Progress in Nuclear Energy最新文献

{"title":"Study on the influence of heating power on reflooding process in rectangular narrow channels","authors":"Dan Wu ,&nbsp;Jiayue Zhou ,&nbsp;Qianlong Zuo ,&nbsp;Changwen Liu ,&nbsp;Jian Deng","doi":"10.1016/j.pnucene.2025.105981","DOIUrl":"10.1016/j.pnucene.2025.105981","url":null,"abstract":"<div><div>It is well known that the reflooding process involves particularly complex two-phase flow and heat transfer characteristics, making both experimental studies and the development of analytical models challenging. Over the past few decades, extensive experimental and simulation research had been conducted internationally on the reflooding characteristics in rod bundles, circular tubes, and so on. However, the flow and heat transfer characteristics of reflooding process in rectangular narrow channels still needs further investigation. The most severe challenge is whether the cooling capacity is sufficient to remove decay heat, thereby preventing the wall temperature from continuously rising, especially in the precursory cooling region where the quench front has not yet arrived. This paper focused on experimental studies of reflooding process in rectangular narrow channels. The main emphasis were on the differences in key parameters such as the quench front velocity, the wall temperature changes under different heating power. Additionally, visualization experiments were conducted to observe droplets and changes of flow patterns, which would help to understand the impact mechanism of heating power on the reflooding process. It was found that the effect of heating power was very complex. If the heating power was too low, steam generated by vaporization was insufficient to carry enough liquid to the downstream of the quench front, leading to poor precursory cooling capacity. If the heating power was too high, the overall cooling rate would be slower due to the rapid evaporation of droplets and the difficulty of droplets contacting with the wall. In the high superheat region, heat transfer coefficient increased with the increase of heating power under the same wall superheat. It should be pointed out that the convective heat transfer coefficient of single-phase steam was very small, the heat transfer coefficient was even less than 100W/m²K at the beginning of reflooding process within the parameter ranges we were studying. In addition, the visualization research results showed that Inverted annular flow rather than annular flow or annular dispersed flow was the main flow pattern in the quench front region during the reflooding process of rectangular narrow channels. And the higher the heating power, the greater the probability of droplets appearing in the precursor cooling region at the beginning of reflooding process.The findings of this study will enhance our understanding on the reflooding characteristics of rectangular narrow channels. Further experimental research especially focusing on on droplet size and motion characteristics will be conducted in the future, and on this basis, a set of numerical analysis models for the reflooding process in rectangular narrow channels will be developed.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105981"},"PeriodicalIF":3.2,"publicationDate":"2025-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144903673","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 the characteristics of annular fluid film in a locally deformed narrow rectangular channel","authors":"Fenze Wang ,&nbsp;Quanyao Ren ,&nbsp;Haidong Liu ,&nbsp;Yi Xiao ,&nbsp;Hongxing Yu ,&nbsp;Deqi Chen","doi":"10.1016/j.pnucene.2025.106000","DOIUrl":"10.1016/j.pnucene.2025.106000","url":null,"abstract":"<div><div>The local deformation of plate fuel assembly due to neutron irradiation will lead to the deterioration of heat transfer in the flow channel. Studying the characteristics of annular flow films in narrow rectangular channels under localized deformation provides a foundation for understanding the heat and mass transfer and the hydrodynamic characteristics of the locally deformed narrow flow channel. In this study, the annular flow liquid film characterization experiments were conducted in a narrow rectangular channel under three conditions: no local deformation, ellipsoidal deformation, and pillow-shaped deformation. Various parameters of the annular flow liquid films, including liquid film thickness, disturbance wave height, and disturbance wave wavelength, were obtained using a high-speed camera acquisition system and advanced image processing techniques. The effects of flow conditions and local deformation on the liquid film parameters of each annular flow were analyzed. As the radial size of the local deformation increases relative to the liquid film thickness, the viscous bottom layer of the liquid film disintegrates, resulting in a reduction of the total liquid film thickness. When the gas-phase velocity is low, the local deformation exerts a greater shear force on the disturbance wave, prolonging the action time and increasing the likelihood of wave disintegration. Under conditions of no channel deformation, the disturbance wave wavelength is negatively correlated with the gas-phase velocity; conversely, under localized channel deformation, the wavelength is positively correlated with the gas-phase velocity. Thus, the effect of localized deformation on wavelength is bidirectional. Local deformation impact factor functions for liquid film thickness, disturbance wave height, and disturbance wave wavelength have been derived, with the error of these impact factor functions maintained within ±30 %.