Annals of Nuclear Energy最新文献_第8页

Influence of ferric oxide (Fe2O3) content on the mechanical strength and radiation attenuation capacity of concrete 三氧化铁（Fe2O3）含量对混凝土机械强度和辐射衰减能力的影响

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

Annals of Nuclear Energy

Pub Date : 2026-01-10 DOI: 10.1016/j.anucene.2025.112105

Mohamed Hasabelnaby , Mohammad Marashdeh , K.A. Mahmoud , Reham M. Abd El Rahman , Hanan Akhdar , Ghada Salaheldin , Mohammad Y. Hanfi

Radiation shielding materials are crucial for nuclear, industrial, and medical situations where shielding against ionizing radiation is a concern. In this study, the effect of ferric oxide (Fe₂O₃) incorporation on the mineralogical, physical, mechanical, and gamma-ray shielding properties of Portland cement-based concretes is studied. Concrete samples were made by replacing fine aggregate with Fe₂O₃ (0–40 wt%) in various amounts and were examined for mineralogy using XRD, elemental composition using XRF, density, porosity, and water absorption, compressive strength, elastic modulus, and compression through gamma-ray attenuation. The results showed that on average Fe₂O₃ incorporation led to higher concrete density (2.51–2.69 g/cm³), lower porosity of concrete, and a more than 38 % and 36 % reduction in water absorption, and improved gamma-shielding performance, with average increases in linear attenuation coefficient (LAC) of ∼7 % at energies of 0.511 and 0.662 MeV. Nevertheless, average compressive strength declined from 9.75 MPa (control) to 3.75 MPa (40 wt% Fe₂O₃) and the elastic modulus from 15.6 GPa to 9.7 GPa which are not strong load bearing results. Regression analysis produced predictive models (R² > 0.95) relating Fe₂O₃ amount to density, porosity, and strength to allow for performance estimation for design. These results confirmed that Fe₂O₃ concretes, while not viable for structural load bearing, would still serve as effective non-structural shielding materials for medical and nuclear applications. Based on the data obtained from the Fe₂O₃ study, the upper limit of enhancement through Fe₂O₃ was assigned to CON30 to CON40, though these levels represent some of the highest attenuation values detected in relation to mechanical degradation. Therefore, the Fe₂O₃ doping group concretes do not uphold their suitability for load bearing applications but do give considerable merit for use as non-load bearing radiation shielding materials across medical, research, and nuclear facilities.

辐射屏蔽材料对于需要屏蔽电离辐射的核、工业和医疗环境至关重要。在本研究中，研究了氧化铁（Fe2O3）掺入对硅酸盐水泥基混凝土的矿物学、物理、机械和伽马射线屏蔽性能的影响。用不同量的Fe2O3（0-40 wt%）代替细骨料制成混凝土样品，用XRD检测矿物学，用XRF检测元素组成，通过伽马射线衰减检测密度、孔隙率、吸水率、抗压强度、弹性模量和压缩率。结果表明，Fe2O3的掺入提高了混凝土的密度（2.51 ~ 2.69 g/cm3），降低了混凝土的孔隙率，降低了38 %和36 %的吸水率，并改善了γ屏蔽性能，在能量为0.511和0.662 MeV时，线性衰减系数（LAC）平均增加了7 %。然而，平均抗压强度从9.75 MPa（对照）下降到3.75 MPa(40 wt% Fe2O3)，弹性模量从15.6 GPa下降到9.7 GPa，不是强承载结果。回归分析产生了预测模型（R2 >； 0.95），将Fe2O3的数量与密度、孔隙率和强度联系起来，以便对设计进行性能估计。这些结果证实，Fe2O3混凝土虽然不能用于结构承重，但仍然可以作为有效的非结构屏蔽材料用于医疗和核应用。根据从Fe2O3研究中获得的数据，通过Fe2O3增强的上限被指定为CON30到CON40，尽管这些水平代表了与机械降解相关的一些最高衰减值。因此，Fe2O3掺杂基团混凝土不能维持其在承载应用中的适用性，但在医疗、研究和核设施中作为非承载辐射屏蔽材料的使用确实具有相当大的优点。

