Chinese Fusion Engineering Test Reactor Research Articles

Numerical simulation of <inline-formula><tex-math id="Z-20231101091527">\begin{document}$\boldsymbol \alpha$\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_Z-20231101091527.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_Z-20231101091527.png"/></alternatives></inline-formula> particle slowing-down process under CFETR scenario

The high-energy <i>α</i> particles produced by deuterium-tritium fusion are the primary heating source for maintaining high temperatures in future tokamak plasma. Effective confinement of <i>α</i> particles is crucial for sustaining steady-state burning plasma. The initial energy of <i>α</i> particles is <inline-formula><tex-math id="M1">\begin{document}$ 3.5 {\text{ MeV}} $\end{document}</tex-math><alternatives><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_M1.jpg"/><graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_M1.png"/></alternatives></inline-formula>. According to theoretical calculations, it takes approximately 1 second to slow down <i>α</i> particles through Coulomb collisions to an energy range similar to the energy range of the background plasma. In the slowing-down process, some <i>α</i> particles may be lost owing to various transport processes. One significant research problem is how to utilize <i>α</i> particles to effectively heat fuel ions so as to sustain fusion reactions in a reactor. Assuming local Coulomb collisions and neglecting orbital effects, a classical slowing-down distribution for <i>α</i> particles can be derived. However, considering the substantial drift orbit width of <i>α</i> particles and the importance of spatial transport, numerical calculations are required to obtain more accurate <i>α</i> particle distribution function. In this study, the particle tracer code (PTC) is used to numerically simulate the slowing-down process of <i>α</i> particles under different scenarios in the Chinese Fusion Engineering Test Reactor (CFETR). By combining particle orbit tracing method with Monte Carlo collision method, a more realistic <i>α</i> particle distribution function can be obtained and compared with the classical slowing-down distribution. The results show significant differences between this distribution function and the classical slowing-down distribution, particularly in the moderate energy range. Further analysis indicates that these disparities are primarily caused by the strong radial transport of <i>α</i> particles at these energy levels. The research findings hold profound implications for the precise evaluating of ability of <i>α</i> particles to heat the background plasma. Understanding and characterizing the behavior of <i>α</i> particles in the slowing-down process and their interaction with the plasma is critical for designing and optimizing future fusion reactors. By attaining a deeper comprehension of the spatial transport and distribution of <i>α</i> particles, it becomes possible to enhance the efficiency of fuel ion heating and sustain fusion reactions more effectively. This study establishes a foundation for subsequent investigations and evaluation of <i>α</i> particles as a highly efficient heating source for fusion plasmas.

Design Consideration of Cavity and MIG of a 170/230-GHz Dual-Frequency Megawatt-Class Gyrotron for CFETR

In this article, a dual-frequency, megawatt-class (MW-class) gyrotron for Chinese Fusion Engineering Testing Reactor (CFETR) is analyzed, which can operate individually at frequencies of 170 and 230 GHz with the operating modes of TE+33,12 and TE+44,16, respectively. A triode-magnetic injection gun (MIG) is investigated for this dual-frequency gyrotron. With the help of the time-dependent, multimode, self-consistent code, the startup scenarios are investigated with considering the realistic electron beams. The results show that the gyrotron can steadily operate at the operating modes with MW-class under the transverse velocity spreads from 3% to 7% together with the beamwidth from 3 to 5 times Larmor radii. When considering the velocity spreads of 7% and the electron radial thickness of 4 times Larmor radii, the output power of TE+33,12 mode can reach 875 kW with a beam voltage of 70 kV and a current of 48 A, and the output power of TE+44,16 mode can reach 805 kW with a beam voltage of 65 kV and a current of 45 A.

