Nuclear Engineering and Design最新文献

{"title":"Validation of NekRS pebble bed modeling for low flow scenarios","authors":"Michael Seneca ,&nbsp;Haomin Yuan ,&nbsp;Tri Nguyen ,&nbsp;David Reger ,&nbsp;Victor Coppo Leite ,&nbsp;Luiz Aldeia ,&nbsp;Joseph Seo ,&nbsp;Seth Macias ,&nbsp;Dezhi Dai ,&nbsp;Elia Merzari ,&nbsp;Yassin Hassan","doi":"10.1016/j.nucengdes.2025.114693","DOIUrl":"10.1016/j.nucengdes.2025.114693","url":null,"abstract":"<div><div>This study validates the high-fidelity NekRS Large Eddy Simulation (LES) framework for modeling conjugate heat transfer in low-flow pebble bed reactor scenarios, using the SANA experiments as a benchmark. Verification began by comparing the MOOSE Net Radiation Method to a 1D analytical model for a tall-cavity case, yielding Nusselt numbers within 1% relative difference. Next, the NekRS P1 approximation was compared to MOOSE in a single-pebble case, where temperature and velocity profiles agreed within 5% relative difference. Validation proceeded with the 283-pebble bed experiments at Texas A&amp;M, in which NekRS velocity profiles fairly matched experimental measurements despite uncertainties in pebble placements. Finally, full conjugate heat transfer simulations of the SANA pebble beds (5 kW helium–graphite and helium–alumina cases) were performed, with the optimized configuration producing a mean temperature error of 1.56%. The effective chamfer thermal conductivity (representing conductive contact between pebbles) introduced significant uncertainty in modeling the SANA experiments; further work will optimize this parameter based on a refined contact resistance model. Simulations will also be extended to higher-power conditions to continue validating conjugate heat transfer in NekRS for pebble-bed reactor designs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114693"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884689","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 and simulation of a free-piston Stirling generator with built-in reactor core for space applications","authors":"Guang Teng ,&nbsp;Jianying Hu ,&nbsp;Qi Lu ,&nbsp;Yanlei Sun ,&nbsp;Baifeng An ,&nbsp;Ercang Luo","doi":"10.1016/j.nucengdes.2025.114729","DOIUrl":"10.1016/j.nucengdes.2025.114729","url":null,"abstract":"<div><div>In this paper, a novel Stirling generator is proposed, in which a reactor core with flow channels replaces the original high-temperature heat exchanger, consequently, the heat transfer loop between the reactor and the Stirling engine is eliminated, enabling an integrated design of the reactor and thermoelectric converter. In order to ensure reactor criticality and protect the linear motor from radiation, an enlarged reactor core, a neutron reflector and an acoustic transmission tube are innovated into the engine. A numerical model is established and simulations are carried out to study the influence of the newly introduced part on the performance. The results show that the enlarged core, newly introduced reflector and acoustic transmission tube introduce additional exergy loss. Through design optimization, the overall thermoelectric conversion efficiency of the new configuration exceeds 39 %, which is only 4.84 % lower than that of the traditional. This demonstrates that the free-piston Stirling generator with built-in reactor core could be a feasible technology roadmap for new nuclear power.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114729"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884682","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":"Random vibration analysis of nuclear power plant structures","authors":"Gizem Çağlar Dal ,&nbsp;Kurtuluş Soyluk","doi":"10.1016/j.nucengdes.2025.114698","DOIUrl":"10.1016/j.nucengdes.2025.114698","url":null,"abstract":"<div><div>In this study, random vibration analysis of a nuclear power plant building under earthquake loading is performed based on a large-magnitude earthquake of Kobe 1995. A typical nuclear power plant structure widely used in China is selected as a numerical model and modeled as a 3D system. Within the scope of the study, random vibration and deterministic analyses were performed on firm, medium, and soft soils to determine the effects of earthquake motions on nuclear power plant systems. In the study, the theory of random vibration analysis based on the filtered white noise (FWN) ground motion model was utilized and it was intended to determine to what extent the FWN model reflects the real earthquake motion. In addition to soil type, the considered power plant system is analyzed for the ground motions showing near-fault and far-fault characteristics. As a result of the study, it is concluded that the FWN ground motion model used to model earthquake ground motion can be used to consider the effect of real earthquakes. It is also underlined that differences in soil type, fault type and analysis methods affect the results for the considered nuclear power plant structure.