Introduction of Fusion Research Activities at SWIP China Fusion Energy Co., Ltd. Southwestern Institute of Physics ZHONG Wulyu 2026.7.20 CONTENTS 01 Introduction of CFEC & SWIP 02 HL-3 Tokamak 03 ITER-Related Projects 04 Fusion Technologies 05 Prospects of Fusion Research 01 Introduction of CFEC & SWIP China National Nuclear Corporation (CNNC) - Founded in 1955, former Ministry of Nuclear Industry - Centrally managed state-owned enterprise approved by the State Council - Key player in the national nuclear science and technology industry - The only comprehensive nuclear industry enterprise among the Fortune Global 500 - Mission: Promoting the development and utilization of nuclear energy, nuclear engineering construction, and the application of nuclear technology. - Total assets has reached approximately 1.7 trillion RMB Yuan. - > 170,000 employees. Southwestern Institute of Physics (SWIP) - Affiliated with China Fusion Energy Co., Ltd (CFEC), CNNC - Earliest institute for fusion energy research in China, founded in 1965 with staff > 2000 - SWIP has developed more than 20 fusion research devices of various magnetic configurations Timeline of devices: - 1965 - 1969: Θ pinch-3 - 1971: LingYun stellarator - 1972: Θ pinch-4, MM-2 - 1975: MM-1 magnetic mirror - 1977: PT tokamak - 1978: MM-3, RF, WeiHuan-1 tokamak - 1979 - 1981: CF-1, MM-5, FY-1 tokamak - 1982: MM-4 - 1984: CF-2, HL-1 tokamak - 1985: WeiHuan-2 - 1987: MM-4U - 1989: MM-2U, RFP (1990) - 1994: HL-1M - 1999 - 2002: HL-2A - 2009 - 2020: HL-3 China Fusion Energy Co., Ltd (CFEC) Based on 60-year fusion energy R&D of SWIP China Fusion Energy Co., Ltd. (CFEC) established in 2025, affiliated to CNNC CFEC established in Shanghai in July 2025 Shareholders include: - China National Nuclear Corporation - China National Petroleum Corporation - National Green Development Fund - Shanghai Future Fusion Energy Technology Co., Ltd - Zheneng Electric Power Co., Ltd - Chengdu Industry Co., Ltd R&D layout with 3 Campuses in 2 Cities Chengdu, Sichuan Province: - Shuangliu District (Baijia): Frontier scientific experiment (HL-3) - Tianfu New Area (Yongxing): Key engineering technology R&D for fusion reactor Shanghai: - Minhang District (Maqiao): • Fusion reactor design, high-temperature superconducting (HTS) magnet, digital tokamak • High-field tokamak with HTS magnet (HL-4) • Construction started in Dec. 2025 and will be completed in 2030 R&D Centers: 1. Fusion Engineering Design Center: - Integrated design of fusion reactors - High temperature superconducting (HTS) magnets R&D - Digital tokamak 2. Fusion Technology R&D Center: - First Wall R&D - Power supplies for special applications - Fuel cycling system - Heating and Current Drive R&D 3. Fusion Scientific Experiment Center: - HL-3 Tokamak - Frontier Scientific Experiments 4. Plasma Technology Application Center: - Plasma Environmental Governance Technology - Plasma Surface Engineering Technology Fusion Scientific Experiment Center (Baijia Campus, Chengdu) (~ 5 hectares) HL-3 Nuclear Fusion Technology R&D Center (Yongxing Campus, Chengdu) (~ 33 hectares) - HQ - H&CD - FW, TBM, Materials Phase One (2022-2026) Interior of Zhengwu Building (HQ) Maqiao Campus in Shanghai For fusion reactor design, HTS magnet, digital tokamak, HL-4… (about 13 hectares) HL-4 02 HL-3 Tokamak Tokamaks developed by SWIP Evolution of Size, Parameters, and n Ti τE: - HL-1 (1984): Radius 1.02 / 0.20 m. First large scientific and engineering fusion device in China. - HL-1M (1994): Radius 1.02 / 0.24 m, Plasma current >300 kA. Plasma fueling by SMBI proposed for the first time. - HL-2A (2002): Radius 1.65 / 0.4 m, Plasma current 0.45 MA. First diverted tokamak in China. Important milestones achieved in fusion field. - HL-3 (2020): Radius 1.78 / 0.65 m, Plasma current 2.5 (3) MA. Largest tokamak in China with highest parameter and advanced divertor. Mission of HL-3 tokamak - CN largest and highest parameter Tokamak to High Performance Plasma Operation > High-performance scenarios compatible with low heat load > Deep understanding of frontier plasma physics > Tests and validation of high heat flux plasma-facing components HL-3 Main