Nuclear Engineering and Technology最新文献_第9页

Localization of a radioactive source with the predictive range-whale optimization algorithm method 预测距离鲸优化算法在放射源定位中的应用

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-25 DOI: 10.1016/j.net.2025.103562

Zongheng Hong , Feiyun Cong , Bo Yang , Zhangyu Chen , Yunlong Niu

{"title":"Localization of a radioactive source with the predictive range-whale optimization algorithm method","authors":"Zongheng Hong , Feiyun Cong , Bo Yang , Zhangyu Chen , Yunlong Niu","doi":"10.1016/j.net.2025.103562","DOIUrl":"10.1016/j.net.2025.103562","url":null,"abstract":"<div><div>Heuristic algorithms can effectively improve the accuracy and efficiency of estimating the status of radioactive sources based on the maximum likelihood estimation method. However, heuristic algorithms often converge prematurely at local optima, compromising stability and making them unsuitable for prolonged monitoring tasks. To address this problem, the Predictive Range-Whale Optimization Algorithm (PR-WOA) method was proposed in this paper. Initially, the predictive location and intensity ranges of the radioactive source were determined by integrating historical prediction results. Subsequently, the initial population was more concentrated within the predictive range to improve the algorithm's capability for local optimization. Finally, an inertia weight was introduced to adaptively adjust the search step, consequently improving its global searching capability and efficiency. The performance of the PR-WOA method was evaluated with the simulation and experimental data. Comparative studies demonstrated that the proposed method significantly improves accuracy and stability in predicting the status of the radioactive source. In the 2022 radioactive source localization experiment in Hangzhou, PR-WOA method achieved an average localization accuracy of 1.30m during the trajectory tracking of a moving radioactive source mounted on the drone. Compared to traditional heuristic algorithms, this method improved localization accuracy by 17.9 % and enhanced localization stability during long-term monitoring by 19.0 %.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103562"},"PeriodicalIF":2.6,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143839582","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

An experimental investigation on the safety performance of dry ice blasting technology 干冰爆破技术安全性能的试验研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-25 DOI: 10.1016/j.net.2025.103561

Shulong Huang , Kewei Jiang , Jinjie Zhao , Lixiao Guo , Dan Luo , Yuhang Zhang , Jialu Li

引用次数: 0

Investigation of the effect of boric acid concentration on neutron reflection coefficient 硼酸浓度对中子反射系数影响的研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-25 DOI: 10.1016/j.net.2025.103564

Farnaz Karimi, Farhad Zolfagharpour, Sara Azimkhani

引用次数: 0

Effect of brazing parameters on the microstructure and mechanical properties of SiC ceramic joint with Zr-Cu-Nb filler metal 钎焊参数对Zr-Cu-Nb钎料SiC陶瓷接头显微组织和力学性能的影响

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-25 DOI: 10.1016/j.net.2025.103547

Bofang Zhou , Zixuan Leng , Hongxia Zhang , Zhaojie Zhang , Wuman Wang

{"title":"Effect of brazing parameters on the microstructure and mechanical properties of SiC ceramic joint with Zr-Cu-Nb filler metal","authors":"Bofang Zhou , Zixuan Leng , Hongxia Zhang , Zhaojie Zhang , Wuman Wang","doi":"10.1016/j.net.2025.103547","DOIUrl":"10.1016/j.net.2025.103547","url":null,"abstract":"<div><div>The excellent properties of SiC ceramic make them possible to replace zirconium alloy and become the 4th generation of nuclear power plant cladding material. In this study, the microstructure and mechanical properties of SiC ceramic joint with Zr-Cu-Nb powder filler metal under different brazing temperature and holding time were investigated. The results show that Zr-Cu-Nb filler metal can effectively braze SiC ceramic at different brazing temperature (950<span><math><mo>∼</mo></math></span>1150 °C) with the holding time of 20 min. The primary phases in the interface reaction layer are mainly ZrC and Zr<sub>2</sub>Si. The shear strength of the joint first increases and then decreases with the brazing temperature, reaching a maximum of 135 MPa at 1050 °C under the holding time of 20 min. The joint primarily fractures at the reaction layer, exhibiting brittle fracture behavior. The thickness of the interface reaction layer increases with the holding time(5<span><math><mo>∼</mo></math></span>60min) under the brazing temperature of 1050 °C. The shear strength of the joint reaches its maximum of 135 MPa at 20 min under the brazing temperature of 1050 °C, with the thickness of the reaction layer being 1.2 <span><math><mi>μ</mi></math></span>m. The joint primarily fractures at the reaction layer, exhibiting brittle fracture behavior.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103547"},"PeriodicalIF":2.6,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682209","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}

引用次数: 0

Development of FEMAXI-ATF for fuel performance modeling of SiC cladded fuel involving PCMI 基于PCMI的SiC包覆燃料性能建模femaxis - atf的开发

