Nuclear Engineering and Technology最新文献_第3页

A theoretical approach to understand the reaction dynamics for producing Technetium isotopes 了解生产锝同位素的反应动力学的理论方法

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-30 DOI: 10.1016/j.net.2025.103672

Amit Chauhan

引用次数: 0

A historical review of high heat flux cooling techniques 高热流密度冷却技术的历史回顾

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-30 DOI: 10.1016/j.net.2025.103725

Ji Hwan Lim, Seung-Hwan Yu

{"title":"A historical review of high heat flux cooling techniques","authors":"Ji Hwan Lim, Seung-Hwan Yu","doi":"10.1016/j.net.2025.103725","DOIUrl":"10.1016/j.net.2025.103725","url":null,"abstract":"<div><div>This review provides a comprehensive historical analysis of high heat flux cooling techniques critical for advanced applications in nuclear energy, aerospace propulsion, and high-performance electronics. We trace the evolution of cooling strategies from early nucleate boiling experiments to recent innovations such as swirl flow devices, Hypervapotron channels, screw tubes, and porous coatings. By synthesizing experimental results and theoretical models, we identify significant enhancements in critical heat flux (CHF) and heat transfer coefficients, with swirl flow techniques offering improvements of 50–100 % over smooth channels and Hypervapotron systems achieving CHF levels up to 30 MW/m<sup>2</sup> under optimal conditions. Our review also highlights that while individual cooling methods exhibit unique performance benefits, an integrated hybrid approach holds promise for future high heat flux cooling systems. However, the translation from laboratory-scale success to practical, industrial-scale applications presents challenges, particularly in system complexity, maintenance, and cost analysis. We conclude that future research must bridge these gaps through interdisciplinary strategies, robust cost-analysis frameworks, and optimization studies. This balanced evaluation underscores both the technical achievements and the practical hurdles that define the current state and future potential of high heat flux cooling technologies.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103725"},"PeriodicalIF":2.6,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144222531","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

Measurement of IBD-like background with plastic scintillator detector array at surface level 表面塑料闪烁体探测器阵列测量类ibd背景

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-30 DOI: 10.1016/j.net.2025.103707

Chao Wang , Gao-Kui He , Hua-Yang Tian , Xiu-Zhang Tang , Jiang-Bin Zhao , Yang Liu , Chun-Yu Gao

引用次数: 0

Reactor physics fast calculation method based on model order reduction and machine learning 基于模型阶数约简和机器学习的反应堆物理快速计算方法

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-30 DOI: 10.1016/j.net.2025.103724

Chen Zhao , Qinyi Zhang , Bin Zhang , Jiangyu Wang , Jiayi Liu , Lianjie Wang , Bangyang Xia , Xiaoming Chai , Xingjie Peng

{"title":"Reactor physics fast calculation method based on model order reduction and machine learning","authors":"Chen Zhao , Qinyi Zhang , Bin Zhang , Jiangyu Wang , Jiayi Liu , Lianjie Wang , Bangyang Xia , Xiaoming Chai , Xingjie Peng","doi":"10.1016/j.net.2025.103724","DOIUrl":"10.1016/j.net.2025.103724","url":null,"abstract":"<div><div>The rapid development of artificial intelligence technology has provided new ideas and methods for reactor physics calculations. Based on AI technology, a fast calculation method for reactor physics has been established, which combines model order reduction and machine learning to address the challenges of excessive parameter quantities in machine learning-based parameter prediction. During the training process, the full-order model is established using the two-step core nuclear design software package TORCH, and the model order reduction theory is applied, which are then trained using the random forest machine learning method. In the prediction process, the basis weight coefficients are rapidly calculated for specific input parameters, and the core distribution results are reconstructed. A reactor physics fast calculation program has been developed and verified using a M310-type pressurized water reactor nuclear power plant with 9522 samples. All results show that the fast calculation method based on model order reduction and machine learning has good computational efficiency and accuracy. The calculation time can be reduced to 0.1 s and the proportion of samples with less than 1 % deviation in various core physics parameters is higher than 90 %.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103724"},"PeriodicalIF":2.6,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144271357","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

