Chinese Fusion Engineering Test Reactor Research Articles

Design and construction of a helium cooling experimental loop

The helium cooling tritium breeding blanket and helium cooling divertor, which produce tritium or heat, are important components in Chinese Fusion Engineering Test Reactor (CFETR) design. The helium cooling system, which function is similar to the primary loop of Pressurized Water Reactor (PWR), is designed to extract energy from blankets and divertors. These helium cooling components have many complex internal flow channels to extract high thermal load. They also have to withstand electromagnetic force, thermal stress and coolant pressure, etc. Therefore, out-of-pile experiments must be carried out as far as possible to verify the design. Under CFETR project support, a helium cooling experiment loop is built to carry out the thermo-hydraulic experiments for these components and to accumulate experience of helium cooling system. According to the cooling requirements of blanket, the loop test section operating parameters are flow rate 2.5kg/s, temperature 550°C and pressure 12MPa. During preliminary commissioning, these parameters have been reached. A high heat flux testing facility will connect to this loop for future experiments.

Corrosion resistance of 9–12Cr ODS steels in pressurized water for fusion application

The blanket design of Chinese fusion engineering test reactor (CFETR) may utilize pressurized water as coolant. The fusion structural materials will endure high temperature, high pressure and result in corrosion in the pressurized water. It is necessary to verify corrosion resistance of the materials. In the present study, three kinds of advanced fusion structural materials with excellent mechanical properties and neutron irradiation resistance, i.e., 9Cr-ODS steel, 9CrODS-Al steel, and 12CrODS-Al steel were immersed in pressurized water at 325 °C under the pressure of 15.5 MPa up to 859 h. Room-temperature tensile properties were investigated for the materials before and after corrosion. The oxidation layers of the steels were analyzed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM). The ODS steels have good corrosion resistance in the pressurized water, especially the 9CrODS-Al and 12CrODS-Al steels. The outer oxide layer is identified as Fe2O3 for all the steels. The inner layer is mainly FeCr2O4 for the 9CrODS steel, and Fe(Al,Cr)2O4 for the ODS steels with Al element which better protect the steels from further corrosion. The relevant properties data can provide reference for the blanket materials choice of CFETR and the future fusion reactors.

Inverse kinematics for the CMOR in the CFETR based on workspace analysis: A novel approach

This paper delves into the intricate field of inverse kinematics in the context of the Chinese Fusion Engineering Test Reactor (CFETR) and its specialized robot, the multi-purpose overload robot (CMOR). Nuclear fusion, a cornerstone for sustainable energy, presents immense potential with benefits like high energy output and minimal environmental impact. The goal of the CFETR is to transform nuclear fusion energy from theory to practical solutions to solve the global energy crisis. The CMOR, with nine degrees of freedom, is pivotal in maintaining in-vessel components damaged by intense radioactive exposure. This study introduces a novel approach to solve the inverse kinematics of the CMOR based on the CFETR's work environment. It involves dividing the working space, identifying target interval based on coordinates, and employing intelligent algorithms for precise and efficient kinematic resolution. Six algorithms were evaluated, highlighting the proposed approach's superior computational speed and accuracy, crucial for real-time monitoring and control in CFETR operations. This work presents a significant advancement in redundant robotic kinematics in nuclear fusion environment.

Magnetic diagnostics layout design for CFETR plasma equilibrium reconstruction

Plasma equilibrium reconstruction provides essential information for tokamak operation and physical analysis. An extensive and reliable set of magnetic diagnostics is required to obtain accurate plasma equilibrium. This study designs and optimizes the magnetic diagnostics layout for the reconstruction of the equilibrium of the plasma according to the scientific objectives, engineering design parameters, and limitations of the Chinese Fusion Engineering Test Reactor (CFETR). Based on the CFETR discharge simulation, magnetic measurement data are employed to reconstruct consistent plasma equilibrium parameters, and magnetic diagnostics’ number and position are optimized by truncated Singular value decomposition, verifying the redundancy reliability of the magnetic diagnostics layout design. This provides a design solution for the layout of the magnetic diagnostics system required to control the plasma equilibrium of CFETR, and the developed design and optimization method can provide effective support to design magnetic diagnostics systems for future magnetic confinement fusion devices.

