Progress in Nuclear Energy最新文献

{"title":"Multi-parameter optimization of NPP simulation models using enhanced particle swarm method","authors":"Zikang Li ,&nbsp;Hang Wang ,&nbsp;Li Fei ,&nbsp;Minjun Peng ,&nbsp;Zhang Xian ,&nbsp;Gui Zhou","doi":"10.1016/j.pnucene.2025.105671","DOIUrl":"10.1016/j.pnucene.2025.105671","url":null,"abstract":"<div><div>This paper delves into the optimization of simulation models for large-scale complex dynamic systems that couple multiple disciplines such as nuclear physics, heat transfer, and fluid mechanics, within the context of digital transformation in nuclear power. An enhanced particle swarm optimization (PSO) algorithm-based multi-parameter optimization method is proposed. This method integrates various strategies to improve the simulation accuracy of system-level models in replicating the operational characteristics of real systems. The effectiveness of this method is demonstrated through experiments on simulation models of the reactor coolant system and the chemical and volume control system within a full-range simulator. Post-optimization, the errors of key parameters are reduced to within 2%. This approach not only aids researchers in refining parameter design during the model development phase but also enables automatic parameter adjustments based on the actual system status after deployment. It meets the needs for online optimization and rapid tracking of actual system states in the application of nuclear power digital twin models.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105671"},"PeriodicalIF":3.3,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143619888","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 and discussion on the effect of temperature on steam explosion in seawater and water","authors":"Jinkun Min,&nbsp;Yaopeng Gong,&nbsp;Qiang Guo,&nbsp;Chuanjun Li,&nbsp;Yidan Yuan","doi":"10.1016/j.pnucene.2025.105726","DOIUrl":"10.1016/j.pnucene.2025.105726","url":null,"abstract":"<div><div>Seawater may be used for emergency cooling during a severe accident, however this could lead to a steam explosion in seawater. In order to study the effect of seawater on steam explosion, the steam explosion experiments of molten metal in water and seawater at different temperatures and experiments were carried out in seawater with different salinity. Comparison of explosion characteristics was evaluated by using pressure impulse, energy conversion ratio and residence depth. The experiment shows that seawater has a significant effect on the steam explosion process. Steam explosion in seawater occurs earlier and shallower, and can happened over a wider range. An equivalent minimum film boiling temperature theory is proposed through the experiment of vapor film on the surface of solid spheres, which explains the influence of temperature, different phenomenon in seawater and the influence of salinity. The concept of transition zone of the influence of temperature is proposed and the uncertainty in the experiment is explained. These findings have important implications for understanding and predicting steam explosion phenomena in various coolant environments.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105726"},"PeriodicalIF":3.3,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610727","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":"Interfacial behaviors, kinetic analysis and dynamic parameters of disturbance waves in 3 × 3 rod bundle channels","authors":"Guangyuan Jin,&nbsp;Jinghu Bai,&nbsp;Weilian Li,&nbsp;Hongye He,&nbsp;Yandong Hou","doi":"10.1016/j.pnucene.2025.105727","DOIUrl":"10.1016/j.pnucene.2025.105727","url":null,"abstract":"<div><div>To facilitate operational monitoring and accident management in various nuclear power plants, it is essential to investigate the interfacial behaviors, parameter characteristics, and formation mechanisms of annular flow in rod bundle channels. A visual experimental system was developed to record and classify the interfacial behaviors of disturbance waves on side and corner rods. The results indicated that the disturbance waves can be categorized into Small-scale waves, Bag-crest waves, and Ligament-crest waves. The flow map of the two channels indicated that wave structures with lower heights in Channel-1 appeared earlier when the liquid phase velocity remained constant. The peak values in the Power Spectral Density (PSD) results