China Fusion Engineering Test Reactor Research Articles

Preliminary analysis and design of 4kW@4.5K helium refrigerator for CFETR toroidal field magnet test facility

China Fusion Engineering Test Reactor (CFETR) is the next magnetic confinement fusion device to demonstrate fusion energy. To test the cryogenic performance of CFETR Toroidal Field (TF) superconducting magnet, a large-scale helium cryogenic system was under construction to provide Supercritical Helium (SHe), supported by Comprehensive Research Facility for Fusion Technology (CRAFT). In this paper, heat loads of TF magnet and the dedicated distribution system were analysed to define the scale. Process flow of the helium refrigerator has been designed considering the various operation phases. Preliminary thermodynamic cycle of the helium refrigerator have been calculated with the capacity of 4kW@4.5K+20g/s@50K. This 4kW@4.5K helium refrigerator was not only a test facility for CFETR superconducting magnets, but also a prototype of CFETR helium refrigerator.

Estimation of China Fusion Engineering Test Reactor performance and burning fraction in different pellet fueling scenarios by a multi-species radial transport model

Tritium self-sufficiency in future deuterium–tritium fusion reactors is a crucial challenge. As an engineering test reactor, the China Fusion Engineering Test Reactor requires a burning fraction of 3% for the goal to test the accessibility to the future fusion plant. To self-consistently simulate burning plasmas with profile changes in pellet injection scenarios and to estimate the corresponding burning fraction, a one-dimensional multi-species radial transport model is developed in the BOUT++ framework. Several pellet-fueling scenarios are then tested in the model. The results show that the increased fueling depth improves the burning fraction by particle confinement improvement and fusion power increase. Nevertheless, by increasing the depth, the pellet cooling-down may significantly lower the temperature in the core region. Taking the density perturbation into consideration, the reasonable parameters of the fueling scenario in these simulations are estimated as pellet radius injection rate and pellet injection velocity without drift or with high-field-side drift.

Preliminary Structural Design and Analysis of the Horizontal Cold Box for CFETR 25 kW@4.5 K Helium Refrigerator

The China Fusion Engineering Test Reactor cryogenic system mainly provides necessary supercritical helium for cooling superconducting coils and structure to transfer the static thermal load and the dynamic thermal load generated during plasma operation mode. According to the conceptual design of CFETR cryogenic system, the cooling capacity at 4.5 K is about 125 kW, with five 25 kW@4.5 K helium refrigerators in parallel. As one of the main components, a horizontal cold box provides a vacuum and insulating working environment for plate-fin heat exchangers, turbines, adsorbers, control valves and pipes. This paper presents a preliminary structural design for the 25 kW@4.5 K helium refrigerator horizontal cold box. The structural strength and static heat leakage of the cold box including support seats are analyzed through the finite element simulation method. The simulation results further verify the rationality and feasibility of the cold box structure design, and provide a certain reference for the subsequent detailed engineering design.

Numerical study on the flow and heat transfer characteristics of natural convection in CFETR cryostat

The cryostat in the China Fusion Engineering Testing Reactor (CFETR) provides vacuum and low temperature operating conditions for the superconducting magnet system. A high pressure in the cryostat can increase the Rayleigh number (Ra) and cause the natural convection within the continuous hypothesis. A three-dimensional model, using incompressible ideal gas model and realizable k-ε model, is built to simulate the natural convection between the superconducting magnet system and the thermal shield within the temperature range of 4.5-80 K. The distributions of velocity field and heat flux on different surfaces of superconducting magnet system are obtained under different Rayleigh number conditions. The results show that the areas with large natural convection velocity mainly locate near the Central Solenoid (CS) coils and at the bottom of vacuum chamber. When the Rayleigh number ranges from 104 to 1010, the maximum velocity of helium gas decreases, while the heat flux on walls of superconducting magnets increases rapidly, and the heat flux on walls of PF-1, CS and TF coils is more affected by pressure, which needs to be considered in designing the safety system.

