Progress in Nuclear Energy最新文献

{"title":"The flow and heat transfer in the rod bundle channel in the natural circulation system for heaving and rolling conditions","authors":"Jing-Liang Bi,&nbsp;Yan-Ping Huang,&nbsp;Feng Xie,&nbsp;De-Wen Yuan,&nbsp;Jian-Jun Xu,&nbsp;Dian-Chuan Xing","doi":"10.1016/j.pnucene.2025.105827","DOIUrl":"10.1016/j.pnucene.2025.105827","url":null,"abstract":"<div><div>Floating nuclear power plants are continually exposed to sea waves, which leads to rolling and heaving movement of the fuel assembly in the nuclear reactor. The majority of heat transfer experiments were carried out at low pressure, which is far from the operating pressure of nuclear power plants. Moreover, the effects of heaving on thermal hydraulics characteristics were barely experimentally investigated. The impact of heaving and rolling on the mass flux, flow resistance, rod wall temperatures and heat transfer coefficients (HTCs) of single-phase flow in a 3 × 3 rod bundle in a natural circulation loop are acquired experimentally in this research. The experimental findings suggest that the mass flux variation plot is very close to a sinusoidal curve pattern during heaving process. Rolling amplitude has a large effect on the average mass flux and fluctuation amplitudes of mass flux. In the cases where the rolling amplitude reaches a rather large angle, the average mass flux throughout rolling operation descends. The transient wall temperatures and HTCs did not change much during the heaving process, while the fluctuations of HTCs and rod wall temperatures are more drastic in rolling process. As the rolling amplitude reaches 30°, it can be clearly observed that the average HTC throughout the entire rolling operation exceeds the steady state HTC before rolling. The fluctuation of Re in the heaving process is not the main factor of the HTC fluctuation. Instead, the key factor contributing to the HTC fluctuation is the crossflow within the sub-channels.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105827"},"PeriodicalIF":3.3,"publicationDate":"2025-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143922675","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":"Influence of low permeable rocks above deep horizontal boreholes repository on the safety of solidified high level nuclear waste isolation","authors":"V.I. Malkovsky,&nbsp;V.A. Petrov,&nbsp;M.I. Ojovan ,&nbsp;S.V. Yudintsev","doi":"10.1016/j.pnucene.2025.105829","DOIUrl":"10.1016/j.pnucene.2025.105829","url":null,"abstract":"<div><div>We analyze the transport of radionuclides by thermal convection of groundwater from high level radioactive waste (HLW) disposed of in deep horizontal boreholes and the effect of above-located low-permeable geological formations typically represented by clayey rocks. As a rule, the topography-driven groundwater flow at depths of several kilometers is weak, and the resulting flow is governed by convection caused by heat generating HLW packages. We show that thermal convection for horizontal boreholes is much weaker compared with vertical boreholes. Due to this the rate of radionuclides migration from the horizontal boreholes is much lower than from similar disposal of HLW in vertical boreholes.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105829"},"PeriodicalIF":3.3,"publicationDate":"2025-05-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143928112","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":"Experimental study on characteristics of phase distribution in air-water bubbly flow in a rod bundle with spacer grid","authors":"Jiaxing Ren ,&nbsp;Ruohao Wang ,&nbsp;Fangdong Wang ,&nbsp;Weiqiang Xu ,&nbsp;Ruifeng Tian ,&nbsp;Puzhen Gao ,&nbsp;Shouxu Qiao ,&nbsp;Sichao Tan","doi":"10.1016/j.pnucene.2025.105801","DOIUrl":"10.1016/j.pnucene.2025.105801","url":null,"abstract":"<div><div>This study investigates the phase distribution characteristics of upward adiabatic air-water bubbly flow in a vertical 5 × 5 rod bundle with a spacer grid. A four-sensor conductivity probe is used to measure the detailed local cross-sectional distribution of interfacial parameters, including void fraction, interfacial area concentration, Sauter mean diameter, and bubble velocity in twelve bubbly flow conditions at five axial elevations. The core-peak and gap-peak distribution patterns of void fraction and interfacial area concentration in individual subchannels are observed through cross-sectional measurements. The bubble size distribution is relatively uniform, and the bubble velocity peak is consistently located at the subchannel center. The transition from core-peak to gap-peak distributions occurs with increasing superficial liquid velocity and decreasing superficial gas velocity. During this process, a dual peak