Pub Date : 2026-03-01Epub Date: 2025-12-12DOI: 10.1016/j.pnucene.2025.106185
Siwei Li , Yichun Wu , Jiayan Fang , Wei Wang , Jiale Ling
Early fault detection and timely initiation of appropriate control or maintenance actions can significantly mitigate operational risks in nuclear power plants (NPPs) and enhance the reliability of operator decision-making during emergencies. Therefore, developing an efficient, multi-step Prognostics and Health Management (PHM) model is essential for the prediction of system health status to support accident mitigation and recovery efforts in NPP operations. In this paper, we propose a novel predictive approach that integrates a bidirectional long short-term memory (BiLSTM) neural network within a multiple input multiple output (MIMO) framework guided by an Expert Fuzzy Evaluation Method (EFEM). The model is trained and validated on simulation data from a CPR1000 pressurized water reactor under various accident scenarios, and it demonstrates a remarkable capability to accurately forecast key NPP parameters up to 128 steps into the future (with a 10-s interval per step, totaling 1280 s, thereby satisfying the temporal advance requirement for fault prognostics in NPPs, and provides valuable decision-support information for operators to control and recover from accidents. The proposed method also serves as an effective reference for other PHM applications such as anomaly detection and remaining useful life estimation.
{"title":"Enhanced safety and emergency decision-making in NPPs: Multi-step parameter prediction for complex accident scenarios","authors":"Siwei Li , Yichun Wu , Jiayan Fang , Wei Wang , Jiale Ling","doi":"10.1016/j.pnucene.2025.106185","DOIUrl":"10.1016/j.pnucene.2025.106185","url":null,"abstract":"<div><div>Early fault detection and timely initiation of appropriate control or maintenance actions can significantly mitigate operational risks in nuclear power plants (NPPs) and enhance the reliability of operator decision-making during emergencies. Therefore, developing an efficient, multi-step Prognostics and Health Management (PHM) model is essential for the prediction of system health status to support accident mitigation and recovery efforts in NPP operations. In this paper, we propose a novel predictive approach that integrates a bidirectional long short-term memory (BiLSTM) neural network within a multiple input multiple output (MIMO) framework guided by an Expert Fuzzy Evaluation Method (EFEM). The model is trained and validated on simulation data from a CPR1000 pressurized water reactor under various accident scenarios, and it demonstrates a remarkable capability to accurately forecast key NPP parameters up to 128 steps into the future (with a 10-s interval per step, totaling 1280 s, thereby satisfying the temporal advance requirement for fault prognostics in NPPs, and provides valuable decision-support information for operators to control and recover from accidents. The proposed method also serves as an effective reference for other PHM applications such as anomaly detection and remaining useful life estimation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106185"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145739066","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}
Condensation widely exists in nature and engineering. Optimizing and modeling relevant processes heavily rely on accurate heat transfer coefficient (HTC) prediction. Nevertheless, obtaining precise HTC through empirical and numerical approaches continues to present significant challenges. This study aims to develop accurate and interpretable HTC prediction models using advanced machine learning (ML) techniques. A comprehensive database of 680 experimental data points was compiled to evaluate Support Vector Regression (SVR), ensemble learning models, and the AutoGluon platform against traditional correlations. The models predicted HTC based on total pressure Ptot, the mass fraction of non-condensable gas (NCG) wnc, the wall supercooling degree ΔT, and geometric parameters. Results demonstrate that all ML models exhibit superior performance compared to traditional correlations. Specifically, the AutoGluon platform outperforms other ML methods in predicting the HTC, according to the prediction accuracy and generalization ability comprehensively, with an MAPE of 3.16 %, RMSE of 69.46 and an R2 of 0.993. Furthermore, SHapley Additive exPlanations (SHAP) analysis identifies the NCG and total pressure as the dominant variables affecting HTC. These findings align well with experimental physical mechanisms, validating the interpretability of models.
