Annals of Nuclear Energy最新文献

{"title":"Optimal design of marine nuclear power deaerator based on improved particle swarm optimization algorithm","authors":"Jiarui Zhao ,&nbsp;Yanjun Li ,&nbsp;Xu Chen ,&nbsp;Yuan Fu ,&nbsp;Baozhi Sun ,&nbsp;Yuanwei Cao ,&nbsp;Jianxin Shi","doi":"10.1016/j.anucene.2025.111332","DOIUrl":"10.1016/j.anucene.2025.111332","url":null,"abstract":"<div><div>Deaerator is a key equipment in the secondary circuit system, its operating parameters and structural size have a significant impact on the thermal efficiency and arrangement rationality of the system. In this research, a mathematical model of the marine nuclear power deaerator is established, and the influence of thermal and structural parameters on the weight and volume of the deaerator is analyzed. An improved particle swarm optimization algorithm is proposed by introducing Tent chaotic mapping, evolutionary factor and Metropolis criterion, and its performance is verified. Taking the weight and volume minimization of the deaerator as the optimization objectives, the optimal design of the deaerator is carried out using the proposed IPSO algorithm, while satisfying the structural and performance constraints. The optimization results show that the volume and weight of the deaerator can be reduced by 12.979% and 10.213%, respectively, and the feasibility of the optimization design method is proved theoretically.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111332"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579630","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":"Preliminary studies on diversion and misuse for TRISO-fueled heat-pipe-cooled microreactors","authors":"Nelson Snow ,&nbsp;Sweet Shanahan ,&nbsp;Quinton Williams ,&nbsp;Ishita Trivedi ,&nbsp;Emerald Ryan ,&nbsp;Stefano Terlizzi ,&nbsp;Camille Palmer ,&nbsp;Ryan Stewart","doi":"10.1016/j.anucene.2025.111309","DOIUrl":"10.1016/j.anucene.2025.111309","url":null,"abstract":"<div><div>Microreactors are being developed by multiple reactor designers to mass-produce nuclear facilities for worldwide deployment. Including microreactors in a country’s energy portfolio introduces new concerns for international safeguards regarding the number of reactor locations. Microreactors have smaller quantities of nuclear material compared to current light-water reactors, which makes diverting enough material for clandestine purposes difficult for a single reactor. Diversion or misuse of multiple microreactors to obtain a significant quantity introduces a new acquisition pathway. This work examines diversion and misuse scenarios for a realistic heat-pipe-cooled microreactor to understand how these scenarios would effect reactor operations to determine if monitoring key parameters could reduce the burden of inspection on monitoring agencies. We determined misuse caused too drastic of an effect on core operations to be a valid acquisition pathway for the realistic heat-pipe-cooled microreactor. Diversion was explored across varying levels of severity, where potential diversion scenarios could yield one significant quantity of material from between six and eleven microreactors. Through examining the critical control drum angle, excess reactivity, control drum worth, and power distribution changes from nominal were detected over the operational lifetime, which could indicate a divergence from normal operations.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111309"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579676","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Development of a non-intrusive ROM for 5 × 5 rod bundles of PWR using small sample data","authors":"Hao Qian ,&nbsp;Guangliang Chen ,&nbsp;Dong Liu ,&nbsp;Yang Yu ,&nbsp;Yuhang Zhang ,&nbsp;Lixuan Zhang ,&nbsp;Jinchao Li","doi":"10.1016/j.anucene.2025.111347","DOIUrl":"10.1016/j.anucene.2025.111347","url":null,"abstract":"<div><div>This study aims to develop an efficient Reduced-Order Model (ROM) for rod-type fuel assemblies in pressurized water reactors (PWR), essential components of typical PWR cores. This paper integrates Proper Orthogonal Decomposition (POD) with machine learning (ML) for the rapid prediction of the thermal–hydraulic (TH) parameter field in the reactor core, aiding in real-time monitoring and optimization. This paper focuses on validating the model’s feasibility and accuracy for PWR core components operating at 15.5 MPa pressure. To verify the applicability of the POD method in predicting small sample data under actual operational conditions, this study uses actual operational boundaries from nuclear power plants and 121 sets of small sample data as boundary conditions. Extensive Computational Fluid Dynamics (CFD) calculations are