Annals of Nuclear Energy最新文献

{"title":"Research on optimization of steam generator level control in startup process based on intelligent algorithm","authors":"Qi Zhang ,&nbsp;Jiajun Huang ,&nbsp;Wenhao Yu ,&nbsp;Siqi Jin ,&nbsp;Yulong Wang ,&nbsp;Xinyu Wei ,&nbsp;Peiwei Sun","doi":"10.1016/j.anucene.2025.111835","DOIUrl":"10.1016/j.anucene.2025.111835","url":null,"abstract":"<div><div>Automatic steam generator (SG) level control during pressurized water reactor (PWR) startup is crucial for safety. The poor adaptability of fixed-parameter SG level controllers of main feedwater system (TFM), the manual SG level control by auxiliary feedwater system (TFA), and the manual switching between TFA and TFM, have limitations for the startup process automation. An intelligent control scheme is proposed, integrating gain-scheduled proportional-integral-derivative (PID) controller tuning via non-dominated sorting genetic algorithm II and fuzzy logic-based switching scheme. Simulation results demonstrate that the optimized TFM level control scheme reduces the average settling time by 47.3% under load step disturbance conditions, and enhances disturbance rejection. The designed TFA level control system can maintain level stability during startup. The fuzzy controller achieves bumpless switching between systems. The approach reduces average maximum level deviation by 39.2% and average settling time by 13.4% in startup processes, offering a practical solution to enhance PWR automation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111835"},"PeriodicalIF":2.3,"publicationDate":"2025-09-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144932555","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":"NUMERICAL STUDY ON THERMAL BEHAVIOUR AND MODERATOR BOILING IN TWO NUCLEAR CHANNELS OF IPHWR 220 MWE REACTOR AT DIFFERENT HEAT INPUTS DURING LOSS OF COOLANT ACCIDENT (LOCA)","authors":"Raushan Kumar,&nbsp;Chandranshu Kumar Singh,&nbsp;Anurag Singh,&nbsp;Mushtaque Momin,&nbsp;Mukesh Sharma","doi":"10.1016/j.anucene.2025.111847","DOIUrl":"10.1016/j.anucene.2025.111847","url":null,"abstract":"<div><div>Indian PHWRs, derived from CANDU designs, are vulnerable to postulated accident scenarios like LOCA with simultaneous ECCS failure. To investigate such conditions, this study analyses the temperature distribution in the pressure tube (PT) and calandria tube (CT) under voided channel conditions using steady-state thermal analysis and ANSYS Fluent at various heat fluxes (900, 1500, 2500, 3500, and 4500 W/m²). The PT showed negligible circumferential temperature variation, while the CT exhibited significant heating near neighbouring channels. As heat flux increased, the moderator surrounding the CT progressively heated and eventually began to boil beyond 1500  W/m². Boiling was more intense near the lower region of channel 1 and the upper region of channel 2. The trend in boiling behaviour remained consistent across different heat fluxes, although the onset occurred earlier at higher flux levels. These findings highlight critical thermal behaviour under accident-like conditions and can support safety evaluations of reactor designs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111847"},"PeriodicalIF":2.3,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144921909","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 metal fission products on the thermal transport of SiC/PyC nanocomposites: Insight from molecular dynamics simulations","authors":"Qian Wang ,&nbsp;Nan Gui ,&nbsp;Yiyang Luo ,&nbsp;Xingtuan Yang ,&nbsp;Jiyuan Tu ,&nbsp;Shengyao Jiang","doi":"10.1016/j.anucene.2025.111849","DOIUrl":"10.1016/j.anucene.2025.111849","url":null,"abstract":"<div><div>The High-Temperature Gas-cooled Reactor (HTGR) is a generation-IV advanced nuclear reactor, which has a special nuclear fuel design, and the tiny TRISO particle (∼1 mm) is adopted. Each TRISO particle is coated with four layers, and silicon carbide (SiC) and pyrolytic carbon (PyC) are the two main components. The heat transfer process at the SiC/PyC interface is important to compute the temperature distribution inside the TRISO particle, but it is quite difficult to research this phenomenon based on experimental results. However, the Molecular dynamics (MD) method could be seen as a viable simulation scheme in micro-scale phenomena. The non-equilibrium molecular dynamics (NEMD) was employed to compute the temperature profile and interfacial resistance of SiC/PyC nanocomposites. In addition, seven atomic models embedded with noble metal fission products, including silver (Ag), palladium (Pd), and ruthenium (Ru), were built, with both aggregated nanoparticle state and atomically dispersed state being investigated. The phonon density of states was used to quantify differences in heat transfer performance. The results show that the Kapitza resistance of the SiC/PyC interface decreases gradually with increasing temperature due to the weakening of phonon scattering, but the FP atoms could reduce the heat transfer capability of SiC/PyC nanocomposites; the dispersed atoms have a more significant effect than the FP nanoparticle as an integral part. This study reveals the mechanism underlying the fission products’ influence on the heat transfer characteristics of TRISO coating layers, providing a theoretical basis for enhancing the comprehension of their interaction.