Progress in Nuclear Energy最新文献

{"title":"Study on hydrodynamic characteristics of symmetric two-droplet impact on film using lattice Boltzmann method","authors":"Huifang Zhang,&nbsp;Jian Yu,&nbsp;Yapei Zhang,&nbsp;Shihao Wu,&nbsp;Wenxi Tian,&nbsp;Suizheng Qiu,&nbsp;Guanghui Su","doi":"10.1016/j.pnucene.2026.106269","DOIUrl":"10.1016/j.pnucene.2026.106269","url":null,"abstract":"<div><div>Droplet impact on a liquid film is ubiquitous omnipresent and very fundamental in nature and industrial. For instance, in the spray cooling of the lower head of reactor pressure vessels, the method can enhance the safety margin of reactors. Extensive research has been carried out on the vertical impact of multiple droplets or single droplet on liquid films. However, the dynamical characteristics of multiple droplets impacting inclined liquid films remain insufficiently understood. Moreover, simulation approaches have predominantly concentrated on the Volume of Fluid (VOF) method. Therefore, this study attempts to conduct an in-depth numerical investigation of this phenomenon using the lattice Boltzmann method (LBM). A computational model was developed based on the Q3D27 and validated through benchmark cases involving single-droplet impacts on liquid films under both vertical and oblique conditions. The model accurately predicted key characteristics such as the outer diameter of the crown splash and the upstream crown radius. Based on the validated model, simulations of oblique impacts by dual droplets on a thin liquid film were conducted. The interfacial evolution was systematically analyzed, including the formation and development of crown splashes as well as the dynamics of intermediate thin-film jets. Furthermore, the Plateau-Rayleigh instability theory was employed to investigate the breakup mechanisms of liquid columns under varying impact angles and velocities. The fluid dynamic interactions between the two droplets under oblique impact conditions were also examined in detail, revealing complex flow behaviors relevant to multiphase flow dynamics.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106269"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039088","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":"Formation and evolution of radiation defects under low-temperature neutron irradiation","authors":"A.V. Kozlov ,&nbsp;N.V. Glushkova ,&nbsp;K.A. Kozlov ,&nbsp;K.M. Ladeyschikov ,&nbsp;V.L. Panchenko ,&nbsp;A.A. Zyryanova ,&nbsp;E.V. Shabelnikov","doi":"10.1016/j.pnucene.2026.106289","DOIUrl":"10.1016/j.pnucene.2026.106289","url":null,"abstract":"<div><div>The current study presents the way for consistent description of the processes of the radiation defects formation and evolution in conditions of low-temperature neutron irradiation. The topicality of the problem is due to the fact that exposure parameters such as temperature, rate of radiation damage, and damaging dose are not enough for description of the microstructure evolution. The main feature of the method consists in using the statistical model of defects migration. For the cross section descriptions of the interactions between neutrons (depending on their energies) and atoms the analytic expressions, fitting procedure is performed allowing calculations without expensive high-velocity electronic computing systems. The proposed approach was applied to describe the radiation defects evolution in chromium (as BCC structure pattern) exposed to low-temperature neutron irradiation in the reactor IVV-2M. It is shown that calculation results have a good match with experimental data regarding the number of vacancies accumulated in the sample during irradiation obtained using the dilatometer measurements. The developed approach is supposed to be uniform due to application for the description of defects formation and evolution in pure metals as well as in the alloys under irradiation in a wider temperature range in the reactors with various neutron spectra.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106289"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189699","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":"Heat conduction calculation of pebble-bed high temperature gas-cooled reactor based on nodal expansion method","authors":"Aolin Zhang,&nbsp;Dongyu Xu,&nbsp;Yongping Wang,&nbsp;Yuxuan Wu,&nbsp;Chuangye Zhou","doi":"10.1016/j.pnucene.2026.106277","DOIUrl":"10.1016/j.pnucene.2026.106277","url":null,"abstract":"<div><div>The Pebble-Bed High-Temperature Gas-Cooled Reactor (PB-HTGR) features a core structure with porous media characteristics, leading to big differences from Pressurized Water Reactors (PWR) in terms of thermal-hydraulic calculation. For the solid heat conduction calculation of the whole core, mainstream codes, like THERMIX, mainly adopt the finite volume method. There is still room for further optimization of this method's computational efficiency. To address this issue, the Nodal Expansion Method (NEM), originally used for solving the neutron diffusion equation, is applied to the whole-core solid heat conduction calculation of PB-HTGR, aiming to improve the computational efficiency while maintaining the accuracy. A corresponding code TH-NEM is developed. Meanwhile, biquadratic polynomials are used for temperature reconstruction in the meshes. A series of steady-state and transient heat conduction problems are calculated, and the results are in good agreement with the reference. In addition, for the HTR-PM heat conduction problem, compared with the fine-mesh method, the proposed method achieves a speed-up ratio of 4. Finally, TH-NEM is used to simulate the depressurized loss of forced cooling (DLOFC) accident of HTR-PM. The calculated maximum fuel temperature is 1532 °C, which is 40 °C higher than that of the reference code but still within the limitation of 1620 °C.