Annals of Nuclear Energy最新文献_第8页

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-06 DOI: 10.1016/j.anucene.2025.111757

Hossam H. Abdellatif , David Arcilesi , Richard Christensen , Arsen S. Iskhakov

{"title":"Similarity analysis of high-Prandtl surrogate fluids for thermal-hydraulic studies of molten salt reactors","authors":"Hossam H. Abdellatif , David Arcilesi , Richard Christensen , Arsen S. Iskhakov","doi":"10.1016/j.anucene.2025.111757","DOIUrl":"10.1016/j.anucene.2025.111757","url":null,"abstract":"<div><div>Molten salt reactors (MSRs) require high-temperature testing of corrosive fluids like FLiBe and FLiNaK, complicating integral and separate effect experiments. To overcome these limitations, this study employs a similarity-based methodology, grounded in the hierarchical two-tier scaling (H2TS) framework to investigate the performance of Therminol-66, DOWTHERM A, and DOWTHERM RP as low-temperature, high-Prandtl-number surrogates for fluoride salts. We first match each simulant’s Prandtl number to its corresponding salt at the reactor’s mean operating temperature, then derive all relevant non-dimensional groups under forced and natural circulation. Applying this to three MSR designs, KP-FHR, FuSTAR, and FUJI-233Um, yields compact, low-power test loops with zero distortion in all primary groups and negligible Biot-number error. Nusselt-number predictions using standard correlations show that heat transfer is replicated to within 1% over the target Re-Pr range. These results establish a practical, safe, and scalable pathway for high-fidelity thermal-hydraulic experiments supporting future MSR developments.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111757"},"PeriodicalIF":2.3,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144781664","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

Flow and heat transfer characteristics of 3 × 3 rod bundles in natural circulation loop with asymmetrical inclination 非对称倾斜自然循环回路中3 × 3杆束流动传热特性

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-06 DOI: 10.1016/j.anucene.2025.111784

Qiang Lian , Jing-Liang Bi , Si-You Yin , Yu Liang , Qi Lu , Lu-Teng Zhang , Liang-Ming Pan

{"title":"Flow and heat transfer characteristics of 3 × 3 rod bundles in natural circulation loop with asymmetrical inclination","authors":"Qiang Lian , Jing-Liang Bi , Si-You Yin , Yu Liang , Qi Lu , Lu-Teng Zhang , Liang-Ming Pan","doi":"10.1016/j.anucene.2025.111784","DOIUrl":"10.1016/j.anucene.2025.111784","url":null,"abstract":"<div><div>As a most common structure in nuclear reactor core, rod bundle fuel assemblies are widely adopted. With the continuous development of ocean nuclear power platforms, the influence of the marine environment on the transient characteristics of thermal and hydraulic performance in rod bundle should be considered in detail. In this study, the multi-scale coupling method between RELAP5 and Fluent is adopted to obtain both the systematic behavior and the detailed local phenomena for the 3 × 3 rod bundles in the natural circulation loop with asymmetrical inclination. The coupling method is validated against the experimental results of subcooled boiling and natural circulation under inclination. Then, the single-phase and two-phase thermal–hydraulic characteristics of 3 × 3 rod bundles are investigated under asymmetrical inclination of the natural circulation loop. The results show that under single-phase operation the increase of inclination angle makes the gravitational driving force and the system flow rate in the loop decrease accordingly. Due to the asymmetrical inclination, the system flow rate during forward inclination is smaller than that during reverse inclination. In the rod bundle, the inclination leads to the flow mixing enhancement of low-temperature fluid at the positive side of inclination. Under two-phase operation, the loop characteristics of natural circulation under inclined condition are similar to those under single-phase condition with more drastic change. In the rod bundle, the secondary flow on the cross-section is intensified by the increase of inclination angle. The vapor phase migrates and aggregates towards the wall on the opposite side of the inclination, which may lead to heat transfer deterioration in the reactor core.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111784"},"PeriodicalIF":2.3,"publicationDate":"2025-08-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144781661","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

