Nuclear Engineering and Design最新文献_第8页

The verification and validation of the coupled neutronics thermal-hydraulics code, MTRDYN, for steady-state condition of RSG-GAS reactor RSG-GAS反应器稳态条件下中子热工-水力学耦合程序MTRDYN的验证与验证

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

Nuclear Engineering and Design

Pub Date : 2026-01-08 DOI: 10.1016/j.nucengdes.2025.114746

Surian Pinem , Farisy Yogatama Sulistyo , Peng Hong Liem , Sukmanto Dibyo , Wahid Luthfi

The in-house coupled thermal-hydraulic and neutronic code, MTRDYN, has been developed with a three-dimensional capability to solve few-group neutron diffusion equations and thermal-hydraulic parameters for plate type fueled research reactor. The multi-group neutron diffusion equations are addressed through neutron flux factorization within an adiabatic kinetic equation. Heat conduction in the fuel element was computed using the finite difference method, with the heat transfer restricted to the radial direction approximation. This study aims to evaluate the accuracy of the MTRDYN in calculating the behavior of RSG-GAS reactor during steady-state operation. The calculated core parameters include excess reactivity, power peaking factor (PPF), fuel cladding temperature, and coolant temperatures. The coolant and cladding temperature obtained from MTRDYN were validated against measured data from instrumented fuel elements (IFE) located at various positions within the core. The calculated excess reactivity for the first and sixth cores differed from experimental results by −160 pcm and 20.0 pcm, respectively. The total control rod reactivity showed a maximum error of 3.9 % compared to experimental results. No significant differences in kinetic parameters were found compared to the RSG-GAS safety analysis report (SAR). The calculated fuel cladding temperatures showed a maximum deviation of 5.78 %. Based on these calculations, the MTRDYN code demonstrates sufficient accuracy in determining the steady-state neutronic and thermal-hydraulic parameters of the RSG-GAS reactor.

开发了内部热工-中子耦合程序MTRDYN，该程序具有求解板式燃料研究堆的少群中子扩散方程和热工参数的三维能力。在绝热动力学方程中，通过中子通量分解来求解多群中子扩散方程。采用有限差分法计算了燃料元件的热传导，并将传热限制在径向近似。本研究旨在评估MTRDYN在计算RSG-GAS反应器稳态运行行为时的准确性。计算的堆芯参数包括过量反应性、功率峰值因子（PPF）、燃料包层温度和冷却剂温度。从MTRDYN获得的冷却剂和包层温度与位于堆芯内不同位置的仪表燃料元件（IFE）的测量数据进行了验证。计算得到的第一堆和第六堆的超反应性与实验结果分别相差- 160 pcm和20.0 pcm。总控制棒反应性与实验结果的最大误差为3.9%。与RSG-GAS安全分析报告（SAR）相比，动力学参数无显著差异。计算得到的燃料包壳温度最大偏差为5.78%。基于这些计算，MTRDYN程序在确定RSG-GAS反应堆的稳态中子和热工参数方面具有足够的准确性。

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引用次数: 0

Evaluating primary circuit pressure reduction strategy in VVER-1000/V446 nuclear reactor during SBO severe accident with MELCOR 1.8.6 基于MELCOR 1.8.6的SBO严重事故中VVER-1000/V446核反应堆一次回路减压策略评价

