Annals of Nuclear Energy最新文献_第5页

Fuzzy variable-parameter output feedback control method based on axial spatial distribution parameter model of nuclear reactor power 基于核反应堆功率轴向空间分布参数模型的模糊变参数输出反馈控制方法

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-02-05 DOI: 10.1016/j.anucene.2026.112181

Airan Dang, Bowen Tu, Xiuchun Luan

{"title":"Fuzzy variable-parameter output feedback control method based on axial spatial distribution parameter model of nuclear reactor power","authors":"Airan Dang, Bowen Tu, Xiuchun Luan","doi":"10.1016/j.anucene.2026.112181","DOIUrl":"10.1016/j.anucene.2026.112181","url":null,"abstract":"<div><div>Both the time variation and spatial distribution of nuclear reactor power are research objects in the field of nuclear reactor control. For the time variation of power, the control objective is to align the actual power change with the desired power change, namely achieving load following. For the spatial distribution of power, the control objective is to prevent overshoot when the axial peak power moves. Meeting both objectives is crucial for the safe operation of nuclear reactors. Therefore, the distributed parameter model is established in this paper based on the neutron diffusion equation to meet this requirement, as it is capable of simultaneously observing both power variation and axial spatial power distribution. Typically, two control objectives require two actuators, forming a dual-input dual-output (DIDO) system. However, the coupling between the two control rods complicates controller design significantly. Simulations of the movement of a single group of control rods have shown that it also influences both power time variation and spatial distribution. Thus, this paper designs a fuzzy variable-parameter output feedback controller for a single group of control rods, and simulation results verify the feasibility of this method. Compared with the DIDO system, the single-group control rod approach, while slightly inferior in performance, offers a substantial reduction in complexity.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112181"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186141","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Sealing behavior of bolted flange system with double metallic C-rings during prolonged high-temperature operation 双金属c形环螺栓法兰系统在长时间高温工况下的密封性能

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-02-06 DOI: 10.1016/j.anucene.2026.112182

Yijun Chai , Yunzhou Wu , Wenhao Yu , Qin Ma , Xiongwei Yang , Yueming Li

{"title":"Sealing behavior of bolted flange system with double metallic C-rings during prolonged high-temperature operation","authors":"Yijun Chai , Yunzhou Wu , Wenhao Yu , Qin Ma , Xiongwei Yang , Yueming Li","doi":"10.1016/j.anucene.2026.112182","DOIUrl":"10.1016/j.anucene.2026.112182","url":null,"abstract":"<div><div>Bolted flange joint, serving as a primary pressure barrier in nuclear reactors, is essential to operational safety. Double metallic C-ring shows potentials for next Generation reactors, while its sealing performance during prolonged high-temperature operation remains inadequately characterized. To this end, a numerical methodology incorporating essential factors such as bolt preload, high-temperature creep behavior, structural contact, and multi-physical coupling, is developed. The fundamental material creep behavior is characterized through high-temperature stress relaxation experiments. The investigation emphasizes evolutionary characteristics of contact stress and rebound behavior of the double C-rings across loading stages, while a higher bolt preload does not always improve sealing performance; conversely, its synergistic effect with internal medium pressure can lead to complete failure of the inner seal. Therefore, the inner ring experiences more significant rebound, indicating vulnerability for sealing failure. Additionally, the correlations among bolt preload, linear load, and rebound magnitude are quantitatively described. Results show that throughout long-term stress relaxation service, relaxation of both the linear load and bolt preload occurs predominantly within the first 120 h, at a rate below 5%, and stabilizes thereafter. This demonstrates that the bolted flange system can maintain a high preload level even after prolonged high-temperature service under the current operating conditions. The proposed framework could be critically useful in assessing the sealing status and the effective sealing duration of the flange system.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112182"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186193","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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Robust and fault-tolerant control of MSBR reactor using a hybrid QFT–PID–LSTM framework with disk margin analysis 基于QFT-PID-LSTM混合框架和磁盘裕度分析的MSBR反应器鲁棒容错控制

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-10 DOI: 10.1016/j.anucene.2026.112131

Achu Govind K.R.