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引用次数: 0

Effect of molten salt redox states on the chemical behavior of Tellurium: A machine learning molecular dynamics study 熔融盐氧化还原态对碲化学行为的影响：一种机器学习分子动力学研究

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

Annals of Nuclear Energy

Pub Date : 2026-01-09 DOI: 10.1016/j.anucene.2026.112122

Jingxiang Cao , Guifeng Zhu , Huiqin Yin , Linbing Jiang , Xinmei Yang , Jie Qiu , Wenguan Liu

Tellurium (Te) is the primary cause of intergranular embrittlement in structural materials of molten salt reactors (MSR). This study investigates the chemical behavior of Te in FLiBe molten salt under different redox states, and finds that the redox states have a substantial impact on the chemical behavior of Te. Under reducing redox conditions of the molten salt, Te can stably exist as an anion and preferentially forms bonds with positively charged Th or U atoms. In neutral or mildly oxidizing environments, Te atoms are more likely to aggregate and form Te–Te bonds, which facilitates nucleation in the molten salt and promotes their adsorption onto the alloy surface. Under strongly oxidizing conditions, Te tends to exist in a cationic form and may be present as tellurium fluoride gas. This study reveals the possibility of inhibiting Te-induced intergranular embrittlement in MSRs by adjusting the redox state of the molten salt.

碲是熔盐堆结构材料晶间脆化的主要原因。本研究考察了不同氧化还原状态下Te在FLiBe熔盐中的化学行为，发现氧化还原状态对Te的化学行为有很大的影响。在熔盐还原氧化还原条件下，Te可以稳定地以阴离子形式存在，并优先与带正电的Th或U原子形成键。在中性或轻度氧化环境中，Te原子更容易聚集形成Te - Te键，这有利于熔盐中的成核，并促进其吸附在合金表面。在强氧化条件下，Te倾向于以阳离子形式存在，并可能以氟化碲气体的形式存在。本研究揭示了通过调节熔盐的氧化还原状态来抑制te诱导的MSRs晶间脆化的可能性。

引用次数: 0

Transient multiphysics simulations with pin power reconstruction in the Griffin reactor physics code Griffin反应堆物理代码中pin功率重建的瞬态多物理场模拟

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

Annals of Nuclear Energy

Pub Date : 2026-01-09 DOI: 10.1016/j.anucene.2025.112111

Shikhar Kumar , Changho Lee , Vincent Laboure , Yeon Sang Jung , Stefano Terlizzi , Yaqi Wang , Javier Ortensi

This work introduces the pin power reconstruction capability available in the Griffin reactor physics code. This capability is implemented in an unstructured mesh framework, and the methods introduced are applied to the 2D SIMBA reactor core, which has assemblies and pins arranged in a hexagonal lattice. Since this reactor has a non-Cartesian geometry and also operates in the thermal spectrum, a general approach to pin power reconstruction is adopted, where SPH-based equivalence is leveraged to preserve assembly-wise reaction rates, while computing full-core form functions to preserve pin-wise fission production rates within the fuel pins of the reactor core. In a 2D microreactor benchmark problem, this pin power reconstruction approach was shown to reproduce pin powers compared to the Serpent2 Monte Carlo code for fixed temperature conditions and control drum rotation angles, yielding a core-wide RMS error level of 0.6% and a maximum absolute pin error of 2.3%. In addition, a tabulated library of multigroup cross sections, SPH factors, and form functions was generated to demonstrate the applicability of pin power reconstruction to a thermal feedback problem. Finally, a control drum transient was successfully simulated, showcasing the application of pin power reconstruction in a transient multiphysics feedback problem.

