Annals of Nuclear Energy最新文献

{"title":"Novel designs and performance appraisal of non-uniformly heated annuli for heat transfer augmentation with supercritical carbon dioxide","authors":"Ashok Kumar Gond,&nbsp;Amaresh Dalal,&nbsp;Dipankar N. Basu","doi":"10.1016/j.anucene.2025.111393","DOIUrl":"10.1016/j.anucene.2025.111393","url":null,"abstract":"<div><div>Deterioration of heat transfer and possible appearance of high temperatures are common concerns in supercritical flow channels, inspiring several innovating modifications in traditional tubes. Despite the relevance of annular geometry and rod bundles in high-power applications, similar efforts are quite rare in configurations beyond the conventional channels. Present study aims at addressing this research gap by proposing and analyzing two novel designs of supercritical flow channel with variable flow areas. Annular channels with continually increasing or decreasing heater diameter within a shell, accordingly allowing axial variation in heat flux based on the local thermal conditions, facilitate converging or diverging flow paths. Five different taper angles are contemplated for either of the geometries, and numerical simulations are performed to envisage their relative performances during energy addition with supercritical carbon dioxide as the working medium. Both the designs are able to demonstrate marked improvements in overall thermalhydraulic response, with the converging duct being identified as the more preferable choice. Local heat transfer coefficient at the channel exit can be as much as 126% greater for a converging channel compared to a plain annular one, without any significant rise in pressure losses. A converging channel is also able to maintain lower temperature levels throughout the flow path, and the effects are more prominent with greater tapering. Diverging channel also records considerable gain in heat transfer, but may experience higher local temperatures and augmented pressure losses. Buoyancy effect is found to be dominant within short entrance region, while flow acceleration governs the extent of interactions in later segments. Converging duct is able to sustain temperature close to pseudocritical value within the buffer region, which can be a primary reason for its superiority. The disparity between both configurations are less significant at lower flow rates and higher powers, and the use of diverging channel can be recommended only for large power-to-flow-rate ratios, which can limit temperature ranges better than converging channel only for such specific cases.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111393"},"PeriodicalIF":1.9,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738351","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":"An intelligent control rod movement strategy for boron-free reactor core using multi-layer perceptron machine learning model","authors":"M. Hosseinllu ,&nbsp;M. Abbasi ,&nbsp;O. Safarzadeh ,&nbsp;F. Dehghani","doi":"10.1016/j.anucene.2025.111405","DOIUrl":"10.1016/j.anucene.2025.111405","url":null,"abstract":"<div><div>The boron-free operation of Small Modular Reactor (SMR) core requires efficient approaches to manage excess reactivity throughout prolonged operational cycles. Adjusting Control Rods (CRs) is the only way to compensate reactivity and regulate<!--> <!-->the reactor power during the operational cycle of boron-free cores. Therefore, developing a proper CR movement strategy throughout the cycle length is crucial for boron-free cores. This study aims to apply data mining methods within machine learning approach to forecast critical CR positions at each burnup level of the boron-free core using a Multi-Layer Perceptron (MLP) model. To achieve this goal, the design of the boron-free core Control Banks (CBs) and their computations, are investigated. Furthermore, the effects of CR movement on core neutron physics parameters are considered.</div><div>The regression values for training, testing, and all datasets are calculated. The results indicate that the prediction of critical CR movement strategy is properly done by the developed MLP model. The trained MLP model operates extremely quickly (less than 1 sec) and can serve as a quick support model for forecasting CR movement strategy. The forecasting results of the developed model, based on known and unknown data, verify a high correlation between forecasted and real values, demonstrating that the performance of the developed model is good and has high accuracy.