reflect the type and energy of various waves. The highest peak value gradually increases as the wave structure transitions from SS, BC, LC to LC-ll waves. The increasing texture lines with the highest values in the RP results indicate a distinct periodicity of BC waves. Furthermore, the periodicity of BC waves is more pronounced on the corner rod than on the side rod. The MPE value increases as wave types change from Small-scale to Single-crest waves. Additionally, the value of liquid film thickness are generally higher than that of pressure drop. The wave height, velocity, and frequency under different flow conditions were presented and compared. The results indicated that variations in dynamic parameters are closely related to the characteristics of disturbance wave structures.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105727"},"PeriodicalIF":3.3,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143600947","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":"Full-scale demonstration of human reliability analysis framework for TRIGA research reactor","authors":"Wasin Vechgama ,&nbsp;Jinkyun Park ,&nbsp;Saensuk Wetchagarun ,&nbsp;Anantachai Pechrak ,&nbsp;Weerawat Pornroongruengchok ,&nbsp;Kampanart Silva","doi":"10.1016/j.pnucene.2025.105718","DOIUrl":"10.1016/j.pnucene.2025.105718","url":null,"abstract":"<div><div>The Thailand Institute of Nuclear Technology (TINT) and the Korea Atomic Energy Research Institute (KAERI) together developed the human reliability analysis (HRA) framework for the TRIGA research reactor based on the EMpirical data-Based crew Reliability Assessment and Cognitive Error analysis (EMBRACE) and TAsk COMplexity (TACOM) methods with the Human Reliability data EXtraction (HuREX) database for extracting human errors and estimating the human error probabilities (HEPs) of actions during the implementation of emergency operating procedures (EOPs). This study provides a full-scale demonstration of how to use the HRA framework through the Thai Research Reactor-1/Modification 1 (TRR-1/M1) case study to extract human errors of actions and estimate the overall HEPs of the steps and tasks in EOPs. Human error of Type C was mainly gathered as the data source from observations of TRR-1/M1's emergency training. Application of the TACOM method systematically and consistently improved the EOPs by providing how to identify the primitive tasks of human errors using the double-column procedure, the same standard as in nuclear power plants. The HRA event tree from the Technique for Human Error Rate Prediction (THERP) was used to consider human errors without machine or system failures to represent realistic human errors in the estimation of nominal human error probabilities (NHEPs). The highest NHEP task was found to be Task 8 of the loss of coolant accident (LOCA) EOP in error of omission (EOO) mode at 6.929E-01 due to multiple manual operations of valves. Additionally, consideration of performance shaping factors (PSF) significantly increased the HEPs when compared to their NHEPs due to the effects of the complexity of the required task and subjective stress.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105718"},"PeriodicalIF":3.3,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610726","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Power-maximization of an irreversible simple Brayton cycle space nuclear power plant","authors":"Lingen Chen ,&nbsp;Tan Wang ,&nbsp;Yanlin Ge ,&nbsp;Huijun Feng","doi":"10.1016/j.pnucene.2025.105708","DOIUrl":"10.1016/j.pnucene.2025.105708","url":null,"abstract":"&lt;div&gt;&lt;div&gt;A simple Brayton cycle space power plant includes two parts: closed Brayton cycle with a compressor, a turbine and two heat exchangers, and radiator panel to dissipate heat to cosmic space. A model of simple irreversible closed Brayton cycle space power plant is established by utilizing finite-time thermodynamics herein, cycle thermal efficiency and cycle power output are deduced and optimized. When heat transfer areas of two heat exchangers and radiator panel are &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mn&gt;15.7&lt;/mn&gt;&lt;msup&gt;&lt;mi&gt;m&lt;/mi&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;F&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;/msub&gt;&lt;mo&gt;=&lt;/mo&gt;&lt;mn&gt;122.4&lt;/mn&gt;&lt;msup&gt;&lt;mi&gt;m&lt;/mi&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/msup&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, and