Nuclear Engineering and Design最新文献

{"title":"Enhancing uncertainty analysis: POD-DNNs for reduced order modeling of neutronic transient behavior","authors":"Yijun Zhang ,&nbsp;Wenhuai Li ,&nbsp;Sitao Peng ,&nbsp;Jinggang Li ,&nbsp;Ting Wang ,&nbsp;Qingyun He ,&nbsp;Tao Wang ,&nbsp;Haoliang Lu ,&nbsp;Ling Zeng","doi":"10.1016/j.nucengdes.2025.113969","DOIUrl":"10.1016/j.nucengdes.2025.113969","url":null,"abstract":"<div><div>In reactor safety analysis, sensitivity analyses on critical parameters are essential for ensuring the reliability of safety conclusions, particularly regarding transient behavior, which often requires time-consuming computations. Developing surrogate models presents a promising solution. This paper extends the Proper Orthogonal Decomposition-Radial Basis Function (POD-RBF) framework to the 3D Light Water Reactor Core Transient Benchmark (3DLWRCT) for control rod ejection accidents. The primary aim is to simulate transient behavior under random perturbations in the macroscopic neutronic cross-sections of fuel assemblies.</div><div>Our results indicate that the traditional POD-RBF approach struggles to accurately reconstruct the highly nonlinear transient system, whether through spatiotemporal folding or spatial data reduction. To overcome these challenges, we enhance the model by integrating Deep Neural Networks (DNNs) and employing the Tree-structured Parzen Estimator for optimal neural network architecture selection. This improved approach significantly increases the accuracy of the surrogate models, demonstrating its feasibility and effectiveness. The integration of DNNs offers a deeper understanding of complex interactions within the reactor core, effectively capturing nonlinearities and yielding reliable predictions even under uncertainty.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113969"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143520205","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":"VVER long-term operation – A review based on the material studies results from past and ongoing EU-supported research projects","authors":"Vladimir Slugen ,&nbsp;Jana Simeg Veternikova ,&nbsp;Maria Domankova ,&nbsp;Matus Gavalec ,&nbsp;Jana Petzova ,&nbsp;David Slnek ,&nbsp;Mykola Dzubinsky ,&nbsp;Oleksii Shugaylo ,&nbsp;Ildiko Szenthe ,&nbsp;Ferenc Gillemot ,&nbsp;Maksym Zarazovskii ,&nbsp;Jari Lydman ,&nbsp;Radim Kopriva ,&nbsp;Szabolcs Szavai ,&nbsp;Ondrej Srba","doi":"10.1016/j.nucengdes.2025.113949","DOIUrl":"10.1016/j.nucengdes.2025.113949","url":null,"abstract":"<div><div>Safe and long-term operation (LTO) of nuclear power plants (NPP) with Water-Water Energetic Reactors (VVER) is essential for several central and eastern European countries to keep their energy supply security. During the last decades, several EU-supported projects in framework projects or Horizon2020 schemes have focused on the material degradation of pressurised water reactors (PWRs), especially VVERs. This paper aims to extract the most important results dominantly from EURATOM projects which support or could limit the long-term operation of this Soviet design of PWRs. The accent is given on ongoing projects such as FRACTESUS, STRUMAT-LTO, APAL, and ENTENTE. Actually, running project DELISA-LTO (2022–2026) comprises not only results from previous research in this area but also incorporates wide material databases coming from nuclear power plant V-1 in Jaslovské Bohunice (Slovakia), which was shut down after 28 years of operation and provides wide-range of original service-aged materials specimens for profound material studies having in mind lifetime extension of VVERs. Consolidated knowledge focused mainly on the reactor pressure vessel (RPV) material studies is summarised and supported by relevant references.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113949"},"PeriodicalIF":1.9,"publicationDate":"2025-03-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143526997","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":"[NURETH-20] Evaluation on mitigation performance of flooding safety system under hypothetical loss of coolant accident in Korean i-SMR with MELCOR code","authors":"Chang Hyun Song ,&nbsp;Jae Hyung Park ,&nbsp;JinHo Song ,&nbsp;Sung Joong Kim","doi":"10.1016/j.nucengdes.2025.113950","DOIUrl":"10.1016/j.nucengdes.2025.113950","url":null,"abstract":"<div><div>In response to the growing interest in Small Modular Reactors (SMRs) globally, many countries are actively pursuing the development of high-power SMRs, based on their inherent advantages including enhanced safety, grid flexibility, and potential for hydrogen production. Among these endeavors, an i-SMR with an electrical power output of 170 MWe has been under development since 2021 by Korea Hydro &amp; Nuclear Power Co., Ltd. in Republic of Korea. The i-SMR has established ambitious top-tier requirements, such as core damage frequency less than 1.0 × 10⁻⁹/module-year and large early release frequency less than 1.0 × 10⁻¹⁰/module-year, and emergency planning zone within nuclear power plant site boundary. All of which is extremely challenging and necessitates innovative safety systems with exceptional reliability. In this context, this study proposed a Flooding Safety System (FSS) as a novel safety system, and its mitigating performance under a hypothetical accident scenario was evaluated by using MELCOR code for validating the efficacy of this conceptual approach. To conduct the accident analysis, a MELCOR input model for the i-SMR was developed. The initiating event assumed was the stuck open of depressurization valve, leading to the discharge of coolant from the primary system into the metal containment vessel, which can be deemed as a loss of coolant accident. The findings in this study revealed that timely operation of the FSS can prevent core damage, thus validating its crucial role to assuring the integrity and reliability of the i-SMR.