本文介绍了Griffin反应堆物理代码中可用的引脚功率重构能力。该功能在非结构化网格框架中实现，并将所介绍的方法应用于二维SIMBA反应堆堆芯，该堆芯的组件和引脚排列在六边形晶格中。由于该反应堆具有非笛卡尔几何形状，并且也在热谱中运行，因此采用了一种通用的针功率重建方法，其中利用基于sph的等效来保持装配方向的反应速率，同时计算全芯形式函数以保持反应堆堆芯燃料针内针方向的裂变产生速率。在一个二维微反应器基准问题中，与Serpent2蒙特卡罗代码相比，该引脚功率重建方法在固定温度条件和控制转鼓旋转角度下再现了引脚功率，产生了0.6%的核心范围的均方根误差水平，最大绝对引脚误差为2.3%。此外，还生成了多组截面、SPH因子和形式函数的表格库，以证明引脚功率重构对热反馈问题的适用性。最后，对一个暂态控制鼓进行了成功的仿真，展示了引脚功率重构在瞬态多物理场反馈问题中的应用。

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引用次数: 0

Transient thermal diffusion analysis and failure prediction in heat-pipe-cooled reactors 热管冷却堆瞬态热扩散分析及失效预测

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

Annals of Nuclear Energy

Pub Date : 2026-01-09 DOI: 10.1016/j.anucene.2026.112120

Jiaqing Zhang, Xiao Zhang, Wenxiao Chu, Qiuwang Wang

Heat pipe cooled reactors (HPCRs) rely on thermal conduction to transfer heat from the reactor core, where the thermal reliability becomes a critical concern. Studies on the temperature dynamic response due to random heat pipe (HP) failures and the prediction of specific heat pipes through temperature distribution analysis are the key challenges. This study investigates the spatial thermal diffusion mechanism and temperature dynamic response characteristics in the HPCR core during random HP failure processes using experimental and numerical methods. Moreover, the Random Forest algorithm method is introduced to predict HP failure locations. Results indicate that boundary HP failure (HP-A) exhibits a broader critical failure diffusion radius of 65.1 mm and diffusion angle of 190°, while central HP failure (HP-D) causes minimal disturbance and results in more uniform temperature gradient distributions. Correspondingly, the dynamic response time constant and response delay time are employed to quantitatively feature the temperature field evolution during HP failure. For HP-A, the time constant and response delay time are 5040 s and 170 s, respectively, compared to 10,950 and 550 s for HP-D. Additionally, two patterns with single and dual HP failures and four HP failure orientations are predicted by the Random Forest algorithm method. Results demonstrate the prediction accuracy of 97.1 %, with a failure time prediction error ranging from −0.7 % to 1.6 %.

热管冷却堆依靠热传导从堆芯传递热量，其热可靠性成为一个关键问题。研究随机热管（HP）失效时的温度动态响应以及通过温度分布分析预测比热管是关键挑战。采用实验和数值方法研究了高压随机失效过程中HPCR芯内的空间热扩散机制和温度动态响应特征。在此基础上，引入随机森林算法对HP故障位置进行预测。结果表明，边界HP失效（HP- a）的临界失效扩散半径为65.1 mm，扩散角为190°，而中心HP失效（HP- d）的扰动最小，温度梯度分布更均匀。相应的，采用动态响应时间常数和响应延迟时间来定量表征高温高压失效时温度场的演变。HP-A的时间常数和响应延迟时间分别为5040 s和170 s， HP-D的时间常数和响应延迟时间分别为10950 s和550 s。此外，利用随机森林算法预测了单HP故障和双HP故障的两种模式以及四种HP故障方向。结果表明，预测精度为97.1%，故障时间预测误差为- 0.7% ~ 1.6%。

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引用次数: 0

Evaluation of ENDF/B-VIII.0 nuclear data for criticality calculations using machine learning and the SHAP interpretability method ENDF/B-VIII的评价。使用机器学习和SHAP可解释性方法进行临界计算的0核数据

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

Annals of Nuclear Energy

Pub Date : 2026-01-08 DOI: 10.1016/j.anucene.2026.112126

M. Hadouachi , K. Laazouzi , O. Belhaj , H. El Yaakoubi , A. Arectout , Abdelhamid Nouayti , H. Boukhal , E. Chakir , T. El Bardouni

In nuclear reactor criticality and stability studies, nuclear data uncertainties can significantly influence integral parameters such as the effective multiplication factor and neutron flux, which are directly linked to reactor safety margins and operational performance. It is therefore essential to quantify the impact of nuclear data uncertainties on reactor calculations. In this work, machine learning techniques were applied to identify the nuclear data that have the greatest impact on criticality calculations. For this purpose, sensitivity profiles, combined with other benchmark characteristics, were used as input features for various machine learning algorithms to predict the bias

Δ k_{eff}

. In order to interpret the model’s predictions, a SHAP (SHapley Additive exPlanations) analysis was applied to determine which reactions had the greatest influence on (k

_{eff}

) bias. The results highlight that nuclear data for nuclides such as ²³⁹Pu, ²³⁵U, ²³³U, ²³⁸U, ¹²C, and ¹H are the most important parameters related to a high

Δ k_{eff}

.

