Fusion Engineering and Design最新文献

Corroborating VNA and thermal measurements of transmission loss on the DIII-D ECH waveguide system DIII-D ECH波导系统传输损耗的VNA和热测量的确证

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-20 DOI: 10.1016/j.fusengdes.2026.115626

M.P. Ross , K.A. Thackston , A. Dupuy , Y. Gorelov , N. de Boucaud , P. Nesbet , A. Torrezan , Z. Bayler , N. Watson , J. Anderson , J.P. Squire

Electron cyclotron heating (ECH) and current drive (ECCD) will play a large role in tokamak-based fusion reactors. At the DIII-D tokamak, 110 GHz microwaves injected into the plasma can provide core heating and current drive as well as impurity control, neoclassical tearing mode mitigation, and breakdown assistance. Understanding the physics of these processes relies on accurate estimates of injected ECH power. DIII-D’s ECH system consists of six MW-class Microwave Power Products (MPP) gyrotron microwave sources. Operating the gyrotrons far from the tokamak removes them from magnetic field interference, so 31.75 mm inner-diameter corrugated waveguides transmit the microwave power the 80 m from the gyrotrons to steerable launchers in the tokamak chamber. Estimates of injected power rely on knowing the generated power at the source and then subtracting transmission loss. Conventional transmission loss measurements based on calorimetric dummy loads are onerous and only possible during extended maintenance periods. This work examines two tools that provide more flexibility for the transmission loss measurements. A resistive temperature detector (RTD) array installed along a waveguide measures heat lost to the transmission line, and low power time domain reflectometry (TDR) measurements with a vector network analyzer (VNA) allows loss measurements without burdensome hardware modifications.

电子回旋加热（ECH）和电流驱动（ECCD）将在托卡马克聚变反应堆中发挥重要作用。在DIII-D托卡马克上，注入等离子体的110 GHz微波可以提供核心加热和电流驱动，以及杂质控制、新经典撕裂模式缓解和击穿辅助。了解这些过程的物理原理依赖于对注入ECH功率的准确估计。DIII-D的ECH系统由六个mw级微波功率产品（MPP）回旋管微波源组成。在远离托卡马克的地方运行回旋管可以使其免受磁场干扰，因此，31.75 mm内径的波纹波导将微波功率传输到距离回旋管80米的托卡马克室内的可操纵发射器。注入功率的估计依赖于知道源处的发电功率，然后减去传输损耗。传统的基于量热虚拟负载的传输损耗测量是繁重的，并且只有在延长的维护期间才有可能。这项工作考察了两种为传输损耗测量提供更大灵活性的工具。沿着波导安装的电阻式温度检测器（RTD）阵列可以测量传输线的热损失，而带有矢量网络分析仪（VNA）的低功率时域反射仪（TDR）测量可以在不进行繁琐的硬件修改的情况下进行损耗测量。

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引用次数: 0

Mechanical analysis of tail structures in the next-generation fully superconducting tokamak CS HTS 下一代全超导托卡马克cshts尾结构力学分析

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-24 DOI: 10.1016/j.fusengdes.2026.115632

Yi Yu , HouXiang Han

The Tail structure is a crucial component of high-temperature superconducting coils, providing both mechanical load transfer and reliable electrical connectivity between conductor leads. In this study, we designed and analyzed a Tail structure for a next-generation fully superconducting tokamak. We performed a coupled multi-physics analysis using finite element software to evaluate the mechanical performance of the proposed system under operational loads. The simulation revealed a maximum stress intensity of 1065.9 MPa, while linearized stress amplitudes remained within the allowable design limits, confirming adequate structural safety, and the minimum fatigue life of the structural components exceeds 127,000 cycles. We also developed and validated specialized welding fixtures through a series of welding experiments. All the welds successfully passed the quality inspections and met the technical specifications required for the procedure qualification, offering essential technical support for the construction of superconducting fusion devices.

