Annals of Nuclear Energy最新文献

{"title":"Plutonium diversion detection in molten salt reactors via gamma emitter signature","authors":"Alvin J.H. Lee,&nbsp;Tomasz Kozlowski","doi":"10.1016/j.anucene.2026.112149","DOIUrl":"10.1016/j.anucene.2026.112149","url":null,"abstract":"<div><div>In Molten Salt Reactors (MSRs), the easy accessibility of the fuel salt is a potential avenue for plutonium diversion and safeguard inspectors must have the means to detect such diversion to limit nuclear proliferation. In this work, we developed a mathematical basis to translate the radioactivities of select gamma emitting fission products into the fissile isotope ratios of the reactor, and described a methodology to detect plutonium diversion. This work identified ^138mCs/^134mI as the species pair that can predict the fissile isotope ratios with good accuracy while under chemical effects such as reprocessing. Simulated diversion cases demonstrated good accuracy of around 0.96% discrepancy from the Monte Carlo estimate for ^138mCs/^134mI, and around 0.034% for other potential pairs when less limiting conditions were considered (e.g., finite but slow precursor removal vs decay). The detection methodology contributes to the existing MSR safeguards efforts, which is necessary for the successful deployment of MSRs.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112149"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035494","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 on the characteristics of the flow inside a reactor for different injet angle of inlet nozzle","authors":"Tiancai Tan,&nbsp;Lixia Gao,&nbsp;Hongwei Qiao,&nbsp;Jianzhong Ma,&nbsp;Litao Liu","doi":"10.1016/j.anucene.2026.112194","DOIUrl":"10.1016/j.anucene.2026.112194","url":null,"abstract":"<div><div>CFD simulation is an important means to study the flow field inside a reactor. In this paper, in order to analyze the effects of different injet angle of inlet nozzle on the characteristics of the flow inside a reactor with three inlets and three outlets, a CFD simulation is conducted. The effects of different injet angle of inlet nozzle on the flow characteristics in inlet nozzle, downcomer and lower plenum of pressure vessel, and on the flow distribution at core inlet are investigated. Simulation results reveal that the injet angle of inlet nozzle has a significant effect on the internal flow characteristics of reactor. The fluid velocity of the inlet nozzle becomes more and more uneven if the injet angle gradually increases. The coolant forms a rotational flow around the downcomer and a large rotating fluid in low plenum due to the elbow near inlet nozzle. The flow distribution at the core inlet becomes unstable or even deteriorates due to the big bending angle of the inlet pipe.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112194"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186631","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":"Implementation of a projection-based control rod de-cusping method in the Finite Element Neutronic Code FENNECS","authors":"Romain Henry,&nbsp;Jérémy Bousquet,&nbsp;Armin Seubert","doi":"10.1016/j.anucene.2026.112116","DOIUrl":"10.1016/j.anucene.2026.112116","url":null,"abstract":"<div><div>This paper presents a new feature of the Finite Element Method (FEM) code FENNECS (Finite ElemeNt NEutroniCS) for modelling reactivity control systems. The interface between materials within a finite element (usually referred to as a mixed element) is modelled using a flux-weighting method. While the method has demonstrated its accuracy in modelling the vertical movement of control rods in traditional Light Water and Fast Reactors (LWR and FR), it has limitations in modelling the rotation of control drums.</div><div>The projection-based cusping treatment is another method that defines an effective homogenized cross-section for the mixed element. Unlike the flux-weighting method, this method does not involve any approximations. Instead, it exactly solves the weak form of the neutron diffusion equation.</div><div>In order to illustrate the appropriate implementation of the method in the code, three exercises were solved. A comparison with the legacy flux-weighting model was performed, highlighting the benefits of the projection-based de-cusping method.</div><div>In every case, if it is not completely removed, the cusping effect is mitigated, enabling the production of a solution compatible with nuclear safety analysis. Furthermore, it has been demonstrated that the projection-based method clearly outperforms the flux and volume weighting method in terms of accuracy.</div><div>In terms of runtime, the projection-based method has demonstrated an average reduction of 40% for control rod exercises, while control drum exercises have shown a reduction of 15%.