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114698"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884688","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":"Dynamic response of stainless steel-concrete-stainless steel containment structures in nuclear power plants under internal explosions","authors":"Yahang Tang ,&nbsp;Shoulong Zhang ,&nbsp;Jinhe Gao ,&nbsp;Qingrui Lu ,&nbsp;Hao Jiang","doi":"10.1016/j.nucengdes.2025.114699","DOIUrl":"10.1016/j.nucengdes.2025.114699","url":null,"abstract":"<div><div>With the increasingly stringent safety requirements for containment structures, Steel-Concrete-Steel (SCS) composite has become the preferred structural solution, which is further enhanced in safety by the incorporation of stainless steel. A comprehensive investigation was conducted into the dynamic performance of Stainless Steel-Concrete-Stainless Steel (SSCS) composite containment structures under internal explosive loading conditions. To ensure numerical simulation reliability, quasi-static and dynamic mechanical tests were conducted to obtain stainless steel constitutive parameters, which were then used to calibrate the material model. A refined finite element model of the SSCS composite structure was subsequently developed and validated against existing experimental data. Based on this validated model, a full SSCS composite containment structure under internal explosion loading was analyzed. A parametric study evaluated the influence of various factors on containment dynamic performance. By combining the equivalent Single-Degree-Of-Freedom (SDOF) method with multi-parameter regression analysis, a predictive formula for the SSCS containment Damage Enhancement Factor (DEF) was proposed. Results indicate that SSCS composite structures exhibit superior blast resistance compared to conventional SCS composite structures. Notably, increasing the steel plate thickness significantly mitigates the dynamic response of the containment dome. The proposed empirical formula accurately describes the relationship among the damage enhancement factor, scaled distance, and steel plate thickness, demonstrating strong applicability in engineering.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114699"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145799346","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":"Assessment of fluid-to-fluid similarity criteria to evaluate deterioration of heat transfer in flows of supercritical fluids","authors":"Chukwudi Azih","doi":"10.1016/j.nucengdes.2025.114683","DOIUrl":"10.1016/j.nucengdes.2025.114683","url":null,"abstract":"<div><div>In nuclear reactors, heat exchangers, and power cycles operating at supercritical thermodynamic conditions, heat transfer deterioration is a concern for the design and safety of primary heat removal components. Such deterioration can cause localized excursions in surface temperature, potentially compromising the integrity of the pressure boundary that contains the pressurized coolant. Therefore, accurately predicting flow conditions that lead to this deterioration is essential. To address this, numerous studies have investigated heat transfer at supercritical fluid states using various working fluids, including water. Building on previously published similarity criteria, this work evaluates deteriorated heat transfer predictive criteria against a more comprehensive dataset of over 480 tests. Unlike conventional approaches that normalize based on pressure at the critical state, the updated similarity criteria define surrogate fluid states using state-space similarity with peak Prandtl number, which arguably has more relevance to the flow. Additionally, dynamic flow criteria were examined by comparing the Richardson number with other buoyancy parameters and relevant ranges of Reynolds numbers were specified. The assessments were conducted for tubular geometries under upward flow, downward flow, and horizontal flow conditions. Particularly strong results are observed for the upward flow dataset, which comprises more than half of the database tests. Based on these evaluations, new predictive correlations for deteriorated heat transfer are proposed, applicable across broad ranges of Reynolds numbers and buoyancy parameters and fluids.