appears temporarily before converging into a stable single peak in the subchannel edge. The effects of the spacer grid on phase distribution are discussed in detail. Cross-flow causes a shift in the local peaks and enhances bubble exchange between subchannels, resulting in a more uniform distribution of the entire cross-section. The mixing vanes significantly increase the overall void fraction within a very short distance, approximately 1–2 <em>D</em>_<em>h</em> downstream of the spacer grid. Using the improved cross-sectional area-averaging method, the one-dimensional distribution of interfacial parameters is obtained to analyze the transport characteristics in rod bundles. The spacer grid causes an increase in interfacial area concentration by promoting bubble breakup, as confirmed by the axial distribution of bubble size. The combined effects of the spacer grid are critical for predicting thermal-hydraulic parameters in nuclear safety analysis. Based on the present experimental study, its influence gradually diminishes around 20 <em>D</em>_<em>h</em> downstream in bubbly flow.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105801"},"PeriodicalIF":3.3,"publicationDate":"2025-05-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143916386","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":"3D imaging of gas bubbles in nuclear waste containers via Muon Scattering Tomography","authors":"Mohammed Mhaidra ,&nbsp;Ahmad Alrheli ,&nbsp;Dominic Barker ,&nbsp;Chiara De Sio ,&nbsp;Daniel Kikoła ,&nbsp;Anna Kopp ,&nbsp;Patrick Stowell ,&nbsp;Lee Thompson ,&nbsp;Jaap Velthuis ,&nbsp;Michael Weekes ,&nbsp;Magdalena Dobrowolska","doi":"10.1016/j.pnucene.2025.105803","DOIUrl":"10.1016/j.pnucene.2025.105803","url":null,"abstract":"<div><div>Due to perpetual radiation exposure, radioactive waste stored in bituminized containers can generate various gases, which may pose significant safety and handling concerns for storage facilities. This paper introduces a novel 3D imaging method that uses muon scattering tomography to detect and map gas bubbles. Our study demonstrates that this technique can detect gas volumes as small as 0.55 L at a 95% confidence level after 27 days of muon exposure in a 200 L bituminized waste container. Our approach is versatile and can be applied to various waste management scenarios and other industries for gas and void control.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105803"},"PeriodicalIF":3.3,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143912480","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":"Implementation of simultaneous mechanical- and chemical-spectral shift control mechanisms on the VVER-1200 assembly fueled with UO2 and MOX fuels","authors":"Behram Melikkendli","doi":"10.1016/j.pnucene.2025.105817","DOIUrl":"10.1016/j.pnucene.2025.105817","url":null,"abstract":"<div><div>In the present study, simultaneous mechanical- and chemical-spectral shift control mechanisms are implemented on the VVER-1200 assembly. The chemical regulation is applied by varying the volume fraction of heavy water in mixed <span><math><mrow><msub><mrow><mi>H</mi></mrow><mrow><mn>2</mn></mrow></msub><mi>O</mi><mo>/</mo><msub><mrow><mi>D</mi></mrow><mrow><mn>2</mn></mrow></msub><mi>O</mi></mrow></math></span> moderator during the burnup. The mechanical spectral shift regulation is based on the burnup-dependent change of the volumetric ratio of the moderator to fuel. It also requires modifying the structural characteristics of the reference fuel assembly. For this aim, without any change in the total fissile mass, the spectral-shift control fuel assembly (SSC-FA) is introduced by reducing both the fuel rod pitch and fuel rod numbers. This modification, in turn, increases the unit-cell numbers, where some of the unit-cells are allocated to the movable spectral shift control rods (SSCRs) filled with <span><math><msub><mrow><mi>ThO</mi></mrow><mrow><mn>2</mn></mrow></msub></math></span>. To implement the suggested simultaneous mechanisms, the burnup steps of a certain length are considered, and the optimum volume fraction of heavy water and optimum insertion heights of the SSCRs in each burnup step are calculated. The implementation of the suggested methodology shows a significant improvement in both cycle burnup and conversion ratio compared to those of the reference assembly. These improvements also increase by decreasing the burnup step length. The proposed method is also applied to the equivalent MOX-fueled SSC-FA. It is observed that the use of MOX reduces the heavy water requirement and also provides slightly greater improvement in both the cycle burnup and conversion ratio compared to <span><math><msub><mrow><mi>UO</mi></mrow><mrow><mn>2</mn></mrow></msub></math></span>-fueled SSC-FA. Finally, for the considered cases, burnup-dependent variations of the fuel compositions, temperature reactivity coefficients, pin power peaking factors, and kinetics parameters are calculated and compared with each other.