Annals of Nuclear Energy最新文献_第6页

Effect of Bi2O3 on structural, optical and radiations shielding properties of glass system Bi2O3对玻璃体系结构、光学和辐射屏蔽性能的影响

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-17 DOI: 10.1016/j.anucene.2025.112074

Emran Eisa Saleh , Dua’a Anis Taya , Adel Ahmed Saeed , Mohammed A. Algradee , Mohammed M. Damoom , Essam Banoqitah

This research focuses on the synthesis and characterization of glass compositions containing Bi₂O₃ as a dopant for enhanced radiation shielding capabilities. The compositions, formulated as (80-x)B₂O₃ + 10Na₂O + 10BaCl₂ + xBi₂O₃ with x values of 0, 15, 30, and 45 mol%, were subjected to X-ray diffraction analysis to examine their structural properties. The X-ray diffraction analysis indicated that samples Bi0.0 and Bi0.15 exhibit an amorphous structure, while samples Bi0.30 and Bi0.45 display crystallization. Furthermore, the density of the samples was determined using the Archimedes method and the results reveled an increase from 2.5 to 5.4 g/cm³. The optical properties of the prepared glass samples were measured using UV–VIS spectrophotometer within the spectral range of 200–1200 nm. The efficiency of radiation shielding was evaluated through the Phy-x program software, providing insights into the attenuation of radiation across the different Bi₂O₃ doping levels. The findings illustrate a notable rise in the mass attenuation coefficient from 13.32 to 87.74 cm²/g correlating with the increased Bi₂O₃ content. Additionally, the effective atomic number exhibited an escalation from 30.75 to 76.96 cm²/g. The effective electron density demonstrated an increase in the glass samples, rising from 10.3 × 10²³ to 13.2 × 10²³ el/g for samples Bi0.0 and Bi0.15. In contrast, the effective electron density exhibited a decrease in the Bi0.30 and Bi0.45 samples, declining from 10.3 × 10²³ to 8.2 × 10²³ el/g. Simultaneously, the half-value layer witnessed a decrease from 0.021 to 0.001 cm, while the mean free path concurrently decreased from 0.030 to 0.002 cm. Moreover, the fast neutron removal cross section was examined, the results show the values fluctuating between 0.093 and 0.105 for glass samples.

本研究的重点是合成和表征含有Bi2O3作为增强辐射屏蔽能力掺杂的玻璃组合物。配制成（80-x）B2O3 + 10Na2O + 10BaCl2 + xBi2O3， x值分别为0、15、30和45 mol%，用x射线衍射分析其结构性质。x射线衍射分析表明，样品Bi0.0和Bi0.15表现为非晶结构，而样品Bi0.30和Bi0.45表现为结晶结构。此外，用阿基米德法测定了样品的密度，结果显示从2.5 g/cm3增加到5.4 g/cm3。用紫外-可见分光光度计在200 ~ 1200 nm光谱范围内测定了所制备的玻璃样品的光学性能。通过Phy-x程序软件评估了辐射屏蔽的效率，从而深入了解了不同Bi2O3掺杂水平下的辐射衰减。结果表明，随着Bi2O3含量的增加，质量衰减系数从13.32增加到87.74 cm2/g。有效原子序数由30.75增加到76.96 cm2/g。在Bi0.0和Bi0.15样品中，有效电子密度从10.3 × 1023 el/g上升到13.2 × 1023 el/g。相比之下，Bi0.30和Bi0.45样品的有效电子密度呈下降趋势，从10.3 × 1023 el/g下降到8.2 × 1023 el/g。同时，半值层从0.021 cm减小到0.001 cm，平均自由程从0.030 cm减小到0.002 cm。此外，对玻璃样品的快中子去除截面进行了测试，结果表明该值在0.093 ~ 0.105之间波动。

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引用次数: 0

Fault diagnosis strategy for nuclear power plant CVS based on multi-model fusion ensemble learning 基于多模型融合集成学习的核电厂CVS故障诊断策略

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-17 DOI: 10.1016/j.anucene.2025.112075