Fabrication and performance examination on explosively welded ODS-Cu / 316 L clad plate for CFETR divertor heat sink

Copper alloys are among the most promising candidate materials for divertor heat sinks owing to their remarkable advantages. However, it is challenging to use CuCrZr alloys as heat sink materials because of the harsh loading environment expected for plasma-facing components (PFCs) in future fusion reactors. Explosive welding is an effective method for manufacturing high-quality oxide dispersion strengthened copper (ODS-Cu) / 316 L stainless steel (SS) clad plates. Non-destructive testing results revealed a soundness bonding interface between ODS-Cu and 316 L without any detectable defects or cracks. In this study, optical microscopy and scanning electron microscope analysis revealed that the bonding interface did not contain any cavities and a typical wave-bonding morphology was observed. Moreover, the tensile stress of the bonding interface was higher than that of the ODS-Cu parent metal, and there was an increment in hardness on both sides near the bonding interface. The helium leak testing results indicate that the welded ODS-Cu / 316 L exhibits soundness without deformation or leakage at cyclic room and high temperatures. The results prove that the explosive welding method is a practical method for manufacturing high-quality ODS-Cu / 316 L clad plate that meet the design requirements and can provide critical references for the engineering design of the China Fusion Engineering Test Reactor (CFETR) divertor heat sink.

Parametric decay instabilities of lower hybrid waves on CFETR

The lower hybrid current drive is a potential candidate for sustaining plasma current in tokamak steady-state operations, which could be used in China Fusion Engineering Test Reactor (CFETR) with input power up to a few tens of megawatts. Such high input power could trigger the well-known parametric instabilities (PIs) at the plasma edge affecting the propagation and absorption of the lower hybrid pump waves. By analytically solving the nonlinear dispersion relation describing PIs, an explicit expression of the PI growth rate is obtained and analyzed in detail. It is found that pressure is the key parameter determining the PI characteristics. Ion sound quasi-mode is the dominant decay channel in the low-pressure regime, while the ion cyclotron quasi-mode (ICQM), as well as its harmonics, becomes dominant in the intermediate regime. In the high-pressure regime, only one mixed channel is found, which is related to Landau damping by free-streaming ions. Analytical expressions of growth rates of these decay channels are also obtained to show the parameter dependence at different pressure limits. The above analytical results are used to estimate the PIs on a typical profile of CFETR, and verified by corresponding numerical calculations. ICQM is found to be the strongest decay channel with a considerable growth rate for CFETR.

Seismic Analysis of the Vacuum Vessel for CFETR

The China Fusion Engineering Test Reactor (CFETR) is the next tokamak device in China to bridge the gaps between ITER and the DEMOnstration nuclear fusion reactor (DEMO). The CFETR vacuum vessel (VV) was designed to remove nuclear heating, provide safety shielding, and maintain a high-quality vacuum environment. Seismic load is considered one of the most relevant accidental events affecting the structural integrity of the VV. In order to investigate the resistance of the CFETR VV against seismic load, finite element models of the VV were built. In this paper, equivalent static and response spectrum analyses were carried out to calculate displacements and stress fields aiming to check the response of the CFETR VV against a foreseen seismic load. The resulting stresses are lower than the allowable limits and satisfy the design requirements.

Conceptual Design of CFETR Divertor Dome for Remote Handling

The design of the China Fusion Engineering Test Reactor (CFETR) device is one of the main tasks in the next stage of research on fusion energy in China. The CFETR divertor is a key in-vessel plasma-facing component, which will suffer from sputtering erosion and exfoliation caused by intense plasma bombardment, severe heat fluxes, and neutron irradiation. Due to neutron activation of in-vessel components, remote handling (RH) technology has been considered as one of the essential issues of the CFETR divertor. In this article, a new concept of an integral hybrid divertor blanket for the CFETR is proposed, which is feasible for installation and disassembly in CFETR vacuum vessel by RH. Dome is an important part of the CFETR divertor, which comprises 80 modules, and each module is composed of flat-type plasma-facing units (PFUs), steel supports (SSs), and RH compatible structures. Every Dome module is connected to the CFETR blanket by RH compatible structures. This article mainly carries out the conceptual design of Dome (option), degree of freedom (DOF) analysis, stability analysis, RH process planning, and flow field analysis of cooling scheme, which will provide a solid foundation for its subsequent thermal-hydraulic analysis.