Particle-transport/depletion/activation/source-term coupling analysis for CFETR with NECP-MCX

As a bridge between ITER and Chinese DEMO, the design of Chinese Fusion Engineering Test Reactor (CFETR) is being carried out. The tritium breeding ratio (TBR) and the ShutDown-Dose-Rate (SDDR) are two significant neutronics parameters for CFETR and other fusion reactors. However, in the previous researches, the variation of the TBR with time was analyzed without consideration of the activation of beryllium. In the analysis of the SDDR, conventional methods assumes that the neutron spectra are not changed in the whole operation duration. To analyze the TBR and the SDDR with higher precision, the functionalities of the particle-transport/depletion/activation/source-term coupling analysis have been developed based on NECP-MCX, which is a Monte-Carlo-Deterministic particle-transport code developed by the Nuclear Engineering Computational Physics (NECP) Lab at Xi'an Jiaotong University. The transmutation solver NECP-Erica, developed also by the NECP Lab, is coupled with NECP-MCX internally. The integrated analysis code, which inherits the name NECP-MCX, is verified by several fusion-related experimental benchmarks. The verification results show the reliability of NECP-MCX. After the verification, the particle-transport/depletion/activation/source-term coupling analysis of CFETR is performed with NECP-MCX. Numerical results show that ignorance of the activation of beryllium overestimates the TBR of CFETR by 1.42% after an operation time of 16 years with a capability factor of 0.5. In the analysis of the SDDR, the ignorance of spectrum variation brings ∼2.0% discrepancy in the activity and decay heat of the structure materials.

Analysis of Existing and Proposed Maintenance Deployment Systems Toward DEMO MPD Development

This article reports on a study of previously existing or proposed maintenance deployment systems with similar functions and structure to that of the proposed multipurpose deployer (MPD) for demonstration (DEMO fusion power plant project). The current MPD design iteration consists of a boom deployment system that is ~30 m long and can support a payload of ~1000 kg, while still being able to access the DEMO vacuum vessel through a 2.78 m high by 1.08 m wide port. The purpose of this work is to benefit from previous experience by comparing the mechanical attributes and performance of systems as well as their advantages and disadvantages and any issues encountered to bring design input to MPD design development. The following systems were investigated: Joint European Torus (JET) in-vessel remote handling booms, telescopic articulated remote mast (TARM), Next European Torus (NET) experimental device for in-torus handling (EDITH), Tokamak Fusion Test Reactor (TFTR) Maintenance Manipulator, and Snake-like Robot Arms in Nuclear Environments. Systems that are currently in development for ITER and Chinese Fusion Engineering Test Reactor (CFETR) were also investigated according to their latest available design iterations. This article concludes that these systems, comprising of articulating links to form long-reach slender structures, give rise to challenges with their payload, stiffness, and control. The straight boom style system would be the most suitable design for the current tasks that a DEMO MPD is expected to perform. However, there is no particularly strong candidate without first fully defining the requirements and constraints that a DEMO MPD must adhere to.

Prediction and Validation for Welding Deformation of the Upper Port Stub

The vacuum vessel (VV) as an ultrahigh vacuum pressure vessel is one of the essential components of the Chinese Fusion Engineering Test Reactor (CFETR) device. VV has strict position tolerance. The VV port stub is a typical large-scaled welded structure manufactured by electron beam welding (EBW) technique. Welding has the most significant influence on deformation. Therefore, it is essential to analyze welding deformation. A systematic welding simulation method aimed at welding deformation controlling for a large EB welded structure is proposed in this article. First, the dynamic process of EBW of 50-mm-thick ultralow carbon austenitic stainless steel was simulated based on the thermal elastoplastic theory. The weld forming mechanism of high energy density welding of the thick plate was revealed. Then the inherent strain extracted from the results of the simulation of local welding was introduced to the simulation of tailor welding of the port stub. Finally, optimal welding sequence and clamping conditions were obtained, and the reliability of the simulation was verified by the profile measurement results of the inner shell of the upper port stub.

