China Fusion Engineering Test Reactor Research Articles

Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept

For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum vessel (VV). The break leads to helium inject into the VV. After the accident occurs, the helium release leads a plasma disruption and pressure increase in the VV. This paper presents the analysis results of the in-vessel LOCA accidents in CFETR and gives the accident sequence compare with the break in the FW and header pipe. The accident analysis and influence compares can provide a reference for the iterative design.

3D modeling of a double-driver ion source considering ion magnetization: an investigation of plasma symmetry modulation methods

A three-dimensional fluid model of a double-driver negative hydrogen ion source for China Fusion Engineering Test Reactor (CFETR) neutral beam injection is developed. In this model, the magnetic filter field is generated by 16 permanent magnets, which are surrounded by a soft iron. In order to accurately describe the transportation of charged species in the presence of strong magnetic field, both the electron magnetization and ion magnetization are taken into account, and the accuracy of the model has been proved by comparison with experimental data. By employing this model, the spatial distributions of the plasma parameters have been investigated, and three methods are proposed to optimize the symmetry at the bottom of the expansion region of a double-driver source. The results indicate that by adjusting the power of Driver I while keeping the power of Driver II constant, the symmetry of the electron density and negative hydrogen ion density could be improved. Furthermore, the inclusion of partition improves the symmetry of the electron temperature and density but has no impact on the regulation of the negative hydrogen ion density distribution. Finally, the application of magnetic shield can not only improve the symmetry of the electron density and negative hydrogen ion density, but also increase their densities at the bottom of the expansion region.

Mechanical and physical properties of modified N50 steel at cryogenic temperatures

China Fusion Engineering Test Reactor (CFETR) has more advanced design parameters compared to the International Thermonuclear Experimental Reactor (ITER), particularly demanding a high yield strength (over 1500 MPa at 4.2 K) for jacket material in the cable-in-conduit conductor (CICC). Modified N50 austenitic steel, due to its excellent mechanical properties at liquid helium temperature, has been identified as a promising candidate material for the jacket material. However, while some research focuses on the mechanical properties of modified N50 steel at cryogenic temperatures, little is known about the cryogenic physical properties that are critical for conduit jacket applications. In this study, a modified N50 steel was prepared and characterized in terms of tensile properties and physical properties at cryogenic temperatures. We tested the tensile properties of the modified N50 at 4.2 K, 77 K, and room temperature (RT). Additionally, we measured the thermal conductivity, thermal expansion, specific heat capacity, and magnetization of the modified N50 from 4 K to 300 K. These results could provide a more comprehensive reference for applying the modified N50 steel in jacket materials for the CFETR.

An extension of NASCA for SDDR analysis of mobile activated components with global variance reduction

During the plasma operation period of fusion reactor facilities, neutrons generated by DT reaction can penetrate deeply into structures and induce strong activation. Highly activated components have to be transferred from the on-load position to the hot cell for the purpose of maintenance. This movement produces a high-dose field, originating from the mobile radiation source, which could affect the functionality of important devices or pose risks to the workers. Therefore, it is imperative to compute the shutdown dose rate (SDDR) maps while transporting activated components in the facility and develop an optimized solution for this calculation process. In this study, we established a new extension of the NASCA method combined with the OTF GVR method to compute SDDR maps. NASCA-OTF is a method developed based on the shutdown dose simulation code NASCA to ensure automatic modification of the geometry structure on the fly. In the meantime, it uses OTF method for speeding up the Calculation. Thus, this modified method produces convergent SDDR maps corresponding to different positions of activated components within a single Monte Carlo (MC) simulation. The NASCA-OTF was verified by comparison with R2S method through benchmark analysis. The method was also used for SDDR analysis of the mobile blanket sector for the China Fusion Engineering Test Reactor (CFETR), and the results indicated that the NASCA-OTF is feasible for computing the SDDR resulting from the movement of activated components.