low temperature heat sink is &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = 490 &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;K&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, cycle power of initial design scheme is &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = 33.72 &lt;span&gt;&lt;math&gt;&lt;mtext&gt;kW&lt;/mtext&gt;&lt;/math&gt;&lt;/span&gt;. When three area distributions (&lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;f&lt;/mi&gt;&lt;mi&gt;H&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;f&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;f&lt;/mi&gt;&lt;mi&gt;R&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;) are optimized and &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = 490 &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;K&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, the maximum cycle power is &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = 34.75 &lt;span&gt;&lt;math&gt;&lt;mtext&gt;kW&lt;/mtext&gt;&lt;/math&gt;&lt;/span&gt;, with an increase of about 3.05% compared with &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;. When &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;T&lt;/mi&gt;&lt;mi&gt;L&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; is further optimized, the double maximum cycle power is &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;mo&gt;,&lt;/mo&gt;&lt;mn&gt;2&lt;/mn&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; = 39.45 &lt;span&gt;&lt;math&gt;&lt;mtext&gt;kW&lt;/mtext&gt;&lt;/math&gt;&lt;/span&gt;, with an increase of about 13.53 % compared with &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;, and an increase of about 17 % compared with &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt;. The curve between &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and the corresponding efficiency &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;o&lt;/mi&gt;&lt;mi&gt;p&lt;/mi&gt;&lt;mi&gt;t&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; is loop-shape one, that is, there is the maximum optimal efficiency &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;o&lt;/mi&gt;&lt;mi&gt;p&lt;/mi&gt;&lt;mi&gt;t&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/math&gt;&lt;/span&gt; and the corresponding power output &lt;span&gt;&lt;math&gt;&lt;mrow&gt;&lt;msub&gt;&lt;mi&gt;P&lt;/mi&gt;&lt;msub&gt;&lt;mrow&gt;&lt;mo&gt;(&lt;/mo&gt;&lt;msub&gt;&lt;mi&gt;η&lt;/mi&gt;&lt;mrow&gt;&lt;mi&gt;o&lt;/mi&gt;&lt;mi&gt;p&lt;/mi&gt;&lt;mi&gt;t&lt;/mi&gt;&lt;/mrow&gt;&lt;/msub&gt;&lt;mo&gt;)&lt;/mo&gt;&lt;/mrow&gt;&lt;mi&gt;max&lt;/mi&gt;&lt;/msub&gt;&lt;/msub&gt;&lt;/mrow&gt;&lt;/m","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105708"},"PeriodicalIF":3.3,"publicationDate":"2025-03-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143610785","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":"Research on fast prediction of boiling flow parameters in rod bundle of NPP based on efficient CFD-ROM methods","authors":"Jinhong Mo ,&nbsp;Xiaomeng Dong ,&nbsp;Yujie Xu ,&nbsp;Anqi Xu ,&nbsp;Yang Yu ,&nbsp;Ming Yang","doi":"10.1016/j.pnucene.2025.105723","DOIUrl":"10.1016/j.pnucene.2025.105723","url":null,"abstract":"<div><div>Historically, Computational Fluid Dynamics (CFD) has been widely used to verify the flow dynamics in rod bundle channels. Nevertheless, the iterative calculation and time consumption make it impractical for the application of digital twin that require efficiency. In order to tackle this difficulty, we suggest utilizing a data-driven order reduced model (ROM) to rapidly predict flow fields of rod bundle channels. Previous research on Reduced Order Models (ROM) for rod bundle channel has primarily focused on the single-phase state, with little emphasis on the two-phase state. Thus, the study focuses on the ROM of rod bundle channel in a two-phase state. First, a Computational Fluid Dynamics (CFD) data set is processed using the Proper Orthogonal Decomposition (POD) algorithm to identify important modes. Then, a Back-Propagation Neural Network (BPNN) model is trained as the agent model, due to its strong ability to fit non-linear relationship between input and output. The structure of BPNN is optimized. Ultimately, the ROM model is utilized to predict the