Annals of Nuclear Energy最新文献

{"title":"Multi-objective optimization of a beam dump for Deuterium-Deuterium neutron source","authors":"Jingru Chen ,&nbsp;Mingfei Yan ,&nbsp;Zhanfei Liu ,&nbsp;Xubin Zhang ,&nbsp;Lucheng Yang ,&nbsp;A.K. Jadi ,&nbsp;Huasi Hu","doi":"10.1016/j.anucene.2025.111859","DOIUrl":"10.1016/j.anucene.2025.111859","url":null,"abstract":"<div><div>Experimental background significantly affects data quality in neutron experiments, with backscattered neutrons and gamma-rays from the back wall being the primary sources, especially in confined environments. To suppress this, a multilayer beam dump composed of high density polyethylene (HDPE), lead (Pb), boron carbide (B₄C), and steel was designed for the Deuterium-Deuterium (DD) neutron source facility for neutron imaging. A multi-objective optimization method combining the non-dominated sorting genetic algorithm (NSGA-II) and Monte Carlo simulation code MCNP6 was employed. The energy spectra and spatial distributions of backscattered neutrons and gamma-rays were obtained using MCNP6 for three configurations: without a beam dump, with the original design, and with the optimized design. Compared to the original design, the optimized beam dump reduced neutron flux by 23%, gamma-ray flux by 22%, and total mass by 58%, confirming that the multi-objective optimization method of NSGA-II combined with MCNP6 is an effective tool for beam dump design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111859"},"PeriodicalIF":2.3,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145026716","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 study of 3D-Printable polymers for beta interaction behaviours via monte carlo simulations","authors":"Kevser A. Hışıroğlu,&nbsp;Özgür Akçalı,&nbsp;Ozan Toker,&nbsp;Orhan İçelli","doi":"10.1016/j.anucene.2025.111864","DOIUrl":"10.1016/j.anucene.2025.111864","url":null,"abstract":"<div><div>The rapid advancement of 3D printing technologies has revolutionised production in many sectors due to the versatile properties of printable materials. Thermal, mechanical and functional performance of 3D printable components make them suitable for use in radiation-intensive environments such as betavoltaic cells and clinical phantoms. This study investigates the behaviour of 3D-printable materials, including ABS, PLA, PETG, TPU, and UV resins<strong>,</strong> under exposure to various beta radiation sources, such as ³H, ⁶³Ni, ¹⁴C, ¹⁴⁷Pm, ⁹⁹Tc, ⁹⁰Sr, and ⁹⁰Y, through a comparative analysis of their beta radiation attenuation efficiency, energy absorption, and dose transmission characteristics. TPU exhibited a surface flux reduction of approximately 72.3 % for tritium (³H) and 69.8 % for nickel-63 (⁶³Ni) at a thickness of 0.1 µm, demonstrating better attenuation performance for low-energy beta radiation. A new metric, Half Absorption Thickness (HAT), was introduced to quantify the shielding capacity more precisely; the lowest HAT was found for TPU under ³H (0.08 mm), and the highest for PETG under ⁹⁹Tc (1.12 mm). The derived logarithmic models (R² &gt; 0.99) enable estimation of shielding efficiency as a function of material thickness and beta energy. Among all tested combinations, ABS exhibited the best attenuation performance against high-energy beta particles, with a maximum flux reduction of 87.1 % for ⁹⁰Sr and a corresponding HAT of 0.48 mm, establishing it as the most efficient barrier material in the studied energy range. Conversely, UV-cured resin demonstrated the highest beta transmission, particularly under ³H and ⁶³Ni exposure, making it a suitable candidate for applications where controlled dose delivery is desired. These findings contribute to the literature by offering a comparative evaluation of the attenuation performance of 3D printable materials against different beta radiation sources and by providing a basis for material selection in beta applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111864"},"PeriodicalIF":2.3,"publicationDate":"2025-09-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145026715","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":"Development of a real-time interactive physics simulator for the Dalat Nuclear Research Reactor of Vietnam","authors":"Truong Hoang Tuan,&nbsp;Cao Thanh Long,&nbsp;Ho Manh Dung,&nbsp;Huynh Dong Phuong,&nbsp;Nguyen Hoang Nhat Khang","doi":"10.1016/j.anucene.2025.111843","DOIUrl":"10.1016/j.anucene.2025.111843","url":null,"abstract":"<div><div>This paper presents the development of DalatSim, a real-time interactive physics simulator for the Dalat Nuclear Research Reactor (DNRR) in Vietnam. DalatSim is a modified version of RELAP5 that dynamically computes and predicts reactor transients. It integrates a reactor point kinetics solver based on SUNDIALS, which has been validated against published benchmarks. Experimental results confirm the simulator’s effectiveness in modeling operational procedures under normal conditions and hypothetical transient scenarios. Comparisons with recorded DNRR power data demonstrate strong agreement. The simulator is designed to support operator training, cost-effectiveness education, and the preservation of reactor operational knowledge and experience.