尾部结构是高温超导线圈的关键组成部分，提供机械负载传递和导体引线之间可靠的电气连接。在这项研究中，我们设计并分析了下一代全超导托卡马克的尾部结构。我们使用有限元软件进行了耦合多物理场分析，以评估所提出系统在运行载荷下的机械性能。模拟结果表明，最大应力强度为1065.9 MPa，线性化应力幅值保持在设计允许范围内，结构安全可靠，结构部件的最小疲劳寿命超过12.7万次。我们还通过一系列焊接实验开发和验证了专用焊接夹具。所有焊缝均顺利通过质量检验，达到工序合格要求的技术规范，为超导聚变装置的建设提供了必要的技术支持。

引用次数: 0

Design and performance analysis of a high-current pulse inductor for fusion magnet quench protection 熔合磁体猝灭保护用大电流脉冲电感器的设计与性能分析

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-29 DOI: 10.1016/j.fusengdes.2026.115641

Wei Tong , Meng Xu , Hua Li , Zhenhan Li , Zhiquan Song , Peng Fu

Superconducting magnets in fusion devices are at the risk of quench during operation. Once quench occurs, the stored electromagnetic energy rapidly turns to Joule heat, potentially damaging the magnets severely. To protect these magnets, the Quench Protection System (QPS) employs high-power DC breaker to quickly interrupt the magnet current and transfer the quench energy, which relies on the LC circuit resonance to generate reverse pulsed current for creating a current zero-crossing point. This paper introduces the design and performance analysis of a high-current pulse inductor with toroidal helical structure. Its unique structure confines the magnetic field inside the coil, reducing interference and ensuring stability under pulse current. Targeting 20 μH inductance and 80 kA rated pulse current, the structural design of the inductor is performed firstly. Further, the Finite Element Method (FEM) is used to conduct a systematic simulation analysis of its electromagnetic, structural, and thermal performance. Finally, a prototype is also manufactured and subjected to LR parameter measurement and pulse high-current testing. Both simulation and test results show that the inductor has excellent magnetic confinement capability, outstanding structural stability, and reasonable temperature rise control ability.

核聚变装置中的超导磁体在运行过程中存在淬灭的危险。一旦发生淬灭，储存的电磁能量迅速转化为焦耳热，可能会严重损坏磁铁。为了保护这些磁体，猝灭保护系统（QPS）采用大功率直流断路器快速中断磁体电流并传递猝灭能量，猝灭能量依靠LC电路谐振产生反向脉冲电流，形成电流过零点。本文介绍了一种环形螺旋结构的大电流脉冲电感器的设计和性能分析。其独特的结构限制了线圈内部的磁场，减少了干扰，确保了脉冲电流下的稳定性。首先以电感量为20 μH，额定脉冲电流为80 kA为目标，进行了电感器的结构设计。在此基础上，采用有限元法对其电磁性能、结构性能和热性能进行了系统的仿真分析。最后，制作了样机并进行了LR参数测量和脉冲大电流测试。仿真和试验结果表明，该电感具有优良的磁约束性能、良好的结构稳定性和合理的温升控制能力。

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引用次数: 0

Consideration of density measurement using toroidal interferometer and polarimeter on JA DEMO 在JA DEMO上使用环面干涉仪和偏振仪测量密度的思考

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-20 DOI: 10.1016/j.fusengdes.2026.115638

K. Iwasaki , S. Sugiyama , Y. Ohtani , Y. Sakamoto

The toroidal interferometer and polarimeter (TIP) have been investigated as density diagnostics for JA DEMO. A model for the interferometer and polarimeter phase shifts incorporating the finite electron temperature effect has been implemented into a plasma control simulation code to generate the synthetic phase shift signals. The laser wavelength is set to

10.6 μ

m identical to that used in ITER. The finite temperature effect is significant. The deviation between the estimated line-averaged densities obtained with and without accounting for finite temperature effects reaches approximately 7% for the interferometer and 10% for the polarimeter along lines of sight near the magnetic axis, and decreases to a few percent near the outer edge. Density feedback control has been performed, and a comparison is made between the line-averaged densities with and without correction for the temperature effect. When the temperature effect is neglected, the density is underestimated, leading to an increase in the actual density. Consequently, the fusion output increases, resulting in an error of up to 11% when using the central viewing chords. Correction of the density error caused by the finite electron temperature has been carried out using TIP alone by taking the difference between the interferometer and polarimeter signals. The results show that it can reduce the density error to below 1%.