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

Nuclear Engineering and Design

Pub Date : 2026-01-07 DOI: 10.1016/j.nucengdes.2025.114744

Abolhasan Nabavi , Ahmad Pirouzmand , Kamal Hadad

In this study, the MELCOR 1.8.6 code is used to simulate the progression of an SBO severe accident over a 24-h period in a VVER-1000/V446 nuclear reactor. Multiple scenarios are analyzed to assess the effectiveness of the primary circuit pressure reduction strategy, which is triggered by operator action when the core exit temperature reaches a predefined setpoint. Sixteen temperature setpoints are considered: 350, 375, 400, 425, 450, 475, 500, 525, 550, 575, 600, 615, 635, 650, 675, and 700 °C. For each setpoint, four modes of operator action delay are evaluated: one with no delay, and three with delays of 110, 600, and 1397 s. In each mode, key parameters are examined, including fuel failure time, lower head failure time, containment failure time, in-vessel hydrogen generation, total radioactive release, and core degradation progression. The variation in these parameters across different setpoints highlights the importance of selecting an optimal setpoint for each delay mode and time interval. Also, the correct estimation of “the time required to the appropriate countermeasure actions” up to 70 min after the SBO, plays a significant role for selection of the optimal setpoints. If the estimation of this time, is less than 4 h, the optimal temperature setpoint, irrespective of the operator delay, is 350 °C. Conversely, depending upon the operator delay and the estimation of “the time required for the appropriate countermeasure actions”, various temperature setpoints are employed.

在本研究中，采用MELCOR 1.8.6代码模拟了VVER-1000/V446核反应堆的SBO严重事故在24小时内的进展。为了评估主回路减压策略的有效性，研究人员分析了多种场景，当堆芯出口温度达到预定设定值时，操作人员会触发主回路减压策略。考虑了16个温度设定值：350、375、400、425、450、475、500、525、550、575、600、615、635、650、675和700°C。对于每个设定值，评估了操作员动作延迟的四种模式：一种是无延迟，三种延迟分别为110、600和1397秒。在每种模式下，检查关键参数，包括燃料失效时间、下水头失效时间、安全壳失效时间、容器内氢气生成、总放射性释放和堆芯降解进程。这些参数在不同设定值上的变化突出了为每个延迟模式和时间间隔选择最佳设定值的重要性。此外，正确估计SBO后70分钟内“采取适当对策行动所需的时间”，对最佳设定值的选择起着重要作用。如果该时间的估计小于4小时，则无论操作延迟如何，最佳温度设定值为350°C。相反，根据操作员的延迟和“适当对策行动所需的时间”的估计，不同的温度设定值被采用。

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引用次数: 0

Experimental study of bubble growth behavior for helical cruciform fuel under high pressure and mass flux conditions 高压和高质量流量条件下螺旋十字形燃料气泡生长行为的实验研究

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

Nuclear Engineering and Design

Pub Date : 2026-01-07 DOI: 10.1016/j.nucengdes.2026.114749

Zhongkai Mei , Desheng Jin , Jun Chen , Qinglong Wen , Xing Lei , Yong Shen

Helical cruciform fuel (HCF) is an innovative fuel element design with promising potential for the application in small modular reactors and advanced Generation IV reactor systems. The geometric structure of HCF differs largely from conventional fuels, which consequently changes the flow field and alters bubble behavior on the HCF surface. Particularly, bubble growth behavior should be particularly paid attention to because of its underlying influence on heat transfer process. In this study, typical bubble growth behavior of HCF is visualized by the high-speed camera, while the effect of heat flux, mass flux, inlet pressure and subcooling degree on bubble growth is analyzed. The results revealed that in given conditions, the bubble tends to persist on the surface of HCF rod for a considerable duration and its growth behavior differs significantly from that observed in the context of other regular fuels. With high inlet pressure, the shape deformation of growing bubbles on the surface of HCF can be neglected in the turbulent flow field. The influence of turbulent flow field on bubble growing process mainly manifests on the cooling of the heating rods, rather than disrupting the force balance of the growing bubbles. Incorporating the pressure ratio and an adjusted growth index, the modified bubble growth rate model is developed and validated against experimental data.

螺旋十字形燃料（HCF）是一种创新的燃料元件设计，在小型模块化反应堆和先进的第四代反应堆系统中具有广阔的应用前景。HCF的几何结构与传统燃料有很大的不同，从而改变了流场，改变了HCF表面的气泡行为。特别是，气泡的生长行为应特别注意，因为它对传热过程的潜在影响。本研究利用高速摄像机对HCF的典型气泡生长行为进行了可视化，分析了热流密度、质量流密度、进口压力和过冷度对气泡生长的影响。结果表明，在一定条件下，气泡会在HCF棒表面持续相当长的时间，其生长行为与在其他常规燃料环境下观察到的明显不同。在高进口压力下，在湍流流场中，HCF表面生长气泡的形状变形可以忽略不计。湍流流场对气泡生长过程的影响主要表现在加热棒的冷却上，而不是破坏气泡生长的力平衡。结合压力比和调整后的生长指数，建立了改进的气泡生长速率模型，并通过实验数据进行了验证。