Yihan Huang , Daogang Lu , Yu Liu , Chunjing Song

This study aims to address the challenge of fault diagnosis in the Chemical and Volume Control System (CVS), a critical auxiliary system in nuclear power plants. The CVS plays a pivotal role in ensuring the overall safety of these plants by regulating water chemistry, maintaining volume balance, and controlling reactivity—core functions that directly impact operational safety. Existing monitoring methods rely heavily on manual intervention and experience, resulting in suboptimal real-time performance and difficulty detecting minor faults. Such limitations can lead to delayed responses and pose significant risks to nuclear safety. To overcome these challenges, this study proposes a multi-model fusion ensemble learning method that integrates XGBoost, Random Forest, and LSTM for the first time using a hybrid boosting and stacking strategy for CVS fault diagnosis. Optimal parameters were determined through optimization, employing a 15-s time window and a 0.75:0.25 data split, based on training with a full-range simulation dataset. The results demonstrate that the model achieved a test accuracy of 99.99% (standard deviation < 0.0001), representing a 1.2% improvement over the single-model voting method, with a prediction time of 15 ms per instance—thereby meeting real-time monitoring requirements. This study offers an efficient and precise diagnostic solution for CVS, supporting intelligent safety management in nuclear power plants and providing significant theoretical and practical value for the intelligent upgrading of the nuclear energy industry.

摘要本研究旨在解决核电厂关键辅助系统—化学与容积控制系统（CVS）的故障诊断问题。CVS通过调节水化学、保持体积平衡和控制反应性等直接影响运行安全的核心功能，在确保这些工厂的整体安全方面发挥着关键作用。现有的监测方法严重依赖人工干预和经验，导致实时性能欠佳，难以检测小故障。这种限制可能导致反应迟缓，并对核安全构成重大风险。为了克服这些挑战，本研究首次提出了一种多模型融合集成学习方法，该方法将XGBoost、随机森林和LSTM结合起来，采用混合增强和叠加策略用于CVS故障诊断。在全范围模拟数据集训练的基础上，采用15 s的时间窗和0.75:0.25的数据分割，通过优化确定最优参数。结果表明，该模型的测试精度达到99.99%（标准差<； 0.0001），比单模型投票方法提高了1.2%，预测时间为每实例15 ms，从而满足实时监控要求。本研究为CVS提供了一种高效、精准的诊断解决方案，为核电厂的智能化安全管理提供了支撑，为核能产业的智能化升级提供了重要的理论和实践价值。

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引用次数: 0

Adaptation of data-driven stratified sampling method to characterization of decommissioning radioactive waste 数据驱动分层抽样方法在退役放射性废物表征中的应用

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-16 DOI: 10.1016/j.anucene.2025.112067

Seungbin Yoon , Younghun Kweon , Sangwon Lee , Woo Nyun Choi , Hee Reyoung Kim

A data-driven stratified sampling code toolkit for radioactive waste characterization during decommissioning of nuclear power plants was developed and evaluated. The toolkit integrates statistical clustering algorithms (K-means, hierarchical clustering, and Gaussian mixture models) for strata definition with quantitative evaluation of sampling representativeness and 3D visualization of sampling locations. Using synthetic voxel-based datasets generated from prescribed contamination distributions, the toolkit was systematically validated through Monte Carlo simulation. The results show that data-driven stratified sampling substantially reduces estimation bias and sampling variance compared with conventional grid-based “hotspot” sampling. By providing a structured framework to optimize sampling strategies and estimation accuracy, the proposed approach supports cost-effective, transparent, and reliable regulatory decision-making for decommissioning radioactive waste management.

开发并评估了核电厂退役期间放射性废物表征的数据驱动分层抽样代码工具包。该工具包集成了用于地层定义的统计聚类算法（K-means，分层聚类和高斯混合模型），以及采样代表性的定量评估和采样位置的3D可视化。使用从规定的污染分布生成的基于合成体素的数据集，通过蒙特卡罗模拟系统地验证了该工具包。结果表明，与传统的基于网格的“热点”抽样相比，数据驱动的分层抽样大大降低了估计偏差和抽样方差。通过提供一个结构化的框架来优化采样策略和估计精度，所提出的方法为退役放射性废物管理提供了具有成本效益、透明和可靠的监管决策。

