Nuclear Engineering and Design最新文献

{"title":"Improving the accuracy of soil-structure interaction analysis through the generalized subtraction method","authors":"Hyeok-Ju Lee ,&nbsp;Oh-Sung Kwon ,&nbsp;Jae-Min Kim","doi":"10.1016/j.nucengdes.2025.114174","DOIUrl":"10.1016/j.nucengdes.2025.114174","url":null,"abstract":"<div><div>This study proposes the Generalized Subtraction Method (GSM) to improve the accuracy of Soil-Structure Interaction (SSI) analysis, which is crucial for seismic design of large structures such as nuclear power plants. Although the existing Subtraction Method (SM) used in the SASSI program is advantageous in terms of computational efficiency, the method has limitations that can cause abnormal responses in high-frequency regions. To address this issue, this study introduces a method of defining additional interaction nodes in the excavated soil to shift the spurious fundamental natural frequency of the excavated soil with fixed boundary conditions at the interaction nodes to above the maximum frequency of interest. By adjusting the fundamental natural frequency through iterative eigenvalue analysis, the proposed method provides stable and accurate SSI analysis results even in high-frequency regions. Numerical analysis results for two example models showed that the GSM achieved a similar level of accuracy to the Direct Method (DM) while using fewer interaction nodes than the existing Modified Subtraction Method (MSM).</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114174"},"PeriodicalIF":1.9,"publicationDate":"2025-06-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270735","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 properties of glass fiber filter material for positive zeta potential of coolant in nuclear power plant","authors":"Tao Liu,&nbsp;Haixia Kong,&nbsp;Yongguo Li,&nbsp;Jia Wang,&nbsp;Mang Wang,&nbsp;Xin Chen,&nbsp;Xu Shi,&nbsp;Shuwei Liang,&nbsp;Yingzhong Shuang,&nbsp;Zhixin Liu","doi":"10.1016/j.nucengdes.2025.114203","DOIUrl":"10.1016/j.nucengdes.2025.114203","url":null,"abstract":"<div><div>The microscopic particles present in the primary coolant of a nuclear power plant, when activated, have radioactivity that can affect the safety of the plant’s operation and the radiation dose of the workers. Traditional filtration materials only use mechanical interception to achieve the removal of small particles, which causes the initial filtration resistance to increase continuously, reducing the service life of the filter material. The element leaching rate is also an important indicator of the performance of the filter material. Since the filter material itself leaches Si⁴⁺, it will combine with Ca²⁺, Mg²⁺, and Al³⁺ ions in the coolant to form negative temperature coefficient silicates, which can form dense deposits at the hottest part of the fuel element bar, affecting the reactivity and fuel performance of the core. The leached SO₄²⁻ can cause intergranular corrosion on pipe materials, which has a significant impact on the service life of equipment.</div><div>In this paper, nano-scale Al powder (50 nm) was used as raw material to successfully prepare AlOOH crystal by one-step hydrothermal method, which was grafted on the surface of glass fiber to prepare glass fiber filter material with high positive Zeta potential, and the method combining mechanical filtration and electrostatic adsorption was realized to remove small particles. At the same time, the leaching of Si⁴⁺ and SO₄²⁻ was decreased. The experimental results showed that the filtration effect of the modified filter material for 0.45 μm particle size increased from 86.83 % to 99.95 %, and the 14-day Si⁴⁺ leaching concentration decreased by 73.17 % and the 14-day SO₄²⁻ leaching concentration decreased by 84.43 %.</div><div>This research helps advance the development of advanced materials that can withstand the harsh environment of nuclear energy systems, providing a reliable solution to ensure coolant purity and the safety of the entire reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114203"},"PeriodicalIF":1.9,"publicationDate":"2025-06-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270736","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":"Investigation of external heat transfer characteristics in a partially submerged vertical heat exchanger tube bundle","authors":"Hui Liang ,&nbsp;Zongwen Hu ,&nbsp;Qiwei Tian ,&nbsp;Chuan Lu ,&nbsp;Hongxing Yu ,&nbsp;Xiaxin Cao ,&nbsp;Ming Ding ,&nbsp;Zhongning Sun","doi":"10.1016/j.nucengdes.2025.114234","DOIUrl":"10.1016/j.nucengdes.2025.114234","url":null,"abstract":"<div><div>A passive residual heat removal heat exchanger (PRHR HX) is commonly employed in nuclear reactors to dissipate decay heat from the reactor core during design extension conditions. In the later stages of such accidents, the coolant level in a storage tank may drop, leading to partial exposure of the heat transfer tubes. To investigate the effect of reduced coolant levels on shell-side heat transfer performance in a vertical tube bundle, an experimental setup was designed, featuring four coolant levels: full, 4/5, 2/3, and 1/2. Saturated steam was introduced