Annals of Nuclear Energy最新文献

英文中文

Pin-resolved ex-core detector response function methodology based on the 2D/1D high-fidelity adjoint neutron transport calculation 基于二维/一维高保真伴随中子输运计算的引脚分辨力前核探测器响应函数方法

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

Annals of Nuclear Energy

Pub Date : 2026-01-27 DOI: 10.1016/j.anucene.2026.112169

Zhinan Xie , Chen Hao , Wen Yin

Accurate calculation of the detector response function is critical in power reconstruction using ex-core detectors. The Monte Carlo and 3D discrete ordinates methods have been applied to calculate the ex-core detectors response function. However, since ex-core detectors are located far from the reactor core, the large neutron flux gradients between the core and the ex-core detector, as well as the weak contribution of in-core neutrons to the detector response, lead to significant limitations of the computational efficiency and accuracy for the Monte Carlo method. And for 3D discrete ordinates method, the resolution of the detector response function is limited due to the homogenization approximations. Therefore, achieving efficient and accurate whole-core, pin-resolved detector response function calculations remain a significant challenge. To address the challenges of computational efficiency and resolution inherent in conventional methods, the 2D Method of Characteristics / 1D Nodal Expansion Method coupling method with multi-group coarse mesh finite difference method acceleration is used to carry out high-fidelity adjoint transport calculations, enabling direct pin-resolved detector response function calculation. Numerical verification is performed using the 2D EPRI-9 model, the 3D C5G7 model and the low temperature heating reactor. The results demonstrate that the 2D/1D method can accurately and efficiently compute pin-resolved detector response function, achieving well agreement with Monte Carlo results.

探测器响应函数的精确计算是利用脱芯探测器进行功率重构的关键。采用蒙特卡罗法和三维离散坐标法计算了前核探测器的响应函数。然而，由于前堆芯探测器距离反应堆堆芯较远，堆芯与前堆芯探测器之间的中子通量梯度较大，以及堆芯中子对探测器响应的贡献较小，导致蒙特卡罗方法的计算效率和精度受到很大限制。而对于三维离散坐标法，由于均匀化近似，探测器响应函数的分辨率受到限制。因此，实现高效、准确的全核、引脚解析检测器响应函数计算仍然是一个重大挑战。为了解决传统方法固有的计算效率和分辨率问题，采用二维特征法/一维节点展开法耦合多组粗网格有限差分法加速进行高保真伴随输运计算，实现直接针分辨探测器响应函数计算。采用二维EPRI-9模型、三维C5G7模型和低温加热反应器进行数值验证。结果表明，二维/一维方法可以准确、高效地计算引脚分辨探测器响应函数，与蒙特卡罗结果吻合较好。

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引用次数: 0

Reinvestigating the off-grid project priorities of small-scale nuclear reactors using an enhanced integrated fuzzy decision support system 基于增强型集成模糊决策支持系统的小型核反应堆离网工程优先级再研究

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

Annals of Nuclear Energy

Pub Date : 2026-01-27 DOI: 10.1016/j.anucene.2026.112160

Serhat Yüksel , Hasan Dinçer , Merve Acar , Edanur Ergün , Serkan Eti

Small-scale nuclear reactors represent a promising option for providing reliable and continuous energy in off-grid and isolated regions; however, limited investment budgets necessitate a clear prioritization of project objectives. This study addresses the need to identify the most critical project priorities and the most suitable off-grid energy applications for small-scale nuclear reactor deployments. To this end, a novel decision-making framework is developed by integrating artificial intelligence-based expert weighting with advanced fuzzy modeling techniques to effectively manage uncertainty and incomplete evaluations. The proposed approach enables a systematic assessment of strategic priorities without being tied to a specific reactor technology. The results indicate that security supported by passive safety systems is the most influential project priority, followed by cost effectiveness and operational flexibility. When alternative off-grid applications are evaluated, steady energy supply for rural industry fields emerges as the most appropriate option due to its strong and balanced performance across safety, economic, and operational dimensions. These findings highlight the interdependence between technical design considerations and application-level decisions. Overall, the study provides practical insights for policymakers and project managers by identifying strategic priorities that can enhance the effectiveness, feasibility, and long-term viability of small-scale nuclear energy investments in off-grid contexts.

