Chinese Fusion Engineering Test Reactor Research Articles

Design of the Water-Cooled Ceramic Breeder blanket for CFETR

The Water-Cooled Ceramic Breeder (WCCB) blanket is one of the blanket candidates for the Chinese Fusion Engineering Testing Reactor (CFETR). In the current version, the WCCB blanket was designed to fulfill the requirements of operating in multiple fusion power modes (200 MW, 500 MW, 1 GW and 1.5 GW) of CFETR while satisfying the tritium self-sufficiency. Structural designs were carried out for typical inboard and outboard modules. Adequate neutron shielding capability and acceptable thermo-mechanical performance were achieved according to the analyses. The global Tritium Breeding Ratio (TBR) was 1.165 with full blanket coverage. It reduced to 1.052 considering the ports occupied by auxiliary heating systems and diagnostic systems. Eventually, the TBR increased to 1.123 by applying the approach of divertor blanket. Besides, two independent cooling systems were applied to adapt to different fusion powers. One cooling system provided coolant for the first wall, the other for breeder zones. The corresponding Primary Heat Transfer System (PHTS) was designed and main operation parameters were obtained under the fusion power of 200 MW∼1.5 GW.

Systematic design of snake arm maintainer in nuclear industry

This paper develops a remote handling system with a Snake Arm Maintainer (SAM) as the end effector, which aims to meet the maintenance needs of various narrow working spaces of the Chinese Fusion Engineering Test Reactor (CFETR). The multi-joint snake-arm is designed with the principle of layered drive to realie the angle synchronisation of adjacent joints, which reduces the complexity of the drive control system, and reduces the volume and weight while ensuring a high spatial curvature. The drive box is connected to the Multi-Purpose Overload Robot (CMOR) through a quick-change connector. The worm gear mechanism realies the reverse folding of the SAM and close to the CMOR, enabling access to the narrow vacuum chamber. Two working methods of SAM are designed based on trajectory planning and trajectory tracking. Finally, based on the actual environment of the CFETR vacuum chamber, simulation experiments were carried out on the rationality of the structural design of the SAM and the correctness of the working methods, which verified the feasibility of working in a narrow vacuum chamber.

Optimizations of CFETR steady state H-mode scenario with localized reversed shear enhanced internal transport barrier

Both a fully noninductive steady state operation scenario and a hybrid scenario with fusion power ∼ 1 GW and fusion gain >10 are being considered to fulfill the mission of a Chinese fusion engineering testing reactor. Compared to the hybrid scenario, plasma current is generally lower in steady state operation, so that better confinement and stabilization of MHD instability introduced by higher normalized beta (possibly beyond the ideal MHD limit without a wall) are required to achieve the same fusion performance. Integrated modeling is used to find candidate scenarios to match both these requirements at the same time. By creating a localized strong reversed magnetic shear using radio frequency wave driven current, a strong off-axis internal transport barrier is formed, so that the target fusion power and fusion gain are achieved for Chinese fusion engineering testing reactor steady state operation. Further optimizing the location of the reversed magnetic shear by modifying radio frequency wave launch parameters can keep the normalized beta below the ideal MHD no-wall limit while the fusion power remains beyond 1 GW. Based on this finding, several combinations of heating and current drives are proposed with fusion gain close to 12.5.

SOLPS-ITER simulations of high power exhaust for CFETR divertor with full drifts

One of the major challenges for the Gigawatt-class Chinese Fusion Engineering Testing Reactor (CFETR) is to efficiently handle huge power fluxes on plasma-facing components , especially the divertor targets. This work investigates the effects of two candidate radiation impurity species, argon (Ar) and neon (Ne), with two different divertor geometries (baseline and long leg divertor geometry) on the reduction of steady-state power load to divertor targets in CFETR by using the SOLPS-ITER code package with full drifts and kinetic description of neutrals. The modeling results show clearly that increasing the seeding rate of Ar or Ne with fixed fueling gas D2 injection rate reduces the target electron temperature and heat flux density for the baseline divertor geometry, which can be reduced further by higher D2 injection rate. With a high impurity seeding rate, partial detachment with steady-state power load at the divertor target below the engineering limit of 10 MW m−2 is demonstrated. In addition, the radiation efficiency for Ar is better than that for Ne. Increasing the divertor leg length reduces the electron temperature and heat load at the targets. This modeling, therefore, suggests that a long leg divertor design with Ar seeding impurity is appropriate to meet the CFETR divertor requirements.