引用次数: 0

Measurement of 35Cl(n,p0)35S cross sections at 0.565 MeV, 1.2 MeV, and 5 MeV with a Lithium-6 enriched CLYC scintillator 用富集锂6的CLYC闪烁体测量35Cl(n,p0)35S在0.565 MeV、1.2 MeV和5mev下的截面

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-15 DOI: 10.1016/j.anucene.2025.112065

So Kamada , Masayuki Hagiwara

The cross sections for the

^{35} Cl (n, p_{0})^{35} S

reaction were measured at neutron energies of 0.565, 1.2, and 5.0 MeV using a Li-6 enriched CLYC scintillator with pulse shape discrimination (PSD) at the monochromatic neutron field of the Facility of Radiation Standards (FRS), Japan Atomic Energy Agency (JAEA). The present results were compared with the JENDL-5 and ENDF/B-VIII.1 nuclear data evaluations and with other recent measurements. At 0.565 MeV, the measured cross section of 6.8(1.2) mb is more than an order of magnitude higher than both evaluations, but remains consistent with the average trend of recent measurements. At 1.2 and 5.0 MeV, the cross sections of 43.6(3.6) mb and 44.3(2.6) mb are systematically lower than the evaluations, but in better agreement with JENDL-5 than with ENDF/B-VIII.1. Over the entire energy range, the measured cross sections agree with recent measurements within uncertainties, whereas the evaluations generally fall outside the experimental bounds. These discrepancies at fission-relevant energies reinforce the need for updated chlorine evaluations for use in fast reactor applications.

在日本原子能机构（JAEA）辐射标准设施（FRS）的单色中子场中，利用具有脉冲形状识别（PSD）的富集Li-6的CLYC闪烁体，测量了35Cl(n,p0)35S反应在0.565、1.2和5.0 MeV中子能量下的截面。本研究结果与JENDL-5和ENDF/B-VIII进行比较。1 .核数据评价和其他最近的测量。在0.565 MeV处，测量到的6.8(1.2)mb的截面比这两个评估值高出一个数量级以上，但与最近测量的平均趋势保持一致。在1.2 MeV和5.0 MeV下，43.6(3.6)mb和44.3(2.6)mb的截面比评估值低，但与JENDL-5比ENDF/B-VIII.1更符合。在整个能量范围内，测量的截面与不确定度内的最近测量一致，而评估通常落在实验界限之外。在与裂变有关的能量上的这些差异加强了在快堆应用中更新氯评价的必要性。

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引用次数: 0

System simulation and evaluation for an integral natural circulation small modular reactor 整体式自然循环小型模块化反应堆系统仿真与评价

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-15 DOI: 10.1016/j.anucene.2025.112059

Zhenliang Zhang, Xianping Zhong

The natural circulation integral pressurized water reactor (NC-iPWR) is a small modular reactor (SMR) that integrates key components such as the reactor core, steam generator, and pressurizer within a single reactor pressure vessel (RPV), achieving a compact configuration and offering significant potential for enhancing inherent safety. However, this high level of integration also introduces strong couplings that result in nonlinear system behavior and increased sensitivity to operating conditions. To address these challenges, this study develops a system-level simulation model that incorporates a multi-node core heat transfer model, a moving-boundary steam generator model, and a multi-region non-equilibrium pressurizer model. Steady-state validation against benchmark data from the NuScale design demonstrates the model’s capability to accurately reproduce the representative thermal–hydraulic characteristics of NC-iPWR. Results from uncertainty and sensitivity analysis indicate that the initial downcomer temperature, gap size, and initial primary pressure are the key parameters affecting steady-state performance. Under transient disturbances, the system exhibits favorable load-following capability and maintains inherent safety margins.

