A divertor plasma configuration design method for tokamaks

Guo, Youming; Xiao, Bingjia; Liu, Lei; Yang, Fei; Wang, Yuehang; Qiu, Qinglai

doi:10.1088/1674-1056/25/11/115201

Cited by 6 publications

(8 citation statements)

References 10 publications

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“…Unlike the QSF equilibrium with lower X-points, the QSF configuration with upper X-points needs high PF5 and PF13 currents, which may exceed the PF current limitation. [5] Thus, based on a typical QSF equilibrium (shoot 71543 at 3 s) with I p = 250 kA, the new equilibrium is designed with PF5 and PF13 currents set as target functions and elongation, top triangularity dtop, and squareness set as variables. The design result is shown in Table 4.…”

Section: Methodsmentioning

confidence: 99%

“…[4] The simulation and experimental results of QSF discharges show that the poloidal flux distribution is not steady around the X-point region. [5] Additionally, the formation of QSF configuration with lower X-point needs high currents of PF6 and PF14, which may exceed the limitation of the coils. QSF configuration design and control remain challenging.…”

Section: Introductionmentioning

confidence: 99%

“…In the EAST QSF configuration design, it is found that the desired PF6 and PF14 currents usually exceed the PF current limitation (12 kA/turn). [5] Obviously, this is not suitable. By modifying the plasma shape, the PF currents can be changed.…”

Section: Introductionmentioning

confidence: 99%

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Plasma shape optimization for EAST tokamak using orthogonal method

Chen¹,

Bao²,

Fu³

et al. 2019

Chinese Phys. B

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It is necessary to reduce the currents of poloidal field (PF) coils as small as possible, during the static equilibrium design procedure of Experimental Advanced Superconductive Tokamak (EAST). The quasi-snowflake (QSF) divertor configuration is studied in this paper. Starting from a standard QSF plasma equilibrium, a new QSF equilibrium with 300 kA total plasma current is designed. In order to reduce the currents of PF6 and PF14, the influence of plasma shape on PF coil current distribution is analyzed. A fixed boundary equilibrium solver based on a non-rigid plasma model is used to calculate the flux distribution and PF coil current distribution. Then the plasma shape parameters are studied by the orthogonal method. According to the result, the plasma shape is redefined, and the calculated equilibrium shows that the currents of PF6 and PF14 are reduced by 3.592 kA and 2.773 kA, respectively.

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Section: Methodsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Plasma shape optimization for EAST tokamak using orthogonal method

Chen¹,

Bao²,

Fu³

et al. 2019

Chinese Phys. B

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“…The TSC studies the evolution of magnetic field in a rectangular computational domain using the Maxwell MHD equations for the plasma, coupled with the boundary conditions to the circuit equations for the poloidal field (PF) coils. [11,12] It could generate evolution data during the whole discharge process at a time interval of 1 millisecond. The high time density and stability of the data make it suitable for incipient neural network model training.…”

Section: Generation Of Training and Testing Databasementioning

confidence: 99%

Estimation of plasma equilibrium parameters via a neural network approach*

et al. 2019

Self Cite

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Plasma equilibrium parameters such as position, X-point, internal inductance, and poloidal beta are essential information for efficient and safe operation of tokamak. In this work, the artificial neural network is used to establish a non-linear relationship between the measured diagnostic signals and selected equilibrium parameters. The estimation process is split into a preliminary classification of the kind of equilibrium (limiter or divertor) and subsequent inference of the equilibrium parameters. The training and testing datasets are generated by the tokamak simulation code (TSC), which has been benchmarked with the EAST experimental data. The noise immunity of the inference model is tested. Adding noise to model inputs during training process is proved to have a certain ability for maintaining performance.

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“…[9,22] Besides the RT-EFIT, [2] the P-EFIT provides another routine real-time plasma equilibrium reconstruction method, which has better spatial resolution and customized modules for plasma control in EAST. The P-EFIT supports a snowflake configuration, the MIMO plasma shape control experiments [23] in EAST, and has some preliminary results in DIII-D. [24]…”

Section: -6mentioning

confidence: 99%

Fast parallel Grad–Shafranov solver for real-time equilibrium reconstruction in EAST tokamak using graphic processing unit

2017

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To achieve real-time control of tokamak plasmas, the equilibrium reconstruction has to be completed sufficiently quickly. For the case of an EAST tokamak experiment, real-time equilibrium reconstruction is generally required to provide results within 1ms. A graphic processing unit (GPU) parallel Grad-Shafranov (G-S) solver is developed in P-EFIT code, which is built with the CUDA™ architecture to take advantage of massively parallel GPU cores and significantly accelerate the computation. Optimization and implementation of numerical algorithms for a block tri-diagonal linear system are presented. The solver can complete a calculation within 16 µs with 65×65 grid size and 27 µs with 129×129 grid size, and this solver supports that P-EFIT can fulfill the time feasibility for real-time plasma control with both grid sizes.

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A divertor plasma configuration design method for tokamaks

Cited by 6 publications

References 10 publications

Plasma shape optimization for EAST tokamak using orthogonal method

Plasma shape optimization for EAST tokamak using orthogonal method

Estimation of plasma equilibrium parameters via a neural network approach*

Fast parallel Grad–Shafranov solver for real-time equilibrium reconstruction in EAST tokamak using graphic processing unit

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