Activation analysis and radwaste assessment of CFETR

The materials in the fusion system will be activated by the neutrons generated from D-T fusion reactions, causing the material with activity and decay heat at a variety level among the system. Predicting the amount of radwaste produced in the Chinese Fusion Engineering Testing Reactor (CFETR) and the time-evolution feature of the radwaste is of great importance to the minimization of waste and determine the management strategy. The purpose of this paper is to minimize the quality and quantity of the activated solid waste as low as reasonably achievable via source control, recycling and reuse, clearance, optimization of the management at the designing phase of CFETR. A detailed 3D model for CFETR is created to calculate the neutron spectra and fluxes of all non-void cells of the model. The inventory calculation for each cell with the given irradiation scenario and the cooling scheme was then carried out with FISPACT-II. The extensive calculation, data analysis, and post-processing are performed using a data processing code to calculate the component averaged activation responses and the waste classifications. The results suggest that the in-vessel materials exposed to high level neutron fluxes can hardly meet the clearance and low-level waste requirements, but they can be recycled with an advanced remote handling system shortly after shutdown or a conservative remote handling system after temporary storage within 100 years. The vacuum vessel, port plug, cryostat, and coils can be recycled with a conservative remote handling system with temporary storage. Especially, the port plug, cryostat, and central solenoid are clearable with temporary storage of 100 years. The influence of the operation time of in-vessel components on the radwaste severity and quantity is also evaluated. The severity of the radwaste increases when the operation time increases from 1000 to 3000 MWY, then decreasing with operation times longer than 3000 MWY. The detailed operation plan should be determined with comprehensive consideration of the important factors, including the influence on the radwaste severity. The allowed impurity level is also estimated, finding that the Uranium in the breeding zone and the SS316L used in divertor should be blamed for the intermediate-level waste production of blanket and divertor.

CRAFT TF Winding Line

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” shape, consists of a winding pack (WP) enclosed in the coil case. The WP is composed of 3 sub-WPs, called low, medium, and high field WPs. Limited by the unit length of the conductor, each sub-WP contains 1–4 subwindings connected by joints. In this article, the layout of the winding line, function, and main features of each unit are introduced, followed by the calculation of the winding trajectory. Based on these results, the detail design of the winding line has been completed, all the winding machines are under manufacturing now.

Decomposition and estimation of the pulsed active power of the CPSS in the CFETR tokamak reactor based on the FFT and improved Prony methods

The Chinese Fusion Engineering Test Reactor (CFETR) aims to bridge the gap between the International Thermonuclear Experimental Reactor and the demonstration fusion power plant. Its coil power supply system (CPSS) will be connected to an external high-voltage power grid for producing large pulsed active power and reactive power to feed its superconducting coils. To analyze the impact of its active power on the connected high-voltage grid, the active power signal should be modeled initially. Herein, the characteristics of the active power of the CPSS are analyzed by comparing with the pulsed load, photovoltaic power, and wind power. It is modeled as a combination of periodic and decaying components, which are decomposed by the fast Fourier transform (FFT) and improved Prony methods, respectively. The proposed Prony constraint is diagnosed, and the mode size is changed to avoid exceeding the parameters. Both the calculated signal-to-noise ratio and fitting curves show the high accuracy of the proposed estimation procedure. Numerical results show that the number of selected FFT signals has no distinct influence on the estimation accuracy. The optimum Prony mode size of the decaying component estimation may be half the actual signal size when both the computing time and accuracy are considered.

Tritium transport code development based on COMSOL and MATLAB and its application to the WCCB blanket

Water cooled ceramic breeder (WCCB) blanket was proposed as one of the candidate blanket schemes of the Chinese Fusion Engineering Test Reactor (CFETR). In this paper, a code simulating tritium transport in WCCB blanket and its external loops has been developed, where zero and two dimensional models are integrated to get more reliable results. The purge gas flow in the breeder zone, heat transfer in the blanket and tritium isotopes mass transport are considered. The isotope exchange in the purge gas and coolant loop are also taken into account in the model. The two dimensional model of the WCCB blanket and zero dimensional model of external loops are integrated to calculate tritium related quantities. The amount of tritium inventory in different components, tritium losses, tritium extracted by the tritium extraction system (TES) and purified by the coolant purification system (CPS) are obtained. The sensitivity analysis about key parameters has also been conducted.

