Nuclear Engineering and Design最新文献

{"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}

{"title":"Effect of free falling condition on melt jet breakup length in partially flooded cavity","authors":"Woo Hyun Jung ,&nbsp;Hyun Sun Park","doi":"10.1016/j.nucengdes.2025.113923","DOIUrl":"10.1016/j.nucengdes.2025.113923","url":null,"abstract":"<div><div>When a severe accident occurs and proceeds to the failure of the reactor pressure vessel, the fuel–coolant interaction commonly occurs under the partially flooded cavity. The jet acceleration and the air entrainment phenomena, which were caused by the existence of the air space, resulted in different jet breakup lengths compared to the fully flooded cavity condition. Since the jet breakup length is an important parameter regarding the debris bed coolability, its precise prediction has been a significant issue for decades. This study investigated the effect of the free fall on the jet breakup length by observing the Froude number change and the air cavity generation according to the free fall height. Therefore, the Saito correlation was re-expressed as a function of the falling height emphasizing the influence of the jet acceleration in the air. Also, the size prediction model for the air cavity was developed as a function of the volume and the velocity of the jet head (bulge) based on the energy conservation between the bulge and the air cavity. By calculating the actual jet breakup length (the breakup length only from a melt-water interaction) and comparing it with the literature data, the influence of the jet acceleration and the air entrainment was highlighted indicating the necessity of the elaborate observation of the melt jet at the air in the fuel–coolant interaction experiments.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113923"},"PeriodicalIF":1.9,"publicationDate":"2025-02-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480255","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":"Test program and numerical studies on reinforced concrete slabs impacted by liquid-filled missiles","authors":"Christian Heckötter ,&nbsp;Ari Vepsä ,&nbsp;Jürgen Sievers","doi":"10.1016/j.nucengdes.2025.113891","DOIUrl":"10.1016/j.nucengdes.2025.113891","url":null,"abstract":"<div><div>Reinforced concrete (rc)-structures are used to protect vital parts of nuclear facilities against external hazards. These include malevolent or accidental airplane crash (APC). In this context kerosene mass may significantly contribute to loading as well as structural damage of the target rc-structure. Past research has been limited regarding this issue and few experimental data were available. Therefore, a test series called L-series dealing with impact of water-filled missiles on rc-slabs was carried out in the framework of Phase III of the multinational research project IMPACT at Technical Research Centre of Finland (VTT). In additional tests impact forces were recorded with a force plate system. Test results are used to validate numerical models for LS-DYNA and ANSYS AUTODYN software.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113891"},"PeriodicalIF":1.9,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143474389","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 multiphysics tightly coupled model for a lead cooled fast reactor concept and application to statistical calculation of hot channel factors","authors":"Y. Yu ,&nbsp;H. Park ,&nbsp;A. Novak ,&nbsp;E. Shemon","doi":"10.1016/j.nucengdes.2025.113915","DOIUrl":"10.1016/j.nucengdes.2025.113915","url":null,"abstract":"<div><div>A tightly coupled multiphysics code system is established using the MOOSE framework for hot channel factor (HCF) evaluation on a Lead Fast Reactor (LFR) concept. The coupled system is driven by the Griffin multiphysics coupling capability under which the MOOSE Heat Transfer module and NekRS computational fluid dynamics solver are coupled for conjugate heat transfer using the Cardinal application. The coupled capability is demonstrated on an LFR assembly model based on materials and geometry of a prototypical lead-cooled fast reactor design by Westinghouse Electric Company, LLC. Moreover, the work integrates the Multiphysics Object Oriented Simulation Environment (MOOSE) Stochastic Tools Module (STM) to perform calculations for statistical analysis of HCF. The coupling strategy and workflow demonstrated in this paper is not only useful for predicting accurate hot channel factors for different kinds of advanced reactors but also for other engineering applications such as control rod worth assessment, generation of high-fidelity database for Artificial intelligence (AI)/machine learning (ML) training, design optimization and multi-resolution modeling.