自然循环整体式压水堆（NC-iPWR）是一种小型模块化反应堆（SMR），它将堆芯、蒸汽发生器和稳压器等关键部件集成在一个反应堆压力容器（RPV）中，实现了紧凑的配置，并为提高固有安全性提供了巨大的潜力。然而，这种高水平的集成也引入了强耦合，导致非线性系统行为和对操作条件的敏感性增加。为了解决这些挑战，本研究开发了一个系统级仿真模型，该模型包含一个多节点核心传热模型、一个移动边界蒸汽发生器模型和一个多区域非平衡增压模型。根据NuScale设计的基准数据进行的稳态验证表明，该模型能够准确再现NC-iPWR的代表性热工特性。不确定性和敏感性分析结果表明，初始降流器温度、间隙尺寸和初始主压力是影响稳态性能的关键参数。在瞬态扰动下，系统表现出良好的负载跟踪能力，并保持固有的安全裕度。

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引用次数: 0

Identification of neutron production mechanisms and spectra in implosion and heating experiments of laser fusion 激光聚变内爆和加热实验中中子产生机制和光谱的鉴定

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-13 DOI: 10.1016/j.anucene.2025.112061

A.Youssef , R. Kodama

Mechanisms of neutron production and their spectra in implosion and heating experiments of laser fusion using CD targets were investigated. In implosion experiments, the neutron spectrum is only 2.45 MeV neutrons due to the thermonuclear fusion mechanism of the D(d,n)³He reaction. In heating experiments, accelerated deuterium and carbon ions induce beam fusion mechanism of the ¹²C(d,n)¹³N and D(¹²c,n)¹³N reactions with neutron yield of each reaction being higher than that of the beam fusion of the D(d,n)³He reaction. Deuteron break-up mechanism is possible through the ¹²C(d,np)¹²C and D(d,np)D reactions. Photodissociation mechanism of deuteron and carbon can occur via the ¹²C(γ,n)¹¹B and D(γ,n)¹H reactions. A detailed analysis of the measured and calculated neutron spectrum using a 3D Monte Carlo code has identified the proper mechanisms involved in neutron production. The relative participation of each mechanism in the overall neutron yield was determined. The deuteron break-up, deuteron electro-disintegration and photonuclear reactions were discussed but their contributions in the neutron production process are negligible under our irradiation conditions.

研究了CD靶激光聚变的内爆和加热实验中中子产生的机理及其光谱。在内爆实验中，由于D(D,n)3He反应的热核聚变机制，中子谱仅为2.45 MeV中子。在加热实验中，加速的氘和碳离子诱导了12C(d,n)13N和d (12C,n)13N反应的束聚变机制，每个反应的中子产率都高于d (d,n)3He反应的束聚变。通过12C(d,np)12C和d (d,np) d反应可以实现氘核分裂机制。氘核与碳的光解作用机制可通过12C(γ,n)11B和D(γ,n)1H反应进行。利用三维蒙特卡罗程序对测量和计算的中子谱进行了详细分析，确定了中子产生的适当机制。确定了各机制在总中子产率中的相对作用。讨论了氘核分裂、氘核电解体和光子核反应，但在我们的辐照条件下，它们对中子产生过程的贡献可以忽略不计。

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引用次数: 0

Adaptive Kriging model based on short-term memory enhancement for optimization of sealing performance in heat transfer tube plugs 基于短时记忆增强的自适应Kriging模型优化换热管塞密封性能

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-13 DOI: 10.1016/j.anucene.2025.112062

Zhiyong Hu , Chunming Fu , Fuchun He , Dewen Tang , Shaoyong Huo , Kaifeng Huang , Chao Jiang

In this paper, an adaptive Kriging surrogate model enhanced by a Historical-augmented Lipschitz Sampling Process (HLIP) is proposed for the optimization of steam generator heat transfer tube plugging. The HLIP method extends the Lipschitz Sampling (LIP) framework by introducing two key improvements. One key feature is a short-term memory mechanism that reduces redundant sampling. Another is the strategic use of Kriging process variance to guide global search. By combining these two mechanisms, the HLIP method forms an integrated strategy that effectively overcomes the sample clustering issue inherent in LIP and significantly improves sampling efficiency. Validation using benchmark functions, ranging from 1D to 10D, demonstrates that HLIP achieves markedly lower Normalized Root Mean Square Error (NRMSE) than established methods, such as LIP, EI, and TEAD. Furthermore, the adaptive Kriging model is successfully applied to optimize the plug expansion performance of a specific SG heat transfer tube, significantly enhancing plug joint quality while ensuring sealing integrity.