conducted to collect data for constructing a snapshot matrix and further developing the ROM. A data-driven approach is employed to explore the relationship between POD coefficients and boundary conditions, achieving rapid prediction of the TH parameter field under any combination of boundary conditions. The study compares the predictive performance of four machine learning algorithms: Radial Basis Function Neural Networks (RBFNNs), K-Nearest Neighbors Regression (KNN), Random Forest Regression (RF), and Gradient Boosting Regression (GBR). Results show that all ML algorithms can construct ROM and predict the TH parameter field with reasonable accuracy for test data. Among them, Random Forest Regression achieves the highest precision, with an average Mean Square Error (MSE) of approximately 3.24 × 10⁻⁴ and a coefficient of determination (R²) of around 0.96. Furthermore, the average prediction time for the detailed three-dimensional parameter field per component is only 15.0 s.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111347"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579678","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":"Parametric investigation of mercury cell electrolysis for isotopic separation of Li using coupled numerical methods","authors":"J. Ahmadi ,&nbsp;M. Sharifi ,&nbsp;M. Aghaie ,&nbsp;S. Yousefi-Nasab ,&nbsp;A. Arzhangmehr","doi":"10.1016/j.anucene.2025.111336","DOIUrl":"10.1016/j.anucene.2025.111336","url":null,"abstract":"<div><div>Fusion reactors will be one of the major resources of future energy. The fuel required for these reactors is deuterium–tritium (D-T). Tritium, which is not naturally available, is produced by the collision of neutrons with ⁶Li in the environment of these reactors. Various compounds, including lithium hydroxide (LiOH) solution, can be used to extract ⁶Li. The natural abundance of ⁶Li is approximately 7.5%, which is insufficient for producing tritium used in fusion reactors and necessitates further enrichment. Various methods, including electrolysis, can be employed to increase ⁶Li enrichment. In this research, the electrolysis exchange (ELEX) method has been used to extract ⁶Li from LiOH with a favorable separation factor. The simulation of this process using the design of a mercury cell has been conducted through Comsol 6.1 software. This geometry (mercury cell) is a sloping space for continuous movement of mercury. Mercury, which is actually the cathode in this system, absorbs lithium isotopes, especially ⁶Li, due to the mobility of this isotope compared to ⁷Li, and increases the percentage of ⁶Li abundance at the output of the system (LiHg). Electrochemical physics, involving the solution of Nernst-Planck equations, and turbulent flow physics, governed by the Navier-Stokes equations, have been employed to carry out the electrolysis process. It should be noted that, given the significance of diffusion coefficient values in the process of isotope separation by electrolysis, these values have also been determined through molecular dynamics simulations. Finally, and based on the parametric study that has been conducted on all the parameters affecting the electrolysis process, a separation coefficient of about 1.05 has been obtained. This separation coefficient, which has also been compared with experimental references, has been stabilized after approximately 3 h of operation.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111336"},"PeriodicalIF":1.9,"publicationDate":"2025-03-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579680","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":"Mathematical modelling of bubble formation at top-submerged nozzles placed inside a quiescent liquid","authors":"Soumya Sarkar ,&nbsp;Nirvik Sen ,&nbsp;K.K. Singh","doi":"10.1016/j.anucene.2025.111281","DOIUrl":"10.1016/j.anucene.2025.111281","url":null,"abstract":"<div><div>Bubble size is an important variable that determines the efficiency of air induced liquid mixing, which, due to its passive nature, is commonly used in processes involving radioactive species. In majority of cases, in air induced mixing devices used in radiochemical plants, air is introduced through top-submerged nozzles. This necessitates fundamental understanding of the phenomenon of bubble formation at top-submerged nozzles placed inside a liquid pool. In this study, a mathematical model is developed by carrying out force balance to estimate the diameter of the bubbles generated at the tip of a top-submerged nozzle submerged in a quiescent liquid. The model is validated with the experimental data. The parametric analysis carried out with the developed model provides detailed understanding of the effects of geometric, operating parameters and properties of the liquid medium on bubble formation phenomenon.