Progress in CFETR Power Stations Concept Study

While EAST experiment was running in 2012, the project of the China fusion engineering test reactor (CFETR) concept design was started. This ITER-like tokamak system will be the second full superconducting tokamak in China based on EAST technology. In phase I, it has 50–200 MW heat output for demonstrating power generation. The fusion power stations contain complete structure of fusion power plant (FPP) which do not appear in the ITER and huge HV substation which receives power from the 500 kV transmission grid for powering its pulsed power electric network (PPEN) and steady-state electric power network. Furthermore, its structure of turbine generator of FPP is similar to that of a nuclear power station of the pressurized-water reactor. This paper describes the typical CFETR loads and put forward the requirements of short circuit capacity of HV grid. It analyzes different strategies of putting the generator power to the grid, i.e. on the 500 kV grid for future DEMO power structure or 66 kV medium-voltage local grid for self-use. In period between twice burning plasma, conceptual solutions are presented to maintain thermal circuit operation.     

Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices.     

Preliminary Calculation of Electromagnetic Loads on Vacuum Vessel of HL-2M

For a robust design of vacuum vessel of HL-2M, the electromagnetic (EM) loads have to be understood clearly. In this paper, some crucial transient events, such as plasma major disruptions (MDs), vertical displacement events (VDEs), fast discharge of toroidal field (TF) coils, have been investigated to evaluate the eddy currents and EM forces on vacuum vessel and in-vessel components. The results show that the eddy currents depend strongly on the current decay time, and the maximum toroidal eddy current flowing in the whole vessel can reach up to 2.4 MA during MDs that is close to the plasma current. Large symmetric radial forces and a net vertical force on vessel shells could be caused by these transient events. Combination of eddy currents in in-vessel components and toroidal field could twist the copper plates and other internal parts, however, if these plates are supported and connected carefully, the twist moments will not have a big e®ect on the vessel shells and vessel support.     

Electromagnetic disruption analysis of 2.45 GHz LHW antenna in EAST under different plasma configurations

The 2.45 GHz lower hybrid wave (LHW) antenna is one of the key components for plasma heating and current drive on experimental advanced superconducting tokamak (EAST). In the lower hybrid current drive (LHCD) experiment, the microwave power is delivered to the plasma through the LHW antenna. During a plasma disruption, the eddy currents are induced in the antenna because of plasma current decay. These induced currents interact with the strong static magnetic field to produce forces and torques in the antenna which are one of key factors determining the design of the antenna. Therefore, this paper presents the key results of a transient electromagnetic (EM) analysis of the antenna during disruption events under different plasma configurations. Two plasma centered disruption scenarios are taken into account: exp quench and linear quench. The analysis was performed with MAXWELL, a computer code based on the finite element method. All the results are presented and discussed which will offer guidance for the design and manufacture of the antenna in future.     

Electrical connectors for blanket modules in ITER

Blanket electrical connectors (E-straps, ES) are low-impedance electrical bridges crossing gaps between blanket modules (BMs) and vacuum vessel (VV). Similar ES are used between two parts on each BM: the first wall panel (FW) and shield block (SB). The main functions of E-straps are to: (a) conduct halo currents intercepting some rows of BM, (b) provide grounding paths for all BMs, and (c) operate as electrical shunts which protect water cooling pipes (branch pipes) from excessive halo and eddy currents. E-straps should be elastic enough to absorb 3-D imposed displacements of BM relative VV in a scale of ±2 mm and at the same time strong enough to not be damaged by EM loads. Each electrical strap is a package of flexible conductive sheets made of CuCrZr bronze. Halo current up to 137 kA and some components of eddy currents do pass through one E-strap for a few tens or hundreds milliseconds during the plasma vertical displacement events (VDE) and disruptions. These currents deposit Joule heat and cause rather high electromagnetic loads in a strong external magnetic field, reaching 9 T. A gradual failure of ES to conduct Halo and Eddy currents with low enough impedance gradually redistributes these currents into branch pipes and cause excessive EM loads. When branch pipes will be bent so much that will touch surrounding structures, the Joule heating in accidental electrical contact spots will cause local melting and may lead to a water leak.The paper presents and compares two design options of E-straps: with L-shaped and Z-shaped elastic elements. The latter option was developed in 2012 on the basis of more thoughtful analysis of bi-directional cyclic loading conditions influencing a fatigue lifetime. Detail comparative simulations of current and field patterns and subsequent analysis of the fatigue strength and technological assessment allowed make a final choice for the E-strap design in ITER.     

