School of Aerospace, Xi’an Jiaotong University,
28 West Xianning Road, Xi’an, 710049, China
Since 1970s, China has constructed a serious of small and medium size testing devices for the controllable nuclear fusion, such as CT-6B, HT-6B, HT-6M, KT-5C, HL-1 and HL-1M. Using these devices, a lot of scientific experiments on the plasma physics and other related topics are performed and some remarkable achievements have been obtained. In 1990s, superconductor Tokamak HT-7 has been built, which significantly upgrades the physical experiments level of China. Especially from the starting of this century, a fully superconducting Tokamak device (EAST) has been constructed in the Institute of Plasma Physics of Chinese Academy of Sciences (ASIPP), and a non-circle Tokamak device (HL-2A) has been built in the South West Institute of Physics (SWIP) of China National Nuclear Corporation. These devices have significantly enhanced the plasma parameters and the device operation level. Now, serious of plasma physics related researches are conducted actively in these institutions based on these devices. [1~5]
EAST (Experimental Advanced Superconducting Tokamak) is the first fully superconductor Tokamak device in the world which is independently designed and constructed by experts in China (Fig.1). The features of the device and its major technical parameters are as follows: 16 large D type superconducting Toroidal Field (TF) coils are used to generate central magnetic flux density as large as 3.5 T; 12 Poloidal Field (PF) superconducting magnets can produce flux change as large as ΔΦ ≥10 VS, which makes possible a plasma current of million Amperes; Continued plasma current can keep as long as 1000 seconds, and the plasma can reach 100 million degree of temperature by using a large power heating system.
The other major profile of the EAST device includes: Height of the structure: 11 m, Diameter of the structure: 8 m, Weight of the structure: 400 tons. The Device consists of 6 major subsystems: Vacuum Vessel, PF magnets, TF magnets, Inner and Outer thermal shielding, Outer Vacuum Dewar and the Supporting System. To support the operation of the system, a large scale cryogenic system for providing liquid Helium, a system of high power pulsed electricity source and circuit, testing system for large superconducting magnets, a large computer system for control, data acquisition and processing, a Mega-Watt class lower hybrid wave current driving and microwave heating system, large super vacuum system, and advanced diagnosis systems etc.
Researches on key physical problems related to a stable and advanced Tokamak fusion reactor are carrying out with use of EAST, including: advanced stable plasma operation mode, the confinement and transportation behaviors of plasma, interaction between first wall and plasma under the condition of stable divertor, on-line control of stable plasma and operation safety, new generation techniques for reactor heating and diagnosis. The construction and physical researches can provide experience for the ITER project. EAST is the only device in the next ten years that can provide platform for the advanced long pulse stable operation and non-circular cross-section plasma. It will play a key role for the physical research projects on the development of high quality stable plasma, which can contribute the ITER project and consequently, the development of controllable fusion energy.
Fig.1 An overview of the fully superconductor Tokamak device, EAST
Fig.2 An overview of HL-2A
HL-2A has been finished its construction at the end of 2002 by the SWIP. It is an experimental Takamak device of non-circular cross-section with advanced divertor configuration. The major target of the device is to perform experiments to improve the plasma confinement under condition of high plasma parameters. At the same time, by using its special divertor structure, the experimental researches on the frontier physics problems and related engineering issues will be conducted. The major parameters of the HL-2A are those listed in the Table 1.
The major experiments on plasma physics performed in HL-2A are as follows: Optimization of magnetic field configuration, Acquisition of variable cross-section plasma with triangle configuration, Divertor physics including control of plasma particle and energy, transportation of boundary plasma, researches on SOL physics and advanced divertor configuration, plasma confinement, studies on physics and methods to establish advanced plasma configuration and to improve the plasma confinement based on plasma parameter control, researches on two step heating and current driving, including heating technique of neutral beam injection, driving current and control of plasma rotation, Ion Cyclotron Resonance Heating (ICRH) and fast wave current driving, Lower Hybrid wave current driving, and their synergetic effects with neutral beam injection and ICR, etc. Now, the upgrade of HL-2A to HL-2M is also considered.