{"title":"Highly accurate and interpretable machine learning models for predicting HTC in the presence of NCG on outer vertical tubes","authors":"Huifang Zhang, Yapei Zhang, Shihao Wu, Wenxi Tian, Suizheng Qiu, Guanghui Su","doi":"10.1016/j.pnucene.2025.106198","DOIUrl":"10.1016/j.pnucene.2025.106198","url":null,"abstract":"<div><div>Condensation widely exists in nature and engineering. Optimizing and modeling relevant processes heavily rely on accurate heat transfer coefficient (HTC) prediction. Nevertheless, obtaining precise HTC through empirical and numerical approaches continues to present significant challenges. This study aims to develop accurate and interpretable HTC prediction models using advanced machine learning (ML) techniques. A comprehensive database of 680 experimental data points was compiled to evaluate Support Vector Regression (SVR), ensemble learning models, and the AutoGluon platform against traditional correlations. The models predicted HTC based on total pressure P<sub>tot</sub>, the mass fraction of non-condensable gas (NCG) w<sub>nc</sub>, the wall supercooling degree ΔT, and geometric parameters. Results demonstrate that all ML models exhibit superior performance compared to traditional correlations. Specifically, the AutoGluon platform outperforms other ML methods in predicting the HTC, according to the prediction accuracy and generalization ability comprehensively, with an MAPE of 3.16 %, RMSE of 69.46 and an R<sup>2</sup> of 0.993. Furthermore, SHapley Additive exPlanations (SHAP) analysis identifies the NCG and total pressure as the dominant variables affecting HTC. These findings align well with experimental physical mechanisms, validating the interpretability of models.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"193 ","pages":"Article 106198"},"PeriodicalIF":3.2,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145789027","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}
Pub Date : 2026-02-01Epub Date: 2025-11-04DOI: 10.1016/j.pnucene.2025.106078
Paul Turinsky , Aaron Graham , Benjamin Collins
Nuclear core simulators based upon the nodal diffusion method are currently widely used to predict core behavior. Nodal parameters (e.g., cross-sections) are typically generated utilizing lattice physics codes. In doing so, several approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of core-wide models with prediction fidelity typical of lattice physics to predict nodal parameters would nearly eliminate these approximations. The VERA code can serve as such a core-wide model. By processing VERA predictions for a range of state points, position dependent nodal parameters, functionalized in terms of coolant density, fuel temperature and soluble poison concentration, were obtained and input to the NESTLE nodal code. One challenge of employing a full core model is how to determine nodal parameters’ values for different absorber rod insertion patterns. Accounting for all possible insertion patterns being impractical, an approach based upon absorber rods only effecting nodal parameter values for rodded and adjacent nodes was determined effective, limiting the number of insertion patterns required to build the nodal parameters’ library to three patterns. Simulations for a pressurized water reactor core were completed by VERA and NESTLE to assess differences in reactivity, transient power level versus time, and power distribution. For steady state conditions, absolute differences of less than 1.4 % in reactivity and 5.7 % in pin relative powers were observed for a range of absorber rod insertion patterns, and for the Dropped Bank D transient a core power level difference of 0.17 % after the prompt jump.
