Fusion Engineering and Design最新文献_第8页

Managing machine safety challenges for over 1 MA H-mode plasma operation on the HL-3 tokamak 在HL-3托卡马克上管理超过1ma h模式等离子体操作的机器安全挑战

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

Fusion Engineering and Design

Pub Date : 2025-12-10 DOI: 10.1016/j.fusengdes.2025.115576

Xiaolong Liu, Qinghua Ren, Jie Xu, Hongbin Xu, Yong Lu, Jilai Hou, Wenyu Huang, Ruibao Jia, Jin Wang, Long Zhang, Xiaoquan Ji

The HL-3 tokamak has achieved plasma currents exceeding 1 MA in high-confinement H-mode operations, pushing the boundaries of fusion research and offering new scientific insights. However, this advancement poses significant engineering challenges related to machine safety due to increased thermal loads and electromagnetic forces. To address these challenges, an integrated real-time machine monitoring and safe operation system has been developed, featuring real-time monitoring, interlock protection, data analysis, and centralized control. This system oversees critical components such as coils, vacuum vessel, plasma-facing elements, and auxiliary systems like cooling and gas injection. A hydraulic preload system optimizes preload forces on the toroidal field coils, reducing mechanical stresses and enhancing structural integrity. To mitigate risks from uncontrolled plasma disruptions, a Shattered Pellet Injection (SPI) system has been implemented. This system injects pellets composed of hydrogen, deuterium, neon, or argon to rapidly dissipate plasma energy, significantly reducing mechanical stresses and potential damage. Operational results at plasma currents up to 1.6 MA demonstrate the effectiveness of these measures, with successful monitoring and control of displacements and accelerations in vital components. These advancements have effectively addressed safety challenges, allowing high-performance operation without compromising machine integrity. Future work will focus on safely increasing plasma currents beyond 1 MA and achieving higher toroidal magnetic fields, addressing new engineering challenges while maintaining safety as the paramount priority.

HL-3托卡马克在高约束h模操作中实现了超过1毫安的等离子体电流，推动了聚变研究的界限，并提供了新的科学见解。然而，由于热负荷和电磁力的增加，这一进步带来了与机器安全相关的重大工程挑战。为了应对这些挑战，开发了一套集实时监测、联锁保护、数据分析和集中控制为一体的机器实时监测和安全操作系统。该系统监督关键部件，如线圈，真空容器，等离子体面元件和辅助系统，如冷却和气体注入。液压预紧系统优化了环形磁场线圈上的预紧力，减少了机械应力，提高了结构的完整性。为了降低不受控制的等离子体中断的风险，已经实施了破碎颗粒注入（SPI）系统。该系统注入由氢、氘、氖或氩组成的颗粒，以迅速消散等离子体能量，显著降低机械应力和潜在损伤。在高达1.6 MA的等离子体电流下的运行结果证明了这些措施的有效性，并成功地监测和控制了重要部件的位移和加速度。这些进步有效地解决了安全挑战，在不损害机器完整性的情况下实现了高性能操作。未来的工作将集中在安全地增加等离子体电流超过1毫安，并实现更高的环向磁场，在保持安全为首要任务的同时解决新的工程挑战。

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引用次数: 0

An optimization approach for cryogenic distillation-based protium removal systems in magnetic fusion energy fuel cycles 磁聚变能燃料循环中基于低温蒸馏的脱丙系统的优化方法

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

Fusion Engineering and Design

Pub Date : 2025-12-09 DOI: 10.1016/j.fusengdes.2025.115534

Alex D. Somers, P. Arron Rowell, Collin R. Malone, Holly B. Flynn, George K. Larsen

