Annals of Nuclear Energy最新文献

{"title":"A non-negative Lasso Orthogonal Matching Pursuit method for gamma radiation field reconstruction with sparse measurement data","authors":"Longji Qiu,&nbsp;Nan Chao,&nbsp;Yong-kuo Liu,&nbsp;Zongzhen Shi,&nbsp;Joseph Daniel","doi":"10.1016/j.anucene.2026.112147","DOIUrl":"10.1016/j.anucene.2026.112147","url":null,"abstract":"<div><div>This paper proposes a gamma radiation field reconstruction method based on sparse detection data. By integrating compressed sensing (CS) theory with clustered transport theory, a sensing matrix for radiation field reconstruction is constructed. A novel Non-negative Lasso Orthogonal Matching Pursuit (NNLasso-OMP) algorithm is developed, combining the high efficiency and flexibility of OMP with the overfitting resistance of Lasso. To evaluate the proposed method, three simulation scenarios are conducted, using Monte Carlo simulation results as reference benchmarks. The reconstruction performance of NNLasso-OMP is compared with that of OMP, inverse distance weighting (IDW) and 3DCNN algorithms. Results show that the average relative error (ARE) of NNLasso-OMP remains below 10% across all scenarios, achieving a reconstruction success rate (SR) exceeding 95%, while accurately identifying source locations. The proposed NNLasso-OMP method outperforms both OMP and IDW, demonstrating its effectiveness in achieving high-quality gamma radiation field reconstruction from sparse measurements.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112147"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035571","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":"Study on Monte Carlo iterative K-source (IKS) method for frequency domain neutron noise calculation","authors":"Xuran Yang ,&nbsp;Qi Zheng ,&nbsp;Liangzhi Cao ,&nbsp;Qingming He ,&nbsp;Hongchun Wu","doi":"10.1016/j.anucene.2026.112150","DOIUrl":"10.1016/j.anucene.2026.112150","url":null,"abstract":"<div><div>The solution to the neutron noise equation is critical for on-line reactor monitoring and fault diagnosis via neutron noise analysis. This study applies the Monte Carlo Iterative K-Source (IKS) method to frequency domain neutron noise calculations, aiming to enhance computational efficiency and clarify its distinctions from the traditional fixed-source direct method. The theoretical derivation of the IKS method is performed and this method is implemented in the NECP-MCX code. The numerical validation is carried out based on a simplified UOX 2D fuel assembly and an adapted IAEA 3D PWR benchmark. The numerical results demonstrate that the IKS method achieves significant efficiency gains in the plateau frequency region, with a higher Figure of Merit (FOM) compared to the direct method—particularly in complex 3D geometries, where it exhibits more balanced relative standard deviation (RSD) distributions. The proposed convergence criteria, integrating Shannon entropy of real and imaginary parts and phase via stochastic oscillators, effectively judges iteration convergence, with phase emerging as a critical indicator for noise problems. However, the IKS method faces challenges at extreme frequencies where the <em>k</em>_s <!-->factor approaches 1.0: convergence slows markedly, requiring large particle numbers per cycle to suppress<!--> <em>k</em>_s <!-->fluctuations, and result accuracy becomes sensitive to<!--> <em>k</em>_s <!-->instability due to the<!--> <!-->1/(1- <em>k</em>_s)<!--> <!-->normalization. This work establishes the IKS method as a promising alternative for efficient neutron noise simulation, providing technical support for reactor monitoring and fault diagnosis applications.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"231 ","pages":"Article 112150"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146186639","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":"Influence of ferric oxide (Fe2O3) content on the mechanical strength and radiation attenuation capacity of concrete","authors":"Mohamed Hasabelnaby ,&nbsp;Mohammad Marashdeh ,&nbsp;K.A. Mahmoud ,&nbsp;Reham M. Abd El Rahman ,&nbsp;Hanan Akhdar ,&nbsp;Ghada Salaheldin ,&nbsp;Mohammad Y. Hanfi","doi":"10.1016/j.anucene.2025.112105","DOIUrl":"10.1016/j.anucene.2025.112105","url":null,"abstract":"<div><div>Radiation shielding materials are crucial for nuclear, industrial, and medical situations where shielding against ionizing radiation is a concern. In this study, the effect of ferric oxide (Fe₂O₃) incorporation on the mineralogical, physical, mechanical, and gamma-ray shielding properties of Portland cement-based concretes is studied. Concrete samples were made by replacing fine aggregate with Fe₂O₃ (0–40 wt%) in various amounts and were examined for mineralogy