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111281"},"PeriodicalIF":1.9,"publicationDate":"2025-03-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143579629","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":"Global sensitivity analysis using Gaussian process surrogate models & impact of thermophysical properties uncertainties on the steady-state operation of molten salt reactors","authors":"Fadel M. Nasr ,&nbsp;Mauricio Tano ,&nbsp;Piyush Sabharwall ,&nbsp;Yousry Azmy","doi":"10.1016/j.anucene.2025.111304","DOIUrl":"10.1016/j.anucene.2025.111304","url":null,"abstract":"<div><div>Measurement uncertainties in the thermophysical properties of molten salts affect the predicted steady-state operation of molten salt reactors (MSRs). We aim to quantify that impact using three MOOSE-based multiphysics models of MSRs: the molten-salt fast reactor (MSFR), the molten-salt reactor experiment (MSRE), and the molten-chloride reactor experiment (MCRE). For each model we perturb the fuel salt’s thermophysical properties using Gaussian and uniform distributions, then we analyze the impact on the statistical distributions of various quantities of interest (QoIs) that are representative of the steady-state operation of MSRs. These QoIs can be grouped into two categories: those related to energy and those related to flow. Our results demonstrate that while the resulting uncertainties depend on the considered QoI, they are predominantly influenced by density uncertainty. Specifically, for the MSFR, heat capacity and density have the greatest impact on uncertainty. In the MSRE, the key contributors are heat capacity and dynamic viscosity, while in the MCRE, thermal conductivity, dynamic viscosity and density are the most significant sources of uncertainty. In general, for energy-related QoIs in the MSFR and MSRE cases, the observed QoI distributions were Gaussian regardless of the distribution chosen for the input thermophysical properties, i.e., uniform or Gaussian. For the flow-related QoIs, the distribution of the QoI is dependent on the prior distribution. However, interestingly, multimodal distributions are observed for the energy-related QoIs of MCRE, suggesting a possible bifurcation behavior in the multiphysics model. Refining the mesh for the MCRE model causes the multimodal distributions to disappear suggesting that the bifurcation is spurious, nonphysical. This article establishes a more general framework for uncertainty quantification in MSRs that accounts for the nonlinearity of the underlying models.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111304"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563278","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":"A higher order implicit stair-tailored scheme for the modified Burgers’ equation","authors":"V.P. Shyaman,&nbsp;A. Sreelakshmi,&nbsp;Ashish Awasthi","doi":"10.1016/j.anucene.2025.111284","DOIUrl":"10.1016/j.anucene.2025.111284","url":null,"abstract":"<div><div>In the quest of constructing a user-friendly handy algorithm for the numerical solutions of the modified Burgers’ equation, an adaptive finite point method is carefully orchestrated with no compromise on precision. The proposed algorithm has been designed so as the localized behavior of the analytic solutions is innately inherited to the numerical solutions. The original equation first undergoes a linearization post which the algorithm revolves around an implicit 4-point stair-shaped framework. Within this structure, the solution of the linearized equation at each node in the advanced temporal level is expressed as the linear combination of the nodal solutions at the current and previous temporal levels. Apart from being conditionally stable, consistent, converging, and rapid the method above and beyond replicates the exact solutions on coarse meshes even when the kinematic viscosity close in to zero.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111284"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563279","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":"Simulation of the effect of partial and total blockage in a subchannel on flow features distribution and interactions in sodium-cooled 5 × 5 rod bundles","authors":"Ayodeji A. Ala ,&nbsp;Zhu Feng ,&nbsp;Liu Junyan ,&nbsp;John Njoroge ,&nbsp;Bin Ye","doi":"10.1016/j.anucene.2025.111327","DOIUrl":"10.1016/j.anucene.2025.111327","url":null,"abstract":"<div><div>Subchannel flow obstruction accident is one of the safety considerations for sodium-cooled fast reactors. A computational fluid dynamics simulation was performed to study the flow of liquid sodium in a 5 × 5 fuel assembly