Fusion Engineering and Design最新文献_第6页

Efficient calculation of magnetic fields from ferromagnetic materials near strong electromagnets, and application to stellarator coil optimization 强电磁铁附近铁磁材料磁场的有效计算及其在仿星器线圈优化中的应用

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-10 DOI: 10.1016/j.fusengdes.2026.115627

Matt Landreman , Humberto Torreblanca , Antoine Cerfon

In fusion reactor design, steels under consideration for the blanket are ferromagnetic, so the steel’s effect on the plasma physics must be examined. For efficient calculation of these fields, we can exploit the fact that the ferromagnetic material gives a small perturbation relative to the fields from the electromagnetic coils and plasma. Moreover the magnetization is saturated due to the strong fields in typical fusion systems. These approximations significantly reduce the nonlinearity of the problem, so the magnetic materials can be described by an array of point dipoles of known magnitude, oriented in the direction of the coil and plasma field. The approach is verified by comparison to finite-element calculations with commercial software and shown to be accurate. As no linear or nonlinear solve is required, only evaluation of Biot–Savart-type integrals, the method here is significantly simpler to implement than other methods, and extremely fast. The method is compatible with arbitrary CAD geometry, and also allows rapid computation of the magnetic forces. We demonstrate adding the ferromagnetic effects to free-boundary magnetohydrodynamic equilibrium calculations, assessing the effect on plasma physics properties such as confinement and stability. Moreover, it is straightforward to differentiate through the model to get the derivative of the field with respect to the electromagnet parameters. We thereby demonstrate gradient-based coil optimization for a quasi-isodynamic stellarator in which the field contribution from a ferromagnetic blanket is included. Even a significant steel volume is found to have little impact on the plasma physics properties, with the main effects being a slight destabilization of ballooning modes and a radial shift of the edge islands due to decrease in rotational transform. Both of these issues are corrected by the minor reoptimization of the coil shapes to account for the field from the steel.

在核聚变反应堆设计中，考虑用于包层的钢是铁磁性的，因此必须检查钢对等离子体物理的影响。为了有效地计算这些场，我们可以利用铁磁材料相对于电磁线圈和等离子体产生的场有一个小的扰动这一事实。此外，在典型的核聚变系统中，由于强磁场的作用，磁化是饱和的。这些近似极大地降低了问题的非线性，因此磁性材料可以用已知量级的点偶极子阵列来描述，这些点偶极子在线圈和等离子体场的方向上取向。通过与商业软件的有限元计算对比，验证了该方法的准确性。由于不需要线性或非线性解，只需要求biot - savart型积分，因此该方法比其他方法实现起来简单得多，而且速度非常快。该方法兼容任意CAD几何形状，也允许快速计算磁力。我们演示了将铁磁效应添加到自由边界磁流体动力学平衡计算中，评估了对等离子体物理特性（如约束和稳定性）的影响。此外，通过模型求导得到电场对电磁铁参数的导数是很简单的。因此，我们展示了基于梯度的准等动力仿星器线圈优化，其中包括铁磁包层的场贡献。研究发现，即使钢的体积很大，对等离子体物理特性的影响也很小，主要影响是气球模式的轻微不稳定和由于旋转变换的减少而导致的边缘岛的径向位移。这两个问题都可以通过对线圈形状的微小重新优化来纠正，以考虑来自钢的磁场。

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引用次数: 0

The effect of minority heating on the electron temperature profile recovery using ICRH for future real-time control applications in tokamak plasmas 少数加热对利用ICRH恢复电子温度分布的影响，用于未来托卡马克等离子体的实时控制应用

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-17 DOI: 10.1016/j.fusengdes.2025.115555

M. Cappelli , A. Cardinali , V.K. Zotta , G. Pucella , M. Brambilla , S. Gabriellini , R. Gatto , M. Zerbini , L. Garzotti , D. Van Eester , JET contributors , WPTE Team

