Nuclear Engineering and Design最新文献_第2页

Thermo-mechanical coupled analysis and probabilistic safety evaluation of the prestressed concrete containment vessel under accident pressure 事故压力下预应力混凝土安全壳热-力耦合分析及概率安全性评价

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-27 DOI: 10.1016/j.nucengdes.2026.114760

Hua Rong , Xuan Zhang , Yajing Shen , Shuai Tan , Xinglang Fan , Yang Du , Yifeng Feng

This study investigates the prestressed concrete containment vessel (PCCV) of the Qinshan Phase II nuclear power plant, focusing on the nonlinear thermo-mechanical coupling behavior and probabilistic safety evaluation of its load capacity under accident pressure conditions. An energy-based elastoplastic damage constitutive model for concrete at elevated temperatures was employed to develop a refined finite element model of the PCCV, comprehensively considering the effects of temperature gradients, material degradation, and thermo-mechanical coupling. Numerical simulation reveals the whole process of mechanical behaviour of the containment from elastic response, damage evolution to final failure. The results show that the stress concentration around the equipment opening was caused by the geometric discontinuity, which became the weak link of the structure and the first area of damage. To further quantify the influence of the variability in concrete strength, a stochastic damage response analysis was conducted based on the probability density evolution theory (PDET). The probabilistic evolution of damage and displacement responses at key locations was obtained, and the time-dependent reliability of the structure under different damage thresholds was evaluated. The results indicate that the randomness of concrete material strength significantly affects the damage propagation path and failure mode of the containment structure. The proposed analysis framework provides a theoretical and numerical foundation for risk assessment and reliability-based design of nuclear containment structures under accident conditions.

以秦山二期核电站预应力混凝土安全壳为研究对象，重点研究了事故压力条件下预应力混凝土安全壳的非线性热-力耦合行为及其承载能力的概率安全评价。采用基于能量的高温混凝土弹塑性损伤本构模型，综合考虑温度梯度、材料降解和热-力耦合的影响，建立了PCCV的精细有限元模型。数值模拟揭示了安全壳从弹性响应、损伤演化到最终破坏的全过程力学行为。结果表明，设备开口周围的应力集中是由几何不连续引起的，该区域成为结构的薄弱环节和首当其冲的损伤区域。为了进一步量化混凝土强度变异性的影响，基于概率密度演化理论（PDET）进行了随机损伤响应分析。得到了关键位置损伤和位移响应的概率演化，并评估了不同损伤阈值下结构的时变可靠度。结果表明，混凝土材料强度的随机性对围护结构的损伤传播路径和破坏模式有显著影响。所提出的分析框架为核安全壳结构在事故条件下的风险评估和可靠性设计提供了理论和数值基础。

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引用次数: 0

Modelling of MSRE graphite temperature in porous-medium multi-physics simulations 多孔介质多物理场模拟中MSRE石墨温度的建模

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-26 DOI: 10.1016/j.nucengdes.2026.114757

S. Amirkhosravi , A. Scolaro , F. van Niekerk , M.H. du Toit , A. Pautz

MSRs stand out as prominent candidates among advanced reactor designs, addressing the global demand for safer and more sustainable nuclear energy. Accurate multi-physics modelling is essential for the advancement of MSR technology, particularly for understanding the thermos-hydraulic behaviour of graphite under irradiation. This study focuses on developing and implementing a high-fidelity methodology within the GeN-Foam code to model graphite temperature distribution within porous-medium multi-physics simulations, using the MSRE as a benchmark. The approach combines thermal-hydraulic and neutronic modelling by using Serpent-generated cross-section data as inputs for the Gen-Foam neutronic solver. Validation against MSRE measurements performed at ORNL benchmark data confirms the framework's reliability. The axial temperature distribution yields a Mean Absolute Percentage Error (MAPE) of 1.09%, while the radial distribution shows a MAPE of 0.62%. The average graphite temperature of 935.6 K is consistent with the ORNL reference value of 936.4 K under steady-state conditions.