Tritium inventory reduction for fusion power plants is central to the successful adoption of fusion energy. The advent of direct internal recycling in deuterium-tritium fusion fuel cycle design has led to significant reduction in startup tritium inventory estimates for fusion power plants but requires an isotope rebalancing and protium removal (IRPR) system to ensure acceptable isotopic fuel composition. Cryogenic distillation is a potential solution for near-term deployment in an IRPR system due to its demonstrated performance in similar operating regimes. Using protium removal in the fuel cycle as the primary performance metric, this paper presents an optimization methodology for a single-column continuous cryogenic distillation-based IRPR system for a 500 MW_fus magnetic fusion device. Distillation column optimization was performed using the CryOgenic Distillation For Isotopic Separation of Hydrogen (CODFISH) code developed at Savannah River National Laboratory. The distillation column design presented maximizes direct recycling of hydrogen isotopes from the fusion chamber exhaust to the fueling system while minimizing IRPR system steady-state tritium inventory. The optimized IRPR distillation column presented achieves direct recycling of 60 % of the hydrogen isotopes in the fusion chamber exhaust with an estimated steady-state system tritium inventory <30 g. The optimized IRPR distillation column operation was then used to estimate the design requirements for a detritation column to treat the IRPR system effluent stream.

减少核聚变发电厂的氚库存是成功采用核聚变能源的关键。氘-氚聚变燃料循环设计中直接内部循环的出现导致聚变发电厂启动氚库存估算显著降低，但需要同位素再平衡和除protium （IRPR）系统来确保可接受的同位素燃料组成。由于低温蒸馏在类似的操作系统中具有良好的性能，因此它是近期在IRPR系统中部署的潜在解决方案。以燃料循环中protium的去除为主要性能指标，提出了一种500 MWfus磁聚变装置单柱连续低温蒸馏IRPR系统的优化方法。精馏塔优化使用萨凡纳河国家实验室开发的氢同位素分离低温蒸馏（CODFISH）代码进行。精馏塔设计最大限度地从聚变室废气中直接回收氢同位素到燃料系统，同时最大限度地减少IRPR系统稳态氚库存。所提出的优化的IRPR精馏塔可以直接回收聚变室废气中60%的氢同位素，估计稳态系统氚库存为30 g。然后利用优化后的IRPR精馏塔操作来估计处理IRPR系统出水的精馏塔的设计要求。

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引用次数: 0

A robust predictive tool for the yield strength of irradiated RAFM steels for fusion engineering applications 用于核聚变工程应用的辐照RAFM钢屈服强度的可靠预测工具

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

Fusion Engineering and Design

Pub Date : 2025-12-09 DOI: 10.1016/j.fusengdes.2025.115572

YanBang Tang

Reduced-activation ferritic/martensitic (RAFM) steels are primary candidates for structural components in future fusion reactors, where their performance is critically dictated by radiation-induced hardening. Predicting this complex, multi-variable phenomenon is essential for material design and reactor safety. In this study, we developed a robust, automated machine learning (AutoML) framework to predict the yield strength of RAFM steels under various service conditions. A comprehensive database containing 1843 data points was established, incorporating 26 input features spanning chemical composition, irradiation dose, irradiation temperature, helium concentration, and mechanical testing temperature. After establishing a performance benchmark with over 40 conventional algorithms, the AutoML framework was deployed to systematically construct a multi-layer stacked ensemble model. This advanced approach demonstrated superior and more stable predictive fidelity across multiple unseen test sets, surpassing all individual models. Permutation-based feature importance analysis revealed that irradiation conditions, particularly dose and testing temperature, are the most dominant factors governing yield strength. Among the alloying elements, tungsten (W), molybdenum (Mo), and tantalum (Ta) were identified as the most critical contributors. This work not only provides a high-precision predictive tool for assessing the mechanical integrity of RAFM steels but also offers quantitative, data-driven insights into the hierarchy of factors controlling radiation hardening.