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106277"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189225","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 investigation of methyl iodide removal in venturi scrubber","authors":"Jawaria Ahad,&nbsp;Masroor Ahmad,&nbsp;Amjad Farooq,&nbsp;Khalid Waheed,&nbsp;Shafiq-ur Rehman,&nbsp;Naseem Irfan","doi":"10.1016/j.pnucene.2026.106268","DOIUrl":"10.1016/j.pnucene.2026.106268","url":null,"abstract":"<div><div>Methyl iodide is a hazardous and volatile radioactive product that can be released during severe accidents in nuclear power plants, posing significant risk to human health and the environment. Therefore, efficient removal of methyl iodide is crucial for ensuring public safety and environmental protection. Due to its high volatility and hazardous nature, methyl iodide is difficult to handle, which has hindered extensive research into its removal using wet scrubbers. This study investigates the removal of methyl iodide using a venturi scrubber. Performance of a venturi scrubber is highly dependent on its operational parameters. So, influence of various parameters like liquid level height, gas flowrate, superficial velocity &amp; gas holdup was studied. Results showed that the studied parameters had a positive impact on removal of methyl iodide. Additionally, the effect of different additives such as water, sodium hydroxide, sodium thiosulphate and Aliquat-336 was investigated. Best results were obtained from the combination of sodium hydroxide, sodium thiosulphate and Aliquat-336 and maximum removal efficiency of &gt;98 % was obtained. This study also addresses the revolatilization of methyl iodide, considering its volatile nature by studying the effect of dose, pH and temperature. Results indicated that low dose, higher pH and temperatures up to 80 <span><math><mrow><mover><mspace></mspace><mo>̊</mo></mover><mi>C</mi></mrow></math></span> can reduce revolatilization. Addition of Aliquat-336 was observed to be particularly effective in this regard. Furthermore, a mass transfer model for venturi scrubber was developed to evaluate the effect of afore-mentioned additives on methyl iodide removal. Good agreement was observed between experimental and simulation results. This research would be beneficial in understanding the influence of different parameters on removal of methyl iodide to achieve optimum removal efficiency in order to keep the environment safe and clean and for facilitating the development of nuclear energy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106268"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189701","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 chattering-free fuzzy adaptive sliding mode controller based on a nonlinear two-point kinetics model for load-tracking of nuclear reactor","authors":"Hailemichael Guadie Mengsitu ,&nbsp;Xiuchun Luan ,&nbsp;Hetao Sun ,&nbsp;Fabiano Gibson Daud Thulu","doi":"10.1016/j.pnucene.2026.106254","DOIUrl":"10.1016/j.pnucene.2026.106254","url":null,"abstract":"<div><div>In this study, a T-S type fuzzy logic controller-based chattering-free adaptive sliding mode controller is proposed to attain smooth and precise power-level control and disturbances rejection capability for the VVER-1000 reactor core. To accomplish the purpose, a nonlinear two-point kinetics model with three groups of delayed neutron precursors is developed. An adaptive T-S type fuzzy logic controller is considered to eliminate the chattering problem inherent in a conventional sliding mode controller, in which a continuous switching function replaces the discontinuous switching signum function. Stability analysis is guaranteed using the Lyapunov synthesis approach. To validate the designed controller, several simulations are conducted under various transient conditions. Furthermore, the designed controller is compared with a traditional sliding mode controller, a conventional PID controller with fixed parameters, and a fuzzy controller with standalone configurations. The results demonstrate that the proposed controller strategy exhibits effective load-tracking performance and adaptive disturbance rejection ability under load-following operations. It is adapted to different working conditions, effectively reduces overshoot and settling time, and produces a smooth control input deprived of the chattering problem in the actuator. It achieved an accurate estimation of most unmeasured states and removed the premise that all system states are quantifiable, which makes more sense from a practical standpoint, and it was able to estimate the majority of unmeasured states accurately. Furthermore, the system outputs, normalized axial offset, and axial xenon oscillation index remain within acceptable ranges based on a constant axial offset power-distribution strategy.