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引用次数: 0

A fouling prediction method for analyzing deposition distribution in double-side once through steam generator with helical tubes 一种分析螺旋管双面直通蒸汽发生器沉积分布的结垢预测方法

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

Nuclear Engineering and Design

Pub Date : 2026-01-07 DOI: 10.1016/j.nucengdes.2025.114743

Dongdong Wang, Shengwei Fu, Meng Jiao, Jiahang Jiang, Jiacheng Jiang

During the operation of nuclear reactors, corrosion particles originating from various materials progressively accumulate on the surfaces of the heat transfer tubes in the steam generator, thereby inducing fouling. This fouling not only diminishes the heat transfer efficiency of the heat exchanger but may also result in pipe blockage. This study establishes a comprehensive fouling prediction methodology specifically for the secondary side of double-side once through steam generator (OTSG) with helical tubes, enabling the prediction of fouling deposition distribution on heat transfer tube surfaces after long-term operation. A systematic analysis was conducted to evaluate the impact of circumferential non-uniformity on fouling deposition. The findings demonstrate that the unique crescent-shaped flow channel in OTSGs induces significant circumferential non-uniformity, which substantially influences the circumferential distribution of fouling deposition and consequently leads to the formation of helical band-shaped deposition patterns. Although more severe fouling deposition occurs under high-power conditions and on straight tube surfaces, these cases exhibit reduced susceptibility to circumferential non-uniformity effects.

在核反应堆运行过程中，来自各种材料的腐蚀颗粒逐渐积聚在蒸汽发生器的换热管表面，从而产生污垢。这种结垢不仅降低了换热器的换热效率，而且可能导致管道堵塞。本研究针对带螺旋管的双面一过蒸汽发生器（OTSG）二次侧污垢建立了综合预测方法，实现了对长期运行后传热管表面污垢沉积分布的预测。系统分析了周向不均匀性对污垢沉积的影响。研究结果表明，OTSGs中独特的新月形流道引起了显著的周向非均匀性，这极大地影响了污垢沉积的周向分布，从而导致螺旋带状沉积模式的形成。尽管在高功率条件下和直管表面会发生更严重的污垢沉积，但这些情况对周向非均匀性影响的敏感性较低。

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引用次数: 0

Derivation and validation of a turbulent Prandtl number model for LBE in rod bundle channels 棒束通道中LBE湍流普朗特数模型的推导与验证

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

Nuclear Engineering and Design

Pub Date : 2026-01-07 DOI: 10.1016/j.nucengdes.2025.114736

Ruizhi Qi , Hui Liang , Xingguang Zhou , Yueyao Zhang , Xiaoqing Chen , Hongxing Yu , Chuan Lu , Pengrui Qiao , Dalin Zhang

Lead‑bismuth eutectic (LBE), as a candidate coolant for Gen-IV reactor types, shows favorable neutronic performance and thermal safety features. However, the molecular Prandtl number (Pr) of LBE is far less than unity, which suppresses turbulent temperature fluctuations. Consequently, the Reynolds analogy—which assumes similarity between turbulent momentum and heat transport—becomes invalid. Therefore, in Reynolds-averaged Navier–Stokes (RANS) based computational fluid dynamics (CFD) simulations of turbulent heat transfer in LBE, the turbulent heat flux term in the energy equation cannot be closed by assuming a turbulent Prandtl number (Pr_t) ≈ 1. At present, the most widely adopted engineering approach is to apply turbulent Prandtl number models for turbulent heat flux closure. Nevertheless, most existing turbulent Prandtl number models are derived from simplified geometries such as circular pipes and parallel-plate channels, and their applicability to rod bundle channels, which are prevalent in nuclear power systems, is limited. This paper proposes a method for deriving a turbulent Prandtl number model based on the Lyon–Martinelli correlation and the corresponding experimental Nu–Pe correlation. CFD simulations with the derived turbulent Prandtl number model show good agreement with experimental results, demonstrating its potential for practical engineering applications.