小型核反应堆是在离网和偏远地区提供可靠和持续能源的有希望的选择；然而，有限的投资预算需要明确项目目标的优先次序。本研究解决了确定最关键的项目优先级和最适合小型核反应堆部署的离网能源应用的需要。为此，将基于人工智能的专家权重与先进的模糊建模技术相结合，开发了一种新的决策框架，以有效地管理不确定性和不完整的评估。拟议的方法能够在不依赖于特定反应堆技术的情况下对战略优先事项进行系统评估。结果表明，被动安全系统支持的安全性是最具影响力的项目优先级，其次是成本效益和操作灵活性。当评估替代性离网应用时，稳定的农村工业领域能源供应成为最合适的选择，因为它在安全、经济和运营方面具有强大而平衡的性能。这些发现突出了技术设计考虑和应用程序级决策之间的相互依赖关系。总体而言，该研究通过确定可以提高离网环境下小型核能投资的有效性、可行性和长期可行性的战略重点，为政策制定者和项目经理提供了实用的见解。

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引用次数: 0

Operational strategies for nuclear district heating systems in extremely cold climates 极冷气候下核区域供热系统的运行策略

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

Annals of Nuclear Energy

Pub Date : 2026-01-27 DOI: 10.1016/j.anucene.2026.112162

Ruo-Jun Xue , Han-Wen Liang , Hao-Fang Chong , Min-Jun Peng

As energy structures evolve and environmental standards rise, nuclear energy shows clear advantages in district heating, particularly in cold regions. To ensure consistent quality and stability of supply, accurate and responsive load-following capability is essential for the effective use of nuclear energy in the heating sector.

This study presents a simulation model for a 400 MW pool-type nuclear heating reactor, accompanied by the development of a comprehensive system simulation platform using Python and Computational Fluid Dynamics (CFD) for both one-dimensional (1D) and three-dimensional (3D) coupled analyses. The research systematically examines the operational parameters and strategies of the nuclear heating system under extreme cold climate conditions. The simulation results indicate that under the constant-temperature heating mode, after accounting for the thermal delay characteristics of the reactor pool, the number of power adjustments required during the 139-day heating season is reduced by 22 instances. Under the variable-temperature heating mode, the independent heating configuration of this reactor can satisfy the thermal demand during the heating period in cold regions.

随着能源结构的演变和环境标准的提高，核能在区域供热方面显示出明显的优势，特别是在寒冷地区。为了确保供应的质量和稳定性，准确和响应的负荷跟踪能力对于在供热部门有效利用核能至关重要。本研究提出了400mw池式核加热堆的仿真模型，并利用Python和计算流体动力学（CFD）开发了一个综合系统仿真平台，用于一维（1D）和三维（3D）耦合分析。该研究系统地考察了在极端寒冷气候条件下核加热系统的运行参数和策略。仿真结果表明，在恒温加热模式下，考虑反应堆池热延迟特性后，139天采暖季所需的功率调整次数减少了22次。在变温加热模式下，该反应器的独立加热配置可以满足寒冷地区采暖期间的热需求。

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引用次数: 0

A deep learning-based prognostic approach for predicting PWR degradation and remaining useful life using GNN-PTC-LSTM 基于GNN-PTC-LSTM的基于深度学习的压水堆退化和剩余使用寿命预测方法

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

Annals of Nuclear Energy

Pub Date : 2026-01-27 DOI: 10.1016/j.anucene.2026.112172

Shadman Ahmad Khattak , Liu Yong-Kuo , Liu Yu-Kun , Gao Jia-Rong , Shi Zhou-Xin , Liu Ji

Early fault detection and prediction are key for ensuring the proper functioning and safety operation of complex nuclear systems especially in nuclear power production. Remaining Useful Life (RUL) prediction in Pressurized Water Reactors (PWRs) is one of the crucial parameters to prevent gradual degradation and disaster. In this paper a hybrid model composed of graph neural network,physics topology constraint and long short term memory network (GNN-PTC-LSTM) is proposed for fault prognostic and remaining useful life prediction based on the nuclear power plant using 2 loop PWR PCTRAN datasets. The proposed framework employs Graph Neural Networks (GNNs) to capture the spatial dependencies between reactor subsystems while integrating a PTC-LSTM module that incorporates physical topology and plant dynamics as temporal sequence learning constraints. In contrast to traditional LSTMs which are purely statistical model the physics-informed PTC-LSTM integrates a priori understanding of domain knowledge to mitigate non-physical predictions models outputs that violate known physical/topological constraints and reduce false alarms due to spurious data correlations. Additionally, One-Class SVM is used for anomaly detection in multivariate telemetry data to allow early discovery of abnormal behaviours. The proposed GNN-PTC-LSTM framework achieved an overall fault classification accuracy of 99.1%, an early warning accuracy of 98.2%, and competitive RUL prediction performance with a mean absolute error (MAE) of 0.0042 and root mean square error (RMSE) of 0.0105 under simulated PWR accident scenarios.