msr在先进反应堆设计中脱颖而出，满足了全球对更安全、更可持续核能的需求。精确的多物理场建模对于MSR技术的进步至关重要，特别是对于理解辐照下石墨的热-水力行为。本研究的重点是在GeN-Foam代码中开发和实现高保真度方法，以MSRE为基准，在多孔介质多物理场模拟中模拟石墨温度分布。该方法结合了热工和中子建模，使用蛇形生成的横截面数据作为Gen-Foam中子求解器的输入。在ORNL基准数据上进行的MSRE测量验证证实了该框架的可靠性。轴向温度分布的平均绝对百分比误差（MAPE）为1.09%，径向温度分布的平均绝对百分比误差为0.62%。稳态条件下石墨平均温度为935.6 K，与ORNL参考值96.4 K一致。

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引用次数: 0

A state-of-the-art review of R&D for the supercritical water-cooled reactor technology. Part II materials & chemistry 超临界水冷堆技术研究进展综述。第二部分：材料与化学

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-25 DOI: 10.1016/j.nucengdes.2026.114789

K. Khumsa-Ang , M. Fulger , A. Sáez Maderuelo , A. Toivonen , M. Sipova , D. Marusakova , L. Zhang , J. Macák , R. Novotny

This document presents a summary of the most relevant research and development (R&D) carried out to support the development of the only generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF).The coolant of the proposed reactor operates at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of high-grade process heat. Several collaborations have been established to support this technology under the GIF umbrella as well as through other international avenues; as a result, the development work is bolstered by a collective effort between numerous R&D institutions across Asia, Europe, and North America. The Joint European Canadian Chinese development of Small Modular Reactor Technology (ECC-SMART) collaborative project was established to encompass the design and pre-licensing requirements as well as a roadmap demonstrating the safe operation of the supercritical water small modular reactor (SCW-SMR).

One of the main challenges in the material and component aspect is the selection and qualification of a fuel cladding material that can withstand supercritical water conditions (beyond 374 °C and 22.1 MPa). The aim of the materials testing work package (WP2) in the ECC-SMART project is to achieve a deep understanding of the corrosion behavior of selected candidate cladding materials. Over 750 corrosion specimens were tested including those under nominal SCW-SMR operating conditions and also at simulated accident conditions. This article summarizes the findings from the study of corrosion behavior of non-irradiated and pre-irradiated candidate materials and from the study of the effect of chemistry and changes in the chemical properties of SCW.

本文件概述了为支持第四代国际论坛（GIF）认可的唯一第四代水冷堆的开发而进行的最相关的研究与开发（R&；D）。所建议的反应堆的冷却剂在超临界水条件下运行，允许提高工厂的热力学效率和生产高级工艺热。在GIF的框架下以及通过其他国际途径，已经建立了若干合作来支持这项技术；因此，开发工作得到了亚洲、欧洲和北美众多研发机构之间的集体努力的支持。建立了欧洲、加拿大、中国联合开发小型模块化反应堆技术（ECC-SMART）合作项目，以涵盖设计和预许可要求，以及展示超临界水小型模块化反应堆（SCW-SMR）安全运行的路线图。材料和部件方面的主要挑战之一是燃料包壳材料的选择和鉴定，该材料能够承受超临界水条件（超过374°C和22.1 MPa）。ec - smart项目中材料测试工作包（WP2）的目的是深入了解选定候选包层材料的腐蚀行为。超过750个腐蚀样本进行了测试，包括在SCW-SMR的名义运行条件下和模拟事故条件下的腐蚀样本。本文综述了未辐照和预辐照候选材料的腐蚀行为研究，以及化学效应和化学性质变化的研究结果。

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引用次数: 0

Ultra-real-time model reduction for nuclear reactor primary circuit calculation 核反应堆一次回路计算的超实时模型简化

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-24 DOI: 10.1016/j.nucengdes.2026.114781