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111405"},"PeriodicalIF":1.9,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738352","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":"BTRR model development and verification for Monte Carlo simulation","authors":"Saad Islam ,&nbsp;Mohammad Abdul Motalab","doi":"10.1016/j.anucene.2025.111419","DOIUrl":"10.1016/j.anucene.2025.111419","url":null,"abstract":"<div><div>The Bangladesh Atomic Energy Commission’s TRIGA Mark II research reactor (BTRR), located near Dhaka, has been operational since September 1986, primarily for training, education, radioisotope production, and research. Currently, the Reactor Physics and Engineering Division is enhancing its analysis methods by employing modern tools like high-fidelity Monte Carlo analysis and introducing advanced codes and cross-section libraries. This study details the new reactor model implemented by means of the OpenMC code using the nuclear cross-section data libraries based on the ENDF/B-VIII.0 and ENDF/B-VII.1, focusing on calculating critical reactor parameters such as core excess reactivity, shutdown margin, control rod worth, power peaking factors, etc. The accuracy of the model verified by comparisons with previous calculations and experimental data showed good agreement.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111419"},"PeriodicalIF":1.9,"publicationDate":"2025-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143747658","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":"Effect of hydration at various temperatures on the thermal conductivity of compacted Indian bentonites","authors":"Banavath Prasad Nayak,&nbsp;Ramakrishna Bag","doi":"10.1016/j.anucene.2025.111413","DOIUrl":"10.1016/j.anucene.2025.111413","url":null,"abstract":"<div><div>Thermal conductivity of bentonite is one of the key parameters influencing thermal behavior of geological repositories, where bentonite can serve as a buffer material. ​This study experimentally investigated the effects of hydration period with temperature (27℃→80℃→100℃→120℃) on the thermal conductivity of bentonites, compacted with varying water content and dry density. It was observed that thermal conductivity increased with hydration period at 27℃ for specimens with lower water content, whereas it decreased for those with higher water content for given dry density. However, it increases with temperature, increased significantly during early hydration period and tended slow rate increase after 5 days of hydration for temperatures above 80℃. The change in thermal conductivity with temperature during hydration was minimal for dry specimens, whereas it changed significantly in specimens with higher water content. Hydration of compacted bentonites with various temperatures decreased inter-aggregate pore fractions, increasing particle contact area, confirmed by microstructure investigations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111413"},"PeriodicalIF":1.9,"publicationDate":"2025-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738350","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 and validation of NECP-SARAX: An advanced neutron analysis code system for accelerator-driven subcritical systems","authors":"Wenjie Chen ,&nbsp;Xianan Du ,&nbsp;Xunchao Zhang ,&nbsp;Youqi Zheng ,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2025.111404","DOIUrl":"10.1016/j.anucene.2025.111404","url":null,"abstract":"<div><div>An accelerator-Driven Subcritical System (ADS) is a facility that utilizes high-energy spallation neutrons to transmute minor actinide nuclides and long-lived fission products, supporting the “partitioning and transmutation” strategy. However, the energy of spallation neutrons can reach hundreds of megaelectron volts, which represents a huge challenge to the deterministic neutron transport methods. To address this challenge, NECP-SARAX, an advanced neutron analysis code system, has been developed by the Nuclear Engineering Computational Physics Laboratory, Xi’an Jiaotong University. To apply NECP-SARAX code in the ADS, the influence of high-energy spallation neutrons on the system was evaluated as the first step. Next, SARAXLIB-HE, a high-energy neutron database, was established based on high-energy evaluation databases JENDL4.0-HE, JENDL-HE-2007, and IAEA-ADS-HE. To address neutron leakage and the impact of external sources, the NECP-SARAX code has been enhanced by incorporating a fixed source transport solver for fission materials during the process of generating few-group cross-sections. Finally, verifications of assembly and reactor benchmarks demonstrated that the SARALIB_HE database and the SARAX code are suitable for calculating and analyzing the ADS.