The effect of storage conditions and computer model simplification on the thermal state of the RBMK-1500M2 cask 储存条件和计算机模型简化对RBMK-1500M2木桶热状态的影响

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-29 DOI: 10.1016/j.net.2025.103729

Kęstutis Račkaitis , Povilas Poškas , Robertas Poškas , Hussam Jouhara

{"title":"The effect of storage conditions and computer model simplification on the thermal state of the RBMK-1500M2 cask","authors":"Kęstutis Račkaitis , Povilas Poškas , Robertas Poškas , Hussam Jouhara","doi":"10.1016/j.net.2025.103729","DOIUrl":"10.1016/j.net.2025.103729","url":null,"abstract":"<div><div>Interim storage of spent nuclear fuel is a very important part of the overall nuclear power generation cycle. At Ignalina NPP, spent nuclear fuel is stored in interim storage facilities in specially designed casks before being transferred to a geological repository. The internal structure of spent nuclear fuel casks and the processes involved are quite complex. Therefore, simplifying and optimizing simulations for evaluating decay heat removal from the cask is worthwhile. In this paper, the effect of computer model simplifications on the thermal characteristics of the CONSTOR RBMK-1500/M2 cask stored in building-type and open-type storage facilities is presented. The modeling was carried out using the ANSYS Fluent code. The analysis showed a substantial impact of solar insolation. Also, in the case of the homogenization of the SNF load in the basket, higher temperatures are obtained compared with the case when detailed modeling of the internal basket structure is performed. Hence, it was demonstrated that the homogenization model can be used in safety assessment as a conservative approach for the modeling of decay heat removal from the cask.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103729"},"PeriodicalIF":2.6,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144263782","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

Strategy to safely enable X-ray computed tomography examination of highly radioactive tristructural isotropic nuclear fuel 对高放射性三结构各向同性核燃料进行x射线计算机断层扫描的安全策略

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-28 DOI: 10.1016/j.net.2025.103726

William Chuirazzi , Jayson Bush , Brian Gross , Matthew Bryant , Ken Clark , Miles Cook , Joshua Burtenshaw , Joseph Price , Swapnil Morankar , McKenzie Blattner , Raegan Landon , Keth Galloway , Justin Stanger , Ruger Stamos , Justin Duke , Camren Watt , John Stempien

{"title":"Strategy to safely enable X-ray computed tomography examination of highly radioactive tristructural isotropic nuclear fuel","authors":"William Chuirazzi , Jayson Bush , Brian Gross , Matthew Bryant , Ken Clark , Miles Cook , Joshua Burtenshaw , Joseph Price , Swapnil Morankar , McKenzie Blattner , Raegan Landon , Keth Galloway , Justin Stanger , Ruger Stamos , Justin Duke , Camren Watt , John Stempien","doi":"10.1016/j.net.2025.103726","DOIUrl":"10.1016/j.net.2025.103726","url":null,"abstract":"<div><div>Nondestructive post-irradiation examination of nuclear fuels and materials is useful in collecting data to inform commercial licensing of new nuclear concepts. This work details the planning, shielding design, and workflow of an X-ray computed tomography examination of an irradiated AGR-5/6/7 tristructural isotropic fuel compact with a dose rate of 1318 R/h on contact from both β and ɤ-ray radiation with 120 R/h of the dose rate coming exclusively from ɤ-rays. Rather than waiting for radioactive decay to reduce the dose rate from the sample, careful implementation of specially developed shielding and work controls enabled some examinations of the irradiated fuel in a timelier manner, which benefits the ongoing qualification and commercialization cases for this fuel by reducing the timeline from irradiated to examination. This timeframe reduction is important as data from these experiments can be more quickly available to inform modeling efforts and subsequent experiments. While this work represents the hottest radiological sample that has been openly transported and examined at Idaho National Laboratory's Irradiated Materials Characterization Laboratory to date, it is anticipated that this work and its lessons learned will facilitate future examinations of similar or even more radioactive specimens.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103726"},"PeriodicalIF":2.6,"publicationDate":"2025-05-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144184974","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

Sorption of U(IV) on MX-80, illite, shale and limestone in Na-Ca-Cl solutions 在Na-Ca-Cl溶液中，MX-80、伊利石、页岩和石灰岩对U（IV）的吸附