using XRD, elemental composition using XRF, density, porosity, and water absorption, compressive strength, elastic modulus, and compression through gamma-ray attenuation. The results showed that on average Fe₂O₃ incorporation led to higher concrete density (2.51–2.69 g/cm³), lower porosity of concrete, and a more than 38 % and 36 % reduction in water absorption, and improved gamma-shielding performance, with average increases in linear attenuation coefficient (LAC) of ∼7 % at energies of 0.511 and 0.662 MeV. Nevertheless, average compressive strength declined from 9.75 MPa (control) to 3.75 MPa (40 wt% Fe₂O₃) and the elastic modulus from 15.6 GPa to 9.7 GPa which are not strong load bearing results. Regression analysis produced predictive models (R² &gt; 0.95) relating Fe₂O₃ amount to density, porosity, and strength to allow for performance estimation for design. These results confirmed that Fe₂O₃ concretes, while not viable for structural load bearing, would still serve as effective non-structural shielding materials for medical and nuclear applications. Based on the data obtained from the Fe₂O₃ study, the upper limit of enhancement through Fe₂O₃ was assigned to CON30 to CON40, though these levels represent some of the highest attenuation values detected in relation to mechanical degradation. Therefore, the Fe₂O₃ doping group concretes do not uphold their suitability for load bearing applications but do give considerable merit for use as non-load bearing radiation shielding materials across medical, research, and nuclear facilities.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112105"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145924527","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":"Neutronic characteristics of a partially damaged reactor model with varying numbers of damaged fuel assemblies","authors":"Hoang Hai Nguyen","doi":"10.1016/j.anucene.2026.112171","DOIUrl":"10.1016/j.anucene.2026.112171","url":null,"abstract":"<div><div>This study examined the effects of the moderator-to-fuel volume ratio, fuel debris shape, and the number of damaged fuel assemblies on the neutronic features of a partially damaged reactor, where the fuel assemblies in the center region collapse into debris and the fuel assemblies in the outer region are kept unchanged. The partially damaged reactor model was based on the Watts Bar Nuclear 1 reactor. The investigations were performed by the Serpent code. The findings show that in cases where the debris is surrounded by intact fuel assemblies, the change of k_eff depends on the geometry of the debris. Conversely, in scenarios where the debris is not fully encircled by intact fuel assemblies, the geometry of the debris has a negligible impact on the k_eff. Additionally, the number of neutrons entering and leaving the debris determines how its shape affects the k_eff.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112171"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035493","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":"Calculation speed and precision of activation simulation using detailed and homogenized geometries","authors":"Norikazu Kinoshita ,&nbsp;Takuma Noto ,&nbsp;Kazuaki Kosako ,&nbsp;Yuki Sasaki ,&nbsp;Kazuyuki Torii ,&nbsp;Makoto Inagaki","doi":"10.1016/j.anucene.2026.112159","DOIUrl":"10.1016/j.anucene.2026.112159","url":null,"abstract":"<div><div>We investigated several specific activities produced at a location close to the reactor core of the research reactor of KUR, by comparing experimental results with the results of simulations using MCNP5 and ORIGEN codes. In the simulation using MCNP5, we applied a detailed geometry that precisely describes the interior of the nuclear reactor and simplified geometries that homogenized materials in the nuclear fuel assembly by maintaining the bulk density and bulk chemical composition. The calculation time increased with increasing number of cells in the geometry input. The specific activities were not affected by the homogenization of the geometry. We confirmed that nuclides produced from trace elements in concrete can be simulated with an accuracy of approximately 20% at locations relatively close to the nuclear core. In the case of KUR, the homogenization of the geometries can contribute to shortening the calculation time without degrading the precision.