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106254"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981521","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":"Diagnosis of water hammer conditions in multi-pipeline system using data-driven and genetic algorithm approach","authors":"Qingyin Zeng ,&nbsp;Cheng Peng ,&nbsp;Jiang Wu ,&nbsp;Jian Deng","doi":"10.1016/j.pnucene.2026.106274","DOIUrl":"10.1016/j.pnucene.2026.106274","url":null,"abstract":"<div><div>While existing experimental investigations and numerical simulations have preliminarily elucidated the multi-factor coupled mechanism governing Condensation-Induced-Water-Hammer (CIWH), the conventional threshold-based alarm approach predominantly relies on single-parameter criteria, e.g. typically pressure, which fails to effectively capture the early warning signs of CIWH. Concurrently, the genetic algorithm, recognized for its applicability in complex systems, remains relatively underutilized in the rapid identification of hazardous operating conditions, resulting in a technical gap where mechanistic understanding is disconnected from practical prevention and control requirements. This study investigates CIWH in the feedwater pipelines of the secondary circuit deaerator in nuclear power plants, where subcooled feedwater mixes with steam. A numerical simulation method based on the NUMAP code is developed to capture the underlying dynamics. Pilot study reveals that feedwater flow rate and pipe diameter jointly regulate the balance between inertial and frictional forces. This interaction shapes flow distribution and void fraction, which in turn influence the intensity of CIWH. Building upon this insight, a systematic analysis is conducted to quantify the effect of feedwater flow rate and pipe diameter on pressure variation rates, with peak values occurring at approximately 84.5 kg/s and 506 mm, respectively. To further identify hazardous operating conditions, a genetic algorithm (GA) is employed with different fitness functions. The results demonstrate that the relative change rate outperforms other metrics, whereas the mean absolute change and standard deviation show certain deviations, and the coefficient of variation is the least effective. This study confirms the effectiveness of the genetic algorithm in identifying hazardous operating conditions of CIWH under complex coupled scenarios and provides a feasible approach for predictive safety control and operational risk assessment in nuclear power plants.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106274"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039079","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":"RELAP5-3D simulation of natural circulation start-up and station blackout benchmark for a NuScale-like iPWR","authors":"Vincenzo Zingales ,&nbsp;Francesco D'Auria ,&nbsp;Yassin A. Hassan","doi":"10.1016/j.pnucene.2026.106265","DOIUrl":"10.1016/j.pnucene.2026.106265","url":null,"abstract":"<div><div>This paper discusses the key features of integral Pressurized Water Reactors (iPWRs) including Helical Coil Steam Generators (HCSGs) and associated correlations. The objective is to highlight design challenges and system level modeling strategies.</div><div>A RELAP5-3D ver. 4.4.2 nodalization of a NuScale-like iPWR was created based on publicly available literature and, where necessary, design assumptions. The model was qualified under steady-state conditions against the NuScale Final Safety Analysis Report approved as part of the Design Certification Application (DCA) in 2020.</div><div>Because HCSGs are prone to instabilities, a start-up procedure was simulated to test the model response across a range of operating parameters. Primary flow and temperature results have been found to be consistent with DCA data at reduced power. However, at power levels below 60%, Type-II Density Wave Oscillations (DWOs) occurred in the HCSGs tubes. Analysis of subcooling and phase change numbers informed of potential mitigation strategies; however, stable performance at low power could not be obtained together with steam superheat and constant primary average temperature.</div><div>The main outcome of the present paper is the presentation of a collaborative benchmark internal to the Texas A&amp;M University in which a station blackout scenario for a NuScale-like iPWR was simulated. Results obtained from RELAP5-3D were compared with those from TRACE, the NuScale Simulator, and DCA data, demonstrating strong qualitative agreement but highlighting quantitative discrepancies primarily due to geometric and correlation differences among the different models. RELAP5-3D simulations have specifically highlighted the occurrence of Type-I DWO if the ECCS is not timely activated.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106265"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039089","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":"Two-group interfacial area concentration model for downward dispersed two-phase flows","authors":"Shaoting Jia,&nbsp;Takashi Hibiki","doi":"10.1016/j.pnucene.2026.106271","DOIUrl":"10.1016/j.pnucene.2026.106271","url":null,"abstract":"<div><div>The two-fluid model is widely employed in thermal-hydraulic simulations of gas-liquid two-phase flows. Its predictive capability is strongly dependent on the accurate estimation of the interfacial area concentration (IAC), a key parameter governing the transfer of mass, momentum, and energy across the interphase. In dispersed flows, bubbles vary widely in size and morphology, which significantly influence their dynamics and contributions to IAC. Those variations necessitate a two-group classification: Group 