Nuclear Materials and Energy最新文献_第4页

Development of ITER first wall heat load feedback control 开发 ITER 第一壁热负荷反馈控制装置

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-21 DOI: 10.1016/j.nme.2024.101781

Federico Pesamosca , Timo Ravensbergen , Richard A. Pitts , Domenico Frattolillo , Mattia Erroi , Quentin Deliege , Peter C. de Vries , Luca Zabeo , Luigi Pangione , Ivo S. Carvalho , Anna Vu

Real-time monitoring and control of thermal loads on tungsten plasma-facing components is mandatory for ITER to avoid their overheating during plasma discharges. For the Start of Research Operations (SRO) phase, dedicated First Wall Heat Load Control (FWHLC) functions are included in the ITER Plasma Control System (PCS) to prevent the development of excessive plasma heat loads on the inertially cooled first wall (FW). In this work, we propose a FWHLC design with a model-based approach, which features a simplified thermal model for the FW armour. This control-oriented model simulates the thermal response of ITER FW to slow changes in plasma equilibrium and energy, which during ITER operation will be monitored by the real-time wide angle infrared camera system. This allows computationally efficient runs for rapid controller performance assessment but can also assist operation scenario design by pre-empting potential heat load issues. FWHLC is designed to react to measured and predicted FW temperatures overcoming predesigned limits with real-time variations of the plasma shape or auxiliary power references, aiming to reduce thermal loads before a premature plasma termination is required to protect the FW. The new scheme is tested in closed loop simulations, where perturbations to nominal ITER low current scenarios are introduced in the wall thermal model to trigger the response of FWHLC, supporting the effectiveness of the proposed policy.

实时监测和控制面向等离子体的钨组件的热负荷是热核实验堆的必修课，以避免它们在等离子体放电期间过热。在研究运行开始（SRO）阶段，ITER 等离子体控制系统（PCS）中包含了专门的第一壁热负荷控制（FWHLC）功能，以防止惯性冷却的第一壁（FW）上产生过多的等离子体热负荷。在这项工作中，我们提出了一种基于模型的 FWHLC 设计方法，其特点是简化了 FW 装甲的热模型。这个以控制为导向的模型模拟了热核实验堆 FW 对等离子体平衡和能量缓慢变化的热响应，在热核实验堆运行期间，广角红外摄像系统将对这些变化进行实时监测。这样就可以通过高效的计算运行来快速评估控制器性能，还可以通过预先防范潜在的热负荷问题来协助运行方案设计。FWHLC 的设计目的是对测量和预测的 FW 温度做出反应，通过等离子体形状或辅助功率基准的实时变化来克服预先设计的限制，从而在需要过早终止等离子体以保护 FW 之前降低热负荷。新方案在闭环模拟中进行了测试，在墙热模型中引入了对名义 ITER 低电流情景的扰动，以触发 FWHLC 的响应，从而证明了所建议策略的有效性。

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引用次数: 0

Access to an ELM-suppressed X-point radiator regime in TCV snowflake minus configurations 在 TCV 雪花负配置中访问 ELM 抑制的 X 点辐射器系统

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-20 DOI: 10.1016/j.nme.2024.101784

H. Reimerdes , C. Theiler , M. Bernert , B.P. Duval , O. Février , S. Gorno , D. Hamm , K. Lee , O. Pan , A. Perek , L. Simons , G. Sun , A. Thornton , K. Verhaegh , Y. Wang , C. Wüthrich , M. Zurita , the TCV team , the EUROfusion tokamak exploitation team

