Nuclear Engineering and Design最新文献

{"title":"Effect of spray flow rate on pressure and temperature distribution in SMR containment","authors":"Jinglin Cao,&nbsp;Xuefeng Lyu,&nbsp;Fenglei Niu,&nbsp;Jialei Chen","doi":"10.1016/j.nucengdes.2025.114005","DOIUrl":"10.1016/j.nucengdes.2025.114005","url":null,"abstract":"<div><div>In the case of a direct vessel injection (DVI) line break in small modular reactors, the key to ensuring reactor safety is a timely and effective suppression of the pressure and temperature rise in the containment. In this paper, GASFLOW was utilized to analyze the influence of the internal spray mass flow rate on the pressure suppression and to compare 3D temperature distribution in the containment following the double-ended DVI line rupture. Results showed that the spray system significantly reduced the pressure and temperature in the containment in the initial phases. As the accident progressed, the impact of varying spray flow rates on the containment pressure and temperature gradually diminished, temperature distribution became more uniform, and the condensation effect of the spray ultimately stabilized. These findings substantiate the efficacy of the spray system and reveal a positive correlation between spray flow rates and more evident pressure suppression and cooling effects.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114005"},"PeriodicalIF":1.9,"publicationDate":"2025-03-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682449","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}

{"title":"Uncertainty-aware prediction of Peak Cladding Temperature during extended station blackout using Transformer-based machine learning","authors":"Tran C.H. Nguyen ,&nbsp;Aya Diab","doi":"10.1016/j.nucengdes.2025.113984","DOIUrl":"10.1016/j.nucengdes.2025.113984","url":null,"abstract":"<div><div>Accurate prediction of the Peak Clad Temperature (PCT) may be used to evaluate the efficacy of operator mitigation actions during extended Station Blackout (SBO) scenarios. In this study, we propose a two-stage machine learning (ML) framework that integrates classification and regression to forecast PCT. While the classification stage identifies whether mitigation efforts succeed or fail, the regression stage provides precise multi-step PCT predictions. Our framework leverages advanced ML models, including Transformer architectures, Attention mechanism, and Long Short-Term Memory (LSTM) networks, alongside the Best Estimate Plus Uncertainty (BEPU) approach. To account for the underlying uncertainty and generate confidence intervals, we incorporate Monte Carlo (MC) Dropout. By integrating BEPU with machine learning and uncertainty quantification, our model produces reliable temperature forecasts despite the system’s inherent complexity and nonlinearity with R² values exceeding 0.98 for 60-, 120-, and 240-step time frames. Notably, the LSTM-Transformer model proves to be the most effective, even for longer prediction horizons. The developed framework serves as a powerful real-time decision support tool for operators, for accurate prediction and effective mitigation of critical conditions like extended SBO events.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 113984"},"PeriodicalIF":1.9,"publicationDate":"2025-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682373","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}

