Nuclear Engineering and Design最新文献

{"title":"Study on the base isolation design and parameter optimization analysis of friction pendulum bearings for reactor building in swimming pool-type low-temperature heating reactor","authors":"Yingying Gan ,&nbsp;Xiaoying Sun ,&nbsp;Pengxiang Dong ,&nbsp;Ziqiao Liu","doi":"10.1016/j.nucengdes.2025.114739","DOIUrl":"10.1016/j.nucengdes.2025.114739","url":null,"abstract":"<div><div>The swimming pool-type low-temperature heating reactor (SPLTHR) is a single-unit small heating reactor that can serve as an alternative to fossil energy. The peak ground acceleration (PGA) for the safe shut- down earthquake (SSE) at the proposed site is up to 0.5 g in horizontal direction. To ensure seismic safety and improve economic efficiency of the reactor, the Friction Pendulum (FP) bearing is employed for the base isolation design of the reactor building. Firstly, a three-dimensional finite element model (FEM) of the reactor building is established. The layout scheme of the base isolation layer is designed. Subsequently, a parameter optimization analysis about the equivalent radius of curvature and dynamic friction coefficient of the FP bearing is conducted to achieve the optimal isolation performance for the reactor building. Finally, the acceleration response spectrum (ARS) in three directions were compared between the base-isolated system and non- isolated system at the same place. The acceleration reduction rate was defined to quantified the isolation performance. The study results indicate that the base isolation layer using 28 FP bearings with load capacity in axial direction of 15,000 kN and 8 viscous damper can meet the design requirement. The dynamic friction coefficient of the FP bearing has a more significant influence on the isolation performance than the equivalent radius of curvature. In general, a larger equivalent radius of curvature and a smaller dynamic friction coefficient result in better isolation performance. The ARS in horizontal direction of the superstructure in the non-isolated system completely envelops that of the base-isolated system. The seismic response of the base-isolated system shows a substantial reduction in the dominant frequency and a significant decrease in the ARS of the superstructure in horizontal direction. The maximum reduction rates for zero-period acceleration (ZPA) and peak acceleration can reach up to 75.0 % and 85.4 %, respectively, demonstrating excellent isolation performance. Compared to the ARS in vertical direction of the non-isolated system, the base-isolated system has a lower dominant frequency, a leftward shift in peak acceleration response (with lower peak frequency), and an insignificant increase in peak values. It is recommended to focus on the seismic response of key equipment which is sensitive to the vertical frequency ranges of the base-isolated system, and to implement appropriate local vertical isolation measures if necessary.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114739"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146081255","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":"Modelling of MSRE graphite temperature in porous-medium multi-physics simulations","authors":"S. Amirkhosravi ,&nbsp;A. Scolaro ,&nbsp;F. van Niekerk ,&nbsp;M.H. du Toit ,&nbsp;A. Pautz","doi":"10.1016/j.nucengdes.2026.114757","DOIUrl":"10.1016/j.nucengdes.2026.114757","url":null,"abstract":"<div><div>MSRs stand out as prominent candidates among advanced reactor designs, addressing the global demand for safer and more sustainable nuclear energy. Accurate multi-physics modelling is essential for the advancement of MSR technology, particularly for understanding the thermos-hydraulic behaviour of graphite under irradiation. This study focuses on developing and implementing a high-fidelity methodology within the GeN-Foam code to model graphite temperature distribution within porous-medium multi-physics simulations, using the MSRE as a benchmark. The approach combines thermal-hydraulic and neutronic modelling by using Serpent-generated cross-section data as inputs for the <strong>Gen-Foam</strong> neutronic solver. Validation against MSRE measurements performed at ORNL benchmark data confirms the framework's reliability. The axial temperature distribution yields a Mean Absolute Percentage Error (MAPE) of 1.09%, while the radial distribution shows a MAPE of 0.62%. The average graphite temperature of 935.6 K is consistent with the ORNL reference value of 936.4 K under steady-state conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114757"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146078927","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 of nuclear fuel loading patterns for a PWR with Wasserstein generative adversarial networks","authors":"Anderson Alvarenga de Moura Meneses ,&nbsp;Lenilson Moreira Araujo","doi":"10.1016/j.nucengdes.2026.114765","DOIUrl":"10.1016/j.nucengdes.2026.114765","url":null,"abstract":"<div><div>The Loading Pattern (LP) design is part of the nuclear fuel management of a Nuclear Power Plant (NPP). The design of an LP includes the permutation of fuel assemblies, as well as calculations performed with reactor physics codes, aiming to producing energy with the satisfaction of constraints such as those related to safety. From a