Nuclear Engineering and Design最新文献

{"title":"High-fidelity and high-resolution simulation of two different rod ejection accidents in a NuScale-like small modular reactor with conventional and accident tolerant fuels","authors":"Zsolt Soti ,&nbsp;Paul Van Uffelen ,&nbsp;Arndt Schubert ,&nbsp;Ville Valtavirta ,&nbsp;Riku Tuominen ,&nbsp;Heikki Suikkanen ,&nbsp;Ville Rintala ,&nbsp;Andre Gommlich ,&nbsp;Emil Fridman ,&nbsp;Yurii Bilodid ,&nbsp;Luigi Mercatali ,&nbsp;Victor Hugo Sanchez-Espinoza","doi":"10.1016/j.nucengdes.2025.114183","DOIUrl":"10.1016/j.nucengdes.2025.114183","url":null,"abstract":"<div><div>This work presents a high-fidelity pin-by-pin simulation approach for a NuScale-like Small Modular Reactor core during a rod ejection accident (REA). We coupled 3D Monte Carlo neutron transport (Serpent), subchannel thermal–hydraulic (SUBCHANFLOW) and fuel performance (TRANSURANUS) codes using the Interface for Code Coupling (ICoCo), which is part of the EU’s Salome open source platform. To resolve fuel intra-assembly details, we simulated all the fuel rods and channels, subdividing them into axial slices and transferred calculated data between the codes using scalar fields saved in memory variables. Two different REA scenarios were modelled, and the behaviour of fresh-loaded cores with conventional UO₂ fuel with Zr-4 cladding and accident tolerant fuel (ATF) materials, U₃Si₂ fuel with FeCrAl cladding, were analysed. In both scenarios, the control rod was ejected within 0.1 s, followed by a SCRAM after two seconds. In the first moderate scenario, the control rod ejection occurred at 75% of the nominal power, whereas in the second accident scenario, it occurred at hot zero power (HZP) conditions. In the first scenario, the power increase was around 25%, while in the HZP case it amounted up to 600% and 300% of the nominal power for the core loaded with UO₂ and ATF-fuel and cladding, respectively. Detailed calculations were conducted on a High-Performance Computer (HPC). The results demonstrated the robustness and flexibility of the coupled code system, providing full-core behaviour and rod-level safety parameters and predicting as needed during the safety analysis support of the licensing processes. This paper outlines the system setup, presents rod-level results and underlines the usefulness to assess the performance of SMR-cores loaded with different fuel types under various REA scenarios. In the scenarios considered, we did not observe significant fuel rod deformations, and the core loaded with ATF-fuel and cladding showed a large margin to melting.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114183"},"PeriodicalIF":1.9,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169834","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":"Enhancing computational efficiency of Bayesian Inference by identifying the intensity measure range to update seismic fragility curves","authors":"Mrinal Jyoti Mahanta ,&nbsp;Saran Srikanth Bodda ,&nbsp;Abhinav Gupta ,&nbsp;Jeong-Gon Ha","doi":"10.1016/j.nucengdes.2025.114151","DOIUrl":"10.1016/j.nucengdes.2025.114151","url":null,"abstract":"<div><div>The United States Nuclear Regulatory Commission (US NRC) has established stringent criteria for the acceptance of new Standard Plant designs. These criteria require that the fragility curves have a high degree of confidence, especially in the low-probability regions. However, conventional methods of developing seismic fragility curves require a substantial number of dynamic analyses, which can be computationally intensive. To address this challenge, the Bayesian framework offers a more efficient and effective solution. Bayesian Inference enables the integration of prior knowledge with newly acquired data to refine the data-generating process. This manuscript presents a systematic Bayesian framework to update the seismic fragility curve of structures, systems, and components (SSCs). The efficiency of the framework is illustrated using an application to the seismic fragility of a concrete shear wall. Concrete Damage Plasticity Model (CDPM) is used to characterize the nonlinear behavior of concrete. The seismic fragility curve developed with our proposed approach aligns closely with those generated through conventional nonlinear simulations while significantly reducing the computational cost.