{"title":"Automating equipment identification in nuclear engineering drawings","authors":"Issam Hammad ,&nbsp;Mishca de Costa ,&nbsp;Ameneh Boroomand ,&nbsp;Muhammad Anwar","doi":"10.1016/j.nucengdes.2025.114002","DOIUrl":"10.1016/j.nucengdes.2025.114002","url":null,"abstract":"<div><div>Engineering drawings are critical assets in the nuclear industry, essential for the design, construction, and maintenance of facilities like the Darlington Nuclear Generating Station (DNGS). Manual processes for identifying equipment within these drawings are time-consuming and error-prone, affecting operational efficiency and safety compliance. This paper presents design methodologies to build an Intelligent Drawing Query (IDQ) system, leveraging Cloud Base Artificial Intelligence (AI) including Optical Character Recognition (OCR) technologies to automate equipment identification of tags within nuclear engineering drawings. The paper evaluates and compares the extraction efficiency of cloud-based OCR services including Microsoft’s Azure OCR and Azure Document Intelligence (DI). Additionally, the paper explores best practices to maximize the extraction efficiency. Moreover, the paper explores the potential of OpenAI’s multimodal GPT-4 model for additional detection tasks. Such automation reduces human error, enhances workflows, and ensures compliance with safety and regulatory standards.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 114002"},"PeriodicalIF":1.9,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143683353","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Numerical analysis on restoring stable status in natural circulation loop","authors":"Yao Yao ,&nbsp;Tao Zhou ,&nbsp;Dongli Huang ,&nbsp;Jianyu Tang ,&nbsp;Zefeng Wang ,&nbsp;Shilei Dun","doi":"10.1016/j.nucengdes.2025.114001","DOIUrl":"10.1016/j.nucengdes.2025.114001","url":null,"abstract":"<div><div>Natural circulation flow instability is a common phenomenon in nuclear reactor systems, especially in components such as passive safety systems, reactor vessel downcomers, and steam generators. In general, this kind of instability is undesirable as it can jeopardize nuclear system safety, leading to fatigue damage, problems of system control, and heat transfer deterioration. It is very crucial to evaluate impact factors of restoring the stable status of natural circulation since reducing the duration of instability or restoring the stable status at the early stage of instability will prevent reactor systems from potential failures and risks. Despite the significance of system stability, the majority of the literature has focused on different impact factors of instability onset, while few has discussed the restoration conditions. This manuscript investigates conditions to restore stable status of natural circulation, including various axial power factor distributions, multiple parallel channel types, and inlet subcooling. The working condition of the studied natural circulation loop is under 10 MPa. Insights and suggestions are provided in this manuscript on enhancing the safety and reliability of nuclear reactors by optimizing operating conditions and designs for two-phase natural circulation systems. Numerical analysis employs the RELAP5 system code to model natural circulation loop with two types of channels, one with a single channel and the other with parallel channels, based on a well-validated natural circulation test facility. Criteria of instability onset and restoring stable status are defined by the amplitude and period of mass flow rate. Key responses, including mass flow rate, time of restoring stable status, duration of instability, and flow regime are examined. Results indicate that the uniform power distribution in parallel channels with high inlet subcooling will postpone the instability onset and shorten the duration of instability, with which condition will effectively help loop to restore stable status.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 114001"},"PeriodicalIF":1.9,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143683354","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}

{"title":"Modeling of the irradiation-induced multi-scale mechanical behaviors for surrogate FCM pellets with interfacial cracking","authors":"Zekun Li ,&nbsp;Jing Zhang ,&nbsp;Feng Yan ,&nbsp;Shurong Ding ,&nbsp;Qisen Ren","doi":"10.1016/j.nucengdes.2025.113999","DOIUrl":"10.1016/j.nucengdes.2025.113999","url":null,"abstract":"<div><div>The simulation method and analysis code to investigate the irradiation-induced thermo-mechanical behaviors of surrogate FCM pellets are established, incorporating the cohesive model at the interface of the surrogate kernel with the buffer layer. The innovative volume growth strain model is adopted to correlate the anisotropic shrinkage and creep deformations of the solid skeleton with the macroscale volumetric growth of the buffer layer under external hydrostatic pressures. The predictions of the pellet swelling and the microstructure information agree well with the experimental results, validating the developed models and simulation strategy. It is indicate that: (1) a gap with a width of ∼22.98 μm is generated at the fast neutron fluence of 7.50 × 10²⁵ n/m²; (2) the predicted maximum tensile stress of ∼1894 MPa for the SiC layer implies that its tensile strength is particularly high; the tensile strength of the SiC matrix might exceed ∼289 MPa; (3) the volumetric swelling of the surrogate TRISO particle is mainly contributed by the outward displacements of the buffer layer after interfacial cracking; (4) without considering the anisotropic skeleton creep contribution on the macroscale volumetric growth of the buffer layer, the peak shrinkage strain of the buffer layer could be twice higher due to the enhanced hydrostatic pressure, accompanied by the reduced current porosity and the enlarged gap width; the maximum skeleton tensile stress will increase by ∼60.37 %. This study offers insights into the irradiation-induced thermo-mechanical behaviors of surrogate FCM pellets, supplying a foundation for further research on FCM fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113999"},"PeriodicalIF":1.9,"publicationDate":"2025-03-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143683352","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}