本文提出了一种基于历史增强Lipschitz采样过程（HLIP）的自适应Kriging代理模型，用于蒸汽发生器换热管堵塞优化。HLIP方法通过引入两个关键改进扩展了Lipschitz采样（LIP）框架。一个关键特性是减少冗余采样的短期记忆机制。另一个是战略性地使用克里格过程方差来指导全局搜索。将这两种机制结合起来，HLIP方法形成了一个综合策略，有效地克服了LIP固有的样本聚类问题，显著提高了采样效率。使用基准函数（从1D到10D）进行验证，表明HLIP比现有方法（如LIP、EI和TEAD）实现了明显更低的标准化均方根误差（NRMSE）。此外，成功应用自适应Kriging模型对某型SG换热管的塞头膨胀性能进行了优化，在保证密封完整性的同时显著提高了塞头质量。

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引用次数: 0

Fabrication PVA/cellulose/tungsten composite aprons for X-ray shielding: Structural, attenuation, and mechanical evaluation x射线屏蔽用聚乙烯醇/纤维素/钨复合胶圈的制造：结构、衰减和力学评价

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-13 DOI: 10.1016/j.anucene.2025.112063

Dahlang Tahir, Ardiansyah Ardiansyah, Rifqah Nurul Ihsani, Muhammad Azlan, Faradiba Tsani Arief, Heryanto Haryanto, Bualkar Abdullah, Paulus Lobo Gareso

The increasing demand for lead-free radiation protection materials has driven the exploration of environmentally friendly alternatives with effective X-ray shielding performance. This study develops cellulose–tungsten hybrid composites by varying tungsten mass fractions to optimize attenuation efficiency and mechanical properties. The composites were fabricated using a microwave-assisted drying method and characterized through FTIR, XRD, mechanical testing, and X-ray attenuation analysis in the 60–80 keV range. FTIR confirmed cellulose’s functional groups alongside the incorporation of tungsten, while XRD revealed increased crystallinity and crystal size with higher tungsten content. The CHO/W-1.00 composition achieved the highest µ (0.61 cm⁻¹), µ_m (2.46 cm²

∙

g⁻¹), and the lowest HVL and TVL, indicating superior shielding capability. Mechanically, CHO/W-0.75 exhibited the best balance of tensile strength (1.96 MPa) and elongation (10.26 %), highlighting the trade-off between stiffness and flexibility. These findings demonstrate that cellulose–tungsten composites offer a lightweight, non-toxic, and sustainable alternative as aprons with promising applications in industrial radiation protection.

对无铅辐射防护材料日益增长的需求推动了对具有有效x射线屏蔽性能的环保替代品的探索。本研究通过改变钨的质量分数来开发纤维素-钨杂化复合材料，以优化衰减效率和机械性能。采用微波辅助干燥法制备了复合材料，并通过FTIR、XRD、力学测试和60-80 keV范围内的x射线衰减分析对其进行了表征。FTIR证实了纤维素的官能团随钨的加入而增加，而XRD显示纤维素的结晶度和晶体尺寸随钨含量的增加而增加。CHO/W-1.00组合物达到最高µ（0.61 cm−1），µm (2.46 cm2∙g−1)，最低HVL和TVL，表明具有较好的屏蔽能力。力学上，CHO/W-0.75表现出抗拉强度（1.96 MPa）和伸长率（10.26%）的最佳平衡，突出了刚度和柔韧性之间的权衡。这些发现表明，纤维素-钨复合材料提供了一种轻质、无毒、可持续的围裙替代品，在工业辐射防护中具有广阔的应用前景。

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引用次数: 0

Thermal-hydraulic analysis of helical coil once-through steam generator for lead-based cooled fast reactor using a program based on two-fluid model 采用基于双流体模型的程序对铅基冷却快堆螺旋盘管式蒸汽发生器进行热水力分析

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-13 DOI: 10.1016/j.anucene.2025.112064