铅铋共晶（LBE）作为第四代反应堆类型的候选冷却剂，具有良好的中子性能和热安全特性。而LBE的分子普朗特数（Pr）远小于1，抑制了湍流温度波动。因此，假设湍流动量和热输运相似的雷诺类比就失效了。因此，在基于reynolds -average Navier-Stokes （RANS）的LBE湍流传热计算流体动力学（CFD）模拟中，不能通过假设湍流普朗特数(Prt)≈1来关闭能量方程中的湍流热流通量项。目前，工程上采用最广泛的方法是采用湍流普朗特数模型进行湍流热流闭合。然而，大多数现有的湍流普朗特数模型都是从圆形管道和平行板通道等简化几何形状中推导出来的，它们对核电系统中普遍存在的棒束通道的适用性有限。本文提出了一种基于Lyon-Martinelli相关和相应的实验Nu-Pe相关推导湍流普朗特数模型的方法。利用所建立的湍流普朗特数模型进行的CFD模拟结果与实验结果吻合较好，证明了该模型在实际工程应用中的潜力。

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引用次数: 0

Three-dimensional crack reconstruction from Beam–Particle Model for CFD-based leakage assessment 基于cfd泄漏评估的梁-颗粒模型三维裂缝重建

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

Nuclear Engineering and Design

Pub Date : 2026-01-06 DOI: 10.1016/j.nucengdes.2025.114718

Omar Najjar , Thomas Heitz , Cécile Oliver-Leblond , Jean-Louis Tailhan , Giuseppe Rastiello , Frédéric Ragueneau

Accurate crack characterization in reinforced concrete structures, such as the containment walls of 1300 MWe nuclear power plants, is essential for reliable air leakage estimation. The complexity of crack patterns — including their path, opening, and surface roughness — significantly influences predictions of air leakage. Traditional modeling approaches, such as applying Poiseuille’s law to simplified geometries, depend on indirect parameters like the tortuosity coefficient, which is challenging to predict accurately and has limited applicability, thereby increasing uncertainty. This study introduces a novel numerical post-processing tool based on the Beam–Particle Model (BPM) approach, aimed at enhancing the precision of crack predictions derived from finite element simulations. It employs graph theory to detect micro-crack paths and reconstruct a possible macro-crack geometry, thus improving the representation of crack characteristics. The generated data can be integrated into computational fluid dynamics (CFD) simulations for more precise airflow modeling or used to calibrate simplified methods, such as Poiseuille’s law, with numerically obtained crack descriptors. The fracture geometries produced by the tool in the Brazilian splitting test are validated against optical microscopic observations, demonstrating the tool’s ability to capture crack tortuosity. In contrast, the PSD-based assessment of the Hurst exponent on BPM-generated crack surfaces yields non-physical values, so surface roughness could not be quantified numerically due to BPM computational time constraints. In addition, the current reconstruction algorithm is limited to a single dominant crack path and cannot handle crack branching or macro-crack turnover, although it can detect separated macro-cracks. These results establish a benchmark for future efforts to incorporate detailed crack geometries into CFD-based air leakage estimates, underscoring the tool’s potential to reduce uncertainty in such predictions.

在钢筋混凝土结构中，如1300mwe核电站的安全壳，准确的裂缝表征对于可靠的空气泄漏估计至关重要。裂纹模式的复杂性——包括它们的路径、开口和表面粗糙度——显著地影响着空气泄漏的预测。传统的建模方法，如将泊泽维尔定律应用于简化几何，依赖于弯曲系数等间接参数，难以准确预测，适用性有限，从而增加了不确定性。本文介绍了一种基于束-粒子模型（BPM）方法的新型数值后处理工具，旨在提高基于有限元模拟的裂纹预测精度。它利用图论来检测微裂纹路径并重建可能的宏观裂纹几何，从而改善裂纹特征的表示。生成的数据可以集成到计算流体动力学（CFD）模拟中，以实现更精确的气流建模，或者用于校准简化方法，例如用数值获得的裂缝描述符校准泊泽维尔定律。该工具在巴西劈裂测试中产生的裂缝几何形状通过光学显微镜观察进行了验证，证明了该工具捕捉裂缝弯曲度的能力。相比之下，基于psd的对BPM生成的裂纹表面的Hurst指数的评估产生非物理值，因此由于BPM计算时间的限制，表面粗糙度无法进行数值量化。此外，目前的重构算法虽然可以检测到分离的宏观裂纹，但仅限于单一的主裂纹路径，无法处理裂纹分支或宏观裂纹翻转。这些结果为未来将详细的裂缝几何形状纳入基于cfd的空气泄漏估计奠定了基础，强调了该工具在减少此类预测中的不确定性方面的潜力。

