Southwestern Institute Of Physics Research Articles

Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China

ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.

The research progress of an E//B neutral particle analyzer

An // neutral particle analyzer (NPA) has been designed and is under development at Sichuan University and Southwestern Institute of Physics. The main purpose of the // NPA is to measure the distribution function of fast ions in the HL-2A/3 tokamak. The // NPA contains three main units, i.e. the stripping unit, the analyzing unit and the detection unit. A gas stripping chamber was adopted as the stripping unit. The results of the simulations and beam tests for the stripping chamber are presented. Parallel electric and magnetic fields provided by a NdFeB permanent magnet and two parallel electric plates were designed and constructed for the analyzing unit. The calibration of the magnetic and electric fields was performed using a 50 kV electron cyclotron resonance ion source (ECRIS) platform. The detection unit consists of 32 lutetium-yttrium oxyorthosilicate (LYSO) detector modules arranged in two rows. The response functions of , hydrogen ions ( , and ) and for a detector module were measured with Am, Cs and Eu sources together with the 50 kV ECRIS platform. The overall results indicate that the designed // NPA device is capable of measuring the intensity of neutral hydrogen and deuteron atoms with energy higher than 20 keV.

Composite effect of helium ion irradiation and transient thermal shock on recrystallized pure W and W-Y2O3 alloys

Tungsten (W) is the material of choice for the thermal nuclear fusion reactors of the future. In this research, pure W (PW) and W-Y2O3 alloys made by the SWIP (Southwestern Institute of Physics) wet chemical process are chosen. Simulation of long-term extreme operating conditions of fusion reactor materials by annealing and recrystallization of forged tungsten materials. The thermal shock resistance of recrystallized PW and W-Y2O3 alloys to radiation by helium (He) ions at a flux of 1.4 × 1022 ions/(m2s) was investigated. The surface microscopic pictures of recrystallized W and W-Y2O3 alloys were analyzed using scanning electron microscopy (SEM) and a laser surface profiler under the conditions of helium ion irradiation and electron beam transient thermal shock. After helium ion irradiation, the surface of PW and W-Y2O3 alloys exhibits a fuzzy structure. Under the same 0.33 GW/m2 thermal shock condition, the average width of the main cracks of PW and W-Y2O3 alloys significantly increases after helium ion irradiation, indicating that helium ion irradiation can reduce the thermal shock resistance of the surface of recrystallized W materials.

A new scheme of plasma control system based on real-time Linux cluster for HL-2M

HL-2M is a medium-sized tokamak constructed by the Southwestern Institute of Physics (SWIP) in China. A new plasma control system (PCS) based on the software framework of DIII-D PCS for HL-2M has been proposed to replace the original loose system. The new hardware system proposed in this paper consists of three computer nodes and peripheral equipment. Data acquisition is completed by a standing server and a reflective memory network that passes back data. The software architecture of DIII-D PCS is generally retained, but some interfaces are added to adapt to the existing system. In addition, the Linux operating system has also been upgraded and optimized so that real-time tasks can achieve better performance. The initial system involves some basic control algorithms satisfying early experiments. The new PCS has been preliminarily verified through simulation and can output control commands as expected in the integrated environment. This paper focuses on two aspects of DIII-D PCS on a new device, i.e., performance and interface adaptation. More efforts will be made to add more algorithms and use the system in real experiments.

Control and Protection Strategy of Toroidal Field Magnet Power Supply System in HL-2M

The HL-2M device is a new generation tokamak at the Southwestern Institute of Physics that is used to develop nuclear fusion research. To confine the plasma, the toroidal field (TF) coil of HL-2M is needed to output 140-kA direct current to generate a 2.2-T magnetic density flux. Therefore, the TF power supply system was developed. It is powered by two flywheel pulsed motor-generator sets, and a cophase counter parallel connection bridge rectifier was used to provide more current. According to the TF power supply system topology, the amplitude of the magnetic flux density of the TF coil can be tuned by adjusting the excitation system of the flywheel pulse generator; the control strategy and mathematical formula are given. Then, the TF power supply protection system is established, and protection strategies under different faults are designed. The control and protection performance of the TF power supply system has been successfully applied and verified in the first plasma discharge of the HL-2M experiment.