研究了环形干涉仪和偏振仪（TIP）作为JA DEMO的密度诊断。考虑有限电子温度效应的干涉仪和偏振仪相移模型被实现到等离子体控制仿真代码中，以产生合成相移信号。激光波长设置为与ITER相同的10.6μm。有限温度效应是显著的。在考虑有限温度效应和不考虑有限温度效应的情况下，估计的线平均密度之间的偏差在靠近磁轴的视线方向上，干涉仪达到约7%，偏振仪达到约10%，在靠近外缘处减小到几个百分点。进行了密度反馈控制，并对温度效应进行了校正和不校正后的线平均密度进行了比较。当忽略温度效应时，密度被低估，导致实际密度增加。因此，融合输出增加，导致使用中央观察弦时误差高达11%。利用干涉仪和偏振仪信号的差值，对有限电子温度引起的密度误差进行了单独的TIP校正。结果表明，该方法可将密度误差降低到1%以下。

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引用次数: 0

Investigation of the ultra-wideband polarizer for high power millimeter wave system 大功率毫米波系统超宽带偏振器的研究

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-01-15 DOI: 10.1016/j.fusengdes.2026.115630

Feng Zhang, Mei Huang, Gangyu Chen, He Wang, Wanxin Zheng, Jieqiong Wang, Guoyao Fan, Cheng Chen

As a crucial component of the electron cyclotron resonance heating (ECRH) system, the polarizer primarily serves to change the polarization characteristics of millimeter wave. In this study, an ultra-wideband polarization strategy based on two polarizers for the ECRH system is presented. By employing two identical sinusoidally grooved polarizers at a designated incident angle of 60°, it is possible to attain arbitrary polarization can be attained across an ultra-wideband frequency ranging from 99 GHz to 189 GHz. A ultra-wideband polarizer was devised and evaluated, and computational results indicate that the arbitrary polarization efficiency of the proposed method reaches at least 99.94%. According to this analysis, nearly every desired polarization state can be realized using the presented polarization strategy.

偏振器作为电子回旋共振加热（ECRH）系统的关键部件，主要用于改变毫米波的偏振特性。本文提出了一种基于双极化器的ECRH系统超宽带极化策略。通过在指定的60°入射角上使用两个相同的正弦波槽偏振器，可以在99 GHz至189 GHz的超宽带频率范围内获得任意偏振。设计并评估了一种超宽带偏振器，计算结果表明，该方法的任意极化效率达到99.94%以上。根据这一分析，使用所提出的极化策略几乎可以实现所有期望的极化状态。

引用次数: 0

Temperature-dependent ATR-FTIR calibration using a two-state spectral model for deuterium detection in water–manuscript draft- 温度依赖的ATR-FTIR校准，使用双态光谱模型用于水-手稿草案中的氘探测

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-02-07 DOI: 10.1016/j.fusengdes.2026.115656

Edouard Amadou, Jean Jules Fifen, Mama Nsangou, Pierre Doba

Tritiated water will be present in many fusion demonstration facilities, in various forms and at highly heterogeneous concentrations. The implementation of fast and reliable measurement methods is therefore a critical requirement for process monitoring, radiological safety, and tritium traceability. This study evaluates the performance of attenuated total reflectance Fourier-transform infrared (ATR)-FTIR spectroscopy for the quantification of deuterium in aqueous solution, used here as a non-radioactive analogue of tritiated water (HTO). Measurements were conducted over a wide temperature range (0–74 °C) to replicate variable operational conditions. An excellent linear correlation (R² > 0.99) was observed between infrared absorbance and D₂O concentration for the ν₁ and ν₃ vibrational modes, validating the applicability of the Beer-Lambert law under these conditions. Detection limits between 5.4 × 10⁻³ and 7.4 × 10⁻³ mol·mL⁻¹ were obtained, in accordance with IUPAC guidelines. While these values remain above the thresholds required for trace detection, they are compatible with the elevated concentrations expected in certain fusion subsystems. Lastly, the bending mode ν₂, although not calibrated in this work, exhibits promising thermal stability and may serve as a reference band in future applications. This study thus provides a realistic basis for the integration of ATR-FTIR spectroscopy into in situ tritium detection systems. p { margin-left: -0.01in; margin-right: -0.01in; text-indent: 0.32in; margin-bottom: 0.1in; direction: ltr; color: #000,000; line-height: 115%; text-align: justify; orphans: 2; widows: 2}p.western { font-family: "Cambria", serif; font-size: 12pt}p.cjk { font-family: "Cambria"; font-size: 12pt}p.ctl { font-family: "Cambria"

