Nuclear Engineering and Design最新文献

{"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":"2026-01-20","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":"Theoretical model of fluid-structure interaction and prediction of fluidelastic instability for wire-wrapped fuel rods in axial flow","authors":"Shusheng Dai ,&nbsp;Xiaochang Li ,&nbsp;Yu Zhang ,&nbsp;Ruifeng Tian ,&nbsp;Jiming Wen ,&nbsp;Sichao Tan","doi":"10.1016/j.nucengdes.2026.114786","DOIUrl":"10.1016/j.nucengdes.2026.114786","url":null,"abstract":"<div><div>Wire-wrapped fuel rod bundles are commonly employed in Generation IV fast reactors. However, the swirl and perturbations induced by the helical wires can significantly enhance fluid-structure interaction and increase the risk of instability, and traditional models still struggle to capture the response characteristics. This study establishes a high-order theoretical model of fluid-structure interaction for wire-wrapped fuel rods in axial flow, grounded in the fundamental principles of Newtonian mechanics. Building upon the dynamic framework of a bare rod, the model incorporates the additional effects of wire-wrap mass, stiffness, and drag, enabling a concise yet systematic representation of wire-wrap effects. A dimensionless parameter system is established for nondimensionalization, and the Galerkin method is applied for discretization. The resulting system matrix is solved computationally to obtain the first wet-mode natural frequency and the critical instability velocity. The predicted results show good agreement with numerical simulations, with relative frequency errors below 8%, validating the model's reliability. Parameter sensitivity analysis is further performed to elucidate the effects of wire-wrap diameter and pitch on the critical instability velocity, and its coupled regulatory mechanisms on system stability under different rod diameters and lengths. The results indicate that increasing the wire-wrap diameter intensifies flow disturbances and reduces the critical instability velocity; a critical range of wire-wrap pitch exists that leads to the lowest system stability, displaying non-monotonic characteristics; and variations in rod parameters influence the wire-wrap effect on system stability, with smaller diameters or longer rods promoting instability, diameter reduction enhancing coupling, and length increase weakening it.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114786"},"PeriodicalIF":2.1,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036024","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 state-of-the-art review of R&D for the super critical water-cooled reactor technology. Part I economics, thermalhydraulics, safety and licensing","authors":"Armando Nava Dominguez ,&nbsp;Chukwudi Azih ,&nbsp;Alberto D'Ansi Mendoza España ,&nbsp;Hussam Zahlan ,&nbsp;Guido Mazzini ,&nbsp;Alis Musa-Ruscak ,&nbsp;Sara Kassem ,&nbsp;Andrea Pucciarelli ,&nbsp;Walter Ambrosini ,&nbsp;Fabian Wiltschko ,&nbsp;Ivan Otic ,&nbsp;Tamás Varju ,&nbsp;Attila Kiss ,&nbsp;Pan Wu ,&nbsp;Elena Poplavskaia","doi":"10.1016/j.nucengdes.2026.114767","DOIUrl":"10.1016/j.nucengdes.2026.114767","url":null,"abstract":"<div><div>This study presents a summary of the most relevant research and development (R&amp;D) carried out to support the development of the only Generation IV water-cooled reactor endorsed by the Generation IV International Forum (GIF). The coolant of the proposed reactor is operated at supercritical water conditions, allowing for an increase in thermodynamic efficiency of the plant and the production of higher-grade process heat. Several collaborations have been established to support the development of this technology under the GIF umbrella, as well as through other international avenues. Therefore, the development is bolstered by a collective effort between numerous R&amp;D institutions across Asia, Europe, and North America. Globally, the R&amp;D programs have been methodologically executed in phases, namely: fundamental R&amp;D, validation and verification of assumptions used in R&amp;D analyses, and pre-conceptualization of supercritical water-cooled reactors (SCWRs).</div><div>This article summarizes the recent R&amp;D work performed to support the development of the SCWR technology on thermalhydraulics and safety, economics and licensing. Furthermore, R&amp;D highlights, identified knowledge gaps, conclusions, and recommendations are presented.