T-15MD

Overview

The T-15MD is a tokamak research facility operated by the National Research Center "Kurchatov Institute" in Russia. It represents a significant modernization of the Soviet-era T-15 tokamak, repurposed to address key physics and engineering challenges for next-generation fusion devices like ITER and DEMO. Its primary research mission is focused on plasma-material interactions, particularly the development and testing of advanced divertor concepts and plasma-facing components (PFCs) under high heat flux conditions. The device is designed to handle stationary heat loads on the divertor targets of up to 20 MW/m².

A defining characteristic of T-15MD is its hybrid magnet system. It uniquely combines a water-cooled copper toroidal field (TF) coil with the large, cryogenically cooled niobium-tin (Nb₃Sn) superconducting poloidal field (PF) coils salvaged from the original T-15 machine. This configuration provides a flexible platform for studying long-pulse, high-performance plasma scenarios without the cost of a fully superconducting TF system. The facility's research program is central to Russia's national fusion strategy and serves as a domestic platform for training personnel and testing technologies relevant to its contributions to the international ITER project.

Physics / Mechanism

The T-15MD operates on the magnetic confinement principle of a tokamak, using a combination of toroidal and poloidal magnetic fields to confine a high-temperature deuterium plasma in a toroidal (doughnut-shaped) vacuum vessel. The device has a conventional aspect ratio (R/a ≈ 2.2), placing it between smaller spherical tokamaks and larger, lower-aspect-ratio machines like JET or ITER.

Key design parameters include a major radius (R₀) of 1.48 m, a minor radius (a) of 0.67 m, and a toroidal magnetic field (B₀) on-axis of up to 2.0 T. It is designed to sustain a plasma current (Iₚ) of up to 2 MA for pulse durations of several seconds. The plasma shape is D-shaped with an elongation (κ) of up to 1.8, enabling advanced confinement regimes such as the H-mode.

The auxiliary heating and current drive systems are central to its mission. The planned system is a multi-megawatt combination of:

• Electron Cyclotron Resonance Heating (ECRH): A system of gyrotrons provides microwave power (initially ~4 MW, planned up to 9 MW) to heat electrons and drive non-inductive current. This is crucial for plasma startup, heating, and MHD stability control.
• Ion Cyclotron Resonance Heating (ICRH): An antenna system delivers radio-frequency waves to heat plasma ions, with a planned power of up to 6 MW.
• Neutral Beam Injection (NBI): A high-energy neutral particle beam is planned to provide the bulk of the heating power, up to 15 MW, and drive significant plasma current.

The divertor is a key area of innovation. T-15MD is designed to test multiple divertor configurations, including a conventional single-null and advanced concepts aimed at spreading the heat load and facilitating plasma exhaust. The divertor targets are designed for extreme thermal loads, a critical research area for the viability of a future fusion power plant.

Historical development

The history of T-15MD is rooted in the Soviet fusion program. The original T-15 tokamak, which began operation in 1988, was a pioneering device and one of the world's first large tokamaks to utilize superconducting magnets for its toroidal field coils. It achieved a toroidal field of 3.6 T and a plasma current of 1 MA. However, its operation was curtailed in the mid-1990s due to economic difficulties following the dissolution of the Soviet Union.

For nearly two decades, the device remained dormant. In the early 2010s, a decision was made not to recommission the aging T-15 but to execute a comprehensive rebuild and modernization project, creating a new machine designated T-15MD. The "MD" signifies "Modernized Divertor," highlighting the new research focus. The project was led by the Kurchatov Institute and became a cornerstone of Russia's updated federal fusion program.

The modernization was extensive. While the large superconducting PF coils and some infrastructure were retained, a new vacuum vessel, a resistive copper TF coil system, and entirely new divertor, diagnostic, and heating systems were designed and installed. This approach allowed for the creation of a modern, capable machine at a fraction of the cost of a completely new facility.

Construction and assembly progressed through the late 2010s. The physical startup of the T-15MD was achieved on May 18, 2021, with the successful generation of its first plasma. This milestone marked the return of a large-scale tokamak facility to operation in Russia and was a significant event for the national and international fusion communities.

Current status

As of early 2026, T-15MD is in its operational phase, conducting experimental campaigns and undergoing staged upgrades to its auxiliary systems. The initial years of operation (2021-2025) focused on commissioning the device, characterizing ohmic plasmas, and bringing the first ECRH heating systems online. These experiments successfully demonstrated stable plasma control and achieved initial target parameters.

The current research program is expanding to include higher-power operations as more heating systems are commissioned. The focus is on divertor physics, including heat load management, detachment studies, and testing of novel PFC materials. Experiments are being conducted to explore the synergy between the different heating methods (ECRH, ICRH) and to develop stable, long-pulse H-mode scenarios.

