Nuclear Engineering and Design最新文献

{"title":"Impact of γ irradiation and thermal aging on the swelling pressure of GMZ bentonite","authors":"Wei Liu ,&nbsp;Dong Liang ,&nbsp;Zhongtian Yang ,&nbsp;Chao Gao ,&nbsp;Jingli Xie ,&nbsp;Meilan Jia","doi":"10.1016/j.nucengdes.2025.114012","DOIUrl":"10.1016/j.nucengdes.2025.114012","url":null,"abstract":"<div><div>To investigate the combined effects of γ irradiation and thermal aging on the swelling pressure of buffer materials used in the geological disposal of high-level radioactive waste, Gaomiaozi (GMZ) bentonite from Inner Mongolia was herein selected. Samples were irradiated by γ rays and exposed to a combination of γ irradiation followed by thermal aging at 90℃. Swelling pressure tests were then conducted. Results showed that irradiation and thermal aging reduced the swelling pressure of GMZ bentonite. Notably, after irradiating and storing at (25 ± 3)℃ for 360 d, the swelling pressure increased by 51.59 % compared with that of the γ-irradiated sample; the swelling pressure decreased by 75.9 % for the γ irradiation–thermal sequentially aged sample (heated for 180 d). Partial chemical-bond destruction, the transformation of structure trivalent to divalent iron, and the water redistribution in the bentonite were the primary drivers of the decrease in the swelling pressure of GMZ bentonite.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114012"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706418","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":"Assessment and predicting the axial power distribution effect on the thermal-mechanical parameters of the NuScale nuclear reactor core loaded with TVS-2 M fuel assemblies as well as axial Offset optimizing for load-following operation","authors":"M.H. Zahedi yeganeh,&nbsp;G.R. Ansarifar,&nbsp;H.Zayermohammadi Rishehri","doi":"10.1016/j.nucengdes.2025.114009","DOIUrl":"10.1016/j.nucengdes.2025.114009","url":null,"abstract":"<div><div>This study evaluates and examines the thermal–mechanical behavior of a NuScale reactor core which utilizes TVS-2 M hexagonal fuel assemblies. The efficiency of the fuel rods is validated using the FRAPCON code. Initially, the reactor’s core is modeled with the MCNP code to locate the control banks. The design phase ensures the capability to shut down the reactor in two scenarios. In the Hot Zero Power (HZP) scenario, MCNP simulation reveals a sub-critical state with a multiplication factor of 0.94481 ± 0.00023. In the Cold Zero Power (CZP) scenario, the multiplication factor of 0.9935 ± 0.00023 confirms the adequacy of control assemblies. Subsequently, a thermal–mechanical analysis is conducted on the fuel rod over 1330 days, confirming its acceptable design and operational effectiveness in the core. Also, one of the parameters that can be examined during reactor control and load-following operations is Axial Offset (AO). Therefore, the study investigates the impact of AO on fuel rod’s thermal–mechanical changes. The MCNP code was used to simulate control rod inputs and obtain power distribution data for each AO deviation. Based on assessments regarding the association between AO and the thermal–mechanical characteristics of fuel, it has been determined that the impact of power distribution increases significantly over time, particularly towards the end of the operational period. Afterward, based on FRAPCON results, an artificial neural network (ANN) estimator is developed to predict thermal–mechanical parameters at the beginning of the cycle (BOC). The ANN proves to be a powerful method for estimation. By employing the ANN estimator and exploring different cost functions based on thermal–mechanical parameters, the optimal AO is determined using a genetic algorithm, which enhances the reactor’s performance, particularly in load-following operations. The attained optimal AO value for various cost functions are as follows: −0.10316, −0.19635, and −0.25817. This approach allows for the selection of the most efficient AO, leading to improved performance of the NuScale reactor core loaded with TVS-2 M hexagonal fuel assemblies. Indeed, optimization of AO is very important and useful for load-following operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114009"},"PeriodicalIF":1.9,"publicationDate":"2025-03-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143706419","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":"Analyses of the MELCOR capability to simulate integral PWR using passive systems in a DBA scenario","authors":"M. Principato ,&nbsp;F. Giannetti ,&nbsp;M. Imperatori ,&nbsp;M. D’Onorio ,&nbsp;M. Garcia ,&nbsp;L.E. Herranz ,&nbsp;A. Bersano ,&nbsp;F. Mascari","doi":"10.1016/j.nucengdes.2025.114004","DOIUrl":"10.1016/j.nucengdes.2025.114004","url":null,"abstract":"<div><div>This paper is focused on a thermal–hydraulic transient analysis of a generic 300 MWe integral Pressurized Water Reactor (iPWR) based on a passive safety mitigation strategy and dry containment. The main target of the paper is to analyze the MELCOR code’s capability to simulate the operation of the passive systems in Small Modular Reactor (SMR) configuration and the consequent plant behavior. The nodalization and the modeling approach used in the MELCOR code have been described in the present work, and a double-ended rupture of the Direct Vessel Injection (DVI) line has been postulated as initiating event.</div><div>The results of the analysis demonstrated that the MELCOR code can qualitatively replicate the dominant phenomena that drive the passive mitigation strategy’s operation. Additionally, the core reflooding was made possible by the operation of the available safety systems, which was adequate to prevent severe accident conditions during the entire simulated transient. The activity has been developed in the framework of SASPAM-SA Horizon Euratom project as the base for further derivative activities focused on assessing the capability of the MELCOR code to simulate the phenomena taking place in postulated plausible severe accident scenarios in iPWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114004"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682450","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":"Justification of RCCAs lifetime extension at operating Ukrainian NPPs. Summary calculations","authors":"Valeriy Zuyok ,&nbsp;Oleksandr Mazurok ,&nbsp;Oleg Godun ,&nbsp;Mykola Chaikovskyi ,&nbsp;Anton Makarenko ,&nbsp;Vadym Ivanov ,&nbsp;Valodymyr Zigunov ,&nbsp;Mykhaylo Tretyakov","doi":"10.1016/j.nucengdes.2025.114008","DOIUrl":"10.1016/j.nucengdes.2025.114008","url":null,"abstract":"<div><div>One of the ways to temporarily meet the needs of NPPs for rod cluster control assemblies (RCCAs) is to extend their lifetime. The justification for extending the lifetime of RCCAs should be based on the justification of mechanical reliability of control rods (CRs) and RCCAs in general, fulfillment of the criteria for physical efficiency of RCCAs, thermotechnical characteristics of RCCAs, taking into account changes in materials characteristics during the entire previous period of operation in the reactor core.</div><div>In the course of the research, the operational data (RCCAs positions and power unit capacity) for each of the 61 RCCAs of all 13 VVER-1000 power units for the last almost 10 years were analyzed and systematized. The maximum fast neutron flux and ¹⁰B burnup were calculated for each RCCA, and the margin before reaching the limit values was also calculated.</div><div>The results of the performed research enabled to extend the lifetime of almost all RCCAs in VVER-1000 reactors of Ukrainian NPPs from 25,500 to 38,000 h in the automatic control group and from 75,600 to 113,500 h in the reactor scram group, except for twelve RCCAs from RNPP Unit 4 and six RCCAs from SUNPP Unit 1. An individual (separate) assessment of the lifetime was performed for these RCCAs, throughout which the operational integrity criteria for the maximum value of fast neutron flux in the lower part of CR cladding and ¹⁰B burnup will be met.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114008"},"PeriodicalIF":1.9,"publicationDate":"2025-03-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682451","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":"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}