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 106000"},"PeriodicalIF":3.2,"publicationDate":"2025-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144903674","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":"Application of the half-boundary method to solving the neutron transport equation in 1D cylindrical coordinate","authors":"Yang Liu ,&nbsp;Lei Xue ,&nbsp;Hangyu Shi ,&nbsp;Zhiyi Ni ,&nbsp;Liangzhi Cao ,&nbsp;Xiaoping Ouyang","doi":"10.1016/j.pnucene.2025.105996","DOIUrl":"10.1016/j.pnucene.2025.105996","url":null,"abstract":"<div><div>The neutron distribution within a nuclear reactor core plays a crucial role in nuclear engineering, directly influencing the safe operation of nuclear reactors. The neutron transport equation provides a fundamental approach to determine this distribution. This study applies the half-boundary method (HBM) to solve the neutron transport equation in cylindrical coordinates. By deriving mathematical relationships among discrete nodal values, the HBM establishes explicit correlations between boundary conditions and neutron flux at arbitrary spatial points throughout the model. Compared to traditional finite difference methods, the HBM only requires iterative calculations on boundary values, thereby improving computational accuracy while reducing both execution time and memory requirements. In this paper, the HBM discretization and derivation processes are described in detail. The sensitivity analysis to assess the influence of varying the spatial and angular discretization parameters is made. Convergence analysis demonstrates that spatial discretization achieves second-order accuracy in the radial direction, while angular discretization exhibits first-order convergence. Three numerical test cases are presented to verify the HBM by comparing its results with the Monte Carlo method, showing their consistency, high accuracy, and credibility.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105996"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902420","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 fuel alternatives to achieve breeding capability in the FUJI-U3-(0) reactor","authors":"Maíra Cesário A. Lobo ,&nbsp;Marcelo V. Silva ,&nbsp;Diego M.E. Gonçalves ,&nbsp;João Manoel L. Moreira ,&nbsp;Daniel Artur P. Palma ,&nbsp;Francky Roger A. da Silva ,&nbsp;Giovanni L. de Stefani","doi":"10.1016/j.pnucene.2025.105958","DOIUrl":"10.1016/j.pnucene.2025.105958","url":null,"abstract":"<div><div>In this study, the core physics of the FUJI-U3-(0) small modular molten salt reactor was simulated using the SERPENT code. The cylindrical thermal reactor core was modeled with three concentric regions, each containing distinct fuel channel radii embedded within hexagonal graphite moderator rods. Five molten salt fuel compositions were analyzed, varying the fissile and fertile materials and their respective proportions. After criticality was achieved, key reactor performance metrics, including conversion ratio and waste production, were evaluated. The results led to discussion of the feasibility of a thermal molten salt breeder reactor with different fuel types. Furthermore, APTh1000 and ANGRA-I nuclear waste was incorporated into selected fuel compositions. For APTh1000 waste, an economic assessment of recycled fuel was performed, which revealed that it is more cost-effective and environmentally advantageous compared to fresh fuel.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105958"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902421","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 flow and heat transfer characteristics of the core rod bundle region in a space gas-cooled microreactor based on similarity analysis method","authors":"Yangguang Zhang ,&nbsp;Daogang Lu ,&nbsp;Qiong Cao ,&nbsp;Xiaotian Wang ,&nbsp;Chao Xu ,&nbsp;Yuzhou Wang ,&nbsp;Henghua Liang","doi":"10.1016/j.pnucene.2025.105998","DOIUrl":"10.1016/j.pnucene.2025.105998","url":null,"abstract":"<div><div>In the design process of space gas-cooled microreactors, it is essential to investigate the flow and heat transfer characteristics of fuel rod bundles to ensure the rationality and feasibility. Helium-xenon gas mixtures are commonly employed as coolants in such reactors. However, existing studies about helium-xenon gas mixture flow and heat transfer characteristics were mainly performed with circular, annular, or rectangular cross-sectional channels, while few with rod bundle regions. Furthermore, current studies predominantly rely on numerical simulations, while few on experiments due to the high cost of helium-xenon gas mixture. In order to obtain