氚化水将出现在许多核聚变示范设施中，以各种形式和高度不均匀的浓度。因此，实施快速可靠的测量方法是过程监测、放射安全和氚可追溯性的关键要求。本研究评估了衰减全反射傅立叶变换红外(ATR)-傅里叶变换红外光谱（ftir）用于定量水溶液中氘的性能，这里用作氚化水（HTO）的非放射性类似物。测量在很宽的温度范围内（0-74°C）进行，以复制可变的操作条件。在ν₁和ν₃振动模式下，观测到红外吸光度与D₂O浓度之间具有良好的线性相关性（R²> 0.99），验证了Beer-Lambert定律在这些条件下的适用性。根据IUPAC的指导方针，得到了5.4 × 10⁻³和7.4 × 10⁻³mol·mL⁻¹之间的检测限。虽然这些值仍然高于痕量检测所需的阈值，但它们与某些融合子系统中预期的浓度升高是相容的。最后，弯曲模式ν₂，虽然在本工作中没有校准，但表现出很好的热稳定性，可以作为未来应用的参考波段。因此，该研究为ATR-FTIR光谱集成到原位氚检测系统中提供了现实基础。P {margin-left: -0.01in；margin-right: -0.01;indent: 0.32;margin-bottom: 0.1;方向:ltr;颜色:# 000000;行高:115%;text-align:证明;孤儿:2;寡妇:2}p。西部{font-family: "Cambria", serif；字体大小:12 pt} p。cjk {font-family: "Cambria"；字体大小:12 pt} p。ctl {font-family: “Cambria”

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引用次数: 0

Analysis and modelling of Inner Fuel Cycle performance using exhaust bypass and Direct Internal Recycling 采用排气旁通和直接内循环的内燃料循环性能分析与建模

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-02-11 DOI: 10.1016/j.fusengdes.2026.115668

Federico Hattab , Yuri Igitkhanov , Vincenzo Narcisi , Alessia Santucci , Fabio Giannetti , Giulia Valeria Centomani , Paul A. Staniec , Richard Kembleton , Thomas Giegerich

Fusion power plants require robust fuel cycle (FC) architectures that minimize tritium inventories while managing impurity build-up and isotopic imbalances. This work investigates the performance of a Inner Fuel Cycle (IFC) architecture based on the Direct Internal Recycling (DIR) concept, with an additional bypass loop for recycling of exhaust gases utilized in gas puffing. Particular focus is given to fuel dilution due to impurity accumulation and deuterium–tritium (D–T) imbalance. A new Julia-based dynamic fuel cycle modeller, MINERVA (Modelling and Integration of Nuclear fusion Energy Reactor fuel cycle for Versatile Analysis), is introduces and used to evaluate the performances of the proposed architecture and for understanding the dynamics and criticalities of a DIR-based FC. Protium build-up is identified as a potential challenge, with accumulation becoming problematic at high separation efficiencies without dedicated removal systems. Two reactor case studies are analysed, EU DEMO 2018 and Gauss Fusion’s GIGA reactor. Results demonstrate that the proposed architecture effectively manages impurity concentrations below 1% for protium while maintaining optimal D–T ratios through active control systems. The proposed architecture achieves significant reductions in external fuel requirements, with effective conversion ratios growing exponentially with DIR separation efficiency. The bypass loop successfully provides the majority of gas puffing requirements without causing excessive impurity accumulation. This work establishes a foundation for advanced fuel cycle optimization studies essential for the development of commercial fusion power plants.

核聚变发电厂需要强大的燃料循环（FC）架构，以最大限度地减少氚库存，同时管理杂质积累和同位素失衡。这项工作研究了基于直接内部回收（DIR）概念的内燃料循环（IFC）架构的性能，该架构具有用于气体膨化中废气回收的附加旁路回路。特别关注的是由于杂质积累和氘氚（D-T）不平衡造成的燃料稀释。介绍了一种新的基于julia的动态燃料循环建模器MINERVA（用于通用分析的核聚变能反应堆燃料循环建模和集成），并用于评估所提出的体系结构的性能，并用于理解基于dir的FC的动态和临界性。Protium积聚被认为是一个潜在的挑战，在没有专用去除系统的情况下，在高分离效率下积聚会成为一个问题。分析了两个反应堆案例研究，欧盟DEMO 2018和高斯聚变的GIGA反应堆。结果表明，所提出的结构有效地将杂质浓度控制在1%以下，同时通过主动控制系统保持最佳的D-T比。所提出的结构显著降低了外部燃料需求，有效转化率随着DIR分离效率呈指数级增长。旁路回路成功地提供了大部分的气体膨化要求，而不会造成过多的杂质积累。这项工作为先进的燃料循环优化研究奠定了基础，这对商业核聚变发电厂的发展至关重要。