Recent EAST Experimental Results and Systems Upgrade in Support of Long-Pulse Steady-State Plasma Operation

In recent years, EAST has achieved a 60-s scale steady-state plasma operation (SSO), which is electron dominant heated by the electron cyclotron wave and low hybrid wave. Substantial progresses have been made in the development of long-pulse operation, such as active control of radiative divertor, edge localized mode (ELM) suppression, and He plasma experiment. To further develop the steady-state operation scenario and study the related critical physical issues, the upgrade of EAST has been completed in 2020. The auxiliary heating systems were rearranged and upgraded with a total heating power capability of 33 MW. The 2.45-GHz lower hybrid current drive (LHCD) launcher has been updated from full active multijunction (FAM) to passive active multijunction (PAM) to avoid damage caused by plasma–material interaction. The injection power capability of electron cyclotron resonance heating (ECRH) system is upgraded to 1.4 MW. The upgraded ion cyclotron resonance frequency (ICRF) antenna injects low parallel wavenumber ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k_{\vert \vert }$ </tex-math></inline-formula> ) spectra with good single pass absorption. The neutral beam injection (NBI) system is moved from F-port to D-port and is reoriented from counterclockwise to clockwise injection. The armor material of lower divertor has been upgraded from graphite to tungsten improving the steady-state heat handling capability from 2 to 10 MW/m2. Based on the upgraded pumping system and improved conductance, the effective pumping speed for <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$D_{2}$ </tex-math></inline-formula> of the EAST lower divertor increased twice. With all these upgrades, 100-s scale steady-state operation ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T_{e0} &gt;10$ </tex-math></inline-formula> keV) and 1056-s high-performance operation with 1.73-GJ injection energy were achieved in 2021 campaigns. The experience in high-performance long-pulse operation obtained on EAST will lay a foundation for the ITER and Chinese Fusion Engineering Testing Reactor (CFETR) operation.

Research on fault diagnosis method for hydraulic system of CFETR blanket transfer device based on CNN-LSTM

Conducting fault diagnosis on the hydraulic system of the blanket transfer device Mover in the Chinese Fusion Engineering Test Reactor (CFETR) is a key technical issue that needs to be addressed urgently. In this article, a CNN (Convolutional Neural Networks)-LSTM (Long Short-Term Memory) deep learning model-based method is proposed for fault diagnosis, combining the advantages of feature extraction of the CNN model with the advantages of the LSTM model for time series data processing. Therefore, this model shows a " multi-perspective" property, greatly improving its ability to extract features from data. In the fault diagnosis experiment under the condition of four typical faults, the proposed model has the highest accuracy of 98.56% on the test set and good efficiency in computation time compared to the other three models. This method provides some insights for future research on the Prognostics and Health Management (PHM) of the Mover's hydraulic system and the CFETR's remote handling intelligent operational decision system.

Parametric analysis of influence factors to the in-vacuum vessel LOCA for CFETR based on RELAP5

The water-cooled blanket and water-cooled divertor located inside the vacuum vessel (VV) are two crucial components of the Chinese Fusion Engineering Testing Reactor (CFETR). The flash vaporization from high temperature and high pressure water pressurizes the VV when in-VV Loss of Coolant Accident (in-VV LOCA) occurs. As the first confinement barrier of the CFETR, the integrity of the VV is required to be guaranteed to contain radioactive materials. Parametric analysis of the in-VV LOCA is carried out to analyze the key influence factors on the pressure increasing rate and leaked water mass so as to provide basis for safety facility designs of the CFETR. The water-cooled components, VV and vacuum vessel pressure suppression system are modeled using RELAP5. The pressure increasing rate and leaked water mass are analyzed under circumstances with different discharge pressures, VV initial pressures and the breach areas. Moreover, the pressure increasing rate increases with the discharge pressure, VV initial pressure and breach area increased. The leaked water mass and its variation both increase with breach area increased.

