TGYRO has been updated to compute the total radiation loss, including impurity line radiation, dynamically. The ability to dynamically update the impurity radiation is important for reactor modeling in order to obtain more precise estimates of fusion performance. The presently implemented formulae are the (modified) Post 1977 fits to the cooling rate Lz used in ONETWO. These routines have been extracted and made fully self-contained and portable. Future improvements to include ADAS-based or other tabulated data should now be straightforward to implement and check.

Understanding the plasma response to the applied resonant magnetic perturbations (RMPs) is important for controlling the edge localized modes (ELMs) in tokamak H-mode plasmas. Linear and quasi-linear plasma response modelling offers key insights into (i) the ELM control coil current optimization, and (ii) the side effects of RMPs on the plasma transport. Such an investigation has recently been carried out for ITER, where 7 different plasma scenarios are considered covering both the pre-fusion power and the fusion power operation phases. Utilized in the modeling are the toroidal MARS-F/Q codes which have been validated against DIII-D and other tokamak experiments. Optimal coil current configurations are identified for ELM control in ITER for all these 7 scenarios, based on the linear plasma response computed by MARS-F. Utilizing the initial value quasi-linear code MARS-Q, it is found that the applied RMP fields have minor effects on the plasma toroidal flow, in particular on the core flow, in all 7 scenarios. The RMP effect on the plasma density is generally weak too in these ITER plasmas, except the 7.5 MA/2.65 T and the 7.5 MA/5.3 T scenarios, where large effect on the plasma density is found in the modeling assuming the RMP coil current configurations that maximize the ELM control. This work, which also resulted in major contributions to a journal paper (https://doi.org/10.1103/PhysRevLett.125.155001) as reported in the recent ITER newlines (https://www.iter.org/newsline/-/3502?utm_campaign=whatsnew_weekly&utm_medium=email&utm_source=20%20Oct%202020&utm_content=featured), can help guide the ELM control in the future ITER operations.

The coding style of the OMFIT project has been formatted according to the `black` Python standard, using the 140 character limit. The update involved formatting over half a million lines of code split over 3700 files. OMFIT is a community effort, with over 100 contributors from over 30 institutions. Standardizing OMFIT's coding style is beneficial for improving the readability of the source code across components developed by different contributors. The OMFIT team is confident that this standardization will help ease code review and lower the barrier for new developers to contribute to the project.

A novel flux-surface parameterization has been developed that is suitable for local MHD equilibrium calculations with strongly-shaped flux surfaces. The method is based on a systematic expansion in a small number of intuitive shape parameters, and reduces to the well-known Miller D-shaped parameterization in the limit where some of the coefficients are set to zero. The new parameterization is valid for up-down asymmetric plasmas and provides an improvement to the Miller form. Simultaneously, the method is rapidly convergent and requires only about half the number of shape parameters as a general Fourier representation in the pedestal. The new approach has been implemented in GACODE for use with NEO, CGYRO, and TGYRO. It allows one to carry out edge neoclassical and turbulence simulations with the accuracy of a brute-force Fourier parameterization, but with the convenience of an explicit and intuitive Miller-type model.

High energy, relativistic, runaway electrons (REs) formed during plasma disruption have been shown to be capable of locally melting the metallic wall in present day tokamak devices such as JET. The situation can be even more severe in ITER if REs are not efficiently mitigated or avoided. In a study supported by internal funding, a passive helical coil, which produce 3D magnetic field perturbations due to inductively generated currents during the plasma disruption is being designed to dissipate the RE current when a large and fast change in the magnetic flux occurs. Such passive coils may be particularly useful for reducing RE seeding in future large devices. Utilizing the recently developed MARS-F REORBIT module, REs loss modeling was performed for a DIII-D discharge, where the geometry of helical coils was optimized to produce the best spectrum for the applied 3D vacuum field. Assuming a 100 kA-turn helical coil current, up to 70% of REs are found to be lost to the DIII-D limiting surface within less than 1 msec. The loss fraction is found to be sensitive to the plasma equilibrium current, the assumed coil current as well as the coil geometry. With the same coil geometry and current, modeling shows a higher loss fraction at increased plasma current. The coil geometry that maximizes the magnetic field line stochasticity produces stronger RE losses during the initial stage of the simulation, but does not necessarily result in the largest loss fraction in steady state. For the modeled cases, the majority of the lost REs hit the high field side of the limiting surface. This study helps to quantify the RE mitigation efficiency by 3D passive helical coils. Similar coil design strategies and RE loss modeling can be applied for future large devices.