Real-time control using Ion Cyclotron Resonance Heating (ICRH) has been proposed in JET operational scenarios to counteract temperature hollowing effects. Specifically, in cases of hollow electron temperature profiles, central ion cyclotron resonance heating could be employed to restore temperature peaking based on real-time Electron Cyclotron Emission (ECE) data. ICRH has been utilized to optimize the plasma ramp-down process, correcting the discharge's end and preventing plasma disruption. Before designing the real-time controller, it is necessary to carefully evaluate the ability of the ICRH to recover the temperature profile by depositing the power emitted in the desired way. For this purpose, the presented work conducted simulations of a JET discharge to evaluate power deposition using a full wave code (TORIC). To quantify the power transferred from hydrogen ions to electrons, a quasi-linear analysis was conducted. The effects of ICRH application on the power balance were assessed through predictive transport analysis using the JINTRAC suite of codes. The integrated study's findings demonstrate the potential of utilizing ICRH alongside ECE measurements for real-time control of the electron temperature profile, offering valuable insights for future plasma control strategies and advanced tokamak operation.

利用离子回旋共振加热（ICRH）对射流进行实时控制，以抵消温度空穴效应。具体来说，在空心电子温度分布情况下，基于电子回旋发射（ECE）实时数据，可以采用中心离子回旋共振加热来恢复温度峰值。ICRH已被用于优化等离子体下降过程，纠正放电结束并防止等离子体中断。在设计实时控制器之前，有必要仔细评估ICRH通过以期望的方式沉积发出的功率来恢复温度分布的能力。为此，本研究利用全波码（TORIC）对JET放电进行了模拟，以评估功率沉积。为了量化氢离子向电子传递的能量，进行了准线性分析。通过使用JINTRAC代码套件进行预测输运分析，评估ICRH应用对功率平衡的影响。综合研究结果表明，利用ICRH和ECE测量实时控制电子温度分布的潜力，为未来的等离子体控制策略和先进的托卡马克操作提供了有价值的见解。

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引用次数: 0

Timing control system design of stellarator based on virtual instrument 基于虚拟仪器的仿星器定时控制系统设计

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-08 DOI: 10.1016/j.fusengdes.2026.115625

Jiajing Hua, Puqiong Yang, Xianghui Yin, Yushan Zhou, Yulin Wang

CN-H1 is a three-cycle quasi-spiral symmetric star simulator. The timing control system needs to manage the starting, running and closing sequence of the coil power supply, heating system, diagnosis system and other subsystems in a unified way, and the clock error is required to be controlled within ±2 microseconds to ensure the coordinated operation of each subsystem. This paper designs and implements a distributed timing control system centered on virtual instruments in response to the strict requirements for high-precision and multi-channel timing control during the plasma discharge of stellarators. The system takes LabVIEW FPGA as the core execution module and realizes hardware-level parallel logic by using its graphical programming. Aiming at the complex timing logic of the stellarator, a hybrid programming model of "state machine - event" was designed to decompose the complex discharge process into configurable time segments, so as to achieve the purpose of real-time switching and dynamic adjustment of waveform segments during the experiment. The test results show that the nanosecond-level absolute delay of the system signal is completely transparent to millisecond-level applications. Meanwhile, the waveform synchronization error of any two channels of the system is better than 55ns. The system can stably generate complex timing waveforms with millisecond-level cycles, and the timing accuracy and synchronization performance significantly exceed those of conventional solutions. This design has the advantages of multi-channel output, simple operation and strong real-time performance. It fully meets the timing control requirements during the discharge of CN-H1 and has important engineering application value.

CN-H1是一个三圆准螺旋对称星模拟器。定时控制系统需要对线圈电源、加热系统、诊断系统等子系统的启动、运行、合闸顺序进行统一管理，并要求时钟误差控制在±2微秒以内，以保证各子系统的协调运行。针对仿星器等离子体放电过程中对高精度、多通道定时控制的严格要求，本文设计并实现了以虚拟仪器为核心的分布式定时控制系统。该系统以LabVIEW FPGA为核心执行模块，利用其图形化编程实现硬件级并行逻辑。针对仿星器复杂的时序逻辑，设计了一种“状态机-事件”混合规划模型，将复杂的放电过程分解为可配置的时间段，从而达到实验过程中波形段实时切换和动态调整的目的。测试结果表明，系统信号的纳秒级绝对延迟对毫秒级应用是完全透明的。同时，系统任意两个通道的波形同步误差均优于55ns。该系统可以稳定地产生毫秒级周期的复杂时序波形，其时序精度和同步性能明显优于传统方案。本设计具有多通道输出、操作简单、实时性强等优点。完全满足CN-H1放电时的定时控制要求，具有重要的工程应用价值。

