Fusion Engineering and Design最新文献_第6页

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

PAV-2: a new mock-up to investigate niobium membrane-PAV performances optimization in PbLi systems PbLi系统中研究铌膜- pav性能优化的新模型PAV-2

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.2025.115554

Francesca Papa , Ciro Alberghi , Vincenzo Claps , Daniele Martelli , Alessandro Venturini

Permeator Against Vacuum was confirmed in 2023 as the reference technology for the tritium extraction and removal (TER) system of the Water-Cooled Lithium-Lead Breeding Blanket (WCLL BB) due to its overall better performances and higher Technology Readiness Level. The manufacturing and the first characterization of a PAV with a niobium membrane in a shell and tube configuration with U-tubes (PAV-ONE mock-up) was recently performed at ENEA Brasimone R.C., demonstrating that this technology can be satisfactorily employed in PbLi. This paper will present the design of a new PAV test section with niobium membrane to be installed in the TRIEX-II facility. The objective of the new mock-up is to investigate the correlation between the extraction flux and different parameters to optimize the future design of the technology. In particular, PAV-two will allow to deeply examine the influence of turbulence, vacuum pressure, and surface conditions on the hydrogen transport in the system and, therefore, on the performances of the technology. The simple and flexible design of PAV-two will enable the discrimination of each parameter’s impact on the extracted flux in a repeatable and reliable manner.

2023年，由于其整体性能更好、技术成熟度更高，反真空渗透技术被确定为水冷锂铅增殖毯（WCLL BB）氚提取和去除（TER）系统的参考技术。最近在ENEA Brasimone研究中心进行了带u型管壳管结构（PAV- one模型）的铌膜PAV的制造和首次表征，证明了该技术可以令人满意地应用于PbLi。本文将介绍安装在TRIEX-II设施中的新型铌膜PAV试验段的设计。新模型的目的是研究萃取通量与不同参数之间的相关性，以优化该技术的未来设计。特别是，pav - 2将允许深入研究湍流、真空压力和表面条件对系统中氢输运的影响，从而影响该技术的性能。pav - 2的设计简单灵活，可重复可靠地识别各参数对提取通量的影响。

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引用次数: 0

Benchmarks of iron nuclear data for fusion neutron sources 聚变中子源铁核数据基准

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

Fusion Engineering and Design

Pub Date : 2025-12-04 DOI: 10.1016/j.fusengdes.2025.115548

Saerom Kwon , Chikara Konno , Shogo Honda , Shunsuke Kenjo , Satoshi Sato

To assess the accuracy of iron data in the latest nuclear data libraries, mainly FENDL-3.2b, for accelerator-based fusion neutron source designs, we analyzed the QST/TIARA iron experiment with quasi mono-energy neutrons of 40 and 65 MeV, and the JAEA/FNS iron experiment with DT neutrons, by using the Monte Carlo code MCNP6.2. As the results, we found the following issues: 1) the calculation result with FENDL-3.2b underestimated the measured neutron flux in the continuous energy range (10 - 60 MeV) by 40 % in the TIARA experiment with 65 MeV neutrons, 2) it tended to underestimate the measured neutron flux above 10 MeV by 20 % at a depth of 70 cm and overestimate that below 10 keV by 30 % up to a depth of 40 cm in the FNS experiment. We modified the FENDL-3.2b iron data to investigate these issues and identified underlying remarks: 1) the non-elastic and inelastic scattering data of ⁵⁶Fe in FENDL-3.2b underestimated the measured neutron flux above 10 MeV, 2) the (n,np) data of ⁵⁶Fe in FENDL-3.2b overestimated the measured neutron flux above 10 MeV, and 3) the inelastic scattering and (n,2n) data of ⁵⁶Fe and the inelastic scattering data of ⁵⁷Fe in FENDL-3.2b caused the overestimation of the measured neutron flux below 10 keV. These issues of ^56,57Fe in FENDL-3.2b should be improved.

