Nuclear Engineering and Technology最新文献_第7页

Performance analysis of a two-step calculation procedure based on Monte Carlo and pin-wise diffusion methods for PWR core design 压水堆堆芯设计中基于蒙特卡罗和引脚扩散两步计算方法的性能分析

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-22 DOI: 10.1016/j.net.2025.103596

Changhyun Lim , Sung Joon Kwon , Jooil Yoon

{"title":"Performance analysis of a two-step calculation procedure based on Monte Carlo and pin-wise diffusion methods for PWR core design","authors":"Changhyun Lim , Sung Joon Kwon , Jooil Yoon","doi":"10.1016/j.net.2025.103596","DOIUrl":"10.1016/j.net.2025.103596","url":null,"abstract":"<div><div>This study introduces an efficient two-step calculation procedure for PWR core design by integrating Monte Carlo and pin-wise diffusion methods. The methodology combines Monte Carlo's high-fidelity cross-section generation with pin-wise diffusion's computational efficiency to model neutron flux and power distribution in reactor cores. The approach incorporates Super Homogenization (SPH) factors to enhance neutron flux heterogeneity modeling, addressing the complexities of modern reactor designs with advanced burnable absorbers and control rod strategies.</div><div>Verification using the APR1400 benchmark demonstrates accuracy comparable to whole-core transport codes while maintaining computational efficiency. The methodology is also applied to innovative Small Modular Reactors (i-SMR), particularly evaluating cores with advanced fuel management and soluble boron-free operations. Results show accurate predictions of neutron flux and power distributions in i-SMR cores incorporating advanced burnable absorbers like HIGA (Highly Intensive and Discrete Gadolinium/Alumina Burnable Absorber). The approach effectively addresses i-SMR-specific challenges, including maintaining reactor criticality during extended operational periods. Through optimized parallelization, 3D reactor calculations are completed within seconds, ensuring practical applicability in various operational scenarios.</div><div>This methodology represents a significant advancement in reactor core analysis, offering a high-precision, computationally efficient solution for modern PWR and i-SMR core designs, while maintaining exceptional accuracy in predicting core physics parameters.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103596"},"PeriodicalIF":2.6,"publicationDate":"2025-03-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143715045","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

Study on the operation scheme of a pool-type lead-bismuth fast reactor 池型铅铋快堆运行方案研究

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-21 DOI: 10.1016/j.net.2025.103600

Muping Li, Aodi Sun, Ru Zhang, Peiwei Sun, Xinyu Wei

{"title":"Study on the operation scheme of a pool-type lead-bismuth fast reactor","authors":"Muping Li, Aodi Sun, Ru Zhang, Peiwei Sun, Xinyu Wei","doi":"10.1016/j.net.2025.103600","DOIUrl":"10.1016/j.net.2025.103600","url":null,"abstract":"<div><div>The small lead-bismuth fast reactor has become one of the main reactor types in the Generation IV reactor because of its high inherent safety and easy miniaturization. However, the complex operation environments and variable working conditions make traditional control logic inadequate for load-following requirements. Additionally, the lack of operational experience with lead-bismuth fast reactors complicates control system design. To accelerate the development of small lead-bismuth fast reactors and ensure their safe and economical operation, this paper establishes and verifies a pool-type lead-bismuth fast reactor model and analyzes its dynamic characteristics under different operation modes by using time-domain analysis methods. Nine different small pool-type lead-bismuth fast reactor operation schemes were designed, and the operation schemes were analyzed based on the coupling characteristics of the lead-bismuth fast reactor. The best operation schemes are obtained under both turbine-leading and reactor-leading operation modes. This paper offers guidance for designing operation schemes of lead-bismuth fast reactors.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103600"},"PeriodicalIF":2.6,"publicationDate":"2025-03-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143705038","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

A sourceless efficiency calibration model of airborne γ-spectrometry based on Geant4 基于Geant4的机载γ能谱无源效率定标模型

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-20 DOI: 10.1016/j.net.2025.103594

Cong Mei , Chen Fu , Wei Zhang , Jiangkun Li , Xudong Liang , Xue Wu , Qibin Luo , Yaxin Yang

{"title":"A sourceless efficiency calibration model of airborne γ-spectrometry based on Geant4","authors":"Cong Mei , Chen Fu , Wei Zhang , Jiangkun Li , Xudong Liang , Xue Wu , Qibin Luo , Yaxin Yang","doi":"10.1016/j.net.2025.103594","DOIUrl":"10.1016/j.net.2025.103594","url":null,"abstract":"<div><div>The calibration facility process for the airborne γ-spectrometer, which is built at the Shijiazhuang Dagou Village Airport of China, is time-consuming and labor-intensive. In this paper, a sourceless efficiency calibration model based on the Monte Carlo method using the Geant4 software has been established. The proposed model allows for the acquisition of the γ-spectrum response curve of the airborne γ-spectrometer above a cylindrical γ-radiation source containing uranium, thorium, and potassium elements in any proportion. It is applied to perform Compton scattering calibration and height attenuation correction for UGRS-10 airborne gamma spectrometer. The relative errors between the calibration parameters obtained by Geant4 simulation and those obtained by UGRS-10 airborne gamma spectrometer at Daguocun airport calibration facility are between 3 % and 27 %. The experiment demonstrates the Monte Carlo-based sourceless efficiency calibration model is suitable to calibrate airborne γ-spectrometers and provides technical support for airborne γ-spectrometry.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103594"},"PeriodicalIF":2.6,"publicationDate":"2025-03-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143760308","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