computational perspective, it is an NP-hard combinatorial problem solved with success by optimization metaheuristics. With the breakthrough of generative Artificial Intelligence (AI), the immediate question is whether LPs can be designed within such paradigm. In the present article, a methodology is proposed for training and applying Wasserstein Generative Adversarial Networks (WGANs) for automatic generation of LPs of a Pressurized Water Reactor. With the application of the methodology to the benchmark IAEA-2D, WGANs generated LPs satisfying the safety constraints and objectives proposed. Thus, WGANs can learn implicit probability distributions of nucler fuel and automatically design high-quality LPs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114765"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036031","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":"Optimizing iPWR SMR core design: a power peaking factor analysis of annular fuel rods using MCNP5","authors":"Fatima Ghandour ,&nbsp;Salah Hamieh ,&nbsp;Ziad Francis","doi":"10.1016/j.nucengdes.2026.114768","DOIUrl":"10.1016/j.nucengdes.2026.114768","url":null,"abstract":"<div><div>This study investigates the neutronic performance of dual cooled annular fuel rods in the CAREM 25 integral Pressurized Water Reactor (iPWR), a small modular reactor (SMR), using MCNP5 Monte Carlo simulations. The motivation is to reduce power peaking factors (PPFs) and enhance thermal-hydraulic safety margins by adopting an annular fuel geometry with internal and external cooling. Three annular fuel configurations with 100%, 95%, and 93% fuel loading were analyzed and compared to the conventional solid fuel design. Geometric transformations were performed analytically—introducing, for the first time, closed-form equations for the inner and outer radii of annular fuel rods—to maintain the fuel-to-coolant volume ratio while limiting fuel mass reduction to ≤10%. The results show that the total PPF decreased by up to 27.45% in the 95% fuel loading case, dropping from 2.404 (solid design) to 1.744. Additionally, the effective multiplication factor (Keff) was reduced from 1.12445 to 1.09512, enhancing reactor controllability. The 95% loading configuration emerged as the optimal design, balancing neutronic performance and safety. These findings demonstrate that annular fuel can significantly flatten the power distribution and improve the safety profile of iPWR SMRs without compromising core performance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114768"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036073","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":"Computational modeling of graphite degradation in molten salt reactors: Role of infiltration","authors":"Veerappan Prithivirajan ,&nbsp;Benjamin Spencer ,&nbsp;Joseph Bass ,&nbsp;Somayajulu L.N. Dhulipala ,&nbsp;Daniel Schwen ,&nbsp;Mustafa K. Jaradat","doi":"10.1016/j.nucengdes.2026.114796","DOIUrl":"10.1016/j.nucengdes.2026.114796","url":null,"abstract":"<div><div>Molten salt reactors (MSRs) often employ graphite as a moderator and reflector. An important challenge for deploying graphite in these reactors is that, due to limited experimental data, our understanding of graphite’s structural integrity in molten salt environments remains incomplete. This study addresses heat generation from fuel-bearing salt that has infiltrated open pores in the graphite, driven primarily by pressure differentials. This is one of multiple identified physical and chemical mechanisms through which molten salt could potentially degrade graphite. Thermally driven stresses are quantified using the Molten-Salt Reactor Experiment (MSRE) graphite moderator elements as a case study. Finite element simulations predict stress distributions at varying infiltration levels, indicating that thermal stresses increase with higher infiltration. Rare-event simulations using the parallel subset simulation framework identify the combinations and corresponding ranges of input parameters that lead to stresses above a specified threshold. In particular, combinations involving high infiltration amounts, high power density, and low thermal conductivity tend to induce the highest stresses. Under the inputs and assumptions considered in this work, the magnitudes of the thermally driven stresses are quite low, with a very low likelihood of causing failure due to exceeding the graphite’s tensile strength. Additionally, rare-event simulations were performed for two more scenarios: a scaled-up moderator geometry and a localized hotspot in the original geometry. Both cases resulted in increased susceptibility to failure, though not to a detrimental extent. Furthermore, the combined effects of irradiation and infiltration-induced thermal stresses were evaluated. The results showed that thermal stresses from infiltration were negligible compared to those caused by irradiation. The findings of such a study are inherently component-specific, but the methodology presented here could be used for similar assessments of salt-infiltration effects in other graphite components.