Diffusion bonding and interface structure of advanced carbide-dispersion-strengthened Cu and oxide-dispersion-strengthened W

Metallurgical bonding was achieved between carbide-dispersion-strengthened Cu (CDS-Cu) and oxide-dispersion-strengthened W (ODS-W) after diffusion bonding via an interface tailoring method of anodization and deoxidation of ODS-W combined with the introduction of Cu wetting layer. The shear strength of the CDS-Cu/ODS-W joints fabricated by the interface-tailoring bonding is 124 MPa, which is higher than that of 99 MPa of the joint prepared by direct bonding. The increase in strength is due to the achievement of the metallurgical bonding between CDS-Cu/ODS-W, the increase of the debonding energy and the increase of propagation distance of the cracks. After metallurgical bonding, the second phase particles of WC in CDS-Cu and Y2O3 in ODS-W were dispersed at the phase boundary to reduce the lattice strain energy. This work paves the way for the preparation of advanced plasma-facing components comprised of high-performance CDS-Cu and ODS-W for China Fusion Engineering Test Reactor (CFETR).

Basic Design and Progress of the Spliced Joint for CFETR TF Coil

High-performance superconducting joints are essential for the continuous and stable operation of superconducting magnet systems to generate stable magnetic fields, because the local heating from resistive losses can cause the superconductor to enter an unstable regime and result in quench. The China fusion engineering test reactor has five internal joints for each toroidal field coil. To provide a lower heat load of the joints to the cryogenic system, the direct current (dc) resistance of the internal joint connection must be below 0.3 nΩ. At the same time, the structure of the joint jacket should be convenient for winding. The internal joints must fit a space envelope equivalent to the original conductor cross section and must have enough structural strength and a low pressure drop. In the basic design, two trial joints were manufactured and tested at ASIPP. This article provides a description of the design, simulation, and manufacturing process of the trial joint, and finally obtained the test results with electrical resistances of 0.057 and 0.076 nΩ.

Preliminary accident analyses of ex-vessel LOCA and TES pipe break for CFETR HCCB blanket system

Safety analysis is one of the most important parts in the design of nuclear power reactor. It is a necessary step for China Fusion Engineering Test Reactor (CFETR) to make the research in safety accidents analysis of main coolant systems and tritium systems. The consequences of nuclear accidents can be used to prove the safety of plant and optimize the systems design. Now, the main design parameters of CFERR have been determined, and the conceptual design of the main systems have been done. This paper presents two design basis accidents (ex-vessel loss of coolant accident (LOCA) and TES pipe break) analysis results in primary helium cooling system (HCS), and tritium extraction system (TES). According to the accident analysis results, the radioactive release assessments are given. The analysis results can provide reference for the following parameter optimization design of CFETR systems.

Progress on neutronic analysis for CFETR

The neutron-induced irradiation field is a key problem in a fusion reactor related to nuclear responses, shielding design, nuclear safety and thermo-hydraulic analysis. To support the system design of the China Fusion Engineering Test Reactor (CFETR), a comprehensive analysis of the irradiation field has been conducted in support of many newly developed advanced tools. This paper first summarizes recent progress on the related effort to develop neutronics code including the geometry conversion tool cosVMPT, Monte Carlo variance reduction technology and ‘on-the-fly’ global variance reduction. Such developed tools have been fully validated and applied to the CFETR nuclear analysis. The neutron irradiation has been evaluated on the CFETR water-cooled ceramic breeder blanket, divertor, vacuum vessel (VV), superconductive coils and four kinds of heating systems: electron cyclotron resonance heating, ion cyclotron resonance heating, low hybrid wave and neutral beam injection. The nuclear responses of tritium breeding ratio, heating, irradiation damage and the hydrogen/helium production rate of material have been analyzed. In the case of neutron damage and overheating, deposition on the superconductive coils and VV, the interface and the shielding design among heating systems, and the blanket and other systems, has been initialized. The results show the shielding design can meet the requirement of coil and VV after several iterated neutronics calculations.

