Abstract Achieving tritium self-sufficiency is a critical challenge for future fusion power plants.
The BABY 1L experiment, part of the LIBRA project at MIT, aims to benchmark tritium breeding and release in molten salt breeder systems under deuterium-tritium (DT) neutron irradiation.
Building on the initial \SI{100}{mL} campaign, BABY 1L introduces a tenfold increase in breeder volume, improved thermal and gas handling systems, and enhanced neutron diagnostics, including a proton recoil telescope.
We report on results from four irradiation experiments using sealed-tube DT neutron generators, with tritium collected by water bubblers measured via liquid scintillation counting.
Experimentally determined Tritium Breeding Ratios (TBRs) were compared to OpenMC neutronics simulations, showing very good agreement.
The measured TBR values demonstrate a six-fold improvement over the \SI{100}{mL} experiments, largely attributed to the increased solid angle and improved measurement fidelity.
We also investigate tritium release dynamics and identify diffusion-limited transport as the dominant regime in the salt volume in the temperature range 630-750 \si{\celsius}.
Additionally, we observe that the introduction of hydrogen in the helium carrier gas significantly accelerates tritium release, consistent with an isotopic exchange mechanism.
All analysis is conducted through the open-source \texttt{libra-toolbox} \cite{libra-toolbox}, which streamlines simulation, data processing, and validation across experimental campaigns.
These results provide critical insights into the design and operation of future liquid breeder systems and demonstrate the maturity of the BABY platform as a testbed for tritium breeding studies.

Abstract Tungsten has demonstrated a competitive figure of merit in its application to plasma-facing components (PFCs) of fusion reactors. During service, the material is exposed to high temperatures and high-level displacement damage. A common interest is fostered in the nuclear materials community to address the issue of defect evolution at operating temperatures, and how they recover throughout service. During maintenance, the application of in situ thermal repair technologies is tempting, featuring attractive efficiency in defect removal via an optimal selection of post-irradiation annealing (PIA) parameters. In previous studies, we examined the role of PIA temperature, PIA duration, and initial defect concentration on defect evolution, and redefined the damage recovery stages for tungsten, but this was done from a room-temperature heavy-ion irradiation perspective; see Wang et al (2023 J. Nucl. Mater. 581 154454), Wang et al (2024 Acta Mater. 273 119942). In this study, the scope is expanded to displacement damage saturation induced by heavy-ions at high temperatures, relevant to the service conditions of tungsten-based PFCs. The damage microstructure evolution in response to varied irradiation temperatures ( T Irr ) and PIA temperatures ( T PIA ) was assessed via transmission electron microscopy and Doppler broadening positron annihilation spectroscopy. Irradiation hardening was evaluated via nano-indentation. A scientific framework is proposed to guide thermal healing of displacement damage in tungsten via PIA treatment. It was ineffective when T PIA ⩽ T Irr . An adverse effect of PIA-induced secondary hardening occurred when T PIA (stage III) > T Irr (stage III). The optimal PIA scheme was confirmed when T PIA (stage IV) > T Irr (stages III–IV), eluding PIA-induced secondary hardening and minimizing PIA-enhanced recrystallization.

Abstract The plasma energy and particle flows in double-null configurations during MAST L-mode discharges are investigated using the 3D plasma turbulence code STORM. The modelling reproduces key phenomena, such as in-out and up-down heat load asymmetries. Turbulent energy transport, driven by ballooning-like instabilities, dominates the radial energy flux across the last closed flux surface, with over 90% entering the scrape-off layer on the low-field side (LFS). In disconnected geometry (the separation between the two X-points in double-null configurations δr sep ≠0), part of that LFS radial flux is transported to the high field side targets via the secondary X-points, causing the in-out power asymmetry to peak in connected geometries (δ r sep = 0). Differences between lower double null (LDN) and upper double-null (UDN) configurations arise due to the upward electron ∇B drift and clock-wise poloidal E × B drift, leading to higher collisionality and stronger turbulence near the separatrix in LDN but a shorter heat flux decay length. Poloidal energy fluxes to different divertors exhibit in-out asymmetries, with more energy flowing to the primary outer divertor in LDN and UDN. Additionally, the clockwise E × B drift in the primary private flux regions (PFRs) redistributes energy between primary inner-outer divertors, reducing the heat load on LDN’s primary outer target while increasing it in UDN. Thus, for the same |δ r sep |, the total heat loads on primary outer targets in LDN and UDN become comparable. These findings provide insights into plasma and energy transport in double-null configurations, with implications for optimizing divertor performance in fusion reactors.

Abstract At the onset of detachment, measurements show that radial turbulent particle transport ( Γ ~ r ) increases by a factor of 2 − 4 above the dissipative region in the near scrape-off layer (SOL) of both single-null (SN) and snowflake (SF) divertors. This leads to a fourfold increase in the density decay length, with broadened density profiles and flattened electron temperature ( T e ) profiles along the divertor leg. These profiles are convected uniformly through the divertor by parallel flows, with parallel heat transport shifting from conduction- to convection-dominated. Within the dissipative divertor, the flattening and collapse of the plasma potential ( V p ) profiles weaken electric field gradients, reducing the E × B drifts. Turbulent spreading upstream of the dissipative region thus emerges as the dominant mechanism for radial transport at the detachment onset. Radial and poloidal drifts are shown to move respectively ∼ 10 %–20% and ∼ 15 %–30% of the target-normal particle and heat flux delivered to the outer strike point in attached conditions, but their contribution drops to a few percent at the detachment onset.