the experimental data, an experiment is designed using air instead of helium-xenon gas mixture based on similarity analysis between air and helium-xenon gas mixture. Consequently, in this paper, a similarity analysis method by differential equation analysis method for the flow and heat transfer of air and helium-xenon gas mixture is established. According to the boundary conditions obtained from the similarity analysis theory, the flow and heat transfer experiment is conducted under different operating conditions to obtain the temperature distribution and the pressure drop variations within the core. Through analyzing the experimental results of helium-xenon gas mixture converted from air by similarity theory criteria, a set of semi-empirical correlations for helium-xenon gas mixture flow and heat transfer is proposed by using a six-rod average methodology, demonstrating errors within 10 %. The derived correlations exhibit conservatism, ensuring reliability in engineering applications. This study can validate the reactor core structural design and provide critical experimental references for the design of space gas-cooled microreactors.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105998"},"PeriodicalIF":3.2,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902422","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 effect of cladding elliptical deformation on the thermal-hydraulic performance of fuel assemblies","authors":"Jianghao Yang ,&nbsp;Mei Huang ,&nbsp;Yaodi Li","doi":"10.1016/j.pnucene.2025.105995","DOIUrl":"10.1016/j.pnucene.2025.105995","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105995"},"PeriodicalIF":3.2,"publicationDate":"2025-08-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902535","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":"Thermal shock behavior of PHWR fuel cladding subjected to LOCA integral test","authors":"T.K. Sawarn ,&nbsp;Shefali Shukla ,&nbsp;Suparna Banerjee ,&nbsp;P.P. Nanekar","doi":"10.1016/j.pnucene.2025.105984","DOIUrl":"10.1016/j.pnucene.2025.105984","url":null,"abstract":"<div><div>During a hypothetical loss of coolant accident (LOCA) in a pressurised heavy water reactor (PHWR), cladding temperature may rise up to 1500 °C due to coolant positive void coefficient of reactivity. At such a high temperature, β-Zr phase absorbs significant amount of oxygen and hydrogen. This paper provides novel data on the synergistic effects of oxygen, hydrogen and prior β-Zr layer thickness on the crack initiation and propagation during thermal shock in a LOCA scenario. Integral LOCA experiments were performed on the PHWR cladding. The results showed the presence of non-uniform distribution of oxygen, hydrogen and prior β-Zr layer thickness along the circumference and axial direction of cladding. Significant amount of localised deformation and oxidation in the burst region lead to a high oxygen pick up and thinning of prior β-Zr layer, making it a highly brittle region. Substantial amount of hydrogen (up to 5000 wppm) was picked up by the cladding away from the burst opening due to inner surface oxidation by steam. Two types of hydride morphologies were noticed, hydride blisters near the burst tip and fine hydride platelets away from it. Based on the integral tests, threshold concentration of oxygen in the prior β-Zr for crack initiation ahead of the burst tip was observed to be 1.2 wt%. Hydrogen played an indirect role of increasing the embrittlement of cladding by enhancing the oxygen solubility in the β-Zr. Hydrogen has minimal role on the thermal shock failure below 900 wppm. Circumferentially averaged more than cladding half thickness of prior β-Zr layer prevents the crack propagation.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105984"},"PeriodicalIF":3.2,"publicationDate":"2025-08-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144890340","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":"Multi-step time series prediction for nuclear power plants based on variational mode decomposition and gated recurrent units","authors":"Qiming Yang ,&nbsp;Minghan Yang ,&nbsp;Shuai Chen ,&nbsp;Zhulan Zhang ,&nbsp;Zicheng Liang ,&nbsp;Jianye Wang","doi":"10.1016/j.pnucene.2025.105990","DOIUrl":"10.1016/j.pnucene.2025.105990","url":null,"abstract":"<div><div>Accurate prediction of reactor sensor data and fault diagnosis is critical for ensuring the safe and stable operation of nuclear reactors. It can effectively enhance reliability, reduce the risk of failure, and ensure the stable operation of nuclear power plants. However, the nuclear fission process and neutron flux vary over time, and physical quantities such as reactor temperature, pressure, and flow are influenced by factors like load, power regulation, and coolant flow, which result in nonlinear fluctuations. These fluctuations make sensor data complex and non-stationary, making it difficult for traditional methods to extract useful features effectively. To address this, we propose VMD-Bi-GRU, a novel hybrid model combining Variational Mode Decomposition (VMD) and Bayesian-optimized Bidirectional Gated Recurrent Units (Bi-GRU). VMD