Nuclear Engineering and Design最新文献

{"title":"Numerical analysis of a water-LBE interaction experiment: Sensitivity analysis, inverse uncertainty quantification and uncertainty propagation","authors":"Qing Zhou,&nbsp;Xin’an Wang,&nbsp;Feng Mao","doi":"10.1016/j.nucengdes.2025.114035","DOIUrl":"10.1016/j.nucengdes.2025.114035","url":null,"abstract":"<div><div>Accurately predicting reactor behavior during Steam Generator Tube Rupture (SGTR) events in Lead-Bismuth Cooled Fast Reactors (LBFRs) is critical for ensuring safety and reliability. This study employs a two-dimensional numerical simulation model to analyze water-LBE interactions during SGTR events with an experiment conducted at the China Nuclear Power Technology Research Institute (CNPRI). A comprehensive sensitivity analysis identified key input parameters that significantly influence pressure responses within the primary pool. Utilizing the Input Parameter Range Evaluation Methodology (IPREM), uncertainty ranges for these parameters were systematically determined. Monte Carlo method, guided by Wilks’ formula, propagated these uncertainties to generate uncertainty bands for pressure responses. The simulation results demonstrated that the uncertainty bands effectively encompassed the observed pressure transients, confirming the model’s reliability.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114035"},"PeriodicalIF":1.9,"publicationDate":"2025-04-08","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799002","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":"Bottom-up levelized cost estimation of low-enriched and low-pressure nuclear batteries","authors":"Gyutae Park,&nbsp;Jacopo Buongiorno,&nbsp;Koroush Shirvan","doi":"10.1016/j.nucengdes.2025.113936","DOIUrl":"10.1016/j.nucengdes.2025.113936","url":null,"abstract":"<div><div>Nuclear batteries (NBs), also known as microreactors, can potentially reduce development and deployment timeline for nuclear energy. However, their lack of economy-of-scale challenges their ability to achieve reasonable cost of energy production. We developed a cost analysis tool that can guide designers to the key attributes that enable cost reduction for NBs. These attractive features are embedded into two NBs for the near-term markets (low enriched uranium fuel, proven experience, and well-known core materials)—the sodium-cooled graphite moderated thermal reactor and the organic-cooled water-moderated thermal reactor. Individual reactor concepts and their point design are presented. Both systems operate at low pressure, which further simplifies the design and operation. The levelized cost of each NB plant operating as a single-unit plant at 15 MWth, are compared for a successful first-of-a-kind production (FOAK) unit, and the Nth-of-a-kind (NOAK) unit. Then, three NB cost-reduction schemes are explored: power uprates, co-siting and equipment sharing, and multi-batch fueling. Based on the levelized costs, the 1-unit, sodium-cooled and organic-cooled NBs for both FOAK and NOAK units would be competitive only in remote markets. However, through a combination of the three NB cost-reduction strategies, the organic-cooled NB would become competitive also in larger markets in the U.S. We find that the most effective parameters in the order of reducing NB’s costs are: 1) higher power reactor designs, 2) multiple-reactor-unit-plants, and 3) batched-fueling scheme.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 113936"},"PeriodicalIF":1.9,"publicationDate":"2025-04-07","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143785934","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":"Conceptual core design study of a pipe type transportable molten salt fast reactor","authors":"Seong Jun Yoon,&nbsp;Tongkyu Park,&nbsp;Sung Kyun Zee,&nbsp;Jae Uk Seo,&nbsp;Yubin Go","doi":"10.1016/j.nucengdes.2025.114033","DOIUrl":"10.1016/j.nucengdes.2025.114033","url":null,"abstract":"<div><div>This study presents a new design for a transportable micro molten salt reactor (MSR) that diverges from conventional upright cylindrical configurations by utilizing a horizontally elongated, thin-walled pipe structure. This novel design aims to facilitate secure transportation and enhance the formation of a molten salt flow field within the reactor core This innovative design aims to facilitate secure transport and enhance the formation of a molten salt flow field within the reactor core. Emphasizing compactness for container loading, the reactor maximizes reflector efficiency and integrates adaptable control mechanisms suitable for its configuration. The horizontally elongated pipe reactor concept allows for the optimal arrangement of subsystems, enhancing vibration safety during transportation by lowering the overall system’s center of gravity, which, in turn, improves durability against vibrations and external impacts.