Abstract Minimization of fast-ion losses induced by the 3D effects has been a topic of key importance in the context of controlling edge localized modes (ELMs) using resonant magnetic perturbations (RMPs). In this paper, numerical simulations have been carried out to explore the parameter space where ELM control and fast-ion confinement can be balanced. The ELM control capability, as represented by edge magnetic stochasticity, exhibits different dependencies on the RMP spectrum compared to fast-ion losses, thereby creating an optimization window. Selecting the optimal spectrum can reduce the fast-ion loss by a factor of two. Parameter scanning results reveal that the fast-ion loss is primarily attributed to global kink-like magnetic perturbations, which can be amplified by increased plasma pressure. In contrast, the edge magnetic stochasticity is strongly associated with the layer physics and influenced by plasma resistivity and rotation. This fundamental difference provides a basis for addressing the compatibility issue between the control effect and the minimization of fast-ion losses. The spectral dependencies of magnetic stochasticity and fast-ion losses respectively depict strong and negligible dependence on q 95 within a narrow range, whereas on a larger scale, they exhibit similar linear variations. Applying high- n RMP represents a more effective strategy than low- n RMP in minimizing fast-ion loss. The results presented here may provide insights into the optimization of RMP configurations and plasma parameters with respect to the NBI confinement in future studies.

Abstract Gyrotrons having a high output-frequency of more than 200 GHz will be necessary for future fusion devices, such as the DEMO reactor. At 170 GHz, gyrotrons for ITER achieved a high power of 1 MW with a high electrical efficiency of approximately 50% (oscillation efficiency of approximately 30%) in a long pulse operation. At frequencies higher than 200 GHz, no gyrotrons have achieved ITER-like performance, which this study attempted to demonstrate using a prototype multi-frequency gyrotron for 170 GHz, 203 GHz, and 236 GHz. At 203 GHz, an output power of 0.87 MW was achieved at the gyrotron window with an oscillation efficiency of 25% and an electrical efficiency of 41% in a long pulse operation of 100 s. Although this performance is lower than that of the ITER gyrotrons at 170 GHz, this study represents the first demonstration that ITER-like performance is achievable at frequencies above 200 GHz. In addition, short 1-ms pulse tests at 236 GHz achieved an output power of 1 MW.

Abstract Guiding-centre (GC) and full-orbit (FO) simulations of the beam-injected fast ion distribution and the corresponding neutron emissivity have been carried out for magnetohydrodynamics-quiescent NSTX plasmas, using a combination of ASCOT5 and DRESS, to assess the suitability of the GC approximation for fast ions in NSTX. It was found that GC and FO simulations predicted substantially different steady-state distributions in both position and velocity space and different neutron emissivity profiles, leading to a 15% reduction in the predicted global neutron rate for FO relative to GC. These changes accompany a higher magnetic moment in FO, and correspond to a change in particle orbits from co-passing to trapped and stagnation orbits. ASCOT5 was also benchmarked against TRANSP/NUBEAM with input loaded entirely from TRANSP/NUBEAM output files, with agreement found between the GC simulations when finite Larmor radius (FLR) corrections were omitted. ASCOT5 FO and TRANSP/NUBEAM with FLR produced fast ion distributions which differed in localised regions, but predicted global neutron rates which agree within 3%.

Abstract Mechanisms of destabilization of toroidal Alfvén eigenmodes (TAEs) in tokamaks are analyzed with the aim to reveal those leading to modes propagating in the direction opposite to plasma current, i.e. counter-propagating modes (ctr-TAE). Plasmas with fast-ions sources [such as neutral beam injection (NBI), ion cyclotron resonance heating, fusion reactions] and without them (Ohmic discharges) are considered. A particular NSTX-U experiment with NBI, where co- and counter-propagating TAEs were observed simultaneously (Podestà et al 2018 Nucl. Fusion 58 082023), is considered. It is concluded that both types of TAEs occurred because their destabilization was caused by the velocity anisotropy of beam ions, which overrode effects of spatial inhomogeneity of these ions.

Abstract During Phase I of WEST operation, one of the lower divertor sectors, known as the test sector, was partially equipped with actively cooled ITER-like plasma-facing units (PFUs) made of 35 monoblocks (MBs). A visual inspection of the test sector after the C4 campaign revealed changes on the surface condition of several MBs located in the outer strike area. In this work, one of the MBs of interest, MB 28 from PFU#13, was extensively investigated focusing on fuel retention, material migration (erosion, transport, redeposition), and surface morphology. The plasma exposed surface of MB 28 exhibited four distinct zones along the toroidal direction. In the first zone, extending up to 4 mm from the leading edge (LE), the surface showed signs of erosion, with periodically spaced cracks oriented perpendicular to the LE. The second zone exhibited the highest material deposition (B, N, Fe and Cu) and fuel (D and He) retention, with the maximum deposit thickness reaching approximately 1.5 μ m. In the third zone, the deposition gradually decreased and accumulated in linear areas. The fourth zone, near the trailing edge, was marked by the maximum hydrogen ( H ) and carbon ( C ) deposition. In this zone, linear deposition areas rich in O and B were detected, running parallel to the edge. A clear correlation was observed between the deposition of B, N, Fe and Cu and the retention of D and He.