decomposes raw sensor signals into physically interpretable intrinsic mode functions (IMFs), effectively isolating noise and enhancing feature extraction. The Bi-GRU network then leverages its bidirectional temporal modeling capability for multi-step prediction. Crucially, Bayesian optimization automates hyperparameter tuning, maximizing model generalizability. Evaluated on CPR1000 simulator data, VMD-Bi-GRU achieved average RMSE reductions of 5.4 % and 55.2 % across 1–20 prediction time horizons for the first and second experimental datasets, respectively On the second experimental dataset, VMD-Bi-GRU reduces 20-step prediction MAE by 46 %. The reconstructed predictions are highly consistent with the original data (<span><math><mrow><msup><mi>R</mi><mn>2</mn></msup></mrow></math></span> &gt;0.7), enabling early anomaly detection and reflecting the reactor status with higher fidelity. This framework provides a reliable foundation for intelligent scheduling and predictive maintenance of nuclear power plants.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105990"},"PeriodicalIF":3.2,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144886070","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":"Challenges of small modular reactors: A comprehensive exploration of economic and waste uncertainties associated with U.S. small modular reactor designs","authors":"Philseo Kim ,&nbsp;Allison Macfarlane","doi":"10.1016/j.pnucene.2025.105989","DOIUrl":"10.1016/j.pnucene.2025.105989","url":null,"abstract":"<div><div>This article presents a comprehensive analysis and review of the levelized cost of electricity (LCOE) and nuclear waste generation from small modular reactors (SMRs) designed in the United States. The analysis presented here offers a perspective on the uncertainties surrounding the calculation of LCOE for SMRs. Initially, the paper discusses and calculates the capital costs, front-end fuel cycle costs, operating and maintenance expenses, and nuclear waste production of four distinct types of SMRs—boiling light water, pressurized light water, high-temperature gas-cooled, and sodium fast reactor types— based on designs by four different manufacturers in the United States. Our review and analysis indicate there is considerable uncertainty in LCOE estimations of SMRs, particularly emphasizing that non-light water designs are likely to be more expensive in fuel cycle costs and have greater uncertainties in nuclear waste management compared to light water reactor designs. This comprehensive analysis of SMRs significantly broadens our understanding of their economic complexities and the implications these pose for deployment in the foreseeable future. Lastly, our estimates suggest that both SMRs and large nuclear reactors will require revolutionary approaches to reduce the capital cost overruns to be cost competitive with other energy sources.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105989"},"PeriodicalIF":3.2,"publicationDate":"2025-08-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144886071","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":"Revolutionizing radioactive waste resin treatment: A comprehensive review of advanced technologies and future directions","authors":"Yue-Lin Wang ,&nbsp;Yun Xue ,&nbsp;Yang-Hai Zheng ,&nbsp;Xin Liu ,&nbsp;Qing-guo Zhang ,&nbsp;Fu-Qiu Ma ,&nbsp;Mi-Lin Zhang ,&nbsp;Ren-Yan Xie ,&nbsp;Sheng Chang ,&nbsp;Yong-De Yan","doi":"10.1016/j.pnucene.2025.105987","DOIUrl":"10.1016/j.pnucene.2025.105987","url":null,"abstract":"<div><div>Nuclear-grade ion exchange resins are essential for managing the water chemistry within the primary circuit of nuclear power plants. Once these resins have reached the end of their useful life, they are typically not regenerated, and their disposal presents a significant challenge due to radioactive contamination. Spent ion exchange resins represent a substantial portion of low- and intermediate-level radioactive solid waste generated during the decommissioning of nuclear facilities. Various treatment methods have been developed to address this waste, which can be broadly classified into oxidation decomposition methods and non-oxidation decomposition methods. Oxidative decomposition approaches include acid digestion, supercritical water oxidation, Fenton processes, steam reforming, incineration, pyrolysis and gasification plasma treatment, and molten salt oxidation. Conversely, non-oxidation decomposition methods encompass cementation, bituminization, plastic solidification, vitrification, hot Super Compression, high Integrity Container. This paper reviews the fundamental principles underlying these treatment methodologies along with recent research advancements while discussing their respective advantages and limitations. The emphasis is placed on the process flow and processing capacity of various methods in practical applications.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"190 ","pages":"Article 105987"},"PeriodicalIF":3.2,"publicationDate":"2025-08-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144864499","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}