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105817"},"PeriodicalIF":3.3,"publicationDate":"2025-05-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143905868","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 and sensitivity analysis of core flow distribution optimization using Monte-Carlo method and Wilks' formula","authors":"Zixuan Wang ,&nbsp;Yifan Meng ,&nbsp;Yan Wang","doi":"10.1016/j.pnucene.2025.105828","DOIUrl":"10.1016/j.pnucene.2025.105828","url":null,"abstract":"<div><div>The core flow distribution optimization is of great significance in enhancing reactor performance and safety. However, the use of uncertainty quantification and sensitivity analysis to assess the reliability of flow distribution optimization schemes is rare. Previous uncertainty studies mainly focus on safety-related reactor parameters. In this paper, the optimization objectives (the maximum-to minimum and the maximum-to-average temperature difference at the core outlet) and the safety constraint (MDNBR) for flow distribution were considered as the system outputs of interest. Taking twelve input uncertain parameters into account, Monte Carlo method and Wilks' method were utilized to quantify the output uncertainty of the optimized system. The results showed that, even under the most restrictive conditions, the optimized system still significantly outperformed the unoptimized state, with MDNBR remaining well above the safety limit. Therefore, the credibility of the optimization scheme was confirmed. Furthermore, through adjusting the order of the Wilks’ statistics and multiple trials, we compared the performance of the two uncertainty analysis methods. Lastly, sensitivity analysis based on Monte Carlo results was performed using Pearson, Spearman, Kendall, and partial rank correlation coefficients. The most influential parameter on core outlet temperature nonuniformity and the hottest channel temperature was the radial power distribution. On the other hand, axial power distribution, system flow, core power, and radial power distribution exhibit significant correlations with MDNBR.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105828"},"PeriodicalIF":3.3,"publicationDate":"2025-05-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143903814","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":"Statistical evaluation comparing strength of new and aged nuclear grade HEPA filters","authors":"J. Matthew Barnett ,&nbsp;Mary Bliss ,&nbsp;Scott K. Cooley ,&nbsp;Kurt J. Haunreiter","doi":"10.1016/j.pnucene.2025.105818","DOIUrl":"10.1016/j.pnucene.2025.105818","url":null,"abstract":"<div><div>Nuclear grade high-efficiency particulate air (HEPA) filter service life is generally limited to 10 years from the date of manufacture at Department of Energy facilities. Filter life extension offers a substantial cost savings and decreased facility downtime. This study compared tensile strength results between unused new and unused aged filters (7, 10, and 13 years old) to discern age related differences. Unused HEPA filters were disassembled. The filter sheets were preconditioned in a controlled environment for 24 h to resolve any atmospheric history of the filter media and to remove any effects of hysteresis on the results. Media used in HEPA filters are fragile, and damage occurs outside of the sample area when the material is cut. Basis weight measurements were recorded. Machine direction (MD) and cross direction (CD) tensile strength specimens were taken from the top or bottom of each filter pleat. Tensile strength specimens were also obtained of the pleat fold in the machine direction. Burst strength specimens were taken from across each filter pleat from top to bottom. Analysis of the raw data suggests that differences in tensile strength between new and aged HEPA filters are not significant. These results complement previous studies that have evaluated the relationships between the postulated 10-year HEPA filter lifespan, other performance indicators, and assessed cohort of aged filters as a function of normal operations. Those results along with the tensile strength results reported here suggest that HEPA filters, when properly maintained, can reasonably operate safely and efficiently for extended periods beyond 10 years when managed with well-defined performance and operational requirements.