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112159"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035496","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 spatiotemporal EIM for long-term xenon dynamics prediction in commercial reactor cores","authors":"Ziyi Zheng ,&nbsp;Lichen Yuan ,&nbsp;Xianping Zhong ,&nbsp;Lianjie Wang ,&nbsp;Jiangyu Wang ,&nbsp;Deng Pan ,&nbsp;Jun Luo ,&nbsp;Helin Gong","doi":"10.1016/j.anucene.2026.112167","DOIUrl":"10.1016/j.anucene.2026.112167","url":null,"abstract":"<div><div>For the safe and stable operation of nuclear reactors, accurate prediction of xenon-induced power oscillations is crucial. However, existing prediction methods are constrained by limited sensor coverage, high computational complexity, and substantial algorithmic cost. This study proposes a space–time prediction framework based on the Empirical Interpolation Method (EIM). A large set of parameterized space–time simulations is first generated; unlike typical reduced-basis approaches that only process spatial snapshots, each full space–time solution is treated as one snapshot in a greedy EIM procedure. This yields interpolation bases and a set of space–time magic points, whose temporal locations are restricted to past instants with available measurements. During operation, observations at these magic points are assimilated through a low-dimensional linear system to reconstruct the space–time evolution of the physical fields and to predict future states. The proposed approach is assessed under various power conditions of the HPR1000 reactor core, where it demonstrates superior accuracy and efficiency compared with traditional methods, as well as robustness to measurement noise. Its stability in long-term predictions indicates strong potential as a practical tool for real-time monitoring and prediction of reactor core behavior.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112167"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035569","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":"Identification of the containment heating mechanism and temperature distribution by high-temperature gas leakage under severe accident conditions","authors":"Restu Kojo","doi":"10.1016/j.anucene.2025.112009","DOIUrl":"10.1016/j.anucene.2025.112009","url":null,"abstract":"<div><div>This study evaluates the thermal-hydraulics in the containment of boiling water reactors caused by superheated gas leakage under severe accident conditions. The research involved characterizing the heat transfer paths, which included identifying superheated gas leakage positions and heat release from the reactor pressure vessel boundaries, classifying the containment volumes, and categorizing potential failure sections of the containment. Based on these insights, the heat transfer paths during severe accidents were clarified, and accident scenarios considering leakage from the safety relief valve and traversing in-core probe tubes were selected as representative scenarios.</div><div>A three-dimensional computational fluid dynamics (CFD) model of the Mark I containment was developed to evaluate the thermal-hydraulics of an entire drywell of the containment. Special attention was given to modeling the detailed structures of the containment top head flange, radiation from the reactor pressure vessel upper head, condensation in the suppression pool, and heat release from the containment top head to the reactor well. The CFD analyses focused on two scenarios: safety relief valve leakage and traversing in-core probe tube leakage, which can result in significant temperature distribution in the upper and lower drywell, respectively. This study identified the high temperature location on the containment boundary with higher possibility of failure by high-temperature gas leakage under severe accident conditions based on temperature distribution obtained by the present detailed three-dimensional CFD analysis.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112009"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146035576","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":"Design and performance analysis of high-power lead-bismuth cooled micro reactor","authors":"Yiming Xiong,&nbsp;Ren Li,&nbsp;Jilin Sun,&nbsp;Yuandong Zhang,&nbsp;Genglei Xia,&nbsp;Minjun Peng","doi":"10.1016/j.anucene.2026.112115","DOIUrl":"10.1016/j.anucene.2026.112115","url":null,"abstract":"<div><div>With the wide range of sea applications of reactors, it is necessary to develop a high-power LBE reactor with a small volume and high-power output. A process of core design is proposed in this paper for conducting rapid iterations. With this process and a series of design criteria and guidelines, the concept of the High-Power Lead-Bismuth Cooled Micro Reactor (HLCMR) is designed. This design adopts low-enriched fuel and two sets of independent shut-down control systems, which are arranged in a triangular pattern to achieve higher power density and lower power peak. After iterative design, the design power is 12 MW, and the lifetime is at least 5 years with the core power density of 76.72 MW/m³. The power distribution shows that the highest power peak is 1.444 in the operating state. The temperature field and reactivity are also calculated to evaluate the safety and reliability of this design. The results show that all parameters of the HLCMR meet the thermal–hydraulic and control design requirements.