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引用次数: 0

Particle and thermal transport in JET Helium and Hydrogen-Helium H-mode plasmas JET氦和氢氦h模等离子体中的粒子和热输运

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-02-13 DOI: 10.1016/j.fusengdes.2026.115664

I. Voitsekhovitch , M. Poradzinski , D. Taylor , A. Chomiczewska , H. Dudding , I. Ivanova-Stanik , D. King , M. Maslov , C. Roach , JET contributors , the EUROfusion Tokamak Exploitation Team

Fusion performance in a tokamak-reactor strongly depends on the confinement of thermalised α-particles (Helium (He) ash) in the core plasma region. Consequently, the development of He particle transport models and their validation in present experiments is an important step towards a more accurate prediction of fusion power production in future devices. In the absence of a computationally fast well-validated theory-based transport models for He, the empirical Bohm-gyroBohm (BgB) model is tested here for the first time to our knowledge in the predictive self-consistent temperature and density simulations of JET H-mode He and Hydrogen (H) - He discharges. The thermal confinement in JET He plasmas is found to be well below the Deuterium (D) BgB model reference – this result is qualitatively consistent with the observation of reduced global thermal confinement in He discharges observed on ASDEX Upgrade, Cmod, DIII-D and EAST tokamaks compared to the confinement of D plasmas. The “Helium” version of the BgB model including the re-calibrated BgB thermal diffusivity and the He particle diffusion coefficient defined as a fixed fraction of the thermal electron diffusivity is proposed here. This model is validated in the JET discharges performed at different toroidal magnetic fields, plasma densities, wall materials (Carbon and ITER-like wall) and main ion compositions. Strong reduction of He particle transport with the increase of magnetic field has been found in JET discharges. However, the simulations of the He ash accumulation in the future high-field tokamak-reactor ARC with the model validated in JET predict a tolerable amount of He content in the burn phase in the investigated parameter space, with a weak impact on the fusion power production. Similar conclusion has been drawn for the H-mode EU-DEMO scenario by extrapolating the JET He particle transport model to this device.

托卡马克反应堆的聚变性能在很大程度上取决于核心等离子体区热化α-粒子（氦（He）灰）的约束。因此，He粒子输运模型的发展及其在现有实验中的验证是朝着更准确地预测未来设备中聚变功率产生的重要一步。在缺乏计算快速且经过验证的基于理论的He输运模型的情况下，我们首次在JET H模式He和氢(H) - He放电的预测自洽温度和密度模拟中测试了经验Bohm-gyroBohm （BgB）模型。JET He等离子体中的热约束远低于氘(D) BgB模型的参考值，这一结果与在ASDEX Upgrade、Cmod、DIII-D和EAST托卡马克上观察到的He放电的整体热约束比D等离子体的约束减少的观察结果在质量上是一致的。本文提出了“氦”版本的BgB模型，包括重新校准的BgB热扩散系数和定义为热电子扩散系数固定分数的He粒子扩散系数。该模型在不同环向磁场、等离子体密度、壁材（碳和iter样壁）和主要离子成分下的JET放电中得到了验证。在JET放电中，He粒子输运随磁场的增大而显著降低。然而，利用JET验证的模型对未来高场托卡马克反应堆电弧中的He灰积累进行了模拟，预测在所研究的参数空间中，燃烧阶段的He含量是可以容忍的，对聚变发电的影响很小。通过将JET He粒子输运模型外推到该装置上，对h模EU-DEMO场景也得出了类似的结论。

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引用次数: 0

Novel divertor target structural design for extreme heat load: FEM Simulation and experimental validation 极端热负荷下新型导流器靶结构设计：有限元模拟与实验验证

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-02-04 DOI: 10.1016/j.fusengdes.2026.115653

Yu-zhong Jin , You-yun Lian , Jian-bao Wang , Fan Feng , Cheng-ming Tu , Dong zhao , Hao Wang , Zong-jian Chai , Zi-jie Wang , Xiang Liu

Designed heat flux as 10 MW m⁻² during stationary operation can be foreseen for ITER divertor tungsten, while the heat flux would be increased to even 20 MW m⁻² for future fusion devices. Design of advanced divertor target structure with enhanced cooling ability and prolonged lifetime is a popular research direction at present. In this paper, a new divertor target structure that combines flat tile concept and monoblock concept has been proposed. Finite Element Method (FEM) has been used to compare the thermal response as well as the service lifetime between ITER-like divertor monoblock and the novel divertor structure. Low-cycle fatigue damage and thermal creep rupture has been considered during simulation. A linear rule criterion has been applied to roughly estimate the creep-fatigue interaction on the armor material. The results show the novel divertor structure owns better heat transfer capability and it is expected to own longer service lifetime than ITER-like monoblock. Furthermore, high heat flux experiments have been conducted to verify its heat removal ability and structural reliability.