Subwinding Pack Embedding Experiment of CRAFT TF Coil

The Comprehensive Research Facility for Fusion Technology (CRAFT), national big science facility in China, is focused on exploring and building key technologies and prototype systems for Chinese fusion engineering test reactor (CFETR). The toroidal field (TF) coil is one of the most important parts of CRAFT, targeting at manufacturing a prototype TF coil for CFETR. The TF coil is gigantic “D” shaped, consists of a winding pack (WP) and coil case. The WP is composed of 3 sub-WPs, called low (LF), medium (MF), and high field (HF) WPs, which are embedded to form the WP. The embedding gap between neighboring sub-WPs ranges from 10 mm at the inboard leg to 30 mm at the outboard leg and 120 mm at the helium inlet regions. The LF, MF, and HF WPs are wound separately by NbTi, conventional Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn, and high property Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn cable in conduit conductor. The fabrication of MF and HF WPs follows the routine of “Wind and React” approach; therefore, the maximum strain on the conductor shall not exceed ±0.1% during the whole embedding process. In this article, the embedding system is introduced, the strain of the conductor is analyzed, the embedding process is verified by 1/4 scale MF and HF model WPs. The results of analysis and experiment imply that the conceived embedding process is fully engineering feasible.

Implementation of CO2 and PbLi as working fluids in RELAP5/MOD3.3 towards accident analysis of COOL blanket for CFETR

The supercritical CO2 cOoled Lithium-Lead (COOL) blanket is an advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). The COOL blanket has dual coolant. Supercritical CO2 (S-CO2) is used to cool the structural components, while lithium-lead (PbLi) is self-cooling in the breeding zones. To verify the reliability of the COOL blanket, it is essential to conduct the accident analysis. The system code RELAP5/MOD3.3 is developed with implementation of the two working fluids of the COOL blanket, namely PbLi and CO2. The outboard blanket segment of the COOL blanket is modeled using the modified RELAP5. The plasma pulse operation and fusion power excursion transient caused by Multifaceted Asymmetric Radiation From the Edge (MARFE) are simulated to study the thermal dynamic behavior of the COOL blanket. Furthermore, sensitivity analysis of Loss of Flow Accident (LOFA) is carried out. The results show that the outlet temperature of PbLi under steady-state operation meets its design requirements, and the temperature of the structural material does not exceed its temperature limit. The thermal hydraulic parameters have pulsed characteristics under plasma pulse operation. The MARFE event can cause sharp temperature rise of the First Wall (FW) but the material remains below its temperature limit. In case of LOFA of the CO2 system, it is suggested to restore CO2 mass flow rate in time to prevent the FW from overheating when the outlet temperature of PbLi is 973 K. The results of the preliminary accident analysis can provide the basis for the design and optimization of the blanket structure and system.

Research on the Dynamic Characteristics of the CFETR Vacuum Vessel Based on the Modal Synthesis Method

The vacuum vessel is the core component of the Chinese Fusion Engineering Testing Reactor (CFETR); its main function is to remove nuclear heating, provide safety shielding, and maintain a high-quality vacuum environment. Therefore, the safety of the vacuum vessel is of great significance to the CFETR, and examining its dynamic performance is necessary. However, the conventional finite element method takes too long to perform the dynamic analysis of the vacuum vessel, which greatly reduces the efficiency of the design and analysis. Based on the modal synthesis method, this study uses ANSYS software to establish a substructure model of the CFETR vacuum vessel. A modal analysis and harmonic response analysis are conducted, and their results are compared with those of the conventional finite element model. The results show that the substructure model not only has the same accuracy as conventional finite element models, but that it also greatly reduces the time of dynamic calculation.

Research on Nondestructive Examination of Jacket Section for CFETR TF Coil

The Chinese Fusion Engineering Testing Reactor (CFETR) is aiming to bridge the gap between the International Thermonuclear Experimental Reactor (ITER) and the first commercial fusion power plant, a necessary complement of the ITER. The requirement for sustaining long burn duration as specified in the duty time CFETR mission necessitates the use of superconducting magnets. Totally 16 toroidal field (TF)D-shaped coils which are of cable-in-conduit conductor type are designed. All the subcomponents have very demanding requirement to withstand the severe environments. This paper gives the nondestructive examination (NDE) method developed for the jacket of TF conductor which is made of 316L stainless steel. Based on the linear elastic fracture mechanics, maximum acceptable defect sizes in the TF jacket material have been defined. Phased array ultrasonic testing(PAUT)method and eddy current test (ECT) method were developed for the circular-in-square jacket inspection. For PAUT method, focus laws were calculated and optimized. For ECT method, specific ECT probe was designed and manufactured. Benchmark experiments were carried out to study the NDE reliability.