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引用次数: 0

Numerical simulation of a supercritical heat-exchanger using the lattice Boltzmann method 超临界换热器的晶格玻尔兹曼数值模拟

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

Nuclear Engineering and Design

Pub Date : 2026-01-06 DOI: 10.1016/j.nucengdes.2026.114753

Vinicius Akyo Matsuda , Ivan Talão Martins , Julia Akiko Sassa , Luben Cabezas-Gómez , Tiago Augusto Moreira

Recently, supercritical heat-exchangers have recovered the attention of researches to applications in GenIV supercritical water-cooled reactors (SCWR). As supercritical fluids have high heat capacity and low viscosity, they present an efficient heat exchange and reduce the overall energy requirements, resulting in more compact devices. However, when crossing the pseudocritical temperature, as it occurs for the water in both the core and in the primary heat exchanger, these fluids experience high variations in properties, which can affect the heat transfer and must be well understood. In this paper, a modified lattice Boltzmann method (LBM) to deal with supercritical flows near the pseudocritical point is proposed. It is based on modifications of existing models for conjugated heat transfer problems, enhancing the numerical scheme to deal with temperature-dependent properties. The new proposed model was employed to investigate a heat exchanger operating with water as both a hot (supercritical) and cold fluid, mimicking one possible scenario for the primary heat exchanger in a SCWR. The effects of the supercritical–subcritical transition were evaluated by performing a parametric evaluation of the hot fluid mass flow rate. The results evidenced the multi-dimensionality of the supercritical–subcritical transition, with the fluid flow presenting high thermal property variations in both

x

and

y

directions. It elucidates the importance of considering both temperature-dependent properties and more robust numerical methods when dealing with supercritical flows near the pseudocritical point.

近年来，超临界换热器在GenIV型超临界水冷堆（SCWR）中的应用重新受到研究的关注。由于超临界流体具有高热容量和低粘度，因此它们具有高效的热交换和降低总体能量需求，从而使设备更紧凑。然而，当越过假临界温度时，就像水在堆芯和主热交换器中发生的那样，这些流体的性质会发生很大的变化，这可能会影响传热，必须很好地理解。本文提出了一种改进的晶格玻尔兹曼方法（LBM）来处理伪临界点附近的超临界流动。它基于对现有的共轭传热问题模型的修正，增强了处理温度相关特性的数值格式。新提出的模型被用于研究将水作为热（超临界）和冷流体运行的热交换器，模拟SCWR中主热交换器的一种可能情况。通过对热流体质量流量进行参数化评价，评价了超临界-亚临界过渡的影响。结果证明了超临界-亚临界转变的多维性，流体流动在x和y方向上都表现出很大的热性质变化。它阐明了在处理伪临界点附近的超临界流动时考虑温度相关性质和更可靠的数值方法的重要性。

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引用次数: 0

Integer programming approach for nuclear power plant cable routing based on problem decomposition and workspace approximation 基于问题分解和工作空间逼近的核电厂电缆布线整数规划方法

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

Nuclear Engineering and Design

Pub Date : 2026-01-06 DOI: 10.1016/j.nucengdes.2025.114732

Zihao Tu , Xinzhi Zhou , Yudi Zhu , Wenjing Liu , Xiaoping Yue , Yu Ma , Qian Ning