为了评估最新核数据库（主要是FENDL-3.2b）中铁数据在加速器聚变中子源设计中的准确性，我们利用蒙特卡罗程序MCNP6.2对QST/TIARA含40和65 MeV准单能中子的铁实验和JAEA/FNS含DT中子的铁实验进行了分析。作为结果,我们发现以下问题:1)计算结果与fendl - 3.2 b低估了测量中子通量在连续能量范围(10 - 60 MeV) 40%头饰试验65伏中子,2)倾向于低估了测量中子通量高于10兆电子伏20% 70厘米的深度和高估,低于10 keV 30%到40厘米的深度fn实验。我们修改了fendl - 32 b铁数据来调查这些问题，并确定了潜在的评论：1) FENDL-3.2b中56Fe的非弹性和非弹性散射数据低估了10mev以上的中子通量，2)FENDL-3.2b中56Fe的（n,np）数据高估了10mev以上的中子通量，3)FENDL-3.2b中56Fe的非弹性散射和（n,2n）数据以及57Fe的非弹性散射数据导致了10kev以下中子通量的高估。FENDL-3.2b中56,57fe的这些问题需要改进。

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引用次数: 0

Mathematical modeling and design of a microwave polarizer for DTT ECRH applications DTT ECRH微波偏振器的数学建模与设计

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

Fusion Engineering and Design

Pub Date : 2025-12-03 DOI: 10.1016/j.fusengdes.2025.115552

Sofia Bertolami , Franco Di Paolo , Alessandro Bruschi , Francesco Fanale , Alessandro Moro , Saul Garavaglia , Gustavo Granucci , Afra Romano , Alessandro Simonetto

In an Electron Cyclotron Resonance Heating (ECRH) system, to efficiently couple the signal power to the plasma, the signal wave polarization must be accurately matched to the plasma conditions at the plasma boundary. However, the millimeter-wave radiation from the power source (gyrotron) is normally linearly polarized: consequently, some kind of polarization matching is required. This study focuses on the design of a grating polarizer with sinusoidal grooves for the 170 GHz ECRH system, with an application specifically intended for the Divertor Tokamak Test (DTT), currently under construction in Frascati, Italy. To enable the generation of all possible output polarization states, a pair of polarizer mirrors will be employed and integrated into the Quasi-Optical (QO) transmission line connecting the gyrotrons to the Electron Cyclotron (EC) waves launchers. The primary objective of this study is to describe an analytical tool capable of providing detailed insights into the polarization characteristics of the reflected electric field resulting from the interaction between the incident wave and the polarizer. Additionally, the proposed program tool calculates the precise combinations of rotation angles required for the polarizers to achieve the desired output polarization states. The accuracy and reliability of the model’s prediction have been validated by comparing them with simulations conducted using commercial electromagnetic software.

在电子回旋共振加热（ECRH）系统中，为了有效地将信号功率与等离子体耦合，必须精确匹配等离子体边界处的信号波极化情况。然而，来自电源（回旋管）的毫米波辐射通常是线极化的，因此需要进行某种极化匹配。本研究的重点是为170 GHz ECRH系统设计一种带有正弦沟槽的光栅偏振器，该偏振器专门用于目前正在意大利弗拉斯卡蒂进行的托卡马克转向器测试（DTT）。为了能够产生所有可能的输出偏振状态，将使用一对偏振镜并将其集成到连接回旋加速器和电子回旋加速器（EC）波发射器的准光学（QO）传输线中。本研究的主要目的是描述一种分析工具，该工具能够提供详细的见解，以了解由入射波和偏振器之间的相互作用产生的反射电场的偏振特性。此外，所提出的程序工具计算所需的旋转角度的精确组合，以实现所需的输出偏振状态的偏振器。通过与商用电磁软件模拟结果的对比，验证了模型预测的准确性和可靠性。

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引用次数: 0

Directives, codes, standards and other requirements applicable to the design and manufacture of components in the ITER project 适用于ITER项目中元件设计和制造的指令、规范、标准和其他要求

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

Fusion Engineering and Design

Pub Date : 2025-12-03 DOI: 10.1016/j.fusengdes.2025.115550

A.A. Shoshin

The description of the main regulatory documents applied in the design and construction of the elements of the international thermonuclear reactor ITER in France is given, their main requirements are presented. Significant difficulties with the design and manufacture of components arise because ITER is a nuclear facility under French law. The French classification of pressure equipment (otherwise called 'pressurized equipment') in nuclear facilities is considered, examples of ITER diagnostic port equipment are given. The difficulties arising from the application of these regulatory documents are shown. The main rules and requirements developed by the ITER Organization itself for vacuum equipment and mechanical components are listed. The main industry standards used in this project are reviewed. One possible solution that could facilitate the development and construction of fusion reactors is to develop regulations specifically for fusion plants.