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112116"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145975288","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 tuning-free POD-FV-ROM with automatic boundary enforcement for practical thermal-hydraulic applications","authors":"Yang Di ,&nbsp;Zhang Chunyu ,&nbsp;Lin Jiming ,&nbsp;Ding Peng","doi":"10.1016/j.anucene.2026.112163","DOIUrl":"10.1016/j.anucene.2026.112163","url":null,"abstract":"<div><div>Reduced order models (ROMs) have been widely adopted to accelerate high-fidelity simulations while retaining essential predictive accuracy. However, two gaps persist for POD-FV-ROMs in the context of thermal-hydraulic problems. First, when Dirichlet and Neumann boundaries coexist, a robust and effective strategy to enforce boundary constraints in the ROM is still missing. Second, the ROM performance has not been convincingly demonstrated on practical three-dimensional cases with complex geometries. To close these gaps, a ROM combining POD-Galerkin projection with supremizer stabilization and POD-RBF interpolation was investigated, together with a penalty formulation for boundary treatment. Two novel <em>a priori</em> estimators are proposed to determine penalty factors (PFs) without manual tuning, one based on residuals relevant to projected snapshots and one derived from optimization to a domain-wise error. The entire ROM framework was assessed on an 2 × 2 helical cruciform fuel assembly discretized into around 24 million cells with parameterized boundary conditions. The results demonstrated that, under appropriate PFs, the ROM delivered satisfying accuracy while achieving a speed up of five orders of magnitude.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112163"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035530","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":"Performance of multiple-type reference electrode oxygen sensors in LBE","authors":"Ruixian Liang,&nbsp;Hui Li,&nbsp;Huiping Zhu,&nbsp;Hao Wu,&nbsp;Haicai Lyu,&nbsp;Zulong Hao,&nbsp;Yang Liu,&nbsp;Fang Liu,&nbsp;Fenglei Niu","doi":"10.1016/j.anucene.2026.112129","DOIUrl":"10.1016/j.anucene.2026.112129","url":null,"abstract":"<div><div>The performance of various oxygen sensors in liquid lead–bismuth eutectic (LBE) alloy varies significantly across different temperature ranges. Therefore, it is necessary to establish a comprehensive temperature-related calibration database to achieve real-time dynamic calibration and compensation for sensor measurements. In this paper, multiple types of oxygen sensors have been developed based on 8YSZ ceramic tubes. The air reference (LSCF/Air, LSM/Air and Ag/Air) oxygen sensors and metal/metal oxide (Cu/Cu₂O, Fe/Fe₃O₄, Ni/NiO, Bi/Bi₂O₃, and In/In₂O₃) reference oxygen sensors were tested under different temperature variations to obtain their operating characteristics in different temperature ranges. The air reference oxygen sensors have been demonstrated to exhibit excellent response speed, accuracy and stability within the range of 205 ∼ 550℃. The metal/metal oxide reference oxygen sensor is more suitable for applications in the medium to high temperature range (≥350<!--> <!-->℃). It provides reference data for the operation of non-isothermal lead–bismuth system oxygen sensors.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112129"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035495","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":"Numerical study on the thermo-hydraulic behaviors of the Directionally-Alternated wire-wrapped fuel assembly in lead-cooled fast reactors based on SSTSAS k-ω-kθ-εθ four-equation model","authors":"Yunxiang Li ,&nbsp;Runsheng Yang ,&nbsp;Yuefeng Guo ,&nbsp;Xingkang Su ,&nbsp;Yuping Zhou ,&nbsp;Jian Hong ,&nbsp;Yuxing Liu ,&nbsp;Zinan Huang ,&nbsp;Xin Su ,&nbsp;Youpeng Zhang ,&nbsp;WenJun Hu ,&nbsp;Long Gu","doi":"10.1016/j.anucene.2026.112119","DOIUrl":"10.1016/j.anucene.2026.112119","url":null,"abstract":"<div><div>The study of the thermo-hydraulic behavior of liquid lead–bismuth in wire-wrapped fuel assemblies is of great significance for the safety design of the CiADS subcritical reactor. The four-equation turbulence model, which introduces both dynamic and thermal turbulence time scales to transport the turbulent Prandtl number, offers improved numerical accuracy in simulating the heat transfer of liquid lead–bismuth eutectic. To further enhance the heat-transfer performance of LBE within the fuel assembly, the arrangement and rotational direction of the spacer wires in the lead-cooled fast reactor fuel assembly were optimized, and a directionally-alternated wire-wrapped fuel assembly was proposed. In this work, a customized CFD solver named LBE4EqnFoam was developed based on the open-source platform OpenFOAM. Using the developed solver, detailed