Zelong Zhao , Honghang Chi , Yuchen Xie , Yahui Wang , Yu Ma

Ultra-real-time simulation is crucial for ensuring the safe operation and control of nuclear power plants, as it enables rapid prediction and response to thermal-hydraulic behavior under accident conditions. This study proposes an ultra-real-time thermal-hydraulic modeling approach for the reactor primary circuit based on an intrusive reduced-order model (ROM). The governing equations of all components are discretized using a finite difference scheme, and variables for establishing ROMs are selected from these discretized equations to eliminate nonlinear terms. The transient solutions obtained from the initial 5% of time steps, calculated by the full-order model, served as snapshots, from which characteristic modes are extracted using the proper orthogonal decomposition. By projecting the discretized governing equations of each component onto the characteristic mode space, ultra-real-time thermal-hydraulic ROMs are constructed for each component. The integration of these ROMs for all components resulted in a comprehensive ultra-real-time model (URTM) of the primary circuit, capable of predicting system evolution. Simulation results demonstrated that the URTM achieves ultra-real-time performance while maintaining a maximum relative error of less than 0.15% for key thermal-hydraulic parameters.

超实时仿真是确保核电站安全运行和控制的关键，因为它可以快速预测和响应事故条件下的热工水力行为。提出了一种基于侵入式降阶模型（ROM）的反应堆一次回路超实时热工建模方法。采用有限差分格式对各分量的控制方程进行离散化，并从这些离散化方程中选择建立rom的变量，消除非线性项。由全阶模型计算的前5%的时间步长得到的瞬态解作为快照，利用适当的正交分解从中提取特征模态。通过将各部件的离散化控制方程投影到特征模态空间上，构建了各部件的超实时热液rom。将这些rom集成到所有元件中，形成了一个全面的主电路超实时模型（URTM），能够预测系统的演变。仿真结果表明，URTM在实现超实时性的同时，对关键热液参数保持了小于0.15%的最大相对误差。

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引用次数: 0

The validation of fission response function neutron transport code FLASH in AP1000 reactor core AP1000堆芯裂变响应函数中子输运代码FLASH的验证

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-24 DOI: 10.1016/j.nucengdes.2026.114759

Honglong Li , Xinxiang Long , Yunxin Zhang , Donghao He , Xiaojing Liu

FLASH, based on the fission response function (FRF) theory, is a high-fidelity and low-cost adaptive neutronics code. In this study, FLASH is validated in the AP1000 pressurized water reactor whole-core problem at hot zero power condition. It features a highly heterogeneous core arrangement with enrichment zoning, WABA and IFBA utilization. To accurately simulate the reactor core, an FRF database comprising 20 distinct assembly types was developed, and environmental factors were employed to account for local effects. Compared with Monte Carlo reference calculation, the FLASH AP1000-3D whole-core calculation yielded a

k_{e f f}

error of 236 pcm and a RMS error in the pin-wise fission rate distribution of 0.99%. The entire core calculation was completed in less than 2 min.

FLASH是一种基于裂变响应函数（FRF）理论的高保真、低成本自适应中子码。在本研究中，FLASH在AP1000压水堆热零功率条件下的全堆芯问题中进行了验证。它的特点是具有富集带、WABA和IFBA利用的高度非均匀的核心排列。为了准确地模拟反应堆堆芯，开发了一个包含20种不同组件类型的FRF数据库，并采用环境因素来解释局部影响。与蒙特卡罗参考计算相比，FLASH AP1000-3D全芯计算在引脚方向裂变率分布上的keff误差为236 pcm， RMS误差为0.99%。整个岩心计算在不到2分钟的时间内完成。

引用次数: 0

CIGMA experiments on integral phenomena related to thermal hydraulics in a reactor containment vessel and building during a severe accident 严重事故中反应堆安全壳和建筑物内热工力学相关整体现象的CIGMA实验

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-23 DOI: 10.1016/j.nucengdes.2026.114787

Satoshi Abe, Ari Hamdani, Shu Soma, Ryosuke Hangai, Masashi Ohmori, Akihiko Ohwada, Toshihito Ohmiya, Yasuteru Sibamoto