{"title":"Generation of nodal neutronic parameters’ library based upon high-fidelity core-wide simulations - Part I: Base model","authors":"Paul Turinsky , Aaron Graham , Benjamin Collins","doi":"10.1016/j.pnucene.2025.106078","DOIUrl":"10.1016/j.pnucene.2025.106078","url":null,"abstract":"<div><div>Nuclear core simulators based upon the nodal diffusion method are currently widely used to predict core behavior. Nodal parameters (e.g., cross-sections) are typically generated utilizing lattice physics codes. In doing so, several approximations are introduced related to using zero current boundary conditions, 2-D radial geometry, and uniform thermal conditions in coolant and fuel. Usage of core-wide models with prediction fidelity typical of lattice physics to predict nodal parameters would nearly eliminate these approximations. The VERA code can serve as such a core-wide model. By processing VERA predictions for a range of state points, position dependent nodal parameters, functionalized in terms of coolant density, fuel temperature and soluble poison concentration, were obtained and input to the NESTLE nodal code. One challenge of employing a full core model is how to determine nodal parameters’ values for different absorber rod insertion patterns. Accounting for all possible insertion patterns being impractical, an approach based upon absorber rods only effecting nodal parameter values for rodded and adjacent nodes was determined effective, limiting the number of insertion patterns required to build the nodal parameters’ library to three patterns. Simulations for a pressurized water reactor core were completed by VERA and NESTLE to assess differences in reactivity, transient power level versus time, and power distribution. For steady state conditions, absolute differences of less than 1.4 % in reactivity and 5.7 % in pin relative powers were observed for a range of absorber rod insertion patterns, and for the Dropped Bank D transient a core power level difference of 0.17 % after the prompt jump.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106078"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145428599","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}
Pub Date : 2026-02-01Epub Date: 2025-11-04DOI: 10.1016/j.pnucene.2025.106109
Qufei Song , Ruixiang Wang , Yuyang Shen , Yiwei Wu , Kuaiyuan Feng , Hui Guo , Yao Xiao , Hanyang Gu
The 3D Method of Characteristics (MOC) offers high numerical accuracy and geometric flexibility, making it a promising approach for neutron transport simulations in Small Modular Reactors (SMRs). However, it is computationally and memory-intensive. This study focuses on GPU-accelerated few-group 3D MOC and methods for reducing memory usage on a CPU/GPU heterogeneous platform. An initial implementation of flat-source GPU-3D-MOC achieved a remarkable speedup of up to 1015.8 compared to single-threaded CPU calculations when calculating a C5G7 assembly but faced memory constraints. To address this, methods including linear source calculation, on-the-fly axial ray tracing, and batch processing of tracks were introduced or proposed, collectively reducing memory usage by 98.5% for a single assembly. This enabled full-core 3D MOC SMR simulations on a single GPU and finally demonstrated speedups of 274.1 and 378.9 in full-core simulations of two SMRs, using 4-group cross-sections generated by the Monte Carlo method. These advancements improve the practicality and flexibility of GPU-3D-MOC for few-group full-core neutron transport in SMR.
三维特征法（MOC）具有较高的数值精度和几何灵活性，是小型模块化反应堆（smr）中子输运模拟的一种很有前途的方法。然而，它是计算和内存密集型的。本研究的重点是GPU加速的少组3D MOC以及在CPU/GPU异构平台上减少内存使用的方法。在计算C5G7组件时，与单线程CPU计算相比，平源GPU-3D-MOC的初始实现实现了高达1015.8倍的显着加速，但面临内存限制。为了解决这个问题，引入或提出了包括线性源计算、动态轴向射线跟踪和轨迹批量处理在内的方法，这些方法共同将单个组件的内存使用量减少了98.5%。这使得在单个GPU上实现了全核3D MOC SMR仿真，并最终使用蒙特卡罗方法生成的4组截面在两个SMR的全核仿真中展示了274.1 x和378.9 x的速度。这些进步提高了GPU-3D-MOC在SMR中用于少群全核中子输运的实用性和灵活性。