{"title":"Prediction of steam generator liquid level under main steam line break accident based on wavelet decomposition combined with deep learning","authors":"Biaoxin Wang,&nbsp;Yuang Jiang,&nbsp;Mei Lin,&nbsp;Qiuwang Wang","doi":"10.1016/j.nucengdes.2025.113998","DOIUrl":"10.1016/j.nucengdes.2025.113998","url":null,"abstract":"<div><div>Liquid level monitoring is essential for maintaining the safe operation of nuclear power circuits. During a Main Steam Line Break (MSLB) accident, significant fluctuations in the liquid level within the steam generator pose challenges for traditional measurement methods, which often fail to accurately capture the true liquid level. This study conducted experiments of MSLB accidents under controlled conditions, with parameters including heating power ranging from 8 to 16 kW, break pressures from 0.05 to 0.1 MPa, and relative break sizes between 20 % and 100 %. In selected conditions, rolling motions were introduced to simulate marine environments. Wavelet decomposition was utilized to extract features at varying frequency levels, and deep learning models were employed to predict each component. The proposed approach achieved a prediction accuracy of 88.3 %, outperforming direct predictions from raw data with improvements of 21.9 % in Mean Squared Error (<em>MSE</em>), 12.3 % in Mean Absolute Error (<em>MAE</em>), and 10.0 % in the coefficient of determination (<em>R</em>²). The detail component cD1 was found to have the most significant impact on overall prediction accuracy, highlighting it as a key parameter for further optimization. Furthermore, the use of wavelet-decomposed data significantly reduced computational complexity, enhancing time efficiency. These results demonstrate the effectiveness of the proposed method in improving prediction accuracy and operational efficiency, offering valuable support for the safe management of nuclear power systems during MSLB accidents.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113998"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621076","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}

{"title":"High-temperature oxidation of Zr-4 and Zr-1Nb-O alloys: Influencing factors, oxidation behaviors and mechanisms","authors":"Huanteng Liu ,&nbsp;Donghai Xu ,&nbsp;Guanyu Jiang ,&nbsp;Xueling Fan ,&nbsp;Guangyi Liu","doi":"10.1016/j.nucengdes.2025.113979","DOIUrl":"10.1016/j.nucengdes.2025.113979","url":null,"abstract":"<div><div>Under loss-of-coolant accidents, the oxidation rate of zircaloy rapidly increases, which can cause cladding failure and pose serious safety risks. Thus, enhancing the oxidation resistance of zircaloy is of utmost importance to ensure the safe utilization of a nuclear power. This work provides a comprehensive review on influencing factors and mechanisms for high-temperature oxidation performance of Zr-4 and Zr-1Nb-O alloys. These factors mainly include alloying composition, oxidizing atmosphere, oxidation temperature, pre-oxidation, irradiation, and hydrogen absorption. Oxidation kinetics, behavior, and mechanisms in steam, O₂ and air are thoroughly discussed, and a comparative analysis of oxidation kinetics is presented. Overall, the addition of Nb enhances the oxidation resistance of zircaloy. In air, the oxidation rate of zircaloy is notably faster compared with that in steam and O₂ environments due to the formation of ZrN. At elevated temperatures, the critical size of zirconia increases, leading to a phase transition and a reduction in the volume fraction of monoclinic zirconia. The phase transition makes the zirconia oxide layer crack and less stable. Pre-oxidation at low temperatures in O₂ or steam significantly improves the oxidation resistance of samples. The formation of oxides during the oxidation process of zircaloy is controlled by O^2– diffusion. The breakaway oxidation of zircaloys occurs as a result of the transformation from tetragonal to monoclinic phase of zirconia, as well as stress relaxation of oxides and evolution of oxide morphology when reaching critical thickness.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113979"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621077","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}

{"title":"Study on off-design performance of a hydrogen and electricity joint production system utilizing a very-high-temperature reactor","authors":"Hang Ni ,&nbsp;Xinhe Qu ,&nbsp;Ping Zhang ,&nbsp;Ekaterina Sokolova ,&nbsp;Han Zhang ,&nbsp;Khashayar Sadeghi ,&nbsp;Wei Peng","doi":"10.1016/j.nucengdes.2025.113987","DOIUrl":"10.1016/j.nucengdes.2025.113987","url":null,"abstract":"<div><div>This study introduces a hydrogen and electricity joint production system that utilizes a very-high-temperature reactor and integrates the iodine–sulfur cycle with the steam Rankine cycle. The system can operate under off-design conditions by adjusting the helium mass flow rates in the main and secondary loops, using constant pressure operation (CPO) or sliding pressure operation (SPO) for the power generation loop. The system’s performance is investigated at partial reactor thermal power. The recommended reactor thermal power load ratios range from 63.90 % to 100 % for CPO and from 64.29 % to 100 % for SPO, with the lower limit determined by the steam generator’s hot-end temperature difference. The hydrogen production rate and the system’s electrical power output both decline with a lower load ratio. Within the recommended load ratios, the hydrogen production rates for CPO and SPO range from 129.44 mol/s to 200 mol/s. With a lower load ratio, the power generation efficiency declines, while the hydrogen-electricity efficiency and system’s exergy efficiency first rise and then fall. At a fixed load ratio, the power generation efficiency, hydrogen-electricity efficiency, and system’s exergy efficiency are higher using SPO than those using CPO, indicating better off-design performance using SPO.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113987"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143628098","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}