ICRF heating schemes for the HL-2M tokamak

The HL-2M tokamak is a new medium-sized tokamak at SouthWestern Institute of Physics. Two of its key missions are to achieve 10 keV ion temperature and investigate the behavior of energetic particles relevant to burning plasmas. A 6 MW ion cyclotron range of frequencies (ICRF) heating power is embedded in the next upgrade program of HL-2M. In order to facilitate the engineering design of the ICRF system, this paper analyses the main ICRF heating schemes for HL-2M, in terms of ion heating and energetic ion generation in particular. D(H) minority heating and the 2nd harmonic D will act as the main ion heating schemes, for which the optimal RF frequency range 27–33 MHz, antenna parallel wavenumber k // ∼ 8 m−1 are proposed and strong single pass absorption is expected under typical HL-2M plasma parameters. Full wave simulations carried out via TORIC/steady-state Fokker–Planck quasilinear solver and TRANSP codes suggest that by adopting three ion scheme or synergetic heating on neutral beam injection D ions by the 2nd harmonic D, energetic ions with energy at MeV level can be produced. This study shows that ICRF heating could play significant roles in ion heating, energetic ion generation in HL-2M.

A newly constructed Linear Experimental Advanced Device LEAD

The Linear Experimental Advanced Device (LEAD) has been constructed for comprehensive studies on plasma turbulence, plasma-material interaction and new techniques for tokamaks at Southwestern Institute of Physics (SWIP). The vacuum vessel of this linear device is 3055 mm long. It consists of three parts, i.e. the source chamber, the linear magnetized confinement chamber and the plasma-material interaction (PMI) chamber. Kinds of plasma sources, including radio-frequency (RF) sources and a hot cathode source, have been developed to produce steady-state plasma. Wave-mode helicon plasmas can be excited by a large-size four-ring-antenna RF source when the input power is higher than 180 W. The typical electron density is higher than 1019/m3. Transfer from distinct narrowband drift waves to broadband turbulence is observed by means of adjusting the axial magnetic field, which makes LEAD a suitable device to investigate plasma turbulence. Auxiliary heating systems, including a 2.45 GHz Electron Cyclotron Resonance and an ion heating source, and a 1.5 MV ion accelerator are to be installed on LEAD. It promises a particle flow of over 1023/m2s and a heat load of several MW/m2, which makes LEAD a suitable platform for not only fundamental plasma physics but also PMI research.

An allocation method of reliability goal for a compact fusion device

Currently Southwestern Institute of Physics (SWIP) is performing the conceptual design of a compact fusion device. With the top-level objectives and operation plan determined, the top-level reliability goal has been established. Once the reliability goal is established, the goal must be translated into subsystems or main functions. Currently, there is no relevant literature to study the allocation method suitable for the fusion device systematically. This paper provides an allocation method suitable for the currently designed compact fusion device, namely Analytic Hierarchy Process (AHP), which combines qualitative and quantitative analysis. It uses expert experiences to deeply understand the nature, influencing factors, and internal relationships between complex subsystems. By establishing the hierarchical structure model and constructing the judgment matrix, the distribution weight of each subsystem is given, and the allocation of the reliability goal is completed.

Effect of irradiation temperature on radiation hardening of CLF-1 steel

A kind of reduced-activation ferritic-martensitic steel, CLF-1, was developed by the Southwestern Institute of Physics, China, as a candidate structural material for fusion and fourth-generation fission reactors. This work reports the results from the nanoindentation and transmission electron microscopy study performed on this steel after irradiation with 2.75 MeV Fe11+ ions up to 1 and 10 dpa at temperatures of 300 - 500 °C. An increase in hardness was observed in the damaged layer of samples irradiated at 400 °C, while the samples irradiated at 500 °C exhibited a softening of the damaged layer. The hardening can be primarily, although not exclusively, attributed to the increase of <100>-type dislocation loops observed at the increasing irradiation temperature. Nano-sized cavities around the dislocation loops and along dislocation lines were also observed, explaining the lattice swelling after irradiation at 500°C. The research results provide a valuable reference for the development of irradiation-hardening-resistant Fe-Cr alloys for advanced nuclear energy systems.