Cable harness routing is a critical component in the layout design of complex electromechanical systems in nuclear power plants, where its accuracy and safety directly impact the operational reliability of nuclear island buildings. However, current mainstream metaheuristic algorithms suffer from inherent limitations in solution quality control and physical interpretability. Meanwhile, traditional edge-driven multi-commodity network flow integer programming models face computational infeasibility due to the combinatorial explosion of decision variables, particularly struggling to meet the stringent engineering constraints of radiation shielding, safety clearances, and hierarchical routing in nuclear power scenarios. To address these bottlenecks, this study proposes a node-driven Binary Integer Linear Programming model (BILP) combined with a Problem Decomposition and Workspace Approximation Algorithm (PD&WA), achieving a balance between mathematical rigor and computational feasibility for nuclear power cable routing problems. Validation through industrial-scale nuclear island building scenarios in nuclear power plants (involving 10 groups of cables with different voltage classes and protection categories) shows that under equivalent time budgets, the proposed strategy reduces the number of decision variables by 99.6 % compared to the original model. The objective function values are decreased by 81.9 %, 51.2 %, and 79.5 % relative to the original model, literature benchmarks, and metaheuristic algorithms, respectively. Theoretical analysis and experimental data confirm that the near-globally optimal solutions generated by this framework within finite time can be directly applied to nuclear power cable layout design, establishing a novel paradigm that integrates mathematical rigor and computational feasibility for high-dimensional constrained path planning problems in nuclear reactor systems.

电缆线束布线是核电站复杂机电系统布置图设计中的关键组成部分，其准确性和安全性直接影响到核岛建筑的运行可靠性。然而，目前主流的元启发式算法在解的质量控制和物理可解释性方面存在固有的局限性。同时，由于决策变量的组合爆炸，传统的边缘驱动多商品网络流整数规划模型面临计算上的不可行性，特别是难以满足核电场景中辐射屏蔽、安全间隙和分层路由等严格的工程约束。为了解决这些瓶颈，本研究提出了一种节点驱动的二进制整数线性规划模型（BILP），结合问题分解和工作空间近似算法（PD&；WA），实现了核电电缆布线问题的数学严谨性和计算可行性之间的平衡。通过核电站工业规模核岛建设场景（涉及10组不同电压等级和保护类别的电缆）的验证表明，在相同的时间预算下，与原始模型相比，所提出的策略减少了99.6%的决策变量数量。与原始模型、文献基准和元启发式算法相比，目标函数值分别下降了81.9%、51.2%和79.5%。理论分析和实验数据证实，该框架在有限时间内生成的近全局最优解可直接应用于核电电缆布设设计，为核反应堆系统高维约束路径规划问题建立了一个集数学严密性和计算可行性于一体的新范式。

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引用次数: 0

Modeling of the irradiation embrittlement dependent fracture strength and mode I fracture toughness for ceramics 陶瓷辐照脆化相关断裂强度和I型断裂韧性的建模

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

Nuclear Engineering and Design

Pub Date : 2026-01-05 DOI: 10.1016/j.nucengdes.2025.114738

Feilong Zhang , Weiguo Li , Yanli Ma , Zhaoliang Qu , Liming Chen , Wenli Pi

Quantitative characterization of irradiation embrittlement effect is important to ensure the service safety of ceramic materials in irradiation environments. Based on the Li's Principle of Energy Equivalence, taking into account the equivalent relationship between the irradiation energy absorbed by the material and the strain energy, theoretical models for irradiation embrittlement dependent fracture strength and mode I fracture toughness of ceramics have been developed, respectively. These models reveal a quantitative relationship between irradiation fluence and fracture strength/mode I fracture toughness, enabling the prediction of irradiation data over a wide range based on limited irradiation experimental data. The predicted results of both models are well validated by all available experimental data, including two groups of fracture strength data and four groups of mode I fracture toughness data. These models offer an effective theoretical approach for the convenient prediction of the mechanical properties of ceramic materials under wide irradiation fluence ranges. This effectively reduces reliance on difficult irradiation experiments, while providing a theoretical basis for the structural design and reliability assessment of reactor ceramics, as well as for regulatory and cost optimization aspects.