Mesh-free perturbation source method using the closest point pair method in k-eigenvalue Monte Carlo perturbation calculations 无网格摄动源方法采用最接近点对法进行k-特征值蒙特卡罗摄动计算

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-05 DOI: 10.1016/j.anucene.2025.111787

Toshihiro Yamamoto , Hiroki Sakamoto

{"title":"Mesh-free perturbation source method using the closest point pair method in k-eigenvalue Monte Carlo perturbation calculations","authors":"Toshihiro Yamamoto , Hiroki Sakamoto","doi":"10.1016/j.anucene.2025.111787","DOIUrl":"10.1016/j.anucene.2025.111787","url":null,"abstract":"<div><div>In <em>k</em>-eigenvalue calculations using the Monte Carlo method, the perturbation source method (PSM) is a robust technique for perturbation calculations due to changes in cross　sections or geometry. The PSM requires consideration of changes in the fission source distribution resulting from perturbations. Thus far, the region where the fission source exists has been discretized into small bins, with the fission sources integrated in each bin. This method is not necessarily desirable as the arbitrariness and approximations introduced by discretization can affect the results. In response, this study introduces a new mesh-free method by employing the closest point pair approach in which a fission source generated by perturbation source is moved to the closest fission source and combined there. While involving approximation due to the movement of the fission sources, this mesh-free method circumvents the issues associated with discretization. For perturbations in regions with low fission source density (i.e., near the outer boundary), significant errors arise due to the increased movement distance. To address this issue, we introduce a new method to improve the accuracy of this approach by forcibly increasing the fission sources near the perturbation region. The improved PSM (IPSM) can successfully reproduce results that are comparable to those of the conventional PSM and the reference solution.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111787"},"PeriodicalIF":2.3,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773036","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 critical benchmark experiment of beryllium reflector 铍反射器关键基准实验的验证

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-05 DOI: 10.1016/j.anucene.2025.111763

Yanliang Chang , Huayun Shen , Bin Zhong , Pei Sun , Liang Xing , Bo Gao , Yuting Wei , Changlin Lan

{"title":"Verification of critical benchmark experiment of beryllium reflector","authors":"Yanliang Chang , Huayun Shen , Bin Zhong , Pei Sun , Liang Xing , Bo Gao , Yuting Wei , Changlin Lan","doi":"10.1016/j.anucene.2025.111763","DOIUrl":"10.1016/j.anucene.2025.111763","url":null,"abstract":"<div><div>Critical benchmark experiments constitute the cornerstone for validating and refining nuclear data libraries. Recent advancements, however, have revealed fundamental incompatibilities within multiple beryllium-reflected critical configurations - a paradox that challenges conventional nuclear data validation paradigms. This study pioneers a diagnostic framework integrating sensitivity analysis with uncertainty quantification to establish a similarity metric for Be-related benchmark systems. A paradigmatic analysis of the PMF021 benchmark reveals a critical contradiction: while nuclear data modifications could theoretically induce maximum k<span><math><msub><mrow></mrow><mrow><mi>e</mi><mi>f</mi><mi>f</mi></mrow></msub></math></span> variations of 278.7 pcm, the observed 1135 pcm discrepancy between configurations exceeds theoretical adjustment limits by a factor of 4.1. This breakthrough necessitates two fundamental needs in nuclear data practices: (1) Introduction of a quantitative reliability index for benchmarking conflicting configurations, (2) Mandatory re-evaluation protocols for existing Be-reflected benchmarks. Our findings establish a new parameter, providing an essential discriminator for identifying inconsistent benchmarks that risk misleading the nuclear data community.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111763"},"PeriodicalIF":2.3,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144781659","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

Levels and risk due to uranium content in drinking water of the Chakrata region of Garhwal Himalaya, India 印度加尔瓦尔喜马拉雅查克拉塔地区饮用水中铀含量的水平和风险

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-05 DOI: 10.1016/j.anucene.2025.111778