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111404"},"PeriodicalIF":1.9,"publicationDate":"2025-03-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738334","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":"Inversion of nuclear accident source terms combining Bayesian method with machine learning","authors":"Zhonghao Li ,&nbsp;Bo Cao ,&nbsp;Dingping Peng ,&nbsp;Qingyue You ,&nbsp;Xuewei Miao ,&nbsp;Zhouliang Chen","doi":"10.1016/j.anucene.2025.111418","DOIUrl":"10.1016/j.anucene.2025.111418","url":null,"abstract":"<div><div>Source term inversion plays a critical role in consequence assessment during nuclear accidents. This study investigates three Bayesian approaches for machine learning-based source term inversion, with datasets generated by the radioactive nuclide atmospheric dispersion program RADC. Firstly, a Bayesian neural network (BNN) was designed using Python to predict the release rate of I-131 during nuclear accidents. The BNN achieved a mean absolute percentage error (MAPE) of 7.92%, showing significantly higher accuracy and robustness compared to a backpropagation neural network (BPNN) with an identical structure and under the same accident scenarios. Secondly, both BNN and BPNN are further enhanced through Bayesian optimization, achieving a significantly improved MAPE values of 5.78% and 3.80% respectively. Furthermore, The Monte Carlo Dropout method was used to generate confidence intervals for the BPNN, offering an uncertainty analysis approach for conventional neural networks in source term inversion. However, its uncertainty analysis showed greater fluctuations compared to BNN.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111418"},"PeriodicalIF":1.9,"publicationDate":"2025-03-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143738349","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":"Novel geopolymer materials for fast and thermal neutron shielding","authors":"Jianxin Zhou ,&nbsp;Alexander Fields ,&nbsp;Shaffer Bauer ,&nbsp;Ali Ozer ,&nbsp;Waltraud M. Kriven ,&nbsp;Angela Di Fulvio","doi":"10.1016/j.anucene.2025.111388","DOIUrl":"10.1016/j.anucene.2025.111388","url":null,"abstract":"<div><div>In this work, we developed novel geopolymer (GP) composites for neutron shielding applications by dispersing up to 50 wt<span><math><mtext>%</mtext></math></span> polyethylene (PE) and boron carbide (B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C) powders into a potassium-based GP (KGP) matrix. We measured their attenuation coefficients to fast neutrons using a spontaneous fission ²⁵²Cf source and a fusion deuterium–tritium generator and to thermal neutrons using a moderated ²⁵²Cf source. The PE-based KGP were designed to reduce the neutron primary energy through neutron–hydrogen collisions, while ¹⁰B in B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C was exploited to capture and stop thermal neutrons. The fast-neutron shielding properties of 50 wt<span><math><mtext>%</mtext></math></span> PE-KGP were superior, with an attenuation coefficient to fission neutrons up to 30% higher than that of high-density concrete. The B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C-based KGP formulation with B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C at 25 wt<span><math><mtext>%</mtext></math></span> and 50 wt<span><math><mtext>%</mtext></math></span> outperforms commercial boron-loaded flexible materials by approximately 18% and 50%, respectively. In simulation, we combined KGP with PE and B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C dispersants. A prototypical material comprised of 50 wt<span><math><mtext>%</mtext></math></span> PE-KGP and 50 wt<span><math><mtext>%</mtext></math></span> B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C-KGP significantly outperformed barite in reducing the dose due to fission-based spectra, similar to those found at nuclear reactors. Additionally, KGP shows excellent adhesive properties, high-temperature resistance, and can be molded into any size or shape, as well as 3D or 4D printed. Therefore, PE and B<span><math><msub><mrow></mrow><mrow><mn>4</mn></mrow></msub></math></span>C-loaded KGP and associated manufacturing techniques can be effectively employed in nuclear power plants and other facilities that rely on intense neutron sources, significantly enhancing their safety and cost-effectiveness.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111388"},"PeriodicalIF":1.9,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143724802","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}

{"title":"Investigation of sodium compatibility of oxide ceramics for development of Advanced core Catcher for SFRs","authors":"Prabhat Kumar Shukla ,&nbsp;Hemanth Rao E ,&nbsp;E. Vetrivendan ,&nbsp;Vani Shankar ,&nbsp;Anish Kumar ,&nbsp;M. Muthuganesh ,&nbsp;Muthu Ambika ,&nbsp;Sanjay Kumar Das ,&nbsp;PK Chaurasia ,&nbsp;B.K. Sreedhar","doi":"10.1016/j.anucene.2025.111360","DOIUrl":"10.1016/j.anucene.2025.111360","url":null,"abstract":"<div><div>Core Catcher (CC) is a passive safety device, provided in nuclear reactors for safe retention of core debris resulting from a hypothetical core melt accident. For Sodium cooled Fast Reactors (SFR), in-vessel CC is preferred for containing radioactivity within the reactor vessel. Design of CC for new generation of SFRs consider whole core retention. To protect the CC against corium impingement and limiting the temperature of structural members within safe limit, a refractory sacrificial lining is being developed. Among several requirements, long term compatibility of sacrifical materials with liquid sodium under normal operation