Nuclear Engineering and Design最新文献

Preliminary design and neutronic characterisation of a 200 kWt HALEU fueled heat-pipe reactor for space applications 用于空间应用的200kwt高浓铀燃料热管反应堆的初步设计和中子特性

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-11 DOI: 10.1016/j.nucengdes.2025.114521

Riccardo Boccelli , Andrea D’Ottavio , Stefano Lorenzi , Angelica Peressotti , Maria Antonietta Perino , Marco Enrico Ricotti , Lorenzo Tutolo

{"title":"Preliminary design and neutronic characterisation of a 200 kWt HALEU fueled heat-pipe reactor for space applications","authors":"Riccardo Boccelli , Andrea D’Ottavio , Stefano Lorenzi , Angelica Peressotti , Maria Antonietta Perino , Marco Enrico Ricotti , Lorenzo Tutolo","doi":"10.1016/j.nucengdes.2025.114521","DOIUrl":"10.1016/j.nucengdes.2025.114521","url":null,"abstract":"<div><div>Nuclear reactors represent a key technology for advancing space exploration and utilisation, providing a reliable, continuous, and environmentally independent energy source essential for long-duration missions beyond Earth. This capability is crucial for ensuring uninterrupted operation of spacecraft systems, sustaining lunar habitats, and supporting energy-intensive scientific experiments. Nuclear energy, particularly fission, is renowned for its high energy density and reliability, making it a key element in the design of space missions where mass minimisation is a critical requirement for feasibility and economic viability. While designs like KRUSTY have demonstrated compactness and simplicity, it is desirable to develop reactors with higher power levels, lower enrichment, and compatibility with non-proliferation concepts.</div><div>The objective of this work is to propose a preliminary design for a heat-pipe reactor conceived for both lunar surface applications and low-power electric propulsion and to conduct a comprehensive neutronic analysis. The proposed reactor combines the simplicity of a KRUSTY-type reactor with increased power to 200 kWt (40-50 kWe assuming 20%–25% conversion efficiency) and reduced enrichment (from HEU to HALEU). The preliminary design is characterised by an epithermal spectrum, a nominal operating temperature between 1000 K and 1100 K, and has a mass of 1139 kg, resulting in a specific power of 176 Wt/kg. A neutronic analysis was performed to characterise the reactor, extracting information such as power and flux distributions, feedback coefficients, reactivity control, burnup, and addressing safety aspects, demonstrating that the reactor maintains sufficient reactivity margin under certain accidental criticality condition. All the criticality calculations were performed using OpenMC Monte Carlo code.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114521"},"PeriodicalIF":2.1,"publicationDate":"2025-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266807","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}

引用次数: 0

Effect of thermal ageing on mechanical, fatigue and fracture properties of nuclear power plant piping materials: a review 热老化对核电站管道材料力学、疲劳和断裂性能的影响

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-11 DOI: 10.1016/j.nucengdes.2025.114528

Shabna Shirin , S. Vishnuvardhan , Keerthy M. Simon

引用次数: 0

Experimental study on seismic sloshing of storage water for spent fuel pool in NPPs 核电站乏燃料池储水地震晃动试验研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-11 DOI: 10.1016/j.nucengdes.2025.114531