早期故障检测和预测是保证复杂核系统正常运行和安全运行的关键，特别是在核电生产中。压水堆剩余使用寿命（RUL）预测是防止压水堆逐渐退化和发生灾害的关键参数之一。本文利用2环压水堆PCTRAN数据集，提出了一种由图神经网络、物理拓扑约束和长短期记忆网络组成的混合模型（GNN-PTC-LSTM），用于核电厂的故障预测和剩余使用寿命预测。提出的框架采用图神经网络（gnn）来捕获反应堆子系统之间的空间依赖关系，同时集成PTC-LSTM模块，该模块将物理拓扑和植物动态作为时间序列学习约束。与纯统计模型的传统lstm相比，物理信息灵通的PTC-LSTM集成了对领域知识的先验理解，以减轻违反已知物理/拓扑约束的非物理预测模型输出，并减少由于虚假数据相关性引起的误报。此外，将一类支持向量机用于多变量遥测数据的异常检测，以便及早发现异常行为。所提出的GNN-PTC-LSTM框架在模拟压水堆事故场景下，总体故障分类准确率为99.1%，预警准确率为98.2%，竞争性RUL预测性能的平均绝对误差（MAE）为0.0042，均方根误差（RMSE）为0.0105。

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引用次数: 0

Ray Adaptive Stochastic Transport (RASTr): Importance sampling based variance reduction for characteristics method transport solvers 射线自适应随机输运（RASTr）：基于重要性抽样的特征法输运解的方差缩减

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

Annals of Nuclear Energy

Pub Date : 2026-01-26 DOI: 10.1016/j.anucene.2026.112152

Owen Mylotte, Benoit Forget

The Random Ray Method (TRRM) neutron transport solver is a robust variation of the Method of Characteristics (MOC) for which ray sampling is uniform in space and angle, as opposed to the typical fixed quadrature cyclical ray tracking. However, there are several classes of problems for which the implicitly assumed uniform sampling distribution of TRRM may not be optimal. This work introduces the Ray Adaptive Stochastic Transport (RASTr) method, which computes statistical weights for an arbitrary spatially and angularly biased ray sampling distribution to provide variance reduction for problems where non-uniform sampling is desirable. The RASTr algorithm is implemented and tested on 1D and 2D test cases with demonstrated improvement in statistical relative uncertainty.

随机射线法（TRRM）中子输运求解器是特征法（MOC）的一种鲁棒变体，它的射线采样在空间和角度上是均匀的，而不是典型的固定正交周期射线跟踪。然而，有几类问题，隐式假设的均匀抽样分布的TRRM可能不是最优的。这项工作介绍了射线自适应随机传输（RASTr）方法，该方法计算任意空间和角度偏置的射线采样分布的统计权重，为需要非均匀采样的问题提供方差减少。RASTr算法在一维和二维测试用例上进行了实现和测试，证明了统计相对不确定性的改善。

引用次数: 0

Plutonium diversion detection in molten salt reactors via gamma emitter signature 利用伽玛辐射源特征探测熔盐反应堆中的钚转移

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

Annals of Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.anucene.2026.112149

Alvin J.H. Lee, Tomasz Kozlowski

In Molten Salt Reactors (MSRs), the easy accessibility of the fuel salt is a potential avenue for plutonium diversion and safeguard inspectors must have the means to detect such diversion to limit nuclear proliferation. In this work, we developed a mathematical basis to translate the radioactivities of select gamma emitting fission products into the fissile isotope ratios of the reactor, and described a methodology to detect plutonium diversion. This work identified ^138mCs/^134mI as the species pair that can predict the fissile isotope ratios with good accuracy while under chemical effects such as reprocessing. Simulated diversion cases demonstrated good accuracy of around 0.96% discrepancy from the Monte Carlo estimate for ^138mCs/^134mI, and around 0.034% for other potential pairs when less limiting conditions were considered (e.g., finite but slow precursor removal vs decay). The detection methodology contributes to the existing MSR safeguards efforts, which is necessary for the successful deployment of MSRs.