在固定运行时，ITER转导器钨的设计热流密度可预见为10 MW m−2，而未来聚变装置的热流密度将增加到20 MW m−2。设计具有增强冷却能力和延长使用寿命的先进导流器靶结构是目前研究的热门方向。本文提出了一种结合平面砖概念和单块概念的新型导流靶结构。采用有限元方法对iter型导流器单体结构与新型导流器结构的热响应和使用寿命进行了比较。模拟中考虑了低周疲劳损伤和热蠕变断裂。采用线性准则对装甲材料的蠕变-疲劳相互作用进行了粗略估计。结果表明，新型导流器结构具有更好的换热性能，与类似iter的单体导流器相比，具有更长的使用寿命。并进行了高热流密度试验，验证了其排热能力和结构可靠性。

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引用次数: 0

A confinement ventilation system zoning mode of nuclear fusion facilities based on equilibration time 基于平衡时间的核聚变设施约束通风系统分区模式

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

Fusion Engineering and Design

Pub Date : 2026-04-01 Epub Date: 2026-02-05 DOI: 10.1016/j.fusengdes.2026.115643

Junhao Rong, Bin Guo, Jiansheng Hu, Muhammad Salman Khan, Jinxuan Zhou, Yuqing Tang, Gang Wu

In tritium-handling fusion facilities, the heating, ventilation, and air conditioning (HVAC) system must not only satisfy thermal comfort and indoor air quality requirements but also ensure dynamic confinement to prevent the spread of radioactive materials, which is also called confinement ventilation system (CVS). According to ISO 16646 and ISO 17873, negative-pressure zoning within the CVS is adopted to control the risk of unconfined tritium release. However, for auxiliary rooms located inside tritium-controlled areas without any radioactive source terms, there remains no clear zoning strategy. Improper zoning of auxiliary rooms may lead to difficulties in achieving equilibration or increase the risk of CVS instability. To address this issue, a three-layer nested dilution-ventilation model was established based on the dilution ventilation equation and mass conservation principles. Case analyses show that increasing the volume of auxiliary spaces within the nested negative-pressure system does not vary the final equilibration concentration or total radioactive quantity, but it extends the equilibration time. Besides, equilibration time shows the linear increasing variations with V/q_v in each zones. Conversely, setting an auxiliary space to positive or normal pressure reduces the total radioactive quantity. Breaking the nest negative pressure chain and significantly increasing the equilibration time in low concentration area will be the two main drawback. Finally, three different zoning recommendations for auxiliary rooms are evaluated, based on the indicators consideration of CVS system safety, flow-rate, and ventilation equilibration time. The findings offer practical guidance for ventilation zoning design and commissioning in tritium-bearing fusion buildings.

在氚处理聚变设施中，暖通空调（HVAC）系统不仅要满足热舒适和室内空气质量要求，还要保证动态约束，防止放射性物质扩散，也称为约束通风系统（CVS）。根据ISO 16646和ISO 17873，在CVS内采用负压分区来控制无限氚释放的风险。然而，对于没有任何放射源条件的氚控制区内的辅助房间，仍然没有明确的分区策略。辅助室分区不当可能导致难以达到平衡或增加CVS不稳定的风险。针对这一问题，基于稀释通风方程和质量守恒原理，建立了三层嵌套稀释通风模型。实例分析表明，增加巢式负压系统内辅助空间的体积不会改变最终的平衡浓度或总放射线量，但会延长平衡时间。平衡时间随各区域的V/qv呈线性增加的变化。相反，将辅助空间设置为正压或正压可减少总辐射量。破坏巢负压链和显著增加低浓度区域的平衡时间将是两个主要缺点。最后，基于CVS系统安全性、流量和通风平衡时间等指标，对辅助用房的三种不同分区建议进行了评价。研究结果对含氚核聚变建筑通风分区设计和调试具有实际指导意义。

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引用次数: 0