{"title":"Robust and fault-tolerant control of MSBR reactor using a hybrid QFT–PID–LSTM framework with disk margin analysis","authors":"Achu Govind K.R.","doi":"10.1016/j.anucene.2026.112131","DOIUrl":"10.1016/j.anucene.2026.112131","url":null,"abstract":"<div><div>Reliable reactive power regulation in Molten Salt Breeder Reactor (MSBR) cores is essential for safety and efficiency. Conventional controllers often exhibit poor robustness and fault tolerance under uncertainties, nonlinear dynamics, and disturbances. To address these limitations, this work introduces a Quantitative Feedback Theory (QFT)-based proportional integral derivative (PID) controller enhanced with Long Short-Term Memory (LSTM) modeling. The controller is designed in two stages. First, a QFT-based PID controller is synthesized to guarantee stability and robust performance across plant uncertainty sets. Robustness, disturbance rejection, and tracking are enforced as frequency-domain inequalities. In the second stage, an LSTM network is integrated to adaptively tune PID gains in real time. This ensures that predictions inherently satisfy robust stability, tracking, and disturbance rejection constraints. A composite performance-driven loss further biases the network toward minimizing integral absolute error (IAE), overshoot, and settling time while preserving robustness. The QFT-PID-LSTM controller achieved faster rise time, shorter settling time, negligible steady-state error, and lower control effort compared to existing approaches. Quantitative indices showed reductions of nearly 90% in performance metrics and 80% in statistical measures. Disk margin analysis confirmed stability, while Monte Carlo simulations demonstrated tightly bounded error distributions. The controller also maintained stability and accurate tracking under both sensor and actuator faults, confirming strong robustness and fault tolerance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112131"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145924528","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub 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

{"title":"Transient multiphysics simulations with pin power reconstruction in the Griffin reactor physics code","authors":"Shikhar Kumar , Changho Lee , Vincent Laboure , Yeon Sang Jung , Stefano Terlizzi , Yaqi Wang , Javier Ortensi","doi":"10.1016/j.anucene.2025.112111","DOIUrl":"10.1016/j.anucene.2025.112111","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112111"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915172","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

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

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-27 DOI: 10.1016/j.anucene.2026.112169

Zhinan Xie , Chen Hao , Wen Yin

{"title":"Pin-resolved ex-core detector response function methodology based on the 2D/1D high-fidelity adjoint neutron transport calculation","authors":"Zhinan Xie , Chen Hao , Wen Yin","doi":"10.1016/j.anucene.2026.112169","DOIUrl":"10.1016/j.anucene.2026.112169","url":null,"abstract":"<div><div>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.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112169"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074248","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Analysis of fracture conditions of Cr-coated Zr alloy claddings under LOCA conditions calculated using FEMAXI fuel performance code 利用FEMAXI燃料性能程序计算了cr包覆Zr合金在LOCA条件下的断裂情况

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-11 DOI: 10.1016/j.anucene.2026.112114

Vu-Nhut Luu, Yoshinori Taniguchi, Yutaka Udagawa, Yudai Tasaki, Jinya Katsuyama

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Neutron Data Evaluation in the n + 241,243Am reactions below 200 MeV 200 MeV以下n + 241,243Am反应的中子数据评价

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-17 DOI: 10.1016/j.anucene.2026.112127

Xinwu Su , Yongli Xu , Yinlu Han

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Appropriate selection of objective functions with constraints for successive multicycle optimization of the HPR1000 reactor core using deep learning enhanced multi-objective genetic algorithm 基于深度学习增强型多目标遗传算法的HPR1000堆芯连续多周期优化目标函数的选择

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-02-04 DOI: 10.1016/j.anucene.2026.112128

Muhammad Kamran Butt , Samia Nasir , Muhammad Ahmad , Muhammad Talha Bin Humayun , Ahmer Riaz , Chenghui Wan , Liangzhi Cao

{"title":"Appropriate selection of objective functions with constraints for successive multicycle optimization of the HPR1000 reactor core using deep learning enhanced multi-objective genetic algorithm","authors":"Muhammad Kamran Butt , Samia Nasir , Muhammad Ahmad , Muhammad Talha Bin Humayun , Ahmer Riaz , Chenghui Wan , Liangzhi Cao","doi":"10.1016/j.anucene.2026.112128","DOIUrl":"10.1016/j.anucene.2026.112128","url":null,"abstract":"<div><div>Once the core exits from the equilibrium reloading scheme, the nuclear core designer needs to perform multicycle fuel reloading pattern optimization (MCO). MCO can be performed either as successive multicycle optimization, also known as cycle-by-cycle optimization, or simultaneous multicycle optimization, wherein a set of cycles is optimized simultaneously. In this research, successive multicycle optimization (SCO) is performed for five consecutive cycles of the HPR1000 reactor using deep learning and multi-objective genetic algorithm (MOGA). Ten suitable objective functions were tested. These objective functions were carefully subjected to certain constraints such that the reactor safety criteria are fulfilled in each cycle of MCO. The results of SCO were compared with the reference equilibrium reloadings, and it was concluded that the SCO reloading patterns were safer than the equilibrium reloading patterns by having more margins in the reactor safety criteria at the cost of an average cycle length of 1.6 EFPDs per cycle.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112128"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186142","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0

