Nuclear Engineering and Design最新文献

{"title":"Study on scaling analysis methods based on Lead-Bismuth loop with natural circulation","authors":"Jianing Xu ,&nbsp;Zhen Wang ,&nbsp;Rui Pan ,&nbsp;Shichao Zhang ,&nbsp;Qiusun Zeng ,&nbsp;Jie Yu","doi":"10.1016/j.nucengdes.2025.114346","DOIUrl":"10.1016/j.nucengdes.2025.114346","url":null,"abstract":"<div><div>The natural circulation characteristics of the Lead-Bismuth Eutectic (LBE) enhance the passive safety of the lead–bismuth fast reactors. The scaled-down experimental facilities play an important role in further evaluating the natural circulation of lead–bismuth. Traditional studies of the scaling methods are carried out for water-cooled reactors. To provide the theory for the design of the scaled experimental facilities and further verify the traditional scaling methods for LBE flow, multiple methods such as H2TS and DSS were applied to an LBE loop in this study. A simplified single-phase natural circulation case was established using the Relap5 code and the corresponding criteria were obtained based on the different scaling methods. The study analyzed the steady and transient cases and the characteristics of several methods were compared. The results indicate that the H2TS and DSS methods can accurately simulate the steady cases of natural circulation in the LBE loops. For the dynamic processes with power changes, the DSS method can simulate the initial stage of the dynamic process more accurately than the H2TS method when the length ratio is 0.25. As the length ratio increases, the difference between several methods decreases and the dynamic deviation of all scaled-down cases compared to the prototype is reduced. In addition, when the power changes nonlinearly, the model obtained by the DSS method is more consistent with the prototype case than the H2TS method.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114346"},"PeriodicalIF":2.1,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144722926","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":"Statistical techno-commercial analysis of nuclear cogeneration projects: The case of potable water production by nuclear desalination of seawater using small modular reactors (SMRs)","authors":"Rupsha Bhattacharyya","doi":"10.1016/j.nucengdes.2025.114358","DOIUrl":"10.1016/j.nucengdes.2025.114358","url":null,"abstract":"<div><div>Small modular nuclear reactors (SMRs) deployed for nuclear cogeneration represent an opportunity for extending the contribution of nuclear energy beyond low emissions intensity electricity supply alone, thus addressing the global clean energy transition, climate change mitigation and adaptation programs. The thermal and electrical energy produced by these reactors can be used for desalination of brackish or seawater to produce potable water, thereby avoiding carbon emissions from use of fossil fuel derived heat and electric power and alleviating water stress in many regions. In this study, a statistical approach based on Monte Carlo simulations using input parameter distributions derived from literature data is used to study the techno-commercial features of SMR based nuclear desalination plants. The key metric of interest is the calculated levelized cost of freshwater production. It is found that the best case levelized cost of electricity from SMRs is expected to be $ 67–119/MWh(e) for the initial or first of its kind SMRs with power of about 300 MW(e) and expected to be deployed in the 2030 s, decreasing to $ 33–84/MWh(e) by 2050 for the n^th of its kind SMRs to be available in the 2050 s. This translates into corresponding expected potable water costs of $ 0.81–1.31/m³ from reverse osmosis plants and $ 1.09–2.62/m³ from multi-effect distillation plants. The optimistic values are found to be broadly comparable with nuclear desalination using current generation large reactors.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114358"},"PeriodicalIF":2.1,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144723467","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental investigation on Ni foam-supported Pt-Pd catalyst of a passive catalytic recombiner for hydrogen risk mitigation","authors":"Jushang Zhang ,&nbsp;Qihui Zhao ,&nbsp;Tianming Man ,&nbsp;Yunhe Zhao ,&nbsp;Zehua Guo ,&nbsp;Ming Ding","doi":"10.1016/j.nucengdes.2025.114355","DOIUrl":"10.1016/j.nucengdes.2025.114355","url":null,"abstract":"<div><div>During a severe accident in nuclear power plant, hydrogen explosion is one of the main reasons for the failure of nuclear power plant containment. The passive autocatalytic recombiners (PARs) are considered H₂ central system for emergency gas removal. In this study, we prepared two Ni foam-supported Pt-Pd catalysts: the pine needle-like Pt-Pd nano-dendrites catalysts by the impregnation method and the Pt-Pd nanosheet catalysts by the electrodeposition method. The foam metal, specifically Ni foam, can serve as an effective catalytic support which provides more reaction sites with its exceptional porosity structure. We tested their catalytic performance by evaluating their surface morphology, hydrogen conversion, heat distribution, and stability. The catalytic reaction was performed in a flow channel which simulates