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引用次数: 0

Optical boundary reconstruction with visible/infrared integrated imaging systems on the HL-3 tokamak HL-3托卡马克上可见光/红外综合成像系统的光学边界重建

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-31 DOI: 10.1016/j.fusengdes.2025.115609

M.Y. He , J.M. Gao , X.Q. Ji , T.F. Sun , A. Wang , B.T. Cui , H.L. Du , J.X. Li , L. Liu , G.Z. Hao

The plasma boundary has been reconstructed using integrated multispectral optical imaging systems on the HL-3 tokamak, with particular emphasis on the divertor region. In addition to the mid-plane visible imaging system, which is commonly used to reconstruct the main plasma boundary, a new lower divertor visible and infrared imaging system has been developed to reconstruct the locations of the X-point and the strike points. It increases the accuracy of reconstructed plasma boundary, achieving precision of approximately 10 mm. Simulation results demonstrate that the averaged reconstructed error of the optical boundary is within a few millimeters. Finally, the reconstructed optical plasma boundary shows strong potential for applications in plasma diagnostics and equilibrium analysis.

利用HL-3托卡马克上的集成多光谱光学成像系统对等离子体边界进行了重建，重点研究了转向器区域。除了通常用于重建等离子体主边界的中平面可见成像系统外，还开发了一种新的下分流器可见和红外成像系统，用于重建x点和打击点的位置。它提高了重建等离子体边界的精度，达到了约10 mm的精度。仿真结果表明，光学边界的平均重构误差在几毫米以内。最后，重建的光学等离子体边界在等离子体诊断和平衡分析方面显示出强大的应用潜力。

引用次数: 0

Design and simulation of a beam extraction and acceleration system for a small high-intensity neutron generator 小型高强度中子发生器束流提取与加速系统的设计与仿真

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-30 DOI: 10.1016/j.fusengdes.2025.115606

Jingtian Xu , Wen Wang , Qi Yang , Minghuang Wang , Shiyou Yang , FDS Consortium

A small high-intensity neutron generator imposes constraints on the beam extraction and acceleration system, which can deliver a deuterium/tritium (D/T) mixed ion beam with an energy of more than 200 keV and a current of more than 126 mA to the target within a limited space of 500 mm in length and 60 mm in radius. The beam at the target should have a spot radius of less than 40 mm and a peak current density of less than 50 A/m². In this work, beam transport simulations were conducted using the IBSIMU code. A preliminary design was obtained by iteratively optimizing the electrode geometry, which enables the transportation of a 200 keV, 126 mA D-T beam with a spot radius of 37 mm and a peak current density of 49.2 A/m² at the target. Based on this design, the effects of key geometric parameters, including the extraction gap, extraction aperture radius, acceleration gap, and acceleration aperture radius, on the beam spot radius and peak current density, were systematically analyzed. The results indicate that the variations in the extraction and acceleration gaps significantly affect the beam focusing condition, thus exerting a strong influence on the beam spot size and beam distribution. Under-focused transport conditions are more favorable for meeting the design requirements of the neutron generator. Variations in the extraction aperture radius and acceleration aperture radius do not modify the beam focusing condition and only marginally affect the beam spot and density, thereby allowing fine adjustments to be made according to practical requirements.

一个小型的高强度中子发生器对束流提取和加速系统施加了限制，该系统可以在长度为500mm、半径为60mm的有限空间内向目标输送能量超过200kev、电流超过126ma的氘/氚（D/T）混合离子束。激光束的光斑半径应小于40mm，峰值电流密度应小于50a /m²。在这项工作中，使用IBSIMU代码进行了波束输运模拟。通过对电极几何形状的迭代优化，得到了一个初步的设计方案，该设计方案可使200 keV、126 mA、光斑半径为37 mm、峰值电流密度为49.2 A/m2的D-T束流在目标处传输。在此基础上，系统分析了提取间隙、提取孔径半径、加速间隙和加速孔径半径等关键几何参数对光束光斑半径和峰值电流密度的影响。结果表明，提取间隙和加速间隙的变化会显著影响光束聚焦条件，从而对光束光斑大小和光束分布产生强烈影响。欠聚焦输运条件更有利于满足中子发生器的设计要求。提取孔径半径和加速孔径半径的变化不会改变光束聚焦条件，只会对光束光斑和密度产生轻微影响，可以根据实际需要进行微调。