低活化铁素体/马氏体（RAFM）钢是未来核聚变反应堆结构部件的主要候选材料，其性能在很大程度上取决于辐射诱发硬化。预测这种复杂的、多变量的现象对于材料设计和反应堆安全至关重要。在这项研究中，我们开发了一个鲁棒的自动机器学习（AutoML）框架来预测RAFM钢在各种使用条件下的屈服强度。建立了包含1843个数据点的综合数据库，包含化学成分、辐照剂量、辐照温度、氦浓度、力学测试温度等26个输入特征。在建立了40多种传统算法的性能基准后，部署AutoML框架系统地构建了多层堆叠集成模型。这种先进的方法在多个未见过的测试集上证明了优越和更稳定的预测保真度，超越了所有单个模型。基于排列的特征重要性分析表明，辐照条件，特别是剂量和试验温度，是影响屈服强度的最主要因素。在合金元素中，钨(W)、钼（Mo）和钽（Ta）是最重要的贡献者。这项工作不仅为评估RAFM钢的机械完整性提供了高精度的预测工具，而且还为控制辐射硬化的因素层次提供了定量的、数据驱动的见解。

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引用次数: 0

Evaluating solid-state neutron detectors for measuring 14 MeV neutrons at high temperatures 评价高温下测量14mev中子的固态中子探测器

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

Fusion Engineering and Design

Pub Date : 2025-12-08 DOI: 10.1016/j.fusengdes.2025.115586

Q. Potiron , C. Destouches , M. Houry , O. Llido , A. Lyoussi , C. Reynard-Carette , L. Dubus , P. Malard , P. Legou , B. Cheymol

Silicon Carbide 4H Polytype (4H-SiC) and Diamond wide bandgap semiconductors are promising detector materials for fusion environments. Threshold energy nuclear reactions provide information on the energy of impinging fast neutrons and the combination of low intrinsic carrier concentration with high thermal conductivity makes these semiconductors suitable for high-temperature applications, especially for neutron monitoring in tritium production through ITER breeding blankets. While the carrier properties of SiC and Diamond offer interesting charge collection dynamics from room temperature up to 200 °C, the stability of their detection performance at high temperatures above 200 °C remains to be confirmed. To investigate this, we conducted a measurement campaign in a fast neutron field representative of fusion reactors at the GENESIS (Generator of Neutrons for Science and IrradiationS) research platform of LPSC (Laboratoire de Physique Subatomique et de Cosmologie) laboratory in Grenoble, France. Both 4H-SiC and Diamond sensors were irradiated with 14 MeV fast neutrons from a D-T neutron generator while encapsulated in a heating device, recording current signals from room temperature up to 500 °C. Using a direct measurement method of charge carrier collection dynamics as a function of applied bias voltage and temperature by pulse shape analysis provided information on velocity drift and collected charge. The results offer a first representative study of charge carrier mobility behavior with increasing temperature up to 500 °C. The stability of performance in terms of CCE (charge collection efficiency) has been demonstrated for SiC from room temperature up to 500 °C, while Diamond experiences a CCE drop of 60% between 200 °C and 300 °C.

碳化硅4H多型（4H- sic）和金刚石宽禁带半导体是很有前途的聚变环境探测器材料。阈能核反应提供了关于撞击快中子能量的信息，低固有载流子浓度和高导热率的结合使这些半导体适合于高温应用，特别是用于通过ITER增殖毯生产氚的中子监测。虽然SiC和金刚石的载流子性质在室温至200°C的范围内提供了有趣的电荷收集动力学，但它们在200°C以上的高温下检测性能的稳定性仍有待证实。为了研究这一点，我们在法国格勒诺布尔LPSC（物理、亚原子和宇宙实验室）实验室的GENESIS（用于科学和辐射的中子发生器）研究平台上的聚变反应堆代表的快中子场进行了测量活动。将4H-SiC和Diamond传感器封装在加热装置中，用来自D-T中子发生器的14 MeV快中子照射，记录室温至500°C的电流信号。通过脉冲形状分析，直接测量电荷载流子收集动态随外加偏置电压和温度的变化，提供了速度漂移和电荷收集的信息。结果提供了第一个有代表性的研究电荷载流子迁移行为随着温度升高到500°C。SiC的CCE（电荷收集效率）性能在室温至500°C范围内保持稳定，而金刚石的CCE在200°C至300°C范围内下降60%。

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引用次数: 0

Design and feasibility verification of a novel metal lip seal for large-diameter detachable vacuum interfaces in fusion reactors 一种新型大直径可拆卸真空界面金属唇密封的设计与可行性验证