Shubham Sharma , Krishna Pal Singh , Abhishek Joshi , Rohit Mehra , Taufiq Ahamad , Vikrant Thakur , A.A. Bourai , R.C. Ramola

{"title":"Levels and risk due to uranium content in drinking water of the Chakrata region of Garhwal Himalaya, India","authors":"Shubham Sharma , Krishna Pal Singh , Abhishek Joshi , Rohit Mehra , Taufiq Ahamad , Vikrant Thakur , A.A. Bourai , R.C. Ramola","doi":"10.1016/j.anucene.2025.111778","DOIUrl":"10.1016/j.anucene.2025.111778","url":null,"abstract":"<div><div>Radiological and chemical toxicity analysis in water is vital to ensure the safety of public well-being. Present study examines the concentration of uranium in drinking water samples from the Chakrata region of Uttarakhand, India, alongside key physiochemical properties (i.e., pH, EC, and TDS). A total of 100 drinking water samples were collected and analyzed by using LED fluorimetry. The measured uranium concentrations ranged from 0.54 to 2.63 µg/L, with an average of 1.32 µg/L. The corresponding annual effective doses varied between 4.69––22.67 µSv/y, averaging 11.42 µSv/y. In most of the samples, the uranium levels were found to be within the safe limits recommended by the AERB and the WHO, respectively. Additionally, pH, TDS (Total Dissolved Solids) and EC (Electrical Conductivity) measurements provide insight into the overall water quality in the region. Furthermore, LADD (Lifetime Average Daily Dose), ELCR (Excess Lifetime Cancer Risk) and HQ (Hazard Quotient) values were estimated for a better understanding of radiological and chemical toxicity. These findings bring out the importance of ongoing monitoring efforts to safeguard public health and ensure access to safe drinking water.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111778"},"PeriodicalIF":2.3,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773033","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

Direct numerical simulation of square rod bundle channel and evaluation of common turbulence models 方杆束通道的直接数值模拟及常用湍流模型的评价

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-05 DOI: 10.1016/j.anucene.2025.111755

Heng Miao , Yu Ma , Jie Li, Yahui Wang

{"title":"Direct numerical simulation of square rod bundle channel and evaluation of common turbulence models","authors":"Heng Miao , Yu Ma , Jie Li, Yahui Wang","doi":"10.1016/j.anucene.2025.111755","DOIUrl":"10.1016/j.anucene.2025.111755","url":null,"abstract":"<div><div>Simulation of flow and heat transfer in rod channels is essential for nuclear reactor safety. Its complexity poses significant challenges for computational implementations. This paper presents a numerical study on the thermal-hydraulic behavior in rod channels using the Direct Numerical Simulation (DNS) method. The flow and temperature distributions in a 3 × 3 square rod bundle with a Reynolds number of 22876 are investigated. The flowing and heat transfer characteristics in different regions are analyzed. A comparison of various turbulence models, including the <span><math><mrow><mi>k</mi><mo>−</mo><mi>ɛ</mi></mrow></math></span> models, <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> models, and the Reynolds Stress models, are conducted to evaluate their performance in predicting the velocity and temperature distributions. The results show that the Stress-BSL model and SST <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> model exhibit better overall performance. This work can provide some data reference for the subsequent detailed flow and heat transfer analysis of fuel rod bundles, and provide some insights for the improvement of turbulence models.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111755"},"PeriodicalIF":2.3,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773035","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

Formation mechanism of localized corrosion area for pre-oxidized high-silicon ferritic/martensitic steel immersing in the oxygen-saturated liquid lead–bismuth eutectic (LBE) 预氧化高硅铁素体/马氏体钢浸在饱和氧铅铋共晶（LBE）液中的局部腐蚀区形成机理

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-05 DOI: 10.1016/j.anucene.2025.111779