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111843"},"PeriodicalIF":2.3,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019882","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":"A nodalization study on modeling the containment and reactor pool of an integral PWR","authors":"Tuomo Sevón","doi":"10.1016/j.anucene.2025.111866","DOIUrl":"10.1016/j.anucene.2025.111866","url":null,"abstract":"<div><div>This study investigates the impact of nodalization of the containment and reactor pool of a small integral Pressurized Water Reactor (iPWR) using the MELCOR 2.2 code. The research focuses on Design 1 of the EU-funded SASPAM-SA project, featuring a containment partially submerged in the reactor pool. Two accident scenarios were analyzed: a Design Basis Accident (DBA) and a severe accident. The simulations were conducted with four different nodalizations: a detailed base model, a single-volume containment model, a single-volume pool model, and a single-volume model for both the containment and the pool. The results indicate that while detailed nodalization can simulate temperature stratification, its effect on the overall accident simulation results, pressures and fission product releases is limited. The findings suggest that differences between single-volume and more detailed nodalizations are relatively small, with the detailed nodalization providing slightly lower containment pressures during the DBA scenario due to more efficient heat transfer.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111866"},"PeriodicalIF":2.3,"publicationDate":"2025-09-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019881","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":"A pile noise experiment in current mode in the 37 fuel assemblies core configuration of the nuclear research reactor LR-0","authors":"Romain Boffy ,&nbsp;Benoit Geslot ,&nbsp;Eliott Lerouge ,&nbsp;Laurent Pantera ,&nbsp;Vlastimil Juricek ,&nbsp;Michal Košt'ál","doi":"10.1016/j.anucene.2025.111836","DOIUrl":"10.1016/j.anucene.2025.111836","url":null,"abstract":"","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111836"},"PeriodicalIF":2.3,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145010367","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":"Analysis of heat threshold in contact melting process","authors":"Wenzhen Chen,&nbsp;Junjie Ma","doi":"10.1016/j.anucene.2025.111851","DOIUrl":"10.1016/j.anucene.2025.111851","url":null,"abstract":"<div><div>In the processes of core meltdown under severe reactor accidents and nuclear waste self-burial, the main heat source is radioactive decay heat. Melting will cease when the heat generated by the source is insufficient to melt the surrounding medium. For contact melting induced by a decay heat source under the third-type boundary condition, the concept of heat threshold is proposed. This paper analyzes and calculates the heat threshold and the time to reach this threshold during the melting process of a spherical heat source under different conditions, discusses the main factors affecting the heat threshold. It is found that for contact melting with a heat source of finite heat capacity or decay heat, there will be cases of melting cessation or heat threshold; the heat threshold is mainly affected by the initial temperature of the heat source under severe reactor accidents; the rock thermal conductivity has a significant impact on the melting distance and heat threshold during the nuclear waste self-burial process.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111851"},"PeriodicalIF":2.3,"publicationDate":"2025-09-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145019880","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":"Choosing the right reactor technology: A key step to strengthen nuclear security in West Africa’s nuclear energy plans","authors":"S.A. Birikorang ,&nbsp;Y. Mansung ,&nbsp;C. Marianno ,&nbsp;M. Fuhrmann ,&nbsp;S. Adu","doi":"10.1016/j.anucene.2025.111837","DOIUrl":"10.1016/j.anucene.2025.111837","url":null,"abstract":"<div><div>This study examines the critical role of national capacity and technology selection in enhancing nuclear security within West Africa’s emerging nuclear energy landscape. An assessment of regional countries revealed varied levels of institutional resilience, security preparedness, and commitment to international safeguards, underscoring the necessity for robust national frameworks. Employing a fuzzy logic-based model, three Small Modular Reactor (SMR) designs, NuScale, EM²-HTGR, and LFTR-MSR, were evaluated across 11 nuclear security indicators spanning two fuel cycle stages. Among them, NuScale ranked highest with a final score of 4.8038, followed by EM²-HTGR (3.9600) and LFTR-MSR (3.8542). NuScale’s superior performance confirms its suitability for the region. The study recommends a regional systems-thinking approach that integrates technical reliability, capacity building, and policy harmonization to ensure secure nuclear deployment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111837"},"PeriodicalIF":2.3,"publicationDate":"2025-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145004010","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":"A fast and high-precision gamma-ray buildup factor prediction model based on ensemble machine learning","authors":"Haowei Wang ,&nbsp;Xingfu Cai ,&nbsp;Minjun Peng ,&nbsp;Miao Yang ,&nbsp;Hongyi Yao","doi":"10.1016/j.anucene.2025.111826","DOIUrl":"10.1016/j.anucene.2025.111826","url":null,"abstract":"<div><div>While the Geometric Progression (GP) fitting formula achieves near-unbiased accuracy