在熔盐堆（MSRs）中，燃料盐的易获取性是钚转移的潜在途径，保障监督检查员必须具备检测这种转移的手段，以限制核扩散。在这项工作中，我们开发了一个数学基础，将选定的伽马发射裂变产物的放射性转化为反应堆的可裂变同位素比率，并描述了一种检测钚转移的方法。本研究发现138mCs/134mI是在后处理等化学作用下，能较准确预测可裂变同位素比值的物质对。模拟的分流情况表明，当考虑较少的限制条件（例如，有限但缓慢的前体去除与衰变）时，与蒙特卡罗估计的138mCs/134mI的误差约为0.96%，与其他潜在对的误差约为0.034%。检测方法有助于现有的MSR保障工作，这是成功部署MSR所必需的。

引用次数: 0

Numerical investigation of wall effects on cross flow over inline tube bundles with various pitch-to-diameter ratios 不同节径比直列管束横向流动壁面效应的数值研究

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

Annals of Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.anucene.2026.112168

Yifan Zhou, Houjian Zhao, Yang Liu

Shell and tube heat exchangers are widely used in nuclear engineering and the petrochemical industries. In the current investigation, cross flow over inline tube bundles with various pitch-to-diameter ratios is simulated by SST k-ω-γ. The mesh near the shear layer region is refined due to the large velocity gradient. The effects of the bounding wall, end wall, and pitch ratio on time-averaged and transient flow fields are systematically analyzed. The increased streamwise pitch results in the impinging point shifting to near θ = 0°. The increased transverse pitch results in a larger influence on side passages. The recirculation region near the end wall is attenuated, resulting in reduced drag and large velocity magnitude. The separated vortices after the middle tubes sway into the main flow. There is a separation vortex near the bounding wall due to the entrainment of separate vortices.

管壳式换热器广泛应用于核工程和石油化工行业。在目前的研究中，利用SST k-ω-γ模拟了不同节径比的直列管束的横向流动。切变层区域附近的网格由于速度梯度较大而进行了细化。系统分析了边界壁、端壁和螺距比对时均流场和瞬态流场的影响。增大的顺流螺距导致碰撞点移至θ = 0°附近。横向节距的增加对侧通道的影响更大。端壁附近的再循环区域被衰减，导致阻力减小和速度大小增大。中间管道进入主流后的分离涡。由于分离涡的夹带，在边界壁附近存在分离涡。

引用次数: 0

Neutronic characteristics of a partially damaged reactor model with varying numbers of damaged fuel assemblies 不同数量燃料组件损坏的部分损坏反应堆模型的中子特性

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

Annals of Nuclear Energy

Pub Date : 2026-01-24 DOI: 10.1016/j.anucene.2026.112171

Hoang Hai Nguyen

This study examined the effects of the moderator-to-fuel volume ratio, fuel debris shape, and the number of damaged fuel assemblies on the neutronic features of a partially damaged reactor, where the fuel assemblies in the center region collapse into debris and the fuel assemblies in the outer region are kept unchanged. The partially damaged reactor model was based on the Watts Bar Nuclear 1 reactor. The investigations were performed by the Serpent code. The findings show that in cases where the debris is surrounded by intact fuel assemblies, the change of k_eff depends on the geometry of the debris. Conversely, in scenarios where the debris is not fully encircled by intact fuel assemblies, the geometry of the debris has a negligible impact on the k_eff. Additionally, the number of neutrons entering and leaving the debris determines how its shape affects the k_eff.

本研究考察了慢化剂与燃料体积比、燃料碎片形状和受损燃料组件数量对部分受损反应堆中子特征的影响，其中中心区域的燃料组件坍塌成碎片，而外部区域的燃料组件保持不变。部分损坏的反应堆模型是基于瓦茨巴核1号反应堆。调查是由毒蛇代码执行的。研究结果表明，在碎片被完整的燃料组件包围的情况下，keff的变化取决于碎片的几何形状。相反，在碎片没有被完整的燃料组件完全包围的情况下，碎片的几何形状对keff的影响可以忽略不计。此外，进入和离开碎片的中子数量决定了碎片的形状如何影响keff。

引用次数: 0

Performance of multiple-type reference electrode oxygen sensors in LBE LBE中多类型参比电极氧传感器的性能

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

Annals of Nuclear Energy

Pub Date : 2026-01-23 DOI: 10.1016/j.anucene.2026.112129

Ruixian Liang, Hui Li, Huiping Zhu, Hao Wu, Haicai Lyu, Zulong Hao, Yang Liu, Fang Liu, Fenglei Niu