Xinyu Cao , Junhua You , Yangpeng Zhang , Lijian Rong , Desheng Yan

{"title":"Formation mechanism of localized corrosion area for pre-oxidized high-silicon ferritic/martensitic steel immersing in the oxygen-saturated liquid lead–bismuth eutectic (LBE)","authors":"Xinyu Cao , Junhua You , Yangpeng Zhang , Lijian Rong , Desheng Yan","doi":"10.1016/j.anucene.2025.111779","DOIUrl":"10.1016/j.anucene.2025.111779","url":null,"abstract":"<div><div>To improve the effectiveness of pre-oxidation treatment in preventing lead–bismuth corrosion, this study investigates the underlying causes of localized corrosion areas when the pre-oxidized high-silicon ferritic/martensitic steel were immersed in the oxygen-saturated liquid LBE. The pre-oxidized samples were submerged in oxygen-saturated LBE for 20 h, 200 h and 1000 h at 600 °C respectively. Subsequently, the microstructure of the localized corrosion areas was finely characterized and analyzed. The findings reveal that a small number of thin or discontinuous areas exist within the outermost Cr oxide layer of the pre-oxidized film. In these specific areas, the pre-oxidized film exhibits diminished resistance to the intrusion of Pb and O as well as the outward diffusion of Fe. With the continuous erosion of LBE, O and Fe would rapidly diffuse laterally and longitudinally along numerous interfaces within the loose oxide layer, ultimately forming localized corrosion areas.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111779"},"PeriodicalIF":2.3,"publicationDate":"2025-08-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144773034","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

Hydrogen diffusion layer within solid-state lithium target for compact accelerator-driven neutron source: Irradiation damage, hydrogen concentration, thermal properties and fabrication 紧凑加速器驱动中子源固态锂靶内的氢扩散层：辐照损伤、氢浓度、热性能和制造

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-04 DOI: 10.1016/j.anucene.2025.111781

Yupeng Xie , Xiaobo Li , Fanxi Zhang , Yaocheng Hu , Yixin Si , Xuanyu Meng , Qiuyu Sun , Yaru Wang , Chen Chen , Xiaozhi Zhang , Sheng Wang

{"title":"Hydrogen diffusion layer within solid-state lithium target for compact accelerator-driven neutron source: Irradiation damage, hydrogen concentration, thermal properties and fabrication","authors":"Yupeng Xie , Xiaobo Li , Fanxi Zhang , Yaocheng Hu , Yixin Si , Xuanyu Meng , Qiuyu Sun , Yaru Wang , Chen Chen , Xiaozhi Zhang , Sheng Wang","doi":"10.1016/j.anucene.2025.111781","DOIUrl":"10.1016/j.anucene.2025.111781","url":null,"abstract":"<div><div>This study focused on investigation of the hydrogen diffusion layer (H-D layer) in the solid-state lithium target for compact accelerator-driven neutron source through numerical simulation and experimental results. Simulations based on Monte Carlo and finite element method were employed to explore the influence of H-D layers (tantalum and vanadium) on the irradiation damage resistance and cooling performance of the solid-state lithium target. For cooling performance, the multi-channel target design with a 1 mm channel width and 47 channels, coupled with an H-D layer, demonstrated the best cooling efficiency. Fabrication of the H-D layer involved depositing a 20 μm Ta layer on a Cu substrate by magnetron sputtering at a power of 300 W, exhibiting excellent surface properties with the lowest surface roughness of 13.9 nm, and a dominant β-phase structure as identified by XRD. Additionally, the influence of the Ta film structure and sputtering power on thermal conductivity was analyzed.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111781"},"PeriodicalIF":2.3,"publicationDate":"2025-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144766777","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

Development of core design optimization process: Feasibility study of multivariable optimization via integrated sequential analyses of neutronics, thermal-hydraulics, and fuel integrity evaluation 核心设计优化过程的发展：通过中子、热工和燃料完整性评估的综合序列分析进行多变量优化的可行性研究

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-04 DOI: 10.1016/j.anucene.2025.111754