The Fukushima Daiichi accident underscored the urgent need to understand complex thermal-hydraulic phenomena governing containment integrity and gas mixture distribution during a severe accident. In response, the Japan Atomic Energy Agency (JAEA) established the CIGMA (Containment InteGral Measurement Apparatus) facility, a flagship large-scale installation capable of high-temperature, high-pressure experiments with a steam-air‑helium gas mixture. This paper presents key findings from a comprehensive experimental campaign with CIGMA. The JT-SJ series demonstrated the effectiveness of external surface cooling in suppressing top head flange overheating. The CC-SP series revealed spray-induced mixing mechanisms that rapidly homogenize flammable stratifications. The CC-PL series identified condensation processes of the gas mixture that are decisive for containment cooling strategies. Finally, the CC-SJ series provided insights into inter-compartment gas transport relevant to the multi-stage explosions in Unit 3 of Fukushima Daiichi. These results establish a high-fidelity experimental database, offering benchmarks for CFD validation and advancing the development of robust hydrogen mitigation and accident management strategies worldwide.

福岛第一核电站事故强调了迫切需要了解在严重事故中控制安全壳完整性和气体混合物分布的复杂热水力现象。为此，日本原子能机构（原子能机构）建立了CIGMA（安全壳整体测量装置）设施，这是一个旗舰大型装置，能够用蒸汽-空气-氦气混合物进行高温高压实验。本文介绍了CIGMA综合实验活动的主要发现。JT-SJ系列证明了外表面冷却在抑制顶封法兰过热方面的有效性。CC-SP系列揭示了喷雾诱导的混合机制，可以快速均匀化可燃分层。CC-PL系列确定了对安全壳冷却策略具有决定性作用的气体混合物的冷凝过程。最后，CC-SJ系列提供了与福岛第一核电站3号机组多级爆炸有关的隔间间气体输送的见解。这些结果建立了一个高保真度的实验数据库，为CFD验证提供了基准，并推动了全球范围内稳健的氢缓解和事故管理策略的发展。

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引用次数: 0

Scaling sensitivity and data applicability analysis of small-scale integral test to the ESBWR small break loss-of-coolant accidents ESBWR小破口失冷事故小尺度积分试验的尺度敏感性及数据适用性分析

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-22 DOI: 10.1016/j.nucengdes.2026.114788

Xueyan Zhang, Yixuan Zhang, Jun Yang

Employing system-code simulations, the study investigates scaling effects on transient simulations, uncertainty and sensitivity quantification, and data applicability analysis between the Purdue University Multi-Dimensional Integral Test Assembly and the Economic Simplified Boiling Water Reactor. The analysis focuses on refining Best Estimate Plus Uncertainty methodologies for nuclear reactor safety by evaluating the impact of scaling between test facilities and their prototypes. Transient simulation results highlight the importance of precise scaling in replicating system behaviours during small break loss-of-coolant accidents, with particular emphasis on the discrepancies in dynamic responses. In uncertainty and sensitivity quantification, variations due to scaling notably influence the magnitude and correlation of critical parameters such as the Depressurization Valve discharge coefficient and core thermal hydraulic diameter, underscoring the necessity for accurate model scaling in safety assessments. Furthermore, data applicability analysis, bolstered by dimensionless number evaluations, reveals essential insights into the extent to which scaled experimental data can mirror prototype phenomena, thereby emphasizing the pivotal role of scaling in experimental setups for nuclear safety analysis. Collectively, these findings advance the accuracy of predictive safety evaluations and contribute significantly to the enhancement of nuclear safety standards and methodologies.

采用系统代码模拟，研究了普渡大学多维积分测试组件与经济简化沸水反应堆之间的瞬态模拟、不确定性和敏感性量化以及数据适用性分析的尺度效应。分析的重点是通过评估试验设施及其原型之间的尺度影响来改进核反应堆安全的最佳估计加不确定性方法。瞬态模拟结果强调了在小的冷却剂中断损失事故中精确缩放系统行为的重要性，特别强调了动态响应的差异。在不确定性和敏感性量化中，由于尺度变化而引起的变化显著影响关键参数的大小和相关性，如减压阀排放系数和岩心热液直径，强调了在安全评估中精确模型尺度的必要性。此外，在无量纲数评估的支持下，数据适用性分析揭示了缩放实验数据在多大程度上可以反映原型现象的基本见解，从而强调了缩放在核安全分析实验设置中的关键作用。总的来说，这些发现提高了预测性安全评价的准确性，并对提高核安全标准和方法作出了重大贡献。

