Structural Design Study for ITER Upper ELM Coils

ITER ELM coils are important parts of in-vessel coils and they are mounted on the vacuum vessel and behind the blanket module. They consist of three sets of coils, referred to as the upper, mid, and lower coils. In order to verify the structural design feasibility and find the better structure for upper edge localized modes (ELM) coil, two different variants of coil support structures are studied under the electromagnetic load, thermal and other loads. Results show that besides the bracket of variant 2 does not meet the fatigue criteria, the conductor, jacket and bracket of the two structures can meet the static, fatigue and crack threshold criteria and both of them are valid and feasible. In addition, the better structure is chosen for upper ELM coil.     

Rapid Thermal-Hydraulic Analysis and Design Optimization of ITER Upper ELM Coils

ITER edge localized mode （ELM） coils are important components of the in-vessel coils （IVCs） and they are designed for mitigating or suppressing ELMs. The coils located on the vacuum vessel （VV） and behind the blanket are subjected to high temperature due to the nuclear heat from the plasma, the Ohmic heat induced by the working current and the thermal radiation from the environment. The water serves as coolant to remove the heat deposited into the coils. Based on the results of nuclear analysis, the thermal-hydraulic analysis is performed for the preliminary design of upper ELM coils using a rapid evaluation method based on 1D treatment. The thermal-hydraulic design and operating parameters including the water flow velocity are optimized. It is found that the rapid evaluation method based on 1D treatment is feasible and reliable. According to the rapid analysis method, the thermal hydraulic parameters of two water flow schemes are computed and proved similar to each other, providing an effective basis for the coil design. Finally, considering jointly the pressure drop requirement and the cooling capacity, the flow velocity is optimized to 5 m/s.     

Stress and Thermal Analysis of the In-Vessel RMP Coils in HL-2M

A set of in-vessel resonant magnetic perturbation(RMP) coils for MHD instability suppression is proposed for the design of a HL-2M tokamak.Each coil is to be fed with a current of up to 5 kA,operated in a frequency range from DC to about 1 kHz.Stainless steel(SS) jacketed mineral insulated cables are proposed for the conductor of the coils.In-vessel coils must withstand large electromagnetic(EM) and thermal loads.The support,insulation and vacuum sealing in a very limited space are crucial issues for engineering design.Hence finite element calculations are performed to verify the design,optimize the support by minimizing stress caused by EM forces on the coil conductors and work out the temperature rise occurring on the coil in diferent working conditions,the corresponding thermal stress caused by the thermal expansion of materials is evaluated to be allowable.The techniques to develop the in-vessel RMP coils,such as support,insulation and cooling,are discussed.     

Conceptual Design of Alborz Tokamak Poloidal Coils System

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. One of the most important parts of tokamak design is the design of the poloidal field system. This part includes the numbers, individual position, currents and number of coil turns of the magnetic field coils. Circular cross section tokamaks have Vertical Field system but since the elongation and triangularity of plasma cross section shaping are important in improving the plasma performance and stability, the poloidal field coils are designed to have a shaped plasma configuration. In this paper the design of vertical field system and the magnetohydrodynamic equilibrium of axisymmetric plasma, as given by the Grad-Shafranov equation will be discussed. The poloidal field coils system consists of 12 circular coils located symmetrically about the equator plane, six inner PF coils and six outer PF coils. Six outer poloidal field coils (PF) are located outside of the toroidal field coils (TF), and six inner poloidal field coils are wound on the inner legs and are located outside of a vacuum vessel.     

RAMI Analysis Program Design and Research for CFETR (Chinese Fusion Engineering Testing Reactor) Tokamak Machine

Chinese Fusion Engineering Testing Reactor (CFETR) is a test reactor which shall be constructed by National Integration Design Group for Magnetic Confinement Fusion Reactor of China with an ambitious scientific and technological goal. The reactor has the equivalent scale compared with ITER, but has the complementary function to ITER. CFETR is a demonstration of long pulse or steady-state operation with duty cycle time not less than 0.3–0.5 and the full cycle of tritium self-sustained with TBR not less than 1.2. At the same time it will be exploring options for DEMO blanket and divertor with an easy changeable core by remote handling way. To be able to reach its scientific and technological objectives, as one of technical risks control methods, RAMI analysis need to be done during the hold lifetime of CFETR, from conception design to decommissioning. Base on stating of CFETR lifetime and preliminary operational programme, the RAMI analysis program and process are designed and discussed, it consists of five major steps: (1) functional analysis are performed, (2) calculating reliability block diagrams, (3) analyzing failure mode, effects and criticality analysis, (4) risk mitigation actions are taken to ensure every system is compatibility with RAMI objectives, (5) All the RAMI analysis are integrated as the final RAMI analysis reports to be reviewed in the system final design review. Along with the elements of the analysis the vacuum vessel (VV) system was performed to provide as examples, detailed showing how the CFETR RAMI analysis is carried out. CFETR RAMI analysis guidelines were designed and established, after constantly revised and improved these analysis criteria and programs will become the basis standards for CFETR RAMI analysis. Preliminary RAMI analysis of CFETR VV system was obtained, which will be updated with the VV system design progresses.     

