N. Cao

Validation of nonlinear gyrokinetic simulations of L- and I-mode plasmas on Alcator C-Mod

New validation of global, nonlinear, ion-scale gyrokinetic simulations (GYRO) is carried out for L- and I-mode plasmas on Alcator C-Mod, utilizing heat fluxes, profile stiffness, and temperature fluctuations. Previous work at C-Mod found that ITG/TEM-scale GYRO simulations can match both electron and ion heat fluxes within error bars in I-mode [White PoP 2015], suggesting that multi-scale (cross-scale coupling) effects [Howard PoP 2016] may be less important in I-mode than in L-mode. New results presented here, however, show that global, nonlinear, ion-scale GYRO simulations are able to match the experimental ion heat flux, but underpredict electron heat flux (at most radii), electron temperature fluctuations, and perturbative thermal diffusivity in both L- and I-mode. Linear addition of electron heat flux from electron scale runs does not resolve this discrepancy. These results indicate that single-scale simulations do not sufficiently describe the I-mode core transport, and that multi-scale (coupled ele...

Electron critical gradient scale length measurements of ICRF heated L-mode plasmas at Alcator C-Mod tokamak

A profile for the critical gradient scale length (Lc) has been measured in L-mode discharges at the Alcator C-Mod tokamak, where electrons were heated by an ion cyclotron range of frequency through minority heating with the intention of simultaneously varying the heat flux and changing the local gradient. The electron temperature gradient scale length (LTe−1 = |∇Te|/Te) profile was measured via the BT-jog technique [Houshmandyar et al., Rev. Sci. Instrum. 87, 11E101 (2016)] and it was compared with electron heat flux from power balance (TRANSP) analysis. The Te profiles were found to be very stiff and already above the critical values, however, the stiffness was found to be reduced near the q = 3/2 surface. The measured Lc profile is in agreement with electron temperature gradient (ETG) models which predict the dependence of Lc−1 on local Zeff, Te/Ti, and the ratio of the magnetic shear to the safety factor. The results from linear Gene gyrokinetic simulations suggest ETG to be the dominant mode of turbul...

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

Abstract The ARC pilot plant conceptual design study has been extended beyond its initial scope [B. N. Sorbom et al., FED 100 (2015) 378] to explore options for managing ∼525 MW of fusion power generated in a compact, high field (B0 = 9.2 T) tokamak that is approximately the size of JET (R0 = 3.3 m). Taking advantage of ARC’s novel design – demountable high temperature superconductor toroidal field (TF) magnets, poloidal magnetic field coils located inside the TF, and vacuum vessel (VV) immersed in molten salt FLiBe blanket – this follow-on study has identified innovative and potentially robust power exhaust management solutions. The superconducting poloidal field coil set has been reconfigured to produce double-null plasma equilibria with a long-leg X-point target divertor geometry. This design choice is motivated by recent modeling which indicates that such configurations enhance power handling and may attain a passively-stable detachment front that stays in the divertor leg over a wide power exhaust window. A modified VV accommodates the divertor legs while retaining the original core plasma volume and TF magnet size. The molten salt FLiBe blanket adequately shields all superconductors, functions as an efficient tritium breeder, and, with augmented forced flow loops, serves as an effective single-phase, low-pressure coolant for the divertor, VV, and breeding blanket. Advanced neutron transport calculations (MCNP) indicate a tritium breeding ratio of ∼1.08. The neutron damage rate (DPA/year) of the remote divertor targets is ∼3–30 times lower than that of the first wall. The entire VV (including divertor and first wall) can tolerate high damage rates since the demountable TF magnets allow the VV to be replaced every 1–2 years as a single unit, employing a vertical maintenance scheme. A tungsten swirl tube FLiBe coolant channel design, similar in geometry to that used by ITER, is considered for the divertor heat removal and shown capable of exhausting divertor heat flux levels of up to 12 MW/m2. Several novel, neutron tolerant diagnostics are explored for sensing power exhaust and for providing feedback control of divertor conditions over long time scales. These include measurement of Cherenkov radiation emitted in FLiBe to infer DT fusion reaction rate, measurement of divertor detachment front locations in the divertor legs with microwave interferometry, and monitoring “hotspots” on the divertor chamber walls via IR imaging through the FLiBe blanket.

Applying X-ray Imaging Crystal Spectroscopy for Use as a High Temperature Plasma Diagnostic.

هديكچ هلاقم . يكشزپ نارود رد مسينردم هدس ود زا يلك دروخرب رد و نآ يفسـلف ياههاگديد اب قبطنم يرگ يدام هفسلف اب لبق برغ ناهج زور , درانربدولك تايرظن صوصخب , و دـش دلوتم و تملاس هب نآ شرگن ضـحم كـيژولويب هاگدـيد يراـميب ) يتخانش تسيز ( دوب . كـيدزن اب ندـش هـب هـلق ياـه كرد تسيز يتخانش ناسنا زا نارود زا رذگ و مسينردم نديسر و هب نارود سـپ مسـينردم زا ) post-modernism ( يفاـكان ، ندوب ياههاگديد تسيز يتخانش لح رد تلاكشم ناسنا و هاـگن رد ـب ه يتملاـس يراـميب و هدـيدرگ رتراكـشآ تـسا . تاقيقحت نادنمشناد ددعتم كي زا تشگزاب و وس هعماج هـب شزرا ياه يناسنا يوس زا رگيد , قفا ياه ينامرد يديدج ار هدوشگ تسا . يكشزپ نيزگياج ) Alternative Medicine ( هدنيازف دشر يا هتفاي و هفسلف تسـيز يتخانـش فـيرعت رد يتملاس و يراميب ناسنا اب دروخرب و ياج هب ار دوخ هفسلف لك يرگن عماـج اي يرـگن ) Holistic view ( هداد تـسا . رد نيا هفسلف هاگن هب ناسنا تسـيز داعبا رد يتخانـش يـنهذ ، ، يعامتجا يحور ، يـنيد ــ ) يوـنعم ( تلاماـعت و ، ناسـنا اـب هعومجم ناهج لخ تق ـ ب ه ناوـنع طيـحم يگدـنز وا , دروـم هجوت هتفرگ رارق تسا . تـسا يتدـم هلوقم نيا هك اجنآ زا و ماظن و يلاع شزومآ يملع لفاحم ثحب عوضوم ام روشك رد يتشادهب تامدخ هئارا هـتفرگ رارـق يناـمرد , اـب ييانـشآ هاگديد نيا يساسا لوصا و هچخيرات ، ميهافم , يـم يرورض دومن .X-ray spectra provide a wealth of information on high temperature plasmas; for example electron temperature and density can be inferred from line intensity ratios. By using a Johann spectrometer viewing the plasma, it is possible to construct profiles of plasma parameters such as density, temperature, and velocity with good spatial and time resolution. However, benchmarking atomic code modeling of X-ray spectra obtained from well-diagnosed laboratory plasmas is important to justify use of such spectra to determine plasma parameters when other independent diagnostics are not available. This manuscript presents the operation of the High Resolution X-ray Crystal Imaging Spectrometer with Spatial Resolution (HIREXSR), a high wavelength resolution spatially imaging X-ray spectrometer used to view hydrogen- and helium-like ions of medium atomic number elements in a tokamak plasma. In addition, this manuscript covers a laser blow-off system that can introduce such ions to the plasma with precise timing to allow for perturbative studies of transport in the plasma.