the working condition of PARs under varying flow velocity, inlet temperatures, and hydrogen concentrations. At the flow velocity of 0.17 m/s and the hydrogen concentration ranging from 1 % to 4 %, the average hydrogen conversion rate for catalyst (impregnation method) is 20.98 %. This is 2.6 % higher than that of catalyst (electrodeposition method). Furthermore, the highest conversion peak of 51.96 % was noted at the hydrogen concentration of 4 % and the flow velocity of 0.05 m/s. The maximum temperatures on the catalyst surface ranged between 236.2 and 366.2 °C during the reaction, and this could be sustained at a hydrogen flow velocity of 0.17 m/s for several hours. These catalysts are expected to solve the shortcomings of current PAR catalysts in terms of hydrogen conversion and hotspot elimination.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114355"},"PeriodicalIF":2.1,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144723471","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":"Advancing inherent safety in small modular reactors: A CFD-based analysis of the PIUS concept","authors":"Ju Hun Jung,&nbsp;In Cheol Bang","doi":"10.1016/j.nucengdes.2025.114343","DOIUrl":"10.1016/j.nucengdes.2025.114343","url":null,"abstract":"<div><div>Inherent safety for advanced small modular reactors has been significantly considered as one of the major targeted design features for the reliability and safety of nuclear reactors. As one of the candidates, the Process Inherent Ultimate Safety (PIUS) is noteworthy to enhance reactor safety representing accident mitigation performance without core meltdown as well as power excursion in emergency scenarios. These kinds of reactors can ensure the integrity of the nuclear core, components, and systems with the application of the PIUS concept. The analytical methodology was conducted to analyze the thermal–hydraulic behavior of the PIUS concept for the SMR scale by modeling a separate effect test loop with a computing method, such as a CFD tool, in this work. The model included key components, density locks, in which the thermal stratification phenomenon is actively generated. In normal operation, the density locks blocked the cold borated water flowing into the core region like valve functions. In transient cases, thermal stratification passively collapses since the momentum of the primary coolant dissipates or exceeds due to the coolant pump conditions. This passive injection of cold water triggers the reactor shutdown, and the primary system can be cooled by a long-term cooling process with global temperature behavior by natural circulation. The results indicate the feasibility of the PIUS concept with light-water-cooled technology for advanced SMR designs. This work is expected to provide research grounds for the ultimately safe SMR design with innovative safety technology.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114343"},"PeriodicalIF":1.9,"publicationDate":"2025-07-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144713347","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":"Simulation of UNCL experimental device for the reactor fuels based on reactor Monte Carlo code RMC","authors":"Yuanhao Gou ,&nbsp;Conglong Jia ,&nbsp;Dacai Zhang ,&nbsp;Zhechuan Tan ,&nbsp;Kan Wang","doi":"10.1016/j.nucengdes.2025.114316","DOIUrl":"10.1016/j.nucengdes.2025.114316","url":null,"abstract":"<div><div>The neutron multiplicity counting method is a non-destructive testing (NDT) technique that analyzes the properties of materials by measuring neutron emission events from nuclear materials. This method does not require the destruction or alteration of the material itself, making it widely used in the detection and safety management of nuclear materials. This article is based on the self-developed Reactor Monte Carlo code RMC to perform calculations and analysis of the UNCL(uranium neutron coincidence collar) series devices. First, calculations were performed on the UNCL device to compute neutron multiplicity countings under Californium-252 and AmLi neutron sources, and the results were compared with the simulation results of MCNP. Next, modeling calculations were conducted on the UNCL-II device, computing the neutron multiplicity countings under the same neutron sources, while considering the fuel assemblies containing Gd in the UNCL-II. The results were also compared with experimental results and the simulation results of MCNP. The neutron multiplicity count rate calculated by RMC is in good agreement with the experimental values and MCNP results, which proves the correctness of the calculated results by RMC. In addition, the effects of different fitting curves on the mass inversion results and their uncertainties are compared. At the same time, the impact of uncertainties in the polyethylene density on the results was also analyzed. The calculations show that the uncertainty in the polyethylene density has a non-negligible effect on the neutron multiplicity count rate and mass inversion results.