to the inner side of the tubes, while distilled water served as the coolant on the outer shell side. High-speed imaging and precise thermal measurements enabled the observation and analysis of three distinct heat transfer regions: nucleate pool boiling, liquid film evaporation, and steam convection. A new method was proposed to calculate heat transfer coefficients, accounting for the complex phenomena in the liquid film evaporation region at low coolant levels. The results indicate that shell-side heat transfer performance does not degrade linearly with decreasing coolant levels. At the 4/5 and 2/3 coolant levels, intensified thermal disturbances from water jet impacts and liquid film falling improved heat transfer. Even at the 1/2 coolant level, the heat removal capacity remained over 75 % of that at full submergence, due to effective phase-change heat transfer in the liquid film region. This study concludes that PRHR HX systems can maintain effective heat removal even at reduced coolant levels, driven by film evaporation and droplet impingement mechanisms. These findings offer valuable insights for the thermal design and safety analysis of passive cooling systems in advanced nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114234"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270711","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 of operational stability on reactor coolant pump with staggered guide vane under coast-down condition","authors":"Daoxing Ye ,&nbsp;Rui Deng ,&nbsp;Junlin Chen ,&nbsp;Junyu Zhang ,&nbsp;Jiaming Han ,&nbsp;Mengke Wu ,&nbsp;Binhao Liu","doi":"10.1016/j.nucengdes.2025.114227","DOIUrl":"10.1016/j.nucengdes.2025.114227","url":null,"abstract":"<div><div>To explore the impact of staggered guide vane on operational stability of reactor coolant pump during coast-down condition, numerical methods were employed to study the reactor coolant pump equipped with staggered guide vane. Research shows that during the entire coast-down process, staggered guide vane can significantly reduce the pressure pulsation of impeller at impeller rotational frequency and blade passing frequency; decreasing the amplitude of pressure pulsation in volute and guide vanes, making the intensity distribution of pressure pulsation more uniform. However, in the later stages of coast-down, due to the deterioration in the match between impeller and vane, these improvements diminish. Staggered guide vane can make the radial force distribution more uniform and reduce the radial force amplitude at the impeller rotational frequency, but this improvement decreases from 14.2% to 7.2% as coast-down progresses.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114227"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270734","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":"Probabilistic fracture mechanics analysis of a HTGR reactor pressure vessel under several typical transient conditions","authors":"Bowen Li,&nbsp;Haitao Wang,&nbsp;Heng Peng","doi":"10.1016/j.nucengdes.2025.114219","DOIUrl":"10.1016/j.nucengdes.2025.114219","url":null,"abstract":"<div><div>Probabilistic fracture mechanics (PFM) has been increasingly used in the structural integrity evaluation of reactor pressure vessels (RPVs) in light water reactors (LWRs). For high temperature gas-cooled reactor (HTGR), the working load of its RPV is different from that of LWR in terms of temperature, pressure, transient and fast neutron fluence. In addition, there are differences in the safety requirements associated with RPV. In this paper, PFM analysis of RPV of a 290MWth pebble-bed modular HTGR under several typical transient conditions is carried out. To simulate the manufacturing conditions of RPVs under different ASME rules, it is assumed that the flaw information of RPV along the wall thickness has different levels of manufacturing quality. In addition, the contribution of different types and regions of flaws in RPV to the conditional probability of initiation (CPI) and the conditional probability of failure (CPF) is investigated. Numerical results indicate that due to the low level of fast neutron fluence and slow transient development, the CPF of the RPV of HTGR is extremely low even under conservative assumptions of flaw size.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114219"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270712","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":"3D numerical investigation of fragmentation of wood’s metal in water and estimation of particle size distribution","authors":"M. Chandra Kumar ,&nbsp;A. Jasmin Sudha","doi":"10.1016/j.nucengdes.2025.114223","DOIUrl":"10.1016/j.nucengdes.2025.114223","url":null,"abstract":"<div><div>During a severe accident in a nuclear reactor, the molten nuclear fuel may interact with the liquid coolant in the form of jet in the reactor vessel. In this work hydrodynamic fragmentation of the Wood’s metal which is a simulant material for corium in water is studied. The 3D numerical model is developed in the open-source DNS solver Basilisk code and validated with experimental results found in literature for low temperature Wood’s metal