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

Fusion Engineering and Design

Pub Date : 2025-12-08 DOI: 10.1016/j.fusengdes.2025.115571

Qingzhou Yu , Genmu Shi , Shilin Chen , Zhaoxi Chen , Qingxi Yang , Hao Xu

This paper presents a novel metal seal designed for large-diameter detachable vacuum interfaces in future fusion reactors, addressing stringent requirements of high irradiation, reliability, and repeated assembly. The design features a dual-lip mating structure incorporating an angular adjustment mechanism, which enhances tolerance to the flatness of the flange surface while ensuring sealing repeatability. An optimization model based on response surface methodology (RSM) is constructed to systematically evaluate the influence significance and sensitivity of key structural parameters on the gap between lips and pretightening force, and to determine the optimal parameter set. Simulation results demonstrate that the optimized seal achieves excellent lip contact and exhibits favorable elastic response characteristics. Further simulations considering weld-trim-induced lip shortening suggest that the seal structure supports approximately 10 repeated sealing cycles, meeting the requirements for reliable sealing performance under repeated disassembly scenarios in future fusion reactors. To validate manufacturability and structural robustness, a 6 m sector prototype is fabricated via multi-roller cold bending and ring forming, and its geometric accuracy and magnetic characteristics are evaluated. Experimental results show precise lip dimensions within design tolerances and low relative magnetic permeability. Furthermore, sealing performance tests under ideal alignment and induced misalignment (up to 5 mm) demonstrate that the lip seal maintains effective contact and excellent vacuum tightness, with maximum local leakage rate below 5 × 10^-11 Pa·m³/s and average leakage rate per unit length below 1 × 10^-10 Pa·m³/(s·m) for helium. This study confirms the reliability, manufacturability, and practical applicability of the proposed lip seal, providing technical guidance for the design and implementation of large-scale vacuum seals in fusion reactors.

本文提出了一种新型的金属密封，用于未来聚变反应堆的大直径可拆卸真空界面，以满足高辐照、可靠性和重复组装的严格要求。该设计采用了带有角度调节机构的双唇配合结构，提高了对法兰表面平整度的容忍度，同时保证了密封的重复性。建立了基于响应面法（RSM）的优化模型，系统评价了关键结构参数对唇间隙和预紧力的影响重要性和敏感性，确定了最优参数集。仿真结果表明，优化后的密封具有良好的唇部接触性能和良好的弹性响应特性。进一步的模拟表明，考虑到焊接装饰引起的唇缩短，密封结构支持大约10次重复密封循环，满足未来聚变反应堆在反复拆卸情况下可靠密封性能的要求。为了验证其可制造性和结构稳健性，通过多辊冷弯和环形成形工艺制作了一个6 m扇形原型，并对其几何精度和磁性特性进行了评估。实验结果表明，唇形尺寸在设计公差范围内精度高，相对磁导率低。此外，在理想对中和诱导错中（不超过5 mm）下的密封性能试验表明，唇形密封保持了有效的接触和良好的真空密封性，氦气的最大局部泄漏率低于5 × 10-11 Pa·m3/s，单位长度的平均泄漏率低于1 × 10-10 Pa·m3/（s·m）。本研究证实了所提出的唇密封的可靠性、可制造性和实用性，为聚变反应堆大型真空密封的设计和实施提供了技术指导。

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引用次数: 0

Improvements to standard diagnostic preparation and data-quality monitoring in the TCV tokamak 改进了TCV托卡马克的标准诊断准备和数据质量监测

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

Fusion Engineering and Design

Pub Date : 2025-12-08 DOI: 10.1016/j.fusengdes.2025.115578

P. Molina-Cabrera, F. Pastore, A. Frank, L. Simons, A. Tourneur, C. Yildiz, B. Vincent, K. Verhaegh , C. Marini , M. Wensing, A. Ianchenko, A. Balestri, S. Ernst, S. Coda, U. Sheikh, TCV team

