Nuclear Engineering and Design最新文献

{"title":"A novel method about nuclear hydrogen production by methanol steam reforming for Small Modular Reactor (SMR)","authors":"Ruiyang Liu ,&nbsp;Zhiyi Peng ,&nbsp;Bin Du ,&nbsp;Huang Zhang ,&nbsp;Huaqiang Yin ,&nbsp;Shanfang Huang ,&nbsp;Yan Wang ,&nbsp;Shengwei Tang ,&nbsp;Chang Zeng ,&nbsp;Danrong Song","doi":"10.1016/j.nucengdes.2025.114444","DOIUrl":"10.1016/j.nucengdes.2025.114444","url":null,"abstract":"<div><div>Hydrogen is an important vector for industry as well as a fuel energy carrier. Nuclear hydrogen production offers distinct advantages, including low carbon emissions and high yield. Among nuclear technologies, Small Modular Reactor (SMR) stands out due to the feature like flexible deployment, short construction period, and enhanced safety features. Hence, hydrogen production using SMR could broaden its application scope and improve its economy viability. However, the output temperature of SMR, such as ACP100, is about 230 °C to 260 °C. The traditional method of hydrogen production using SMR is water electrolysis, which has low efficiency as compared to thermochemical method. Methanol Steam Reforming (MSR) for hydrogen production could work in the range of 180°C to 320°C, which matches the range of the output temperature of SMR. In this work, two kinds of hydrogen production systems by MSR using SMR were developed. Parameter sensitivity analysis of the systems, and the optimal reaction conditions were discussed. These results show that the suggested values of reaction temperature range, reforming pressure, and water-to-methanol mass ratio are 240 °C to 250 °C, 1.1 MPa, and 0.6 to 0.63, respectively. Then, MSR and water electrolysis using SMR were compared in terms of economics, which shows that the MSR (with an average hydrogen production cost of 0.30 US $/Nm³ H₂) is more economical than water electrolysis (with an average hydrogen production cost of 0.49 US $/Nm³ H₂). This work provides a new method of hydrogen production using SMR, which could highly improve its integrated applications.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114444"},"PeriodicalIF":2.1,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097502","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":"High-fidelity forced convection simulations of the University of Wisconsin–Madison air-cooled reactor cavity cooling system","authors":"Sinan Okyay ,&nbsp;David Reger ,&nbsp;Victor Coppo Leite ,&nbsp;Elia Merzari ,&nbsp;Paolo Balestra ,&nbsp;Gerhard Strydom","doi":"10.1016/j.nucengdes.2025.114388","DOIUrl":"10.1016/j.nucengdes.2025.114388","url":null,"abstract":"<div><div>Among the various types of advanced nuclear technologies that exist, high-temperature gas-cooled reactors (HTGRs) stand out for their co-generation capabilities and exceptional passive safety systems. HTGRs rely on a reactor cavity cooling system (RCCS) to remove decay heat from the reactor pressure vessel (RPV) during the course of long-term transients. Several studies have focused on experimental and numerical modeling of the RCCS; however, the literature reveals a notable gap when it comes to high-fidelity simulations such as direct numerical simulations (DNS) or large eddy simulation (LES) models of the RCCS. An LES model was developed for forced convection inside the RCCS so as to obtain a high-fidelity representation of the RCCS at a scale never before attempted. The LES model was compared against three different Reynolds-averaged Navier–Stokes (RANS) models (Realizable <span><math><mrow><mi>k</mi><mo>−</mo><mi>ϵ</mi></mrow></math></span>, <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> SST, and Reynolds stress model [RSM]). This comparison verified the robustness and accuracy of the turbulence modeling approach of the RANS models for the RCCS. Finally, based on the comparison between the LES and the RANS models, the <span><math><mrow><mi>k</mi><mo>−</mo><mi>ω</mi></mrow></math></span> SST RANS model of the experimental facility was developed and compared against the available experimental data.