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引用次数: 0

Effect of FDS uncertainty in fire simulations of nuclear power plants under different ventilation conditions 不同通风条件下核电厂火灾模拟中FDS不确定性的影响

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-21 DOI: 10.1016/j.nucengdes.2025.114733

María González-Alvear , Mariano Lázaro , Daniel Alvear , Eugenia Morgado , Miguel Ángel Jiménez , David Lázaro

Fire Dynamics Simulator (FDS) is a well-known fire computer model, which has been widely applied for different scenarios. In particular, several standards and guidelines support its use in fire safety engineering approaches in nuclear plants. Although the uncertainty of the FDS model has been analysed and collected in literature, the influence of each input parameter has not yet been fully addressed.

Some of these previous contributions were based on the Benchmark Exercise No. 3 of the International Collaborative Fire Model Project (NUREG 6905). Moreover, the Best Practice Guidelines of NEA/CSNI/R (2014)11 were used as the reference to analyse the influence of the boundary conditions on simulation results. This work aims to study the impact of selected key fire dynamics parameters on the simulations of that scenario, updating previous findings.

Since this fire scenario involves three horizontal cable trays and one vertical cable tray, it is of special interest for nuclear power plants. Moreover, it is relevant to analyse the influence of input parameters on cable ignition. A sensitivity analysis was conducted, to evaluate the most important parameters for the selected scenario, focussing on ventilation and the thermal properties of the cables such as conductivity, specific heat, density, emissivity.

The results show the influence of each parameter in the surface temperature and heat flux in the different cable trays. Consequently, this enables the authors to formulate some recommendations for defining fire scenarios when applying fire safety engineering principles in nuclear power plants.

火灾动力学模拟器（Fire Dynamics Simulator， FDS）是一种广为人知的火灾计算机模型，已广泛应用于不同的场景。特别是，一些标准和指南支持在核电站的消防安全工程方法中使用它。虽然FDS模型的不确定性已经在文献中进行了分析和收集，但每个输入参数的影响尚未得到充分解决。其中一些先前的贡献是基于国际协同火灾模型项目（NUREG 6905）的第3号基准演习。并参考NEA/CSNI/R(2014)11的最佳实践指南，分析边界条件对仿真结果的影响。这项工作旨在研究选定的关键火灾动力学参数对该情景模拟的影响，更新先前的研究结果。由于这种火灾场景涉及三个水平电缆桥架和一个垂直电缆桥架，因此对核电站来说特别有趣。此外，分析输入参数对电缆点火的影响也是有意义的。进行了敏感性分析，以评估所选方案的最重要参数，重点关注通风和电缆的热性能，如电导率、比热、密度、发射率。结果显示了各参数对不同电缆桥架表面温度和热流密度的影响。因此，这使作者能够在核电站应用消防安全工程原则时制定一些确定火灾情景的建议。

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引用次数: 0

Gas-liquid flow characteristics through a Micron scale orifice of failed fuel pin in Lead-bismuth cooled reactors 铅铋冷却堆失效燃料销微米级孔内气液流动特性研究

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-21 DOI: 10.1016/j.nucengdes.2026.114772