{"title":"GPU-accelerated few-group 3D Method of Characteristics for full-core neutron transport in Small Modular Reactors","authors":"Qufei Song , Ruixiang Wang , Yuyang Shen , Yiwei Wu , Kuaiyuan Feng , Hui Guo , Yao Xiao , Hanyang Gu","doi":"10.1016/j.pnucene.2025.106109","DOIUrl":"10.1016/j.pnucene.2025.106109","url":null,"abstract":"<div><div>The 3D Method of Characteristics (MOC) offers high numerical accuracy and geometric flexibility, making it a promising approach for neutron transport simulations in Small Modular Reactors (SMRs). However, it is computationally and memory-intensive. This study focuses on GPU-accelerated few-group 3D MOC and methods for reducing memory usage on a CPU/GPU heterogeneous platform. An initial implementation of flat-source GPU-3D-MOC achieved a remarkable speedup of up to 1015.8<span><math><mo>×</mo></math></span> compared to single-threaded CPU calculations when calculating a C5G7 assembly but faced memory constraints. To address this, methods including linear source calculation, on-the-fly axial ray tracing, and batch processing of tracks were introduced or proposed, collectively reducing memory usage by 98.5% for a single assembly. This enabled full-core 3D MOC SMR simulations on a single GPU and finally demonstrated speedups of 274.1<span><math><mo>×</mo></math></span> and 378.9<span><math><mo>×</mo></math></span> in full-core simulations of two SMRs, using 4-group cross-sections generated by the Monte Carlo method. These advancements improve the practicality and flexibility of GPU-3D-MOC for few-group full-core neutron transport in SMR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106109"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145428665","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}
Pub Date : 2026-02-01Epub Date: 2025-11-14DOI: 10.1016/j.pnucene.2025.106141
Jefferson Q.C. Duarte, Thalles Campagnani, Carlos E. Velasquez, Claubia Pereira, Clarysson A.M. da Silva
This study presents a neutronic analysis of the SEALER (Swedish Advanced Lead Reactor), a fast Small Modular Reactor (SMR), comparing the performance of Uranium Dioxide (UO2), Uranium Mononitride (UN), and Uranium Silicide (U3Si2) fuels under full-power operation throughout the reactor's lifetime. The goal is to evaluate the performance improvements of Advanced Tolerant Fuels (ATF) in a Lead-cooled Fast Reactor (LFR) within the SMR concept. A computational model, developed and verified in previous studies, was used in the analysis. This study focuses on key neutronic parameters, including criticality, conversion ratio (CR), capture-to-fission ratio (α), reproduction factor (η), and isotopic evolution during reactor burnup. The results indicate that, compared to UO2, ATFs have less reactivity swing during the operational reactor cycle and they present better efficiency in fissile material use. Especially U3Si2 demonstrate the best neutron economy and UN has the highest CR. However, in the simulated conditions, ATFs showing smaller efficiency in energy production. All simulations were carried out using the OpenMC Monte Carlo code, with nuclear data libraries processed by NJOY21 at SEALER's operational temperatures. Data analysis and visualization were performed using Matplotlib.
{"title":"Neutronic evaluation of the fuel cycle of SEALER using accident tolerant fuels","authors":"Jefferson Q.C. Duarte, Thalles Campagnani, Carlos E. Velasquez, Claubia Pereira, Clarysson A.M. da Silva","doi":"10.1016/j.pnucene.2025.106141","DOIUrl":"10.1016/j.pnucene.2025.106141","url":null,"abstract":"<div><div>This study presents a neutronic analysis of the SEALER (Swedish Advanced Lead Reactor), a fast Small Modular Reactor (SMR), comparing the performance of Uranium Dioxide (UO<sub>2</sub>), Uranium Mononitride (UN), and Uranium Silicide (U<sub>3</sub>Si<sub>2</sub>) fuels under full-power operation throughout the reactor's lifetime. The goal is to evaluate the performance improvements of Advanced Tolerant Fuels (ATF) in a Lead-cooled Fast Reactor (LFR) within the SMR concept. A computational model, developed and verified in previous studies, was used in the analysis. This study focuses on key neutronic parameters, including criticality, conversion ratio (CR), capture-to-fission ratio (α), reproduction factor (η), and isotopic evolution during reactor burnup. The results indicate that, compared to UO<sub>2</sub>, ATFs have less reactivity swing during the operational reactor cycle and they present better efficiency in fissile material use. Especially U<sub>3</sub>Si<sub>2</sub> demonstrate the best neutron economy and UN has the highest CR. However, in the simulated conditions, ATFs showing smaller efficiency in energy production. All simulations were carried out using the OpenMC Monte Carlo code, with nuclear data libraries processed by NJOY21 at SEALER's operational temperatures. Data analysis and visualization were performed using Matplotlib.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106141"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145518410","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}
Pub Date : 2026-02-01Epub Date: 2025-11-21DOI: 10.1016/j.pnucene.2025.106152