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114151"},"PeriodicalIF":1.9,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169835","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":"Reverse application of the Andrews protocol for assessing risks in a hypothetical nuclear accident","authors":"Angelica C. Nogueira ,&nbsp;Hugo F. Menossi ,&nbsp;Rafael M. Hespanhol ,&nbsp;Andre C. Tavares ,&nbsp;Rodrigo C. Curzio ,&nbsp;Tercio Brum ,&nbsp;Edson R. Andrade","doi":"10.1016/j.nucengdes.2025.114188","DOIUrl":"10.1016/j.nucengdes.2025.114188","url":null,"abstract":"<div><div>This study introduces a novel approach to assessing radiobiological damage by employing an inverted version of the classical Andrews protocol (or Andrews nomogram) within the context of nuclear accidents. The methodology consists of two stages of computational simulation: The first stage simulates the release of radioactive materials from a pressurized water reactor (PWR), concentrating on scenarios of operational instability. In contrast, the second stage utilizes HotSpot Health Physics software to model the environmental dispersion of these materials. By estimating Total Effective Dose Equivalents (TEDE) at various locations and under different atmospheric stability conditions, we reverse the traditional sequence of the Andrews nomogram, which typically estimates lymphocyte counts based on radiation dose exposure. This innovative framework facilitates the development of a TEDE versus a Count curve, thereby enhancing the ability to anticipate the health implications following radiological incidents. The results highlight the importance of accurate modeling in predicting biological effects on populations exposed to radioactive releases and provide a tool to support responses and help design coping strategies enhancing the situational awareness of healthcare teams in nuclear emergencies.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114188"},"PeriodicalIF":1.9,"publicationDate":"2025-05-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169836","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":"Low cycle fatigue behavior of T91 ferritic-martensitic steel hollow specimen in oxygen saturated lead–bismuth eutectic at 200 ∼ 450 °C","authors":"Jianhua Ding ,&nbsp;Jibo Tan ,&nbsp;Ziyu Zhang ,&nbsp;Xinqiang Wu ,&nbsp;Xiang Wang ,&nbsp;Wei Ke","doi":"10.1016/j.nucengdes.2025.114185","DOIUrl":"10.1016/j.nucengdes.2025.114185","url":null,"abstract":"<div><div>The low cycle fatigue behavior of T91 ferritic-martensitic steel hollow specimens were investigated in air and with oxygen saturated lead–bismuth eutectic (LBE) filled at 200 ∼ 450 °C. A clear fatigue endurance “trough” was observed with LBE-filled hollow specimens at 300 ∼ 350 °C, while the fatigue lives in air at different temperatures are comparable. The fracture surfaces with LBE-filled at 200 ∼ 350 °C were characterized by quasi-cleavage cracking, which indicates the occurrence of liquid metal embrittlement. By contrast, oxide scale on surface with LBE-filled at 400 ∼ 450 °C inhibited fatigue crack initiation. The fatigue crack initiation mechanisms at different temperatures are discussed.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114185"},"PeriodicalIF":1.9,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169833","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 high order numerical method for 2D time-fractional neutron diffusion equation characterizing neutron transport in nuclear reactors","authors":"Jianxiong Cao ,&nbsp;Xiaodong Zhao ,&nbsp;Pradip Roul","doi":"10.1016/j.nucengdes.2025.114145","DOIUrl":"10.1016/j.nucengdes.2025.114145","url":null,"abstract":"<div><div>In this article, we investigate a numerical method for solving a two-dimensional time-fractional neutron diffusion equation that incorporates three Caputo fractional derivatives in time. First, the existence, uniqueness, and solution representation are established. We then develop an efficient numerical scheme using the linear Galerkin finite element method for spatial discretization and a second-order convolution quadrature for temporal discretization. Error estimates for both semidiscrete and fully discrete schemes are derived. Our proposed numerical method is demonstrated to achieve second-order convergence in both space and time. Numerical experiments are conducted to validate the theoretical results. Neutron fluxes are computed for both short and long simulation times to analyze the behavior of neutron diffusion under varying orders of the fractional derivative. Additionally, a comparison is made with the method presented in Roul et al. (2020) to highlight the advantages of our proposed approach.