Methodology and preliminary verification of generating heterogeneous multigroup microscopic cross-section libraries for neutron transport codes based on OpenMC 基于OpenMC的中子输运码非均质多群微观截面库生成方法及初步验证

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-19 DOI: 10.1016/j.net.2025.103590

Bowen Cui , Guohua Chen , Xiaofeng Jiang

{"title":"Methodology and preliminary verification of generating heterogeneous multigroup microscopic cross-section libraries for neutron transport codes based on OpenMC","authors":"Bowen Cui , Guohua Chen , Xiaofeng Jiang","doi":"10.1016/j.net.2025.103590","DOIUrl":"10.1016/j.net.2025.103590","url":null,"abstract":"<div><div>Advancements in reactor technology, particularly Generation IV and modular reactors, have introduced new challenges on the neutronics analysis due to their complex geometries and spectra. This study addresses these challenges by developing a methodology to generate heterogeneous multigroup microscopic cross-section libraries for three-dimensional neutron transport calculations using the OpenMC Monte Carlo code. The approach involves two-dimensional transport calculations in OpenMC for various fuel pins or supercells, generating multigroup microscopic cross-section libraries for isotopes relevant to burnup, temperature, and moderator density. These cross-sections are then post-processed and used in three-dimensional core neutron transport calculations with the CRANE deterministic code. This method combines the high accuracy of Monte Carlo methods with the computational efficiency of deterministic approaches. Preliminary 2D verification was conducted using benchmark problems, including PWR fuel assemblies from the VERA series, a fast reactor pin, a 3600 MWth subassembly, and a 1000 MWth metallic fuel core. Results indicate that the coupled OpenMC/CRANE method accurately captures reactivity and isotopic evolution during burnup, suggesting potential improvements in accuracy and efficiency for neutronic simulations of advanced reactor designs.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103590"},"PeriodicalIF":2.6,"publicationDate":"2025-03-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143681685","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

Concept of understandable diagnostic cause visualization with explainable AI and multilevel flow modeling 可理解的诊断原因可视化概念，可解释的人工智能和多层次流建模

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-17 DOI: 10.1016/j.net.2025.103589

Ji Hyeon Shin , Jung Sung Kang , Jae Min Kim , Seung Jun Lee

{"title":"Concept of understandable diagnostic cause visualization with explainable AI and multilevel flow modeling","authors":"Ji Hyeon Shin , Jung Sung Kang , Jae Min Kim , Seung Jun Lee","doi":"10.1016/j.net.2025.103589","DOIUrl":"10.1016/j.net.2025.103589","url":null,"abstract":"<div><div>In nuclear power plants, operators can face cognitive workloads when diagnosing abnormal events due to the need to monitor numerous parameters and consider hundreds of potential scenarios. Artificial intelligence technologies have been proposed to support this process by providing diagnostic results; however, their lack of transparency can lead to out-of-the-loop unfamiliarity and distrust, hindering effective decision-making. To address these challenges, this study introduces a novel concept to enhance the understandability and trustworthiness of diagnostic support systems through Explainable Artificial Intelligence (XAI). The first method in the proposed concept rearranges monitoring parameters based on system structures to reflect parameter relationships. The second method refines explanations from XAI using Multilevel Flow Modeling (MFM) to ensure consistency with physical flow, and it visualizes diagnostic cause components on a plant map. By filtering out incomprehensible information and visualizing intuitive diagnostic causes, the system enables operators to identify expected causes of diagnostic results directly on the NPP map at the component or system level. This approach provides explainable and comprehensible support information, fostering trust in the system and improving diagnostic efficiency in abnormal situations.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103589"},"PeriodicalIF":2.6,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682065","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

Application of the smoothed particle hydrodynamics method to the neutron diffusion equation 光滑粒子流体力学方法在中子扩散方程中的应用

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-17 DOI: 10.1016/j.net.2025.103592