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-24 DOI: 10.1016/j.net.2025.103558

Yoshihiro Kubo, Akifumi Yamaji

{"title":"Development of FEMAXI-ATF for fuel performance modeling of SiC cladded fuel involving PCMI","authors":"Yoshihiro Kubo, Akifumi Yamaji","doi":"10.1016/j.net.2025.103558","DOIUrl":"10.1016/j.net.2025.103558","url":null,"abstract":"<div><div>The SiC cladded fuel is being developed as a candidate for the accident tolerant fuel of light water reactors, in which the cladding wall consists of the inner SiC/SiC layer and the outer monolithic (mSiC) layer. The FEMAXI-ATF fuel performance modeling code is being developed based on the experience of FEMAXI-7 with the focus on evaluation of the pellet-cladding mechanical interaction (PCMI). In the preceding study, the capability of FEMAXI-ATF to show the PCMI failure limit of the SiC cladded fuel was successfully demonstrated by modeling the SiC/SiC layer with a continuum finite calculation element and reducing Young's modulus of the failed element to practically zero, so that the stress on the cladding wall is redistributed among the remaining cladding wall elements. In this study, FEMAXI-ATF has been further developed to consider pseudo-ductility of SiC/SiC and improvement to the temperature dependent SiC swelling model. The new FEMAXI-ATF was applied in the BWR normal operation conditions to show a possible scenario. That is, the SiC cladded fuel may experience partial failure of the SiC/SiC layer during the reactor shutdown depending on the swelling characteristics of SiC. Consequently, PCMI may induce large mechanical load on the mSiC layer in the subsequent operation.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103558"},"PeriodicalIF":2.6,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682201","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}

引用次数: 0

Insights into irradiation creep coefficient in nuclear graphite from machine learning 基于机器学习的核石墨辐照蠕变系数研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-23 DOI: 10.1016/j.net.2025.103559

Yuzhou Wang , Derek Tsang , Yibo Zhang , Qiang Zhang , Fei Zhu , Ligang Song , Xianfeng Ma

{"title":"Insights into irradiation creep coefficient in nuclear graphite from machine learning","authors":"Yuzhou Wang , Derek Tsang , Yibo Zhang , Qiang Zhang , Fei Zhu , Ligang Song , Xianfeng Ma","doi":"10.1016/j.net.2025.103559","DOIUrl":"10.1016/j.net.2025.103559","url":null,"abstract":"<div><div>Understanding irradiation induced creep in nuclear graphite is critical for the service life extension of current reactor fleet and the technological advancement of next generation nuclear reactors. Nevertheless, qualifying a new graphite grade with respect to irradiation creep requires years of testing and expensive facilities for experiments. Here for the first time, we applied machine learning (ML) algorithms to investigate the irradiation creep coefficient in the secondary stage of graphite creep in hope of gaining new insights and expediting the qualification process. Four ML models were trained on a small dataset with temperature and materials properties as input. The gradient boosting regression model exhibits the best predicting performance. The ML models indicate that temperature and Young's modulus are the most important parameters in the determination of creep coefficients while the rest properties have much weaker impact. These findings align with previous theories and corroborate a creep mechanism governed by dislocation climb, demonstrating the potential of ML in improving the workflow of graphite qualification for advanced reactors.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103559"},"PeriodicalIF":2.6,"publicationDate":"2025-02-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682202","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}

引用次数: 0

Manufacturing B4C-Al dispersion absorber plates using the picture-frame technique 用相框技术制造B4C-Al分散吸收板

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-22 DOI: 10.1016/j.net.2025.103555

Sérgio Ricardo Catharino de Mello, Mauricio Moralles, Marco Antonio Stanojev Pereira, Frederico Antonio Genezini, Elita Fontenele Urano de Carvalho, Michelangelo Durazzo

{"title":"Manufacturing B4C-Al dispersion absorber plates using the picture-frame technique","authors":"Sérgio Ricardo Catharino de Mello, Mauricio Moralles, Marco Antonio Stanojev Pereira, Frederico Antonio Genezini, Elita Fontenele Urano de Carvalho, Michelangelo Durazzo","doi":"10.1016/j.net.2025.103555","DOIUrl":"10.1016/j.net.2025.103555","url":null,"abstract":"<div><div>The rising global demand for nuclear energy emphasizes the importance of neutron absorber materials, essential for spent fuel storage and other nuclear applications. This study focuses on manufacturing neutron absorber plates using the picture-frame technique, incorporating varying B<sub>4</sub>C volume contents on an aluminum matrix in the plates' meat. Three configurations were explored, and the plates underwent radiographic, mechanical, microstructural, and neutronic evaluations. The results confirmed the viability of producing neutron absorber plates with reduced dimensions using the picture-frame method. Material properties were comparable to those of commercial neutron absorbers, indicating scalability for larger sizes, and facilitating the eventual deployment of this product. Plates with 45 vol% B<sub>4</sub>C demonstrated favorable neutron shielding properties, achieving 100 % attenuation in a thermal neutron beam with a 4.0 mm thick B<sub>4</sub>C-Al dispersion. However, adding B<sub>4</sub>C to the aluminum matrix caused a decline in mechanical performance, with elongation values around 2 %, making these plates unsuitable for structural use, similar to other commercial materials in this category.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103555"},"PeriodicalIF":2.6,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682139","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}