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114145"},"PeriodicalIF":1.9,"publicationDate":"2025-05-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144169832","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":"Research on predicting the diffusion of radionuclide in complex building scenarios based on a hybrid deep learning model","authors":"Yuheng Li ,&nbsp;Tao Wang ,&nbsp;Jintao Wang ,&nbsp;Wanxiao Guo ,&nbsp;Weiyi Li ,&nbsp;Hongbo Qiu ,&nbsp;Yue Lin ,&nbsp;Yilin Fang ,&nbsp;Minghua Lv","doi":"10.1016/j.nucengdes.2025.114187","DOIUrl":"10.1016/j.nucengdes.2025.114187","url":null,"abstract":"<div><div>In the event of an extreme nuclear accident, predicting the dispersion of radionuclide is critical for nuclear emergency response. Traditional atmospheric dispersion models struggle to balance accuracy and time efficiency, failing to meet the demands of nuclear emergency situations. Therefore, this study proposes an innovative hybrid deep learning model—Dual CNN-LSTM. On the Indianapolis dataset, the model demonstrates favorable predictive performance, with a coefficient of determination(R² = 0.6351, RMSE = 0.0495, training time = 1919.00 s, prediction time = 1.11 s).The study also found that incorporating Gaussian plume results into the input features reduced the model’s performance in complex scenarios. Through simulation validation, the plume shapes produced by the model were found to be highly consistent with experimental data, indicating that atmospheric stability significantly affects concentration peaks and providing a scientific basis for relevant decision-making.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114187"},"PeriodicalIF":1.9,"publicationDate":"2025-05-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144147536","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":"Preliminary study on the impact of Am-241 buildup on the safety of a medium-sized oxide-fueled SFR","authors":"T. Sathiyasheela ,&nbsp;Rita John ,&nbsp;A. Riyas","doi":"10.1016/j.nucengdes.2025.114171","DOIUrl":"10.1016/j.nucengdes.2025.114171","url":null,"abstract":"<div><div>In fast reactor, based on Plutonium (Pu) fuel, initially the required Pu is derived from the thermal reactor. Once the fast breeder reactor is commissioned then, it can breed more fissile fuel than it consumes, with that further more fast reactors can be commissioned. The vector composition of Pu, depends on the discharged burn up of the fuel from the thermal reactor. Among the Pu isotopes, half-life of Pu-241 is 14.33 years. Since the half-life of Pu-241 is 14.33 years and if the discharged fuel (from the thermal reactor) is kept under cooling for long duration, then Pu-241 would decay into Am-241, wherein the Pu-241 is fissile and Am-241 is an absorber of neutron. Though, Am-241 is removed before making the fuel pellets of fast reactor, still analyses are carried out to find out its effect on safety as it is not possible to remove the Am-241 in the already fabricated fuel pin. This study is important, as the presence of Am-241 reduces the delayed neutron fraction and enhances the void co-efficient. With that, when Am-241 is removed and enrichment normalization is done to get the require power and cycle length, the enriched fuel has more Pu and less U-238. This further reduces the Doppler constant and delayed neutron fraction. Analyses are carried out on a medium sized fast reactor to study the impact of Am-241 buildup on fast reactor safety. From the steady state study, it is concluded that, the change in steady state co-efficient with the decay of Pu-241 to Am-241 in three half-life is less than about 6 %. Under transient, the reactor is going to another steady state under UTOPA, it goes to CDA under ULOFA and the mechanical work potential is found to be less than 1 MJ. From the study, it is possible to conclude that, decay of Pu-241 to Am-241 doesn’t affect the overall safety of a medium sized sodium cooled fast reactor. The buildup of Am-241 depends on the initial amount of Pu-241, static and dynamic results of a fast reactor are influenced by factors such as reactor fuel, enrichment, size, and shape. Hence these results should be independently verified for other reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114171"},"PeriodicalIF":1.9,"publicationDate":"2025-05-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144147535","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":"Applicability of equivalent linear three-dimensional FEM analysis of reactor buildings to the seismic response of a soil–structure interaction system","authors":"Yoshitaka Ichihara ,&nbsp;Naohiro Nakamura ,&nbsp;Kunihiko Nabeshima ,&nbsp;Byunghyun Choi ,&nbsp;Akemi Nishida","doi":"10.1016/j.nucengdes.2025.114160","DOIUrl":"10.1016/j.nucengdes.2025.114160","url":null,"abstract":"<div><div>This paper evaluates the applicability of equivalent linear analysis of reinforced concrete, which uses frequency-independent complex damping with a small computational load, to the seismic design of nuclear power plant reactor buildings. To this end, a three-dimensional finite element method analysis of the soil–structure interaction focusing on nonlinear and equivalent linear seismic behavior of the building embedded in an ideally uniform soil condition (shear wave velocity Vs = 880 m/s) was