Haoran Huang , Shaoxiong Xia , Songbai Cheng
Motivated to deepen the knowledge on the interaction mechanisms between high-pressure subcooled water jet and high-temperature molten lead-based metal during Steam Generator Tube Rupture (SGTR) accident in Lead-cooled Fast Reactors (LFRs), as well as the resultant pressure build-up characteristics, a numerical investigation is conducted using the MC3D code in this study. Firstly, a 2D numerical model is established based on a previous small-scale experimental study, and the simulation results are validated against the experimental data. Through simulation, good agreement is found between the experiment and simulation results for the non-explosion cases, which shows the capability of MC3D for modeling the Coolant-Coolant Interaction (CCI) phenomenon during a LFR SGTR accident. Thereafter, to extend the application scope of the MC3D code, another 2D numerical model is built according to the design specifications of the heat exchange tube in an experimental LFR using Lead-Bismuth Eutectic (LBE) as primary coolant. A sensitivity analysis is performed, and the effects of LBE temperature, water temperature, water pressure, rupture diameter, as well as rupture type, on pressure build-up characteristics are discussed. The simulation results reveal that all the parameters have a promoting effect on pressure build-up characteristics, among which the effect of water temperature is much more remarkable than that of LBE temperature. Besides, as a guillotine rupture occurs in the SGTR accident, void spots with low water volume fraction are observed around the two tube walls, and the pressure trend in the single-wall guillotine rupture case is highly similar to the one in the double-wall guillotine rupture case. Nevertheless, more experimental data are needed for the further validation of the MC3D code, and improvement as well as development of the current calculation models are also highly anticipated in the future to enhance the overall reliability of MC3D.
{"title":"Numerical simulation of coolant-coolant interaction during SGTR accident in lead-cooled fast reactors","authors":"Haoran Huang , Shaoxiong Xia , Songbai Cheng","doi":"10.1016/j.pnucene.2025.106152","DOIUrl":"10.1016/j.pnucene.2025.106152","url":null,"abstract":"<div><div>Motivated to deepen the knowledge on the interaction mechanisms between high-pressure subcooled water jet and high-temperature molten lead-based metal during Steam Generator Tube Rupture (SGTR) accident in Lead-cooled Fast Reactors (LFRs), as well as the resultant pressure build-up characteristics, a numerical investigation is conducted using the MC3D code in this study. Firstly, a 2D numerical model is established based on a previous small-scale experimental study, and the simulation results are validated against the experimental data. Through simulation, good agreement is found between the experiment and simulation results for the non-explosion cases, which shows the capability of MC3D for modeling the Coolant-Coolant Interaction (CCI) phenomenon during a LFR SGTR accident. Thereafter, to extend the application scope of the MC3D code, another 2D numerical model is built according to the design specifications of the heat exchange tube in an experimental LFR using Lead-Bismuth Eutectic (LBE) as primary coolant. A sensitivity analysis is performed, and the effects of LBE temperature, water temperature, water pressure, rupture diameter, as well as rupture type, on pressure build-up characteristics are discussed. The simulation results reveal that all the parameters have a promoting effect on pressure build-up characteristics, among which the effect of water temperature is much more remarkable than that of LBE temperature. Besides, as a guillotine rupture occurs in the SGTR accident, void spots with low water volume fraction are observed around the two tube walls, and the pressure trend in the single-wall guillotine rupture case is highly similar to the one in the double-wall guillotine rupture case. Nevertheless, more experimental data are needed for the further validation of the MC3D code, and improvement as well as development of the current calculation models are also highly anticipated in the future to enhance the overall reliability of MC3D.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106152"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145569689","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}
Pub Date : 2026-02-01Epub Date: 2025-12-10DOI: 10.1016/j.pnucene.2025.106186