Parameters: - Major radius R = 1.78 m - Minor radius a = 0.65 m - Aspect ratio R/a = 2.8 - Plasma current Ip = 2.5 (3) MA - Toroidal field Bt = 2.2 (3) T - Elongation κ = 1.8-2 - Triangularity δ > 0.5 - Heating power: NBI (20) + ECRH (11) + LHCD (4) + ICRH (6) MW Key features Comprehensive operation scenarios: - Baseline Operation: Ip = 2.5 MA, q95 = 3, H98,y2 = 1.0 - Hybrid Operation: Ip = 1.0~1.4 MA, fBS ~40%, fni ~70%, βN > 2.5 - Steady-State Operation: Ip = 1.0-1.2 MA, fully non-inductive current drive, βN > 3.0 Key features - Comprehensive operation scenarios - Flexible magnetic configurations: • Single-null • Double-null • Snow flakes (SF) • SF- • Tripod • Negative triangularity Key features - Comprehensive operation scenarios - Flexible magnetic configurations - High accessibility for experiments: • 128 ports: diagnostics, fueling, heating and current drive systems • Flexible capacity on heat and particle exhaust: actively cooling divertor and strong pumping speed cryopump (38 m³/s) First plasma on Dec. 4th, 2020 - First plasma: >100 kA, >100 ms - Reported by CCTV News on Dec. 4th, 2020 - Selected as: "Top 10 News of Scientific and Technological Progress in China in 2020" Staged research plan of HL-3 I. First plasma (2020): - Ip = 100 kA - Limiter/divertor configuration II. Initial research (~2023): - Ip ≥ 1 MA - H mode discharge - Advanced divertor control III. Integrated research (2025): - Ip ≥ 2 MA - Ti ~ 10 keV - Advanced scenarios - Heat/particle control - Configuration control - AI for fusion IV. Extended research (2030): - Triple product > 10²⁰ m⁻³·keV·s - Fusion reactor grade plasma operation and physics research - Fusion reactor related integration scheme verification - Fusion reactor related critical components/diagnostic tests Heating and current drive systems Plasma heating, current drive, plasma instability control, and plasma start-up assist - Neutral Beam Injection (NBI): 20 MW / 120 keV / 5s (3 systems: 1# in 2023, 2# in 2025, 3# in 2026; 6 MW + 7MW + 7 MW, 120 keV / 40A / 5s) - Electron Cyclotron Resonance Heating (ECRH): 11 MW / 105 GHz, 140 GHz, and 175 GHz / 10s (EL: -20~15° NTM control, UL: for CTS) - Ion Cyclotron Resonance Heating (ICRH): 6 MW / 25–50 MHz / 5s (4×1.5MW) - Lower Hybrid Current Drive (LHCD): 4 MW / 3.7 GHz / 5s (TE01 circle WG technology) Highlights on H&CD design & experiment - ICRF system for the HL-3 tokamak, L. F. Lu, Nucl. Fusion, 10.1088 (2026) - Self-consistent modelling of radio frequency sheath in 3D with realistic ICRF antennas, L. F. Lu, Nucl. Fusion, 126013 (2024) - Development and operational progress of ion source for the second neutral beam injector on HL-3 tokamak, H.Y. Luo, Fusion Eng. Des., 115823 (2026) - Development of the calorimetry target on second neutral beam injection beam line of HL-3 device, B.W. Zhou, Nucl. Tech. 050012 (2024) High performance operation aided by NBI - Early NBI injection during the current ramp optimizes the q profile for ITB formation while facilitating H-mode access, enabling integrated high-performance operation. - Early NBI modifies the q profile and promotes strong ion ITB formation. - High auxiliary heating power enables H-mode access and formation of ETB. - Simultaneous achievement of Ti >10 keV, and a triple product of 0.67×10²⁰ m⁻³·s·keV. Reference: W.L. Zhong, The Innovation. 7 (2026) 101167 eITB Control via Optimized Off-axis ECRH - Systematic optimization of EC power and deposition location established a reliable operating window for reproducible eITB formation. - Optimized ECRH coupling significantly improved electron heating efficiency. - Stable and reproducible eITBs were established using off-axis ECRH. - Peak electron temperatures over 10 keV were achieved with a plasma current range of 0.6 - 1.0 MA. NTM feedback control by ECCD - Real-time neoclassical tearing mode (NTM) feedback control by ECCD - Workflow: NTM identified by ECE -> ECCD injected with particular angle -> NTM suppressed - Architecture: ECE/Mirnov -> RT EFIT -> ECRH antenna angle / CODIS control system -> ECCD -> NTM island shrinking/suppression - Optimal EC antenna angle scanned - Real-time NTM identification and EC