Table 1 the major parameters of the HL-2A device
|Major Radius R(m)||1.64||Safety Factor q||3.3~3.5|
|Minor Radius r(m)||0.54||Continue Time of Plasma ts(s)||5.0|
|Plasma Current Ip(kA)||450||LH Current Driving Power PLHCD(MW)||2|
|Central Magnetic Field B(T)||2.8||Electron cyclotron resonance heating Power PECCD(MW)||~1|
|Plasma Stretch Ratio k||1.3||Ion cyclotron resonance heating Power PICRF(MW)||1|
|Plasma Density (m-3)||8×1021||Neutral beam injection power PNBI(MW)||2~3|
Since joined the ITER program in 2003, China institutions are actively involved to the ITER related researches and developments. According to the agreements, the major packages of ITER engineering construction taken charge by China are as follows: TF Conductors (7%), PF Conductors (69%), Feeders (100%), Correction Coils (100%), HV Substation Materials (100%), AC-DC Converter (62%), Transfer Cask System (50%) and Diagnostics (3.3%). The major institutions in charge of the ITER packages are ASIPP and SWIP. At the same time, China Ministry of Science and Technology has started domestic research projects related to ITER program. The foundation of ITER related domestic research is about the same amount with that of the China investment for ITER construction. The major research topics of domestic researches include: Fusion reaction experimental devices, Superconducting magnets, High power pulse current source, Testing Blanket Modules (TBM), Computer control and data acquisition processing system, Plasma heating techniques, Plasma physics etc.
As the hybrid reactor is operated at a sub-critical state by using external neutron source, the critical accident is impossible in principle that makes the hybrid reactor of better safety feature. In addition, the requirement for the fusion core is much lower than that of the commercial fusion reactor, and is close to the current status of the fusion experiment device. Under the support of national high technology research program (863 Program), ASIPP and SWIP have carried out series of researches related to the conceptual design and related technology developments on the hybrid reactor. The physical plan and design of blanket have been developed, which provide a possible way to realize energy production, tritium transformation, nuclear waste processing, and fission fuel production etc. Among these efforts, the FDS-I design of ASIPP is a typical achievement of hybrid reactor. Its fusion power is 150 MW, the major radius of the fusion reactor is 1.4 mm, minor radius is 1 mm, and neutral wall load is 0.5MW/m2. Dual-cooled waste transmutation blanket made of low activation martensitic steel is adopted together with high pressure helium cooling structure, liquid LiPb self cooling breeder zone etc. Researches on the blanket neutral physics include：Breeder of Tritium, LLMA (Long Lived Minor Actinides) transmutation, LLFP (Long Lived Fission Products) transmutation etc. The rector has good capacity for transmutation of long life actinide elements and nuclear waste of fission products. [6~8]
The design and technical development of blanket module is a key research topic on the development of Tokamak magnetic confinement fusion in China. Both liquid metal blanket (LiPb) and Solid pebble bed blanket are studied. The solid pebble bed breeder blanket and related ITER-TBM development are mainly carried out in the SWIP, and the liquid LiPb breeder blanket and its related ITER-TBM application are in charge by the teams in ASIPP. Both DDDs (Design Description Document) of solid and liquid blanket have been delivered to the ITER International Organization (IO). Now, the R/D on the solid TBM is mainly focused on the breeder material – Lithium Silicate and Lithium Titanate, Helium Cooling System and etc. For the Liquid TMB, the detailed design and analyses on the Dual cooled Waste Transmutation Blanket (DWTB), Dual cooled LiPb blanket, and high temperature LiPb blanket have been carried out. The related key techniques on the Liquid TBM, such as development of Low Activation Steel, MHD effect of liquid LiPb flow, Compatibility of Liquid LiPb and structural material, Experimental Loop of Liquid LiPb, Fabrication and experiments techniques of coating are also actively conducted. [9~15]
In China, the conceptual design of Demonstration Power Plant of Fusion Reactor has been performed for a long time. ASIPP has proposed a conceptual fusion power reactor based on many investigations and analyses on the physics and technologies of ITER and other state of the art of the controllable fusion energy researches. The major parameters of the conceptual design are: Fusion power: 2-3 GW, Electricity Power: 1GW, Normalized βN : about 5, Fusion power gain factor: about 30, Major radius: 6 m, Minor radius: 2 m, Aspect ratio: 3, Average neutron wall-surface load: 3MW/m2, First wall average thermal flux load: 0.5 to 13 MW/m2. The structural material of the blanket module is selected as a low activation steel, and the ODS steel is considered as the surface layer material of the first wall. The high pressure helium cooling structure (8MPa), liquid LiPb metal breeder, and dual cooling of flow circulation and self cooling are adopted in the blanket module. In addition, SiC flow channel inserts or coating are used to reduce the pressure loss of liquid LiPb flow due to MHD effect and to prevent the corrosion of structural material, and on-line Tritium extraction and purification are considered. Through R&D, the detailed design of the reactor core and blanket module are obtained, and the neutronics optimization, economical analysis, structural mechanics analysis are performed. The system safety analysis, Environmental influence analysis, and the thermo-hydraulics design analysis are also conducted. [16~19]
From 2001, studies on the low activation steel, such as the smelting technique, mechanical behaviors testing and micro structure analysis of material, have been conducted in relative research institutions. A typical progress in these researches is the development of the China Low Activation Martensitic Steel – CLAM. After a lot of efforts, it has become an independently developed Reduced Activation Ferritic/Martensitic (RAFM) steel of the highest performance in China. The technology already can fulfill the requirement of ton level smelting, and the product has almost the same performance with those of the other RAFM steel such as EUROFER97, JLF-1T, F82H etc. The researches related to the CLAM development are mainly on: design and optimization of the material composition, researches on the purification smelting preparation technology, material behavior testing and micro structure analysis before irradiation, measurement of irradiation resistance behaviors, welding technology, compatibility with liquid LiPb, material surface coating technology etc. Now, the technology to make tons of CLAM in one time is established, which make the large scale production become possible. The mechanical performance of CLAM is similar or better comparing with some typical RAFM steels in the world, and the irradiation behaviors are better than the RAFM used for comparison. 