The performance of various oxygen sensors in liquid lead–bismuth eutectic (LBE) alloy varies significantly across different temperature ranges. Therefore, it is necessary to establish a comprehensive temperature-related calibration database to achieve real-time dynamic calibration and compensation for sensor measurements. In this paper, multiple types of oxygen sensors have been developed based on 8YSZ ceramic tubes. The air reference (LSCF/Air, LSM/Air and Ag/Air) oxygen sensors and metal/metal oxide (Cu/Cu₂O, Fe/Fe₃O₄, Ni/NiO, Bi/Bi₂O₃, and In/In₂O₃) reference oxygen sensors were tested under different temperature variations to obtain their operating characteristics in different temperature ranges. The air reference oxygen sensors have been demonstrated to exhibit excellent response speed, accuracy and stability within the range of 205 ∼ 550℃. The metal/metal oxide reference oxygen sensor is more suitable for applications in the medium to high temperature range (≥350 ℃). It provides reference data for the operation of non-isothermal lead–bismuth system oxygen sensors.

液态铅铋共晶（LBE）合金中各种氧传感器的性能在不同温度范围内有显著差异。因此，有必要建立一个全面的与温度相关的校准数据库，以实现对传感器测量的实时动态校准和补偿。本文基于8YSZ陶瓷管研制了多种类型的氧传感器。对空气基准氧传感器（LSCF/ air、LSM/ air和Ag/ air）和金属/金属氧化物基准氧传感器（Cu/Cu2O、Fe/Fe3O4、Ni/NiO、Bi/Bi2O3和In/In2O3）在不同温度范围内的工作特性进行了测试。空气参考氧传感器已被证明在205 ~ 550℃范围内具有优异的响应速度、准确性和稳定性。金属/金属氧化物基准氧传感器更适用于中至高温范围（≥350℃）的应用。为非等温铅铋系氧传感器的工作提供了参考数据。

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引用次数: 0

Demonstrating the Osier framework for energy system and nuclear fuel cycle optimization 展示了能源系统和核燃料循环优化的Osier框架

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

Annals of Nuclear Energy

Pub Date : 2026-01-23 DOI: 10.1016/j.anucene.2026.112151

Samuel Gant Dotson , Madicken Munk , Kathryn Dorsey Huff

Energy system optimization models are a class of tools designed to optimize energy planning and are used by energy planners and decision-makers to generate insights that inform energy policy. However, existing tools are challenged by real-world scenarios which require optimization across multiple objectives. In this paper, the multi-objective energy system optimization framework, Osier, is demonstrated. Osier leverages genetic algorithms to calculate a set of co-optimal solutions called a Pareto front. Osier also introduces a novel algorithm to identify a subset of maximally different solutions within the sub-optimal space to address structural uncertainty related to unmodeled objectives. By producing multiple solutions, Osier gives modelers and decision-makers the tools to meaningfully engage with public stakeholders and learn their preferences, thereby attending to issues of procedural and recognition justice. This work verifies Osier’s suitability for energy modeling problems with two in silico experiments. The first set of experiments compare Osier to a more mature energy system optimization model, Temoa, to verify that Osier produces results consistent with known methods. The results for a least-cost optimization with Osier and Temoa show strong agreement, within 0.5% of each other. In the second, Osier reanalyzes a set of nuclear fuel cycle options from the SET tool through the lens of Pareto optimality.

能源系统优化模型是一类旨在优化能源规划的工具，被能源规划者和决策者用来产生为能源政策提供信息的见解。然而，现有的工具受到需要跨多个目标进行优化的现实场景的挑战。本文对多目标能源系统优化框架Osier进行了论证。Osier利用遗传算法来计算一组称为帕累托前沿的共同最优解。Osier还引入了一种新的算法来识别次最优空间中最大不同解的子集，以解决与未建模目标相关的结构不确定性。通过提供多种解决方案，Osier为建模者和决策者提供了有意义地与公众利益相关者接触并了解他们偏好的工具，从而解决程序和认可正义的问题。本文通过两个计算机实验验证了Osier对能量建模问题的适用性。第一组实验将Osier与更成熟的能源系统优化模型Temoa进行比较，验证Osier的结果与已知方法一致。Osier和Temoa的最小成本优化结果显示出很强的一致性，误差在0.5%以内。在第二部分中，Osier通过帕累托最优的视角重新分析了set工具中的一组核燃料循环选项。

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