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-28 DOI: 10.1016/j.net.2025.103720

Shinya Nagasaki , Zhiwei Zheng , Jianan Liu , Jieci Yang , Tianxiao Tammy Yang

{"title":"Sorption of U(IV) on MX-80, illite, shale and limestone in Na-Ca-Cl solutions","authors":"Shinya Nagasaki , Zhiwei Zheng , Jianan Liu , Jieci Yang , Tianxiao Tammy Yang","doi":"10.1016/j.net.2025.103720","DOIUrl":"10.1016/j.net.2025.103720","url":null,"abstract":"<div><div>Uranium has been identified as an element of interest for the safety assessment of a deep geological repository (DGR) for used nuclear fuel. This paper studied the sorption of tetravalent U(IV) onto MX-80 bentonite, illite, shale and limestone in ionic strength 0.1–6 mol/kgw (m) Na-Ca-Cl solutions at pH<sub>m</sub> = 4–10 (for MX-80, illite, shale) and pH<sub>m</sub> = 5–10 (for limestone) under reducing conditions. Sorption of U(IV) on MX-80 increased with pH<sub>m</sub> until pH<sub>m</sub> = 8 and then reached a sorption plateau. Sorption of U(IV) on illite and shale increased from pH<sub>m</sub> = 4 to 5 and then reached a sorption plateau. Sorption of U(IV) on limestone increased with pH<sub>m</sub> up to pH<sub>m</sub> = 7 and then steeply decreased with pH<sub>m</sub>. Sorption of U(IV) on all four solids was not influenced by the ionic strength of Na-Ca-Cl solutions. The sorption of U(IV) onto MX-80 and illite in Na-Ca-Cl solutions was successfully simulated by a 2 site protolysis non-electrostatic surface complexation and cation exchange (2SPNE SC/CE) model using three inner-sphere surface complexation reactions. Values of surface complexation constants (log <em>K</em><sup>0</sup>) were optimized.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103720"},"PeriodicalIF":2.6,"publicationDate":"2025-05-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144263783","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

Examination of SAMEVAL: Software to calculate and evaluate sample sizes for national safeguards inspection in the Republic of Korea SAMEVAL的审查：用于计算和评估大韩民国国家保障检查样本大小的软件

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-28 DOI: 10.1016/j.net.2025.103718

Haneol Lee, Hyun Cheol Lee, Hana Seo, Jung Youn Choi, Hyun Ju Kim, Jinho Ryu

{"title":"Examination of SAMEVAL: Software to calculate and evaluate sample sizes for national safeguards inspection in the Republic of Korea","authors":"Haneol Lee, Hyun Cheol Lee, Hana Seo, Jung Youn Choi, Hyun Ju Kim, Jinho Ryu","doi":"10.1016/j.net.2025.103718","DOIUrl":"10.1016/j.net.2025.103718","url":null,"abstract":"<div><div>As the International Atomic Energy Agency (IAEA) implements the state level approach (SLA) as a safeguards approach, the role of a state system of accounting for and control (SSAC) becomes critically important. The Republic of Korea (ROK), as a member state under the SLA, is implementing an independent sampling plan to strengthen the technical capability of the SSAC. Regarding this sampling plan, this study proposes SAMEVAL: a GUI-based independent sample size calculation and evaluation software. The SAMEVAL software minimizes approximations in the IAEA's \"common inspection on-site software package (CIOSP)\". It can also evaluate the effectiveness of the calculated sample sizes for representative diversion scenarios: a novel capability not offered by CIOSP. This study then demonstrates the SAMEVAL software using 20 benchmark cases of an IAEA technical report (STR-261), and demonstrates the effectiveness of the calculated sample sizes using SAMEVAL for five diversion scenarios with different defect sizes. The effectiveness testing of SAMEVAL evaluates the non-detection probability (<span><math><mrow><mi>β</mi></mrow></math></span>) of the calculated sample sizes for a given diversion scenario based on the Monte Carlo calculation. It then compares the calculated <span><math><mrow><mi>β</mi></mrow></math></span> to the target non-detection probability (<span><math><mrow><msub><mi>β</mi><mi>c</mi></msub></mrow></math></span>) of each case. Our results indicate that SAMEVAL reduces inspection burden by reducing the total sample size or sample sizes with more precise verification methods while maintaining the same level of non-detection probability. With these capabilities, SAMEVAL can be a tool for the sampling plan in national safeguards inspection in the ROK to strengthen the technical capability of the SSAC.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103718"},"PeriodicalIF":2.6,"publicationDate":"2025-05-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144189522","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