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113950"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510666","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":"Impact of hydrogen on iPWR containment Thermal-Hydraulics in severe accidents","authors":"Luke Lebel,&nbsp;Dening Eric Jia,&nbsp;David Grand-Maitre,&nbsp;Tony Clouthier","doi":"10.1016/j.nucengdes.2025.113953","DOIUrl":"10.1016/j.nucengdes.2025.113953","url":null,"abstract":"<div><div>Addressing hydrogen issues is an important part of severe accident management for nuclear accidents, and this is doubly important for water-cooled small modular reactors (SMRs), like integral pressurized water reactors (iPWRs). This study uses a combination of containment thermal–hydraulic experiments and modeling to assess the impact of introducing light non-condensable gases under prototypic accident conditions. The experiments evaluated the gas mixing behaviour and steam condensation heat transfer coefficients in the tertiary steam–air-helium system, extending the unified steam condensation correlations by <span><span>Dehbi (2016)</span></span> past the density difference transition point. The modeling with the GOTHIC thermal-hydraulics code scaled the experimental observations to reactor-relevant accident scenarios, with steam-hydrogen in a closed containment and steam–air-hydrogen in a loss of containment isolation situation. A strong reduction in the heat transfer capacity caused by the presence of non-condensable gases was observed, which disrupt the steam condensation that drives many of the passive cooling loops that form the basis of accident mitigation strategies. Likewise, the observed gas mixing behaviour inferred that hydrogen combustion may be a further issue that needs to be addressed in severe accident management guidelines.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113953"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510665","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":"Prediction of radiation shielding design schemes based on adaptive neural networks","authors":"Qisheng Chen ,&nbsp;Zi-Hui Yang ,&nbsp;Zhong-Yang Li ,&nbsp;Guo-Min Sun ,&nbsp;Shi-Peng Wang ,&nbsp;Yu-Chen Li ,&nbsp;Zhi-Xing Gu ,&nbsp;Fei Li ,&nbsp;Juan Fu ,&nbsp;Gui-Hua Tao","doi":"10.1016/j.nucengdes.2025.113933","DOIUrl":"10.1016/j.nucengdes.2025.113933","url":null,"abstract":"<div><div>As space exploration and space reactor technology continue to advance, radiation shielding design faces numerous challenges, such as space limitations, weight constraints, and the complexity of shielding materials. Traditional design methods typically rely on empirical models validated through Monte Carlo simulations, but these approaches often fail to achieve optimal results. To enhance the radiation protection efficiency of space reactors, this paper proposes a deep neural network model based on the self-attention mechanism to assist in predicting radiation shielding design schemes and to verify its accuracy and practicality. We used SuperMC software to record the relevant safety parameters for the Kilopower reactor under 10,000 different shielding design schemes and calculated the total mass and total radiation dose for these designs, creating a comprehensive dataset. The total mass and radiation dose were used as inputs to the neural network, which then generated the corresponding radiation shielding design schemes. Experimental results show that the model demonstrates high accuracy and strong interference resistance, with the error in total radiation dose and material mass consistently controlled around 3%. Additionally, by combining simulation methods with the self-attention mechanism, the model effectively generates radiation shielding designs suitable for space reactors, providing reliable protection solutions for future space missions. This approach also opens new possibilities for radiation shielding design in other fields.