引用次数: 0

High-abundance Pu-238 produced by nuclear transmutation of liquid MA fuel in reactor 反应堆液态MA燃料核嬗变产生高丰度Pu-238

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-22 DOI: 10.1016/j.net.2025.103554

Dongguo Li

{"title":"High-abundance Pu-238 produced by nuclear transmutation of liquid MA fuel in reactor","authors":"Dongguo Li","doi":"10.1016/j.net.2025.103554","DOIUrl":"10.1016/j.net.2025.103554","url":null,"abstract":"<div><div>The feasibility of mass producing high-abundance <sup>238</sup>Pu using liquid fuel containing minor actinides (MA) in a reactor is discussed. A chemically stable liquid solution or molten salt with MA dissolved, is evenly distributed in the inner or outer region of the reactor core to ensure that the MA materials receive sufficient neutron irradiation and achieve a high transmutation rate. The transmutation products of <sup>241</sup>Am or <sup>237</sup>Np are mainly <sup>238</sup>Pu and fission products. After being burned, the liquid fuel is extracted from the reactor and separated in a reprocessing plant to obtain high-abundance <sup>238</sup>Pu material.</div><div>To maximize the abundance of <sup>238</sup>Pu in reaction products, a thorium-based molten salt fast reactor was adopted. Its inner and outer regions of core are loaded with molten salts LiF + ThF<sub>4</sub>+XF<sub>4</sub> and LiF + ThF<sub>4</sub>, where X is <sup>233</sup>U or <sup>235</sup>U. Fluoride salts of <sup>241</sup>Am, <sup>237</sup>Np, or TRU-MAs (a mixture primarily composed of <sup>241</sup>Am and <sup>237</sup>Np) are added to one of these molten salts. Depending on the different driving fuels and MA targets, the annual output of <sup>238</sup>Pu ranges from 6.9 to 130.2 kg, with abundances varying from 67.7 % to 99.4 %. The annual transmutation rates for single nuclide <sup>241</sup>Am or <sup>237</sup>Np, can reach up to 21.4 % or 20.0 %.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103554"},"PeriodicalIF":2.6,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682068","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}

引用次数: 0

Dynamic characteristics test of safety class cable for nuclear power plants 核电站安全级电缆动态特性试验

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-22 DOI: 10.1016/j.net.2025.103557

Jung-Dae Son , Jae-Hong Kim , Kum-Young Song , Jung-Han Kim

引用次数: 0

Evaluation of accident mitigation capability of passive core and containment cooling systems of i-SMR under representative LOCA and non-LOCA conditions i-SMR被动堆芯和安全壳冷却系统在典型的LOCA和非LOCA条件下的事故缓解能力评估

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-02-22 DOI: 10.1016/j.net.2025.103556

Chang Hyun Song , JinHo Song , Sung Joong Kim

{"title":"Evaluation of accident mitigation capability of passive core and containment cooling systems of i-SMR under representative LOCA and non-LOCA conditions","authors":"Chang Hyun Song , JinHo Song , Sung Joong Kim","doi":"10.1016/j.net.2025.103556","DOIUrl":"10.1016/j.net.2025.103556","url":null,"abstract":"<div><div>The global climate crisis has accelerated efforts to cut greenhouse gas emissions in the energy sector. While renewable energy faces limits in capacity and high costs, nuclear power offers a low-carbon, reliable baseload alternative and has been included in the EU Green Taxonomy since 2023 as a top option for achieving net-zero targets. Meeting heightened safety standards in nuclear power plants (NPPs) and rising electricity demand driven by AI has spurred strong interest in small modular reactors (SMRs) equipped with advanced passive safety systems. A standout example is Korea's innovative SMR (i-SMR), which delivers 170 MWe and is engineered to enhance safety by over 1000 times compared to traditional NPPs, with core damage frequency and early release frequency below 1.0E-09 and 1.0E-10 per module-year, respectively. Using the MELCOR code, this study evaluates the i-SMR's passive cooling systems under LOCA and non-LOCA conditions. The results confirm these systems can maintain reactor core and fuel integrity for 72 h post-accident without operator intervention, showcasing the high efficacy of the i-SMR's passive safety systems.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103556"},"PeriodicalIF":2.6,"publicationDate":"2025-02-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682563","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}

引用次数: 0