</div><div>The reactor’s total dimensions are 194.68 cm in width, 185.84 cm in length, and 133.84 cm in height, incorporating U-shaped geometries with a 40 cm diameter. The single reactor system meets the target reactivity of 1.03 or higher at the beginning of the cycle and is capable of continuous operation at a 10 MWth output for a period of 3 years. By employing linear and U-shaped geometries, this design reduces the overall thickness and length while offering the flexibility to extend the reactor’s length to meet varying output requirements. This work highlights the potential of a transportable and efficient micro MSR to meet the growing demand for distributed sustainable energy solutions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114033"},"PeriodicalIF":1.9,"publicationDate":"2025-04-06","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143783870","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 FeCrAl cladding thermo-mechanical performance: CIEMAT’s contribution to IAEA/CRP ATF-TS","authors":"Pau Aragón,&nbsp;Francisco Feria,&nbsp;Luis E. Herranz","doi":"10.1016/j.nucengdes.2025.114034","DOIUrl":"10.1016/j.nucengdes.2025.114034","url":null,"abstract":"<div><div>This paper provides insights into the response of the advanced technology fuel (ATF) cladding FeCrAl during postulated design basis accident (DBA) and design extension condition without significant fuel degradation (DEC-A) scenarios. Such insights are gained through the development and application of in-house extensions of the FRAPCON/FRAPTRAN fuel performance codes, coupled with the statistical tool DAKOTA, within the framework of a loss-of-coolant accident (LOCA) safety evaluation methodology. While most of the specific FeCrAl models and correlations embedded in these extensions have been documented in the existing literature, the derivation of an instantaneous plasticity model describing the strain-hardening behaviour of FeCrAl alloy C26M is presented for the first time in this paper. The application of the methodology to the DEC-A/LOCA scenario suggests an improved performance of the advanced cladding material, as it maintains its integrity, in contrast to Zircaloy. However, in the DBA/LOCA scenario, no significant differences between these cladding materials were observed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114034"},"PeriodicalIF":1.9,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777470","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Aircraft impact: Coupled dynamic simulations part 1: Modelling aspects","authors":"Pierre Wörndle ,&nbsp;Robert Borsutzky ,&nbsp;Thomas Schubert","doi":"10.1016/j.nucengdes.2025.114046","DOIUrl":"10.1016/j.nucengdes.2025.114046","url":null,"abstract":"<div><div>During the last five decades, the load case “aircraft impact” changed constantly with regard to the definition of the threat, the applied analysis methods, and the required checks; particularly in the analysis of concrete structures in the nuclear sector. On the one hand, the load case was extended from military aircrafts to commercial aircrafts and on the other hand, the rapid development of computer hardware allowed more and more complex numerical simulations with detailed modelling of aircraft and building structure in a coupled analysis. Keeping this development in mind, as well as additional confidential aspects, increasing authorities’ requirements and the lack of detailed normative guidelines, these kinds of specialized analyses can only be performed with deep knowledge not only of the general aspects of the aircraft impact load cases but also of the complex numerical modelling and simulation of these situations.</div><div>This article is the first part of a two-part publication. In terms of content, the two publications focus on the two overarching topics that need to be considered in a typical aircraft impact assessment: on the one hand, the aspects of modeling and structural analysis and, on the other, the assessment of vibrations and the design of mitigation measures. The first part of this paper covers critical modeling aspects within the scope of a coupled structural analysis as well as aspects of assessing the structural damage. For this, this paper summarizes some of the main issues regarding the actual state-of-the-art Aircraft Impact Analysis (AIA) and points out the most discussed topics with focus on numerical modelling aspects in coupled dynamic AIA simulations.</div><div>One of the main conclusions for a state-of-the-art Aircraft Impact Analysis (AIA) is the recommendation of a detailed numerical aircraft model or at least of the center wing box and the attached wings and a coupled dynamic simulation with the impacted structure. With regard to the introduced shear forces, this approach is the only one that allows for the necessary detailed analysis of local impact effects and ultimately leads to a comprehensive assessment of the structural resistance.</div><div>In the second part of this publication, Wörndle and Borsutzky (planned to be published), the aspects of assessment criteria regarding induced vibrations are discussed, supplemented by examples of mitigation methods.