UEDGE modeling of plasma detachment of CFETR with ITER‐like divertor geometry by external impurity seeding

AbstractEfficient handling of high heat flux on the plasma‐facing components, particularly the divertor targets, poses a significant challenge for the Chinese Fusion Engineering Testing Reactor (CFETR) with fusion power of Gigawatt. This work investigates the divertor plasma detachment of CFETR with a standard ITER‐like divertor geometry by neon (Ne) or argon (Ar) impurity seeding using UEDGE code. The cross‐field drifts terms are switched off, and fluid neutral models and a “fixed‐fraction” impurity model are applied to enable efficient simulations for the study of CFETR detachment. In order to reduce the heat load on the divertor targets below the acceptable level (<10 MW/m2), the impurity fraction (f), pumping speed (S), and upstream density are varied to identify the suitable operations window during Ne seeding. The effects of Ne and Ar impurities on the plasma detachment are compared. It is found that with the power across the core‐edge interface PSOL = 200 MW and separatrix density of 2.8 1019 , Ne impurity fraction ≥1.7%, and Ar impurity fraction ≥0.24% can achieve the partial detachment. Achieving similar total radiation power (˜148 MW), the Ne fraction is 2.3% and the Ar fraction is 0.24%. Moreover, the simulation results indicate that Ar exhibits better power radiation efficiency and core compatibility compared with Ne.

Development and verification of tritium transport code based on RELAP5/MOD3.3 with generic model towards COOL blanket

Tritium evaluation is a critical issue in fusion reactors. For the tritium self-sufficiency and safety concern, it is crucial to develop tritium analysis techniques to study its behavior in various conditions including steady states and dynamic states. The supercritical CO2 cOoled Lithium-Lead (COOL) blanket is an advanced blanket concept for the Chinese Fusion Engineering Testing Reactor (CFETR) and one of its major functions is to realize tritium self-sufficiency. Therefore, it is necessary to conduct comprehensive analyses for design and optimization from the perspective of tritium transport. In this work, the tritium transport module is integrated inside RELAP5/MOD3.3. Then the modified RELAP5 facilitates transient analyses to examine coupled thermal hydraulics and tritium transport responses. The code includes 1-D tritium permeation, tritium transport in fluid and tritium extraction, which are associated with thermodynamic parameters to cover the multi-physics processes of the COOL blanket. The modified RELAP5 is verified by the code-to-code benchmark. The results demonstrate the successful integration of the tritium transport capabilities in RELAP5. Furthermore, the code is applied to the heat exchanger and simplified PbLi loop to obtain tritium concentration to highlight its capability in both generic modeling and dynamic analysis.

Preliminary electromagnetic analysis of the COOL blanket for CFETR

The supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been designed as one advanced blanket candidate for the Chinese Fusion Engineering Test Reactor (CFETR). This work focuses on the electromagnetic (EM) loads (Maxwell force and Lorentz force) acting on the COOL blanket, which are important mechanical loads in further structural analysis of the COOL blanket. A 3D electromagnetic analysis is performed using the ANSYS finite element method to obtain EM loads on the COOL blanket in this study. At first, the magnetic scalar potential (MSP) method is used to obtain the magnetic field and the Maxwell force on the COOL blanket. Then, the magnetic vector potential (MVP) method is performed during a plasma disruption event to get the eddy current distribution. At last, a multi-step method is adopted for the calculation of the Lorentz force and the torque. The maximum Lorentz forces of inboard and outboard blanket structural components are 5624 kN and 2360 kN respectively.

The dynamic factor particle swarm optimization (DFPSO) to inverse kinematics of CMOR based on pose decomposition strategy

The Chinese Fusion Engineering Testing Reactor (CFETR) is a superconducting Tokamak device, whose components would get gradually damaged by being susceptible to high temperature, strong magnetic field, radiation, and so on, and should be maintained regularly. The CFETR Multipurpose Overload Robot (CMOR) is a redundant robot designed as a macro-micro arm structure for a gigantic workspace in the blanket and diverter of CFETR's maintenance. Due to the macro-micro arm structure, the robot has a high degree of freedom, and its serial DOF is up to 16 so the traditional algorithm is low efficient for the robot's kinematics. Hence, in this paper, a kinematic decomposition strategy is proposed to allocate or decouple the position aim. Besides, a novel particle swarm optimization (PSO) with dynamic accuracy called dynamic factor particle swarm optimization (DFPSO) is designed to accelerate solving the high-dimensional nonlinear kinematic problem and to improve accuracy. The results of the research show the validation and efficiency of DFPSO in the CMOR system's kinematics. Moreover, it's also a universal kinematic algorithm for all high-DOF robots.