According to a 2023 IAEA report, the facility has demonstrated reliable operation and is systematically increasing plasma current, density, and pulse duration. The diagnostic set is continuously being enhanced to provide high-resolution data on plasma behavior, particularly in the critical edge and divertor regions. T-15MD serves as a key user facility for Russian universities and research institutions, contributing to the development of the next generation of fusion scientists and engineers.

Notable implementations

The T-15MD is a unique national facility operated by the NRC Kurchatov Institute. It is not a commercial endeavor but a government-funded research platform. Its implementation is central to several strategic goals:

1. Russian National Fusion Program: T-15MD is the flagship experimental device of the Russian domestic fusion program. It provides the primary experimental data to support the design of a Russian DEMO-type reactor, known as DEMO-FNS (Fusion Neutron Source).
2. ITER Support: The research conducted on T-15MD directly supports Russia's in-kind contributions to the ITER project. This includes testing of diagnostics, developing plasma control techniques, and investigating plasma-wall interaction phenomena relevant to ITER's operational plan.
3. Materials Science: The facility is a critical testbed for advanced materials developed by Russian institutions, including tungsten-based alloys and composite materials for PFCs. The ability to expose material samples to reactor-relevant heat and particle fluxes is a key capability.
4. International Collaboration: While primarily a national facility, T-15MD is open to international collaboration. It provides a platform for joint experiments and data exchange with fusion laboratories in Europe, Asia, and other regions, often through frameworks established by the IAEA.

Open challenges

Despite its successful startup, T-15MD faces several scientific and engineering challenges that form the core of its research plan:

• High Heat Flux Management: A primary challenge is demonstrating the survivability of divertor components under sustained heat loads of 10-20 MW/m². This involves not only developing robust materials but also perfecting plasma control techniques to mitigate or spread the heat load, such as inducing partial or full divertor detachment.
• Integration of Heating Systems: Achieving the full 30 MW of planned heating power requires the successful integration and synergistic operation of three complex systems (NBI, ECRH, ICRH). Managing the interplay between these systems to control the plasma profile and stability is a significant physics and engineering task.
• Tritium and Neutron Environment: While T-15MD primarily operates with deuterium, future experiments may involve limited deuterium-tritium (D-T) campaigns. This would introduce the challenges of handling tritium fuel and managing the resulting neutron activation of the machine components, requiring remote handling capabilities and a robust safety case.
• Disruption Mitigation: Like all large tokamaks, T-15MD must develop reliable strategies for mitigating plasma disruptions. These events can deposit enormous thermal and electromagnetic loads on the vessel walls, and developing effective mitigation systems, such as massive gas injection, is a critical area of research for protecting the machine and future reactors.

Outlook

The credible 5- to 15-year trajectory for T-15MD involves a phased approach to reaching its full design parameters and research goals. In the near term (2026-2030), the facility will focus on completing the installation and commissioning of its full suite of auxiliary heating systems. This will enable access to high-performance, long-pulse H-mode plasmas, allowing for systematic studies of divertor physics and plasma-wall interactions at heat fluxes approaching DEMO-relevant levels.

In the medium term (2030-2035), T-15MD is expected to become a mature user facility, producing key data for the design of next-generation devices. Research will likely shift towards more integrated challenges, such as steady-state scenario development, control of tungsten impurity transport, and testing of advanced divertor concepts like the snowflake or super-X divertor. Limited experiments with tritium could be conducted in this phase to study alpha particle physics and test components in a neutronic environment.

By 2040, T-15MD will have provided a substantial database for the design of Russia's DEMO-FNS and contributed significantly to resolving key operational issues for ITER. Its role may evolve to become a dedicated test stand for novel components, diagnostics, and materials, ensuring its continued relevance in the global fusion research landscape as larger, more powerful machines come online.

References

1. First plasma in the T-15MD tokamak — Plasma Physics and Controlled Fusion (2022)
2. T-15MD spherical tokamak for studies of plasma–wall interaction — Nuclear Fusion (2015)
3. Status of the T-15MD Project — IEEE Transactions on Plasma Science (2018)
4. Russia's T-15MD tokamak achieves first plasma — ITER Organization Newsline (2021)
5. Overview of the T-15MD experimental program — Journal of Physics: Conference Series (2023)
6. Fusion Devices — International Atomic Energy Agency (IAEA)
7. The electron cyclotron resonance heating system for the T-15MD tokamak: Current status and plans — Fusion Engineering and Design (2019)
8. The Kurchatov institute project T-15MD: A modernized tokamak to support ITER — Problems of Atomic Science and Technology (2012)