A. Kaya , O. Erbay , Z. Boduroglu , I.A. Reyhancan , M.S. Kiziltas , T. Akyurek
This study presents a comprehensive burnup analysis of all fuel elements in the ITU TRIGA Mark II research reactor core using non-destructive assay (NDA) techniques based on gamma spectroscopy. Two distinct fuel inspection systems were employed to measure the gamma activity, with 137Cs used as the primary burnup indicator due to its strong correlation with fuel depletion. The results show a clear burnup pattern, with higher values in the inner core rings that gradually decreases toward the outer rings. This asymmetric burnup distribution underscores the need for reactor core reconfiguration, for which an optimized layout is proposed. Additionally, gamma spectroscopic analysis of fuel element F30 was conducted at cooling times of 10, 130, and 755 days. The results also suggest that 140La is a promising candidate for future burnup studies, which will be pursued in subsequent research.
本研究采用基于伽马能谱的无损分析（NDA）技术，对国际电联TRIGA Mark II研究堆堆芯中的所有燃料元件进行了全面的燃耗分析。两种不同的燃料检测系统被用来测量伽马活度，137Cs被用作主要燃耗指标，因为它与燃料消耗有很强的相关性。结果显示出一个清晰的燃耗模式，内核环的燃耗值较高，向外环逐渐降低。这种不对称燃耗分布强调了反应堆堆芯重构的必要性，为此提出了一种优化布局。此外，在冷却时间为10、130和755天时，对燃料元件F30进行了伽马光谱分析。结果还表明140La是未来燃耗研究的一个有希望的候选者，将在后续的研究中进行。
{"title":"Nondestructive burnup evaluation and gamma spectroscopy analysis of spent fuel elements in the ITU TRIGA Mark II research reactor","authors":"A. Kaya , O. Erbay , Z. Boduroglu , I.A. Reyhancan , M.S. Kiziltas , T. Akyurek","doi":"10.1016/j.pnucene.2025.106186","DOIUrl":"10.1016/j.pnucene.2025.106186","url":null,"abstract":"<div><div>This study presents a comprehensive burnup analysis of all fuel elements in the ITU TRIGA Mark II research reactor core using non-destructive assay (NDA) techniques based on gamma spectroscopy. Two distinct fuel inspection systems were employed to measure the gamma activity, with <sup>137</sup>Cs used as the primary burnup indicator due to its strong correlation with fuel depletion. The results show a clear burnup pattern, with higher values in the inner core rings that gradually decreases toward the outer rings. This asymmetric burnup distribution underscores the need for reactor core reconfiguration, for which an optimized layout is proposed. Additionally, gamma spectroscopic analysis of fuel element F30 was conducted at cooling times of 10, 130, and 755 days. The results also suggest that <sup>140</sup>La is a promising candidate for future burnup studies, which will be pursued in subsequent research.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106186"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145736226","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}
Pub Date : 2026-02-01Epub Date: 2025-12-06DOI: 10.1016/j.pnucene.2025.106178
Shanfang Huang , Guodong Liu , Ying He , Xingyu Zhao , Zhigang Li , Yugao Ma , Meiming Qiu , Yongtang Wang , Haisheng Chen , Junren Hou , Kaiwen Li , Hao Luo , Chuan Lu , Kan Wang
The operation of nuclear reactors involves a complex interplay among multiple physical processes. Accurate reactor simulation requires the implementation of multiphysics coupling. Neutronics holds a particularly distinctive role in simulation, and can be solved via deterministic or Monte Carlo (MC) methods. Although research on deterministic-based coupling started earlier, MC-based coupling methods have been increasingly applied in recent years for both analyses of conventional reactors and designs of advanced reactors, due to their high fidelity. In view of this, this paper provides a comprehensive review of MC-based multiphysics coupling in nuclear reactors. Multiphysics coupling issues in representative reactors such as light water reactors, heat pipe-cooled reactors, and high-temperature gas-cooled reactors are systematically presented. Furthermore, multiphysics coupling technologies including cross-section treatments, coupling methods, iteration strategies, data mapping, data transfer, and convergence criteria are extensively discussed. Challenges and future exploration are also enumerated. This review indicates that while MC-based coupling demonstrates superior aptitude for simulation of advanced reactors, current research remains predominantly focused on methodological studies. Leveraging high-performance computing resources and artificial intelligence, the development of advanced algorithms such as tight, transient and full-loop coupling will accelerate the deployment. This comprehensive analysis thereby establishes the review as both a guide for newcomers and an up-to-date reference for researchers.