Design and research of bilateral teleoperation for CFETR tile assembly using haptic feedback

Bilateral teleoperation is a control strategy used in remote handling (RH) to enable an operator to control a robot or system while perceiving the interactive force in the environment. To enhance the immersive operating experience by reducing time delays, a control system based on OROCOS and DDS has been presented to rapidly establish the bilateral teleoperation system with real-time performance. Additionally, a bilateral control algorithm combining closed-loop pose tracking and force feedback is proposed to improve the pose tracking accuracy of the slave robot and enhance the sense of telepresence. As a part of the pre-study on the tile assembly conducted by CMOR in Chinese Fusion Engineering Test Reactor (CFETR), a lightweight experimental setup has been constructed based on the maintenance scenario in EAST. Relative experiments have been conducted to validate the accuracy and effectiveness of the bilateral teleoperation system. The results demonstrate low latency and smooth control during the bilateral teleoperation process.

Design and Optimization of a New Heterogeneous Printed Circuit Plate Heat Exchanger With Molten Salt Zigzag Passage and Supercritical CO2 Airfoil Fin Passage

Abstract In the fusion power conversion system, a printed circuit heat exchanger (PCHE) between molten salt (MS) and supercritical carbon dioxide (sCO2) transfers huge heat between loops. To improve heat transfer efficiency, a new heterogeneous PCHE with MS zigzag passage and sCO2 airfoil fin passage was proposed. A one-dimensional simulation of the new PCHE was conducted to study the effects of the plate number and the length on its pressure drop, MS mass flowrate, capital cost, and operating cost. Then, a new single objective optimization of the total cost was performed by the genetic algorithm (GA) based on the Chinese Fusion Engineering Testing Reactor (CFETR) parameters. Finally, the new optimal PCHE was compared with the PCHE with MS straight passage and sCO2 airfoil fin passage. The results show that the length and the plate number of the PCHE have an important effect on the pressure drop and its cost. The optimal geometry scheme with the minimum cost is given for the application to CFETR. By comparison with the MS straight-passage PCHE, it is found that the total cost of the new PCHE is reduced by 5.7% and the volume of the heat exchanger is reduced by 10.7%.

Optimization of welding distortion of vacuum vessel for nuclear fusion based on finite element analysis

The vacuum vessel (VV) is a crucial part of the Chinese Fusion Engineering Test Reactor (CFETR). 1/8 VV is comprised of two identical 1/16 VVs with symmetrical properties. Welding at a high density between the sectors will affect the overall profile of the structure and may fail to meet the required fit accuracy. In addition, excessive distortion will influence the subsequent assembly of the VV and the degree of the mounting position of the internal components, which cannot be analyzed directly through experimentation due to the scale of the components. Therefore, before welding, a theoretical analysis must be performed to enhance the welding process and reduce welding distortion. In this investigation, welding distortion of plate butt joints in 1/8 VV was simulated using finite element analysis and experimentally confirmed. Local and global welding sequences and welding constraints were studied utilizing distortion curves and clouds. By optimizing the welding sequence, the overall average distortion can be reduced by approximately 13%, and by applying internal constraints, the average distortion can be reduced by approximately 18%, with the greatest change occurring in transverse distortion, which is reduced by approximately 35%. This paper is intended to provide data support and theoretical guidance for the actual processing of CFETR VV using finite element analysis.

Study on the effect of neutron shielding design on tritium breeding based on water-cooled ceramic blanket for CFETR

The Chinese Fusion Engineering Test Reactor (CFETR) is a magnetic confinement fusion reactor independently designed and developed by China, which is based on the ITER design experience. As one of the candidates for the CFETR, the water cooled ceramic blanket (WCCB) mainly carries out many important tasks, such as tritium breeding and radiation shielding. The high tritium breeding ability is one of the most significant goals of the blanket. For the design of the CFETR, in addition to meeting the requirements of tritium breeding, the design of the blanket must also consider meeting relevant shielding limits. Due to the limited space of the blanket, the improvement in tritium breeding space will inevitably lead to the reduction in neutron shielding space behind the breeding area, and vice versa. Therefore, under the premise of meeting the requirements of neutron shielding, increasing the tritium breeding space as much as possible is the focus of research. In this work, a three-dimensional neutronics model containing the WCCB blanket is developed, and the neutronics performance is calculated based on 1.5 GW fusion power. A set of nuclear analyses are carried out by the MCNP code, including analysis of the neutron wall load, tritium breeding ratio (TBR), and fast neutron fluence of the TF coil. It is found that the shielding space of certain blanket modules could be optimized. After the shielding optimization, the global TBR increased from 1.168 to 1.186, an increase of 0.018 TBR. The current research has important guiding significance for the future design and optimization of the WCCB for the CFETR.