NSTX disruption simulations of detailed divertor and passive plate models by vector potential transfer from OPERA global analysis results

The national spherical torus experiment (NSTX) project is planning upgrades to the toroidal field, plasma current and pulse length. This involves the replacement of the centerstack, including the inner legs of the TF, OH, and inner PF coils. A second neutral beam will also be added. The increased performance of the upgrade requires qualification of the remaining components including the vessel, passive plates, and divertor for higher disruption loads. The hardware needing qualification is more complex than is typically accessible by large scale electromagnetic (EM) simulations of the plasma disruptions. The usual method is to include simplified representations of components in the large EM models and attempt to extract forces to apply to more detailed models. This paper describes a more efficient approach of combining comprehensive modeling of the plasma and tokamak conducting structures, using the 2D OPERA code, with much more detailed treatment of individual components using ANSYS electromagnetic and mechanical analysis. This capture local eddy currents and resulting loads in complex details, and allows efficient non-linear, and dynamic structural analyses.     

Status of ITER blanket attachment design and related R&D

Main function of the ITER blanket system [1], [2], [3] is to shield the vacuum vessel (VV) from nuclear radiation and thermal energy coming from the plasma. Blanket system consists of discrete blanket modules (BM). Each BM is composed of a first wall panel and a shield block (SB). The shield block is attached to the VV by means of four flexible supports and three or four shear keys, through key pads. All listed supports do have parts with ceramic electro-insulating coatings necessary to exclude the largest loops of eddy currents and restrict EM loads. Electrical connection of each SB to the VV is through two elastic electrical straps. Cooling water is supplied to each BM by one coaxial water connector. This paper summarizes the recent evolution of the blanket attachment system toward design solutions compatible with design loads and numbers of load cycles, and providing sufficient reliability and durability. This evolution was done in a frame of pre-defined external interfaces. The ongoing supporting R&D is also briefly described.     

CFETR氦冷陶瓷增殖包层在等离子体主破裂时的电磁结构耦合分析

增殖包层作为中国聚变工程实验堆(China Fusion Engineering Test Reactor,CFETR)的核心部件,承载着能量转换和氚增殖的重要作用。中国科学院等离子体物理研究所在之前增殖包层设计的基础上,又提出了氦冷陶瓷增殖(Helium Cooled Ceramic Breeder,HCCB)包层的概念设计。为评估电磁载荷对HCCB包层结构安全性的影响,借助通用有限元软件ANSYS,研究计算了在等离子体主破裂时包层中产生的感应涡流、洛伦兹力和力矩。通过多物理场耦合分析方法,获取了包层中产生的等效应力和形变位移。结果表明,在等离子体电流指数衰减时,HCCB包层模型上产生的最大等效应力和形变位移满足包层结构设计的要求,同时模拟分析结果也为未来的包层结构优化以及支撑结构设计提供了必要的数据支撑。     

Analysis of an Extreme Scenario in the Vacuum Vessel in KTX

The Vacuum Vessel(VV) system is a vital component of Keda Torus for experiment(KTX).Various accidental scenarios might occur on the VV.In this report,an extreme scenario is assumed and studied:plasma accidental termination during the flat-top stage.Numerical simulations based on finite element are performed as the major tool for analyses.The detailed distributions of eddy and the reaction forces on VV are extracted,and the total eddy current and the maximum reaction force due to electromagnetic load are figured out.In addition,according to the results,the VV can be approximately regarded as a centrally symmetric structure,even though its ports distribution is asymmetric.     