A Water Loop Design for the CRAFT Project towards the Testing of CFETR Water-Cooled Blanket and Divertor

As one of the tasks of the Comprehensive Research Facility for Fusion Technology (CRAFT), a High Heat Flux (HHF) testing device will be built to test the blanket and divertor of Chinese Fusion Engineering Testing Reactor (CFETR). The water loop is a key system of the HHF testing device. The main objective of the water loop is to provide deionized water at specific temperature, pressure, and flow rate for different testing experiments of the water-cooled blanket and water-cooled divertor components. The design of the water loop has been through three major steps. Firstly, the water cooled blanket and divertor were designed and analyzed, in detail, for CFETR. Secondly, thermal hydraulic features of the prototypes were abstracted from the analyses results. Then, the experiment plan was made so that the preliminary design of the water loop was carried out. The third step was the engineering design, which was conducted through cooperation with an industrial enterprise with certifications. At present, the water loop is ready for fabrication and construction. The water loop will be completed, for commissioning operation, by August 2022, as scheduled. After that, the experiments will be carried out step by step and provide solid technical base to CFETR.

A study on nuclear analysis of the divertor region of the CFETR

This paper presents the nuclear analysis performance of the Chinese Fusion Engineering Test Reactor (CFETR) divertor region using the MCNP-5 Monte Carlo N-particles code in a 3D geometry model. We assessed the nuclear responses of the divertor region component systems and evaluated their shielding capability, which can support the development strategy of the physical and engineering design of the CFETR. Model specification based on the latest CAD model of the CFETR divertor has been integrated into the CFETR MCNP reference model with a major/minor radius R = 7.2 m/a = 2.2 m in the 22.5° model, and a fusion-power range of around 1–1.5 GW. The nuclear heating and radiation damage of the divertor system are enhanced compared to that of the ITER and the earlier CFETR design. The initial nuclear responses of the toroidal field coil and vacuum vessel systems showed that the shielding of the current divertor design is not sufficient and optimization work has been carried out. We also carried out calculations and analysis using a hypothetical operating scenario of over 14 years. An excellent improvement in the nuclear performance has been obtained by the improved additional shielding block in the divertor region when referring to the ITER design limit, which can support the design of the future update of the divertor region systems of the CFETR.

Thermal Hydraulic Analysis on the Water Lead Lithium Cooled Blanket for CFETR

A new type of Water Lead Lithium Cooled (WLLC) blanket that adopts the modular design scheme, water cooling the structure components, liquid PbLi as breeder and coolant, and SiC as the thermal insulator between PbLi and structures is under development as a candidate blanket concept for the Chinese Fusion Engineering Test Reactor (CFETR). Based on a poloidal-radial slice model, thermal hydraulic analysis is performed for this blanket to validate the feasibility of design goals. Results show that the present design can achieve the outlet temperature in the range of 600–700 °C, with all the material temperatures safely below the upper limits. A series of sensitivity analyses are also carried out. It indicates that the thermal conductivity (TC) of SiC would have a significant influence on the temperature field, streamlines and pressure drop; that is, lower TC of SiC can maintain the temperature of PbLi at a high level, and induce an increased number of vortices in the liquid PbLi flow as well as a larger pressure drop. On this basis, the joint effects of the TC of SiC and inlet velocity on the performance of blanket thermal hydraulics are analyzed, then the so-called “attainable region” is proposed. Finally, optimization design studies are carried out by decreasing the width of the front channel. Comparison results show that the present design is the most reasonable.