Engineering design and testing of rotary joints’ driving systems for the CFETR multi-purpose overload robot

This study focuses on the engineering design and testing of the rotary joints' driving systems for the China Fusion Engineering Test Reactor (CFETR) Multi-Purpose Overload Robot (CMOR). CMOR is a critical component in the CFETR. The CFETR tokamak serves as an engineering test reactor in China for magnet confinement fusion research. CMOR, a 7-DOF manipulator, is specifically designed for internal maintenance within the CFETR. This poses significant challenges in the design of the joint driving systems. The joint types of CMOR are categorized into three, comprising one translation joint, two swing joints, and four rotary joints. Using rotary joint J6 of CMOR as an example, this paper provides a detailed exploration of the engineering design and testing of the rotary joints' driving systems, offering a comprehensive explanation of their design specifications and principles. Specification and strength assessments are conducted on the engineering design results, meeting all requirements. Furthermore, the paper delves into the testing methods of the driving systems, including the assessment of transmission error and backlash error, two crucial indicators for the driving system. This research contributes valuable insights for future joint designs and testing of the CMOR heavy-load manipulator.

Neutronics analysis of the ion cyclotron resonance heating antenna of the China Fusion Engineering Test Reactor

Ion cyclotron resonance heating (ICRH) is an important auxiliary heating method applied to the China Fusion Engineering Test Reactor, which can effectively heat the ions and electrons in plasma. Owing to the harsh nuclear environment, neutronic analyses are required to verify tritium self-sufficiency and neutron-shielding requirements. In this study, a neutronics analysis of the ICRH antenna was conducted using the COre and System integrated engine for Reactor Monte Carlo (cosRMC) code to estimate the neutron flux, radiation damage, nuclear heating, gas generation rate of key components, and tritium breeding ratio (TBR), providing data support for the subsequent optimization of the shielding design. In addition, the neutron flux of the coils around the antenna was calculated to prevent the entry of neutrons that damage the magnetic field coils through the gaps between the port plugs and antenna, and the shielding effects of the port-plug antenna on the surrounding components were analyzed. Finally, the results obtained using the cosRMC and MCNP codes were compared, which and presented good agreement, thus verifying the reliability of the neutronic analysis using the cosRMC code.

Impact of local plastic deformation on thermo-magnetic instability of high-J c Nb3Sn strand

The future fusion reactor devices built by Institute of Plasma Physics, Chinese Academy of Science will be a compact burning plasma facility to fill the gap between ITER and the China Fusion Engineering Test Reactor, with the mission of studying the deuterium–tritium plasma in steady-state operation. The winding-package (WP) of its toroidal field coil is graded into high-field WP and low-field WP, and the high critical current density (J c) Nb3Sn strand will be applied to the high-field WP based on the current design. The thermo-magnetic instability is the main issue in the application of high-J c Nb3Sn strand, which may induce premature quenching in the low field regions. This issue can be improved by reducing the effective filament diameter (d eff) to suppress flux jumps and increasing the residual resistivity ratio (RRR) to enhance the release of heat generated by flux jumps. However, in the conductor manufacturing process, local plastic deformations (indentation) of strands can impact the sub-element layout and degrade the RRR of the strand, which would increase again the risk on instability. In this study, magnetization measurements and V–I tests were performed on samples with different indentation depths. The hysteresis loss and d eff of the indented sample was obtained from the test results. The impact of local plastic deformation on the thermo-magnetic instability of two types of Nb3Sn strand was confirmed by cross-sectional metallographic observation and quantitative microstructure analysis. It was shown that flux jumps were suppressed at indentation depths below 0.3 mm. Further indentation increase leads to severe flux jumps.

Development of thermo-physical property database and code of fusion materials for CFETR blankets

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap.

Design and analysis of radio frequency window for the China Fusion Engineering Test Reactor ion cyclotron range of frequency heating system.