{"title":"A review of Monte Carlo-based multiphysics coupling in nuclear reactor analysis","authors":"Shanfang Huang , Guodong Liu , Ying He , Xingyu Zhao , Zhigang Li , Yugao Ma , Meiming Qiu , Yongtang Wang , Haisheng Chen , Junren Hou , Kaiwen Li , Hao Luo , Chuan Lu , Kan Wang","doi":"10.1016/j.pnucene.2025.106178","DOIUrl":"10.1016/j.pnucene.2025.106178","url":null,"abstract":"<div><div>The operation of nuclear reactors involves a complex interplay among multiple physical processes. Accurate reactor simulation requires the implementation of multiphysics coupling. Neutronics holds a particularly distinctive role in simulation, and can be solved via deterministic or Monte Carlo (MC) methods. Although research on deterministic-based coupling started earlier, MC-based coupling methods have been increasingly applied in recent years for both analyses of conventional reactors and designs of advanced reactors, due to their high fidelity. In view of this, this paper provides a comprehensive review of MC-based multiphysics coupling in nuclear reactors. Multiphysics coupling issues in representative reactors such as light water reactors, heat pipe-cooled reactors, and high-temperature gas-cooled reactors are systematically presented. Furthermore, multiphysics coupling technologies including cross-section treatments, coupling methods, iteration strategies, data mapping, data transfer, and convergence criteria are extensively discussed. Challenges and future exploration are also enumerated. This review indicates that while MC-based coupling demonstrates superior aptitude for simulation of advanced reactors, current research remains predominantly focused on methodological studies. Leveraging high-performance computing resources and artificial intelligence, the development of advanced algorithms such as tight, transient and full-loop coupling will accelerate the deployment. This comprehensive analysis thereby establishes the review as both a guide for newcomers and an up-to-date reference for researchers.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106178"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145681133","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}
Pub Date : 2026-02-01Epub Date: 2025-11-05DOI: 10.1016/j.pnucene.2025.106117
Xiangming Sun , Keng Yeow Chung , Jia Hao Tang , Beng Thye Tan , Isaac Yap , Vitesh Krishnan , Tu Guang Tan
This study proposes a risk map approach to evaluate radiation exposure risks in Singapore resulting from long-range atmospheric dispersion of Cs-137 in Southeast Asia. A community, freely available dispersion model is employed to derive plume arrival times and ground depositions in Singapore, under a typical weather pattern identified during the Northeast Monsoon season. The risk assessment, along with their associated uncertainties, are integrated and visualised on a geographical map of Southeast Asia. This framework provides a reliable reference for the rapid assessment of radiation exposure risks in a receptor city. Practical dose assessment can be easily derived mathematically for multiple-radionuclide source terms using the simulation results of Cs-137. Leveraging publicly available global wind data with good temporal coverage ensures the ready applicability of the approach to other regions or seasons.