Analysis and Verification of CRAFT TF Coil Turn Insulation Wrapping System

The Comprehensive Research Facility for Fusion Technology (CRAFT), national big science facility in China, is focused on exploring key technologies and building prototype systems for the 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 a gigantic “D” shape, which consists of a winding pack (WP) enclosed in the coil case. The winding pack is composed by embedding three sub-WPs, called low, medium, and high field WPs. The conductor of the high and medium WPs is fabricated from Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn, follow the “winding and reaction” approach. As one of the most critical steps, turn insulation has to be wrapped after the heat treatment of the conductor. Since the superconducting property of Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn is deeply sensitive to the strain, the strain must be strictly controlled during the wrapping process. In this article, the concept design of the wrapping system is introduced. Analysis of strain and construction of the wrapping mathematic model are performed and verified by the wrapping trial. Based on the abovementioned results, the detail design of the wrapping system has been completed and the machine is currently being manufactured.

Neutron characterization analysis of the negative ion-based neutral beam verification facility for CFETR

Neutral beam injection heating with negative ion source is one of the important auxiliary heating system to achieve the high parameter steady- state operation goal for CFETR. A Negative ion-based Neutral Beam Verification Facility (NNBVF), a conceptual design for testing the engineering feasibility of neutral beam injection systems, is proposed for the Chinese Fusion Engineering Test Reactor (CFETR). The neutral-beam injectors used for CFETR will operate with deuterium. Inevitably, a large number of neutrons are generated between beam deuterons and deuterons implanted in the calorimeter. Neutron source is one of the important input data for radiation analysis. The neutron characterization on the calorimeter surface were explored by using a multiaperture Gaussian divergence model and the local mixing model. The total neutron intensity reach up to 1.2 × 1014 n/s for 400 keV deuteron beam. Initial results revealed that the aspect ratio of the neutron beam profile decreases with increasing beam divergence. The shape of the neutron profile on the calorimeter surface remains the same for the beam density distributions. This implies the neutron intensity emitted from the target is suitable for the beam density distribution diagnostic applications.

Multiscenario Electromagnetic Load Analysis of Blanket and VV for CFETR

The evaluation of electromagnetic (EM) loads acting on conductive structures is one of the critical problems for tokamaks. This paper focuses on the main conductive components of the Chinese Fusion Engineering Test Reactor (CFETR), which are vacuum vessel (VV) and blanket components. A multiscenario EM load-calculating simulator is developed. The simulator adopts circuit equations for the calculation of eddy current, magnetic field, and resulting EM loads acting on the VV. And for EM loads acting on the blanket components, an integral parameter method is applied. With the circuit equations and pre-obtained integral parameters implanted in the simulator, EM loads on the CFETR VV and blankets at different scenarios can be evaluated quickly. Results indicate that the simulator can be conveniently used for multiscenario and real-time EM assessment of the tokamak VV and blanket components.

Conceptual design of the supercritical CO2 cooled lithium lead blanket for CFETR

A conceptual design of the supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been proposed as one advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). At present, the COOL blanket is designed to fulfill the requirement of operating under the fusion power of 1.5 GW of CFETR while realizing the tritium self-sufficiency. Structural designs are carried out for outboard and inboard blanket segments and analyses with regard to neutronics and thermo-hydraulics are conducted to evaluate the blanket performance. Results indicate that a comparatively high Tritium Breeding Ratio (TBR) of 1.183 and a PbLi outlet temperature of 600 – 700°C can be achieved for this blanket concept. Besides, a Power Conversion System (PCS) based on the supercritical CO2 recompressing cycle is preliminarily designed for the blanket and the gross thermal efficiency is estimated to be 39% – 46% when assuming the turbine inlet temperature ranges between 550°C and 650°C.

Numerical Analysis of Heat Transfer and Flow Field for CFETR Divertor

Nuclear fusion is now the ultimate need for energy, tokamak is one of the possible devices for controlled nuclear fusion. Based on the advanced tokamak magnetic confinement configuration and the research results of fusion physics in recent decades at home and abroad, Chinese Fusion Engineering Test Reactor (CFETR) will carry out innovative physical design to achieve the international requirements. Divertor is an essential component of the magnetic confinement fusion device, which directly sustains the strong particle flow and high heat load. The cooling system of divertor will directly affect whether divertor can function properly. In this paper, a new structure of divertor suitable for CFETR is designed and simulated, both the flow field analysis and the heat transfer analysis of the model were verified. The results show that the designed divertor structure can meet the system requirements.