Juin-Fu Chai, Fan-Ru Lin, Wei-Hung Hsu, Yi-Jun Kao

{"title":"Experimental study on seismic sloshing of storage water for spent fuel pool in NPPs","authors":"Juin-Fu Chai, Fan-Ru Lin, Wei-Hung Hsu, Yi-Jun Kao","doi":"10.1016/j.nucengdes.2025.114531","DOIUrl":"10.1016/j.nucengdes.2025.114531","url":null,"abstract":"<div><div>Due to the event of Fukushima Daiichi Nuclear Power Plant in 2011, the U.S. Nuclear Regulatory Commission (NRC) requested all U.S. nuclear power plants to conduct seismic hazard re-evaluation per the newest guidance and requirements. Among them, the spent fuel pool (SFP) is considered an important facility. The related evaluation guidance provided in the Electric Power Research Institute EPRI-1025287 report emphasizes failure modes of the SFP that could result in “rapid drain-down”. One of the possible causes is the volume of water splashed out of the pool due to seismic sloshing. In EPRI-1025287 report, a very convenient yet conservative approach for estimating water splash volume was provided. To obtain more accurate results, this study employed a shaking table test of liquid storage tanks to develop a more realistic water splash volume estimation method, based on the relationship between seismic-induced sloshing height and the associated total volume of water splashed out of the tanks. In the development of this method, the use of a single-degree-of-freedom (SDOF) systems to estimate the sloshing time history is also proposed for subsequent splashed volume calculation. Furthermore, the sloshing frequency, which is related to the sloshing height and the splashed water volume, is also a key focus of the study.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114531"},"PeriodicalIF":2.1,"publicationDate":"2025-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266808","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}

引用次数: 0

Experimental study on graphite dust deposition in straight pipes under simulated HTGR operating conditions 高温高温堆模拟工况下直管内石墨粉尘沉积实验研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-11 DOI: 10.1016/j.nucengdes.2025.114526

Shulong Huang , Ziqi Jiang , Lixiao Guo , Jie Kong , Qiyuan Guan , Jialu Li , Yuhang Zhang , Ying Liu , Zhijie Gu

{"title":"Experimental study on graphite dust deposition in straight pipes under simulated HTGR operating conditions","authors":"Shulong Huang , Ziqi Jiang , Lixiao Guo , Jie Kong , Qiyuan Guan , Jialu Li , Yuhang Zhang , Ying Liu , Zhijie Gu","doi":"10.1016/j.nucengdes.2025.114526","DOIUrl":"10.1016/j.nucengdes.2025.114526","url":null,"abstract":"<div><div>The high-temperature gas-cooled reactor (HTGR), especially in the pebble-bed design like China’s HTR-PM, offers key safety and efficiency benefits. However, radioactive graphite dust generated during long-term operation poses safety and maintenance challenges. This study experimentally investigates graphite dust deposition and resuspension in a straight pipeline under varying flow velocities, temperatures, and carrier gases (air and helium). Deposition was measured at upstream, midstream, and downstream locations. Results show that increasing flow velocity significantly reduces deposition, with a 97.6 % decrease at 6 m/s compared to 1.5 m/s and near elimination at 9 m/s. Axial deposition remained consistent, with most dust settling in the front and middle sections. Higher temperatures (25–300 °C) reduced deposition and improved uniformity. Helium, due to its higher viscosity and lower density, led to lower overall deposition than air. Resuspension tests revealed partial particle detachment at high wind speeds (up to 28 m/s), but complete resuspension was not achieved, likely due to strong adhesion forces such as electrostatic interactions. These results provide critical experimental data for CFD model validation and support the development of effective dust control strategies, contributing to safer and more efficient HTGR operation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114526"},"PeriodicalIF":2.1,"publicationDate":"2025-10-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266804","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}

引用次数: 0

Numerical study on the transportation characteristics of nuclide 16N in steam generator 核素16N在蒸汽发生器内输运特性的数值研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114517