Yiyang Luo, Nan Gui, Xingtuan Yang, Shengyao Jiang

{"title":"Application of the smoothed particle hydrodynamics method to the neutron diffusion equation","authors":"Yiyang Luo, Nan Gui, Xingtuan Yang, Shengyao Jiang","doi":"10.1016/j.net.2025.103592","DOIUrl":"10.1016/j.net.2025.103592","url":null,"abstract":"<div><div>A novel meshless method, known as Smoothed Particle Hydrodynamics (SPH), is applied to solve the neutron diffusion equation for the first time. The discrete forms for each term of the neutron diffusion equation and the boundary conditions are presented, along with a discussion on the differences among various discretization strategies. A comprehensive workflow for solving the neutron diffusion equation using SPH is provided, and the properties of several commonly used kernel functions are examined. An in-house code was developed, and the research indicates that not discretizing the constant terms in the diffusion equation but instead using the values from the previous iteration results in higher accuracy and is less sensitive to variations in parameters such as the smoothing length. A similar finding is observed for boundary conditions; when the boundary condition is of the first kind, directly assigning values yields greater accuracy than using kernel function interpolation. Among the compared kernel functions, the super-Gaussian kernel exhibits the best performance, with errors not exceeding 2.38 % under the given conditions, significantly outperforming the Gaussian and quintic spline kernel functions.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103592"},"PeriodicalIF":2.6,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143681684","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

Optimal passivation process for improved spectroscopic grade CdMnTeSe single crystals based on chromic acid 基于铬酸的改进光谱级CdMnTeSe单晶的最佳钝化工艺

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-17 DOI: 10.1016/j.net.2025.103577

Jiwon Seo , Seungho Song , Jangwon Byun , Chaewon Yoon , Taejoon Mo , Younghak Kim , Hansoo Kim , Jung-Yeol Yeom , Beomjun Park

引用次数: 0

The application of Kriging model for three-dimensional core power distribution calibration Kriging模型在三维堆芯功率分配标定中的应用

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-17 DOI: 10.1016/j.net.2025.103591

Tien-Tso Lee, Wu-Hsiung Tung

{"title":"The application of Kriging model for three-dimensional core power distribution calibration","authors":"Tien-Tso Lee, Wu-Hsiung Tung","doi":"10.1016/j.net.2025.103591","DOIUrl":"10.1016/j.net.2025.103591","url":null,"abstract":"<div><div>The safety operation of a nuclear reactor relies on knowing the core power distributions as accurately as possible. PWR adapt the calculated power distributions to obtain accurate power distributions, which requires knowing the calibration factors for the calculated powers of each fuel bundle in the core. The calibration factors are obtained by comparing the measured reaction rates and the calculated reaction rates at the instrumented locations in the core. For the fuel bundles position that do not have measurement data, their calibration factors shall be inferred by using the calibration factor of the instrumented bundles nearby. In this study, we propose using Kriging model to fit the calibration factors of the instrumented bundles and then predict the calibration factors for the un-instrumented bundles. To verify the correctness of the model, we employed a method of repeated random sampling from the dataset to validate the model. After expanding the samples in each group, which originally contained only one plane, to include two or three planes, it was observed that the average RMSE value, as defined by cross-validation, not only significantly decreased but also showed a marked improvement in the predicted calibration factor for most of the validated samples.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103591"},"PeriodicalIF":2.6,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143681682","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

Simulating nuclear fuel inspections: Enhancing reliability through synthetic data 模拟核燃料检验：通过合成数据提高可靠性

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-15 DOI: 10.1016/j.net.2025.103571

J. Knotek , J. Blažek , M. Kopeć

{"title":"Simulating nuclear fuel inspections: Enhancing reliability through synthetic data","authors":"J. Knotek , J. Blažek , M. Kopeć","doi":"10.1016/j.net.2025.103571","DOIUrl":"10.1016/j.net.2025.103571","url":null,"abstract":"<div><div>Visual inspection of nuclear fuel assemblies is critical for assessing fuel reliability and ensuring safe operation. However, the sensitivity of real inspection data, along with its inflexibility and high collection costs, limits its use for research and development (R&D) tasks. These challenges hinder the ability to test and validate new inspection methodologies, making innovation slow and expensive. To address these limitations, we propose the development of synthetic nuclear fuel datasets that simulate fuel assembly inspections. These data sets replicate various defects and degradations in fuel assemblies, providing a controlled environment for hypothesis testing, operator training, and the evaluation of automated inspection techniques. Unlike real-world data, synthetic data offers the advantage of known ground-truth parameters, allowing for rigorous testing and validation. This approach enables the continuous development of inspection technologies, regardless of hardware availability and operational outages in nuclear facilities. By reducing the reliance on costly real-world experiments, synthetic data offers a scalable and flexible solution for the advancement of nuclear fuel inspection methods.</div></div>","PeriodicalId":19272,"journal":{"name":"Nuclear Engineering and Technology","volume":"57 8","pages":"Article 103571"},"PeriodicalIF":2.6,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143682204","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

RUCAS: Software for assessing radiological risk to the public in metallic radioactive waste recycling from decommissioning nuclear facilities RUCAS：用于评估从退役核设施回收金属放射性废物对公众的辐射风险的软件

IF 2.6 3区工程技术

Nuclear Engineering and Technology Pub Date : 2025-03-14 DOI: 10.1016/j.net.2025.103585

Ugyu Jeong , Hyeongjin Byeon , Jaeyeong Park

引用次数: 0