Availability Analysis for CFETR

Availability analysis for Chinese fusion engineering testing reactor (CFETR) is very important at the pre-conceptual phase. Availability apportion are the theoretical basis of system design of CFETR. Availability analysis informs the development of the CFETR overall system and subsystem design. Availability analysis will enable the identification of key subsystems to achieve availability targets. The duty cycle of CFETR should be at least 0.3–0.5. Such design goals require all subsystems of CFETR must have a pretty high availability. The availability of CFETR can be defined by break time analysis results. Analysis results proved that the availability of CFETR is 0.5–0.7. All availability subsystems of CFETR can be apportioned from their relevant mean time before failure (MTBF) and mean time to repair (MTTR) data to meet the availability goals. The relation between reliability, maintainability and availability indicates that a subsystem could have a high availability though its reliability is pretty low. Availability apportionment and analysis indicate that the availability of blanket and divertor should be 0.769–0.91 to meet the design requirements and their availability can be improved by increase the MTBF and reduce the MTTR of blanket and divertor with the development of remote handling or remote maintenance technology for fusion reactor.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

KTX反场箍缩装置纵场线圈结构设计与电磁分析

KTX反场箍缩装置的主要参数介于RFX装置与MST装置之间。反场箍缩的外加纵场需跟随等离子体电流的演化而变化,同时由于RFP中的磁面对于外部特别是等离子体边界处的径向磁场较敏感,所以需外部线圈的磁场更加精细,这对于线圈的磁场分布、误差场以及波纹度等的设计提出了更高的要求。根据KTX物理目标参数要求,提出矩形和楔形截面纵场磁体线圈设计方案,借助有限元软件和程序分析了其电磁场空间分布和结构受力大小。结果表明,6.4°楔形截面方案相比矩形截面方案在控制误差场方面更具有可行性。     

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible.     

Integrated Design System of Toroidal Field Coil for CFETR

Integrating engineering software is meaningful but challenging for a system code of a fusion device.This issue is seldom considered by system codes currently.Therefore,to discuss the issue,the Integrated Design System of TF Coil(IDS-TFC) has been worked out,which consists of physical calculation,CAD,and Finite Element Analysis(FEA).Furthermore,an Integrated and Automatically Optimized Method(IAOM) has been created to address the integration and interfaces.The method utilizes a geometry parameter to connect each design submodule and achieve automatic optimization.Double-objectives optimization has been realized,confirming it is feasible to integrate and optimize engineering design and physical calculation.Moreover,IDSTFC can also serve as a useful reference of integrated design processing for subsequent fusion design.     

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.     

Design Study of Support for ITER ELM Coils

The support is an important part of ITER ELM coils. It should withstand the alternating electromagnetic (EM) force and thermal stresses. Based on the finite element method, 2D and 3D structures of the rigid and flexible support of ITER upper ELM coil in different loads are studied. Results show that the flexible support can reduce the stresses of the conductor and jacket. In the lower level of nuclear heat, two types of supports can be used in the quarter model. In the high level of nuclear heat, the flexible support is needed and 50 mm support is proposed for the quarter model. Considering the EM load, the rigid support has better performance than the flexible support. Therefore, reasonable support can be provided for ELM coil or similar coil according to the thermal expansion and EM load.     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.     

Conceptual Design and Finite Element Analysis of STR-3 Feeder for ITER

The three structure (STR) cooling feeders of the international thermonuclear experimental reactor play an important role in removing heat loads from the coils cases of the superconducting magnets system. In this paper, the materials choice and conceptual design of STR-3 feeder are presented, then the structural analyses are performed based on the finite element method, which is used to evaluate the mechanical characteristics of STR-3 feeder. The results show that the structural design is basically reasonable, but the U bend or S bend structure near terminal box is necessary. Furthermore, the further structural optimization, reducing the heat leakage to cryopipes, radiation of composite insulation materials, overpressure protection and seismic analysis proposals are put forward.