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113915"},"PeriodicalIF":1.9,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480254","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 study on boundary-conditioned hydration impact on bentonite gas permeability with supporting modeling","authors":"Hongyang Ni ,&nbsp;Jiangfeng Liu ,&nbsp;Qi Zhang ,&nbsp;Zhipeng Wang ,&nbsp;Zihao Zhang","doi":"10.1016/j.nucengdes.2025.113940","DOIUrl":"10.1016/j.nucengdes.2025.113940","url":null,"abstract":"<div><div>Deep geological disposal is currently a widely accepted disposal option for high-level radioactive waste, with bentonite serving as a crucial buffer/backfill material due to its self-sealing properties. Upon groundwater infiltration, the bentonite experiences constrained swelling due to surrounding rock confinement, influencing its hydro-mechanical behavior. While extensive research has explored the relationship between hydration and gas permeability under free-swelling conditions, the impact of constrained swelling remains insufficiently understood. In the present study, these properties under both free and constrained swelling conditions are investigated. The findings reveal that constrained swelling conditions hinder the hydration process, leading to reduced porosity and increased water saturation compared to free-swelling conditions. The temporal variation in water content under both boundary conditions can be described by an adsorption-diffusion equation with a strong correlation (coefficients of determination, R² &gt; 0.97, and root mean squared error (RMSE) ≤ 0.0048). Moreover, the van Genuchten (VG) model accurately captures water saturation trends with R² = 0.99 and RMSE = 0.012 for free swelling, and R² = 0.98 and RMSE = 0.013 for constrained swelling). Despite having comparable water content, constrained swelling significantly reduces gas permeability. Additionally, a higher initial equilibrium relative humidity (<em>RH</em>) corresponds to a larger subsequent gas permeability gap, reaching 2.01 × 10⁻¹⁶ m² at 75 % <em>RH</em> and 5.72 × 10⁻¹⁶ m² at 98 % <em>RH</em>. Void ratio and water saturation, two critical parameters influencing gas permeability, undergo substantial changes during the hydration process. Specifically, changes in the available void ratio due to hydration affect gas permeability. By quantifying post-hydration equilibrium changes in void ratio and saturation, a gas permeability model accounting for different boundary conditions is developed, which demonstrates a relatively effective agreement with experimental data (R² = 0.75). This study addresses the knowledge gap on how boundary effects of bentonite hydration influence gas permeability, offering key insights into its long-term performance as a barrier material in nuclear waste repositories.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113940"},"PeriodicalIF":1.9,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143480168","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":"Summary of researches on operational characteristics and safety of molten salt fast reactors based on neutronics and thermal-hydraulics coupling analysis","authors":"Hiroyasu Mochizuki","doi":"10.1016/j.nucengdes.2025.113941","DOIUrl":"10.1016/j.nucengdes.2025.113941","url":null,"abstract":"<div><div>The present paper mainly summarizes the operational and safety characteristics found by the neutronics and thermal-hydraulics coupling analysis for molten-salt fast reactors (MFR). Analysis methods for MFRs have been developed using system codes and Computational Fluid Dynamics (CFD) codes. Traditionally, the analysis has been limited to single-phase flow of the molten fuel salt, but recently analysis has been conducted under two-phase flow conditions with a small amount of helium gas injected to remove fission products (FP). In the MSFR developed by the EU, the void fraction due to two-phase flow is locally distributed, and a method for analyzing this by coupling advanced neutronics calculation methods with a CFD code has been proposed. On the other hand, a method that models all heat transport systems using a system code with reactor point kinetic (PRK) models have also been proposed. In this case, a CFD code is also used to calculate the precise behavior of flow scheme in the reactor. Since MFRs generally do not have control rods, the method of starting up the reactor must be different from that of light water reactors (LWR). Various methods have been proposed, but this paper introduces a startup method that takes advantage of the characteristics of two-phase flow and its negative reactivity. In the analysis of load-following operation, several methods have been proposed to perform time-order load-following operation by actively varying the reactor power, as in conventional LWRs. Recently, a passive load-following operation method has been proposed that fully exploits the characteristics of MFR. In this passive method, the reactor is operated to respond naturally to temperature changes in the heat transport system caused by flow rate changes corresponding to load variations in the energy conversion system, without controlling the fuel and intermediate circuits. This operation method introduces a heat storage tank in the intermediate circuit, which allows load-following operation in all modes from long to short duration. With regard to safety, the paper presents examples of system code and CFD code analysis of the behavior with a safety protection system bypassed for various transients in two-phase flow conditions. These analyses show that molten salt fast reactors always transition to the safe operating range after transients.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"435 ","pages":"Article 113941"},"PeriodicalIF":1.9,"publicationDate":"2025-02-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143474390","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}