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"437 ","pages":"Article 114046"},"PeriodicalIF":1.9,"publicationDate":"2025-04-05","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143777469","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":"Validation of the neutron cross section processing code MGGC3.0 via JOYO-70 reactor physics experiments","authors":"Teng Zhang,&nbsp;Xubo Ma,&nbsp;Xudong Ma,&nbsp;Zhulun Li,&nbsp;Fuxing Wang","doi":"10.1016/j.nucengdes.2025.114045","DOIUrl":"10.1016/j.nucengdes.2025.114045","url":null,"abstract":"<div><div>Fast neutron reactor is a critical design within the Generation IV nuclear reactor systems. In this study, a high-precision neutron cross-section processing code named MGGC3.0 was developed. It directly applies HFG (hyperfine group:∼400000) cross-section data for resonance calculations and utilizes problem-dependent HFG neutron energy spectrum for energy group merging to produce the UFG (ultrafine group:∼2000) cross-section to take into account the complicated resonance self-shielding effect between isotopes. The computation of UFG elastic scattering matrix is expedited through prefabricated scattering function method. For the production of few-group cross section, MGGC3.0 conduct critical buckling searches and employs a two-region approximation for fuel and non-fuel assemblies, respectively. This process calculates the neutron energy spectrum for energy group merging to obtain the few-group cross section. Initially, verification was conducted using three fuel assemblies: MOX, UO2, and U-TRU-Zr. This involved comparing the UFG macroscopic cross-sections produced by MGGC3.0 with those obtained from OpenMC calculations. Subsequently, the code underwent verification using a series of fast reactor benchmarks in ICSBEP. This entailed comparing the eigenvalues computed based on cross sections produced by MGGC3.0 with those calculated by RMC. Lastly, validation of the code was conducted using the JOYO MK-I series zero-power experimental setup. This involved comparing the calculated and experimental values of control rod worth, sodium void reactivity, and fuel replacement reactivity. The computational results of the verification and validation processes indicate that the neutron cross sections produced by the MGCC3.0 code exhibit high accuracy, thereby furnishing precise cross-sectional data for fast reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114045"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767984","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":"Thorium and fully ceramic microencapsulated TRISO fuel neutronics feasibility analysis in a gas cooled fast reactor: Enhancing transmutation of long-lived fission products","authors":"Shohanul Islam","doi":"10.1016/j.nucengdes.2025.114014","DOIUrl":"10.1016/j.nucengdes.2025.114014","url":null,"abstract":"<div><div>This study investigates the feasibility of using Fully Ceramic Microencapsulated (FCM) TRISO fuel and thorium fuel in gas-cooled fast reactor, focusing on enhancing the transmutation of long-lived fission products by performing neutronics analysis using the OpenMC Monte Carlo code. The implementation of FCM and modifications to the TRISO layer aim to decrease the moderation effect of the TRISO fuel and achieve a harder neutron spectrum. Four alternative FCM TRISO fuels were proposed by replacing the porous buffer, inner pyrolytic carbon, and outer pyrolytic carbon layers with SiC, ZrC, TiC, and Si₃N₄ in each case. For thorium fuel, two options were investigated-ThUC and ThPuC. The analysis of neutronics parameters revealed that all models achieved a harder neutron spectrum, with all FCM models displaying more harder neutron spectrum than others. This enhancement in neutron spectra and the robust safety of FCM came with a decrease in cycle length and a marginal increase in the power peaking factor due to a more non-uniform neutron flux. Nevertheless, the FCM models still achieved a satisfactory long core life and maintained power peaking factors within acceptable limits. In contrast, the thorium models, particularly ThUC, demonstrated a longer cycle length and an improved power peaking factor. To completely analyze the viability of all models a comprehensive reactivity parameters calculation was performed including reactivity swing, effective delayed neutron fraction, fuel temperature coefficient, power coefficient of reactivity, control rod worth, and shutdown margin. The findings revealed that all models achieved satisfactory results across all reactivity parameters. Notably, all FCM models exhibited improved power coefficient, control rod worth, and shutdown margin compared to the other models. This comprehensive neutronics analysis suggests that while all proposed models displayed satisfactory neutronics performance, the FCM models showed superior reactivity performance. Notably, the FCM model demonstrated significantly improved transmutation efficiency for four long-lived fission products: Nb-94, Pd-107, I-129, and Sm-151.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114014"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767983","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 review on the state of thermal hydraulics research on air ingress scenarios in High-Temperature Gas-cooled Reactors following a D-LOFC","authors":"Matthew Scheel ,&nbsp;Piyush Sabharwall ,&nbsp;Richard Schultz ,&nbsp;Daniele Ludovisi ,&nbsp;Gianluca Blois","doi":"10.1016/j.nucengdes.2025.113946","DOIUrl":"10.1016/j.nucengdes.2025.113946","url":null,"abstract":"<div><div>With the expectation of near-immediate carbon neutrality, widespread implementation of proven High-Temperature Gas-cooled Reactors (HTGRs) embodies a viable solution pathway given their inherent, passive safety features and high thermal efficiency. This study provides an overview of the current state of research involving the thermal hydraulics associated