TCV’s operating regime with an X-point radiator (XPR) has been broadened by changing the magnetic geometry. XPRs have properties that could make them an attractive power exhaust solution for fusion reactors. These include the conversion of a high fraction of exhaust power into radiation. TCV had previously accessed the XPR regime only with difficulties, as predicted for plasmas where radiative losses are dominated by carbon impurities, that are ubiquitous in TCV. Guided by this theoretical model of the XPR, recent experiments employed TCV’s configurational versatility to demonstrate that XPR access can be facilitated by introducing a second X-point in the vicinity of the separatrix. This configuration, which has a snowflake-minus topology, features a particularly long magnetic connection length from the region just above the X-point to the outer midplane together with a wide geometrical interface with the private flux region that reaches high neutral pressures. Transitioning to this configuration in a high-power H-mode leads to a shift in the radiating region across the separatrix from the divertor to a volume above the X-point, i.e. within the last closed flux surface (LCFS). This displacement of the radiating region is co-incident with the disappearance of edge localised modes (ELMs), while retaining H-mode confinement, a behaviour only, to date, observed in devices with metallic walls. In contrast to observations in these other devices, on TCV, the primary strike points in these configurations remain attached. Detailed measurements of the plasma kinetic parameters inside and outside of the separatrix now challenge the models for access and stability of the XPR and ELMs alike.

TCV 的 X 点辐射器 (XPR) 运行机制已通过改变磁性几何形状而得到拓宽。XPR 的特性可使其成为聚变反应堆的一种有吸引力的功率排气解决方案。其中包括将高比例的排气功率转化为辐射。TCV 先前进入 XPR 状态时遇到了困难，因为根据预测，等离子体的辐射损失主要由碳杂质造成，而碳杂质在 TCV 中无处不在。在这一 XPR 理论模型的指导下，最近的实验利用 TCV 的构型多样性证明，在分离矩阵附近引入第二个 X 点可以促进 XPR 的进入。这种构型具有雪花减拓扑结构，其特点是从 X 点上方区域到外部中平面的磁连接长度特别长，同时与达到高中性压力的私有磁通区域具有宽阔的几何界面。在高功率 H 模式下过渡到这种配置会导致辐射区域从分流器横跨分离矩阵转移到 X 点上方，即最后一个封闭磁通面（LCFS）内。辐射区域的这种移动与边缘局部模式（ELM）的消失同时发生，同时保留了 H 模式约束，迄今为止，只有在具有金属壁的设备中才能观察到这种行为。与在这些其他装置中观察到的情况不同，在 TCV 上，这些配置中的主要撞击点仍然是附着的。对分离矩阵内外等离子体动力学参数的详细测量结果，对 XPR 和 ELM 的进入和稳定性模型提出了挑战。

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引用次数: 0

Establishing the temperature and orientation dependence of the threshold displacement energy in ThO2 via molecular dynamics simulations 通过分子动力学模拟确定二氧化硫阈值位移能的温度和取向依赖性

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-20 DOI: 10.1016/j.nme.2024.101774

Lin-Chieh Yu , Shuxiang Zhou , Miaomiao Jin , Marat Khafizov , David Hurley , Yongfeng Zhang

ThO₂ is a promising fuel for next-generation nuclear reactors. As a critical quantity measuring its radiation tolerance, the dependence of the threshold displacement energy on temperature and crystal orientation in ThO₂ is unclear and established using comprehensive molecular dynamics simulations in this work. For both Th and O primary knock-on atoms (PKAs), the thresholds, denoted as

E_{d}^{Th}

and

E_{d}^{O}

, respectively, are calculated using two different interatomic potentials. Similar temperature and orientation dependence are observed, albeit with some quantitative differences. While on average over all orientations, higher energy is required for Th PKAs than O PKAs to displace atoms, the polar-averaged

E_{d}^{Th}

is significantly lower than that for

E_{d}^{O}

. Further,

E_{d}^{Th}

and

E_{d}^{O}

show different crystal orientation dependence and temperature dependence. Notably, the cubic symmetry in the fluorite structure is followed by

E_{d}^{Th}

, but does not hold for

E_{d}^{O}

because of the existence of two sublattices. The much higher average

E_{d}^{O}

than

E_{d}^{Th}

and their different temperature dependence are interpreted by the distinct recombination rates of Th and O Frenkel pairs in thermal spikes, resulting from the substantially lower migration barriers of O vacancies and interstitials. The recombination of O vacancies and interstitials, both of which are charged, is further enhanced by the Coulomb interaction at small Frenkel pair separations. The new findings are discussed for their generality in fluorite-structured oxides by comparing the results in ThO₂ and UO₂.