As modern fusion experiments continue to push the boundaries of fusion science, the number, complexity, and importance of standard diagnostics have increased. Ensuring the recording of high-quality data from standard diagnostics is a task of great importance, entrusted to Ph.D. students in the TCV tokamak. Students participate in the control room team as the ‘diagnostician of the day’ or diagnosticien du jour (DdJ). This paper presents recent improvements to the DdJ software routines that prepare standard diagnostic settings, display, and automatically monitor the quality of diagnostic data. Recent updates have automated gain preparation in several standard diagnostics, which has led to reduced saturation and minimized signal-to-noise losses in the digitization process. Refactoring has also brought important runtime improvements to automatic data check routines. Lastly, new gain-preparation routines have been implemented that predict changes in plasma temperature due to changes in external electron heating power to better prepare the Thomson Scattering diagnostic, resulting in reduced saturation compared with traditional gain-preparation routines. These improvements have been led by a multi-generational task force: the DdJ-Ninjas.

随着现代核聚变实验不断推动核聚变科学的发展，标准诊断的数量、复杂性和重要性都有所增加。确保从标准诊断中记录高质量数据是一项非常重要的任务，委托给TCV托卡马克的博士生。学生们作为“每日诊断专家”或“每日诊断专家”（DdJ）参加控制室小组。本文介绍了DdJ软件例程的最新改进，该例程准备标准诊断设置，显示和自动监控诊断数据的质量。最近的更新在几种标准诊断中实现了自动增益准备，从而降低了数字化过程中的饱和度并最大限度地减少了信噪比损失。重构还为自动数据检查例程带来了重要的运行时改进。最后，为了更好地制备汤姆逊散射诊断，我们实现了新的增益制备程序，可以预测由于外部电子加热功率的变化而引起的等离子体温度的变化，从而降低了与传统增益制备程序相比的饱和。这些改进是由一个多代任务小组领导的：DdJ-Ninjas。

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引用次数: 0

Simulation of runaway electron generation in the Day-0 scenario of DTT DTT第0天场景下失控电子生成的模拟

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

Fusion Engineering and Design

Pub Date : 2025-12-08 DOI: 10.1016/j.fusengdes.2025.115588

Enrico Emanuelli , Francesco Vannini , Matthias Hoelzl , Nina Schwarz , Eric Nardon , Vinodh Bandaru , Daniele Bonfiglio , Artur Kryzhanovskyy , Giuseppe Ramogida , Fabio Subba , JOREK Team

Formation of Runaway electrons (REs) during tokamak disruptions is a significant challenge in fusion research, as they can locally damage the plasma-facing components by applying thermal loads of tens of MJ per square meter, possibly leading to significant melting. This work investigates the current quench phase of disruptions and the likelihood of RE generation and multiplication in the Day-0 scenario (plasma current

I_{p} = 2

MA) of the Divertor Tokamak Test (DTT), using the non-linear magnetohydrodynamic code JOREK. Our results from 2D (toroidally symmetric) simulations indicate that, in this initial low-current scenario, RE generation is minimal to negligible when the impurities injected through disruption mitigation systems are adequately limited. This suggests that DTT’s early operational phase poses a low RE risk, contributing to operational safety in this regard before transitioning to full power scenarios (

I_{p} = 5.5

MA). In addition to providing an initial RE safety benchmark for DTT, this study lays the groundwork for further research at higher operational currents and for the estimation of heat loads caused by RE beams on plasma-facing components, essential for guiding the design and strategic placement of mitigation elements such as sacrificial limiters.