Zhipeng Liu , Pengrui Qiao , Xueyou Ding , Qinglong Wen

The Helical Coil Once-Through Steam Generator (HCOTSG) is widely used in the design of Lead-based Fast Reactors (LFR) due to its compact structure and high heat transfer efficiency. Based on the two-fluid model, a Transient Analysis Code of Helical coil Once-through Steam generator applicable to LFR, namely TACHOS/LFR, is developed. The model sensitivity analysis aims to verify and ensure the achievement of the optimal combination of constitutive correlations. Steady-state thermal–hydraulic analysis shows that the relative deviation of the steam outlet temperature from the design value is 0.55 %, while the deviation of the Lead-Bismuth Eutectic (LBE) outlet temperature from the design value is 1.2 %. The centrifugal force of the helical tube causes intermittent wetting of the tube wall by the liquid phase, which delays the occurrence of Critical Heat Flux (CHF). As the reactor power level decreases, the length of the saturated boiling section increases, and the coolant undergoes intense boiling phase change, which may lead to inter-tube vibration. The maximum heat flux density reaches 724.13 kW/m², which may induce thermal stress fatigue in the heat transfer tube. Transient calculations indicate that the temperature change at the primary side outlet always lags behind that at the secondary side. The steam pressure drop change is small, posing a limited safety threat to the system. The study may provide a reference for the design and safety analysis of HCOTSG in LFR.

螺旋盘管式蒸汽发生器（HCOTSG）因其结构紧凑、传热效率高而广泛应用于铅基快堆的设计中。基于双流体模型，开发了适用于LFR的螺旋线圈式直流蒸汽发生器暂态分析程序TACHOS/LFR。模型敏感性分析的目的是验证和保证本构相关的最优组合的实现。稳态热水力分析表明，蒸汽出口温度与设计值的相对偏差为0.55%，铅铋共晶出口温度与设计值的相对偏差为1.2%。螺旋管的离心力引起了液相对管壁的间歇性润湿，从而延缓了临界热流密度的发生。随着反应堆功率水平的降低，饱和沸腾段长度增加，冷却剂发生剧烈的沸腾相变，可能导致管间振动。最大热流密度达到724.13 kW/m2，可能导致换热管产生热应力疲劳。瞬态计算表明，一次侧出口的温度变化总是滞后于二次侧出口的温度变化。蒸汽压降变化小，对系统的安全威胁有限。该研究可为LFR中HCOTSG的设计和安全性分析提供参考。

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引用次数: 0

The progress of multiphysics field coupling research on petal-shaped fuel rod 花瓣形燃料棒多物理场耦合研究进展

IF 2.3 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Annals of Nuclear Energy

Pub Date : 2025-12-12 DOI: 10.1016/j.anucene.2025.112011

Binxian He , Shusong Qin , Zequan Huang , Jinyi Cao , Wenchao Zhang , Jianchuang Sun , Jincheng Wang , Lipeng Du , Qian Li , Weihua Cai

As an important pillar for building a clean, low-carbon and high-efficiency energy system, nuclear energy will surely shoulder the key mission of guaranteeing energy stability and security under the trend of declining proportion of traditional energy sources and continuously rising proportion of renewable energy sources. Petal-shaped fuel rod has outstanding performance advantages over conventional fuel rods due to structural and material innovations and has great potential for development. So far, due to the complexity and high cost of experimental studies, the research on petal-shaped fuel rod has mainly focused on the simulation of in-reactor neutron physical, thermo-hydraulic and irradiated thermo-mechanical properties. This paper summarizes the development of petal-shaped fuel rod and the history of its innovative U-50Zr fuel, comprehensively reviews its research progress in neutron physics, thermal hydraulics, and irradiated thermo-mechanical behavior. This paper will help to comprehensively grasp the current research dynamics of petal-shaped fuel rod and provide ideas and directions for further research.

核能作为构建清洁、低碳、高效能源体系的重要支柱，在传统能源比重不断下降、可再生能源比重不断上升的趋势下，必将肩负起保障能源稳定与安全的重要使命。花瓣型燃料棒由于结构和材料的创新，比传统燃料棒具有突出的性能优势，具有很大的发展潜力。到目前为止，由于实验研究的复杂性和高成本，对花瓣状燃料棒的研究主要集中在堆内中子物理、热水力和辐照热力学性能的模拟上。本文综述了花瓣形燃料棒的发展及其创新的U-50Zr燃料的历史，综合评述了花瓣形燃料棒在中子物理、热工力学和辐照热力学行为方面的研究进展。本文将有助于全面掌握花瓣型燃料棒的研究动态，为进一步研究提供思路和方向。

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引用次数: 0