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114796"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146057606","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":"Modeling considerations for passive safety systems in Korean SMRs using system analysis codes","authors":"Seong-Su Jeon,&nbsp;Jungjin Bang,&nbsp;Sang Gyun Nam,&nbsp;Jehee Lee,&nbsp;Youngjae Park,&nbsp;Soon-Joon Hong","doi":"10.1016/j.nucengdes.2025.114747","DOIUrl":"10.1016/j.nucengdes.2025.114747","url":null,"abstract":"<div><div>Numerous Small Modular Reactors (SMRs) are being developed worldwide, and they are equipped with various types of Passive Safety Systems (PSSs). In the Republic of Korea, SMART100 and i-SMR are representative SMRs. SMART100 includes the Passive Safety Injection System (PSIS), the Passive Residual Heat Removal System (PRHRS), and Containment Pressure and Radioactivity Suppression System (CPRSS) while i-SMR is equipped with the Passive Emergency Core Cooling System (PECCS), the Passive Containment Cooling System (PCCS), and the Passive Auxiliary Feedwater System (PAFS). These systems operate based on natural forces such as gravity and buoyancy, performing safety functions without external power or operator action. However, because their operation relies on relatively weak and time-varying driving forces, reliable modeling using system analysis codes is important. In particular, improper simulation of key thermal-hydraulic phenomena such as pressure drop, condensation, boiling, and natural circulation can lead to predictions that deviate considerably from actual performance. To address these concerns, this study reviews the modeling and simulation of PSIS, PAFS, and PCCS in Korean SMRs using various system analysis codes. Based on the authors' extensive experience, detailed modeling considerations are derived to improve the representation of key physical phenomena. Furthermore, the study discusses the importance of robustness evaluation under degraded conditions using the Best Estimate with Performance Issues (BEPI) framework. The insights provided herein are expected to support the credible and technically robust application of system analysis codes to the design and safety assessment of passive safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114747"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981786","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":"Clarifying the trends of hydraulic resistance parameters for supercritical pressure fluids with the aid of CFD","authors":"Md Radwan Ahmed ,&nbsp;Sara Kassem ,&nbsp;Andrea Pucciarelli ,&nbsp;Walter Ambrosini","doi":"10.1016/j.nucengdes.2026.114828","DOIUrl":"10.1016/j.nucengdes.2026.114828","url":null,"abstract":"<div><div>The paper addresses the observed trends of friction factor and shear stress at the wall as a function of operating conditions in supercritical pressure fluids. Previous literature suggests that the friction factor correlations applicable at supercritical pressures, at which fluids show considerable changes in properties while crossing the pseudocritical threshold, should be mostly the same as those for fluids with relatively constant properties in the liquid-like and gas-like regions, though corrections should be instead applied for conditions in between the two extremes. Basing on a previous work performed in past years at the University of Pisa to address this issue and after a review of selected information proposed in the field, some experimental conditions are here addressed with the aid of a CFD code, aiming to consider the behaviour predicted by the available models and to discuss the possible agreement of this information with experimental evidence. In particular, a circular pipe with supercritical carbon dioxide and a 2 × 2 rod-bundle mock-up with supercritical water are considered; in the latter case, the capability of CFD to provide local data of the shear stress at the heated and the unheated walls of the bundle helps in suggesting with sufficient clarity the different influence that bulk and wall temperatures have on friction parameters. The obtained results support the use of the classical correction factors proposed in previous literature to correct isothermal friction factor correlations with density and/or viscosity ratios to account for an observed decrease across the pseudocritical region. This information supports a meaningful explanation of some of the available experimental data that, at the moment, cannot be so detailed as to highlight such local effects, thus proposing a rationale to explain the role of the different parameters influencing hydraulic resistance at supercritical pressure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114828"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190250","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 preferable siliconized graphite with high graphite content in thrust bearings of shaft-sealed main coolant pump","authors":"S.H. Liu ,&nbsp;T.H. Liang ,&nbsp;B.J. Zhang ,&nbsp;J.Q. An ,&nbsp;L. Cai ,&nbsp;M.K. Lei","doi":"10.1016/j.nucengdes.2026.114831","DOIUrl":"10.1016/j.nucengdes.2026.114831","url":null,"abstract":"<div><div>A challenge of achieving high load capacity and low wear loss for siliconized graphite counterparts of the thrust bearings in shaft-sealed nuclear main pump of pressurized water reactor power plant is urgent due to the mixed lubrication under high contact pressure × sliding velocity (<em>pv</em>) factor above 65.0 MPa·m/s in high-temperature and high-pressure water. Wear and