在托卡马克破坏过程中，失控电子（REs）的形成是核聚变研究中的一个重大挑战，因为它们可以通过施加每平方米几十兆焦耳的热负荷局部破坏面向等离子体的组件，可能导致严重的熔化。本研究使用非线性磁流体力学代码JOREK，研究了在转向托卡马克试验（DTT）的第0天场景（等离子体电流Ip=2 MA）中中断的电流灭灭阶段以及RE生成和增殖的可能性。我们的二维（环形对称）模拟结果表明，在初始的低电流情况下，当通过干扰缓解系统注入的杂质得到充分限制时，RE的生成最小到可以忽略不计。这表明，在过渡到全功率情景（Ip=5.5 MA）之前，DTT的早期运行阶段具有较低的RE风险，有助于在这方面提高运行安全性。除了为DTT提供初始的RE安全基准外，本研究还为在更高工作电流下的进一步研究以及对等离子体面向组件的RE光束引起的热负荷的估计奠定了基础，这对于指导诸如牺牲限制器等缓解元件的设计和战略放置至关重要。

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引用次数: 0

Demonstration of enhanced abnormal Plasma Current detection in KSTAR Fast Interlock System KSTAR快速联锁系统中增强异常等离子体电流检测的演示

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

Fusion Engineering and Design

Pub Date : 2025-12-07 DOI: 10.1016/j.fusengdes.2025.115579

Seung-Ju Lee , Myungkyu Kim , Jaesic Hong , Sang-won Yun , Taegu Lee , Sang-hee Hahn , Woong-Ryol Lee , Taehyun Tak

The KSTAR Fast Interlock System (FIS) has the primary role of protecting the devices installed in the vacuum vessel of tokamak such as Plasma Facing Components (PFCs) by immediately stopping the KSTAR heating devices, following the event handling actions of the Plasma Control System (PCS). Furthermore, the FIS assists the PCS event handling operations by redundantly detecting abnormal Plasma Current (IP) events. The initially implemented detection logic for the IP minimum fault event has been successfully evaluated and operated. In this paper, we implement another logic detecting the IP error fault event that the discrepancy between the target IP and the measured IP exceeds the criteria. As the architecture design, we assign more complicated tasks such as the waveform generation to the host server and the error fault-checking task requiring real-time operation to the target controller. Second, the Direct Memory Access (DMA) method for data communication is adopted; thus, the target controller can conduct the detection logic and the data communication in parallel without real-time performance degradation. Third, we design proper timing of the data communication for stable operation. On the host side, we employ ITER Real-Time Framework (RTF) technology for initiating the data communication with precise timing and controlling the precise execution cycle. Finally, we apply the bypass logic to prevent conflict with the same detecting operation of the PCS. We evaluate the functionality of the IP error fault detection logic in the KSTAR plasma experiments.

KSTAR快速联锁系统（FIS）的主要作用是保护安装在托卡马克真空容器中的设备，如等离子体面向组件（pfc），在等离子体控制系统（PCS）的事件处理操作之后立即停止KSTAR加热设备。此外，FIS通过冗余检测异常等离子体电流（IP）事件来协助PCS事件处理操作。初步实现的IP最小故障事件检测逻辑已成功评估并运行。在本文中，我们实现了另一种检测IP错误故障事件的逻辑，即目标IP与被测IP之间的差异超过了标准。在架构设计中，我们将波形生成等较为复杂的任务分配给主机服务器，将需要实时操作的错误故障检测任务分配给目标控制器。其次，采用直接存储器存取（DMA）方式进行数据通信；因此，目标控制器可以并行地进行检测逻辑和数据通信，而不会降低实时性。第三，我们设计了适当的数据通信时序，保证了系统的稳定运行。在主机端，我们采用ITER实时框架（RTF）技术启动数据通信，具有精确的定时和精确的执行周期控制。最后，我们采用旁路逻辑来防止与PCS的相同检测操作发生冲突。我们在KSTAR等离子体实验中评估了IP错误故障检测逻辑的功能。

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引用次数: 0

Preliminary design of the self-cooled lithium-lead SCYLLA blanket for a spherical tokamak 球形托卡马克用自冷锂铅SCYLLA包层的初步设计

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

Fusion Engineering and Design

Pub Date : 2025-12-05 DOI: 10.1016/j.fusengdes.2025.115574

Luigi Candido , Paul Barron , Colin Baus , Italo Godoy-Morison , John McGrady , Minoru Jimma , Satoshi Ogawa , Richard Pearson , Ben Raeves , Taishi Sugiyama , Jun Takamine , Jack Taylor , Luke Taylor-King , Satoshi Ueguchi , Andrew Wilson , Satoshi Konishi