performed for the Kashiwazaki–Kariwa Nuclear Power Plant Unit 7 reactor building. The equivalent linear analysis results correlated well with the nonlinear analysis results of the shear strain, acceleration, displacement, and acceleration response spectrum, demonstrating the effectiveness of the equivalent linear analysis method. Moreover, the equivalent linear analysis results were more conservative than those of nonlinear analysis using the material constitutive law in evaluating the shear strain of the external wall of the reactor building. From this observation, equivalent linear analysis tended to obtain a lower building stiffness than nonlinear analysis under the analysis conditions used in this paper. The equivalent linear analysis calculation results should be conservative in shear strain evaluation for seismic safety.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114160"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144138253","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 investigations addressing steam release in the water pool through a multi-hole sparger and spray water activation in the gas space","authors":"A. Oliva ,&nbsp;D. Paladino ,&nbsp;S. Carnevali ,&nbsp;S. Paranjape ,&nbsp;D. Grishchenko ,&nbsp;P. Kudinov ,&nbsp;S. Mimouni","doi":"10.1016/j.nucengdes.2025.114175","DOIUrl":"10.1016/j.nucengdes.2025.114175","url":null,"abstract":"<div><div>This article presents the experimental results and the phenomenological analyses of the H2P4 tests performed in the PANDA facility, within the OECD/NEA HYMERES project. H2P4 series is characterized by two phases: (i)<!--> <!-->the steam injection in the water pool with development of thermally stratified layer and pressurization of the system, and (ii)<!--> <!-->water spray injection above the pool to depressurize the gas space. This article analyzes two tests that are characterized by different sparger submergence levels. The results indicate that the sparger submergence affects the pool’s thermal behavior, and the pressurization history. A decrease in submergence height leads to a faster pressurization of the vessel. An analysis of the experimental data is carried out using mass and energy balances for the liquid and gas volumes. This analysis contributes to the understanding of the pool phenomena and suggests that mass and energy are lost through vaporization at the liquid–gas interface and a significant amount of steam is condensed at the walls.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114175"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144134187","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":"Thermal-hydraulic analysis of the superheated steam supply system in NHR200-II using refined semi-2D modeling","authors":"Zongyang Li ,&nbsp;Wentao Hao ,&nbsp;Wenwen Zhang ,&nbsp;Weihua Li ,&nbsp;Xingtuan Yang","doi":"10.1016/j.nucengdes.2025.114180","DOIUrl":"10.1016/j.nucengdes.2025.114180","url":null,"abstract":"<div><div>The low-temperature nuclear heating reactor, NHR200-II, is not only capable of providing district heating but also has potential for industrial steam production. To gain deeper insights into the thermal–hydraulic characteristics of its key component—the Horizontal Steam Generator (HSG)—and to verify the system’s energy balance at different power levels, this study employs a more refined numerical modeling approach compared to the previous HSG model in NHR200-II. Using the system code, refined models of the three heat exchange components in the system were developed, as well as the intermediate and reactor core loops. By incorporating additional longitudinal nodes and lateral connections, the model more accurately captures the complex lateral flow characteristics. The simulation results show a high degree of agreement with design values, validating the accuracy and reliability of the refined modeling approach. The study reveals that significant natural circulation occurs within the HSG, accompanied by localized complex flow patterns. These flow characteristics greatly enhance fluid mixing and improve heating efficiency. The system exhibits load-following capability, achieving stable transitions between 50 % and 100 % of rated power in 10 % increments. This research not only confirms the superheated steam supply capability of NHR200-II but also demonstrates the effectiveness of the refined modeling approach in capturing complex thermal–hydraulic phenomena. The results provide a solid foundation for system performance optimization and operational safety assurance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"441 ","pages":"Article 114180"},"PeriodicalIF":1.9,"publicationDate":"2025-05-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144134188","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}