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113933"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143511332","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":"Comprehensive considerations for the co-decontamination and recycling of radioactively contaminated steels","authors":"Mu Zhao ,&nbsp;Yunfei Hua ,&nbsp;Xuzhi Dai ,&nbsp;Wenming Qin ,&nbsp;Xin Wei ,&nbsp;Lin Zhong ,&nbsp;Xuan Zhao","doi":"10.1016/j.nucengdes.2025.113945","DOIUrl":"10.1016/j.nucengdes.2025.113945","url":null,"abstract":"<div><div>This paper thoroughly introduces the one-stop decontamination and reuse process specific to radioactively contaminated steel, highlighting its key characteristics. By leveraging the data from engineering practice, we conduct a detailed analysis of the effectiveness of various decontamination techniques, including crystalline phase temperature difference gradient decontamination, strippable film decontamination, as well as a novel decontamination agent and its corresponding process. Additionally, we explore the properties of the steel products obtained after melting. It is concluded that this one-stop decontamination and reuse process, supported by advanced technologies, realizes the recycling of radioactively contaminated steel.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113945"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510664","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":"Fault detection in thermocouples: Unveiling anomalies with machine learning and signal processing","authors":"Valipi Dinesh Kumar ,&nbsp;Anindya Bhattacharyya ,&nbsp;Rajendra Prasad Behera ,&nbsp;K. Prabakar","doi":"10.1016/j.nucengdes.2025.113955","DOIUrl":"10.1016/j.nucengdes.2025.113955","url":null,"abstract":"<div><div>Reliable data acquisition from installed sensors is crucial for ensuring operational efficiency and safety in industrial settings. Early detection of sensor anomalies is particularly vital in high-integrity applications such as avionics, nuclear reactors, and associated fuel recycling plants, where data reliability directly impacts process and personnel safety. Thermocouples (TCs) are commonly used in critical temperature measurement applications due to their robustness and long history of dependable performance. This paper proposes a data-driven technique to detect TC sheath failure as part of the Operator Support System (OSS) in operating plants, alarm generation, decision support, and predictive maintenance. Additionally, an accelerated aging setup is proposed to simulate sheath failure in TCs and assess its impact on performance characteristics in a controlled environment mimicking the dissolver stage of the Plutonium Uranium Reduction Extraction (PUREX) process in nuclear fuel reprocessing. Our in-situ failure detection approach introduces an application of Empirical Mode Decomposition (EMD) as a data-driven technique to extract sensor noise from true measurement data. The statistical features of the extracted noise signal are then combined with machine learning (ML) based decision-making for early sheath failure detection. This approach is specifically designed for in-situ detection of sheath failure, a primary cause of TC malfunction in corrosive environments. The effectiveness of the proposed method is demonstrated using experimental data from accelerated testing of faulty TCs in a controlled environment. Results show that K-Nearest Neighbor (KNN) and Random Forest (RF) classifiers achieved over 96% classification accuracy under all experimental conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113955"},"PeriodicalIF":1.9,"publicationDate":"2025-02-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510028","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":"Structural design and optimization of two segmented asymmetrical thermoelectric generator for heat pipe cooled reactor application","authors":"Xirui Huang ,&nbsp;Lei Tan ,&nbsp;Huangshiyi Lin ,&nbsp;Xinwen Zhang ,&nbsp;Simiao Tang","doi":"10.1016/j.nucengdes.2025.113948","DOIUrl":"10.1016/j.nucengdes.2025.113948","url":null,"abstract":"<div><div>As deep space and deep sea exploration continues, the requirements for energy systems are constantly increasing. Small reactors, with their high endurance, high reliability and high energy density, are ideal choices. The heat pipe reactor, as a small nuclear reactor with great potential, has received widespread attention from the academic community. Thermoelectric generators (TEG) are commonly used thermalto-electric energy conversion devices in heat pipe reactors and play an important role in static energy conversion. This study employs COMSOL Multiphysics software to conduct finite element simulation analysis on TEG, comprehensively considering factors such as contact resistance, contact thermal resistance, and external resistance that affect TEG performance. The relationship between the leg shape and the thermoelectric power generation performance of two segmented inverted circular truncated cone and circular-X TEG are analyzed and the key factors affecting their thermoelectric conversion performance are summarized. In terms of stress analysis, both types can shift the stress from hot end to cold end, to adapt to the welding process. For two segmented inverted circular truncated cone TEG, when the ratio of the max and min cross-sections is 0.83, it can cause a decrease in thermal stress on the interfaces with few efficiency decrease. For two segmented circular-X TEG, their efficiency increases monotonically as the minimum cross-sectional radius decreases. When the ratio of the max and min cross-sections is 0.83, both the conversion efficiency and the stability of the working state are improved.