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111849"},"PeriodicalIF":2.3,"publicationDate":"2025-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144926169","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":"Verification of multi-scale coupling program for high temperature gas-cooled reactor based on metamorphic testing","authors":"Yin Zhao ,&nbsp;Meng Li ,&nbsp;Ke Zhang ,&nbsp;Xiaohua Yang ,&nbsp;Jie Liu ,&nbsp;Shiyu Yan","doi":"10.1016/j.anucene.2025.111846","DOIUrl":"10.1016/j.anucene.2025.111846","url":null,"abstract":"<div><div>The multi-scale coupling program for high temperature gas-cooled reactors encompasses complex physical phenomena across the microscopic, mesoscopic, and macroscopic level. Owing to the significant development expenses and the complexity of forming precise analytical solutions, making traditional testing methods invalid, verifying multi-scale codes is hindered by the oracle problem. Metamorphic testing is an effective technique to alleviate the oracle problem. This study uses a two-stage verification method grounded in metamorphic relations, following the introduction of code verification in the nuclear domain. Upon identifying 13 metamorphic relations and 1 property based on fundamental physical characteristics, 87 test case pairs successfully revealed two deeply hidden faults undetected by traditional testing methods. The experimental findings indicate that metamorphic testing serves both as a mechanism to evaluate the code correctness and as a technique to increase the number of verification cases. Furthermore, it presents great potential for applications in the verification of nuclear software.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111846"},"PeriodicalIF":2.3,"publicationDate":"2025-08-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144919941","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 and characterization of the vibration fatigue S-N curve for small-diameter 316L stainless steel butt-welded pipes","authors":"Ning Li ,&nbsp;Hongbo Gao ,&nbsp;Lei Lin ,&nbsp;Pin Li ,&nbsp;Shuai Zhou ,&nbsp;Han Liu ,&nbsp;Changning Li ,&nbsp;Decheng Xu ,&nbsp;Yin Liu ,&nbsp;Zhilin Chen","doi":"10.1016/j.anucene.2025.111838","DOIUrl":"10.1016/j.anucene.2025.111838","url":null,"abstract":"<div><div>Vibration fatigue in small-diameter pipes (SDP) represents a key failure mode in industrial pipeline systems, with butt-welded joints constituting the critical weak links. This study addresses the lack of specific component-level vibration fatigue data for 316L stainless steel butt-welded SDP and the insufficient accuracy of traditional evaluation methods (based on base material data with generic Fatigue Strength Reduction Factors), aiming to experimentally obtain and characterize the vibration fatigue S-N curves for these components.The experimental methodology employed an electromagnetic vibration table to conduct component-level resonance bending fatigue tests on 316L butt-welded pipe specimens with an outer diameter (OD) of 16 mm and a wall thickness of 3 mm. Constant amplitude loading was applied at the first-order bending resonance frequency of the specimens (approximately 58 Hz), with stress amplitude monitored and controlled via strain gauges. Failure detection was implemented through gas leakage monitoring. Residual stresses after welding and dynamic stress concentration factors (SCF) were measured using X-ray diffraction (XRD) and digital image correlation (DIC) techniques, respectively. The results successfully established the component-level S-N curve for the joint, determining the median fatigue strength at 10⁷ cycles (162.5 MPa) and the lower limit of design fatigue strength (148.7 MPa). All failures initiated at the weld toe, where SCF values ranged from 1.26 to 1.47. Significant residual compressive stresses were identified on the outer surface of the weld, exhibiting a negative correlation with weld reinforcement height. Under medium and high stress conditions, fatigue life decreased with increasing reinforcement height. Compared to ASME standard reference data, the component-level S-N curves developed in this study exhibited superior fatigue resistance in the medium–high cycle and fatigue limit regions. The component-level S-N curves obtained in this research incorporate the influences of actual weld geometry, residual stress distribution, and microstructural characteristics, providing a more accurate and reliable fatigue assessment basis for 316L butt-welded small-diameter pipes. These findings support optimized design approaches and provide a reference for evaluating the conservatism of standard assessment methods, which has significant implications for improving the safety and reliability of pipeline systems.