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114388"},"PeriodicalIF":2.1,"publicationDate":"2025-09-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097503","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 and transient analysis of a novel type passive residual heat removal system","authors":"Yuepeng Bi ,&nbsp;Genglei Xia ,&nbsp;Chenyang Wang ,&nbsp;Minjun Peng ,&nbsp;Yifan Xu ,&nbsp;Jiebo Wu","doi":"10.1016/j.nucengdes.2025.114446","DOIUrl":"10.1016/j.nucengdes.2025.114446","url":null,"abstract":"<div><div>This study introduces an innovative air–water mixed cooling passive residual heat removal (AWMPRHR) system designed to overcome the challenges associated with long-term core cooling under the constraints of weight and space in marine nuclear power plants (MNPPs). By utilizing sub-zero Celsius temperatures air and water tanks as heat sinks, the RELAP5 code was used to assesses the performance differences between series and parallel configurations of air–water-cooled heat exchangers under conditions of short-term constant decay heat power and long-term cooling. The findings reveal that during constant decay heat power, the parallel configuration offers superior heat removal efficiency. However, in long-term cooling scenarios, the performance disparities between the two configurations are minimal. Comparative analysis with water-cooled and air-cooled systems shows that the AWMPRHR system not only significantly reduces the stack height by 89.84 % but also enhances the coolant subcooling following an SBO accident. Conversely, the air-cooled configuration reduces the tank volume by 72.63 % and ensures long-term core cooling. The minimum coolant subcooling at the core outlet was 36.19 K. The AWMPRHR system demonstrates a more efficient capability for long-term cooling of the core with smaller heat sink requirements.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114446"},"PeriodicalIF":2.1,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097498","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":"Novel design of a multifunctional modular decay tank system with embedded particulate separation capability","authors":"Hongyu Chen,&nbsp;Nan Qian,&nbsp;Xingbo Han,&nbsp;Youshi Zeng,&nbsp;Wei Liu,&nbsp;Xinxin Chu","doi":"10.1016/j.nucengdes.2025.114474","DOIUrl":"10.1016/j.nucengdes.2025.114474","url":null,"abstract":"<div><div>This paper presents a novel, multifunctional modular decay tank system designed for thorium-based molten salt reactors (TMSRs) to enhance off-gas management by integrating particulate separation capabilities. TMSRs, while offering inherent safety advantages and improved neutron economy due to direct fission product release into molten salt, necessitate robust off-gas handling to mitigate operational challenges. Existing off-gas systems are susceptible to particulate clogging in downstream components like adsorption beds, risking system failures and increasing radiation exposure. This research addresses these limitations by proposing a decay tank that not only provides adequate residence time for short-lived isotope decay but also actively separates particulate matter from the gas stream, thereby reducing downstream purification burden and improving system reliability.</div><div>The proposed decay tank is a cubic enclosure with internal baffling that creates a complex, meandering flow path. This design prolongs off-gas residence time, with computational fluid dynamics (CFD) simulations confirming residence times approximating five hours for helium and argon, sufficient for significant radioactive decay. Crucially, the baffle geometry is engineered to induce low-velocity zones, promoting gravitational settling of particulate matter. Lagrangian multiphase modeling simulated the behavior of molten salt particles (1–10 μm) within the flow field. Results demonstrate that the system effectively separates and retains particles with diameters of 2.5 μm and larger, a critical capability for preventing clogging in subsequent filtration and adsorption stages.</div><div>Furthermore, the study investigated the tank’s thermal performance, revealing limited heat transfer capability when configured as a heat exchanger. However, the design strategically utilizes an internal cylindrical “tube-side” region to house essential downstream components such as coolers and charcoal beds. This re-purposing enables a modular off-gas treatment unit concept, where the decay tank acts as a consolidated housing for the entire system. This modular approach significantly simplifies integration, installation, and maintenance, reducing the need for individual component shielding and containment. The consolidated casing acts as a primary containment barrier, enhancing safety and reducing overall system complexity. The design facilitates pre-fabrication of modules, streamlining on-site assembly and reducing installation time and cost. This research pioneers a integrated approach to off-gas