辐照脆化效应的定量表征对于保证陶瓷材料在辐照环境中的使用安全具有重要意义。基于Li’s能量等效原理，考虑材料吸收的辐照能与应变能之间的等效关系，分别建立了陶瓷辐照脆化相关断裂强度和I型断裂韧性的理论模型。这些模型揭示了辐照通量与断裂强度/ I型断裂韧性之间的定量关系，从而能够基于有限的辐照实验数据在大范围内预测辐照数据。2组断裂强度数据和4组I型断裂韧性数据验证了两种模型的预测结果。这些模型为方便地预测大辐照影响范围下陶瓷材料的力学性能提供了有效的理论方法。这有效地减少了对困难的辐照实验的依赖，同时为反应堆陶瓷的结构设计和可靠性评估，以及监管和成本优化方面提供了理论依据。

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引用次数: 0

An integrated 1D3D analysis framework for thermal-hydraulic performance evaluation of a straight-tube once-through steam generator for SMRs smr直管式直流蒸汽发生器热工性能评价的集成1D3D分析框架

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

Nuclear Engineering and Design

Pub Date : 2026-01-03 DOI: 10.1016/j.nucengdes.2026.114750

Haoyuan Yin , Kunwoo Yi , Youngjin Kim , Fang Wang , Peng Su , Shuting Li

In this study, an integrated analysis framework combining a one-dimensional numerical program with a three-dimensional CFD model was developed to evaluate the heat transfer and flow characteristics of an in-vessel once-through steam generator (OTSG) for small modular reactors (SMRs). The one-dimensional model, based on a movable boundary method, is capable of capturing the phase-change location and the axial temperature distribution. Its results were validated against the design data of the B&W OTSG as well as the MARS-KS and Park models, showing good agreement in key parameters such as temperature, heat transfer area, and tube length. The effects of variations in power, mass flow rate, and secondary-side pressure on the heat transfer characteristics were investigated. The results indicate that when the power increases from 100 MW to 300 MW, the secondary-side inlet temperature deviates from the engineeringly feasible range; however, by jointly adjusting the primary and secondary mass flow rates, the inlet temperature can be maintained within a reasonable range, achieving thermal matching under different power levels. When the secondary-side pressure increases from 6.5 MPa to 7.5 MPa, the elevated saturation temperature shortens the superheated region and extends the two-phase region, demonstrating that pressure is a critical control parameter affecting phase-region distribution. Furthermore, after mapping the one-dimensional heat-flux distribution into a non-uniform volumetric heat sink, the three-dimensional CFD model successfully reproduced the characteristic temperature profile featuring a “two-phase plateau followed by a steep rise in the superheated region,” and revealed the patterns of local heat-load concentration and temperature-gradient variation. In contrast, the uniform heat-sink assumption underestimated the local temperature difference and thermal gradient. Overall, the proposed one-dimensional–three-dimensional coupled approach couples a one-dimensional thermal-hydraulic solver with a three-dimensional CFD model and provides a practical basis for the design evaluation and safety assessment of integrated OTSGs in SMRs.

本文采用一维数值计算和三维CFD模型相结合的集成分析框架，对小型模块化反应器（SMRs）的容器内一次性蒸汽发生器（OTSG）的传热和流动特性进行了评估。基于可移动边界法的一维模型能够捕捉相变位置和轴向温度分布。其结果与B&；W OTSG以及MARS-KS和Park模型的设计数据进行了验证，在温度、传热面积和管长等关键参数上显示出良好的一致性。研究了功率、质量流量和二次侧压力对传热特性的影响。结果表明：当功率从100 MW增加到300 MW时，二次侧进口温度偏离工程可行范围；但通过共同调节一次和二次质量流量，可使进口温度保持在合理范围内，实现不同功率水平下的热匹配。当二次侧压力从6.5 MPa增加到7.5 MPa时，饱和温度的升高缩短了过热区域，扩大了两相区域，说明压力是影响相区分布的关键控制参数。此外，在将一维热流密度分布映射到非均匀体积散热器后，三维CFD模型成功地再现了“两相平台后过热区急剧上升”的特征温度分布，并揭示了局部热负荷集中和温度梯度变化的模式。相反，均匀热沉假设低估了局部温差和热梯度。总体而言，本文提出的一维-三维耦合方法将一维热-液求解器与三维CFD模型相结合，为smr一体化otsg的设计评估和安全评估提供了实践依据。

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