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114767"},"PeriodicalIF":2.1,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036056","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":"Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT","authors":"Jacob A. Hirschhorn,&nbsp;Mustafa K. Jaradat,&nbsp;Ryan T. Sweet,&nbsp;Nicolas E. Woolstenhulme,&nbsp;Paul A. Demkowicz,&nbsp;David A. Reger,&nbsp;Paolo Balestra,&nbsp;Gerhard Strydom","doi":"10.1016/j.nucengdes.2026.114773","DOIUrl":"10.1016/j.nucengdes.2026.114773","url":null,"abstract":"<div><div>The United States nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. This work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles’ silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114773"},"PeriodicalIF":2.1,"publicationDate":"2026-01-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036058","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 conjugate heat transfer coupling model for supercritical water flowing in 2 × 2 ballooning ATF rod bundles","authors":"Chengrui Zhang ,&nbsp;Juan Chen","doi":"10.1016/j.nucengdes.2026.114782","DOIUrl":"10.1016/j.nucengdes.2026.114782","url":null,"abstract":"<div><div>The conjugate heat transfer characteristics of ballooning accident-tolerant fuel (ATF) rods in supercritical water were systematically investigated. A two-dimensional numerical multi-physics conjugate heat transfer program was developed for a 2 × 2 rod bundle with blocking sleeve and spacer grids, based on a comprehensive review of experimental and simulation studies on blocked flow, with particular emphasis on correlations for the friction coefficient and convective heat transfer coefficient. The program integrates a two-dimensional thermal conductivity model for ballooning fuel rod, solved by the finite difference method, with the simulation of convective heat transfer between the deformed cladding and coolant using various heat transfer coefficient correlations. Flow distribution between blocked and unblocked channels containing spacer grids is also simulated under supercritical pressure conditions, also employing the finite difference method with different friction coefficient correlations applied. Validation against SWAMUP experimental data shows that the relative error of the calculated axial cladding surface temperature remains within 1.6 %. Subsequently, the effects of blockage ratio, heat flux, and system pressure on heat transfer performance were analyzed. Key findings include: 1) the convective heat transfer coefficient downstream of the blockage region generally increases owing to enhanced turbulence, particularly at low swelling ratio. 2) When the swelling ratio exceeds 55 %, friction pressure drop contributes up to 87.1 % of the total pressure drop, leading to heat transfer deterioration. 3) Higher system pressures result in a greater total pressure drop and more uniform flow distribution between blocked and unblocked channels. 4) Heat transfer deterioration can occur when the inlet temperature approaches the pseudo-critical temperature. This research provides theoretical support for the design optimization and safe operation of SCWR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114782"},"PeriodicalIF":2.1,"publicationDate":"2026-01-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"146036055","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":"2026-01-18","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":"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":"2026-01-16","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":"Risk-informed structural integrity assessment of the HTR-PM reactor pressure vessel","authors":"Bowen Li,&nbsp;Haitao Wang,&nbsp;Yanhua Zheng,&nbsp;Zhengming Zhang,&nbsp;Heng Peng","doi":"10.1016/j.nucengdes.2026.114764","DOIUrl":"10.1016/j.nucengdes.2026.114764","url":null,"abstract":"<div><div>The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) in China represents the world's first demonstration power plant of its kind. As one of the Generation IV reactors, it is designed with inherent safety characteristics and offers potential for optimization in the structural integrity management of its reactor pressure vessel (RPV). Key advantages of the HTR-PM vessel, such as lower fast neutron flux than conventional pressurized water reactors, significantly slow thermal transients and reversed thermal gradients, collectively contribute to reducing both the risk and consequence of its failure. Using probabilistic fracture mechanics (PFM), a quantitative model has been developed to assess flaw initiation, crack propagation, and fracture of the RPV under typical HTR-PM transient and test conditions. This model enables assessment of the RPV failure probability, thereby establishing a basis for risk-informed safety evaluation and for optimization of both manufacturing and in-service inspection (ISI) requirements. Simulation results show that the RPV failure probability of HTR-PM is extremely low in the conservative case of no ISIs during its entire service life. Furthermore, flaws in circumferential weld regions contribute only minimally to the overall failure probability, suggesting potential for further extending inspection intervals or even exempting these areas from periodic inspection.