The Ion Cyclotron Range of Frequency (ICRF) heating system of the China Fusion Engineering Test Reactor (CFETR) is intended to provide plasma heating with a minimum power output of 20MW, which demands the Radio Frequency (RF) window to possess a higher performance requirement. This paper presents the design of an RF window for the CFETR ICRF heating system and focuses primarily on the design and confirmation of its electromagnetic performance. The RF window can be effectively matched in the operating frequency range and has an S11 of under -59 dB. The geometry of the cone type ceramics was optimized to reduce the surface tangential electric field distribution. An analysis of the electric field distribution of the RF window at 50kV indicates that the pressure side was below 2.3 kV/mm and the vacuum side was below 1.3 kV/mm. Furthermore, a transmission line test bench with an open-terminated setup was constructed to conduct withstand voltage tests on the mockup, and the results showed that the mockup could withstand 62kV for 2s and 47kV for 120s.

Comparison of electron-beam-welded joints manufactured by rolled and selective laser melted reduced activation ferritic/martensitic steel

As the core component of China Fusion Engineering Test Reactor (CFETR), the forming and manufacturing of the first wall faces severe challenges due to the constraints of materials, service environment and geometric structure. The microstructure and mechanical properties of welded joints obtained by rolled and laser selective melted (SLMed) reduced activation ferritic/martensitic (RAFM) steel were investigated in depth. The microstructure analysis results showed that the weld metal (WM) was composed of coarse lath martensite and δ-ferrite, while the microstructure were mainly coarse grains in heat affected zone (HAZ) which was easy to break. The average grains size of SLMed RAFM steel (S-RAFMs) was smaller than that of rolled RAFM steel (R-RAFMs), since the rapid melting and solidification of laser in the additive manufacturing process prevented grains to grow. The microstructure of R-RAFMs joint was relatively more uniform and there was no obvious texture and preferred orientation, while the WM of S-RAFMs joint had a weak cubic texture. The results of mechanical properties showed that the S-RAFMs joints exhibited higher strength (UTS = 697.46 MPa) and the R-RAFMs joints exhibited better plasticity (EL = 13.49 %) under the same tensile conditions. The tensile and impact fracture morphologies showed that the fracture mechanism of two joints was ductile-brittle mixed fracture, and the dimples of R-RAFMs joints were finer and denser. It was clearly observed that the hardness of WM (∼390–430 HV0.3) was higher than that of the base metal (BM) (∼190–230 HV0.3) from the hardness distributions along the width directions. The average hardness of BM in R-RAFMs joint (227 HV0.3) was lower than S-RAFMs joint (197.3 HV0.3), which was mainly related to the fine grain strengthening effect during SLM forming. The above results may provide theoretical guidance for the high-performance manufacturing of the first wall and promote the application of SLM and EBW technology in the field of nuclear fusion.

Load study and structural analysis for CFETR divertor OVT

China Fusion Engineering Test Reactor (CFETR) is an important device for studying fusion reactor engineering in China. The Outer Vertical Target (OVT) are high heat load components of the divertors and work in a strong electromagnetic field, so they need sufficient heat removal capability and structural strength. In addition, due to activation of neutrons on materials, the OVT also needs to meet the requirements of Remote Handling (RH) maintenance. CFETR puts forward the concept of hybrid divertor-blanket, which cancels the cassette and adopts a separate structure. The lower part of the divertor is occupied by the breeding blanket, and the divertor target is fixed on the breeding blanket, to improve the RH maintenance efficiency and Tritium Breeding Ratio (TBR). In previous work, the structural design and thermal-hydraulic analysis of the OVT were completed. In this paper, load study and structural analysis was carried out to verify the structural strength of the OVT under experimental test conditions (including liquid pressure, nitrogen baking), normal operation conditions (including the impact of thermal stress under thermal load, gravity, working pressure, etc.) and extreme event conditions (Vertical Displacement Event (VDE) or Major Disruption (MD) event, including the impact of thermal stress under thermal load, gravity, working pressure, electromagnetic load, etc.). The results show that the OVT meets the requirements of heat removal and strength, and effectively promote the processing and manufacturing of the prototype of the OVT.