二氧化硫是下一代核反应堆的理想燃料。作为衡量其辐射耐受性的一个关键量，二氧化钍的阈值位移能与温度和晶体取向的关系尚不清楚，本研究利用全面的分子动力学模拟确定了这一关系。对于 Th 原子和 O 原子的原敲原子（PKAs），使用两种不同的原子间位势计算出了阈值，分别称为 EdTh 和 EdO。尽管存在一些定量差异，但观察到了类似的温度和取向依赖性。虽然从所有取向的平均值来看，Th PKAs 比 O PKAs 需要更高的能量来置换原子，但极性平均值 EdTh 明显低于 EdO。此外，EdTh 和 EdO 表现出不同的晶体取向依赖性和温度依赖性。值得注意的是，EdTh 遵循萤石结构中的立方对称性，而 EdO 则由于存在两个子晶格而不遵循这种对称性。EdO 的平均值远高于 EdTh，而且它们的温度依赖性也不同，这是因为 Th 和 O 的 Frenkel 对在热峰值中的重组速率不同，这是因为 O 空位和间隙的迁移障碍大大降低。O 空位和间隙都是带电的，在较小的弗伦克尔对分离时，它们之间的库仑相互作用会进一步增强它们的重组。通过比较二氧化硫和二氧化铀的结果，讨论了新发现在萤石结构氧化物中的普遍性。

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引用次数: 0

Recent progress in the development of liquid metal plasma facing components for magnetic fusion devices 磁核聚变装置液态金属等离子体面组件的最新研发进展

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-18 DOI: 10.1016/j.nme.2024.101776

J.S. Hu , G.Z. Zuo , L. Li , D.H. Zhang , H.L. Bi , Z.B. Ye , J.H. Pan , S.Y. Dai , X.C. Meng , Z. Sun , M. Ono , Y. Hirooka , D.N. Ruzic

One of the most critical challenges for future fusion reactors is to develop longevity plasma-facing components (PFCs) exposed to extremely high heat and neutron loads. As opposed to those employing solid metals, PFCs with flowing liquid metals (LM) have shown self-healing, heat removal and good impurity control capabilities, all essential to fusion devices. Recently, significant progress in LM-PFC development has been reported globally, with data from several magnetic fusion devices. These studies reveal that LM-PFCs can endure extreme heat fluxes while maintaining plasma compatibility. New design concepts have been proposed and numerically analyzed, advancing models for liquid PFCs in future reactors. Despite existing technical challenges, these developments suggest that LM-PFCs hold promise for future fusion applications.

未来核聚变反应堆面临的最严峻挑战之一，是开发能够承受极高的热量和中子负荷的长寿命等离子体面组件（PFC）。与采用固态金属的元件相比，采用流动液态金属（LM）的 PFC 具有自愈、散热和良好的杂质控制能力，这些都是聚变设备所必需的。最近，全球在 LM-PFC 的开发方面取得了重大进展，并获得了几个磁核聚变装置的数据。这些研究表明，LM-PFC 可以承受极端热通量，同时保持等离子体的兼容性。研究人员提出了新的设计概念，并对其进行了数值分析，从而推动了未来反应堆中液体全氟化碳模型的发展。尽管存在技术挑战，但这些进展表明 LM-PFC 在未来的核聚变应用中大有可为。

引用次数: 0

Designing tungsten armoured plasma facing components to pulsed heat loads in magnetic fusion machines 设计面向等离子体的钨铠装部件，以承受磁核聚变设备中的脉冲热负荷

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-18 DOI: 10.1016/j.nme.2024.101777