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114764"},"PeriodicalIF":2.1,"publicationDate":"2026-01-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981783","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 a foreign object retrieval robot for the internal lower head of a nuclear power plant pressurizer","authors":"Fenghan Ran ,&nbsp;Shiqi Chen ,&nbsp;Shuai Huang ,&nbsp;Xingli Zhu","doi":"10.1016/j.nucengdes.2026.114785","DOIUrl":"10.1016/j.nucengdes.2026.114785","url":null,"abstract":"<div><div>Manual extraction of foreign objects from the lower head of a pressurizer poses significant challenges, including high personnel radiation exposure and operational complexity. This paper offers a comprehensive overview of the design and structure of a foreign object extraction robot for the lower head. It delves into the specific designs of key components like the clamping mechanism, waist and wrist rotating joints, shoulder and elbow swinging joints, and foreign object extraction actuators. Additionally, it provides an in-depth analysis of the robot's reliability. By enabling remote-controlled operation, the robot can extract foreign objects from the lower head without requiring personnel to enter the pressurizer. The findings of this study hold great significance for addressing similar foreign objects related incidents in pressurizers at other power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114785"},"PeriodicalIF":2.1,"publicationDate":"2026-01-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981785","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental study of decontamination factor in pool scrubbing with two-phase flow evolution","authors":"Taizo Kanai ,&nbsp;Miki Saito","doi":"10.1016/j.nucengdes.2026.114771","DOIUrl":"10.1016/j.nucengdes.2026.114771","url":null,"abstract":"<div><div>Pool scrubbing is a key mitigation process in nuclear severe accidents (SAs), in which aerosols and gaseous fission products (FPs) are injected into water and removed through gas–liquid interactions. Although the decontamination factor (DF) depends strongly on particle size and hydrodynamic conditions, systematic DF datasets obtained under high-flow, non-condensing conditions—together with corresponding detailed two-phase flow-field measurements—remain scarce. This study addresses this gap by providing a high-accuracy, size-resolved DF dataset consistent with independently measured flow-field data. A 0.5-m-diameter, 8-m-high test facility was used to reproduce the characteristic severe-accident flow evolution from large globules to bubble breakup and the formation of a fine-bubble swarm. Using Welas, SMPS, and ELPI+, size-resolved DF values for soluble CsI aerosols were obtained over a wide aerodynamic-diameter range (sub-0.05 μm to 10 μm). The mixture water level was varied continuously from 0 to approximately 5 m, enabling DF measurements across evolving flow structures. The DF exhibited clear dependencies on both particle size and water level, and the transition from the injection/breakup region to the swarm region was directly reflected in the DF behavior. These trends were consistent with detailed axial bubble-size evolution measured previously in the same facility.</div><div>An empirical DF correlation was developed as a function of mixture water level and aerodynamic diameter. Comparison of this correlation with the DF measurements, flow-field data, and the mechanistic MELCOR/SPARC90 model showed that, while the major hydrodynamic transitions were consistent with the model, the experimentally observed particle-size dependence and gas-flow-rate independence revealed characteristics associated with strongly turbulent, multidimensional bubble motion in the swarm region. Furthermore, SMPS measurements demonstrated a pronounced DF increase for ultrafine particles (&lt;0.05 μm) due to Brownian diffusion—providing new experimental evidence in a particle-size range for which reliable DF data have been largely unavailable.</div><div>A comparison with insoluble BaSO₄ aerosols generated under identical conditions showed consistently lower DF values for BaSO₄ despite nearly identical particle density, indicating that aerosol solubility has a significant influence on removal efficiency.</div><div>The systematic DF dataset obtained in this study elucidates the governing mechanisms of particle-size-dependent aerosol removal under realistic pool-scrubbing flow regimes. Combined with detailed flow-field measurements from the same facility, these results provide a robust benchmark for improving mechanistic DF models, particularly under high-flow conditions relevant to severe-accident source-term assessments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"449 ","pages":"Article 114771"},"PeriodicalIF":2.1,"publicationDate":"2026-01-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145981784","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}