Design of the HL-2M Support System

The HL-2M tokamak was constructed at Southwestern Institute of Physics. The support system is one of the main subsystems of the device. The magnetic coils and vacuum vessel are located and integrated by the support system. According to its function, the support system is divided into gravity support, support for poloidal field coils, and anti-torque structure for toroidal field coils. The structure design and the finite element analysis of the support system is introduced in this paper.

Design of PLC and EPICS based control system for a deuteron RFQ

Peking University and Southwestern Institute of Physics are jointly developing a new deuteron RFQ to study the migration and deposition of impurities on the first wall of a tokamak facility. This RFQ will accelerate a 10-mA deuteron beam from 40 keV up to 1.5 MeV at a duty factor of 1%. In this paper, the detailed design of the prototype control system based on a Siemens PLC and EPICS/CSS software environment for the RFQ accelerator is presented. The system is composed by many kinds of sub-systems developed using different hardware solutions. The interlock protection mechanism is introduced into the control system. The EPICS Archiver Appliance is used to provide the service of data archiving and retrieval. The system has been successfully implemented for RF conditioning and beam commissioning.

Development of a Radio-Frequency Quadrupole Accelerator for the HL-2A/2M Tokamak Diagnostic System

In order to figure out the migration and deposition of impurities on the first wall of HL-2A/2M tokamak, Peking University and Southwestern Institute of Physics are co-developing a deuteron RFQ as part of the in situ ion-beam diagnostic for the material. The RFQ, which operates at 162.5 MHz, is designed to accelerate a 10-mA deuteron beam from 40 keV up to 1.5 MeV. Key design considerations and the final design parameters are presented. The RFQ has been conditioned at a 1% duty factor for 80 h at RF cavity power of 55 kW. The specific shunt impedance of the cavity is 221 kΩ·m by measuring the bremsstrahlung spectrum. The intrinsic Q-value after the high-power tests measured by the Ring-Down method is 13,780. Beam commissioning has been taken place during the first half of 2021, and the beam measurements include beam current and energy of 2H+ ion. A 10 mA 2H+ beam was successfully accelerated through the RFQ.

RF design and tuning of a deuteron RFQ for detecting impurity deposition on a tokamak

Peking University and Southwestern Institute of Physics are jointly developing a new deuteron Radio Frequency Quadrupole (RFQ) accelerator to study the migration and deposition of impurities on the first wall of a tokamak facility. The RFQ operates at 162.5 MHz with the duty factor of 1%, which can accelerate 10 mA deuteron beam from 40 keV up to 1.5 MeV within 2.2 m length. Such four-vane RFQ is divided into two segments and equipped with 40 tuners in total. In the electromagnetic (EM) design, the dipole stabilizer rods (DSRs) were optimized to obtain a mode separation of 3.1 MHz. The tuners and undercuts were also designed and optimized to fulfill the requirements of the resonant frequency, flat longitudinal field distribution and field tuning. Following the EM design, multi-physics analysis and structure error analysis was carried out. The cavity was fabricated with 99.9% oxygen-free electronic (OFE) copper. The assembly and braze of each RFQ module have been completed, the results of metrological measurements via coordinate measuring machine (CMM) indicate that there was no unexpected deformation. The intrinsic Q-value of the whole cavity is 12834 in the low level radio frequency (RF) measurements, which is 87% of the theoretical simulation with electrical conductivity of 5.8 × 107 S/m. After tuning, the quadrupole perturbative component decreased from 2.6% to 0.6%, and the dipole perturbative components decreased from 6.1% to 1%.