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112115"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145915168","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":"Neutronic calculation route for molten salt fast reactors conception studies","authors":"D. Marjanovic,&nbsp;G. Valocchi,&nbsp;J. Tommasi","doi":"10.1016/j.anucene.2026.112170","DOIUrl":"10.1016/j.anucene.2026.112170","url":null,"abstract":"<div><div>This work investigates the validity and limitations of neutronic modelling routes commonly used in coupled simulations of molten salt fast reactors (MSFRs). This work is carried out as part of the ISAC project funded by the France2030 government program. A two-step deterministic baseline scheme is first developed using APOLLO3® under isothermal, static conditions and benchmarked against Monte Carlo references. The scheme’s accuracy is evaluated across an encompassing set of core configurations, Heavy, Accumulators, and Thermalizing Accumulators. This set aims to capture key neutronic features and assess whether modelling trade-offs are generalizable across diverse reactor behaviours. Once validated, the baseline serves to assess the impact of typical simplifications such as coarse energy group collapsing and diffusion approximations, usually employed in high fidelity multi-physics simulations. Results show that these reduced models can introduce large biases, particularly in configurations with strong spatial and spectral heterogeneities, such as thermalizing accumulators. The study then evaluates whether temperature-induced variations justify iterative neutronic updates or if simplified calculation routes can be used to capture trends and support low-cost rescaling. Key observables such as flux, absorption, and reactivity are examined. Findings support hybrid strategies that reuse accurate baseline results and adjust reactivity using sensitivity-based corrections, offering a promising balance between accuracy and computational cost for MSFR design studies.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112170"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074816","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 deep learning-based prognostic approach for predicting PWR degradation and remaining useful life using GNN-PTC-LSTM","authors":"Shadman Ahmad Khattak ,&nbsp;Liu Yong-Kuo ,&nbsp;Liu Yu-Kun ,&nbsp;Gao Jia-Rong ,&nbsp;Shi Zhou-Xin ,&nbsp;Liu Ji","doi":"10.1016/j.anucene.2026.112172","DOIUrl":"10.1016/j.anucene.2026.112172","url":null,"abstract":"<div><div>Early fault detection and prediction are key for ensuring the proper functioning and safety operation of complex nuclear systems especially in nuclear power production. Remaining Useful Life (RUL) prediction in Pressurized Water Reactors (PWRs) is one of the crucial parameters to prevent gradual degradation and disaster. In this paper a hybrid model composed of graph neural network,physics topology constraint and long short term memory network (GNN-PTC-LSTM) is proposed for fault prognostic and remaining useful life prediction based on the nuclear power plant using 2 loop PWR PCTRAN datasets. The proposed framework employs Graph Neural Networks (GNNs) to capture the spatial dependencies between reactor subsystems while integrating a PTC-LSTM module that incorporates physical topology and plant dynamics as temporal sequence learning constraints. In contrast to traditional LSTMs which are purely statistical model the physics-informed PTC-LSTM integrates a priori understanding of domain knowledge to mitigate non-physical predictions models outputs that violate known physical/topological constraints and reduce false alarms due to spurious data correlations. Additionally, One-Class SVM is used for anomaly detection in multivariate telemetry data to allow early discovery of abnormal behaviours. The proposed GNN-PTC-LSTM framework achieved an overall fault classification accuracy of 99.1%, an early warning accuracy of 98.2%, and competitive RUL prediction performance with a mean absolute error (MAE) of 0.0042 and root mean square error (RMSE) of 0.0105 under simulated PWR accident scenarios.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"230 ","pages":"Article 112172"},"PeriodicalIF":2.3,"publicationDate":"2026-06-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146074817","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}