lubrication of siliconized graphite 42C-44SiC-14Si (vol%) with elevated graphite content were investigated using self-mated pairs in the high-temperature and high-pressure tribometer on a pin-on-disc configuration. Contact pressure <em>p</em> of 0.87 and 2.08 MPa with sliding velocity <em>v</em> from 0.5 to 33.1 m/s supply the <em>pv</em> factors of 0.4–68.8 MPa·m/s in 50 °C and 1.5 MPa water. A tribology-induced oxidation mechanism on SiC of siliconized graphite is proposed under the mixed lubrication. The selective oxidation of SiC characterizes the supersaturation of oxygen, fragmentation of SiC, and amorphization of silicon oxide. Oxidized SiC peaks alternating with valleys of non-oxidized graphite and Si phases facilitate the regular and corrugated-like texture on mated surfaces that promotes hydrodynamic effect. The lower wear depth of 0.12 μm and coefficient of friction (COF) of 0.015 are achieved under 68.8 MPa·m/s. Siliconized graphite 42C-44SiC-14Si with high graphite content provides the solid lubricant under boundary lubrication, and the uniform pressure distribution of water film under mixed lubrication. Therefore, the siliconized graphite as a promising material enables a long service life of thrust bearings in demanding nuclear applications, withstanding high load under operating conditions and offering superior resistance to more startup-shutdown cycles under specific conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114831"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190249","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":"Selection of RANS turbulence model for calculating thermal hydraulics of fuel assemblies of LMC reactors at low Reynolds numbers","authors":"K. Sergeenko ,&nbsp;A. Krutikov ,&nbsp;L. Golibrodo ,&nbsp;A. Zefirova ,&nbsp;M. Shishlenin","doi":"10.1016/j.nucengdes.2025.114741","DOIUrl":"10.1016/j.nucengdes.2025.114741","url":null,"abstract":"<div><div>Liquid metal coolants are considered in various fast neutron reactors. This paper presents the results of DNS calculations of heat and mass transfer processes of liquid metal coolants in the following regions: a parallel-plate channel, a rod bundle at cross-flow, and a channel with a step. The regions are selected in such a way as to describe as comprehensively as possible the thermal hydraulics of the fuel assembly, in which the spacing of the rods is carried out by the spacer grids. Two Reynolds numbers (<em>Re</em> = 2500 and 5000) and three Prandtl numbers (<em>Pr</em> = 0.0043, 0.0154 and 0.0324) are considered. The Prandtl numbers considered correspond to the liquid metal coolants used in fast neutron reactors: sodium, lead and lead bismuth, respectively.</div><div>The results of DNS simulations were compared with the data obtained using various RANS turbulence models (k-ε realizable two layer, k-ω SST, k-ε EB and k-ε V2F). It is shown that the k-ε V2F turbulence model describes the thermal-hydraulic processes in the best way. It recommends use this model for thermal-hydraulic calculations of fuel assemblies in reactor plants with liquid metal coolants at low Reynolds numbers.</div><div>It is shown that for all considered RANS turbulence models, the error in determining the heat transfer increases with increasing Peclet number in the parallel-plate channel. This is due to the features of the mechanisms of turbulent heat transfer for coolants at low Prandtl number.</div><div>The temperature profile is also compared with theoretical data. Satisfactory agreement is shown for all simulations in the area up to y⁺ ≈ 60. This can be used in experimental studies focused on determining the Nusselt number.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114741"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190732","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":"The impact of Rudi J. J. Stamm'ler on the development of the nuclear industry in Argentina","authors":"Eduardo Villarino ,&nbsp;Aldo Ferri","doi":"10.1016/j.nucengdes.2026.114817","DOIUrl":"10.1016/j.nucengdes.2026.114817","url":null,"abstract":"<div><div>This paper examines the technical and methodological contributions of Rudi J. J. Stamm'ler that shaped the historical development of nuclear technology and reactor physics capabilities in Argentina. Through successive missions supported by the <span>International Atomic Energy Agency</span> (IAEA), Rudi Stamm'ler played a decisive role in the training of highly qualified human resources at the <span>Balseiro Institute</span> and in the establishment of national computational capabilities in reactor physics and design. His influence extended to the development of essential calculation tools, the consolidation of independent nuclear design methodologies in Argentina, and the professional growth of engineers who later contributed to major national and international nuclear initiatives. The long-term impact of this work is reflected in the continued use of the computational methods he promoted and in an enduring technical and academic legacy within Argentina's nuclear engineering community.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"450 ","pages":"Article 114817"},"PeriodicalIF":2.1,"publicationDate":"2026-04-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146190254","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}