R. Mitteau, M. Diez, M. Firdaouss

A possible design rule for preventing surface damage from thermal transients to solid tungsten armour is proposed and formulated for the plasma facing components (divertor, first wall) of magnetic fusion machines. The rule is based on combined results from laboratory experiments and operating fusion machines, and fundamental engineering principles such as the heat flux factor (F_HF) and fatigue usage fraction (F_UF). As an example, the rule would allow 2.10⁴ transient heat loads cycles at a F_HF of 10 MJm^-2s^-½ before the lifetime is considered exhausted. The formulation of the rule using engineering principles allows combining loads of different magnitudes and various number of cycles. A practical example of the rule usage is provided, illustrating loads combination and how the rule may contribute to the component geometrical design. The proposed rule is only valid for surface loading conditions, hence is not usable for volumetric loading conditions such as runaway electrons. Setting a budget lifetime and a design rule does not preclude actual plasma operation beyond the design lifetime. It is actually normal that experimental devices explore a larger domain than the one defined at the time of the design.

针对磁核聚变机器的等离子体面部件（分流器、第一壁），提出并制定了防止热瞬态对固体钨铠装造成表面损伤的可行设计规则。该规则基于实验室实验和运行核聚变机器的综合结果，以及热通量系数（FHF）和疲劳使用率（FUF）等基本工程原理。举例来说，在热通量系数为 10 MJm-2s-½ 时，该规则允许 2.104 个瞬态热负荷循环，然后才会认为寿命耗尽。利用工程学原理制定的规则可以将不同量级的负载和不同次数的循环结合起来。我们提供了一个使用该规则的实际例子，说明了载荷组合以及该规则如何有助于组件的几何设计。建议的规则仅适用于表面负载条件，因此不适用于电子失控等体积负载条件。设定预算寿命和设计规则并不排除实际等离子体运行超过设计寿命。实际上，实验装置探索的领域大于设计时定义的领域是很正常的。

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引用次数: 0

Validating reduced models for detachment onset and reattachment times on MAST-U 在 MAST-U 上验证脱离开始和重新附着时间的简化模型

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-16 DOI: 10.1016/j.nme.2024.101765

S.S. Henderson , M. Bernert , D. Brida , G.L. Derks , S. Elmore , F. Federici , J.R. Harrison , A. Kirk , B. Kool , N. Lonigro , J. Lovell , D. Moulton , H. Reimerdes , P. Ryan , J.M. Stobbs , K. Verhaegh , T. van den Doel , T. Wijkamp , O. Bardsley , MAST-U Team , EUROfusion Tokamak Exploitation Team

Two reduced models for predicting detachment onset and divertor reattachment times are validated on MAST Upgrade (MAST-U). These models are essential for future tokamak reactor design, providing rapid calculations based primarily on engineering parameters. The first model predicts detachment onset using a qualifier developed on ASDEX Upgrade (AUG) and later tested on JET, while the second model provides an estimate for the time required for a given transient to burn through the neutral particles in the divertor. Experiments in H-mode plasma scenarios were conducted on MAST-U with double-null and single-null configurations, which involved

D_{2}

fuelling ramps and

N_{2}

seeding. The detachment onset was determined by monitoring divertor parameters, including the target heat flux profile, electron temperature, and electron density, with measurements showing consistency with AUG-derived predictions. Reattachment times were assessed during dynamic vertical shifts of the plasma centroid position, with observations indicating reattachment within milliseconds, consistent with model predictions. Overall, the results confirm the applicability of both reduced models to MAST-U, extending their validation beyond AUG and JET.