Thermohydraulic and thermal fatigue characteristics of W/Cu flat-type microchannel mock-up for CFETR divertor

The divertor of the China Fusion Engineering Test Reactor (CFETR) requires a more efficient heat dissipation structure due to the high demand of heat flux dissipation of up to 20 MW/m2. Microchannel cooling technology excels in thermal management because of large heat transfer area to volume ratio and capability to achieve uniform temperature distribution. However, the heat transfer and thermal fatigue characteristics of W/Cu flat type mock-up with microchannel cooling structures at a heat flux of 20 MW/m2 remains elusive. In this work, we analyzed the thermohydraulic (numerically) and thermal fatigue (experimentally) characteristics of W/Cu flat-type microchannel mock-up. The numerical results show that the thermal loaded area (W) of the microchannel mock-up possesses excellent temperature uniformity and heat transfer performance (the W surface temperature is only ∼685 °C even at a low volumetric flow rate V̇ = 5 L/min). We also compared the heat dissipation efficiency with the hypervapotron cooling channels, and found that the maximum temperature of the W surface and channel wall of the microchannel mock-up are reduced by 18.9 % and 46.7 %, respectively. The thermal stress analysis indicates that the maximum equivalent stress is concentrated in the middle region of the W surface at 15 s, with a value of 357.8 MPa that is below the maximum allowable stress of W material (425 MPa at 650 °C). The as-manufactured W/Cu flat-type microchannel mock-up successfully withstood a thermal fatigue test consisting of 1000 cycles at 20 MW/m2 steady state heat load at V̇ = 30 L/min and Tin = 25 °C. The maximum temperature of W is only 616.2 °C during the test, and some microcracks are detected on the W surface after the test; these microcracks are perpendicular to the heat loading surface, which have negligible influence to the heat transfer of the entire mock-up. This study reveals that the W/Cu flat-type microchannel mock-up is promising in dissipating ultra-high heat flux (steady-state 20 MW/m2) on the divertor target, which might provide valuable insights for the engineering design of the CFETR divertor.

Investigation of alpha-particle transport by Alfvén eigenmodes in CFETR using a simplified diffusion model

The aim of this work is to analyze the alpha particle transport induced by Alfvén eigenmodes (AEs) based on the latest design of China Fusion Engineering Test Reactor (CFETR). Firstly, the stability of AEs is analyzed in order to understand the physical characteristics of AEs in CFETR with latest design parameters. AEs driven by alpha particles are analyzed using the gyrokinetic ion/fluid electron hybrid code GEM. The transport of alpha particles is investigated using a reduced energetic particle transport model based on the resonance broadened quasi-linear model. GEM results show that AEs with many toroidal mode numbers can be destabilized simultaneously by alpha particles in CFETR, and the toroidal mode number of the most unstable mode is . It is found that the excitation threshold of the alpha particle central beta for the mode is substantially below the expected alpha particle beta value of CFETR. All these results indicate that AEs can be driven strongly in CFETR plasmas. Furthermore, simulation results using the reduced transport model show that, due to multiple unstable AEs, the alpha particle density decreases in the core and alpha particles are transported from the core region () to the outer region (). In addition, we find that the minimum value of the safety factor and the central beta value of alpha particle have significant effects on the alpha particle transport, and the AE-induced alpha particle transport level becomes lower with smaller linear growth rates. Finally, it is found that the radial range of the redistributed alpha particle density profile depend on the radial location.