in predicting buildup factors, its parameter determination process is computationally intensive and fails to meet the high real-time demands of radiation field calculations during nuclear accidents. Based on machine learning technology, this paper proposes a highly generalizable Ensemble Machine Learning model that integrates the advantages of multiple models through a stacking algorithm, significantly improving prediction accuracy and generalization performance. This provides an efficient and reliable solution for predicting buildup factors of new materials and greater penetration depths, thereby meeting the real-time and accuracy requirements of complex radiation field calculations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111826"},"PeriodicalIF":2.3,"publicationDate":"2025-09-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145004011","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":"A novel method to estimate the high-order flux moment under the isotropic source assumption","authors":"Jinchao Zhang ,&nbsp;Qian Zhang ,&nbsp;Shuai Qin ,&nbsp;Wei Cao ,&nbsp;Shifu Wu ,&nbsp;Qiang Zhao ,&nbsp;Kun Zhuang","doi":"10.1016/j.anucene.2025.111840","DOIUrl":"10.1016/j.anucene.2025.111840","url":null,"abstract":"<div><div>Collapsing coarse-group high-order scattering matrix using the accurate flux moment is important for the fast reactor calculations. However, current methods for calculating the flux moment are time-consuming because of the treatment of the anisotropic source and the slow convergence rate. To address this issue, this study proposes a novel method to estimate the high-order flux moment based on the isotropic source assumption. The transport equation is solved using the Method of Characteristics transport method under the time-saving isotropic source assumption. The flux moment is calculated with the angular flux saved during the source iterations and is used to collapse coarse-group high-order scattering matrix as an approximation of the actual flux moment.</div><div>To validate the proposed method, two simplified problems are used to assess its performance and two 2-D fast reactor problems are employed to explore its range of applicability. Numerical results for the simplified problems indicate that the high-order scattering matrix collapsed using the proposed method achieves accuracy comparable to that obtained under the anisotropic source assumption. For the 2-D problems, the results demonstrate the effectiveness of the proposed method, with the eigenvalue error within −200 pcm and the maximum relative error of assembly power within 1%. These findings highlight the accuracy and efficiency of the proposed method in predicting the flux moment.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111840"},"PeriodicalIF":2.3,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144996932","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":"Interface capturing simulations of bubble injection dynamics: Effects of surface tension and orifice shape in molten salt coolants","authors":"Farid Ahmed ,&nbsp;Katharina Stapelmann ,&nbsp;Igor A. Bolotnov","doi":"10.1016/j.anucene.2025.111842","DOIUrl":"10.1016/j.anucene.2025.111842","url":null,"abstract":"<div><div>Advanced reactor designs, like Molten Salt Reactors (MSR) are outfitted with a fission gas removal mechanism, such as Xe gas. In this mechanism, bubbles of helium gas are injected into the molten fuel salt to remove the Xe gas. It is important to demonstrate a robust methodology capable of predicting the bubble’s behavior and void fraction within the nuclear reactor to evaluate the efficiency of such systems. However, experimental investigations of bubble dynamics in high-temperature molten salts such as FLiNaK are challenging due to extreme operating conditions and different coolant opacity levels. The present study applies interface capturing simulations (ICS) to model the bubble formation and rise dynamics when injected through oval and circular-shaped orifices with an equivalent diameter of 0.5 mm into quiescent fluid. A finite element method-based flow solver and level set approach is applied for capturing the air/water interfaces. The bubble departure diameter and time, bubble shapes, and deformability factors are evaluated for both the circular and oval-shaped orifices along with different surface tension values. The bubble departure diameter and the temporal evolution of the bubble shapes obtained from the numerical simulation are validated against experimental data. The results indicate that surface tension governs bubble departure: lower surface tension leads to earlier detachment and elongated trajectories. Detachment length is inversely related to surface tension. Moreover, for helium bubbles in FLiNaK, higher surface tension yields stable, smoother trajectories, and less wake instability compared to air bubbles in water. System-level codes as well as multiphase CFD approach rely on closure laws based on average bubble size and other local parameters. The demonstrated and validated ICS approach can help quantify those required parameters based on fluid conditions, properties as well as injector designs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"226 ","pages":"Article 111842"},"PeriodicalIF":2.3,"publicationDate":"2025-09-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144996934","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}