在核反应堆临界和稳定性研究中，核数据的不确定性会显著影响有效乘法系数和中子通量等积分参数，这些参数直接关系到反应堆的安全裕度和运行性能。因此，必须量化核数据不确定性对反应堆计算的影响。在这项工作中，机器学习技术被应用于识别对临界计算影响最大的核数据。为此，灵敏度曲线结合其他基准特征作为各种机器学习算法的输入特征来预测偏差Δkeff。为了解释模型的预测，应用SHAP （SHapley Additive explanation）分析来确定哪些反应对（keff）偏差的影响最大。结果表明，239Pu、235U、233U、238U、12C和1H等核素的核数据是与高Δkeff相关的最重要参数。

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引用次数: 0

Preliminary design and front-end fuel cycle assessment of 200 MWt marine molten chloride fast reactor 200mwt船用氯熔快堆初步设计及前端燃料循环评价

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

Annals of Nuclear Energy

Pub Date : 2026-01-08 DOI: 10.1016/j.anucene.2026.112124

Andika Putra Dwijayanto , Kenji Nishihara , Tomohiro Okamura , Masahiko Nakase

The molten Chloride Fast Reactor (MCFR) emerges as one of the advanced nuclear reactor designs for use in a nuclear marine propulsion. This paper delineates the preliminary design of a 200 MWt Marine MCFR (MMCFR) intended as a propulsion for zero-carbon large container ship, focusing on the neutronic analysis and fuel cycle assessment. The MMCFR employs eutectic 66NaCl-34UCl₃ as the fuel with 19.55 wt% enriched uranium as the initial fuel contained in a BeO-reflected core, operated as a long-lived core and batchwise refuelling to simplify reactivity control and refuelling mechanism in a constrained space. As the MMCFR is designed with a compact core and large initial reactivity, the innovative Partial Fuel Change scheme is proposed to optimise fuel consumption and reduce the strain in the front-end fuel cycle, with Constant Mol or Constant Replacement scenario. Initial reactivity is suppressed using burnable absorber (BA) rods and control drums are used to control the reactivity and core shutdown. Neutronic and depletion calculations for the MMCFR design were performed using Serpent-2 code and ENDF/B-VII.0 library. The optimum front-end fuel cycle was obtained to be Constant Replacement scenario with lowest uranium consumption. Meanwhile, excess reactivity can be maintained below 5% throughout operational time by using BA and control drum, whilst temperature coefficient of reactivity (TCR) is sufficiently negative, ensuring the MMCFR fulfils the safety criteria.

熔融氯化物快堆（MCFR）是用于核动力船舶推进的先进核反应堆设计之一。本文描述了用于零碳大型集装箱船推进的200 MMCFR （MMCFR）的初步设计，重点介绍了中子分析和燃料循环评估。MMCFR采用共晶66NaCl-34UCl3作为燃料，初始燃料为19.55 wt%的浓缩铀，包含在beo反射堆芯中，作为长寿命堆芯运行，批量换料以简化反应性控制和在有限空间内换料机制。由于MMCFR设计具有紧凑的核心和大的初始反应性，因此提出了创新的部分燃料更换方案，以优化燃料消耗并减少前端燃料循环中的应变，采用恒定摩尔或恒定更换方案。使用可燃吸收棒抑制初始反应性，使用控制鼓来控制反应性和堆芯停机。使用Serpent-2代码和ENDF/B-VII进行了MMCFR设计的中子和耗尽计算。0图书馆。得到了最优的前端燃料循环是铀消耗量最低的持续替换方案。同时，通过使用BA和控制鼓，可以在整个运行时间内将过量反应性保持在5%以下，同时反应性温度系数（TCR）足够负，确保MMCFR符合安全标准。