{"title":"A Non-Intrusive Reduced-Order model for rapid analysis of thermal stratification in pressurizer surge line","authors":"Chuzhen Peng,&nbsp;Han Zhang,&nbsp;Yongwang Ding,&nbsp;Lixun Liu,&nbsp;Yingjie Wu,&nbsp;Jiong Guo,&nbsp;Fu Li","doi":"10.1016/j.nucengdes.2025.113996","DOIUrl":"10.1016/j.nucengdes.2025.113996","url":null,"abstract":"<div><div>The pressurizer surge line serves to connect the pressurizer and primary circuit in a PWR (Pressurized Water Reactor) system. However, thermal stratification at its junction can induce distortion and stress, potentially damaging the pipes. Computational Fluid Dynamics (CFD) is a common numerical tool, but its time-intensive nature poses challenges for real-time assessment, especially with multiple parameter variations. To address this issue, we developed a rapid analysis method using a non-intrusive reduced-order model. The experimental design is optimized by incorporating the Generalized Subset Design to minimize sample requirements. The reduced-order model of the temperature field was derived using Proper Orthogonal Decomposition. Off-design cases were predicted using Linear, Radial Basis Function, and Radial Basis Function Neural Network interpolation techniques. The resulting temperature field was utilized for stress analysis in the pipe structure. Results indicate that linear interpolation performs best, with a maximum CvRMSE (Coefficient of Variation of the Root Mean Square Error) of 0.038 for temperature and a maximum RMSE(Root Mean Square Error) of −0.02% in predicting the maximum equivalent stress. The Radial Basis Function interpolation is slightly inferior to linear interpolation. It better fits the thermal stratification region but lacks accuracy in identifying its boundaries. This inaccuracy is more sensitive to equivalent stress, resulting in a maximum stress deviation of −0.08% for sharp boundaries. Additionally, the Radial Basis Function Neural Network is unsuitable for current study due to insufficient sample size, resulting in a maximum stress identification deviation of −3.8%. Finally, the POD coefficient is used as a independent variable to interpolate the maximum Von-Mises stress, and the relative errors were controlled within 5%. This study provides a rapid and accurate method to evaluate the temperature distributions and the maximum stresses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113996"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621075","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}

{"title":"Investigating the possibility of using iridium as a burnable absorber in new fuel pellet designs of VVER-1200 for reactivity management","authors":"Sultan J. Alsufyani ,&nbsp;Nassar Alnassar ,&nbsp;Mohammed Sallah ,&nbsp;Mohamed A.E. Abdel-Rahman ,&nbsp;Naima Amrani ,&nbsp;A. Abdelghafar Galahom","doi":"10.1016/j.nucengdes.2025.113989","DOIUrl":"10.1016/j.nucengdes.2025.113989","url":null,"abstract":"<div><div>Searching for the optimal design of the fuel assembly and the material needed to manage the reactivity in the nuclear reactor is still vital. Therefore, new geometry configurations and materials have been investigated in this work to handle the excess reactivity. Two designs of burnable absorber fuel pellets concentric shell model (CSM) and outer shell model (OSM) have been proposed for VVER-1200 assembly. Four BA materials including Gd₂O₃, Eu₂O₃, Er₂O₃ and Ir₂O₃ have been suggested to be investigated in the proposed fuel pellet designs. Various dimensions and concentrations of the suggested BAs have been studied in the OSM and CSM to obtain the optimum model. From both designs, the optimal models from k_inf behavior point of view have been selected and integral studies have been done on them. The neutronic analysis confirms the effectiveness of using erbium and iridium in the suggested models in managing the excess reactivity of VVER-1200.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"436 ","pages":"Article 113989"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610019","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}