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113948"},"PeriodicalIF":1.9,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143488801","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":"BWR core thermal–hydraulic uncertainty and sensitivity analysis with improved bypass modeling features","authors":"Devshibhai Ziyad ,&nbsp;Agustin Abarca ,&nbsp;Maria Avramova","doi":"10.1016/j.nucengdes.2025.113951","DOIUrl":"10.1016/j.nucengdes.2025.113951","url":null,"abstract":"<div><div>Boiling Water Reactors (BWR) designs have several special features to be considered while analyzing their thermal–hydraulic performance. One of those are the core regions where the coolant is restricted by physical boundaries of coming in contact with fuel rods. These bypass regions can be classified into core bypass (in the core periphery), bundle bypass (between the assemblies), and internal assembly bypass (water rods/channels). Subchannel thermal–hydraulic analysis usually simplifies the modeling of these regions by not accounting for heat transfer to the bypass coolant flow, aiming to be conservative in predicting safety margins to acceptance criteria.</div><div>Since the nuclear industry is embracing economically efficient Best Estimate (BE) simulation methodologies in place of the conservative methodologies, there is a heightened emphasis on the advancements in modeling the BWR core bypass regions in subchannel thermal–hydraulic analyses. Ziyad et al. (2022) have improved the advanced sub-channel code CTF by developing and implementing models for bypass related phenomena in BWRs. For the application of the code in BE analysis, rigorous uncertainty quantification becomes necessary. This involves propagating inherent uncertainties in model inputs for newly developed bypass modeling features in addition to the traditional model inputs. This propagation is important for accurately quantifying uncertainties in System Response Quantities (SRQs) which informs the safety margins and hence has economic incentive.</div><div>In this research, uncertainties in the input parameters are propagated in steady-state simulations through an assembly-resolved full core model and a subchannel-resolved single fuel assembly model of the Peach Bottom Unit 1 at End of Cycle 2. The statistical analysis tool Dakota is used as a driver for CTF, and it is employed for conducting the uncertainty propagation. Random Monte Carlo sampling techniques are utilized for input preparation, while the Spearman correlation metric is employed for sensitivity analysis.</div><div>The sensitivity analysis of the full core model indicates that bypass flow fraction is a strong function of the void fraction in the active core region. This phenomenon is only possible to be captured by employing the bypass modeling which employs pressure equalization across all subchannels. The void fraction prediction is also affected by other bypass modeling features as established by Ziyad et. al. (2022), hence each of the developed features finds its importance in the analysis. It also has been found that isolated single assembly modeling is inadequate to predict thermal–hydraulic conditions in bypass regions as lateral flow between the assembly gaps cannot be captured.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113951"},"PeriodicalIF":1.9,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143488800","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 study on PCHE channels for lead–bismuth eutectic and supercritical carbon dioxide coupled flow and heat transfer","authors":"Xiangyu Fu ,&nbsp;Siwei Cai ,&nbsp;Guodong Qiu ,&nbsp;Jianchuang Sun ,&nbsp;Bin Zhao ,&nbsp;Qian Li ,&nbsp;Weihua Cai","doi":"10.1016/j.nucengdes.2025.113942","DOIUrl":"10.1016/j.nucengdes.2025.113942","url":null,"abstract":"<div><div>The intermediate heat exchanger (IHX) is a key component in lead-cooled fast reactors (LFRs) between the lead–bismuth eutectic (LBE) loop and supercritical carbon dioxide (sCO₂) Brayton cycle. In this paper, the LBE- sCO₂ flow and heat transfer in printed circuit heat exchanger channels is studied. Firstly, for a typical identical semicircular channel on the hot and cold side, the best hot and cold flow rate ratio in the constraint of design temperature and pressure drop requirements was explored. On this basis, the flow and heat transfer characteristics of different channel patterns were comparatively analyzed. The “rectangle-to-two” channel pattern has the largest heat transfer in per unit volume, which is 176 % greater than the typical identical semicircular channels. Finally, the correlation equations of Nusselt number and Fanning friction factor were established. It provides a basis for the flow and heat transfer calculations and structural optimization design of the LFRs.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113942"},"PeriodicalIF":1.9,"publicationDate":"2025-02-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143510621","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}