Yuchen Li, Yanmin Zhou, Haifeng Gu, Yichen Zhang, Shuolei Fan

Fuel cladding in lead‑bismuth eutectic (LBE) cooled reactors may develop micron-scale failures during long-term high-temperature operation. The flow regime of gas submerged jet at the defect orifice directly influences the scrubbing behavior of these fission products. This study investigated a micron-scale failures in LBE system by conducting visualization experiments using inert gas and deionized water as working fluids. The research first systematically compared the criteria for bubbling-jet regime transition between micron-scale and millimeter-scale orifices. The results revealed that the traditional criteria based on the liquid-phase Weber number or Mach number, which are applicable for millimeter-scale orifices failed at micron scale. The phenomenon was primarily attributed to the lower gas momentum from micron-scale orifice, which allowed for continued flow regime evolution even after critical flow was reached. Consequently, a new criterion defined as the product of a density correction factor and the liquid Weber number was proposed to accurately predict the flow regime transition. Furthermore, an empirical correlation for predicting the Sauter mean diameter of the gas bubbles was established based on the SPARC90 bubble size prediction model, with a prediction error within the range of −15% to +15%. Finally, the results were extrapolated to a prototypical lead‑bismuth environment through scaling analysis, which verified a good predictive capability for bubble size generated from micron-scale orifices of this correlation. The research provided theoretical support for two-phase flow regime identification and bubble behavior prediction in the safety analysis of lead‑bismuth reactors.

铅铋共晶（LBE）冷却反应堆的燃料包壳在长期高温运行中可能发生微米级故障。缺陷孔处气体浸没射流的流动状况直接影响着这些裂变产物的洗涤行为。以惰性气体和去离子水为工质，对LBE系统的微米级故障进行了可视化实验研究。本研究首先系统地比较了微米级和毫米级孔口的起泡射流过渡准则。结果表明，传统的基于液相韦伯数或马赫数的判据在微米尺度下失效。这一现象主要是由于来自微米级孔板的气体动量较低，即使在达到临界流量后，也可以继续进行流型演变。因此，提出了密度修正系数与液体韦伯数乘积的新判据来准确预测流型转变。在SPARC90气泡尺寸预测模型的基础上，建立了预测气泡Sauter平均直径的经验相关性，预测误差在- 15% ~ +15%之间。最后，通过尺度分析将结果外推到典型的铅铋环境中，验证了该相关性对微米尺度孔产生的气泡尺寸的良好预测能力。该研究为铅铋堆安全性分析中的两相流型识别和气泡行为预测提供了理论支持。

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引用次数: 0

Design and pilot production of Fully Ceramic Microencapsulated (FCM®) fuels fabricated by sintering 全陶瓷微封装（FCM®）燃料的设计和中试生产

IF 2.1 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design

Pub Date : 2026-01-21 DOI: 10.1016/j.nucengdes.2025.114745

Cameron Hilliard , Ethan Deters , Mason Phillips , Caen Ang

Fully Ceramic Microencapsulated (FCM®) fuels produced by sintering has been demonstrated in a method that does not damage fuel particles, due to assembling particles into layers. Pilot manufacturing data (N = 32) of the designed fuel is presented. Screening of SiC powders indicated that sintered densities are a function of powder size and compaction pressure. Abiding particles are postulated to prevent interparticle infiltration of powder. Image analysis of X-ray Computed Tomography (XCT) and optical microscopy cross-sections show a trend of decreasing matrix density between particles, and density is underestimated by metallographic methods. Interparticle porosity was always observed, limiting matrix density (86–97 %). Appropriate green forming can ensure all fired porosity is closed against He, water, and O₂ penetration. TRISO packing fractions were between 36 and 38 % and the packing has room for improvement. For industry adoption, development of radiation-stable, pressureless sintering of SiC is recommended.

通过烧结生产的全陶瓷微封装（FCM®）燃料已被证明在一种不损坏燃料颗粒的方法中，由于将颗粒组装成层。给出了所设计燃料的中试生产数据（N = 32）。SiC粉末的筛选表明，烧结密度是粉末粒度和压实压力的函数。假定持久的颗粒可以防止粉末的颗粒间渗透。x射线计算机断层扫描（XCT）和光学显微镜的图像分析显示，颗粒之间的基质密度呈下降趋势，而金相方法低估了密度。总是观察到颗粒间孔隙度，限制了基体密度（86 - 97%）。适当的绿色成形可以确保所有的烧成孔隙都是封闭的，不受He、水和O2的渗透。三iso填料分数在36 - 38%之间，填料有改进的空间。为供工业采用，建议发展辐射稳定、无压烧结的碳化硅。

引用次数: 0