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111838"},"PeriodicalIF":2.3,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144917976","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":"Optimization of UHPC with basil plant ash: Impacts on strength, durability, and gamma-ray attenuation","authors":"Sahar A. Mostafa ,&nbsp;Islam N.Fathy ,&nbsp;Alaa A. Mahmoud ,&nbsp;Mohamed A. Abouelnour ,&nbsp;K.A. Mahmoud ,&nbsp;Shaaban M. Shaaban ,&nbsp;Sameh A. Elhameed ,&nbsp;Islam M. Nabil","doi":"10.1016/j.anucene.2025.111825","DOIUrl":"10.1016/j.anucene.2025.111825","url":null,"abstract":"<div><div>This study explored the usability of basil plant ash (BPA) as an eco-friendly supplementary cementitious material for ultra-high-performance concrete (UHPC). This study sought to determine the impact of BPA on the mechanical properties, durability, radiation-shielding ability, environmental sustainability, and economic efficiency of UHPC. The control mix was free of BPA, whereas the investigated mixes were set at 10–40% replacement of cement weight. The concrete composite was examined using scanning electron microscopy(SEM) and X-ray diffraction(XRD) analysis. The radiation protection properties against γ-rays and neutrons were evaluated using Monte Carlo(MC) simulations and the Phy-X online software. The environmental impact was assessed by calculation of carbon dioxide emissions resulting from the concrete production process. An environmental impact assessment revealed that incorporating 30% BPA reduced carbon dioxide emissions by 27% during concrete production. These findings highlight the potential of BPA as an eco-friendly additive to enhance the mechanical, durability, and radiation-shielding properties of UHPC while reducing its environmental impact and production costs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111825"},"PeriodicalIF":2.3,"publicationDate":"2025-08-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144917977","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 and numerical assessment of sodium aerosol behaviour during pool fire in a rectangular chamber","authors":"Amit Kumar ,&nbsp;Parthkumar Rajendrabhai Patel ,&nbsp;Usha Pujala ,&nbsp;E. Hemanth Rao ,&nbsp;Sanjay Kumar Das ,&nbsp;A. John Arul","doi":"10.1016/j.anucene.2025.111831","DOIUrl":"10.1016/j.anucene.2025.111831","url":null,"abstract":"<div><div>Understanding the spatio-temporal transport of the aerosol resulting from a sodium fire is essential for mechanistic assessment of the severe accident source term. To develop mechanistic models for evaluating aerosol transport, the Indira Gandhi Center for Atomic Research (IGCAR) has started a series of experiments to understand the spatio-temporal behaviour of the aerosol in a large volume. These experiments involved the measurement of the temperature, humidity and aerosol characteristics at multiple locations, which improved the understanding of aerosol spatial dispersion during typical sodium fires. In the first experiment, a pool fire was created using 2 kg of sodium in the MINA test chamber to obtain the spatial dispersion of the aerosol over time along with the temperature distribution. Before attempting a detailed mechanistic assessment of aerosol transport, lumped codes were developed in-house to assess the sodium pool, average gas and wall temperatures, pressures, and aerosol characteristics. Results from the experiment indicate spatial heterogeneity in the aerosol concentration across different elevations of the MINA test chamber, and there is a need for CFD-based assessment. The lumped analysis agrees well with the average temperatures, median diameters, and mass concentration. The maximum concentration of sodium aerosols in the chamber is approximately <span><math><mo>∼</mo></math></span> 3.2g/m<span><math><msup><mrow></mrow><mrow><mn>3</mn></mrow></msup></math></span>. The median diameter of the initial aerosol increases from <span><math><mrow><mn>0</mn><mo>.</mo><mn>2</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span> to <span><math><mrow><mn>5</mn><mo>.