Hongming Zhang , Hanrui Qiu , Mingjun Wang , Zili Gong , Wenxi Tian , G.H. Su

{"title":"Numerical study on the transportation characteristics of nuclide 16N in steam generator","authors":"Hongming Zhang , Hanrui Qiu , Mingjun Wang , Zili Gong , Wenxi Tian , G.H. Su","doi":"10.1016/j.nucengdes.2025.114517","DOIUrl":"10.1016/j.nucengdes.2025.114517","url":null,"abstract":"<div><div>The steam generator is a critical component that couples the primary circuit and the secondary circuit of a nuclear reactor. During operation, steam generator tubes are exposed to corrosive environments and high pressures. Since leakage locations are not directly observable and the leakage flow rate from cracks in the heat transfer tubes cannot be measured during operation, leakage incidents are typically diagnosed by sampling emissions from the steam generator exhaust system and analyzing their radioactive activity levels. This study integrates a species transport model into the thermal–hydraulic analysis code STAF (Steam-generator Thermal-hydraulic Analysis code based on Fluent) to investigate the thermal–hydraulic conditions on the secondary side of the steam generator, along with the distribution and concentration evolution of leaked radioactive substance <sup>16</sup>N from the heat transfer tubes, while taking radioactive decay into account. The simulation results reveal the three-dimensional distribution and transient behavior of the leaked substance on the secondary side. Notably, under 50 % power conditions, the equilibrium mass fraction of <sup>16</sup>N exhibits a maximum spatial fluctuation of 38.5 %, indicating that the commonly used assumption of uniform concentration in tube leakage diagnostics may lead to significant errors. The three-dimensional simulation approach developed in this work offers improved accuracy for steam generator tube leakage detection and diagnosis, and provides valuable insights into the transport behavior of radioactive species under realistic operating conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114517"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266801","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}

引用次数: 0

Thermo-neutronic analysis of natural circulation in the ABV small modular reactor under ocean-induced motions 海洋运动下ABV小型模块化反应堆自然循环的热中子分析

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114525

Sayed Abolhasan Nourashrafeddin, Mohsen Shayesteh

{"title":"Thermo-neutronic analysis of natural circulation in the ABV small modular reactor under ocean-induced motions","authors":"Sayed Abolhasan Nourashrafeddin, Mohsen Shayesteh","doi":"10.1016/j.nucengdes.2025.114525","DOIUrl":"10.1016/j.nucengdes.2025.114525","url":null,"abstract":"<div><div>Floating deployment of Small Modular Reactors (SMRs) requires reliable passive cooling performance under dynamic ocean-induced motions. In this work, a comprehensive thermo-neutronic analysis of the ABV reactor was performed to assess the capability of its natural circulation system under inclination, heaving, and rolling conditions. A steady-state coupling between ANSYS CFX and PARCS was implemented, and transient CFD simulations were conducted for dynamic scenarios. Results show that at 30° inclination, the core mass flow decreases by about 9.7 % with a temperature rise of 3.3 K, while at 45° the reduction reaches 19.2 % and 6.9 K, accompanied by a slight decrease in <span><math><mrow><msub><mi>k</mi><mrow><mi>eff</mi></mrow></msub></mrow></math></span> and a shortened cycle length. Under heaving motion, coolant and power oscillations vary between ± 5 % and ± 18 % depending on amplitude and period. Rolling motion produces peripheral flow oscillations with amplitudes below ± 1.5 %, with limited effect on DNBR. These findings demonstrate that ocean-induced motions can significantly affect natural circulation, reactivity feedback, and safety margins, highlighting the necessity of coupled multi-physics modeling and the adoption of power derating and adaptive control strategies for floating nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114525"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267570","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}

引用次数: 0

Oxidation of Pb–Bi eutectic in water-oxygen fluid at elevated temperature and pressure 水-氧流体中铅-铋共晶在高温高压下的氧化