Phenomena Identification and Ranking Table for Thermal-Hydraulic Safety of China Fusion Engineering Test Reactor

The China Fusion Engineering Test Reactor (CFETR) is presently in the engineering design phase, and it is crucial to establish a thermal-hydraulic safety analysis method. First of all, the identification of phenomena in selected accident scenarios affecting reactor safety performance metrics should be focused on with the phenomenon identification and ranking table (PIRT), in which phenomena related to the thermal-hydraulic safety of the vacuum vessel and related systems are screened, including both design-basis accidents and beyond-design-basis accidents. The development of this PIRT exercise for CFETR thermal-hydraulic safety is addressed in this study; specifically, the importance ranking and stage of knowledge (SoK) of phenomena in the selected accidents are evaluated by the PIRT panel. The results of the PIRT analysis revealed that there exist certain safety-critical phenomena for which SoK is relatively low despite their perceived significance in safety performance metrics, which include the phase change of coolant, the migration of the multicomponent gas mixture, the migration of radionuclides within confinement, and the dispersion of gaseous radionuclides in the atmosphere. Ongoing research works and follow-up plans to improve the SoK of phenomena are presented.

Thermal-hydraulic simulation of the CFETR TF coil operation during a plasma scenario: Gandalf-Fluent and 4C code-to-code comparison

Within the framework of the comprehensive research facility for fusion technology (CRAFT) activities, a prototype toroidal field (TF) coil has been designed and will be manufactured at the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), in Hefei. The main purpose is to develop and validate the manufacturing technologies of the large-scale superconducting coils of the China Fusion Engineering Test Reactor (CFETR). To ensure the safe operation of the CFETR device, the Gandalf-Fluent tool and the 4C code are employed to evaluate the performance of the superconducting magnets, and their thermal-hydraulic simulation results are compared with each other. The comparison shows that the thermal coupling between turns and pancakes and the transient model of the thermal coupling to the casing structures, accounted for in 4C but neglected in Gandalf-Fluent, have a significant effect on the temperature profile. Then a detailed reassessment of the minimum temperature margin (ΔTmarmin) is carried out using the 4C code. The results show that the ΔTmarmin is ∼ 1.0 K (if a realistic value for the inter-turn and inter-pancake coupling parameter is assumed), lower than the design requirement of 1.2 K, and it deserves the attention of the designers.

Study on the welding properties of modified N50 CICC jacket for future fusion applications

The modified N50 (N50H) cable-in-conduit conductor CICC jacket developed by China will be applied in China Fusion Engineering Test Reactor (CFETR). Building upon the N50H material, China has developed a corresponding welding material and conducted R&D work on N50H jacket welding. We simulated the N50H jacket welded joint to replicate the entire process of CICC preparation, which includes extrusion, bending, straightening and aging. The magnetic properties of welded joints were evaluated at various temperatures using a physical property measurement system. In addition, the microstructure of the welded joint was studied by various microscopic analysis techniques, and the welded joint showed a stable austenitic structure. Furthermore, mechanical properties of the welded joints were investigated at different temperatures with particular emphasis on their behavior at 4.2 K. The circle-in-square ReBCO conductor jackets welding joint showed a yield strength greater than 1500 MPa, a fracture toughness KIC better than 260 MPa m1/2 and the elongation at break is more than 30 % at 4.2 K. However, a decrease in strength, ductility, and fracture toughness is observed within the welded joints of Nb3Sn conductor jackets after undergoing an aging process. This study will present experimental data on welding joints and discuss the feasibility of N50H as a high-magnetic field jacket material for next-generation fusion reactors.

Electromagnetic load evaluating and optimizing approach of the blanket system during VDEs considering halo current

In most scenarios, the vertical displacement event (VDE) represents the most extreme electromagnetic (EM) event within the tokamak device. The significant EM loads experienced during this event have the potential to compromise the structural stability of in-vessel components. This study investigates the EM loads on the water-cooled ceramic breeder blanket system of China Fusion Engineering Test Reactor (CFETR) using finite element analysis methods in two characteristic events: hot-VDE and cold-VDE. The study discusses the EM load effects resulting from changes in magnetic flux and induced electromotive force, respectively, with a specific focus on halo currents. The results reveal that, with similar current quech time, the difference in EM load on the blanket system during the VDEs primarily depends on the halo currents. When the electrical connection of the back supporting structure (BSS) is open, the halo current path within the blanket system and vacuum vessel (VV) changes, and a substantial portion of the halo current in the blanket system is conducted to the VV via the BSS. Consequently, a portion of the EM load on the blankets and BSS is transferred to the VV due to the transfer of halo current. Inspired by this, the conceptual use of ‘shunts’ is proposed to provide a dedicated circuit for shunting halo currents away from critical device components, such as the VV and blankets. This approach allows for the sharing of EM loads caused by halo currents and reduces the threat posed by halo currents to the structural integrity of these essential components.