flow field of a rod bundle channel under new boundary conditions. The comparison with CFD calculation confirms the effectiveness of the Reduced Order Model (ROM), showcasing its high precision in predicting the temperature, velocity, and void fraction fields within the rod bundle channel. In addition, the ROM model reaches a computation speed that is roughly 10^4 times quicker than standard CFD simulations. Nonetheless, it is noted that the model's predictive accuracy diminishes under conditions of slight subcooled boiling, due to the lack of samples under the condition. An improvement strategy: Subdivide the sampling space region according to various change stages, implement partition sampling.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105723"},"PeriodicalIF":3.3,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591968","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 gap size and inclination angle on Critical Heat Flux (CHF) of pool boiling based on bubble behavior observations","authors":"Yanxu Huang ,&nbsp;Ken Chen ,&nbsp;Junfeng Li ,&nbsp;Zihao Wang ,&nbsp;Kai Wang ,&nbsp;Shixian Wang","doi":"10.1016/j.pnucene.2025.105711","DOIUrl":"10.1016/j.pnucene.2025.105711","url":null,"abstract":"<div><div>In this study, the effect of gap size and inclination angles on the Critical Heat Flux (CHF) on bare copper surfaces is investigated in the context of vertical-facing saturated pool boiling. The research endeavor aims to unravel the intricate relationships between these variables and their collective impact on CHF. Through hypothesis formulation and analysis of the collected experimental data, it has been determined that the effects of the two variables, inclination angle and gap size, on CHF are approximately relatively independent. Based on previous research, experimental data has validated the conclusions that CHF increases with both increasing inclination angle and gap size. Furthermore, an empirical formula has been developed that closely aligns with experimental data, incorporating both contemporary experimental findings and established theoretical models. Comparison with previous data and models has shown that it fits well with the derived formula. The author argues that both variables—inclination angle and gap size—have a significant impact on CHF by modulating the Bubble Film Departure Frequency (BFDF) and the average width of bubbles formed during the boiling process. However, these two variables do not directly affect CHF, instead, they influence CHF by affecting bubble behavior. Bubble behavior is the direct factor in the occurrence of CHF, while gap size and inclination angle are indirect factors. Through analyzing bubble behavior, we can gain a better understanding of how gap size and inclination angle influence the average bubble width and BDFM, as well as how the average bubble width and BDFM affect CHF. This research offers reference for IVR strategy implementation: the In-vessel Corium Retention (IVR) strategy, crucial for mitigating core meltdown accidents, ensures the reactor pressure vessel (RPV) lower head remains intact during severe incidents. The Critical Heat Flux (CHF) on the RPV lower head surface is the main barrier to IVR. Consequently, CHF research is essential for IVR application. While there's extensive research on heating surface inclination and gap size effects on CHF, less focuses on bubble behavior. Instead, they influence CHF by affecting bubble behavior. Bubble behavior is the direct factor in the occurrence of CHF, while gap size and inclination angle are indirect factors. Through analyzing bubble behavior, we can gain a better understanding of how gap size and inclination angle influence the average bubble width and BDFM, as well as how the average bubble width and BDFM affect CHF.