management for TMSRs, offering enhanced safety, improved reliability, and a streamlined design pathway for next-generation nuclear reactors. Future work will focus on optimizing sub-component designs and further evaluating separation efficiency for varying gas and particle compositions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114474"},"PeriodicalIF":2.1,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097499","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 comprehensive study on factors influencing radiation shielding concrete characteristics","authors":"Ashutosh Nanda ,&nbsp;Saubhagya Kumar Panigrahi","doi":"10.1016/j.nucengdes.2025.114466","DOIUrl":"10.1016/j.nucengdes.2025.114466","url":null,"abstract":"<div><div>Radiation shielding concrete (RSC) is a construction material against radiation exposure due to its effective shielding ability, adequate mechanical along with durability characteristics, cost-effectiveness, and abundant versatility. The radiation exposure is detrimental to both living anatomy and structural integrity, hence, the structural radiation shielding through RSC is essential. Being a heterogeneous material, several factors associated with different RSC constituents, like aggregate, binder, additive, and matrix composite, can influence its properties, thus limiting the usefulness of the most sensitive RSC structure, as even minor radiation penetration can have severe consequences for public health and safety. Hence, an attempt has been made to create a comprehensive database on the alternative materials in RSC formation. Literature for review was selected based on a rigorous keyword search through various databases such as ‘SCOPUS’, ‘Google Scholar’, ‘Science Direct’, etc. This manuscript highlights the RSC’s progress and development across the world and discusses the various RSC characteristics through different factors associated with the RSC constituents, influencing the radiation shielding ability of both hydrated RSC (HRSC) and geopolymerised RSC (GRSC) against gamma-rays and neutron particles. Through a map study of these factors and characteristics, the identified primary factors governing the realization of RSC, along with its property development, are high-density aggregate, fiber addition, and nanoparticle addition. The most preferred material constituents for RSC production, along with their optimal dosages, are identified. The maximum attainable RSC attenuation capacity against Gamma-ray and Neutrons using the mentioned most preferred material constituents is 0.30–––0.40 cm⁻¹ and 0.10–––0.20 cm⁻¹, respectively. A comparison is made between the materials utilized in a few prime reactors in India and the preferred constituent materials through this review work. In the event of meager accomplishments in research work concerning factors associated with RSC components, the current review can be treated as a platform for researchers in this field.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114466"},"PeriodicalIF":2.1,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097496","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":"Physical design of a burnup measurement device based on a D–D neutron source","authors":"YiNong Li ,&nbsp;Zheng Wei ,&nbsp;Jun Ma ,&nbsp;Kang Wu ,&nbsp;Chao Han ,&nbsp;XiaoXue Yu ,&nbsp;QiaoYue Jiang ,&nbsp;Ling Yuan ,&nbsp;PeiSheng Zhang ,&nbsp;ZhiYong Deng ,&nbsp;HaoYu Lei ,&nbsp;Yu Zhang ,&nbsp;JunRun Wang ,&nbsp;ZeEn Yao ,&nbsp;XiaoDong Su","doi":"10.1016/j.nucengdes.2025.114464","DOIUrl":"10.1016/j.nucengdes.2025.114464","url":null,"abstract":"<div><div>The nuclear fuel burnup measurement is an important part of the burnup credit, which are closely related to evaluating reactor performance, extending fuel utilization cycles, and achieving fuel breeding effects. In this work, an underwater spent fuel assembly burnup measurement device based on a deuterium–deuterium (D–D) neutron source has been designed. The D–D neutrons which are slowed down by the water, bombard the remaining fissile nuclides in the spent fuel to produce the fission neutron signals. The burnup of the spent fuel is determined by obtaining the signals from “U”-shaped detector array., The burnup data for AFA-3G fuel rods within the range of 5000 to 80,000 MWd/tU were calculated by the ORIGEN-ARP code, and the corresponding nuclear fuel are irradiated by D–D neutrons with the fission behavior and fission neutron signals, which are calculated by the Geant4 code. Based on the calculated results, the response relationship between the neutron-induced fission counts and the burnup of the AFA-3G assembly is established with a strong double exponential relationship, R² = 0.9999, for fuel within the burnup range of 5000–80,000 MWd/tU. Results from the simulation revealed that thermal neutrons constituted 88 % of the detected neutrons, and neutrons originating from assembly fission accounted for over 55 % of the detected signal. The designed burnup measurement device can directly measure the nuclear fuel burnup, which provides a technical solution for burnup credit.