Effects of alpha particles on the CTEM driven zonal flow in deuterium–tritium tokamak plasmas

The effects of fusion-born alpha (α) particles on zonal flow (ZF) driven by collisionless trapped electron mode (CTEM) turbulence are analytically investigated, using gyrokinetic and bounce kinetic theories in the deuterium–tritium (D–T) tokamak plasmas. It is found that ZF growth rate is increased by α particles because of the reduction of polarization shielding as well as enhancement of CTEM instability. The results of this paper are qualitatively consistent with the enhancement of the level of residual ZF by α particles in (Cho and Hahm 2019 Nucl. Fusion 59 066026). The parametric dependence of ZF growth rate is also analyzed. The increment of ZF growth rate is further enhanced by α particles with higher fraction and steeper density profile. Besides, the dependence of ZF growth rate on electron temperature T e could be changed qualitatively by the presence of α particles when T i = T e. Moreover, the difference of ZF growth rates in the presence of α particles with slowing down and equivalent Maxwellian distribution functions is very weak. These results could be very important for accurate prediction of the confinement in the future burning plasmas such as International Thermonuclear Experimental Reactor and China Fusion Engineering Test Reactor.

Design and analysis of the CFETR cryostat

Cryostat is one of the most important systems of the China Fusion Engineering Test Reactor (CFETR) which is a next-generation device in the China fusion research roadmap for the realization of fusion energy. It is a large vacuum tight container (outer diameter: 38 m, height: 39.6 m) that provides 10−4 Pa vacuum environment for the superconductor magnet system and thermal shield system, which will be operated at 4.5 K and 80 K, respectively. The cryostat encloses the entire basic systems of the tokamak and must support all types of load, including gravitational loading, electromagnetic forces, and seismic loads generated from other components and itself. The CFETR cryostat comprises a dome-shaped top lid, two cylindrical sections with circumferential stiffening ribs, a bottom base section, a bottom pedestal, and a support. It is mainly made of stainless steel (SS) 304 L/304. Based on the design requirements, the cryostat has sixteen ports at the top, six ports in the middle, and sixteen ports at the bottom, respectively to provide penetrations for heating, vacuum-pumping, diagnostic, maintenance, etc. In this paper, motivation and main technical parameters of the cryostat are described. Interfaces between the cryostat and other systems are generally presented. Structural analysis, seismic analysis and buckling analysis have been carried out in the normal and abnormal operation cases. The external vacuum cryostat and the vacuum vessel are connected through the bellows in order to absorb relative displacement between them. In addition, the strength check and fatigue life assessment of the bellows of the cryostat are also carried out at the same time. Results live up to the expectation and a technology foundation will be laid for the future construction of the cryostat.

Electromagnetic Analysis of CFETR Vacuum Vessel During Plasma Vertical Displacement Event

The vacuum vessel (VV) of the China Fusion Engineering Test Reactor (CFETR) performs many functions, but its main function is to provide a high-quality, ultra-high-vacuum environment for plasma operation. The electromagnetic load acts on the VV and its ports during a plasma vertical displacement event (VDE). This paper adopts a new simulation method of the plasma current during transient electromagnetic analysis. The distribution of the electromagnetic load on the VV shells, the upper ports, the equatorial ports, and the lower ports is entirely displayed. The structural stress under the electromagnetic load during a VDE was obtained by electromagnetic-structural coupling. According to the structural analysis, the results indicate that the VV can meet the structural strength and stiffness requirement.

Prediction of the energetic particle redistribution by an improved critical gradient model and analysis of the transport threshold

Based on the theory of critical gradient model (CGM) and following the simulation method proposed by Waltz et al. [Nucl. Fusion 55, 123012 (2015)], a combination of TGLFEP and EPtran code is employed to predict the energetic particle (EP) transport induced by Alfvén eigenmodes (AEs). To be consistent with the experiment, recent improvements to the simulation method include consideration of threshold evolution and orbit loss due to finite orbit width. The revised CGM is applied to simulate two DIII-D experimental discharges (#142111 and #153071). It well reproduces the experimental profiles with multiple unstable AEs and large-scale EP transport. Discharge #142111 had previously been simulated using a nonlinear MHD-kinetic code MEGA [Todo et al., Nucl. Fusion 55, 073020 (2015)] with a transport mechanism based on stochasticity induced by overlapping AE. By comparing the simulated EP profiles, we find that the AE transport threshold is approximated by both the MEGA nonlinear stability threshold and the proposed CGM threshold (error <5% for single n and <17% for multiple n simulation). Both of them are larger than the linear stability threshold of the most unstable AE mode by a quantity of the order of the flux needed to sustain EP transport by the background turbulence. We have also applied the improved CGM to simulate the α particle redistribution for a China Fusion Engineering Test Reactor steady state scenario. Because of the clear separation between the AE unstable region and the loss cone, only a moderate α particle loss of ∼9.6% is predicted.