simulations were conducted for both the conventional wire-wrapped fuel assembly and the directionally-alternated wire-wrapped fuel assembly. The results show that LBE4EqnFoam provides highly accurate predictions of LBE flow and heat transfer in complex geometries. The solver has been validated against experimental measurements, showing that the maximum relative error in predicting the coolant temperature is below 2%, while the maximum relative error in predicting the cladding surface temperature is below 3%. Compared with the conventional design, the maximum pressure-drop reduction achieved by the directionally-alternated wire-wrapped fuel assembly is 28.22%, and the reduction at the outlet is 25.56%, which helps decrease the required pump head and the measurement range of pressure sensors. The directionally-alternated configuration also enhances cross-mixing among subchannels, leading to a more uniform temperature field and smaller temperature gradients at the outlet. This improvement is beneficial for reducing thermal fatigue and creep risks in the structural components near the outlet region. Moreover, the directionally-alternated design achieves a global average Nusselt number that is 1.38 times that of the conventional configuration. Furthermore, the directionally-alternated wire-wrapped fuel assembly exhibits superior integrated heat-transfer performance, with the integrated thermal–hydraulic factor improved by 36.45% compared with the conventional configuration.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112119"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035572","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":"Techno-Economic optimization of sandstone uranium Mining: A Case study of uranium content per square meter","authors":"Jiabing Li ,&nbsp;Chuanfei Zhang ,&nbsp;Xiangxue Zhang ,&nbsp;Meifang Chen ,&nbsp;Mingtao Jia","doi":"10.1016/j.anucene.2026.112125","DOIUrl":"10.1016/j.anucene.2026.112125","url":null,"abstract":"<div><div>This study introduces an improved non-dominated sorting genetic algorithm II (INSGA-II) to optimize the boundary delineation of sandstone-type uranium deposits by determining the threshold of uranium content per square meter (UCPSM, kg/m²) for minable units. A multi-objective optimization model was developed to maximize both economic and resource benefits, which was solved using the INSGA-II. Key enhancements include: (1) population initialization via symmetric Latin hypercube design (SLHD); (2) adaptive mutation and crossover parameters. Applied to real data from a Chinese mining area, the model and algorithm demonstrated practical effectiveness. The Pareto solution set derived from the optimization enabled the determination of UCPSM thresholds, supporting a novel mining-area boundary definition method based on aggregating minable units and unlocking the resource and potential economic value of idle uranium deposits. This approach offers a new decision-making tool for sandstone-type uranium mining area design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112125"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035579","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":"Experimental investigation of heat transfer enhancement during quenching using carbon-based nanofluids","authors":"Halima H. Alfailakawi,&nbsp;Shikha A. Ebrahim,&nbsp;Nawaf F. Aljuwayhel","doi":"10.1016/j.anucene.2026.112174","DOIUrl":"10.1016/j.anucene.2026.112174","url":null,"abstract":"<div><div>Nanofluids offer superior thermal properties compared to conventional fluids, making them ideal for heat transfer applications such as cooling systems. This study experimentally investigates the effect of adding carbon-based nanoparticles, such as multi-wall carbon nanotubes (MWCNTs), nanodiamonds (NDs), and graphene nanoparticles (GNPs), on the heat transfer performance in distilled water using a quenching setup at a near-saturation pool and atmospheric pressure. Nanofluids were prepared using a two-step method at concentrations of 0.01%, 0.025%, and 0.05% wt. with Gum Arabic (GA) added as a surfactant to enhance stability. Stability was monitored through photographic analysis, while SEM imaging provided detailed nanoparticle characterization on bare and quenched surfaces.</div><div>Results demonstrated significant improvement in the minimum film boiling temperature (T_min), with nanodiamonds showing the highest enhancement: 29.90% at 0.01%, 23.38% at 0.025%, and 18.75% at 0.05%. GNPs followed, while MWCNTs showed the lowest improvements. The superior performance of NDs is attributed to their high thermal conductivity and the formation of a porous nanoparticle layer on the surface, which increases surface area and disrupts vapor layer formation more effectively. Comparison with existing literature confirms the enhanced performance and highlights the role of nanoparticle shape and concentration. This study provides valuable insight into the literature by investigating in-depth how the same material with various shapes can contribute to improving heat transfer efficiency.