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114537

Oxana N. Fedyaeva, Aleksander P. Grebennikov, Anatoly A. Vostrikov

{"title":"Oxidation of Pb–Bi eutectic in water-oxygen fluid at elevated temperature and pressure","authors":"Oxana N. Fedyaeva, Aleksander P. Grebennikov, Anatoly A. Vostrikov","doi":"10.1016/j.nucengdes.2025.114537","DOIUrl":"10.1016/j.nucengdes.2025.114537","url":null,"abstract":"<div><div>The paper presents the results of an investigation into the oxidation of lead–bismuth eutectic (LBE) by a high-pressure water-oxygen fluid. Tests were conducted by uniformly heating the LBE specimen at a rate of 1 K/min to 873 K in an environment of water vapor, oxygen, and H<sub>2</sub>O/O<sub>2</sub> fluid, as well as by injecting H<sub>2</sub>O/O<sub>2</sub> fluid into the reactor containing the LBE specimen, followed by isothermal holding at a set temperature (623–873 K) for 3 h. The results demonstrate that LBE oxidation is accelerated in both oxygen and H<sub>2</sub>O/O<sub>2</sub> fluid at <em>T</em> > 628 K. Adding water contributes to a multiple increase in the rate of oxidation. The oxidation of LBE by oxygen results in the formation of a dense oxide layer dominated by β-PbO. When oxidized in H<sub>2</sub>O/O<sub>2</sub> fluid, a sponge-like structure forms, whose main component is α-Pb<sub>3</sub>O<sub>4</sub>. The unreacted alloy is enriched in bismuth. Bismuth involvement in oxidation is enhanced at <em>T</em> ≥ 823 K and leads to the formation of plates, as well as feather- and needle-like Bi<sub>2</sub>O<sub>3</sub> structures. The LBE oxidation rate has been found to exhibit a non-monotonic dependence on temperature. It is shown that the process can proceed in the non-activation kinetic mode at 653–723 K. The comparative analysis shows that LBE is more susceptible to oxidation than lead and bismuth individually. These results are important for ensuring the safe operation of lead-cooled nuclear reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114537"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267571","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}

引用次数: 0

Environmentally assisted fatigue design model of thermally aged cast austenitic stainless steel in high-temperature pressurized water 高温加压水中热时效铸造奥氏体不锈钢环境辅助疲劳设计模型

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114504

Yufei Qiao , Hui Zheng , Jibo Tan , Shuangliang Yang , Ziyu Zhang , Jie Li , Xinqiang Wu , Wei Ke

{"title":"Environmentally assisted fatigue design model of thermally aged cast austenitic stainless steel in high-temperature pressurized water","authors":"Yufei Qiao , Hui Zheng , Jibo Tan , Shuangliang Yang , Ziyu Zhang , Jie Li , Xinqiang Wu , Wei Ke","doi":"10.1016/j.nucengdes.2025.114504","DOIUrl":"10.1016/j.nucengdes.2025.114504","url":null,"abstract":"<div><div>Fatigue tests of Z3CN20.09M CASS were carried out in high-temperature pressurized water. The fatigue life of Z3CN20.09M CASS decreased with increasing thermal aging time (0 ∼ 15000 h at 400 °C), while it slightly affected by the dissolved oxygen (＜5 ppb and 500 ppb). Based on the present results and fatigue data from our previous work, a modified Institute of Metal Research (M–IMR) environmental fatigue model considering thermal aging factors on the environmental fatigue correction factor (F<sub>en</sub>) was developed. Compared with Argonne National Laboratory (ANL) and IMR models, the M–IMR model was more accurate in the fatigue life prediction of thermally aged Z3CN20.09M CASSs. The M–IMR model can accurately predict the fatigue life of CASSs in high-temperature pressurized water after thermal aging at temperatures below 450 °C, but not after thermal aging at temperatures above 450 °C, which may be related to microstructure differences caused by thermal aging and the change in the thermal aging mechanism at different temperatures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114504"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145266822","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}

引用次数: 0

Optimizing hydrogen production through SOEC-PeLUIt-40 coupling: a sustainable approach to clean energy generation 通过SOEC-PeLUIt-40耦合优化氢气生产：清洁能源发电的可持续方法

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114499

Sriyono Sriyono , Dedy Priambodo , Marliyadi Pancoko , Topan Setiadipura , Djati Hoesen Salimy , Ign. Djoko Irianto , Sukmanto Dibyo , Mohammad Dhandhang Purwadi , Yus Rusdian Akhmad , Zuhair , Suwoto , Rahayu Kusumastuti , Erlan Dewita , Siti Alimah , Geni Rina Sunaryo , Entin Hartini , Nurul Huda , Farisy Yogatama Sulistyo