</mo><mn>5</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span> after 50 min of sodium fire. The median diameter of the aerosol decreases to <span><math><mrow><mn>3</mn><mspace></mspace><mi>μ</mi><mi>m</mi></mrow></math></span> after 180 min of the sodium fire. The details of the experiments and significant findings are discussed in the present manuscript.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111831"},"PeriodicalIF":2.3,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144911766","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":"Graphite waste classification and disposal cost estimation for high temperature gas and salt reactors","authors":"Liam Hines ,&nbsp;Koroush Shirvan ,&nbsp;Per Peterson ,&nbsp;Lance Snead","doi":"10.1016/j.anucene.2025.111833","DOIUrl":"10.1016/j.anucene.2025.111833","url":null,"abstract":"<div><div>As high-temperature reactor designs progress to demonstration, managing the radioactive wastes from these systems presents unique challenges. This work explores the irradiated graphite source term produced by three reactor designs: The Modular High Temperature Gas reactor (MHTGR), a pebble-bed High Temperature Gas Reactor (pb-HTGR), and a Fluoride-cooled High-temperature Reactor (FHR). We predicted a C-14 concentration of 4.3 Ci/m³ for the MHTGR, 1.2 Ci/m³ for the pebble bed HTGR, and 2.5 Ci/m³ for the gFHR after 20 years of operation. The final C-14 concentration highly depended on the graphite nitrogen impurity, a major precursor for C-14. The C-14 concentration in all reactor types exceeded the 0.8 Ci/m³ threshold, resulting in a Class C waste classification. The costs associated with accepting the graphite after 20 years in a low-level waste disposal facility were projected to be 255 $/kWe for the MHTGR, 248 $/kWe for the pb-HTGR, and 56.8 $/kWe for the FHR.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111833"},"PeriodicalIF":2.3,"publicationDate":"2025-08-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144917975","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":"Noise characteristics and control of regulator spray valve: A hybrid computational and machine learning approach","authors":"Hongwei Xu ,&nbsp;Yufeng Lang ,&nbsp;Jianian Chang ,&nbsp;Xizhuo Le ,&nbsp;Jianfeng Mao","doi":"10.1016/j.anucene.2025.111809","DOIUrl":"10.1016/j.anucene.2025.111809","url":null,"abstract":"<div><div>The flow-induced noise generated during the operation of the regulator spray valve, a critical component for pressure control in pressurized water reactor nuclear power plants, can significantly impact its reliability. In this study, Large Eddy Simulation (LES) is coupled with Lighthill’s acoustic analogy to model the noise-generation mechanism, showing that the sound arising from vortex structures and turbulent pulsations exhibits distinct frequency-domain characteristics. Machine learning is then employed to optimize a multi-objective noise-reduction design for the downstream orifice plate; this yields a 12.4 dB(A) noise reduction while decreasing the flow coefficient by only 7.5%. The optimized configuration increases structural safety without impairing valve operability, offering critical insights for enhancing performance and extending service life.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111809"},"PeriodicalIF":2.3,"publicationDate":"2025-08-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902189","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":"Comparative thermal-hydraulic analysis of parallel and counter flow configurations in a dual fluid reactor mini demonstrator","authors":"Hisham Lotfy Elgendy ,&nbsp;Konrad Czerski","doi":"10.1016/j.anucene.2025.111845","DOIUrl":"10.1016/j.anucene.2025.111845","url":null,"abstract":"<div><div>This article explores the comparative thermal–hydraulic behavior of parallel and counter flow configurations within the mini demonstrator (MD) of the Dual Fluid Reactor (DFR). Utilizing detailed computational fluid dynamics (CFD) simulations, we analyze the heat transfer characteristics, velocity distribution, and swirling effects for both configurations. Given the uniquely low Prandtl number of the liquid lead used in the MD and DFR, the CFD modeling in this study incorporates a variable turbulent Prandtl number. This approach has been validated in our previously published work. The results demonstrate that the counter flow configuration yields higher heat transfer efficiency and more uniform flow velocity, while reducing swirling and mechanical stresses. These findings provide valuable insights for optimizing the DFR’s design, improving safety, and enhancing operational performance.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111845"},"PeriodicalIF":2.3,"publicationDate":"2025-08-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144902188","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}