在 MAST 升级版（MAST-U）上验证了两个用于预测脱离开始时间和岔道重新连接时间的简化模型。这些模型对未来的托卡马克反应堆设计至关重要，主要根据工程参数提供快速计算。第一个模型使用在 ASDEX Upgrade (AUG) 上开发并随后在 JET 上测试的限定符来预测脱离开始的时间，而第二个模型则对给定瞬态烧穿分流器中的中性粒子所需的时间进行估算。在 MAST-U 上用双空和单空配置进行了 H 模式等离子体情景实验，其中涉及 D2 燃料斜坡和 N2 种子。通过监测分流器参数（包括目标热通量曲线、电子温度和电子密度）来确定脱离开始时间，测量结果显示与 AUG 的预测结果一致。在等离子体中心点位置动态垂直移动过程中，对重新附着时间进行了评估，观测结果表明重新附着时间在几毫秒之内，与模型预测一致。总之，这些结果证实了两个简化模型对 MAST-U 的适用性，使其验证范围超出了 AUG 和 JET。

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引用次数: 0

An understanding of the segregation and migration mechanism of point defects in tungsten grain boundaries: An atomic scale simulation 了解钨晶界点缺陷的偏析和迁移机制：原子尺度模拟

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-16 DOI: 10.1016/j.nme.2024.101771

Ya-Wen Li , Xiao-Chun Li , Bai-Chuan Xu , Yilang Mai , Wei Wu , Ziqi Li , Hai-Shan Zhou , Guang-Nan Luo

In this study, molecular dynamics (MD) simulations are employed to investigate the interactions between point defects and grain boundaries (GBs) in tungsten. Firstly, the segregation energy of vacancies (Vs) and self-interstitial atoms (SIAs) to GBs is examined. The results indicate that both Vs and SIAs tend to segregate towards GBs. Notably, the different types of GBs exhibit varying attraction to Vs and SIAs. The interaction radius is applied to investigate the degree of segregation, which is a significant index of GBs’ attraction to Vs and SIAs. The segregation radius of SIAs is typically larger than that of Vs. High-angle grain boundaries (HAGBs) show strong segregation for SIAs, facilitating the aggregation at GBs. Additionally, elastic theory is applied to qualitatively discuss the factors influencing the segregation energy distribution. The differences in atomic free volume caused by hydrostatic stress and the lattice distortion induced by point defects lead to Vs and SIAs occupying different sites. Finally, the nudged elastic band (NEB) method is employed to study the migration of Vs and SIAs near GBs, revealing that migration energy barriers for Vs and SIAs in GBs are much lower than in bulk. GBs facilitate the migration of point defects. Low-angle grain boundaries (LAGBs) present a higher energy barrier for V migration. Vs tend to occupy the compressive region, while SIAs tend to occupy the tensile region. Particularly, in comparison to Vs, SIAs are more likely to segregate to GBs due to higher binding energy, wider interaction radius, and extremely lower diffusion energy barrier. This provides some insights into the segregation and migration of point defects in GBs.

本研究采用分子动力学（MD）模拟来研究钨中点缺陷与晶界（GB）之间的相互作用。首先，研究了空位（Vs）和自间隙原子（SIAs）与晶界的分离能。结果表明，空位和自间隙原子都倾向于向 GB 分离。值得注意的是，不同类型的 GB 对 Vs 和 SIA 的吸引力各不相同。交互半径被用来研究分离程度，它是衡量 GB 对 Vs 和 SIA 吸引力的重要指标。SIA 的偏析半径通常大于 Vs 的偏析半径。高角度晶界（HAGB）显示出对 SIAs 的强烈偏析，促进了 GBs 的聚集。此外，还应用弹性理论定性地讨论了影响偏析能分布的因素。静水压力造成的原子自由体积差异和点缺陷引起的晶格畸变导致 Vs 和 SIA 占据不同的位点。最后，利用裸弹带（NEB）方法研究了Vs和SIA在GB附近的迁移，结果表明Vs和SIA在GB中的迁移能垒远低于在块体中的迁移能垒。GB 有利于点缺陷的迁移。低角度晶界（LAGB）对 V 的迁移具有较高的能量障碍。Vs 倾向于占据压缩区域，而 SIA 则倾向于占据拉伸区域。特别是，与 Vs 相比，SIAs 由于具有更高的结合能、更宽的相互作用半径和极低的扩散能障，更有可能向 GBs 分离。这为了解 GB 中点缺陷的偏析和迁移提供了一些启示。