as well as accident condition are essential. In the present work, alumina and Yttria Stabilized Zirconia (YSZ) have been investigated as candidate materials. Test specimens of alumina and YSZ were exposed to liquid sodium at various temperatures from 400 °C to 950 °C for time durations up to 10000 h. The exposed specimens were investigated for microstructural damage and chemical interaction with sodium. YSZ specimens were found to be intact for temperatures up to 950 °C for 24 h, whereas alumina specimens revealed surface cracking and loss of strength above 850 °C. Though, no detectable chemical interaction with sodium was noticed for alumina and zirconia specimens at 400 °C for a long-term exposure, formation of sodium aluminate and sodium zirconate was observed at higher temperatures.</div><div>YSZ specimens exhibited insignificant degradation at high temperature sodium exposure, while alumina resulted in severe microstructural degradation. Important findings are outlined in the paper.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111360"},"PeriodicalIF":1.9,"publicationDate":"2025-03-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143724540","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":"Research on energy group structure optimization method for pressure water reactor plate-type fuel based on particle swarm algorithm","authors":"Jiannan Li,&nbsp;Ling Chen,&nbsp;Song Li,&nbsp;Yongfa Zhang,&nbsp;Jianli Hao","doi":"10.1016/j.anucene.2025.111422","DOIUrl":"10.1016/j.anucene.2025.111422","url":null,"abstract":"<div><div>The energy group structure significantly affects the accuracy and efficiency of cross-section and transport calculations. Plate-type reactors have significant differences in neutron dynamics compared to traditional rod reactors, which means that traditional energy group structures are not suitable for plate-type reactors. Therefore, it is necessary to explore new energy group structures to adapt to the complex structure of plate-type reactor cores. The paper proposes an optimization method suitable for the energy group structure of plate-type fuel based on the particle swarm algorithm. To ensure the universality of the obtained multi group structure in the plate-type fuel structure, the sensitivity of the plate elements parameters has been analyzed. Multiple examples are set up for optimization, with the average difference in <em>k</em>_∞ between multi-group and coarse-group transport calculations as the objective function, and the optimal energy group boundary position is sought by updating generation by generation. Based on the above method, the 15-group, 29-group, 37-group, and 45-group structures suitable for multi-group transport calculations of pressurized water reactor plate-type cores are optimized. Numerical verification shows that the group structure proposed in this paper has high computational accuracy for various operating conditions of pressurized water reactor plate-type cores, and significantly improves computational efficiency compared to the traditional 361-group structure.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111422"},"PeriodicalIF":1.9,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143724537","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":"Thermal hydraulic deformation analysis of fine 5 × 5 rod bundle fuel assembly based on multi-field coupling","authors":"Yunlong Wang,&nbsp;Rongchang Qiu,&nbsp;Guopeng Liang,&nbsp;Changjie Ying,&nbsp;Weidi Sun,&nbsp;Guan Guan,&nbsp;Chaoguang Jin","doi":"10.1016/j.anucene.2025.111421","DOIUrl":"10.1016/j.anucene.2025.111421","url":null,"abstract":"<div><div>This paper proposes a new multi physics field simulation method that integrates computational fluid dynamics, fluid structure coupling, and heat transfer theory for comprehensive analysis of complex reactions inside nuclear fuel cores. The methodology enables precise modeling of a 5 × 5 rod bundle fuel assembly, considering temperature-dependent material properties and non-uniform thermal convection. A bidirectional coupling analysis is conducted to explore the intricate interactions between thermal, fluid, and solid domains. The transient simulation, spanning 5 s, reveals a maximum displacement of 3.3 μm in fuel rods, with a cloud map highlighting temperature gradients and distinct deformations along radial and height directions. This research presents a feasible and efficient framework for importing and simulating three-dimensional fine models, yielding crucial displacement data pertinent to the thermal–hydraulic performance of fuel components. Its findings carry substantial engineering value for advancing nuclear reactor research, design optimization, and safe operation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"218 ","pages":"Article 111421"},"PeriodicalIF":1.9,"publicationDate":"2025-03-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143724539","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}