Preliminary neutronics/thermal-hydraulics design and analyses of the helium-cooled mixed bed breeder blanket for CFETR under 1.5 GW fusion power

Chinese Fusion Engineering Test Reactor (CFETR) is a tokamak device designed with different power levels to validate fusion engineering technologies in China. Helium-cooled ceramic breeder (HCCB) blanket and Water-cooled ceramic breeder (WCCB) blanket are the main candidates. Compared with WCCB, HCCB generally has a higher tritium breeding ratio (TBR) because helium has smaller neutron absorption cross sections than water. However, the TBR of the HCCB decreases with the increase of the fusion power. In this paper, the preliminary neutronics and thermal-hydraulics design of the Helium-Cooled Mixed bed Breeder (HCMB) blanket is proposed. The blanket structure with a fusion power of 1.5 GW is designed based on the current HCCB design of the CFETR. The HCMB adopts helium as the coolant and the mixed Li4SiO4/Be12Ti as the tritium breeder. The neutronics and thermal-hydraulics performances of HCMB blanket have been analyzed in the paper. The results show that the HCMB blanket design has a TBR up to 1.25 with the thermal-hydraulics requirements satisfied.

Analyses and optimization of the CFETR power conversion system with a new supercritical CO2 Brayton cycle

In this paper, the Chinese Fusion Engineering Testing Reactor (CFETR) power conversion system, with a supercritical CO2 (SCO2) Brayton cycle, is designed, analyzed and optimized. Considering the pulse operation of the reactor, a heat storage loop with high temperature molten salt and low temperature concrete is introduced. Based on the parameters of the first cooling loop, the CFETR power conversion loop is designed and studied. A new SCO2 Brayton cycle for the CFETR dual heat sources, blanket and divertor, is developed and optimized using a genetic algorithm. Compared to other simple and recompression cycles, it is shown that the new SCO2 Brayton cycle combines maximum thermal efficiency with simplicity. Exergy analyses are carried out and show that the exergy destruction rates of turbine and heat exchangers between different loops are the largest due to the large turbine power and the large temperature difference. The exergoeconomic analyses show that the fusion reactor accounts for the main cost, which is the key to the economy of fusion power generation. The following sensitivity analyses show that the hot molten salt temperature has a major influence on the system performance. Finally, several multi-criteria optimization algorithms are introduced to simultaneously optimize the three fitness functions, the cycle thermal efficiency, the system exergy efficiency and the total system product unit cost. It is found that the maximum thermal efficiency, the maximum exergy efficiency and the lowest total system product unit cost can be obtained almost simultaneously for the new CFETR power conversion system, and this optimal operation scheme is presented.

An Approach to Increase the Fusion Tritium Breeding Ratio Using Temperature Field Balance Based on Water-Cooled Ceramic Blanket

An advanced breeding blanket for excellent tritium breeding performance shall be designed to realize the operations of Chinese Fusion Engineering Testing Reactor (CFETR) at steady-state burning tritium fuel in the core plasma for fusion reactions. Tritium release from blanket interior effectively requires a temperature range above than 400 °C. This indicates that the blanket temperature field distribution and its evolution are related to the online tritium breeding ratio (TBR) assessment. In this article, water-cooled ceramic blanket (WCCB) models are built to perform the TBR estimation and the temperature field simulations. It is confirmed that the temperature field affects the TBR variation, which is linked to the blanket cooling pipes (CPs) layout. In addition, the effective volume fraction of tritium release of the blanket model is defined to evaluate the breeding regional proportion whose temperature is higher than 400 °C. The effective volume fraction of tritium release is analyzed for the inlet coolant temperature ranging from 0.6 to 5.0 m/s by an Eulerian-Eulerian two-phase numerical simulation method. The outer surface temperature of CPs can be raised toward 400 °C when inlet velocity is 1.1 m/s. This is a vital approach to reduce the deep temperature gradient in blanket interior, which is beneficial to the tritium release sufficiently. The estimation of critical heat flux (CHF) of first wall (FW) indicates that boiling crisis will not take place in FW. This work is helpful to the engineering design of fusion blanket considering the tritium self-sufficiency in the road toward fusion reactor.