Research on hot isostatic pressing process of CFETR WCCB components

Chinese Fusion Engineering Test Reactor (CFETR) is a superconducting magnet tokamak reactor of equivalent size and complementary functionality to that of ITER. The water-cooled ceramic breeder (WCCB) breeder blanket designed by the Institute of Plasma Physics, Chinese Academy of Sciences is one of the candidate breeder blanket concepts for CFETR. As an important part of the fusion reactor, the WCCB directly determines the safe operation of the fusion reactor. In the WCCB manufacturing technology, the hot isostatic pressing (HIP) process is applied to the manufacture of WCCB due to its unique technical characteristics. This paper introduces the process research and inspection during the development of WCCB components, and the results will be the guidance for the manufacture of WCCB Components.

SOLPS-ITER modeling of CFETR advanced divertor with Ar and Ne seeding

The Chinese Fusion Engineering Testing Reactor (CFETR) is a project proposed by the Chinese fusion community to bridge the gap between ITER and a commercial fusion power plant with fusion power up to 1 GW. The mitigation of divertor target heat fluxes for such a powerful machine is a challenging problem, which might appear to be more severe than in ITER. In the present paper, the results of the CFETR advanced divertor optimization by SOLPS-ITER modeling with full drifts and currents activated are presented. Three divertor geometries, which differ by the distance from the X-point to the strike point on the outer target, are considered. Argon (Ar) and neon (Ne) are compared as seeded impurities. It is demonstrated that for all three geometries and for both radiators it is possible to achieve acceptable divertor heat loads (below 5 MW m−2) without notable fuel dilution (Z eff < 2.5). Impurity compression in divertors and pedestal radiation are compared for two gases. Similar core plasma and divertor conditions, as well as radiated power fraction, may be achieved with 2–3 times less Ar seeding rate than the Ne one. Estimated radiation from the confined region appears to be small compared to the exhaust power. However, in all modeling cases the T e at the far scrape-off layer part of both targets remains significantly above 5 eV, which might cause tungsten (W) sputtering. Further optimization of target shape will be performed to reduce the electron and ion temperature.

Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis

PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.

Critical current degradation behavior in Bi-2212 round wires under cyclic transverse stress

Bi-2212 multi-core round wires without anisotropy can be made into Rutherford cable and cable-in-conduit conductor (CICC) using traditional cable winding technology, which has been widely used in high-field magnets such as fusion reactors. Recently, Bi-2212 is considered to be one of the most promising potential superconducting magnet materials in the Chinese Fusion Engineering Test Reactor (CFETR) magnet. However, Bi-2212’s ceramic structure and Ag/Mg alloy sheath make it susceptible to stress. Due to the complicated working environment of magnet, Bi-2212 round wire is bound to be affected by cyclic electromagnetic stress and thermal stress. Several researchers have reported the performance of Bi-2212 round wire under axial tension strain and tension fatigue. And the impact of indentation on the critical current of Bi2212 round wire has been studied. However, the study of transverse stress characteristics has not been reported. In this work, critical current degradation under cyclic transverse stress for Bi-2212 round wires prepared by different heat treatment processes was studied. The initial critical current Ic0 at 77 K for sample sintered at overpressure with annealing was the highest and Ic0 for sample sintered at atmospheric pressure without annealing was the lowest. The greater transverse fatigue stress will accelerate the process of fatigue.

Radiation shielding design of the CFETR polarimeter interferometer and CO2 dispersion interferometer