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114683"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145799352","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":"On modeling the seismic response of spent nuclear fuel assemblies in vertical dry storage casks","authors":"Fady A. Elshazly ,&nbsp;Elnaz Seylabi ,&nbsp;Lee Joungyeul","doi":"10.1016/j.nucengdes.2025.114719","DOIUrl":"10.1016/j.nucengdes.2025.114719","url":null,"abstract":"<div><div>Dry storage casks (DSCs) have safely contained spent nuclear fuel (SNF) for decades, including at sites in seismically active regions. However, due to the continued absence of a U.S. deep geologic repository for disposal of SNF, they are now expected to remain on-site in dry storage systems for periods of time much longer than originally anticipated. Therefore, it is important to ensure SNF can continue to meet safety regulations while it continues to be stored on-site until it is disposed of. This paper presents a detailed modeling and analysis framework for a mock-up DSC system, including a high-fidelity finite element model of a surrogate fuel assembly. A multi-step approach is proposed to reduce the computational burden of full-system simulations, wherein an auxiliary model is proposed to approximate the boundary conditions for high-resolution fuel assembly models. Comparative studies under various seismic excitations indicate that the multi-step approach accurately reproduces the salient dynamic features, including system-level peak accelerations and localized stress peaks. This strategy achieves over 70% reduction in computational cost relative to a fully detailed DSC model without significant loss of accuracy. Further parametric analyses with an ensemble of ground motions — from Western and Central-Eastern United States sites — demonstrate that the considered fuel assembly, in the absence of degradation effects, maintains stresses less than 100 MPa and 5% damped spectral accelerations less than 8 g.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114719"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145884684","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 of the irradiation embrittlement dependent fracture strength and mode I fracture toughness for ceramics","authors":"Feilong Zhang ,&nbsp;Weiguo Li ,&nbsp;Yanli Ma ,&nbsp;Zhaoliang Qu ,&nbsp;Liming Chen ,&nbsp;Wenli Pi","doi":"10.1016/j.nucengdes.2025.114738","DOIUrl":"10.1016/j.nucengdes.2025.114738","url":null,"abstract":"<div><div>Quantitative characterization of irradiation embrittlement effect is important to ensure the service safety of ceramic materials in irradiation environments. Based on the Li's Principle of Energy Equivalence, taking into account the equivalent relationship between the irradiation energy absorbed by the material and the strain energy, theoretical models for irradiation embrittlement dependent fracture strength and mode I fracture toughness of ceramics have been developed, respectively. These models reveal a quantitative relationship between irradiation fluence and fracture strength/mode I fracture toughness, enabling the prediction of irradiation data over a wide range based on limited irradiation experimental data. The predicted results of both models are well validated by all available experimental data, including two groups of fracture strength data and four groups of mode I fracture toughness data. These models offer an effective theoretical approach for the convenient prediction of the mechanical properties of ceramic materials under wide irradiation fluence ranges. This effectively reduces reliance on difficult irradiation experiments, while providing a theoretical basis for the structural design and reliability assessment of reactor ceramics, as well as for regulatory and cost optimization aspects.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114738"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939570","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":"Evaluating primary circuit pressure reduction strategy in VVER-1000/V446 nuclear reactor during SBO severe accident with MELCOR 1.8.6","authors":"Abolhasan Nabavi ,&nbsp;Ahmad Pirouzmand ,&nbsp;Kamal Hadad","doi":"10.1016/j.nucengdes.2025.114744","DOIUrl":"10.1016/j.nucengdes.2025.114744","url":null,"abstract":"<div><div>In this study, the MELCOR 1.8.6 code is used to simulate the progression of an SBO severe accident over a 24-h period in a VVER-1000/V446 nuclear reactor. Multiple scenarios are analyzed to assess the effectiveness of the primary circuit pressure reduction strategy, which is triggered by operator action when the core exit temperature reaches a predefined setpoint. Sixteen temperature setpoints are considered: 350, 375, 400, 425, 450, 475, 500, 525, 