with air ingress from a depressurized loss of forced cooling (D-LOFC) in HTGRs. Accurately characterizing and predicting the physical phenomena underlying air ingress is of paramount concern, as the integrity of the fuel and core graphite support structures are threatened by the presence of oxygen. Broadly speaking, the air ingress scenario can be delineated into three main stages: (1) Depressurization, (2) Density-Driven Flow, and (3) Natural Convection. In tandem with the underlying fundamental theory, this review collates and synthesizes the existing body of contemporary research concerning the air ingress scenario following a D-LOFC. As evinced by this review, our current understanding and predictive abilities have benefited from extensive research, predominantly concentrated on the rate of air ingestion into the core. Additional research is necessary to holistically capture the phenomenology of an air ingress scenario following a D-LOFC by considering an additional variable: the oxygen content of the ingressing air. The latter variable requires investigation into the complex interactions of the fully integrated system. Additionally, while numerical tools are evolving domestically through the Nuclear Energy Advanced Modeling and Simulation program, a sufficiently validated code remains absent.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 113946"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767987","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 investigation on quenching behavior across spacer grid during reflooding transient","authors":"Long Ji ,&nbsp;Xiaojing Liu ,&nbsp;Wei Xu ,&nbsp;Wei Zeng ,&nbsp;Jie Wang ,&nbsp;Hui He","doi":"10.1016/j.nucengdes.2025.114042","DOIUrl":"10.1016/j.nucengdes.2025.114042","url":null,"abstract":"<div><div>During a Loss Of Coolant Accident (LOCA), the reflooding transient involves complex two-phase flow heat transfer process betweeen dispersed liquid phase, continuous vapor phase and high-temperature wall. The presence of spacer grids along the entire rod bundle has significant effects on the reflooding heat transfer phenomena during reflooding transients by interacting with entrained droplets. Quenching behavior in different regions of the spacer grid during the reflooding transient is studied experimentally using the 2 × 2 rod bundle test facility. The axial and circumferential quenching behaviors of the heater rods upstream and downstream of the spacer grid are analyzed for different reflooding velocities and linear power densities. Experimental results show that earlier occurrence of quenching downstream of the grid spacer is observed under low reflooding velocities and high linear power densities due to the mechanism of droplet breakup by dry spacer grid. On the other hand, at high reflooding velocity and low linear power density, a wetted grid results in increased downstream droplet size and decreased heat transfer performance, causing the linear change of the quench front curve. The experimental results also indicate that the circumferential quenching process of the heater rod upstream and downstream of the spacer grid is inconsistent due to the influence of the inhomogeneous flow pattern and spacer grid wetting conditions. The experimental data is used to support the development and validation of the model of film boiling heat transfer coefficient considering droplet breakup. The comparison results show that the boron equilibrium model can predict the performance of the boron concentration in the rod bundle channel with an accuracy of 5 %.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114042"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767986","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":"Optimization research of heat transfer coefficient prediction model for supercritical water based on Bayesian search algorithm","authors":"Ma Dongliang ,&nbsp;Zhou Tao ,&nbsp;Huang Yanping","doi":"10.1016/j.nucengdes.2025.114036","DOIUrl":"10.1016/j.nucengdes.2025.114036","url":null,"abstract":"<div><div>In order to make better use of machine learning algorithm to perform thermo-hydraulic analysis of supercritical water reactor, different intelligent algorithms are used to optimise the model parameters for predicting the heat transfer coefficient of supercritical water. The accuracy changes of stochastic search and Bayesian search algorithms in predicting the heat transfer coefficient under different parameter spaces are compared and analysed. The results show that the search space and the initial distribution assumption have a large impact on the results. The Bayesian search algorithm is relatively less affected by the search space and parameter distribution assumptions. The prediction accuracy obtained by Bayesian search is 1.25–2.88% higher than that obtained by random search. After optimising the model parameters, the average test accuracy of predicting the heat transfer coefficient of supercritical water is more than 96.4%. At the same time, the spatial distribution characteristics of the optimal parameter points obtained by different search algorithms are analysed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"438 ","pages":"Article 114036"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143767985","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}