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引用次数: 0

Convective heat transfer of high-temperature nitrogen in the geometrically intricate EAST divertor: experimental study and correlation for baking process 几何复杂的EAST导流器中高温氮气的对流换热：烘烤过程的实验研究与关联

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-12 DOI: 10.1016/j.fusengdes.2025.115590

Zhe Liu , Peng Fu , Bin Guo

To enhannce the baking efficiency and improve operational safety of the EAST tokamak, this work experimentally investigates the convective heat transfer characteristics of high-temperature nitrogen flowing through the geometrically intricate and confined passages of the EAST divertor under baking process. A dedicated experiment system was designed and constructed to cover the full operational parameter range of the EAST baking process. The experimental investigations covered full parameter range of the EAST baking process. Results demonstrate that both the convective heat transfer coefficient and the inlet-outlet temperature difference in the test section are significantly governed by the combined effects of the inlet temperature and flow velocity. A modified convective heat transfer correlation was developed by extending the classical formulation, based on statistical analysis of the experimental data. Validation results show that predictions using the modified correlation agree well with experimental measurements, within deviations consistently constrained within

\pm 10 %

. The findings provide essential foundations for baking efficiency optimization in fusion devices.

为了提高EAST托卡马克的焙烧效率和运行安全性，本文对高温氮气在焙烧过程中流经EAST导流器几何形状复杂且密闭通道的对流换热特性进行了实验研究。设计并搭建了一个专门的实验系统，涵盖了EAST烘焙过程的全部操作参数范围。实验研究涵盖了EAST烘烤过程的全部参数范围。结果表明：试验段的对流换热系数和进出口温差受入口温度和流速的综合影响显著；在对实验数据进行统计分析的基础上，对经典公式进行了推广，建立了一个修正的对流换热关系式。验证结果表明，使用修正相关性的预测与实验测量结果吻合良好，偏差一致限制在±10%以内。研究结果为优化核聚变装置的烘烤效率提供了必要的基础。

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引用次数: 0

Comparison of remote access technologies for research facilities using EPICS/CSS. Application to particle accelerator experiments 基于EPICS/CSS的科研设施远程访问技术比较。应用于粒子加速器实验

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-23 DOI: 10.1016/j.fusengdes.2025.115595

Javier Cruz-Miranda , Manuel Rodríguez-Álvarez , Miguel Damas , Iván Casero-Santos , Iván Podadera-Aliseda , José Franco-Campos , Antti Jokinen , André Sancho-Duarte , Javier Díaz

This study compares and proposes new alternatives for remotely connecting to visualize the experiments occurring in a particle accelerator located in Rokkasho, Japan. Three different platforms have been considered for remote access: the existing X2GO client, Guacamole with Control System Studio (CSS), and Phoebus web. While X2GO is a standard client for remote access to a server desktop, the other two platforms are proposed to improve the access, the response time, and the user experience for the researchers The servers for this study and the Operator Interfaces (OPIs) have been placed in our laboratory located in Granada, Spain, and the accelerator data, by means of Process Variables (PVs), were obtained via a VPN. Additionally, these platforms have been tested in two ways: with direct access to the PV data for each connection and using a local EPICS (Experimental Physics Industrial Control System) Gateway. The results prove that these new platforms, with a stable connection to the accelerator, could eventually enhance access to the experiments and balance the load of researchers connecting to the facility. This would allow the international team of researchers to participate in experiments as if they were physically in the control room.

本研究比较并提出了远程连接的新方案，以可视化在位于日本六所所的粒子加速器中发生的实验。考虑了三种不同的远程访问平台：现有的X2GO客户端、Guacamole with Control System Studio （CSS）和Phoebus web。虽然X2GO是远程访问服务器桌面的标准客户端，但我们提出了另外两个平台来改善访问，响应时间和研究人员的用户体验。本研究的服务器和操作员接口（opi）已放置在我们位于西班牙格拉纳达的实验室中，加速器数据通过过程变量（pv）通过VPN获得。此外，这些平台已经通过两种方式进行了测试：直接访问每个连接的光伏数据，以及使用本地EPICS（实验物理工业控制系统）网关。结果证明，这些与加速器稳定连接的新平台最终可以增加对实验的访问，并平衡连接到该设施的研究人员的负载。这将允许国际研究团队参与实验，就好像他们在控制室一样。