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114464"},"PeriodicalIF":2.1,"publicationDate":"2025-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145097500","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 and theoretical study on nonlinear vibration of a reduced-scale PWR fuel assembly with GTR sliding","authors":"Kefei Chen ,&nbsp;Jinming Li ,&nbsp;Jianhang Xu ,&nbsp;Peng Li ,&nbsp;Yiren Yang","doi":"10.1016/j.nucengdes.2025.114443","DOIUrl":"10.1016/j.nucengdes.2025.114443","url":null,"abstract":"<div><div>During the operation of a pressurized water reactor (PWR), the fuel assembly (FA) experiences axial flow-induced vibrations, leading to wear at the contact interfaces between the fuel rod (FR) cladding and the grids. Additionally, during seismic events or loss-of-coolant accidents (LOCAs), the FA may undergo severe deformation, potentially resulting in structural failures. Therefore, establishing a dynamic model of the FA to investigate its structural behavior under these conditions is essential for ensuring reactor safety. Experimental studies have demonstrated that FAs display nonlinear behavior under loading, including bilinear hysteresis in quasi-static experiments and stiffness softening in forced vibration experiments. Previous quasi-static studies have revealed that the stiffness softening and hysteresis behavior of FAs are primarily induced by relative axial sliding and rotational interactions at the grid-to-rod (GTR) joints. Building upon this foundation, the Euler–Bernoulli beam theory with bending-axial deformation coupling is employed to develop a theoretical nonlinear FA dynamic model that extends the previous quasi-static analysis to transient dynamic conditions. To further investigate the structural characteristics of the FA and provide insights for modeling, vibration experiments were conducted on a reduced-scale 3 × 3 FA with both ends rigidly fixed. The theoretical model successfully reproduced the structural response characteristics observed in both quasi-static and dynamic experiments by incorporating relative axial sliding and relative rotational sliding. The stiffness transition phase observed under quasi-static cyclic loading has been demonstrated to play a critical role in simulating the nonlinear dynamic response of the fuel assembly. Furthermore, using the developed nonlinear dynamic model, the GTR fretting (GTRF) wear characteristics of the FA are simulated and analyzed. The simulations identified wear patterns at different GTR joints, with the distribution of high-wear-rate GTR joints aligning with experimental observations. This study provides a reliable structural dynamic modeling framework for FA.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114443"},"PeriodicalIF":2.1,"publicationDate":"2025-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145060723","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":"Isogeometric analysis for flow-induced vibration of U tubes in steam generators","authors":"Junzhe Shen ,&nbsp;Pan Sun ,&nbsp;Jinxiong Zhou","doi":"10.1016/j.nucengdes.2025.114445","DOIUrl":"10.1016/j.nucengdes.2025.114445","url":null,"abstract":"<div><div>This paper presents the first reference on the use of the isogeometric analysis (IGA) for flow-induced vibration (FIV) of steam generator U tubes in pressurized water reactors (PWR). It describes in detail the IGA implementation of the recently proposed semi-analytical time-domain (SATD) model for tube bundles in cross-flow, offering an alternative option to the traditional finite element method (FEM) predominating in the field. Much fewer elements required for spatial discretization, faster convergence, and higher order of smoothness of solution, make IGA an ideal method for FIV of U tubes. A benchmark experiment reported in literature is chosen as the numerical example and a detailed study is performed for both linear and nonlinear dynamic response analysis. Good