ANNEX 1

IEA Technology Collaboration Programme on Tokamak Programmes (CTP TCP)

Annual Briefing 2025

1. Preface and Introduction

The objective of the CTP TCP is to advance the physics and technologies related to toroidal plasmas. This is achieved by strengthening cooperation among tokamak programmes, enhancing the effectiveness and productivity of the research and development (R&D) effort related to the tokamak fusion concept, contributing to and extending the scientific and technology database of toroidal confinement concepts, and providing a scientific and technological basis for the successful development of fusion power. The main accomplishments are integrated in the International Tokamak Physics Activities, related topical group meetings, workshops and joint experiments. The CTP TCP contributes to the development of ITER physics R&D via joint experiments, development of modelling codes, assessment of key subjects and the development of real-time control schemes. The focus is on minimising the risk and maximising the capability of ITER focussing on high performance with the full Tungsten wall, disruption mitigation, Edge Localised Modes mitigation, detached divertor control and burn control.

2. Chair’s report 2025

Membership

At present the CTP TCP has 10 Contracting Parties:

• 6 IEA member countries:

• Australian Nuclear Science and Technology Organisation (ANSTO), Australia

• National Institutes for Quantum and Radiological Science and Technology (QST), Japan

• Korean Ministry of Education, Science and Technology (MEST), Korea

• United States Department of Energy (USDOE), United States

• Ecole polytechnique fédérale de Lausanne (EPFL), Switzerland

• Government of the United Kingdom, United Kingdom

• 1 partner country:

• The Institute for Plasma Research (IPR), India

• 3 International Organizations:

• European Atomic Energy Community (Euratom), European Union

• ITER China Domestic Agency (CNDA)

• ITER Organization

The CTP TCP Executive Committee took note of the proposal to invite Kazakhstan to become a member of the CTP TCP and formal approval will be carried out via written procedure.

At present the CTP TCP has 1 Limited Sponsor

· Commonwealth Fusion Systems based in the United States (US).

Meetings

· 16^th Executive Committee Meeting at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, Friday 5^th December 2025

· 28^th meeting of the International Tokamak Physics Activity Coordinating Committee (ITPA CC)

and 16^th ITPA meeting for Joint Experiments at the ITER Organisation Headquarters, St. Paul-lez-Durance, France, 3^th-5^th December 2025

Participation of the private sector

The CTP TCP Executive Committee, via a Written Procedure that ended on 15^th January 2025, unanimously approved the proposal and process for the use of the CTP TCP Sponsor or Limited Sponsor scheme as a legal framework for the participation of privately funded companies in the ITPA Activities.

The Executive Committee of the CTP TCP, via a Written Procedure that ended on 20^th April 2025 approved the proposal to amend the CTP TCP Implementing Agreement as to include the provisions related to Limited Sponsors.

The Executive Committee of the CTP TCP, via a Written Procedure that ended on 14^th February 2025, unanimously approved the proposal to invite Commonwealth Fusion Systems based in the United States (US) to Join the CTP TCP as a Limited Sponsor. The participation of Commonwealth Fusion Systems based in the United States (US) as a Limited Sponsor to the CTP TCP became effective as of 19^th April 2025.

At the 16^th CTP TCP Executive Committee meeting, the CTP TCP Executive Committee unanimously approved the proposal for the Chinese company ENN Science and Technology Development Co., Ltd. (ENN) to become a Limited Sponsor to the CTP TCP.

CICLOP: A Joint Task of the Co-operation on Tokamak Programmes (CTP) and Stellarators and Heliotrons (SH) Technology Collaboration Programmes (TCPs)

The CICLOP Joint Task between the Co-operation on Tokamak Programmes (CTP) and Stellarators and Heliotrons (SH) Technology Collaboration Programmes (TCPs) was established in 2024. Significant progress has been made in tokamaks and stellarators including very recent achievements in the duration and performance, supported by superconducting coils, actively cooled components, and various types of metallic walls. These achievements represent a crucial step towards bridging the gap between present experiments and ITER, and they provide a solid foundation for the development of steady-state scenarios for DEMO and for future fusion pilot plants.