injection achieved, 3/2 NTM island shrank and finally suppressed when EC deposited near q=1.5 surface - Next step: Real-time angle control to suppress multi-modes. ELM suppression by LHW Injection - LHW suppresses ELMs by enhancing pedestal particle transport and broadening the pedestal, reducing pressure gradient and peeling-ballooning instability. - LHW injection -> enhanced pedestal turbulent transport -> broader pedestal -> decreased pressure gradient -> reduced PB instability -> ELM suppression Reference: Y.M. Zhang, Nucl. Fusion 65 (2025) 096019 Achievements of HL-3 in 2020-2025 - 2020: First Plasma - 2022: 1.15MA plasma - 2023: 1MA H-mode Operation; Advanced Divertor (Snowflake & Tripod) Operation - 2024: 1.5 MA H-mode Operation; H-mode in Tripod Divertor Configuration; High Density Operation (~1.2×nGW) - 2025: Ti = 117 million degrees (>10keV); Te = 160 million degrees (>13keV); Triple product > 0.67×10²⁰ m⁻³·keV·s HL-3 Global Opening - In 2023, HL-3 opened as ITER satellite device to support ITER physics research and operation - In 2024, HL-3 was announced as facility opened to the world for international cooperation Key events: - SWIP and IO signed agreement for deepened cooperation and HL-3 opened as ITER satellite device (2023, IO) - The 17th China International Nuclear Industry Exhibition (Mar., 2024, Beijing) - The 68th IAEA annual general conference side meeting (Sept., 2024, Vienna) HL-3 Joint Experiments - Joint Experiments since 2024 - More than 80 scientists from ITER, Japan, Korea, France, Portugal, US, etc., participated - Focus on high beta plasmas, advanced divertor physics, ETB/ITB regimes, disruption physics, etc. ITPEA Groups in SWIP - Taskforces were set up in accordance with ITPEA Topical Groups - Corresponding Tasks / Synchronized Research / International Collaboration / Joint Experiments - Topical Groups: IOS, MDC, PEP, Diag., DSOL, TC, EP, H&CD, Operation, Magnet, PS and Cryogenics across HL-3 and ITER. 03 ITER-Related Projects ITER Procurement Arrangements SWIP is responsible for 9 PAs of ITER Key Components, including: - First Wall - Shield Block - Divertor Langmuir Probe - Magnet Support - Gas Injection System (GIS) - HCCB TBM-set - Neutron Flux Monitor - Glow Discharge Cleaning Enhanced Heat Flux (EHF) First Wall (FW) Technical development and qualification programme of Enhanced Heat Flux (EHF) First Wall (FW), including small mock-ups, semi-prototype to full-scale prototype (FSP) - Full scale prototype (Beryllium armor) developed, qualification passed in 2023 - Serial Production (57) Contract of W-armour FW signed in Dec. 2025 - Manufacturing Process Qualification approved in June 2026 Milestones: 1st Prototype of FW (beryllium armor, 2023); 1st Prototype of FW (tungsten armour, 2025); Qualification of Tungsten Armour Bonding technology completed (2026). Magnet Support (MS) Manufacturing of ALL ITER Magnet Supports, including 18 sets of Gravity Supports (GS), 108 sets of Poloidal Field Coil Supports (PFCS), and 6 sets of Correction Field Coil Supports (CCS) - ITER’s key structural safety component withstanding 20,000 tons of complex loads - First key components delivered by China, >1600 tons - All products delivered and contract to be closed this year (All Products delivered in Nov. 2023) Shield Block (SB) Technical support to all the 220 SBs - 1st ISO standard in fusion R&D, 'hot helium leak testing method' in 2023 (ISO 4233:2023) - First batch of 48 blocks delivered in 2024 and the rest will be done in 2027 - 2 hot helium leak test facilities developed for test of the blocks Gas Injection System and Glow Discharge Cleaning Gas Injection System (GIS): - Design of the whole ITER GIS system and the manufacture of the ITER GIS except GDS Manifold - First batch of GF GVBs (4 sets) delivered to ITER in May, 2026 Glow Discharge Cleaning (GDC): - All 11 glow discharge cleaning electrodes for ITER - GDC PD Contract passed the acceptance of CNDA in Dec. 2025 DLP and NFM Divertor Langmuir Probe (DLP): - R&D of all 400 sets of Langmuir Probe systems - FDR completed in Dec. 2025 - 100% pass rate of high heat load test Neutron Flux Monitor (NFM): - 4 sets of NFM systems