Numerical Simulation of Airflow and Dust Migration in Vacuum Vessel During Loss of Vacuum Accident for Updated CFETR

A loss of vacuum accident (LOVA) occurs during in-vessel component failure and air ingress. The airflow characteristics of a LOVA are determined by many factors like initial pressure, location of a break, and size of a break and have a great impact on dust migration, which could cause a serious explosion with incoming air and H2. In this paper, a computational fluid dynamics method is adopted, and the k-ε Shear Stress Transport model for airflow and the Discrete Phase Model for dust are used to simulate a LOVA with the updated Chinese Fusion Engineering Test Reactor (CFETR) tokamak device. The effects of initial pressure, break size, and break location on airflow during the LOVA are discussed, and the effects of dust size, break size, and break location on dust migration during the LOVA are investigated as well. The results indicate that the initial pressure and size of a break have a greater impact on airflow of a LOVA than the location of the break and that both the dust size and the characteristics of the airflow have a greater impact on the distribution of the dust. A break located in the upper port has even more dust chaos. This research is the basis for the safety analysis of the CFETR device, and it provides a reference for subsequent studies on dust removal, mitigation of dust explosions, and radioactive substances.

Pre-conceptual studies of a travelling wave array antenna for the EAST

One of the primary problems of ion cyclotron range of frequency (ICRF) systems in magnetic confinement experiments is the coupling of a large amount of radio frequency wave power through the plasma cut-off layer within the voltage limits of the antenna system. Travelling wave array (TWA) antennas have higher coupling than conventional ICRF antennas, which is manifested in a sharp and optimized k // spectrum. As a pre-study of TWA applications in tokamaks, a TWA antenna with six consecutive straps and double-fin capacitors was conceptually designed for the experimental advanced superconducting tokamak (EAST). The antenna geometry was optimized to seek a low reflection coefficient for EAST ICRF heating scenarios. The design and simulation results of the TWA antenna are briefly presented. The results of the frequency sweep in vacuum show that a bandwidth of approximately 3 MHz with S 11 < −30 dB can be obtained. The peak of the k // power spectrum is adjusted to ∼3–4 m−1 at the frequency of 34–36 MHz. In addition, the properties of the power flow and the characteristics of the wave field are also discussed by modelling the plasma facing the TWA antenna using a cold plasma medium. The results in this study may provide some reference and guidance for the study of TWA antennas and other ICRF antennas in magnetic-confined fusion devices like EAST or the Chinese Fusion Engineering Test Reactor.

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The high-energy &lt;i&gt;α&lt;/i&gt; particles produced by deuterium-tritium fusion are the primary heating source for maintaining high temperatures in future tokamak plasma. Effective confinement of &lt;i&gt;α&lt;/i&gt; particles is crucial for sustaining steady-state burning plasma. The initial energy of &lt;i&gt;α&lt;/i&gt; particles is &lt;inline-formula&gt;&lt;tex-math id="M1"&gt;\begin{document}$ 3.5 {\text{ MeV}} $\end{document}&lt;/tex-math&gt;&lt;alternatives&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_M1.jpg"/&gt;&lt;graphic xmlns:xlink="http://www.w3.org/1999/xlink" xlink:href="21-20230700_M1.png"/&gt;&lt;/alternatives&gt;&lt;/inline-formula&gt;. According to theoretical calculations, it takes approximately 1 second to slow down &lt;i&gt;α&lt;/i&gt; particles through Coulomb collisions to an energy range similar to the energy range of the background plasma. In the slowing-down process, some &lt;i&gt;α&lt;/i&gt; particles may be lost owing to various transport processes. One significant research problem is how to utilize &lt;i&gt;α&lt;/i&gt; 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 &lt;i&gt;α&lt;/i&gt; particles can be derived. However, considering the substantial drift orbit width of &lt;i&gt;α&lt;/i&gt; particles and the importance of spatial transport, numerical calculations are required to obtain more accurate &lt;i&gt;α&lt;/i&gt; particle distribution function. In this study, the particle tracer code (PTC) is used to numerically simulate the slowing-down process of &lt;i&gt;α&lt;/i&gt; 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 &lt;i&gt;α&lt;/i&gt; 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 &lt;i&gt;α&lt;/i&gt; particles at these energy levels. The research findings hold profound implications for the precise evaluating of ability of &lt;i&gt;α&lt;/i&gt; particles to heat the background plasma. Understanding and characterizing the behavior of &lt;i&gt;α&lt;/i&gt; 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 &lt;i&gt;α&lt;/i&gt; 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 &lt;i&gt;α&lt;/i&gt; 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.