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引用次数: 0

Design and performance analysis of high-power lead-bismuth cooled micro reactor 大功率铅铋冷却微堆设计与性能分析

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

Annals of Nuclear Energy

Pub Date : 2026-01-08 DOI: 10.1016/j.anucene.2026.112115

Yiming Xiong, Ren Li, Jilin Sun, Yuandong Zhang, Genglei Xia, Minjun Peng

With the wide range of sea applications of reactors, it is necessary to develop a high-power LBE reactor with a small volume and high-power output. A process of core design is proposed in this paper for conducting rapid iterations. With this process and a series of design criteria and guidelines, the concept of the High-Power Lead-Bismuth Cooled Micro Reactor (HLCMR) is designed. This design adopts low-enriched fuel and two sets of independent shut-down control systems, which are arranged in a triangular pattern to achieve higher power density and lower power peak. After iterative design, the design power is 12 MW, and the lifetime is at least 5 years with the core power density of 76.72 MW/m³. The power distribution shows that the highest power peak is 1.444 in the operating state. The temperature field and reactivity are also calculated to evaluate the safety and reliability of this design. The results show that all parameters of the HLCMR meet the thermal–hydraulic and control design requirements.

随着反应器在海上的广泛应用，研制体积小、输出功率大的大功率LBE反应器势在必行。提出了一种快速迭代的核心设计流程。根据这一过程和一系列设计准则和指导方针，设计了大功率铅铋冷却微堆（HLCMR）的概念。本设计采用低浓度燃料和两套独立的停机控制系统，以三角形的形式布置，以达到更高的功率密度和更低的功率峰值。经迭代设计，设计功率为12 MW，寿命至少5年，堆芯功率密度为76.72 MW/m3。由功率分布可知，在运行状态下，最高功率峰值为1.444。计算了温度场和反应性，评价了设计的安全性和可靠性。结果表明，HLCMR的各项参数均满足热液和控制设计要求。

引用次数: 0

Analysis of the blowdown of supercritical carbon dioxide from simple vessel 简单容器中超临界二氧化碳排放的分析

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

Annals of Nuclear Energy

Pub Date : 2026-01-07 DOI: 10.1016/j.anucene.2025.112110

Fengyuan Tian , Minyun Liu , Yanping Huang , Ruohan Zheng , Yangle Wang , Houjun Gong , Yu Tang , Tianzeng Liu , Jinghan Hu , Haohan Yuan , Yuan Zhou

The depressurization accident is one of the key problems during the transport and utilization of S-CO₂. In this paper, an analysis model was developed to analyze the blowdown of S-CO₂ from a simple vessel. The critical mass flow rate agreed well with experiment results, and the maximum error is less than 10%. As for the depressurization model, the homogeneous model could predict pressure and temperature, and the phase separation model and bubble rise model could predict the mass flow rate. Depressurization under different initial parameters, back pressures, breaks were analyzed. Results indicated that when the initial temperature exceeds a certain temperature, the pressure and mass flow rate would change smoothly. When the initial temperature is lower, the depressurization process could be divided into rapid depressurization, flash vaporization, and slow depressurization. The model developed may reflect the characteristics of depressurization and provide a reference for depressurization accidents of the S-CO₂ system.

减压事故是S-CO2输送和利用过程中的关键问题之一。本文建立了一个简单容器S-CO2排放分析模型。临界质量流量与实验结果吻合较好，最大误差小于10%。减压模型中均相模型可以预测压力和温度，相分离模型和气泡上升模型可以预测质量流量。分析了不同初始参数、背压、破断条件下的减压效果。结果表明，当初始温度超过一定温度时，压力和质量流量变化平缓。当初始温度较低时，减压过程可分为快速减压、闪蒸和慢速减压。所建立的模型可以反映出S-CO2系统的降压特性，为S-CO2系统的降压事故提供参考。

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引用次数: 0

Calculation and analysis of self-shielded cross sections for the high-fidelity neutronics calculation of fast reactors 快堆高保真中子计算中自屏蔽截面的计算与分析