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引用次数: 0

Design and analysis of PCHE-type recuperator for helium cooling system in CN HCCB TBS CN HCCB TBS氦冷却系统pche型回热器的设计与分析

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2025-12-27 DOI: 10.1016/j.fusengdes.2025.115603

Zhengning Zhao , Xinghua Wu , Jie Liu , Xingfu Ye , Hong Yang , Xiaoyu Wang

The Helium Cooling System (HCS) is an important ancillary system of the Chinese Helium-Cooled Ceramic Breeder Test Blanket System (CN HCCB-TBS) that provides cooling to remove heat from the fusion reactor blanket during plasma operation. The HCS is an “8″-shaped loop, in which a circulator provides the pressure head for the helium, while two heat exchangers and a heater are arranged to convert heat during operation. The recuperator is positioned at the center of the loop to transfer heat between the cold and hot helium streams, thereby enabling energy recovery and reducing loop energy consumption. Consequently, the design and selection of the recuperator significantly influence the operational stability and energy balance of the HCS. Printed Circuit Heat Exchangers (PCHEs) exhibit superior performance in terms of high-temperature and high-pressure capability, thermal efficiency, compactness, and operational reliability, making them widely applicable in petrochemical and hydrogen energy systems. This article presents the design and analysis of a PCHE-type recuperator based on specified requirements and project experience, providing valuable support for the design and manufacturing of future helium cooling loops and related applications.

氦冷却系统（HCS）是中国氦冷却陶瓷增殖试验包层系统（CN HCCB-TBS）的重要辅助系统，在等离子体运行过程中为聚变反应堆包层提供冷却以去除热量。HCS是一个“8″”形状的回路，其中一个循环器为氦气提供压力头，而两个热交换器和加热器则在运行过程中转换热量。回热器位于回路中心，在冷氦流和热氦流之间传递热量，从而实现能量回收，降低回路能耗。因此，回热器的设计和选择对HCS的运行稳定性和能量平衡有重要影响。印刷电路热交换器（PCHEs）在高温高压性能、热效率、紧凑性和运行可靠性方面表现出优异的性能，广泛应用于石化和氢能系统。本文根据规定的要求和项目经验，介绍了pche型回热器的设计和分析，为未来氦气冷却回路的设计和制造及相关应用提供了有价值的支持。

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引用次数: 0

Neutronic modeling and physical analysis of the neutron yield measurement system for the HL-3 Tokamak HL-3托卡马克中子产率测量系统的中子建模与物理分析

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-07 DOI: 10.1016/j.fusengdes.2026.115614

Zuowei Wen, Lei Feng, Guoliang Yuan, Wei Zhao, Jiawei Shi, Fengzhao Shen

The measurement of neutron yield plays a critical role in fusion power assessment and safety control for tokamak devices. This paper presents a comprehensive performance evaluation and system design of the neutron yield measurement system for the HL-3 tokamak, based on neutronic modeling with MCNP. The reliability of the model was validated against data from in-situ calibration experiments, showing good agreement between simulated and measured absolute detector efficiencies. The system comprises eight fission chambers, designed to cover a neutron yield range from 10¹⁰ to 10¹⁹n/s for both DD and DT discharge scenarios. Simulations were performed to analyze the effects of moderator material, reflected neutron contribution, and plasma displacement on measurement performance. The results indicate that polyethylene as a moderator provides a sufficiently flat sensitivity. The contribution of neutrons reflected by the bio-shielding wall to the detector sensitivity was found to be non-negligible. Plasma displacement has a minimal impact on detector sensitivity and does not significantly alter the system measurement range. Ultimately, the performance of the HL-3 neutron yield measurement system fully meets the physics design requirements and will provide reliable support for neutron yield and fusion power measurement during HL-3 experiments.