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引用次数: 0

Influence of suppressed blistering by heavy ion pre-damage on deuterium retention in tungsten 重离子预破坏抑制起泡对钨中氘保留的影响

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-16 DOI: 10.1016/j.nme.2024.101775

Ting Wang , Yue Yuan , Xiu-Li Zhu , Wangguo Guo , Jipeng Zhu , Shiwei Wang , Long Cheng , Guang-Hong Lu

Surface damage and fuel retention are one of the major threats to the performance of plasma-facing materials (PFMs) in ITER and future fusion reactors. This work aims to investigate the influence of suppressed blistering by heavy ion pre-damage on deuterium (D) retention in tungsten (W). Recrystallized W samples were irradiated with 3.5 MeV iron (Fe¹³⁺) ions at room temperature to create displacement damage with a peak damage level of 0.1 dpa. Afterwards, a series of low-energy (38 eV) D plasma exposures were performed at 500 K. Three exposure fluences below and above the blistering threshold (5 × 10²⁴ ∼ 3 × 10²⁵ D m⁻²) of the pre-damaged W are selected to decouple and couple the damage-induced defects and blistering-induced defects, respectively. Surface observations show that no blisters are formed in un-damaged W after low-fluence D exposure (5 × 10²⁴ D m⁻²), whereas severe blistering (surface coverage ratio: 34.2 %) occurs in the high-fluence case (3 × 10²⁵ D m⁻²). In contrast, only a small number of blisters (4.2 %) are formed in Fe-damaged W when D fluence reaches 3 × 10²⁵ D m⁻². Moreover, Fe pre-damage increases D retention by a factor of 3.33 and 1.20 at the low-fluence (5 × 10²⁴ D m⁻²) and medium-fluence (1 × 10²⁵ D m⁻²) D exposure, respectively. While in the high-fluence case (3 × 10²⁵ D m⁻²), the enhancement effect of D retention in Fe-damaged W is significantly weakened, such that retention is lower for damaged W, probably due to the significant suppression effect on surface blistering and its accompanying defect formation. This work highlights the suppressed blistering-induced D retention by pre-existing damage in W.

表面损伤和燃料滞留是热核实验堆和未来聚变反应堆中面向等离子体的材料（PFMs）性能的主要威胁之一。这项工作旨在研究重离子预损伤抑制起泡对钨（W）中氘（D）保留的影响。在室温下用 3.5 MeV 的铁（Fe13+）离子对重结晶的 W 样品进行辐照，以产生峰值为 0.1 dpa 的位移损伤。选择了低于和高于预损伤 W 的起泡阈值（5 × 1024 ∼ 3 × 1025 D m-2）的三种曝光通量，以分别解耦和耦合损伤引起的缺陷和起泡引起的缺陷。表面观察结果表明，在低荧光 D 暴露（5 × 1024 D m-2）后，未损坏的 W 不会形成水泡，而在高荧光 D 暴露（3 × 1025 D m-2）的情况下，则会出现严重的水泡（表面覆盖率：34.2%）。相反，当 D 通量达到 3 × 1025 D m-2 时，铁损伤的 W 只形成少量水泡（4.2%）。此外，在低通量（5 × 1024 D m-2）和中通量（1 × 1025 D m-2）的 D 暴露下，铁预破坏会使 D 保持率分别提高 3.33 和 1.20 倍。而在高辐照度（3 × 1025 D m-2）情况下，铁损伤 W 中 D 保持率的增强效应明显减弱，因此损伤 W 的保持率较低，这可能是由于对表面起泡及其伴随的缺陷形成具有显著的抑制作用。这项研究凸显了 W 中预先存在的损伤对水泡诱导的 D 保留的抑制作用。

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引用次数: 0

Properties of boron layers deposited during boronisations in W7-X W7-X硼化过程中沉积硼层的特性

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-16 DOI: 10.1016/j.nme.2024.101778