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112174"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146185043","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":"Unsupervised anomaly detection method for nuclear power plants with limited training samples","authors":"Yongkuo Liu ,&nbsp;Longfei Shan ,&nbsp;Jiarong Gao ,&nbsp;Xueying Huang ,&nbsp;Zhen Wang ,&nbsp;Guohua Wu","doi":"10.1016/j.anucene.2026.112195","DOIUrl":"10.1016/j.anucene.2026.112195","url":null,"abstract":"<div><div>Most existing anomaly detection methods are based on unsupervised learning, using normal samples as training data. However, due to limitations in data labeling, accumulation, and collection time, the normal data currently available are often insufficient to encompass the complete variation patterns of normal operating data. Consequently, when confronted with normal samples lying outside the training distribution, the model is prone to generating false positives. To address this issue, we propose a convolutional autoencoder method based on a spatiotemporal attention mechanism. The attention mechanism enables the model to learn the correlations among different features within the normal data, thereby enhancing its generalization and extrapolation capabilities. The proposed method reconstructs the estimated values of all features under normal conditions using the autoencoder, and employs the Euclidean distance as a residual metric to quantify the discrepancy between the input and the reconstructed output. This residual metric is then used to perform anomaly detection. To validate the performance of the proposed model, experiments were conducted using normal and fault data collected from the Fuqing pressurized water reactor simulator under 100% full-power conditions. The test results demonstrate that the proposed method achieves higher anomaly detection accuracy compared with five representative methods, namely DeepSVDD, Autoencoder, PCA, One-Class SVM, and Isolation Forest.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112195"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186636","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 macroscopic depletion method for pebble-bed HTR fuel-cycle analysis with AGREE","authors":"Sefa Bektaş ,&nbsp;Volkan Seker ,&nbsp;Andrew Ward ,&nbsp;Üner Çolak ,&nbsp;Thomas Downar","doi":"10.1016/j.anucene.2026.112166","DOIUrl":"10.1016/j.anucene.2026.112166","url":null,"abstract":"<div><div>Because of the low-carbon generation nature of nuclear energy and its reliability to provide base load electricity, there is a recognized need to consider nuclear reactors as a future source of energy. However, emerging technologies such as the next generation of advanced small modular reactors are being assessed for safety performance which includes the validation and verification of computer codes used to model and simulate reactor transient behavior during anticipated accident scenarios. Most of the current generation of reactor safety codes have been validated primarily for large light-water reactor systems and the unique features and physics of a small modular high-temperature gas-cooled pebble bed reactors can differ significantly from the LWR and require a different set of experimental facilities and a different range of validation data. The objective of this paper was to extend that validation of the AGREE HTR safety analysis code to fuel depletion using the HTR-200 benchmark problem. The depletion capability was developed for AGREE using a quasi-batchwise fuel loading method and applied to the once-through-then-out (OTTO) fuel pass to achieve an equilibrium condition. The depletion analysis is performed using a two-step macroscopic cross section approach for full-core depletion which was implemented in AGREE. The two-step method used both Monte Carlo and deterministic reactor physics methods in which the Serpent Monte Carlo code was generated region-wise cross sections for the AGREE deterministic full core depletion. The spatially homogenized and energy condensed macroscopic cross-sections accounted for the effects of both instantaneous and history variables to include fuel burnup. Validation was performed by comparing AGREE results with both the legacy HTR code VSOP results reported in the benchmark reference documentation and the full-core, temperature-dependent Monte Carlo Serpent simulation results.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112166"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074815","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}