中子产率的测量在托卡马克装置的聚变功率评估和安全控制中起着至关重要的作用。本文介绍了基于MCNP中子模型的HL-3托卡马克中子产率测量系统的综合性能评价和系统设计。根据现场标定实验数据验证了模型的可靠性，结果表明模拟的绝对探测器效率与实测的绝对探测器效率吻合良好。该系统由8个裂变室组成，设计的中子产率范围从1010到1019n/s，适用于DD和DT放电方案。模拟分析了慢化剂材料、反射中子贡献和等离子体位移对测量性能的影响。结果表明，聚乙烯作为慢化剂提供了足够的平坦灵敏度。发现生物屏蔽壁反射的中子对探测器灵敏度的贡献是不可忽略的。等离子体位移对探测器灵敏度的影响很小，也不会显著改变系统的测量范围。最终，HL-3中子产率测量系统的性能完全满足物理设计要求，将为HL-3实验中的中子产率和聚变功率测量提供可靠的支持。

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引用次数: 0

Influence of magnetic fields generated by a magnetized ferritic first wall on surface heat loads from plasma heat flux along magnetic field lines 磁化铁素体第一壁产生的磁场对沿磁力线等离子体热通量表面热负荷的影响

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design

Pub Date : 2026-03-01 Epub Date: 2026-01-08 DOI: 10.1016/j.fusengdes.2025.115601

Yuya Miyoshi, Yushiro Yamashita, Weixi Chen

Localized concentrations of plasma heat flux (comprising charged particles moving along magnetic field lines) can result in excessive thermal loads whose peak values may exceed several MW/m², e.g. at the edges of blanket modules. Such conditions are undesirable and have motivated the development of heat load analysis methods, including magnetic field line tracing within the vacuum vessel (VV), as established in our previous work. In JA DEMO, reduced-activation ferritic martensitic steel is employed for the first wall (FW) due to its superior resistance to neutron irradiation. Here, the strong magnetic field in JA DEMO magnetizes the FW, thereby altering the magnetic field configuration in VV and affecting plasma equilibrium. This modified equilibrium, in turn, influences the magnetization vector in FW. Accurate prediction of magnetic field distribution in VV and the heat load distribution on the FW thus necessitates consideration of this mutual interaction between the magnetized FW and the plasma equilibrium. To address this, a computational code capable of evaluating the effect of FW magnetization under JA DEMO-like conditions is developed. The code iteratively computes a three-dimensional MHD equilibrium consistent with the magnetic field generated by the magnetized FW (

{\vec{B}}_{m}

). Subsequently, plasma heat flux and heat load distributions on the FW are calculated via the magnetic field line tracing. Although

{\vec{B}}_{m}

is relatively weak and it induces unnoticeable changes in plasma equilibrium, it significantly alters the heat load distribution compared to cases neglecting

{\vec{B}}_{m}

. In this research, three distinct patterns of

{\vec{B}}_{m}

influence are identified: (1) cumulative effect of weak

{\vec{B}}_{m}

altering field line trajectories, (2) strong

{\vec{B}}_{m}

modifying field line orbit, and (3) strong

{\vec{B}}_{m}

directly attracting field lines toward the FW. Future work will focus on identifying the specific conditions under which these effects become significant.

局部浓度的等离子体热流（包括沿磁力线移动的带电粒子）可能导致过高的热负荷，其峰值可能超过几MW/m2，例如在包层模块的边缘。这种情况是不可取的，并且推动了热负荷分析方法的发展，包括真空容器（VV）内的磁场线追踪，正如我们之前的工作所建立的那样。在JA DEMO中，由于降低活化的铁素体马氏体钢具有优异的抗中子辐照性能，因此采用了降低活化的铁素体马氏体钢作为第一壁。在这里，JA DEMO中的强磁场磁化了FW，从而改变了VV中的磁场构型，影响了等离子体平衡。这种修正的平衡反过来又影响了FW中的磁化矢量。因此，要准确预测VV内的磁场分布和FW上的热负荷分布，就必须考虑磁化FW与等离子体平衡之间的相互作用。为了解决这个问题，开发了一个能够在JA演示条件下评估FW磁化效果的计算代码。该代码迭代计算与磁化FW （B→m）产生的磁场一致的三维MHD平衡。随后，通过磁场线示踪计算了FW上的等离子体热流密度和热负荷分布。虽然B→m相对较弱，引起等离子体平衡的不明显变化，但与忽略B→m的情况相比，它显著改变了热负荷分布。本研究确定了三种不同的B→m影响模式：(1)弱B→m改变场线轨迹的累积效应，(2)强B→m改变场线轨道，以及(3)强B→m直接吸引场线朝向FW。未来的工作将侧重于确定这些影响变得显著的具体条件。

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引用次数: 0