EMC3–EIRENE simulations of neon impurity seeding effects on heat flux distribution on CFETR

The numerical modelling of the heat flux distribution with neon impurity seeding on China fusion engineering test reactor has been performed by the three-dimensional (3D) edge transport code EMC3–EIRENE. The maximum heat flux on divertor targets is about 18 MW m−2 without impurity seeding under the input power of 200 MW entering into the scrape-off layer. In order to mitigate the heat loads below 10 MW m−2, neon impurity seeded at different poloidal positions has been investigated to understand the properties of impurity concentration and heat load distributions for a single toroidal injection location. The majority of the studied neon injections gives rise to a toroidally asymmetric profile of heat load deposition on the in- or out-board divertor targets. The heat loads cannot be reduced below 10 MW m−2 along the whole torus for a single toroidal injection location. In order to achieve the heat load mitigation (<10 MW m−2) along the entire torus, modelling of sole and simultaneous multi-toroidal neon injections near the in- and out-board strike points has been stimulated, which indicates that the simultaneous multi-toroidal neon injections show a better heat flux mitigation on both in- and out-board divertor targets. The maximum heat flux can be reduced below 7 MW m−2 on divertor targets for the studied scenarios of the simultaneous multi-toroidal neon injections.

Hydrogen isotope inventory evaluation of hydrogen isotopes separation system of CFETR using Aspen Plus simulator

The tritium plant needs to recover unburned deuterium and tritium for recycling to achieve tritium self-sufficiency. The tritium recovered by the water detritiation system (WDS) contains a large amount of protium. High purity deuterium (D) and tritium (T) are required as fuel for fusion reactor while hydrogen (H) always exists as an impurity of the fuel. Therefore, a hydrogen isotope separation system (ISS) is essential for fuel cycle in fusion reactor. ISS separates the recovered H, D, T, and removes the impurity protium (H) to obtain high-purity fuel deuterium and tritium. In this paper, two hydrogen isotope separation systems for the inner fuel cycle system and the outer fuel cycle system of CFETR (China Fusion Engineering Test Reactor) are respectively established according to the disparity in tritium concentration of hydrogen isotopes gas. The tritium inventory in the ISS is expected to be very large, evaluating the tritium inventory of ISS facilitates the evaluation of the minimum initial startup tritium inventory for the operation of the fusion reactor. The chemical simulation software Aspen Plus is used to simulate the separation process of ISS. The purity of products and the distribution of hydrogen isotopes in each distillation column are obtained. Based on the simulated data, the tritium inventory in each distillation column is calculated.

Progress of engineering design of CFETR cryogenic system

The China Fusion Engineering Test Reactor (CFETR) is the next magnetic confinement fusion device in China’s roadmap for realizing commercial fusion energy, aiming to bridge the gaps between the experimental reactor and the demonstration reactor DEMO. As a key subsystem of utilizing fusion power, the CFETR cryogenic system is required to create and maintain low temperature operation conditions for superconducting magnets, CP system and other users. The preliminary engineering design of the CFETR cryogenic system has been proposed. The system has a capacity of 125 kW@4.5 K, 1700 kW@80 K, 6 kW@15 K and 2.5 kW@1.8 K (optional). The dimension of cryodistirbution system has been carried with the length of 85.5 m, width of 82.9 m and height of 40.2 m. The cryogenic control system is based on the CODAC and controlled by PLC.

Progress in engineering design of CFETR toroidal field superconducting magnet

China Fusion Engineering Test Reactor (CFETR) is a national scientific research project which is aimed to initially build fusion energy production up to 2×108 W and to eventually reach relevant power level 1×109 W of DEMO. The major and minor radius of CFETR are 7.2 m and 2.2 m, respectively. As one of the most important components, a full-sized CFETR Toroidal Field (TF) magnet was designed. The height and width of the TF magnet are 21.3 m and 12.4 m separately. To provide 6.5 T at plasma center, TF magnet operates at a current of 95,600 A/turn. In this paper, engineering design of TF coil case (TFCC) is firstly introduced. And the structural design of internal joints, external joints, conductor, and insulation system of TF winding pack (WP) is presented. Based on the thermal load, the cooling system is also designed. Finally, some auxiliary structures are designed to ensure the normal operation of TF magnet.