with partial and entire blockage of a sub-channel. A situation with no blockage was used as a baseline for further investigation into following cases with blockages. The effect of a 75 % and total obstruction in a subchannel on the flow characteristics and heat transfer in the fuel assembly was studied from −5 hydraulic diameters before the blockage to 50 hydraulic diameters after the blockage and compared. The temperature surge of 10 K due to flow through the spacer in the model without blockage increased to 14 K and 24 K in the models with partial obstruction and total blockage, respectively. The blockage-induced resistance coefficient, turbulence kinetic energy distribution, and temperature fluctuation in the adjacent subchannels can be monitored to detect the blockage.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111327"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563276","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":"Coupled reactor physics and thermal-hydraulics analysis of an SCW-SMR reactor concept – Part II: Improvement of the concept and sensitivity analyses","authors":"Tamás Varju,&nbsp;Csenge Antók,&nbsp;Péter Mészáros,&nbsp;Boglárka Babcsány","doi":"10.1016/j.anucene.2025.111221","DOIUrl":"10.1016/j.anucene.2025.111221","url":null,"abstract":"<div><div>In the frame of the EU ECC-SMART project, the pre-conceptual design of a water-cooled small modular reactor operating at supercritical pressure (SCW-SMR) with seven heat-up stages is under development and assessment. The first part of the two-part paper presented the concept design, the models developed at BME in the Apros thermal-hydraulics system code and in the Serpent 2 Monte Carlo reactor physics code, the developed coupling methodology of these models, and the results of the first coupled calculations. This second part of the paper outlines the approaches adopted by the BME research team to improve the original design, primarily to reduce the high cladding temperatures. The effects of the increased mass flow rate and the introduction of inhomogeneous inlet orifices are thoroughly investigated and presented. Finally, a comprehensive sensitivity analysis has also been performed to demonstrate the robustness of the proposed design.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111221"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143696884","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":"Cross-Verification and validation of steady-state OPAL research reactor neutronic models using OpenMC and MCNP5 Codes: A foundation for advanced coupled simulations","authors":"Djahid Lababsa ,&nbsp;Hakim Mazrou ,&nbsp;Djalal Hamed ,&nbsp;Abdelkader Aouaichia ,&nbsp;Hamza Hasnaoui","doi":"10.1016/j.anucene.2025.111324","DOIUrl":"10.1016/j.anucene.2025.111324","url":null,"abstract":"<div><div>This study develops, verifies, and validates high-fidelity steady-state neutronic models of the OPAL research reactor using OpenMC and MCNP5. The models were developed based on a consistent methodology, utilizing identical input data, geometric modeling, and cross-section libraries, minimizing discrepancies from external factors, ensuring a solid foundation for subsequent code-to-code comparisons.</div><div>Verification and validation were conducted through comparisons with results from SERPENT and MCNP6, as well as experimental data. Our results showed strong agreement with the experimental data, including control rod worth calculations within 3% and axial peak factor differences averaging 2% for MCNP5 and 3% for OpenMC. Differences in isothermal temperature and void coefficients were approximately 1 pcm/°C and 7 pcm/%, respectively. Deviations from SERPENT and MCNP6 for few key parameters were within acceptable limits, consistent with reference studies, and attributed to variations in modeling, code differences, and Monte Carlo statistical uncertainties.</div><div>Finally, a comprehensive code-to-code comparison of key neutronic parameters, such as reactivity, flux distributions, kinetic parameters, and power distribution, showed excellent agreement between OpenMC and MCNP5, with relative differences under 1% for most parameters.</div><div>Overall, this study confirms the reliability of the developed models for simulating the OPAL reactor and establishes OpenMC as a viable alternative to MCNP for reactor physics simulations, laying the foundation for future coupling with thermal–hydraulic models.</div></div>","PeriodicalId":8006,"journal":{"name":"Annals of Nuclear Energy","volume":"217 ","pages":"Article 111324"},"PeriodicalIF":1.9,"publicationDate":"2025-03-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143563277","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}