The Engineering Commissioning of the Magnets System On HL-2M Tokamak

HL-2M tokamak is one of largest magnetic confinement fusion devices designed and constructed by Southwestern Institute of Physics (SWIP), China National Nuclear Corporation (CNNC), the first plasma was obtained on December, 4 <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">th</sup> , 2020. The magnets system which plays an important role for the plasma confinement consists of 20 toroidal field (TF) coils and 18 poloidal field (PF) coils. The magnets system commissioning was carried out by two phases: self-commissioning and integrated commissioning. The first phase: self-commissioning including joint resistance commissioning, turn to turn insulation commissioning, ground insulation commissioning, deionized water sealing commissioning was accomplished during the installation and test of the magnets system. The second phase: the integrated commissioning with power supply system, ground system, measurement and data acquisition system, plasma control system, interlock and safety system was accomplished based on the results of the first stage. The total commissioning of magnets system in HL-2M tokamak for the first plasma have been accomplished successfully.

Status of the Domestic Large Helium Refrigeration System of HL-2M Tokamak

HL-2M, the satellite tokamak of ITER (International Thermonuclear Experimental Reactor), has successfully completed the first plasma experiment in December 2020 in Southwestern Institute of Physics, China National Nuclear Corporation (CNNC-SWIP). The 500 W/4.5 K helium cryogenic system used in HL-2M is mainly used to cool down the cryopumps and superconducting magnets. The refrigeration system is the heart components of the cryogenic system. The design and development of the refrigeration system witness our significant cooperation with Fullcryo Company. The refrigeration system has completed the factory commissioning and on-site installation at the end of 2019. This paper presents the performance characteristics of the main key equipment in the system.

Study on Characteristics of the Helium Refrigeration System of HL-2M Tokamak

HL-2M, a new generation of tokamak, has been built continuously in 2020 in Southwestern Institute of Physics, China National Nuclear Corporation, and the first plasma discharge was successfully achieved in 12th, Dec. This tokamak experiment device has high design parameters, and its maximum operating current reaches 3 MA. As a satellite device of ITER, it aims to explore and solve the severe physical and engineering problems. Cryogenic system is an auxiliary system of HL-2M, which mainly meets the stable operation of cryopumps and superconducting magnets. The refrigeration capacity under the condition of Liquid Nitrogen Precooling will up to 500W@4.5K, the liquefaction rate is 160L/h. In this report, we present the main characteristic parameters of the refrigeration system. At present, the system has completed the field assembly, and is expected to start commissioning in June 2021.

Neutronics and shielding design of CFETR HCCB blanket

Tritium breeding blanket (TBB) is one of the most important components of China Fusion Engineering Test Reactor (CFETR) with functions of tritium breeding, heat generating and nuclear shielding. This paper presents the progress on the neutronics and shielding design of the helium cooled ceramic breeder tritium breeding blanket (HCCB TBB) developed by the Southwestern Institute of Physics (SWIP). Based on the latest profile and layout configuration of CFETR TBB, the preliminary neutronics design, optimization and analysis of HCCB TBB have been completed. After several rounds of neutronics design and optimization, the TBR of HCCB TBB is 1.177, and the value drops to 1.101 considering the influence of auxiliary systems, which can meet the design objectives of CFETR. The total nuclear heating of HCCB TBB is 1402 MW with 1500 MW fusion power. Among all the blanket modules, blanket module #3 is subjected to the strongest neutron irradiation. Meanwhile, the overall shielding analysis for TF coils of CFETR has been performed based on the current design of HCCB TBB, and the preliminary shielding design and optimization for the weak shielding areas such as HCCB TBB, divertor and neutral beam injector (NBI) port have been carried out. After shielding optimization, the fast neutron fluence and nuclear heating of all TF coils are below the design limit. By increasing the number of source neutron histories and using some variance reduction techniques in the MCNP calculation, the relative errors of neutronics and shielding analysis results for HCCB TBB are all less than 5%.