The future deployment of commercial fusion energy depends on several critical factors, among which the development of a feasible, safe, and integrated breeding blanket (BB) plays a prominent role. Since the company was founded in 2019, Kyoto Fusioneering (KF) has been developing its capability in advanced blanket design and technology development, focusing efforts on the advancement of its own innovative concept known as SCYLLA (Self-Cooled Yuryo Lithium-Lead Advanced), a self-cooled lithium-lead type blanket using silicon carbide composite (SiC

_{f}

/SiC) as a structural material. Efforts to develop the SCYLLA design have employed a holistic approach focused on component modelling, identification of system interfaces between components and systems, and safety evaluation. In this paper, progress towards an application of the SCYLLA breeding blanket configuration, using a spherical Tokamak reactor as a reference, is reported. The description of the current architecture is provided, focusing on the main modifications to evolve the design from a pre-conceptual configuration to a more robust layout. From the point of view of interfaces and experimental R&D, a lithium-lead loop has also been developed by KF as part of its UNITY-1 facility, based in Kumiyama (Kyoto, Japan). This system includes comprehensive design and modelling of the tritium extraction unit. The chosen modelling strategy and the obtained results are reported in the paper and critically discussed.

未来商业核聚变能源的部署取决于几个关键因素，其中开发一种可行、安全、集成的育种毯（BB）起着突出的作用。自2019年成立以来，京都Fusioneering （KF）一直在开发先进的电毯设计和技术开发能力，专注于推进自己的创新概念SCYLLA（自冷Yuryo锂铅先进型），这是一种使用碳化硅复合材料（SiCf/SiC）作为结构材料的自冷锂铅型电毯。开发SCYLLA设计的努力采用了一种整体方法，侧重于组件建模、组件和系统之间的系统接口识别以及安全评估。本文报道了以球形托卡马克反应堆为参考，在SCYLLA增殖包层结构应用方面的进展。本文提供了当前体系结构的描述，重点介绍了将设计从概念前配置演变为更健壮的布局的主要修改。从界面和实验研发的角度来看，KF也开发了锂-铅回路，作为其位于Kumiyama （Kyoto, Japan）的UNITY-1设施的一部分。该系统包括氚萃取装置的综合设计和建模。本文报告了所选择的建模策略和获得的结果，并对其进行了批判性讨论。

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引用次数: 0

Irradiation resistance properties of WTaVCr alloy coatings WTaVCr合金涂层的耐辐照性能

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

Fusion Engineering and Design

Pub Date : 2025-12-05 DOI: 10.1016/j.fusengdes.2025.115467

Lingmin La , Lin Qin , Guanjie Liang , Lingling Wang

WTaVCr alloy coatings with different elemental contents and W substrates were selected for He ion irradiation experiments (irradiation energy 50 eV, irradiation dose of 1 × 10²⁵ m^-2, irradiation temperature 1273 K). After irradiation, the tungsten substrate showed a "fuzz" structure on the surface, while pinholes and convoluted structures appeared on the surface of WTaVCr. The irradiated WTaVCr alloy coatings exhibited a hardening phenomenon, The W₂₅Ta_23.5V_20.8Cr_30.6 alloy coating exhibits the lowest hardening rate, and the TEM observations of W₂₅Ta_23.5V_20.8Cr_30.6 show that the number density of He bubbles in the alloy was significantly lower than that of pure tungsten, which exhibits excellent resistance to irradiation.

选择不同元素含量的WTaVCr合金涂层和W衬底进行He离子辐照实验（辐照能量50 eV，辐照剂量1 × 1025 m-2，辐照温度1273 K）。辐照后，钨基板表面呈“绒毛状”结构，而WTaVCr表面出现针孔和卷曲结构。W25Ta23.5V20.8Cr30.6合金涂层的硬化率最低，且透射电镜观察表明，W25Ta23.5V20.8Cr30.6合金中He气泡的数量密度明显低于纯钨，表现出优异的耐辐照性能。

引用次数: 0