in Equatorial Port 1#, 7#, 8#, and 17# - Double-layer shielded preamplifier developed with excellent performance - NFM7# support Frame installed at ITER in 2020 Helium-Cooled Ceramic Breeder Test Blanket System (HCCB-TBS) One of major engineering goals of ITER is to test and verify the tritium breeding and energy extraction technologies of fusion reactors - Completed the first TBS conceptual & preliminary design - Developed the first semi-prototype module - First TBM Shield Block developed in 2025; HCCB-TBS PDR held in July 2026 Components: Test Blanket System, Helium cooling system, Tritium system, Test blanket module, TBM Shield Block, Semi-Prototype HCCB TBM Module. ITER Assembly SWIP is one of the China Nuclear Power Engineering (CNPE) consortium members undertaking the ITER Assembly Contracts - 2019: TAC1 - 2024: SMSA - 2024: SMPA - 2025: WPP - 2026: EVA Divertor Cassette Assembly Integration Contract Acceptance, inspection, assembly, packaging, transportation and on-site testing of 58 divertors and on-site acceptance of 205 shielding modules - Divertor components received from JADA - IO, F4E, RFDA, and JADA witnessed the unpacking and inspection of Prototype Divertor components in May - Procedures and specialized tooling for the SB Factory Acceptance Test prepared ITER ELM Power Supply Contract Consortium led by SWIP and joint by top Chinese institutions and enterprises (RXHK, ASIPP, CNPE, CNI23) - 27 sets of adjustable output power supplies for 3 independent ITER in-vessel coils - Contract signed on Jan. 22 - Preliminary design and preparation of Static Magnetic Fields (SMF) Qualification Program is on-going 04 Fusion Technologies Fusion Technologies at SWIP and CFEC Developed kinds of technology to support the fusion research: - Plasma theoretical research and simulation - Diagnostics - Heating (NBI, ECRH) - Power supply - Control - Vacuum - High temperature superconductor - Materials - FW and blanket fabrication - Welding and UDT - Remote handling - Testing platforms Technologies of Diagnostics and Plasmas Fueling Tri-Band High-resolution Spectrometer: - Integrated and compact structure - Simultaneously measuring the profiles of temperature, rotation speed and concentration of carbon and helium ions Supersonic Molecular Beam Injection: - Higher fuelling efficiency, lower neutral recycling - IOS standard approved - Injection speed > 10 Mach - Plasma fueling and control Plasma Control & AI Application - The Fusion Device Operating System (CODIS), integrating control, operation, data, and intelligence, successfully developed and validated on HL-3 - Undertook the IAEA Coordinated Research Project F13022 (AI for Accelerating Fusion R&D) - Two key AI capabilities demonstrated: reinforced learning magnetic control and the large-scale pre-trained model FusionMAE (took over control of magnetic configuration and maintains downstream task performance with missing diagnostics) R&D Platforms for Blanket and Materials - Helium Cooling Experimental Loop - Liquid Metal Integrated Research Platform - Hot Helium Leak Testing System - CNAS-approved Fusion Material Testing Center - ITER-approved High Heat Load Testing platform - Neutron Activation System NBI Platforms for R&D - Full-sized ion source commissioning platform and RF negative ion source platform have been constructed in Tianfu fusion R&D site - Self-developed 7MW NBI beamline, NBI power highest in China - RF negative ion source: Achieving pulse width > 5000s, Max current Density 317A/m², Max beam Energy 202keV - Full-sized commissioning platform (Filament positive ion source 120 keV / 80A) - RF negative ion source platform (RF negative source 400 keV / 20A) RF Platform for R&D ECRH R&D platform: - Self-developed 105GHz gyrotron: 515kW / 1.02s - Multi-frequency, 1000-s, MW gyrotron and transmission line (>100m) test platform (High-Power Long-Pulse 100-Meter-Class Transmission Line System Test Platform, UHV Exhaust Station, Welding Process Platform, Wave Source Test Platform) LHCD R&D platform: - 4MW/3.7GHz high-power dual-tube synthesis test platform - Localized production and validation of high-power klystrons - TE01 circle WG technology