Analytical Analysis of the Electromagnetic Response for the Vacuum Vessel of EAST

During plasma disruptions and vertical displacement events (VDEs), time-varying eddy currents are induced on Vacuum Vessel (VV) and Plasma Facing Components (PFCs) of Tokamak. In this paper, we calculate eddy currents induced forces on VV and inner limiter (one of the PFC) during plasma disruption in Experimental Advanced Superconducting Tokamak (EAST). Various plasma transients—VDEs and major disruptions (MDs) are considered. And the study includes assessments of integral values (net vertical, hoop forces), time evolution, peaks of the force distribution. It is shown that the distribution pattern of the eddy currents for different scenarios differ greatly, therefore the resulting EMFs and torques cause different mechanical response.     

Global computational models for analysis of electromagnetic transients to support ITER tokamak design and optimization

The International thermonuclear experimental reactor (ITER) concept implies a variety of operating modes, design complexity and demand for high reliability. A point of the major concern is the transient electromagnetic (EM) effects. Complex electromagnetic behaviour due to strong inductive coupling, the presence of numerous field sources, and a range of plasma burn scenarios requires careful predictive simulations. Different mathematical models applicable for the design and optimization studies are reviewed. Practical experience in developing detailed global models to investigate eddy currents, EM forces and other EM loads is summarized. Two numerical techniques implemented in the dedicated computer codes are compared, and the validity of relevant models is discussed.     

Simulations of energetic alpha particle loss in the presence of toroidal field ripple in the CFETR tokamak

Energetic alpha particle losses with the toroidal field ripple and the Coulomb collision in the CFETR tokamak have been simulated by using the orbit-following code GYCAVA for the steady-state and hybrid scenarios. The effects of the outer boundary and the ripple amplitude on alpha particle losses have been investigated. The loss fractions and heat loads of alpha particles in the hybrid scenario are much smaller than those in the steady-state scenario for a significant ripple amplitude. Some alpha particles in the plasma core are lost due to the ripple stochastic transport for a large ripple amplitude parameter. The heat loads with the last closed flux surface boundary are different from those with the wall boundary for the CFETR tokamak, which can be explained by typical alpha particle orbits. Discrete heat load spots have been observed in alpha particle loss simulations, which is due to the ripple well loss. The transition of the lost alpha particle behavior from the ripple stochastic diffusion to the ripple well trapping has been identified in our CFETR simulations. The Coulomb collision effect is responsible for this transition.     

Evaluation of electromagnetic loads on various design options of the ITER diagnostic upper port plug during plasma disruptions

Electromagnetic (EM) loads due to eddy current and halo current during plasma disruptions are evaluated for the ITER diagnostic upper port plug. To reduce strong EM loads acting on the port plug fixed to the vacuum vessel like a cantilever beam, three design options have been considered: removal of the diagnostic first wall, slitting of the diagnostic shield module and recess of the port plug. The main focus of the present study is to examine the efficacy of these options in terms of EM loads on the upper port plug. It is found that making slits is more effective than removing the first wall. It is also shown that the upper port plug needs to be recessed to reduce the EM load induced by halo current.     

Design and construction of Alborz tokamak vacuum vessel system

The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.     

Design and Analysis of Fast Response Freezing Valve for EAST Liquid Lithium Limiter Loop

Lithium which can be served as first wall material in tokamak has the function of improving the plasma parameters. Liquid lithium limiter (LLL) system has been planned to be installed in EAST and applied in next campaign of physics experiment. To avoid the large pressure drop of liquid lithium caused by the long distance of feeding pipe in the central shaft, the lithium tank is planned to be installed in the vacuum vessel (VV) and moved with the limiter. A valve will be installed between the lithium tank and the limiter which can stop the lithium feeding immediately if accidents like coolant leaking in the VV occurred. Restricted by the large magnetic field, common valves can’t work normally, and at last freezing valve was chosen to be used. In this paper, the mechanical design of the freezing valve is introduced and the thermal–hydraulic properties of the valve are analyzed. Helium gas (80 K) and water (300 K) can be used as the coolant for the valve. Based on the analysis results, to satisfy the design requirement for the response time of shutting down, the flow rate of the helium gas must be large than 4 g/s (15 m/s), while, for water 2 m/s is enough. Besides, using water for cooling, the thermal affected area is much little than helium gas. At last, water with velocity of 2 m/s is chosen for the freezing valve used in EAST LLL loop, and the thermal-structural analysis shows that the thermal stress is safe for the current design.     