water system with a discrepancy of about 2.3%. Volume of Fluid (VOF) technique is adopted to capture the liquid–liquid interface and Adaptive Mesh Refinement (AMR), to reduce the computational time. From the sensitivity study on mesh refinement levels, refinement level 12 is chosen, striking a balance between accuracy and computational time. The mass and size distributions of the fragments along the lateral and vertical direction are also studied. A parametric study has been carried out by varying the jet initial velocities to study its influence on the break up and particle size distribution. Detached mass fraction is estimated and particle size distribution is obtained for all the cases and a correlation has been developed between the MMD of the fragments and Weber number.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114223"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270710","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 simplified quantification method for seismic risk assessment of nuclear fuel cycle facilities using Clark approximation","authors":"Kotaro Kubo ,&nbsp;Kenji Mori ,&nbsp;Ken Muramatsu","doi":"10.1016/j.nucengdes.2025.114176","DOIUrl":"10.1016/j.nucengdes.2025.114176","url":null,"abstract":"<div><div>Nuclear fuel cycle facilities are important elements for supporting the efficient use of energy resources by establishing a nuclear fuel cycle. However, given that the risks of these facilities are lower than those of nuclear power plants, it is considered reasonable to apply simplified assessment methods when evaluating seismic risks. In this study, a simplified quantification method is proposed for seismic risk assessment at such facilities. Traditional simplified methods have streamlined the assessment process by selecting only representative components, often neglecting others. In contrast, the proposed method simplifies the required computational processes while considering all components by applying Clark approximation. Clark approximation is a mathematical method for approximating the maximum of two normal distributions as a new normal distribution. The proposed method was validated by comparing its seismic probabilistic risk assessment with those performed using Monte Carlo simulations and traditional simplified methods. Results showed that although the proposed method overestimated the high confidence of low probability of failure by a relative difference of 0.15 compared with that of the Monte Carlo method under completely independent condition, the overall plant-level fragility curve was generally within the range of the 5% and 95% confidence fragility curves. The proposed method accounted for the impact of correlated failure, which is critical in seismic risk assessments. Thus, this method enabled the seismic risk assessment of nuclear fuel cycle facilities in a simplified manner without compromising accuracy, potentially contributing to examining risk mitigation measures and developing risk-informed safety regulations for these facilities.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114176"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144262558","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":"Evaluation of Fluid-Structure interaction in multiple support cylinders of a liquid metal fast reactor","authors":"Yu Liu ,&nbsp;Feifan Zhang ,&nbsp;Qiang Liu ,&nbsp;Dexuan Duan ,&nbsp;Chaofan Zhang ,&nbsp;Fei Xie ,&nbsp;Fei Zhao ,&nbsp;Daogang Lu ,&nbsp;Yixian Zhou","doi":"10.1016/j.nucengdes.2025.114220","DOIUrl":"10.1016/j.nucengdes.2025.114220","url":null,"abstract":"<div><div>The support cylinder in a Liquid Metal Fast Reactor (LMFR) is a critical structural component essential for ensuring reactor stability and safety. Under extreme conditions, such as earthquakes, the liquid metal coolant significantly affects the added mass on the support cylinder, changing its modal frequency and vibration response. Previous research on the fluid–structure interaction of fast reactors mainly focused on the sloshing effect in rigid vessels. However, there is a gap in existing models for partially submerged, coupled cylinder systems in LMFRs that incorporate three-dimensional effects. This paper introduces a simplified fluid–structure interaction (FSI) mathematical model for a multi-cylinder system, based on potential flow theory, and employs a generalized single-degree-of-freedom (SDOF) system to simulate the dynamic characteristics of the support cylinder. By comparing experimental and numerical results, the model’s accuracy is validated with an error within 5% across multiple configurations. The paper examines the effects of water level changes, cylinder spacing, height-to-diameter ratio, and shape function choice on the added mass calculation. A three-dimensional correction formula for finite-length cylinders is proposed to improve calculation accuracy. The results show that the added mass coefficient increases with higher water levels and smaller cylinder spacing, while the height-diameter ratio has a nonlinear effect on added mass. This work provides a theoretical analysis and experimental verification of the FSI in LMFR support cylinders, which is important for the reactor’s seismic design and safety evaluation.