- Homepage of the Institute of Plasma Physics of Chinese Academy of Sciences (ASIPP), http://220.127.116.11/asipp-english/index.html
- Homepage of South West Institute of Physics,http://www.swip.ac.cn/cfs/english/
- East Superconductor Tokamak Device, http://cbsf.ihep.ac.cn/hefei-EAST/east-index.htm
- L. Qiu, Fusion Energy and applications, China Science Press, 2008.
- B. Shi, The principle and practice of magnetic confinement fusion, China Nuclear Energy Press, 1999.
- Y. Wu et al. Conceptual Design Activities of FDS Series Fusion Power Plants in China. Fusion Engineering and Design, 2006, 81(23-24): 2713-2718.
- Y. Wu et al., Conceptual Design of fusion driven subcritical reactor, China Journal of Nuclear Science and Engineering, 2004, 24(1), 72-80.
- Y. Wu et al., Conceptual Design of the Fusion-driven Subcritical System FDS-I. Fusion Engineering and Design, 2006, 81, Part B: 1305-1311.
- T. Ihli, L. Giancarli, S. Konishi, S. Malang, F. Najmabadi, A.R. Raffray , A. Sagara, Y. Wu., Review of Blanket Designs for Advanced Fusion Reactors. Fusion Engineering and Design. 2008, 83: 912-919.
- Y. Wu, Design Status and Development Strategy of China Liquid Lithium-Lead Blankets and Related Material Technology, Journal of Nuclear Materials, 2007, 367-370, Part B-2: 1410-1415.
- Y.Wu, H.Wang, Y.Ke and W.W, Progress in Magnetic confinement fusion reactor and Design of ITER TBM . China Review of Atomic Physics,2006, 23(2): 89-95.
- Y. Wu and the FDS Team, Conceptual Design and Testing Strategy of a Dual Functional Lithium-Lead Test Blanket Module in ITER and EAST. Nuclear Fusion, 2007, 47(11): 1533-1539.
- Y.Wu, W.W et al., Strategies Design and Testing of ITER Liquid LiPb TBM in China, Nuclear Science and Technology, 2005, 25(4): 347-360.
- C.P.C. Wong, J-F. Salavy, N.B. Morley, I. Kirillov, Y.H. Kim, Y.C. Wu, S. Tanaka, R. Kumar, Overview of Liquid Metal TBM Concepts and Programs, Fusion Engineering and Design, 2008, 83: 850-857.
- Y.W, Q.Huang, Z.Zhu et al., Progress in Design and Development of Liquid LiPb experimental loops in China, Nuclear Science and Engineering, 2009, 29(2): 161-169.
- C.H. Pan, Y.C. Wu, K.M. Feng, S.L. Liu. DEMO Development Strategy Based on China FPP Program. Fusion Engineering and Design, 2008, 83: 877-882.
- Y. Wu, and FDS Team, Conceptual Design of the China Fusion Power Plant FDS-II. Fusion Engineering and Design, 2008, 83(10-12): 1683-1689.
- Y.Wu, W.W and S.Liu, Conceptual Design of Fussion Nuclear Power Plants, Chin a Nuclear Science and Technology,2005, 25(1): 76-85.
- C.P.C. Wong, V. Chernov, A. Kimura, Y. Katoh, N. Morley, T. Muroga, K.W. Song, Y.C.Wu, M. Zmitko, ITER Test Blanket Module Functional Materials. Journal of Nuclear Materials, 2007, 367-370: 1287-1292.
- Q.Huang, C.Li and Y.Li, Progress of Researches on the Low Activation Martensitic Steel CLAM in China, China Nuclear Science and Engineering, 2007, 27(1): 41-50.