Modelling of the complete heat flux deposition on the CFETR first wall with neon seeding

Steady-state heat flux deposition on the first wall of the Chinese Fusion Engineering Testing Reactor (CFETR) is investigated for 1 GW fusion power and under the updated design scheme. The computation is accomplished by the modelling of two simulation systems: SOLPS-ITER and PFCFlux. Seeding neon (Ne) gas is considered to be one of the possible methods of protecting the CFETR divertor from high heat flux damage. Simulations with different Ne seeding rates were carried out on a standard divertor with a lower single-null configuration to determine the effect of Ne seeding on heat flux deposition on the first wall. A blob-based transport model was used to investigate the effect of enhanced cross-field transport on the CFETR first wall, as with ITER and EU DEMO simulations. This work establishes a potential method for simulating the complete heat flux deposition on the first wall for the CFETR steady state by SOLPS-ITER and PFCFlux. After modelling and energy analysis, the peak value of complete heat flux on each module of the CFETR first wall is obtained, which may provide the basic reference for the CFETR physics and engineering design team. The CFETR first wall module which receives the highest heat flux has also been identified, and Ne seeding has a positive effect on reducing the heat flux on the CFETR first wall from around 2.5 MW m−2 to 1 MW m−2, which is within the engineering requirement.

Conceptual design and heat transfer performance of a flat-tile water-cooled divertor target

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ∼20 MW m−2. In view of such a high heat flux removal requirement, this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology. Rectangular water-cooled channels with a special thermal transfer structure (TTS) are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance (HTP). The parametric design and optimization methods are applied to study the influence of the TTS variation parameters, including height (H), width (W*), thickness (T), and spacing (L), on the HTP. The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes, and the sensitivity is T > L > W* > H. The HTP first increases and then decreases with the increase of T, L, and W* and gradually increases with the increase of H. The optimal design parameters are as follows: H = 5.5 mm, W* = 25.8 mm, T = 2.2 mm, and L = 9.7 mm. The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method. A flat-tile divertor mock-up is developed according to the optimized parameters. In addition, high heat flux (HHF) tests are performed on an electron beam facility to further investigate the mock-up HTP. The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m−2 under the tungsten tile thickness <5 mm and the flow speed ≥7 m s−1. The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ∼13% and ∼30% compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock, respectively. The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m−2 of heat load without visible deformation, damage, and HTP degradation. The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ∼930 °C. The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method, and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor. This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.

The neutron-photon-coupling analysis of the tritium-breeding blanket in CFETR by NECP-MCX

The design of Chinese Fusion Engineering Test Reactor (CFETR) is being carried out. The neutronics analysis of the tritium-breeding blankets in CFETR is one of the essential tasks, and raises requirements for the neutronics-analysis codes such as the ability of simulating the neutron-photon-coupling transport and the ability of obtaining the accurate energy deposition of neutrons and photons. NECP-MCX is a hybrid Monte-Carlo-deterministic code newly developed by the Nuclear Engineering Computational Physics (NECP) Lab at Xi'an Jiaotong University. In order to meet the requirements raised by CFETR, the neutron-photon-coupling transport and other related functionalities are developed in NECP-MCX. The code-to-code verification shows that the results of NECP-MCX agree well with those of other codes. The application of NECP-MCX for the neutron-photon-coupling analysis of the tritium-breeding blankets in CFETR shows that NECP-MCX can obtain more reliable results of photon kerma than MCNP.

Conceptual design of a blanket module integrated with divertor for CFETR

Achieving tritium self-sufficiency (tritium breeding ratio (TBR)≥1.1) is one of the most important missions of the Chinese Fusion Engineering Test Reactor (CFETR). The neutronics analysis shows that the global TBR of the CFETR water-cooled ceramic breeder (WCCB) blanket is 1.147 when the plasma is surrounded by blankets (without any windows), which can meet the tritium self-sufficiency requirement. In the engineering design phase, however, a number of blanket modules are replaced by auxiliary systems. It would lead to decline of TBR and failure of tritium self-sufficiency. Therefore, a new blanket module concept extended to the divertor region and integrated with the divertor is proposed to increase TBR. The blanket module is located at the back of the divertor target plates. This paper presents the design of the proposed blanket module. The feasibility of the blanket design is evaluated considering neutronics, thermal and mechanical performances.