介绍了在法国国际热核反应堆ITER元件设计和建造中应用的主要监管文件，并提出了它们的主要要求。根据法国法律，ITER是一个核设施，因此在设计和制造组件方面存在重大困难。考虑了法国核设施中压力设备（也称为“加压设备”）的分类，给出了ITER诊断端口设备的实例。说明了应用这些规范性文件所产生的困难。列出了ITER组织自己制定的真空设备和机械部件的主要规则和要求。回顾了本项目中使用的主要行业标准。促进核聚变反应堆发展和建设的一个可能的解决方案是制定专门针对核聚变工厂的法规。

引用次数: 0

Preliminary study on the development of the fluidic system for the fabrication of pebble with core–shell structure 核-壳结构卵石制备流体系统开发的初步研究

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

Fusion Engineering and Design

Pub Date : 2025-12-02 DOI: 10.1016/j.fusengdes.2025.115547

Young Ah Park , Yi-Hyun Park , Mu-Young Ahn , Young Soo Yoon

Li₄SiO₄, used as a tritium breeder material in fusion reactors, has high lithium density but suffers from lower mechanical strength compared to another promising material, Li₂TiO₃. In this study, a fluidic system was designed to fabricate core-shell structured pebbles (Li₄SiO₄–Li₂TiO₃) consisting of a Li₄SiO₄ core and a Li₂TiO₃ shell. The system was constructed using a T–shaped fluid flow channel with a double–tube design, enabling the controlled formation of core-shell droplets by adjusting the flow rates of the continuous and dispersed phases. The resulting Li₄SiO₄–Li₂TiO₃ pebbles exhibited a crush load approximately 2.03 times higher than that of single-phase Li₄SiO₄ pebbles, and the stable formation of the core–shell structure was confirmed. This study presents a novel fabrication process with the potential to enhance the mechanical performance of Li₄SiO₄ as a tritium breeder material.

Li4SiO4，用作聚变反应堆中的氚增殖材料，具有高锂密度，但与另一种有前途的材料Li2TiO3相比，机械强度较低。在本研究中，设计了一个流体系统来制造由Li4SiO4核和Li2TiO3壳组成的核-壳结构鹅卵石（Li4SiO4 - Li2TiO3）。该系统采用双管设计的t型流体流道，通过调节连续相和分散相的流速来控制核壳液滴的形成。所得Li4SiO4 - li2tio3鹅卵石的破碎载荷比单相Li4SiO4鹅卵石高约2.03倍，并证实了核壳结构的稳定形成。本研究提出了一种新的制造工艺，有可能提高Li4SiO4作为氚增殖材料的机械性能。

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引用次数: 0

Numerical analysis of 60° V-ribs for helium-cooled high heat flux loaded first wall, challenges and contributions 氦冷高热流负载第一壁60°v型肋的数值分析、挑战和贡献

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

Fusion Engineering and Design

Pub Date : 2025-12-02 DOI: 10.1016/j.fusengdes.2025.115527

Ch. Klein, F. Arbeiter, M Enke

The HCPB (Helium Cooled Pebble Bed) blanket concept for EU DEMO fusion reactors employs high-pressure (8 MPa) helium gas as coolant for the plasma facing first wall (FW). Up-to-date estimates for the total maximum of heat flux go up to 0.73 MW/m². Higher short-term transient loads are possible. 60°-V-shaped ribs show high heat transfer and are thus the subject of the presented studies. Although Scale-Resolving Simulation (SRS) techniques such as LES (Large Eddy Simulation) are able to calculate heat transfer and pressure drop precisely, their application is limited to an individual ribs or a few mm channel segment due to the high required mesh count. Nevertheless, SRS techniques can be used to compare different ribs and to evaluate the performance of RANS (Reynolds-Averaged Navier-Stokes Simulations). Selected RANS models can be used to evaluate the development of secondary flow structures along the channel and compare complete channels with different rib configurations for thermohydraulic performance.