Status report and milestones achieved

ITER

The ITER project has continued making steady progress towards first plasma and the respective sector module sub assembly have started with Sector Module 7 installed in the Tokamak Pit in April 2025, Sector Module 6 installed in the Tokamak Pit in June 2025 and Sector Module 5 installed in the Tokamak Pit in November 2025. Regarding the Central Solenoid Assembly, the 4^th module was stacked in January 2025, the 5^th module stacked in November 2025 and the 6^th module arrived in the Fall 2025, scheduled to be stacked in March 2026.

Europe
Joint European Torus (JET), Axisymmetric Divertor Experiment (AUG) Upgrade, Tokamak à Configuration Variable (TCV), Mega-Amp Spherical Tori (MAST) Upgrade, and Tungsten (W) Environment in Steady-state Tokamak (WEST)

Joint European Torus (JET)

Resources for tokamak exploitation is still being dedicated to the data validation, analysis and interpretative modelling of past JET campaigns. Among the recent analysis was the establishment of an edge turbulence control parameter that unifies in JET the radial density decay of the JET plasma far-Scrape of Layer.

Axisymmetric Divertor Experiment (AUG) Upgrade

Alternative Divertor studies in the new ASDEX Upgrade upper divertor showed no hot spots with both helicities and very good alignment of the Divertor tiles.

Tungsten (W) Environment in Steady-state Tokamak (WEST)

WEST have demonstrated long-pulse operation up to 1337 seconds with a full ITER-grade tungsten divertor, reaching up to 2.6 GJ of injected energy under non-inductive conditions. 73 seconds of plasma operation with a continuous X-Point Radiator was achieved with a peak heat flux reduced by 4 times when compared with single-null divertor configurations.

Tokamak à Configuration Variable (TCV)

Radiation localized around the secondary X-point of the X-Point Target divertor was identified, and operation was achieved with a 6 times reduction in the peak target heat flux and a strong increase in detachment front stability when compared with the single-null divertor configuration.

Japan

JT60-SA

Scientific interest on JT-60SA is increasing with the major goals in 2025 achieved with the operation scenario modelling in preparation of the operation phases Op2 and Op3 and for the transition to Tungsten plasma facing components.

China

Experimental Advanced Superconducting Tokamak (EAST)

The EAST operating window was extended with an ITER-like metal wall with repeatable long-pulse operation with an H-mode edge up to 1066 seconds in fully non-inductive conditions utilizing a water-cooled tungsten lower divertor. EAST achieved double transport barrier operation with a strong ion internal transport barrier with core reversed magnetic shear and an electron transport barrier. A strategy was developed for the Hydrogen and Deuterium density ratio control using Supersonic Molecular Beam Injection fuelling.

HL-3 Tokamak

Negative triangularity configurations were explored for the first time on the HL-3 tokamak. Reduction of ambient turbulence via microtearing mode and energetic-particle-induced geodesic acoustic mode led to improvements in energy and particle confinement on HL-3.

Korea

Korean Superconducting Tokamak Advanced Research (KSTAR)

High-Z impurity transport and accumulation in ITER-relevant conditions were demonstrated on KSTAR using Krypton injection. Krypton injection produced highly radiative plasmas
with small ELMs with approximately 300Hz repetition rate and Neon injection decreased the total radiative power loss and the centrally peaked radiation moved off axis.

United States

Doublet III D-shaped Tokamak (DIII-D)

DIII-D expanded the boundaries of solutions for future fusion devices with increase in shaping and volume leading to record pedestal parameters, high beta poloidal record performance with triangularity increased and with a Greenwall fraction of 1.4, H confinement factor of 1.5 and small edge localised modes in near detached conditions.

National Spherical Torus Experiment Upgrade (NSTX-U)

The NSTX-U recovery project is mostly complete with the only remaining item being the toroidal field and the ohmic heating coil bundle remaining to be installed. NSTX-U operation is expected in December 2026.

India

ADITYA-Upgrade

ADITYA-Upgrade operated with reversed current, zero loop voltage using lower hybrid current drive. The sawtooth duration was modified using gas puffing and the effect of convective transport in the edge region was studied.