting behaviors appear with density increase, in agreement with current profiles in the edge region and the reduction of total driven current, suggesting that more power is deposited in edge region, which does more contribution to hot spot. Above studies offer one possible idea to optimize the edge density so as to satisfy the coupling, mitigate the heat flux in the guard limiter, and improve current drive capability for fusion device.

Two new quartz crystal microbalance (QMB) systems have been installed at different locations in the mid-plane of EAST to measure erosion and deposition of the first wall materials. The two QMBs with different distances from the plasma show complex and distinct erosion and deposition behavior. The QMB in the mid-plane H port closer to the plasma shows a relatively higher net erosion result with an average erosion rate of 1.34 ng/(cm2s). The normal discharges mainly contribute to the total net erosion and the maximum erosion amount reaches up to 110 ng/cm2. Contrarily, a slight net deposition of 0.11 ng/(cm2s) is obtained for the QMB in the mid-plane G port further away from the plasma and the erosion and deposition amount of most discharges are much lower due to lower particle flux. The disruptive discharges lead to a total net deposition rate of 0.48 ng/(cm2s). Measurement results by the QMBs are demonstrated by a good agreement with the post-mortem analysis by Rutherford Backscattering Spectrometry (RBS).

To evaluate the erosion effects of neutral particles on the first wall materials during tokamak operation, aluminum (Al) coated samples were exposed to a sequence of plasma discharges in 2018 and 2019 EAST campaigns. In order to avoid the incidence of charged particles, the samples were installed in a molybdenum (Mo) cylindrical sleeve and positioned far from the main plasma. The erosion thicknesses of Al layer were measured by Rutherford Backscattering Spectrometry (RBS) before and after the plasma exposure. The average net erosion rates of the samples in the two campaigns are around 1.18 × 1013 atoms cm−2 s−1 and 4.64 × 1013 atoms cm−2 s−1, respectively. The different erosion rate may mainly result from different plasma density and heating power. The theoretical calculated erosion rates using the neutral distribution under different density conditions are in fair agreement with the experimental results.

Rare isotope tracer injection represents an established way for following particle transport in fusion research, but a tracer experiment is only as good as the tracer can be followed and separated from the common isotope. Physical models for transport analysis such as ERO2.0 require accurate total amounts of deposition with high spatial resolution. This work describes an extended high throughput NRA post analysis approach together with the surface analysis results of 2 technically rough graphite-tiles of 150x150 mm2 surface area exposed to D and He plasmas in EAST.The applied 1.43 MeV deuteron beam NRA offers percent range accuracy for layer thicknesses ranging from few 10 nm to 20 µm. Optimisation of the beam line and beam optics enables operation with ≤20 nA without reaching critical radiation levels, resulting in a measurement time of 5–10 s per spot. This enabled acquiring in total 3070 NRA points on the central 46 × 46 mm2 cut-outs of each sample and an adjacent part of equal size within 12 h of analysis time.The layers contained 50–60% D, ~2% 12C, and ~3% O. The central 20 mm radius around the injection hole contains 98% of the total 13C deposition. NRA finds up to 5.8 ± 0.9 × 1022 13C/m2 close to the hole with an e-folding fall-off length of about 3–10 mm in every radial direction. In the D discharge 24 ± 2% and 16 ± 2% in He of the injected 13C are found on the tiles with triangle shaped deposition profiles following ExB. The comparison to colour fringe analysis and SIMS reveals the particular strength of D-NRA regarding roughness, thick layers, and the separation of 12C and 13C deposits, but in compatible regions all methods agree. Compared to these, the NRA yields additional information indicating depositing species and physical processes.

Using observations and reanalysis, we find that changes in April precipitation variations in the northwestern US are strongly linked to March Arctic stratospheric ozone (ASO). An increase in ASO can result in enhanced westerlies in the high and low latitudes of the North Pacific but weakened westerlies in the midlatitudes. The anomalous circulation over the North Pacific can extend eastward to western North America, decreasing the water vapor concentration in the air over the northwestern United States and enhancing downwelling in the northwestern US, which results in decreased precipitation there and vice versa for the decrease in ASO. Model simulations using the Whole Atmosphere Community Climate Model version 4 (WACCM4) support the statistical analysis of observations and reanalysis data and further reveal that the ASO influences circulation anomalies over the northwestern US in two ways. Stratospheric circulation anomalies caused by the ASO changes can propagate downward to the troposphere in the North Pacific and then eastward to influence the strength of the circulation anomalies over the northwestern US. In addition, sea surface temperature anomalies over the North Pacific, which may be related to the ASO changes, would cooperate with the ASO changes to modify the circulation anomalies over the northwestern US. Our results suggest that ASO variations could be a useful predictor of spring precipitation changes in the northwestern US.

The OEDGE code is used to model tungsten erosion and transport for experiments with toroidal rings of high-Z metal tiles in the DIII-D tokamak. Such modeling is needed for both experimental and diagnostic design to have estimates of the expected core and edge tungsten density and to understand the various factors contributing to the uncertainties in these calculations. OEDGE simulations are performed using the planned experimental magnetic geometries and plasma conditions typical of both L-mode and inter-ELM H-mode discharges in DIII-D. OEDGE plasma reconstruction based on specific representative discharges for similar geometries is used to determine the plasma conditions applied to tungsten plasma impurity simulations. A new model for tungsten erosion in OEDGE was developed which imports charge-state resolved carbon impurity fluxes and impact energies from a separate OEDGE run which models the carbon production, transport and deposition for the same plasma conditions as the tungsten simulations. These values are then used to calculate the gross tungsten physical sputtering due to carbon plasma impurities which is then added to any sputtering by deuterium ions; tungsten self-sputtering is also included. The code results are found to be dependent on the following factors: divertor geometry and closure, the choice of cross-field anomalous transport coefficients, divertor plasma conditions (affecting both tungsten source strength and transport), the choice of tungsten atomic physics data used in the model (in particular ionization rate for W-atoms), and the model of the carbon flux and energy used for calculating the tungsten source due to sputtering. Core tungsten density is found to be of order 1015m−3 (excluding effects of any core transport barrier and with significant variability depending on the other factors mentioned) with density decaying into the scrape off layer. For the typical core density in the plasma conditions examined of 2 to 4 ×1019m−3, this represents a concentration on the order of 5 × 10−5.