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

Annals of Nuclear Energy

Pub Date : 2026-01-07 DOI: 10.1016/j.anucene.2025.112103

Xiang Li, Zhouyu Liu, Wenjie Chen, Liangzhi Cao, Hongchun Wu

A new high-fidelity neutronics analysis tool for fast reactor calculations is built by coupling the lattice code TULIP, the SN code HYDRA and the high-fidelity neutron transport code NECP-X. TULIP is employed to generate multigroup effective cross sections for assemblies based on the 0-D or 1-D cylindrical model. HYDRA utilizes a reactor core-reflector model to perform group condensation for reflector. The cross sections are passed to NECP-X to conduct a whole-core simulation with explicit geometry description. Verifications were conducted based on the Superphénix 2-D, MET-1000 2-D and JOYO MK-I 3-D core problems. The reference solutions were obtained through Monte Carlo calculations with NECP-MCX. The difference of k_eff was at most 298 pcm for 2-D problems and did not exceed 55 pcm for 3-D problem. The root mean square error of assembly power was a maximum of only 1 %. These results prove the capability of this method in fast reactor calculations.

将晶格码TULIP、SN码HYDRA和高保真中子输运码NECP-X耦合在一起，建立了一个用于快堆计算的高保真中子分析工具。基于0-D或1-D圆柱模型，采用TULIP生成组件的多组有效截面。HYDRA利用反应堆堆芯-反射器模型对反射器进行群凝结。将截面传递给NECP-X进行具有显式几何描述的全核模拟。基于superphacimnix二维、MET-1000二维和JOYO MK-I三维岩心问题进行了验证。利用NECP-MCX进行蒙特卡罗计算得到参考解。二维问题的keff差异最多为298 pcm，三维问题的keff差异不超过55 pcm。装配功率的均方根误差最大仅为1%。这些结果证明了该方法在快堆计算中的能力。

引用次数: 0

Free convective condensation in the presence of noncondensable gases − A review with heat transfer studies 不可冷凝气体存在下的自由对流冷凝-传热研究综述

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

Annals of Nuclear Energy

Pub Date : 2026-01-07 DOI: 10.1016/j.anucene.2025.112099

NK Maheshwari, Divij Kishal

Steam condensation plays a key role in removing heat from the containment in case of a postulated accident in water-cooled nuclear reactors. Steam released into the containment mixes with air present in that environment and condenses in the presence of noncondensable gases (air, helium, etc.) on the containment wall and other structures present in the containment. Advanced reactors design adopts passive containment cooling systems for long term containment cooling during the design basis and severe accident conditions. In this article, the research performed on free convective condensation in the presence of noncondensable gases on the tube outer surface has been reviewed. In the first part of the article, experimental studies have been covered. It is revealed that both the thermal hydraulic and geometrical parameters affect the condensation heat transfer in the presence of noncondensable gases. In the second part, various correlations developed by researchers are discussed accounting for thermal hydraulic, geometric parameters and nondimensional numbers; an assessment of these correlations is performed. In the third part, the theoretical model developed, results obtained and CFD studies performed by previous authors have been discussed. The effects of various parameters are discussed on the basis of experimental work and theoretical model developed. Finally, based on the review and studies performed, a summary is provided.

在水冷核反应堆发生假想事故的情况下，蒸汽冷凝在从安全壳中除去热量方面起着关键作用。释放到安全壳中的蒸汽与该环境中的空气混合，并在安全壳壁和安全壳中存在的其他结构上存在不可冷凝气体（空气，氦气等）的情况下凝结。先进反应堆设计采用被动式安全壳冷却系统，在设计基础和严重事故条件下进行长期安全壳冷却。本文综述了管内外表面存在不凝性气体时自由对流冷凝的研究进展。在文章的第一部分，已经涵盖了实验研究。结果表明，在不凝性气体存在的情况下，热工参数和几何参数对冷凝换热都有影响。在第二部分，讨论了研究人员开发的各种关联，考虑了热水力、几何参数和无因次数；对这些相关性进行评估。在第三部分，理论模型的建立，得到的结果和CFD研究进行了讨论。在实验工作和建立理论模型的基础上，讨论了各种参数的影响。最后，在回顾和研究的基础上，对本文进行了总结。

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引用次数: 0