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114316"},"PeriodicalIF":1.9,"publicationDate":"2025-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144703305","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":"Study on safety of advanced integrated natural circulation reactor of NHR200-II under LOCA condition","authors":"Yan Wang ,&nbsp;Zixuan Wang ,&nbsp;Yifan Meng ,&nbsp;Heng Xie","doi":"10.1016/j.nucengdes.2025.114348","DOIUrl":"10.1016/j.nucengdes.2025.114348","url":null,"abstract":"<div><div>The 200 MW_th nuclear heating reactor (NHR200-II) is a compact modular reactor with exceptional passive safety, making it an economical and dependable energy source for combined heat and power applications in urban and industrial settings. NHR200-II incorporates several advanced design features, including integral arrangement, natural circulation, self-pressurization, an innovative hydraulic control rod drive, and a passive residual heat removal system. Due to the natural circulation and passive safety features of NHR-200 II, the thermal–hydraulic transient response of the reactor system during a postulated Loss of Coolant Accident (LOCA) differs significantly from that of a conventional pressurized water reactor. This distinct response has a profound impact on the containment design and the overall safety assessment of the reactor. In this study, several typical LOCA scenarios for NHR200-II are analyzed using PCNHR—a validated transient analysis code developed by the Institute of Nuclear and New Energy Technology, Tsinghua University. The results indicate that decay heat can be effectively removed, ensuring the safety of the reactor system. Moreover, even under the most severe postulated LOCA conditions, the reactor core remains submerged in residual coolant within the reactor pressure vessel, obviating the need for special emergency core cooling system. These findings confirm the excellent safety features of the NHR200-II design.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114348"},"PeriodicalIF":1.9,"publicationDate":"2025-07-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144703306","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":"Some CEA research activities in support of Fukushima Daiichi fuel debris retrieval","authors":"Christophe Journeau ,&nbsp;Viviane Bouyer ,&nbsp;Arthur Denoix ,&nbsp;Andrea Bachrata ,&nbsp;Laurent Brissonneau ,&nbsp;Ioana Doyen ,&nbsp;Célia Guévar ,&nbsp;Joël Faure ,&nbsp;Emmanuelle Brackx","doi":"10.1016/j.nucengdes.2025.114333","DOIUrl":"10.1016/j.nucengdes.2025.114333","url":null,"abstract":"<div><div>Fuel debris retrieval is one of the important challenges in view of Fukushima Daiichi decommissioning. Indeed, hundreds of tons of fuel debris will have to be cut and collected in the 3 units that have been subject to core meltdown in March 2011. A large R&amp;D effort has been supported by the Japanese stakeholders in support of this retrieval. In this context, CEA has used its severe accident and decommissioning research expertise to launch research activities in support of fuel debris retrieval. A first series of works aims to acquire knowledge about these fuel debris: experiments have been carried out to simulate Molten Core Concrete Interaction or to fabricate fuel debris simulants or depleted uranium-containing prototypes. These samples have been thoroughly analyzed and their properties have been determined. A second line of research deals with fuel debris cutting. CEA has successfully applied and improved its laser cutting technology to fuel debris. Mechanical cutting has also been studied. One of the important safety issues related to debris cutting is the generation of radioactive aerosols and particles. Dedicated research programs have been carried out to characterize these releases and study mitigation techniques.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114333"},"PeriodicalIF":1.9,"publicationDate":"2025-07-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144694989","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":"Scaling analysis of turbulent buoyant jets for thermal striping in sodium fast reactors","authors":"Ryder L. Belgarde,&nbsp;Tiago A. Moreira,&nbsp;Mark H. Anderson","doi":"10.1016/j.nucengdes.2025.114294","DOIUrl":"10.1016/j.nucengdes.2025.114294","url":null,"abstract":"<div><div>Thermal striping is a phenomenon that occurs when a fluid of dissimilar temperature mixes near a surface, leading to convective transfer of periodic temperature fluctuations. These temperature fluctuations can result in high-cycle thermal fatigue and potential mechanical failure. This paper discusses methodologies for designing a reduced-scale experiment that is representative of turbulent buoyant jets or plumes exiting the reactor core of a prototypic reactor in an effort to further understand jet mixing and its effect on structures. The parameters considered to scale the jets include geometric scaling, the Froude number, the Weber number, and jet breakup length. Additionally, previous work found in the literature is used to assess non-dimensional velocity scaling associated with the jet axial distance, supporting the use of Froude number scaling in place of Reynolds similarity. Three distinct jet regimes previously defined in the literature are considered. The application of these regimes that describe the buoyancy and inertial forces acting on the jet is used to propose and design a scaled-down thermal stripping experiment, which will be used in future studies to provide valuable data that are consistent and scalable to a full-scale Sodium Fast Reactor.