Online fault detection and diagnostics for self powered neutron detector along with its cable using hybrid techniques of noise analysis and clustering algorithm 基于噪声分析和聚类算法的自供电中子探测器及其电缆在线故障检测与诊断

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-02-07 DOI: 10.1016/j.anucene.2026.112176

Tahir Kamal Khan , S.R. Shimjith , S. Mukhopadhyay , Ajita Shrivastava

{"title":"Online fault detection and diagnostics for self powered neutron detector along with its cable using hybrid techniques of noise analysis and clustering algorithm","authors":"Tahir Kamal Khan , S.R. Shimjith , S. Mukhopadhyay , Ajita Shrivastava","doi":"10.1016/j.anucene.2026.112176","DOIUrl":"10.1016/j.anucene.2026.112176","url":null,"abstract":"<div><div>Self-Powered Neutron Detectors (SPNDs) are widely used in large nuclear reactors for in-core flux measurement, supporting monitoring, control, and protection functions. However, SPNDs and their associated cables are prone to aging and degradation, making robust fault detection and diagnosis (FDD) essential for ensuring the safety and reliability of nuclear power plant operations. This work presents a novel hybrid methodology that integrates neutron noise analysis with a K-means clustering algorithm for online, non-intrusive and continuous health monitoring and predictive maintenance of SPNDs and their associated cables. Neutron flux within the reactor core exhibits measurable fluctuations in the 0–20 Hz range due to process-related noise. It is shown that SPNDs’ ability to capture these process related noise characteristics can serve as an effective indicator of their health. A comprehensive SPND model, including its cable and amplifier, is first developed to demonstrate the proposed approach through simulations. A new feature extraction technique is then introduced to characterize the dynamic response obtained from noise analysis, allowing clustering-based classification of SPNDs into healthy, degraded, and severely degraded states. The methodology is validated using plant data from an Indian PHWR employing Inconel SPNDs for safety and control applications. Results show strong agreement with existing offline SPND health assessments, confirming the effectiveness of the proposed approach. Analytical results demonstrate that drift in gain due to insulation resistance reduction of SPNDs is quadratically amplified for power spectral density (PSD) of output signal, leading to enhanced fault sensitivity compared to time-domain representations. This helps to identify faults during clustering. Results have been compared with published literature and are found to be with better accuracy and performance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112176"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186191","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

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

Annals of Nuclear Energy Pub Date : 2026-06-01 Epub Date: 2026-01-23 DOI: 10.1016/j.anucene.2026.112151

Samuel Gant Dotson , Madicken Munk , Kathryn Dorsey Huff

{"title":"Demonstrating the Osier framework for energy system and nuclear fuel cycle optimization","authors":"Samuel Gant Dotson , Madicken Munk , Kathryn Dorsey Huff","doi":"10.1016/j.anucene.2026.112151","DOIUrl":"10.1016/j.anucene.2026.112151","url":null,"abstract":"<div><div>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, <span>Osier</span>, is demonstrated. <span>Osier</span> leverages genetic algorithms to calculate a set of co-optimal solutions called a <em>Pareto front</em>. <span>Osier</span> 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, <span>Osier</span> 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 <span>Osier</span>’s suitability for energy modeling problems with two <em>in silico</em> experiments. The first set of experiments compare <span>Osier</span> to a more mature energy system optimization model, <span>Temoa</span>, to verify that <span>Osier</span> produces results consistent with known methods. The results for a least-cost optimization with <span>Osier</span> and <span>Temoa</span> show strong agreement, within 0.5% of each other. In the second, <span>Osier</span> reanalyzes a set of nuclear fuel cycle options from the <span>SET</span> tool through the lens of Pareto optimality.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112151"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035492","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

引用次数: 0