550, 575, 600, 615, 635, 650, 675, and 700 °C. For each setpoint, four modes of operator action delay are evaluated: one with no delay, and three with delays of 110, 600, and 1397 s. In each mode, key parameters are examined, including fuel failure time, lower head failure time, containment failure time, in-vessel hydrogen generation, total radioactive release, and core degradation progression. The variation in these parameters across different setpoints highlights the importance of selecting an optimal setpoint for each delay mode and time interval. Also, the correct estimation of “the time required to the appropriate countermeasure actions” up to 70 min after the SBO, plays a significant role for selection of the optimal setpoints. If the estimation of this time, is less than 4 h, the optimal temperature setpoint, irrespective of the operator delay, is 350 °C. Conversely, depending upon the operator delay and the estimation of “the time required for the appropriate countermeasure actions”, various temperature setpoints are employed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114744"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939571","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":"Three-dimensional crack reconstruction from Beam–Particle Model for CFD-based leakage assessment","authors":"Omar Najjar ,&nbsp;Thomas Heitz ,&nbsp;Cécile Oliver-Leblond ,&nbsp;Jean-Louis Tailhan ,&nbsp;Giuseppe Rastiello ,&nbsp;Frédéric Ragueneau","doi":"10.1016/j.nucengdes.2025.114718","DOIUrl":"10.1016/j.nucengdes.2025.114718","url":null,"abstract":"<div><div>Accurate crack characterization in reinforced concrete structures, such as the containment walls of 1300 MWe nuclear power plants, is essential for reliable air leakage estimation. The complexity of crack patterns — including their path, opening, and surface roughness — significantly influences predictions of air leakage. Traditional modeling approaches, such as applying Poiseuille’s law to simplified geometries, depend on indirect parameters like the tortuosity coefficient, which is challenging to predict accurately and has limited applicability, thereby increasing uncertainty. This study introduces a novel numerical post-processing tool based on the Beam–Particle Model (BPM) approach, aimed at enhancing the precision of crack predictions derived from finite element simulations. It employs graph theory to detect micro-crack paths and reconstruct a possible macro-crack geometry, thus improving the representation of crack characteristics. The generated data can be integrated into computational fluid dynamics (CFD) simulations for more precise airflow modeling or used to calibrate simplified methods, such as Poiseuille’s law, with numerically obtained crack descriptors. The fracture geometries produced by the tool in the Brazilian splitting test are validated against optical microscopic observations, demonstrating the tool’s ability to capture crack tortuosity. In contrast, the PSD-based assessment of the Hurst exponent on BPM-generated crack surfaces yields non-physical values, so surface roughness could not be quantified numerically due to BPM computational time constraints. In addition, the current reconstruction algorithm is limited to a single dominant crack path and cannot handle crack branching or macro-crack turnover, although it can detect separated macro-cracks. These results establish a benchmark for future efforts to incorporate detailed crack geometries into CFD-based air leakage estimates, underscoring the tool’s potential to reduce uncertainty in such predictions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114718"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939572","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":"SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications","authors":"Aaron Epiney ,&nbsp;Gerhard Strydom ,&nbsp;Robert Kile ,&nbsp;Jonathan Barthle ,&nbsp;Izabela Gutowska ,&nbsp;Benjamin Nakhnikian-Weintraub ,&nbsp;Thanh Hua ,&nbsp;Ling Zou ,&nbsp;Jun Fang ,&nbsp;Krishna Podila ,&nbsp;Xianmin Huang ,&nbsp;Qi Chen ,&nbsp;Tariq Jafri ,&nbsp;Geoffrey Waddington ,&nbsp;Peter Pfeiffer","doi":"10.1016/j.nucengdes.2025.114737","DOIUrl":"10.1016/j.nucengdes.2025.114737","url":null,"abstract":"<div><div>Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications.</div><div>The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations.</div><div>This paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"448 ","pages":"Article 114737"},"PeriodicalIF":2.1,"publicationDate":"2026-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145939574","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}