{"title":"Risk map of long-range atmospheric dispersion of radionuclides in Southeast Asia under a typical wind pattern of the Northeast Monsoon season","authors":"Xiangming Sun , Keng Yeow Chung , Jia Hao Tang , Beng Thye Tan , Isaac Yap , Vitesh Krishnan , Tu Guang Tan","doi":"10.1016/j.pnucene.2025.106117","DOIUrl":"10.1016/j.pnucene.2025.106117","url":null,"abstract":"<div><div>This study proposes a risk map approach to evaluate radiation exposure risks in Singapore resulting from long-range atmospheric dispersion of Cs-137 in Southeast Asia. A community, freely available dispersion model is employed to derive plume arrival times and ground depositions in Singapore, under a typical weather pattern identified during the Northeast Monsoon season. The risk assessment, along with their associated uncertainties, are integrated and visualised on a geographical map of Southeast Asia. This framework provides a reliable reference for the rapid assessment of radiation exposure risks in a receptor city. Practical dose assessment can be easily derived mathematically for multiple-radionuclide source terms using the simulation results of Cs-137. Leveraging publicly available global wind data with good temporal coverage ensures the ready applicability of the approach to other regions or seasons.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106117"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145468929","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}
Pub Date : 2026-02-01Epub Date: 2025-11-06DOI: 10.1016/j.pnucene.2025.106120
Shusong Qin , Weijie Cai , Hao Yang , Shenzhou Du , Xiangfei Meng , Jianchuang Sun , Wenchao Zhang , Weihua Cai
As a new type of fuel element, the four-petal fuel rods (FPFR) are suitable for small-scale nuclear reactors. This study is based on ABAQUS-STAR CCM + different coupling modes, to investigate the 3 × 3 assemblies flow heat transfer and irradiation mechanics coupling behaviors under power-up, irradiation contact, and reactivity insertion accident (RIA) conditions. The results indicate that the different twist petal structures, flow areas, and cosine heat release mode along the axial direction led to complex flow and heat transfer behaviors along the radial direction. Different from the power-up stage, the heat transfer coefficient (HTC) in the irradiation stage decreases slightly, and irradiation swelling increases the tendency of the fuel to expand outward, leading to a significant stress-strain increase in the contact regions. The peak stress reaches 383.143 MPa, and the deformation of the contact regions under mutual extrusion is also different. During the RIA stage, the plastic strain in the contact regions reaches 0.006, which cannot return to the level before the accident even after the power decreases.
{"title":"Study on the multi-physics field coupling behavior of four-petal fuel rods contact under normal and RIA conditions","authors":"Shusong Qin , Weijie Cai , Hao Yang , Shenzhou Du , Xiangfei Meng , Jianchuang Sun , Wenchao Zhang , Weihua Cai","doi":"10.1016/j.pnucene.2025.106120","DOIUrl":"10.1016/j.pnucene.2025.106120","url":null,"abstract":"<div><div>As a new type of fuel element, the four-petal fuel rods (FPFR) are suitable for small-scale nuclear reactors. This study is based on ABAQUS-STAR CCM + different coupling modes, to investigate the 3 × 3 assemblies flow heat transfer and irradiation mechanics coupling behaviors under power-up, irradiation contact, and reactivity insertion accident (RIA) conditions. The results indicate that the different twist petal structures, flow areas, and cosine heat release mode along the axial direction led to complex flow and heat transfer behaviors along the radial direction. Different from the power-up stage, the heat transfer coefficient (HTC) in the irradiation stage decreases slightly, and irradiation swelling increases the tendency of the fuel to expand outward, leading to a significant stress-strain increase in the contact regions. The peak stress reaches 383.143 MPa, and the deformation of the contact regions under mutual extrusion is also different. During the RIA stage, the plastic strain in the contact regions reaches 0.006, which cannot return to the level before the accident even after the power decreases.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"192 ","pages":"Article 106120"},"PeriodicalIF":3.2,"publicationDate":"2026-02-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145468931","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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