agreement is achieved between IGA prediction and reported experimental data. The method opens a new avenue for FIV of curved tubes, pipes, and cylinders with complicated forces and boundary conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114445"},"PeriodicalIF":2.1,"publicationDate":"2025-09-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145060722","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":"From the plant layouts to optimized LP and 3D PWR-KWU containment models for combustion risk assessment with GOTHIC 8.3(QA)","authors":"L. Serra ,&nbsp;A. Domínguez-Bugarín ,&nbsp;G. Jiménez ,&nbsp;C. Vázquez-Rodríguez ,&nbsp;M. Braun ,&nbsp;S. Kelm ,&nbsp;L.E. Herranz","doi":"10.1016/j.nucengdes.2025.114459","DOIUrl":"10.1016/j.nucengdes.2025.114459","url":null,"abstract":"<div><div>One of the key objectives of the severe accident management strategies is to preserve containment integrity and to prevent a large release of radioactive products into the environment. To evaluate containment response during a severe accident (SA), two GOTHIC 8.3(QA) models (LP and 3D) of a PWR-KWU containment have been developed in the framework of AMHYCO (EU-funded Horizon 2020 project). The LP and 3D models were compared for the in-vessel phase of a total loss of AC power scenario (SBO), with and without considering Passive Autocatalytic Recombiners (PARs). The two models showed consistent global trends, but the 3D model revealed local variations in hydrogen stratification, condensation, and temperature gradients that were not captured by the LP model. 3D results also highlighted the influence of 3D mesh resolution on stratification and flammability conditions, with finer meshes predicting different hydrogen accumulation flow patterns. As expected, PARs effectively reduced flammable volumes in both models, although 3D models yielded lower recombination rates due to local heterogeneities. Last, this study emphasizes the importance of the post-processing choices made by the user to identify safety relevant conditions with the potential to enhance accident management measures and the positioning of safety systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114459"},"PeriodicalIF":2.1,"publicationDate":"2025-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046865","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":"An investigation on machine learning predictive accuracy improvement and uncertainty reduction using VAE-based data augmentation","authors":"Farah Alsafadi,&nbsp;Mahmoud Yaseen,&nbsp;Xu Wu","doi":"10.1016/j.nucengdes.2025.114433","DOIUrl":"10.1016/j.nucengdes.2025.114433","url":null,"abstract":"<div><div>The confluence of ultrafast computers with large memory, rapid progress in Machine Learning (ML) algorithms, and the availability of large datasets place multiple engineering fields at the threshold of dramatic progress. However, a unique challenge in nuclear engineering is <em>data scarcity</em> because experimentation on nuclear systems is usually more expensive and time-consuming than most other disciplines. One potential way to resolve the data scarcity issue is deep generative learning, which uses certain ML models to learn the underlying distribution of existing data and generate synthetic samples that resemble the real data. In this way, one can significantly expand the dataset to train more accurate predictive ML models. In this study, our objective is to evaluate the effectiveness of data augmentation using variational autoencoder (VAE)-based deep generative models. We investigated whether the data augmentation leads to improved accuracy in the predictions of a deep neural network (DNN) model trained using the augmented data. Additionally, the DNN prediction uncertainties are quantified using Bayesian Neural Networks (BNN) and conformal prediction (CP) to assess the impact on predictive uncertainty reduction. To test the proposed methodology, we used TRACE simulations of steady-state void fraction data based on the NUPEC Boiling Water Reactor Full-size Fine-mesh Bundle Test (BFBT) benchmark. We found that augmenting the training dataset using VAEs has improved the DNN model’s predictive accuracy, improved the prediction confidence intervals, and reduced the prediction uncertainties.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"445 ","pages":"Article 114433"},"PeriodicalIF":2.1,"publicationDate":"2025-09-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"145046862","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}