考虑 H 模式等离子体剖面的托卡马克系统代码核心等离子体功率和电流平衡模型的改进

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design Pub Date : 2025-04-17 DOI:10.1016/j.fusengdes.2025.115022

Shota Sugiyama , Takuya Goto , Hiroyasu Utoh , Yoshiteru Sakamoto

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引用次数: 0

摘要

我们提出了系统代码的模型来计算自举电流、中性束驱动电流和同步辐射损耗功率，并考虑了具有h模基座的等离子体轮廓。通过拟合磁流体动力学平衡计算程序中包含自举电流详细模型的数据，得到了抛物线型和h型等离子体剖面的自举电流分数的标度公式。所建立的中性束驱动电流的简单模型适用于任意的等离子体轮廓和形状。使用CYTRAN模块，仅使用系统代码中的参数，我们展示了通量面平均磁场强度、等离子体表面积和等离子体体积的几何近似，以计算同步辐射损耗功率。通过与详细模型计算结果的比较，验证了模型的有效性。这些模型使系统代码能够一致地考虑等离子体轮廓与核心等离子体功率和电流平衡中的基座，因为平衡方程中的其他数量可以通过计算数量的轮廓并在系统分析可接受的计算时间内对它们进行数值积分来评估。

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Improvement of core plasma power and current balance models for tokamak systems code considering H-mode plasma profiles

We propose the models for the systems codes to calculate the bootstrap current fraction, neutral beam-driven current, and synchrotron radiation loss power, considering the plasma profiles with the H-mode pedestals. The scaling formulae of the bootstrap current fraction are developed for the parabolic and H-mode plasma profiles by fitting the data prepared using a magnetohydrodynamic equilibrium calculation code including a detailed model of the bootstrap current. The developed simple model of the neutral beam-driven current is applicable to arbitrary plasma profiles and shapes. We show the geometric approximation of the flux surface averaged magnetic field strength, plasma surface area, and plasma volume to calculate the synchrotron radiation loss power using the CYTRAN module with only the parameters used in the systems codes. The validity of these models is confirmed by comparing the results to those calculated by the detailed models. These models enable the systems codes to consistently consider the plasma profiles with the pedestals in the core plasma power and current balances because the other quantities in the balance equations can be evaluated by calculating the profiles of the quantities and integrating them numerically within computing time acceptable for the systems analysis.

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来源期刊

Fusion Engineering and Design 工程技术-核科学技术

CiteScore

3.50

自引率

23.50%

发文量

275

审稿时长

3.8 months

期刊介绍： The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.