FW and Blanket Technology Development Progress at SWIP

Blanket, including its first wall (FW), is one of the key components in the fusion reactor, which works in the severe condition with the huge high energy neutron irradiation, high heat flux and electromagnetic load. Facing these technical challenges for the blanket and its FW, Southwestern Institute of Physics is engaging in the development of the blanket technology following the proposed China magnetic confinement fusion development strategy. This paper presents the design and R&D progress of the ITER FW and blanket, Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) for ITER and HCCB tritium blanket for China Fusion Engineering Test Reactor (CFETR). The design of the ITER enhanced heat flux FW has been updated to improve manufacturability and to reduce water leak risk. Key technologies have been qualified by high heat flux tests of small mock-ups and full-scale plasma-facing fingers. New assembly technologies are under development. Following the completion of full-scale prototype qualification, the ITER shielding blanket modules are under manufacturing for series production. The preliminary design of HCCB TBM for ITER is under implementation to improve the manufacturability and inspectability based on the fabrication study. Two prototypes of TBM submodule have been fabricated and tested to validate the feasibility of manufacture procedure, which are finally assembling into one semi-prototype of TBM. The preliminary design of HCCB tritium blanket for CFETR has been completed and the R&D has been started, especially for the special technologies different with the ITER blanket and TBM. Meanwhile, the progress of reduced activation ferritic/martensitic steel and function materials for blanket are also presented.

A large-area planar helicon plasma source with a multi-ring antenna on Linear Experimental Advanced Device (LEAD)

A planar helicon plasma source with a four-ring antenna has been developed on Linear Experimental Advanced Device (LEAD) in Southwestern Institute of Physics. The diameter of the largest antenna ring is 320 mm. This source is located outside of the vacuum chamber without any vacuum interface and injects radio frequency power into the chamber through a 340-mm-diameter quartz window. A low power threshold of 150 W for electromagnetic-mode to wave-mode transfer is experimentally confirmed. A large volume plasma with a density of over 1019 m−3 and a plasma generation efficiency (total number of electrons diveded by input power) of over 30× 1013 W−1 indicate the high performance of this large-area helicon plasma source, promising LEAD a suitable device for fundamental plasma physics and plasma material interaction research.

Preliminary development of a conceptual first wall for DEMO

The first wall is one of the key components in a tokamak, in particular, for DEMO in the future. The conceptual first wall for a DEMO reactor is designed, fabricated and tested in this paper. It has a sandwich-like structure, which consists of a reduced activation ferritic/martensitic (RAFM) substrate, an interlayer and a plasma-facing material of tungsten produced by chemical vapor deposition (CVD-W) due to its characteristics of high density, high purity and superior thermal shock resistance. The interlayer is required to not only have good adhesion between CVD-W and the CLF-1 substrate, which is a type of RAFM developed by the Southwestern Institute of Physics, but also the capability of tritium transport prevention due to the great importance of controlling the tritium buildup in the first wall, improving the fuel efficiency and conforming to the safety regulations of tritium. Titanium nitride (TiN), a well-known tritium barrier, is selected as the interlayer material in this work. In total, eight sandwich-like first wall samples have been fabricated, and thermal cycle tests and plasma exposure tests are implemented. The results of material analysis show that the quality of the TiN coating layer is critical for good adhesion and tritium prevention. At the defects of the TiN interlayer, fluorine (F) penetrated into the CLF-1 during the CVD-W fabrication and, as a result, the sandwich-like structure was broken near the CLF-1 side. For some samples, the TiN interlayer made by CVD has sufficient adhesion as an interlayer between CVD-W and CLF-1 during tests so far. These encouraging test results support further examination of TiN as an interlayer material for W on an RAFM steel as a first wall material. Preliminary results indicate that the design and fabrication of the conceptual first wall are feasible. Further tests, including the improvement of the interlayer quality and the tritium permeation test, are necessary in the future.