{"title":"Optimizing hydrogen production through SOEC-PeLUIt-40 coupling: a sustainable approach to clean energy generation","authors":"Sriyono Sriyono , Dedy Priambodo , Marliyadi Pancoko , Topan Setiadipura , Djati Hoesen Salimy , Ign. Djoko Irianto , Sukmanto Dibyo , Mohammad Dhandhang Purwadi , Yus Rusdian Akhmad , Zuhair , Suwoto , Rahayu Kusumastuti , Erlan Dewita , Siti Alimah , Geni Rina Sunaryo , Entin Hartini , Nurul Huda , Farisy Yogatama Sulistyo","doi":"10.1016/j.nucengdes.2025.114499","DOIUrl":"10.1016/j.nucengdes.2025.114499","url":null,"abstract":"<div><div>The increasing demand for clean and sustainable energy has positioned hydrogen as a promising alternative fuel. Among various production methods, high-temperature electrolysis using Solid Oxide Electrolysis Cell (SOEC) offers significant thermodynamic and economic benefits. This study explores the feasibility of coupling the PeLUIt-40 modular nuclear reactor with SOEC technology to enable efficient hydrogen production while sustaining electricity generation. A thermodynamic simulation using Cycle-Tempo was conducted to evaluate various steam extraction scenarios and identify the optimal conditions for cogeneration. The results indicate that a steam extraction pressure of 3.5 bar provides the most favorable configuration, avoiding thermal crossover in the steam generator and minimizing power losses. Under this condition, integrating PeLUIt-40 with three SOEC units enables hydrogen production of up to 204 kg/h, with a residual net electrical output of 1.26 MWe. These findings demonstrate that nuclear-assisted hydrogen production using PeLUIt-40 can support dual-purpose energy generation and contribute to sustainable industrial decarbonization strategies in Indonesia and beyond.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114499"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267572","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}

引用次数: 0

Study on autonomous search for multiple radioactive leakage sources based on updated infotaxis in nuclear emergency rescue 核应急救援中基于更新信息趋向性的多源放射性泄漏自主搜索研究

IF 2.1 3区工程技术

Nuclear Engineering and Design Pub Date : 2025-10-10 DOI: 10.1016/j.nucengdes.2025.114500

Huidi Li , Chunhua Chen , Yongzhe Zheng , Liwei Chen , Feng Zou

{"title":"Study on autonomous search for multiple radioactive leakage sources based on updated infotaxis in nuclear emergency rescue","authors":"Huidi Li , Chunhua Chen , Yongzhe Zheng , Liwei Chen , Feng Zou","doi":"10.1016/j.nucengdes.2025.114500","DOIUrl":"10.1016/j.nucengdes.2025.114500","url":null,"abstract":"<div><div>Nuclear facilities face leakage risks from natural hazards, human errors, or external attacks, often generating multi-point radioactive leakage sources that produce large-scale dynamic radiation plumes through atmospheric dispersion and multi-source superposition. Unlike orphan source recovery operations (e.g., retrieving displaced or poorly shielded sealed radioactive sources in localized fields), nuclear emergencies require urgent identification of leakage points to enable real-time leakage sources suppression. Based on the Daya Bay nuclear power plant scenario, this study proposes a multi-source radiation leakage inversion model based on an updated infotaxis algorithm, which incorporates the information entropy of superimposed radiation fields from multiple sources. The search path of the mobile detector is optimized by integrating a movement strategy activation function to adjust subsequent positions. Simulation results demonstrate that the hexagonal path unit enhances search efficiency by 21.78% compared to traditional quadrilateral path units. In a scenario involving three radioactive leakage sources, the mobile detector successfully identifies all sources locations through exhaustive grid sampling, achieving an average positioning error of 5.73 m. This approach provides a novel perspective for identifying multiple radioactive leakage sources in nuclear accidents.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114500"},"PeriodicalIF":2.1,"publicationDate":"2025-10-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145267573","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}

引用次数: 0