Progress in Nuclear Energy最新文献

{"title":"Design and full core fuel performance assessment of high burnup cores for 4-loop PWRs","authors":"Assil Halimi ,&nbsp;Yifeng Che ,&nbsp;Koroush Shirvan","doi":"10.1016/j.pnucene.2025.105791","DOIUrl":"10.1016/j.pnucene.2025.105791","url":null,"abstract":"<div><div>Increasing the fuel discharge burnup of current light water reactors (LWRs) promises reductions in fuel cycle and/or operations costs. By assuming a constant core power density, the economic gain is enabled by better fuel utilization and/or an increased capacity factor. In this effort to investigate greater than 62 MWd/kgU maximum rod average burnup for 110+ kW/l core power density, two core designs have been developed for a standard 17x17, 193 fuel assemblies pressurized water reactor (PWR). The levelized unit cost methodology is employed to evaluate fuel cycle, operation and maintenance, and capital cost impacts and to examine the economic viability of both core design pathways. Core design and optimization are performed using the commercial STUDSVIK code package. Fuel performance analysis is realized in full core configuration via auditing FRAPCON4.1, FAST1.2, and the high-fidelity code BISON. To provide a realistic assessment, the core design process takes into consideration best practices in current PWR core design. It features acceptable performance in terms of various core design constraints on maximum allowable peaking and boron concentration. Gadolinia (Gd2O3) is used as a burnable poison with a maximum of 9 wt% concentration while feeding 89 or 77 fuel assemblies in a 3-batch refueling scheme. Full core fuel performance simulation, which allows for characterization of relevant fuel temperatures, plenum pressures, stresses, and strains, is performed with respect to two bounding burnup levels. Such performance is potentially licensable for the 18-month high burnup core (&lt;68 MWd/kgU peak pin), while it is more challenging for the 24-month high burnup core design pathway (&lt;75 MWd/kgU peak pin). Maximum rod plenum pressure is identified as the most limiting fuel performance parameter. While the scope of the present study focuses on the steady-state plus overpower conditions, the acceptability of the new discharge burnup has to be further assessed by considering uncertainties and impacts under accident scenarios in the future.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105791"},"PeriodicalIF":3.3,"publicationDate":"2025-05-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143931318","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":"Comprehensive analysis of oxygen mass transfer and impurity dynamics in oxygen-controlled ELSY lead-cooled reactors","authors":"Dingsheng Lu,&nbsp;Xiao Pang,&nbsp;Yupeng Yang,&nbsp;Chenglong Wang,&nbsp;Dalin Zhang,&nbsp;Suizheng Qiu","doi":"10.1016/j.pnucene.2025.105830","DOIUrl":"10.1016/j.pnucene.2025.105830","url":null,"abstract":"<div><div>In lead-cooled fast reactors, precise oxygen control is critical for mitigating corrosion, maintaining protective oxide layers, enhancing heat transfer, and reducing impurity deposition. However, achieving consistent oxygen management is challenging due to the high cost and limited availability of sensitive sensors, along with difficulties in monitoring oxygen distribution in pool-type reactors. Existing studies often oversimplify oxygen dynamics, neglecting the impacts of temperature gradients and cyclic flows on reactor-wide oxygen behavior and corrosion processes. This study refines traditional models by incorporating corrosion mechanisms and experimental data to analyze oxygen interactions with iron oxide nucleation in oxygen-controlled environments. Using Computational Fluid Dynamics (CFD), the distribution of oxygen and corrosion products is examined under various conditions. Results reveal hypoxic zones in high-flow and reflux regions under anoxic conditions, with inlet oxygen concentrations reaching a minimum at 1715 s. Enhanced gas-phase mass transfer stabilizes oxygen levels, reducing iron release and ensuring uniform distribution over time. Impurity nucleation begins in the cold zone beneath the steam generator at 7960 s, highlighting critical areas for management. By coupling reactor-wide oxygen dynamics and impurity behavior, this study underscores the importance of optimized oxygen management to mitigate corrosion, control impurities, and ensure long-term reactor stability, safety, and efficiency.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105830"},"PeriodicalIF":3.3,"publicationDate":"2025-05-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143937574","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":"Forecasting in-core power distributions in nuclear power plants via a spatial–temporal hierarchical-directed network","authors":"Weiqing Lin ,&nbsp;Xiren Miao ,&nbsp;Jing Chen ,&nbsp;Pengbin Duan ,&nbsp;Mingxin Ye ,&nbsp;Yong Xu ,&nbsp;Hao Jiang ,&nbsp;Yanzhen Lu","doi":"10.1016/j.pnucene.2025.105795","DOIUrl":"10.1016/j.pnucene.2025.105795","url":null,"abstract":"<div><div>As nuclear power plants (NPPs) undertake more peak regulation tasks to handle high new energy penetration and overcapacity, precise forecasting of in-core power distributions is essential for optimal control and safe operation. However, current works lack an effective strategy for predicting high-resolution power distributions and neglect in-core spatial correlations. This study proposes a spatial–temporal hierarchical-directed network (ST-HDN) for forecasting power distributions, whose prediction strategy is guided by the physical model. To characterize spatial correlations and causal relationships among physical quantities, the hierarchical-directed graph is designed and combined with neutron and power signals for input to the ST-HDN. Concretely, the ST-HDN integrates three sub-modules: a temporal-differencing layer to enhance representation of subtle variations; a multi-dilated convolutional network to extract dynamic temporal features; and a graph convolutional network to propagate spatial adjacent information, further predicting power nodes at various positions. The predicted power nodes are post-processed to derive future power distributions. Experiments on two peak regulation scenarios from a real-world NPP illustrate that the ST-HDN outperforms various state-of-the-art methods in 10-, 20-, and 30-min ahead forecasting.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105795"},"PeriodicalIF":3.3,"publicationDate":"2025-05-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143931841","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":"CFD simulation of internal center flow blockage in helical-cruciform fuel assembly","authors":"Z.S. Li ,&nbsp;C. Hu ,&nbsp;B.H. Yan ,&nbsp;L.S. Wang","doi":"10.1016/j.pnucene.2025.105831","DOIUrl":"10.1016/j.pnucene.2025.105831","url":null,"abstract":"<div><div>Helical-cruciform fuel (HCF) rod, an innovative fuel design, has been proposed for its promising thermal-hydraulic performances, which is characterized by contacting at points between rods to achieve self-supporting without space grids. Blockages are most likely to occur at contact point zone for the deteriorated fluidity and a flow blockage accident is caused. Hence, flow characterization of coolant flow in 7 rods HCF assembly with internal center blockage has been carried out by Reynolds-Averaged Navier-Stokes (RANS) based CFD simulation. In this paper, influence of different minimum gap values between petals have been evaluated. Based on this, different turbulence models have been validated for the law of the wall, after which CFD results of axial velocity profiles in the interior and exterior channel have been compared against in-house hydraulic experimental data of 7 HCF rod bundle. Focus has been paid to analyzing hydraulic features in HCF assembly with various blockage configurations, including different thicknesses and blocked percentage of flow channel sectional area. It is observed that special vortex structure formed upstream and downstream in typical blockage conditions. The transverse flow upstream and downstream of blockages is enhanced with increasing blockage thickness and blocked area. The results of the present study would serve as foundation for further in-depth flow blockage incidents analyses in HCF rod assembly.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"186 ","pages":"Article 105831"},"PeriodicalIF":3.3,"publicationDate":"2025-05-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143931308","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":"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}