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114220"},"PeriodicalIF":1.9,"publicationDate":"2025-06-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144270733","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 novel neutron-gamma spectrum-based composite material optimization process for the shielding design of a space reactor","authors":"Izat Khan,&nbsp;Liangzhi Cao,&nbsp;Sohail Ahmad Raza,&nbsp;Muhammad Kamran Butt","doi":"10.1016/j.nucengdes.2025.114215","DOIUrl":"10.1016/j.nucengdes.2025.114215","url":null,"abstract":"<div><div>In this work, a <strong>N</strong>eutron and <strong>G</strong>amma <strong>S</strong>pectrum-based <strong>C</strong>omposite-gradient-shielding-material <strong>O</strong>ptimization <strong>P</strong>rocess (NG-SCOP) has been developed to optimize the shadow shield design of a typical unmanned lithium-cooled small space reactor. This process has been accomplished by coupling an intelligently designed computational algorithm with NECP-MCX through a MATLAB program in an automated manner. The NG-SCOP algorithm designs an optimized shadow shield by adjusting its materials composition based on the relative contributions of neutrons and gamma dose rates in small incremental thicknesses along the entire shield length. Additionally, at each step, this algorithm homogenizes the optimized material composition from all preceding mesh cells up to that point and designs a single homogeneous composite material for the shadow shield structure. This process is repeated across all mesh cells until the fast neutron flux and silicon gamma dose limits for unmanned space reactors are met. The NG-SCOP technique is initially applied to four composite materials, C₂H₄-B₄C-W, C₂H₄-B₄C-Pb, LiH-B₄C-W, and LiH-B₄C-Pb by incorporating them in the shadow shield structure and calculating the thicknesses and masses required to attenuate neutron and gamma fluxes and doses to the required limits. The mass and thickness of LiH-B₄C-W has been further optimized by replacing tungsten, the dense material used for gamma attenuation, with a mixture of different weight fractions of W, B₄C, and LiH. The minimum masses obtained for the cell-wise optimized gradient composite and the homogeneous composite shields for LiH-B₄C-W are 628 kg and 592 kg, respectively. The results show that the NG-SCOP algorithm can design a more compact and lightweight shadow shield and can be adopted to design optimized shielding for any nuclear facility, including space and other portable nuclear reactors, using any combination of constituent materials and adjustments against any shielding design criteria.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114215"},"PeriodicalIF":1.9,"publicationDate":"2025-06-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144262556","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":"Comprehensive investigation of the semi-resolved and resolved LSMPS-DEM for two-dimensional solid particle settling behavior","authors":"Xiaoqiang Guo,&nbsp;Xiaoxing Liu","doi":"10.1016/j.nucengdes.2025.114237","DOIUrl":"10.1016/j.nucengdes.2025.114237","url":null,"abstract":"<div><div>In the late stages of severe accidents in sodium-cooled fast reactors, molten fuel falls into the liquid sodium coolant, forming a debris bed. The coolability of this debris bed directly influences whether the molten pool will undergo reheat, thereby threatening the integrity of the pressure vessel lower head, which is critical for safety. The coolability of the debris bed is closely related to its shape and size, which in turn depend on its formation behavior. Therefore, accurately simulating the formation process of the debris bed is of great importance. Due to its Lagrangian characteristics, the particle method demonstrates significant advantages in simulating the settling and accumulation processes of particle fragments in liquids. Existing studies have predominantly focused on the application of resolved models, while research on semi-resolved models remains largely confined to grid-based methods, and their applicability in the particle method has yet to be systematically explored. To address this research gap, this study conducts an in-depth analysis of the applicable ranges of resolved and semi-resolved models within the particle method. The results show that in the particle method, when the ratio of fluid particle spacing to solid particle diameter is less than or equal to 0.5, the resolved model should be used; when the ratio is equal to 1 or 2, the semi-resolved model is more suitable. These findings suggest that using the resolved model for large particles and the semi-resolved model for small-to-medium particles could enable more detailed and accurate simulations of debris bed formation behavior, providing a foundation for future investigations into mixed particle size conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"442 ","pages":"Article 114237"},"PeriodicalIF":1.9,"publicationDate":"2025-06-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144262557","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}