Verification of reliability of the burnup for PWR spent fuel of OPR1000 type OPR1000型压水堆乏燃料燃耗可靠性验证

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-27 DOI: 10.1016/j.net.2024.103373

Donghee Lee , Yongdeog Kim , Kiyoung Kim , Sunghwan Chung , Taehyeon Kim , Taehyung Na

{"title":"Verification of reliability of the burnup for PWR spent fuel of OPR1000 type","authors":"Donghee Lee , Yongdeog Kim , Kiyoung Kim , Sunghwan Chung , Taehyeon Kim , Taehyung Na","doi":"10.1016/j.net.2024.103373","DOIUrl":"10.1016/j.net.2024.103373","url":null,"abstract":"<div><div>As the saturation of spent fuel pool (SFP) capacity for nuclear power plants starting in 2030 is imminent, the operation of dry storage facilities for spent fuel from pressurized water reactors (PWR) has become urgent. Therefore, it is essential to secure characteristic data on the spent fuel and develop related equipment before these dry storage facilities begin operation. Due to the challenge of inspecting the spent fuel once it is loaded into dry storage casks, it is necessary to identify the characteristics of the spent fuel in advance and secure the information required for the transfer criteria for interim storage and disposal in the future. In particular, among the characteristics data, the burnup of spent fuel is one of the important management items not only for dry storage but also for transportation and disposal. Such burnup varies depending on the operating history, and in reality, nuclear power plants measure and manage the burnup using an in-core instrument (ICI) during operation. This study was conducted to verify the reliability of the burnup of such spent fuel. To this end, comparisons and analyses were conducted among the actual burnup obtained from the power plant, the design burnup calculated from design codes, and the measured burnup of spent fuel actually stored in the storage pool using burnup measuring equipment. This study aimed to establish an evaluation system for the long-term integrity assessment of spent fuel through burnup measurement and comparative analysis for the first time in South Korea, laying the foundation for systematic management.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 9","pages":"Article 103373"},"PeriodicalIF":2.6,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144223566","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

MOUSE software package: challenges and experience of fitting uncertainty management into long-term safety assessment puzzle MOUSE软件包：将不确定性管理融入长期安全评估难题的挑战与经验

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-05-27 DOI: 10.1016/j.net.2025.103721

Valentina Svitelman, Elena Saveleva, Anton Lebedev, Gleb Kiselev, Anton Nuzhny, Alina Rukavichnikova, Evgeny Kolodin

{"title":"MOUSE software package: challenges and experience of fitting uncertainty management into long-term safety assessment puzzle","authors":"Valentina Svitelman, Elena Saveleva, Anton Lebedev, Gleb Kiselev, Anton Nuzhny, Alina Rukavichnikova, Evgeny Kolodin","doi":"10.1016/j.net.2025.103721","DOIUrl":"10.1016/j.net.2025.103721","url":null,"abstract":"<div><div>Deep geological repositories for high-level radioactive waste require comprehensive long-term safety assessments involving numerical simulations of safety-relevant processes. Managing uncertainty across all parts of such a simulation framework is challenging, as general-purpose uncertainty management toolboxes typically lack either the variety of methods or the integration capabilities to simultaneously accommodate all components of the long-term safety assessment framework. Under these circumstances, developing bespoke uncertainty-management software based on up-to-date methods and well-maintained libraries but tailored specifically for the task at hand becomes more efficient than creating various plugins and converters for general-purpose tools. This paper discusses the experience of developing the MOUSE software package to illustrate this path from a model-independent software to the uncertainty-management instrument tailored for the purposes of the numerical safety assessment framework.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 10","pages":"Article 103721"},"PeriodicalIF":2.6,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144189520","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