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105711"},"PeriodicalIF":3.3,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591964","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":"Optimizing nuclear power plant fault diagnosis through an integrated operational support approach","authors":"Yue Yu ,&nbsp;De-shui Han ,&nbsp;Zeyun Yang ,&nbsp;Min-jun Peng ,&nbsp;Zhanguo Ma ,&nbsp;Ning Li","doi":"10.1016/j.pnucene.2025.105722","DOIUrl":"10.1016/j.pnucene.2025.105722","url":null,"abstract":"","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105722"},"PeriodicalIF":3.3,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591969","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 of spacer effects on post-dryout heat transfer in a tight-lattice rod bundle","authors":"Shuo Chen ,&nbsp;Chen Zeng ,&nbsp;Yao Xiao ,&nbsp;Hanyang Gu","doi":"10.1016/j.pnucene.2025.105713","DOIUrl":"10.1016/j.pnucene.2025.105713","url":null,"abstract":"<div><div>Tight-lattice fuel assembly is an advanced design for a compact water-cooled reactor core, but it presents challenges during loss of flow accidents, potentially suffering from critical heat flux (CHF). Downstream of the CHF point, dispersed flow with steam-entrained droplets becomes critical, significantly deteriorating heat transfer abilities. This paper experimentally studies the heat transfer capabilities downstream of spacer grids under post-dryout dispersed flow conditions. Experiments were conducted based on the high-temperature, high-pressure open-loop system at the Reactor Thermal-Hydraulics Laboratory of Shanghai Jiao Tong University. The working medium is water and the conditions tested included pressures ranging from 6 to 10 MPa, mass fluxes between 65 and 200 kg⸱m⁻²⸱s, heat fluxes from 75 to 200 kW⸱m⁻², and inlet qualities between 0.543 and 0.887. The heat transfer enhancement ratio is defined to assess the impact of spacer grids. It is shown that the heat transfer enhancement ratio is influenced by quality, mass flux, and system pressure. Specifically, spacer grids exhibit better heat transfer enhancement capabilities under conditions of low qualities, low mass flow rates, and high pressures. The circumferential wall temperature distribution is quite uniform before dryout, while local hot spots appear near the 30°, 90°, and 180° directions after dryout. Spacer grids exhibit optimal heat transfer enhancement under low qualities, low mass flow rates, and high pressure conditions, with the maximum effect observed at the spacer grid outlet, decreasing exponentially thereafter. Available correlations for spacer grid heat transfer characteristics are evaluated, with a novel correlation according to experimental data proposed. This novel correlation closely aligns with experimental results, maintaining an error range within ±15%.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105713"},"PeriodicalIF":3.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579287","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":"Corrosion mechanisms and differences of Inconel 617 and Incoloy 800H under high-temperature air ingress accident","authors":"Haoxiang Li ,&nbsp;Dapeng Yan ,&nbsp;Guodong Li ,&nbsp;Changqing Tian ,&nbsp;Jiayi An ,&nbsp;Jiaqi Tan ,&nbsp;Wei Zheng ,&nbsp;Bin Du ,&nbsp;Huaqiang Yin","doi":"10.1016/j.pnucene.2025.105725","DOIUrl":"10.1016/j.pnucene.2025.105725","url":null,"abstract":"<div><div>In the high-temperature operation of very high-temperature gas-cooled reactors (VHTRs), \"microclimate reaction\" are one of the key corrosion reactions of concern. This study investigates the corrosion behavior and mechanisms of the two mainstream high-temperature alloys, Incoloy 800H and Inconel 617, under minor air ingress accidents. The main focus is on comparing and analyzing the \"microclimate reaction\" corrosion behavior and mechanisms of the two alloys, as well as exploring changes in their mechanical properties. The results show that during the experiment, both alloys primarily underwent oxidation and \"microclimate reaction\". The double-layer oxide of Cr₂O₃-MnCr₂O₄/MnFe₂O₄ and the SiO₂ layer on the surface of Incoloy 800H provide good resistance to \"microclimate reaction,\" whereas the single-layer Cr₂O₃ oxide on Inconel 617 is not effective in preventing \"microclimate reaction\". The mechanical property test results indicate that the decarburization of the alloy caused by the \"microclimate reaction\" does not significantly affect the overall basic mechanical properties of the alloy on a macroscopic level. Research on the \"microclimate reaction\" should primarily focus on preventing the oxide layer from cracking and reducing the time during which the alloy matrix is unprotected by a stable oxide layer.</div></div>","PeriodicalId":20617,"journal":{"name":"Progress in Nuclear Energy","volume":"184 ","pages":"Article 105725"},"PeriodicalIF":3.3,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143591962","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}