Kazuki Kuwagaki, Erina Hamase, Kenji Yokoyama, Norihiro Doda, Masaaki Tanaka

{"title":"Development of core design optimization process: Feasibility study of multivariable optimization via integrated sequential analyses of neutronics, thermal-hydraulics, and fuel integrity evaluation","authors":"Kazuki Kuwagaki, Erina Hamase, Kenji Yokoyama, Norihiro Doda, Masaaki Tanaka","doi":"10.1016/j.anucene.2025.111754","DOIUrl":"10.1016/j.anucene.2025.111754","url":null,"abstract":"<div><div>The Japan Atomic Energy Agency has been developing an innovative design approach for advanced reactors such as fast reactors, known as Advanced Reactor Knowledge- and AI-aided Design Integration Approach through the whole plant lifecycle (ARKADIA). One task of ARKADIA is to establish a core design optimization process to automatically identify optimal core and fuel design parameters by combining an optimization method and integrated sequential analyses of neutronics, thermal-hydraulics, and fuel integrity evaluations as well as plant dynamics analysis. The optimization process has been developed in stages. In a previous study, the optimal solution consistent with the reference solution was obtained in a simple two-variable optimization problem by focusing only on neutronics. Herein, the optimization process was extended to multivariable optimization, including other analyses. In particular, an integrated sequential analysis system was developed to evaluate thermal-hydraulics and fuel integrity as well as neutronics in the core. The number of core design variables was increased from two to four. The extended optimization process was applied to two problems of three- and four-variable optimization with multiple constraints. In the three-variable problem, the validity of optimization calculation was shown by the optimal solution matched to the reference solution. In the four-variable problem, the solution satisfied all the defined constraints. These results confirmed the feasibility of the core design optimization process combined with integrated analyses up to four variables.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111754"},"PeriodicalIF":2.3,"publicationDate":"2025-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144766776","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

A hybrid L1-Chebyshev-HPM approach for solving fractional neutron diffusion equations with delayed neutrons 求解含有延迟中子的分数阶中子扩散方程的混合l1 - chebyhev - hpm方法

IF 2.3 3区工程技术

Annals of Nuclear Energy Pub Date : 2025-08-01 DOI: 10.1016/j.anucene.2025.111729

Ujwal Warbhe

{"title":"A hybrid L1-Chebyshev-HPM approach for solving fractional neutron diffusion equations with delayed neutrons","authors":"Ujwal Warbhe","doi":"10.1016/j.anucene.2025.111729","DOIUrl":"10.1016/j.anucene.2025.111729","url":null,"abstract":"<div><div>This work presents a novel derivation of a fractional neutron diffusion equation that incorporates memory effects and anomalous transport phenomena observed in nuclear reactor cores. Starting from the classical neutron diffusion model, the derivation extends the framework through the introduction of fractional calculus in the constitutive relations. The resulting formulation accommodates subdiffusive behavior and provides a more accurate description of delayed neutron kinetics. To numerically solve the derived fractional equation, a hybrid algorithm is developed that integrates an L1 finite difference approximation for temporal discretization, Chebyshev spectral collocation for spatial accuracy, and the Homotopy Perturbation Method (HPM) to treat nonlinearity. The method is rigorously analyzed for convergence and stability, yielding exponential spatial convergence and near-first-order temporal accuracy. Comprehensive numerical experiments, including step, ramp, and sinusoidal reactivity cases, demonstrate the superior accuracy and computational efficiency of the proposed scheme compared with traditional methods (e.g., finite element and B-spline collocation (Roul et al., 2020)). Unlike prior works, this framework uniquely combines spectral accuracy with fractional calculus, addressing gaps in modeling anomalous neutron transport in heterogeneous media (Espinosa-Paredes, 2023). This new framework offers a robust tool for reactor dynamics analysis under anomalous diffusion regimes, with potential applications in reactor safety and control by enabling high-fidelity simulations of memory-dependent transport.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"225 ","pages":"Article 111729"},"PeriodicalIF":2.3,"publicationDate":"2025-08-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750694","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