M. Mayer , M. Balden , T. Bräuer , D. Cipciar , C.P. Dhard , P. Drews , S. Elgeti , D. Höschen , C. Killer , D. Naujoks , N. Sandri , J.-H. Schmid-Dencker , L. Vanó , H. Viebke , O. Volzke , W7-X Team

Boronisation was first used for wall conditioning in W7-X during the OP 1.2b operational period, which was characterized by the use of the fine-grain graphite Test Divertor Unit (TDU) and inertial cooling only. After this period, deposited layers were observed on all inner surfaces. Deposited layers were analyzed on 21 inner wall tiles using ion beam analysis methods, the deposited layers consisted mostly of boron with additional carbon and oxygen. During the operational period OP 2.1 with an actively water cooled divertor made of carbon fiber reinforced carbon, different materials were exposed during two individual boronisations using the multi-purpose manipulator. Deposited boronisation layers on the samples were analyzed using nuclear reaction analysis. The deposited layer thicknesses showed some variation depending on substrate material and surface roughness, but a systematic dependence on material and/or roughness was not observed. Under the typical boronisation conditions at W7-X, one A × h (Ampere times hour) of boronisation results in a boronisation layer with a thickness of about 30 ± 15 × 10¹⁵ B-atoms/cm² (about 3 ± 1.5 nm) at the position of the multi-purpose manipulator. The oxygen gettering capacity of the layers is up to 0.5 – 0.9O/B.

硼化首次用于 W7-X 的壁面调节是在 OP 1.2b 运行期间，其特点是使用细粒度石墨试验分流装置（TDU）和仅惯性冷却。之后，在所有内表面都观察到了沉积层。使用离子束分析方法对 21 块内壁瓦上的沉积层进行了分析，沉积层主要由硼组成，还有碳和氧。在运行期 OP 2.1 期间，使用碳纤维增强碳制成的主动水冷分流器，在使用多用途机械手进行的两次单独硼化过程中，暴露出了不同的材料。使用核反应分析法对样品上沉积的硼化层进行了分析。沉积层厚度因基底材料和表面粗糙度的不同而有一定的差异，但没有观察到系统性的材料和/或粗糙度依赖关系。在 W7-X 的典型硼化条件下，1 A × h（安培乘以小时）的硼化作用会在多功能操纵器的位置产生厚度约为 30 ± 15 × 1015 B-atoms/cm2 （约 3 ± 1.5 nm）的硼化层。硼化层的脱氧能力高达 0.5 - 0.9O/B。

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引用次数: 0

Metallic droplet impact simulations on plasma-facing components 金属液滴对面向等离子体部件的冲击模拟

IF 2.3 2区物理与天体物理 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Materials and Energy

Pub Date : 2024-10-15 DOI: 10.1016/j.nme.2024.101748

L. Vignitchouk, S. Ratynskaia, JET Contributors

Multiphase Navier–Stokes simulations of liquid metal droplets colliding with solid plasma-facing components are carried out in conditions representative of magnetic confinement fusion devices. The flow dynamics of the spreading liquid are examined to assess the relative importance of various physical processes in the impact energy budget. Contributions from the initial droplet surface energy and the solidification-induced momentum sink are shown to be of great importance in determining the final geometry of the frozen spatter. Semi-empirical scaling laws available in the literature are adapted to provide robust predictions of the flattening ratio that can be extrapolated to general fusion-relevant impact scenarios.

在磁约束聚变装置的代表性条件下，对液态金属液滴与面向等离子体的固体部件碰撞进行了多相纳维-斯托克斯模拟。研究了扩散液体的流动动力学，以评估各种物理过程在碰撞能量预算中的相对重要性。结果表明，初始液滴表面能和凝固引起的动量汇对确定冷冻飞溅物的最终几何形状非常重要。对文献中的半经验缩放定律进行了调整，以提供对扁平率的可靠预测，并可推断出与核聚变有关的一般撞击情景。

引用次数: 0