The objectives of the present paper are (i) to offer guidance on the range and limits of applicability of numerical methods when dealing with cooling flows in structured channels (sections 1–4) and (ii) to provide results on specific design features of surface structures (sections 5–7) that help designers in implementing thermal-hydraulic efficient yet fabrication friendly structured channels

Challenges like long entrance length and high material properties gradients are shown. Strategies for ribs height reduction with increasing heat transfer and reduction of pressure drop resulting in higher Cooling Performance Numbers (CPN) are found. Thermohydraulic performance of fabrication friendly ribs is calculated along the channel.

用于EU DEMO聚变反应堆的HCPB（氦冷却球床）包层概念采用高压（8 MPa）氦气作为面向第一壁（FW）的等离子体的冷却剂。最新估计的最大总热通量高达0.73 MW/m²。更高的短期暂态负载是可能的。60°v形肋具有高的传热性能，因此是本研究的主题。尽管像LES（大涡模拟）这样的尺度解析模拟（SRS）技术能够精确地计算传热和压降，但由于需要很高的网格数，它们的应用仅限于单个肋或几毫米的通道段。然而，SRS技术可以用来比较不同的肋，并评估RANS （reynolds - average Navier-Stokes simulation）的性能。所选择的RANS模型可用于评估通道沿线二次流结构的发展，并比较不同肋形的完整通道的热水力性能。本论文的目的是(i)在处理结构通道中的冷却流动时，提供关于数值方法的适用范围和限制的指导（第1-4节）；（ii）提供表面结构的特定设计特征的结果（第5-7节），帮助设计师实现热水力高效且制造友好的结构通道。研究了降低肋部高度、增加传热和降低压降从而提高冷却性能数值的策略。沿通道计算了加工友好肋的热工性能。

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引用次数: 0

Nuclear, thermal, and shielding design of test cell in A-FNS A-FNS试验单元的核、热、屏蔽设计

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

Fusion Engineering and Design

Pub Date : 2025-12-02 DOI: 10.1016/j.fusengdes.2025.115549

Shogo Honda, Saerom Kwon, Shunsuke Kenjo, Makoto Oyaidzu, Kentaro Ochiai, Satoshi Sato

Significant quantities of neutrons and gamma rays deduce the nuclear heating on the shielding walls of the test cell (TC) in the A-FNS. This study presents nuclear and thermal analyses of the shielding wall in the A-FNS to evaluate its radiation-shielding effectiveness and cooling capacity. Previous designs incorporated cooling water pipes embedded in the concrete shielding wall for thermal management. However, we determined that this approach failed to maintain concrete temperatures within specified limits due to the low thermal conductivity of concrete and the challenge of ensuring adequate contact between the concrete and piping. We developed a new shielding wall design that eliminates the need for cooling pipes inside the concrete. This updated design integrates an 85 cm of stainless steel 316 L (SS316L) structure containing 20 % cooling water positioned in front of the concrete, which maintains concrete temperatures within acceptable limits. In addition, nuclear assessments of this revised shielding wall structure were conducted to verify its radiation shielding capacity. Results indicate that the effective dose criteria are met at thicknesses of 255 cm for heavy concrete and 335 cm for ordinary concrete when the 85 cm of combing structure of SS316L/water (20 % water) is included.

在A-FNS中，大量的中子和伽马射线推断出测试单元（TC）屏蔽壁上的核加热。本文对A-FNS屏蔽墙进行了核分析和热分析，以评估其辐射屏蔽效果和冷却能力。以前的设计将冷却水管嵌入混凝土屏蔽墙中进行热管理。然而，由于混凝土的低导热性和确保混凝土与管道之间充分接触的挑战，我们确定这种方法无法将混凝土温度保持在规定的范围内。我们开发了一种新的屏蔽墙设计，消除了混凝土内部冷却管道的需要。这个更新的设计集成了一个85厘米的不锈钢316L （SS316L）结构，其中20%的冷却水位于混凝土前面，将混凝土温度保持在可接受的范围内。此外，还对改进后的屏蔽墙结构进行了核评估，验证了其辐射屏蔽能力。结果表明，当SS316L/水（20%水）混合结构为85 cm时，在厚混凝土厚度为255 cm、普通混凝土厚度为335 cm时满足有效剂量标准。

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引用次数: 0