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Affordable Robust Compact reactor is a conceptual design for a Tokamak conceived by Massachusetts Institute of Technology (MIT) researchers. The design of this tokamak is under development and update. One of the key parameters for fusion reactor power plants is the tritium breeding ratio (TBR), which has to guarantee the tritium self-sufficiency. The tritium inventory circulating in a fusion power plant must be minimized. In the meantime, to enhance plant's economics, the amount of tritium generated and stored should be maximized, since it would be used to startup new reactors. Both of the aforementioned trends meet their best in a TBR as high as possible. In this work, ARC tritium breeding ratio is studied and optimized. Taking advantage of Monte Carlo neutron transport codes, several configurations of ARC's blanket and vacuum vessel have been analyzed in order to find the most effective one for a high TBR. The study takes into account different materials for the structure, such as Inconel718, V-15Cr-5Ti and Eurofer97. Moreover, it scans different width of coolant's channels and evaluates the effect of lithium-6 enrichment in the blanket looking for the best configuration in terms of TBR.

Nowadays, Fusion Energy is one of the most important sources under study. During the last years, different designs of fusion reactors were considered. At the MIT, an innovative design was created: ARC, the Affordable Robust Compact reactor. It takes advantage of the innovative aspects of recent progress in fusion technology, such as high temperature superconductors, that permit to decrease the dimensions of the machine, reaching at the same time high magnetic fields. Our main goal is the low-activation analysis of possible structural materials for the vacuum vessel, which is designed as a single-piece placed between the first-wall and the tank that contains the breeding blanket. Due to its position, the vacuum vessel is subject to high neutron flux, which can activate it and cause the reduction of the component lifetime and decommissioning problems. The activation analysis was done also for the liquid breeder FLiBe, compared with Lithium-Lead. Codes used for the low-activation analysis were MCNP and FISPACT-II. The first one is based on a neutronics model and for each component a certain neutron flux is evaluated. For FISPACT-II, the main input is the composition of the analyzed material, the neutron flux and the irradiation time. Results from FISPACT-II are the time behavior of specific activity, contact dose rate. To assess suitable structural materials for the vacuum vessel, low-activation properties were considered. Vanadium alloys turn out to be one of the best alternatives to the present material, Inconel-718. Finally, isotopic tailoring and elemental substitution methods were applied. Here, the composition of each alloy is analyzed and critical isotopes or elements are eliminated or reduced. After the modifications, new simulations are done, and those leading to significant improvements in the final results are highlighted.

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 lev, DOE NNSA Stewardship Science Graduate Fellowship (No. DE-NA0002135), National Science Foundation Graduate Research Fellowship (Grant No. DGE1122374)

The DIONISOS experiment is used to study the impact of RF helium (He) plasma on the surface morphology of tungsten (W) at a frequency of 13.56 MHz. Helium ion energy distributions with a span of 70–75 eV, while still below the sputtering threshold result in nano-tendril bundles (NTBs) and free-standing W whiskers on surfaces at 1020 K. The NTBs are distributed intragranularly with coverage of less than 10% while reaching up to 30 µm normal to the surface for He ion fluence of 7.6 × 10²⁵m⁻² and flux density of 10²²m⁻²s⁻¹. Analysis of the NTB interior and sub-surface structure is provided through focused ion beam cross section. Keywords: Tungsten fuzz; Helium; Nano-tendril; RF sheath, United States. Department of Energy (Award DE-SC00-02060), United States. Department of Energy (Grant DE-FC02-99ER54512)

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., United States. Department of Energy (Contract DE-AC52-07NA27344)

New experiments on Alcator C-Mod reveal that the favorable impurity screening characteristics of the high-field side (HFS) scrape-off layer (SOL), previously reported for single null geometries, is retained in double null configurations, despite the formation of an extremely thin SOL. In balanced double-null, nitrogen injected locally into the HFS SOL is better screened by a factor of 2.5 compared to the same injection into the low field side (LFS) SOL. This result is insensitive to plasma current and Greenwald fraction. Nitrogen injected into the HFS SOL is not as well screened (only a factor of 1.5 improvement over LFS) in unbalanced double-null discharges, when the primary divertor is in the direction of B×∇B. In this configuration, impurity ‘plume’ emission patterns indicate that an opposing E × B drift competes with the parallel impurity flow to the divertor. In balanced double-null plasmas, the dispersal pattern exhibits a dominant E × B motion. Unbalanced discharges with the primary divertor opposite the direction of B×∇B exhibit excellent HFS screening characteristics – a factor of 5 enhancement compared to LFS. These data support the idea that future tokamaks should locate all RF actuators and close-fitting wall structures on the HFS and employ near-double-null magnetic topologies, both to precisely control plasma conditions at the antenna/plasma interface and to maximally mitigate the impact of local impurity sources arising from plasma-material interactions. Keywords: Alcator C-Mod; Impurity screening; Double null; High field side scrape-off layer, United States. Department of Energy (Contract DE-FC02-99ER54512)