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114294"},"PeriodicalIF":1.9,"publicationDate":"2025-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144687119","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 research of spherical cermet fuel element for nuclear thermal propulsion based on multi-physics coupling method","authors":"Linyuan Lu ,&nbsp;Yunhao Wang ,&nbsp;Lihua Guo ,&nbsp;Tianxing Lan ,&nbsp;Wanlin Li ,&nbsp;Yajuan Zhong ,&nbsp;Jun Lin ,&nbsp;Haibin Zhang","doi":"10.1016/j.nucengdes.2025.114334","DOIUrl":"10.1016/j.nucengdes.2025.114334","url":null,"abstract":"<div><div>Nuclear Thermal Propulsion (NTP) has emerged as a promising technology for future manned Mars missions and beyond due to its combination of high specific impulse (800 ∼ 1000 s) and substantial thrust (44.45 ∼ 445 kN) compared to the traditional chemical propulsion (300 ∼ 450 s, 445 ∼ 2225 kN). To address the metallurgical bonding challenges in conventional prismatic ceramic metal (cermet) fuel element, this study proposes a novel spherical cermet fuel element and conducts fuel design research. Key design requirements for NTP fuels are summarized, including a UO₂ loading of approximately 50 %, a peak fuel temperature of less than 3070 K, and strain below a specific threshold based on fuel loss. Methodologically, an algorithm with a high filling rate of random particles is developed by simulating the quasi-physical elastic adjustment process of particles in a confined space. A multi-physics coupling model is established with COMSOL Multiphysics, incorporating heat transfer, thermal expansion, neutronics and burnup, creep, fission gas release and diffusion, irradiation growth and fuel loss. Computational results reveal the following key findings: The coating thickness of coated particles must be maintained below 15 μm to ensure high uranium loading; the fuel diameter should be constrained to 5 ∼ 6 mm to satisfy requirement of heat transfer; the operational duration must be limited to ensure that the strain requirement is met within a reasonable cladding thickness range. Material analysis demonstrates that W matrix with W cladding combination exhibits optimal thermal performance, while the W-25Re matrix with W cladding combination provides optimal mechanical performance; Strain analysis reveals that thermal expansion and matrix/cladding thermal creep account for over 95 % of the total fuel deformation, with UO₂ creep and irradiation growth contributing less than 5 %. The final designed fuel element consists of a tungsten matrix of 5 mm, a cladding thickness of 300 μm, and contains 46 vol% UO₂ in the form of tungsten-coated particles with a diameter of 220 μm and a coating thickness of 10 μm. The designed operating time for fuel is 3 h. During operation, the peak fuel temperature is 2982 K, and the maximum strain is 1.58 %.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114334"},"PeriodicalIF":1.9,"publicationDate":"2025-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144687120","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 numerical investigation on single-phase transverse mixing of the 4 × 4 helical cruciform fuel","authors":"Qi Zhang,&nbsp;Yong Xin","doi":"10.1016/j.nucengdes.2025.114325","DOIUrl":"10.1016/j.nucengdes.2025.114325","url":null,"abstract":"<div><div>In the presented work, the single-phase transverse mixing of the 4 × 4 helical cruciform fuel (HCF) assemblies is studied by Computational Fluid Dynamics (CFD) simulation. The bidirectional mass exchange is detected at rod gap type Ⅰ, while the unidirectional mass exchange is detected at rod gap type ⅠⅠ; these phenomena are termed two-way mixing and one-way mixing, respectively. The tracer method is employed to reveal the two-way mixing and one-way mixing behaviors occurring in the central and peripheral regions of the HCF assembly. In the central regions, the tracer concentration exhibits a gradual variation along the rod length, whereas in the peripheral regions, it varies more abruptly. Besides the concentration evolution of tracer exhibits negligible variation as the Reynolds number increased from 8.83 × 10³ to 4.42 × 10⁴. These findings confirm that the dominant mixing mechanism for HCF assembly is flow sweeping rather than turbulent mixing. The effective mixing coefficient for the HCF assemblies with <em>D_R</em>/<em>D_r</em> ranging from 0.51 to 3.33 and <em>L_HP</em> ranging from 0.5 m to 2.0 m is evaluated. It is found that the effective mixing coefficient increases with <em>D_R</em>/<em>D_r</em>, while it exhibits a linear dependence on the tangent of the rotation angle. In the presented work, effective mixing coefficient for the distinct HCF assemblies ranges between 0.018 and 0.035. An effective mixing coefficient correlation with broad geometrical applicability is proposed for the of HCF assembly.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114325"},"PeriodicalIF":1.9,"publicationDate":"2025-07-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144694988","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}