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105818"},"PeriodicalIF":3.3,"publicationDate":"2025-05-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143902169","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":"Temperature effect on the oxygen control efficiency of LSCF-type oxygen pump for LBE","authors":"Weihao Wu,&nbsp;Huiping Zhu,&nbsp;Tian Wang,&nbsp;Wenxuan Li,&nbsp;Zhangpeng Guo,&nbsp;Ruixian Liang,&nbsp;Hui Li,&nbsp;Lingfeng Yang,&nbsp;Zhenhua Sheng,&nbsp;Fang Liu,&nbsp;Haicai Lyu","doi":"10.1016/j.pnucene.2025.105804","DOIUrl":"10.1016/j.pnucene.2025.105804","url":null,"abstract":"<div><div>As a control component for dissolved oxygen concentration in the coolant of lead-cooled fast reactor, the electrochemical oxygen pump has proven to be effective. Temperature is a critical factor affecting the efficiency of oxygen pumps in controlling oxygen levels. However, there is currently a lack of quantitative research to elucidate the underlying mechanisms of this influence. This paper focuses on the effect of temperature on the oxygen control performance of an oxygen pump. Multi-cycle oxygen control experiments were conducted using an oxygen pump with La_0.6Sr_0.4Co_0.2Fe_0.8O_3+δ(LSCF) as the cathode powder within the temperature range of 350–500 °C. The experimental results indicate that the oxygen control efficiency of the oxygen pump increases with rising temperature. At each temperature within a single oxygenation-deoxygenation cycle, the oxygenation rate was significantly higher than the deoxygenation rate. The oxygenation rate remains relatively constant, while the deoxygenation rate gradually diminishes and approaches zero, ultimately reaching the deoxygenation limit, which increases as the temperature decreases. Additionally, with prolonged oxygen control time, the deoxygenation limit of the oxygen pump shows a certain degree of increase.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105804"},"PeriodicalIF":3.3,"publicationDate":"2025-05-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143898558","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 heat flux on the structural and chemical characteristics of fuel CRUD under pressurized water reactor primary water conditions","authors":"Yunju Lee ,&nbsp;Junhyuk Ham ,&nbsp;Seung Chang Yoo ,&nbsp;Inyoung Song ,&nbsp;Junhyuk Jeong ,&nbsp;Ji Hyun Kim","doi":"10.1016/j.pnucene.2025.105802","DOIUrl":"10.1016/j.pnucene.2025.105802","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105802"},"PeriodicalIF":3.3,"publicationDate":"2025-05-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143891273","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":"Cross-domain fault diagnosis method for nuclear power plant bearings based on deep transfer learning under small sample conditions","authors":"Wenzhe Yin ,&nbsp;Hong Xia ,&nbsp;Enrico Zio ,&nbsp;Xueying Huang","doi":"10.1016/j.pnucene.2025.105792","DOIUrl":"10.1016/j.pnucene.2025.105792","url":null,"abstract":"<div><div>In recent years, methods based on deep learning have attracted attention in the fault diagnosis of rotating machinery in nuclear power plants (NPPs). However, these methods are typically developed under the assumption that sufficient fault samples are available. In practice, rotating machinery in NPPs operate in healthy state most of the time and faults occur rarely and last a relatively short period of time. This work proposes a fault diagnosis method based on deep transfer learning to overcome the issue of small sample conditions in the bearing fault diagnosis task of NPPs. The bearing vibration signals collected by the sensor are converted into a time-frequency map by synchrosqueezed wavelet transforms, they are used as input of the deep convolutional neural network. In the learning phase, the deep learning model first learns domain-related knowledge from real devices, then the model parameters are transferred to the target task, and the model is fine-tuned based on the target domain knowledge. The proposed method was applied to two case studies: bearing fault localization and fault severity assessment. Experimental results demonstrated that, for the fault localization case, the method achieved average accuracy, precision, and F1 score of 95.21 %, 95.35 %, and 95.17 %, respectively, under four small sample conditions (with 10, 20, 30, and 40 samples per category in the training dataset). For the fault severity assessment case, the method attained average accuracy, precision, and F1 score of 95.03 %, 95.45 %, and 94.94 %, respectively, under three small sample conditions (with 3, 5, and 8 samples per category in the training dataset), demonstrating its potential value for NPPs bearing fault diagnosis.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105792"},"PeriodicalIF":3.3,"publicationDate":"2025-04-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143887814","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}