Design and Development of Rogowski Coil Sensors for Eddy Currents Measurement on Toroidal Vessel

In tokamak, eddy currents are produced due to change in plasma positions during plasma instabilities that result into generation of electromagnetic forces on interaction with the induced currents. Measurement of this current is essential to design a mechanical structure that can withstand this force. Principle objective of this paper is the development of Rogowski coil sensor to measure eddy currents on a toroidal vessel. The paper presents an elaborative and practical construction technique of a Rogowski coil. The calibration method for the Rogowski coil is also presented. Rogowski coils as an eddy currents diagnostics are tested and experiments to measure induced currents on the toroidal vessel are performed using the coils. Experimental values of eddy currents are compared with the ANSYS simulation results.     

Structural Safety Assessment of a Tokamak-Type Fusion Facility for a Through Crack to Cause Cooling Water Leakage and Plasma Disruption

A tokamak-type fusion machine has inherent safety associated with plasma shutdown. A small water leak can cause a plasma disruption although there is another possibility to terminate plasma without disruption. This plasma disruption will induce electromagnetic (EM) forces acting in the vacuum vessel (VV). From a radiological safety viewpoint, the VV is designed to form a physical barrier that encloses tritium and activated dust. If the VV can sustain an unstable fracture by EM forces from a through crack to cause the small leak, the structural safety will be assured and the inherent safety will be demonstrated. Therefore, a systematic approach to assure the structural safety is developed. A new analytical model to evaluate the through crack and leak rate of cooling water is proposed, with verification by experimental leak measurements. Based on the analysis, the critical crack length to terminate plasma is evaluated as about 2 mm. On the other hand, the critical crack length for unstable fracture is obtained as about 400 mm. It is concluded that EM forces induced by the small leak to terminate plasma will not cause unstable fracture of the VV; thus the inherent safety is demonstrated.     

Strategy for solving a coupled problem of the electromagnetic load analysis and design optimization for local conducting structures to support the ITER blanket development

The complexity of the electromagnetic (EM) response of the tokamak structures is one of the key and design-driving issues for the ITER. We consider the specifics of the assessment of ponderomotive forces, acting on local components of a large electro-physical device during electromagnetic transients. A strategy and approach is proposed for the operative EM loads modeling and analysis that enables design optimization at early phases of development. The paper describes a method of principal simplification of the mathematical model, based on the analysis and exploiting specific features and peculiarities of the relevant technical problem, determined by the design and operation of the device and system under consideration. The application of the method for predictive EM loads analysis and corresponding numerical calculations are exemplified for the localized ITER blanket components — shield modules. The example demonstrates the efficiency of EM load analysis in complex electromagnetic systems via a set of simplified models with different scope, contents and level of detail.     

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak(EAST)L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer(SOL)widthλ_q and heat spreading 5,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I_p.Strong inverse scaling of the SOL width with I_p has been achieved for both L-mode and H-mode plasmas in the forms ofλ_(q,L-mode)=4.98×I_p~(-0.68)andλ_(q,H-mode)=1.86×I_p~(-1.08).Similar trends have also been demonstrated in the study of heat spreading with S_(L-mode)=1.95×I_p~(-0.542)and S_(H-mode)=0.756×I_p~(-0.872).In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor(CFETR).     

Measurement of Plasma Energy Confinement Time in Presence of Resonant Helical Field in IR-T1 Tokamak

Plasma energy confinement time is one of the main parameters of tokamak plasma and Lawson criterion. In this paper we present an experimental method especially based on diamagnetic loop (toroidal flux loop) for measurement of this parameter in presence of resonance helical field (RHF) in IR-T1 tokamak. For this purpose a diamagnetic loop with its compensation coil constructed and installed on outer surface of the IR-T1. Also in this work we measured the plasma current and plasma voltage from the Rogowski coil and poloidal flux loop measurements. Measurement results of plasma energy confinement time with and without RHF (L = 2, L = 3, L = 2 & 3) show that the addition of a relatively small amount of RHF could be effective for improving the quality of tokamak plasma discharge by flatting the plasma current and increasing the energy confinement time.