A three-wave based laser polarimeter/interferometer and a CO2 laser dispersion interferometer are used to determine the electron and current density profiles on a Chinese fusion engineering test reactor (CFETR). Radiation shielding is designed for the combination of polarimeter/interferometer and CO2 dispersion interferometer. Furthermore, neutronics models of the two systems are developed based on the engineering-integrated design of CFETR polarimeter/interferometer and CO2 dispersion interferometer and the major material components of CFETR. The polarimeter/interferometer and CO2 dispersion interferometer’s neutron and photon transport simulations were performed using the Monte Carlo neutral transport code to determine the energy deposition and neutron energy spectrum of the optical mirrors. The energy depositions of the first mirrors on the polarimeter/interferometer are reduced by three orders with the whole shielding. Since the mirrors of CO2 dispersion interferometer are very close to the diagnostic first wall, shielding space is limited and the CO2 dispersion interferometer energy deposition is higher than that of the polarimeter/interferometer. The dose rate after shutdown 106 s in the back-drawer structure has been estimated to be 83 μSv h−1 when the radiation shield is filled in the diagnostic shielding modules, which is below the design threshold of 100 μSv h−1. Radiation shielding design plays a key role in successfully applying polarimeter/interferometer and CO2 dispersive interferometer in CFETR.

Conceptual design and optimization of an ITER-type ICRF antenna on CFETR

The Chinese Fusion Engineering Testing Reactor (CFETR) plans to use an ITER-type antenna and couple ∼30 MW ion cyclotron range of frequencies (ICRF) power to the plasma. In this paper, the physical design of a CFETR antenna in the midplane port is carried out. Parameter scans were performed to study the optimized toroidal and poloidal numbers of straps as well as the optimized geometric sizes of the straps. The coupling resistance, power spectrum, maximum voltage in the resonant transmission line of the strap and parallel electric field in the antenna vicinity are used to determine the performance of the studied antennas. It is shown that four poloidal substraps (i.e. quadruplets) and six toroidal strap columns arranged in half of the antenna allows the antenna to have the best coupling capability. To improve the coupling capability of the proposed antenna model, local gas puffing methods, as well as various antenna phasings, are studied. It is indicated that the coupling resistance can be increased by a factor of three for all studied antenna phasings when applying the midplane gas puffing with a gas puff rate in the order of 4 × 1023 el s-1. The toroidal phasings suitable for heating include (0, pi, 0, pi, 0, pi) and (0, pi, pi, 0, 0, pi).

Neutronics analyses of COOL blanket for CFETR

In this study, an advanced liquid blanket based on a liquid PbLi technology is proposed, and it is called a supercritical CO2 (S-CO2) cooled Lithium-Lead (COOL) blanket. As a candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), it aims to achieve high tritium breeding and shielding performance. It uses S-CO2 as the coolant for the structural components and PbLi as the tritium breeder and neutron multiplier. To evaluate the neutronics performances of this blanket, a detailed 22.5° three-dimensional neutronics model of CFETR was built using the cosVMPT conversion tool. To enhance the shielding capability of the blanket, a shielding block was designed and added at the back of the manifold to moderate neutrons. Evaluation results of the neutron damage of toroidal field coils (TFC) components confirmed that the blanket meets shielding requirements. In addition, the main neutronics parameters including neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heat, and irradiation damage, were calculated using MCNP5. The results indicated that TBR was 1.205, and the total nuclear heat was 1191.64 MW under the fusion power of 1.5 GW. The peak NWL was 1.97 MW/m2, and the maximum DPA and gas production were 14.80 DPA/FPY, 132.65 He appm/FPY, and 631.30 H appm/FPY.

Overview of the CFETR remote handling system and the development progress

Due to the extreme environment inside the vacuum vessel of Chinese Fusion Engineering Test Reactor (CFETR), the maintenance of core components after shutdown cannot be carried out by human operator or general equipment at the maintenance site. A remote handling (RH) system, which uses the radiation-enhanced automation equipment and is controlled remotely by operators, is adopted. The engineering design of the CFETR RH system has been conducted in the past few years with a novel maintenance strategy of transferring the blanket and divertor components through the vertical upper port. A heavy-load robotic manipulator is illustrated, entering the CFETR vessel through an equatorial port to implement in-vessel maintenance tasks with exchangeable end-effectors and support blanket and divertor RH system. An integrated supervisory control system is being developed to provide a safe, reliable, and user-friendly control interface. Finally, despite all the progress that the CFETR RH team has made, it can be said that the CFETR RH system still faces challenges considering the hostile environment as well as the complex and heavy load maintenance process.