Carbon deposition on tungsten between ELMs was investigated in DIII-D in semi-attached/detached H-mode plasma conditions using fixed outer strike point (OSP) positions. Carbon deposition during plasma exposure of tungsten was monitored in-situ by measuring the reflectivity of the tungsten sample surface. No significant carbon deposition, i.e., without strong variations of the reflectivity, was observed during these experiments including discharges at high densities. In contrast, ERO modeling predicts a significant carbon deposition on the tungsten surface for those high density plasma conditions. The surface reflectivity decreases with methane injection, consistent with increased carbon coverage, as expected. The sweeping of OSP leads to a pronounced increase of the surface reflectivity, suggesting that the strike point sweeping may provide an effective means to remove carbon coating from tungsten surface. The ERO modeling however predicts again a regime of carbon deposition for these experiments. The discrepancies between carbon deposition regime predicted by the ERO model and the experimental observations suggest that carbon erosion during ELMs may significantly affect carbon deposition on tungsten.

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

Tracer injection experiments in TEXTOR with MoF6 and WF6 lead to local deposition of about 6% for Mo and about 1% for W relative to the injected amount of Mo and W atoms. Modelling of these experiments has been done with ERO applying updated data for physical sputtering. The dissociation of the injected molecules has been treated in a simplified manner due to the lack of dissociation rate coefficients. However, with this it was possible to reproduce the observed radial penetration of Mo and W atoms into the plasma. The modelled local deposition efficiencies are about 50% for Mo and 60% for W assuming typical plasma parameters for the experimental conditions used. To reproduce the measured deposition efficiencies an enhancement factor for the erosion of deposited Mo and W has to be assumed (∼10 for Mo and ∼25 for W). Due to the rather low electron temperature Te of these plasma conditions (Te∼15eV at the location of injection), Mo and W are mostly sputtered by impurities whereas sputtering due to deuterium is negligible. A parameter study applying larger electron temperature leads to increased sputtering and thus to reduced local deposition efficiencies of about 30% for Mo and 5% for W. Though, even under these conditions enhanced erosion, albeit with reduced enhancement factors, is needed in the modelling to obtain the small measured deposition efficiencies. keywords: Erosion & Deposition, Sputtering, Impurity transport, TEXTOR, PACS: 52.40.Hf, 52.77.Dq, 52.65.Pp

As High-Z materials will likely be used as plasma-facing components (PFCs) in future fusion devices, the erosion of high-Z materials is a key issue for high-power, long pulse operation. High-Z material erosion and redeposition have been studied using tungsten and molybdenum coated samples exposed in well-diagnosed DIII-D divertor plasma discharges. By coupling dedicated experiments and modelling using the 3D Monte Carlo code ERO, the roles of sheath potential and background carbon impurities in determining high-Z material erosion are identified. Different methods suggested by modelling have been investigated to control high-Z material erosion in DIII-D experiments. The erosion of Mo and W is found to be strongly suppressed by local injection of methane and deuterium gases. The 13C deposition resulting from local 13CH4 injection also provides information on radial transport due to E ×B drifts and cross field diffusion. Finally, D2 gas puffing is found to cause local plasma perturbation, suppressing W erosion because of the lower effective sputtering yield of W at lower plasma temperature and for higher carbon concentration in the mixed surface layer.

The seasonal mean and synoptic fluctuation of the wintertime Taiwan Warm Current (TWC) were investigated using a well-validated finite volume community ocean model. The spatial distribution and dynamics of the synoptic fluctuation were highlighted. The seasonal mean of the wintertime TWC has two branches: an inshore branch between the 30 and 100 m isobaths and an offshore branch between the 100 and 200 m isobaths. The Coriolis term is much larger than the inertia term and is almost balanced by the pressure gradient term in both branches, indicating geostrophic balance of the mean current. Two areas with significant fluctuations of the TWC were identified during wintertime. One of the areas is located to the north of Taiwan with velocities varying in the cross-shore direction. These significant cross-shore fluctuations are driven by barotropic pressure gradients associated with the intrusion of the Taiwan Strait Current (TSC). When a strong TSC intrudes to the north of Taiwan, the isobaric slope tilts downward from south to north, leading to a cross-shore current from the coastal area to the offshore area. When the TSC intrusion is weak, the cross-shore current to the north of Taiwan is directed from offshore to inshore. The other area of significant fluctuation is located in the inshore area between the 30 and 100 m isobaths. The fluctuations are generally strong both in the alongshore and cross-shore directions, in particular at the latitudes 26.5 and 28° N. Wind affects the synoptic fluctuation through episodic events. When the northeasterly monsoon prevails, the southwestward Zhe-Min coastal current dominates the inshore area associated with a deepening of the mixed layer. When the winter monsoon is weakened or the southwesterly wind prevails, the northeastward TWC dominates in the inshore area.