1 (G1) for spherical and distorted bubbles, and Group 2 (G2) for cap, slug, and churn-turbulent bubbles. This study evaluated the predictive performance of several existing two-group IAC models, including IAC models implemented in RELAP5 and TRACE codes, as well as Song and Hibiki's IAC model, using compiled experimental data from vertically downward two-phase flows. The results showed that none of these models could predict the IAC with sufficient accuracy. To address this issue, the contribution of G1 bubbles to the total IAC was analyzed to investigate the flow behaviors for downward two-phase flows. Then, a two-group drift-flux correlation specifically developed for downward flows was introduced to predict the void fractions of G1 and G2 bubbles. Then, a new two-group Sauter mean diameter (SMD) model was formulated by considering the interfacial geometry effect and flow orientation. By integrating the drift-flux correlation and SMD model, an advanced two-group IAC model was proposed for downward two-phase flows. The proposed model demonstrated significantly improved predictive performance for downward dispersed flows, achieving mean absolute relative errors of 31.0 %, 26.7 %, and 27.9 % for G1, G2, and total IACs, respectively.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106271"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189703","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":"Analysis of key nuclides in Xe irradiated by neutrons and their chemical behavior","authors":"Jia Tang ,&nbsp;Yin Hu ,&nbsp;Yi Xia ,&nbsp;Jixue Sui ,&nbsp;Jiajia Yang ,&nbsp;Yunming Chen ,&nbsp;Xiaogen Xiong ,&nbsp;Yi Wang ,&nbsp;Qi Cao","doi":"10.1016/j.pnucene.2026.106291","DOIUrl":"10.1016/j.pnucene.2026.106291","url":null,"abstract":"<div><div>He-Xe binary mixture is a leading candidate for the fourth-generation gas-cooled reactor coolants. this study utilized MJTR to conduct in-pile irradiation experiments on quartz target components filled with Xe gas. The cumulative neutron fluence exceeded 1.0E16 n/cm². And a qualitative to semi-quantitative relationship between Xe impurities yield and neutron fluence was established. Based on the experimental results, the radioactivity derived from neutron irradiation of Xe in specific reactor system was estimated. The chemistry and deposition processes of these impurities was predicted. The results show that CsI and Cs₂Te are thermodynamically formable. Once formed, they are in solid state and unlikely to spontaneously decompose into atoms or ions. At temperatures below 1300–1400 K and 1100–1200 K, CsI and Cs₂Te may spontaneously undergo polymerization reactions to form dimers, which further crystallize, leading to the formation of macroscopic crystals. The main impurity Te would react with Inconel 617, a promising structure material, to form NiTe_0.69. This study deepens the understanding of the physical and chemical behavior of Xe under neutron irradiation and provides valuable guidance for the design and operational strategies of He-Xe coolant-based gas-cooled reactor systems.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106291"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146189712","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 post-dryout heat transfer based on physics-embedded machine learning with Bayesian optimization algorithm","authors":"Meiqi Song ,&nbsp;Zuokai Chen ,&nbsp;Jianhua Xia ,&nbsp;Haozhe Li ,&nbsp;Wei Xu ,&nbsp;Xiaojing Liu","doi":"10.1016/j.pnucene.2026.106263","DOIUrl":"10.1016/j.pnucene.2026.106263","url":null,"abstract":"<div><div>In nuclear power system, encountering the post-dryout heat transfer region can lead to severe heat transfer deterioration. Therefore, it is of great importance to give accurate prediction to post-dryout heat transfer. This study developed a new Physics-Embedded Machine Learning (PEML) framework to predict post-dryout heat transfer, addressing the limitations of traditional \"black box\" models by integrating physical constraints. Thirteen independent dimensionless parameters (e.g., <span><math><mrow><mi>R</mi><msub><mi>e</mi><mtext>TP</mtext></msub></mrow></math></span>, <span><math><mrow><msub><mi>Pr</mi><mi>w</mi></msub></mrow></math></span>), i.e., input features, and Nusselt number <span><math><mrow><mi>N</mi><msub><mi>u</mi><mi>c</mi></msub></mrow></math></span> are derived to present the physical heat transfer mechanism. The <span><math><mrow><mi>N</mi><msub><mi>u</mi><mi>c</mi></msub></mrow></math></span> is embedded into the loss function in proportionality or subtractive relationship, i.e., PEML(<em>Nu</em>_exp/<em>Nu</em>_c) and PEML(<em>Nu</em>_exp-<em>Nu</em>_c). The prediction capability of PEML models are better than traditional correlations. The PEML(<em>Nu</em>_exp/<em>Nu</em>_c) model achieves the best prediction capability with the mean error of 0.0005 and RMS error of 0.007 on the testing dataset from Becker's PDO experiments. It is indicated that increasing the number of input features generally improved model performance, especially the generalizability. The PEML framework successfully embeds heat transfer physics, bridging data-driven models and physical insights, offering a robust prediction tool for heat transfer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"194 ","pages":"Article 106263"},"PeriodicalIF":3.2,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146039078","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}