Fusion Materials - Reduced Activation Ferritic Martensitic (RAFM) steel - CLF-1: Large-scale ingot up to 13 ton. - Plasma facing materials: Advanced W-K alloy (40kg), The high flux performance tolerate 20MW/m², 2000 cycle. - Functional Materials (beryllium pebble & Li4SiO4 pebble): Completed engineering qualification, Li4SiO4 production capacity ≥10kg/batch & >2 t/y (MSM Facility for Li4SiO4). High-Temperature Superconducting Magnet Technology - The production line for HTS stacked armored conductors has been completed - The 300mm large-diameter, 20T HTS solenoid magnet is under development - Showcases HTS Cable, HTC Conductor, and 20T large-bore HTS solenoid magnet designs. Test facilities for high-field magnets research - Testing facilities for HTS strand and large-scale conductor — critical current, electro-mechanical and high field performance (4.2K-77K, 0-12T, 0-100kN, and 0-40kA) - 12T high field magnet (4.2K-77K, 12kA/40kA) - Tensile testing machine - Tensile and twist testing machine - Fatigue testing machine - Conductor manufacturing equipment - Electro-mechanical tests of cable strand and CICC cable Digital engineering (HIL & Virtual Tokamak) - Hardware-in-the-loop testing prototype developed. Introduce real-data disturbance & delay from hardware. Test & optimize plasma control algorithm without real plasma, decreasing VDE risk. - Integrate experimental data to achieve digital visualization of multi-system status. Lay the foundation for further research on digital twins. Industrial Applications of Fusion Technology Key areas: - Plasma Environmental Governance Technology - Plasma Surface Engineering Technology - High-Voltage Power Supply Technology - New and Special Materials Technology Plasma Environmental Governance Technology - Long Yan-1: China’s first domestic plasma high-temperature melting treatment equipment for low-level radioactive waste from nuclear power plants (applied in nuclear power plants in Hainan Province) - Mobile modular waste discharge treatment device for on-site oxidative melting treatment of radioactive combustible and non-combustible solid wastes Plasma Surface Engineering Technology - Plasma Surface Engineering Technology for surface metallization of non-metallic materials and surface strengthening of metallic materials. - This coating technology can significantly improved tool performance in automotive, machine tools, general machinery, aerospace, among others. - Plasma coating equipment developed by SWIP 05 Prospects of Fusion Research Major Scientific, Engineering and Technical Challenges 3 Major Scientific and Technological Challenges: - Steady-state and self-sustained operation of burning plasma - Materials resistant to high-energy neutron and high heat load - Tritium breeding 3 Major Engineering and Technical Challenges: - High field HTS magnets - Plasma operation and control - Heat extraction Innovation Consortium for Controlled Nuclear Fusion - CNNC and SWIP established the Innovation Consortium for Controlled Nuclear Fusion in December 2023. - Its membership has grown to 60, encompassing various entities such as large enterprise groups, universities, and research institutions, covering the entire industrial chain, including: basic technology research, fusion engineering technology development, key material R&D, advanced manufacturing, and current nuclear power construction and operation. Summary 1. To accelerate fusion energy R&D, aiming for its earliest practical and economical application, is currently the core objective of the fusion community. 2. Global fusion development benefits from ITER. SWIP will actively promote the construction and operation progress of ITER. 3. Two new campuses of CFEC and SWIP in two cities will greatly support the R&D of Fusion technologies. The D-T experiments on HL-3 tokamak is the next key mission. 4. The establishment of the Innovation Consortium and CFEC is of strategic significance in fully leveraging and utilizing the technological accumulations and advantages of China's industrial sector, including the experience accumulated by CNNC in nuclear fission energy R&D and application. Thanks for Your Attention!