Thermal Effects on Dynamic Behavior of CFETR Vacuum Vessel

The vacuum vessel (VV) of the China Fusion Engineering Test Reactor (CFETR) was designed to withstand the variable structural loads resulting from dynamic excitation, such as seismic and plasma disruption. Modal analysis, a powerful tool, was used to evaluate the structural dynamic characteristics, such as frequency and mode shape. In addition, the CFETR VV has three different temperature conditions: room temperature (20°C), normal operation temperature (100°C), and baking temperature (200°C). In this paper, in order to investigate the influence of such different temperature conditions to the dynamic behavior of the VV, three independent finite element analysis with the same modal analysis method were performed. According to analysis results, there are obvious thermal effects on the dynamic behavior, such as nature frequency and mode shape, among the different temperature cases for the CFETR VV. Moreover, the results show that the natural frequency for each order decreases as the temperature increases, and the mode shape of the VV also changed with temperature.

The effect of blanket system and port plugs on the harmonic components of error field and optimizing approach investigation

The fusion plasma is sensitive to the penetration of the toroidal mode numbers harmonic components of the error fields (EFs), and the toroidal mode components of the EFs are closely tied to the toroidal distribution of the magnetic field source. The primary source of EFs within the magnetic field sources is the current source, such as the magnet system, while the magnetic field produced by magnetized ferromagnetic materials may also generate EFs. The ferromagnetic field source, such the reduced-activation ferritic-martensitic steel, represent a smaller portion of the EFs source. However, the use of ferromagnetic materials is expected to increase significantly as tokamak research transitions from experimental devices to demonstration and commercial reactors, particularly for the structural materials of in-vessel components and port plugs. Predictably, the EFs induced by ferromagnetic materials will increase and may even surpass the design specifications. In this study, the EFs introduced by the water-cooling ceramic breeder blanket system and the vacuum vessel port plugs of China Fusion Engineering Test Reactor are comprehensively analyzed by an analytical approach and finite element numerical approach, which is based on the Fourier decomposition. The results reveal a clear linear relationship between the EFs formed by the saturated magnetization module and its rigid displacement deviating from periodic symmetry, as well as the same behavior in EFs and overall magnetization intensity. Compared to the assembly error of the blanket system, the EFs that may be introduced by the equatorial ports and plugs are substantial, including larger toroidal mode harmonic components that can lead to lock mode and disruption. Moreover, using the correlation between harmonic components and symmetrical periods, specific toroidal mode components can be selectively shielded.

Key features and highly effective prediction of complete startup from frozen state for high-temperature heat pipe in heat pipe reactor

High-temperature heat pipe is an efficient phase change heat transfer device, widely applied in the space nuclear reactors, the China Fusion Engineering Test Reactor, solar collectors, and the thermal protection of hypersonic vehicles. In order to understand the startup characteristics of the heat pipe and to predict its transient behavior with high accuracy, this paper first experimentally investigated the startup performance of a sodium heat pipe from the frozen state under different heating powers from 534 W to 1897 W. Then, a new compressible flow model for startup from the frozen state was developed and verified, solved the independence problems of mesh number and time-step used in the numerical treatments of most traditional heat pipe codes. Finally, based on the assessed model and experimental data, the startup procedure and the influences of heating power on the working temperature, kick-in time, and temperature difference were analyzed respectively. The comparisons indicated that the new model is extremely informative and effective, and the results provided knowledge for understanding the safe and optimum working regime of a high-temperature heat pipe using in heat pipe reactor.