Nuclear Engineering and Design最新文献

{"title":"Diffusion of solid fission products in UO2 and UO2+x","authors":"Xi Zhou ,&nbsp;Nan Chao ,&nbsp;Caishan Jiao ,&nbsp;Yang Gao ,&nbsp;Chunhui Li ,&nbsp;Yang Zhang","doi":"10.1016/j.nucengdes.2025.114341","DOIUrl":"10.1016/j.nucengdes.2025.114341","url":null,"abstract":"<div><div>The diffusion behaviors of solid fission products Zr (Zr⁴⁺), Ru (Ru⁴⁺), Ce (Ce⁴⁺), Y (Y³⁺), La (La³⁺), Sr (Sr²⁺), and Ba (Ba²⁺) in stoichiometric uranium dioxide (UO₂) and hyperstoichiometric uranium dioxide (UO_2+x) systems have been investigated using density functional theory (DFT) and empirical potential (EP) methods. Solid fission products commonly occupy uranium vacancy trap sites in UO₂ and UO_2+x, with migration occurring via uranium vacancy assisted mechanisms. Five distinct elementary migration mechanisms have been identified. Among these, the impurity-trap exchange mechanism generally exhibits the lowest migration barrier, making it the most active diffusion mechanism. However, the migration energies for U self-diffusion and impurity-trap exchange for Ru⁴⁺ are comparable in UO_2+x systems, causing the most active diffusion mechanism to shift from U self-diffusion to impurity-trap exchange. For Ce⁴⁺, the migration barrier for U self-diffusion consistently remains lower than that for impurity-trap exchange, thereby maintaining U self-diffusion as the most active diffusion mechanism in UO_2+x systems. The diffusivities of Y³⁺, La³⁺, and Ce⁴⁺ are comparable to that of U self-diffusion in both UO₂ and UO_2+x systems, whereas Sr²⁺ and Ba²⁺ exhibit higher diffusivities than U self-diffusion. The diffusivities of fission gas Kr are significantly higher than that of U self-diffusion. Moreover, the diffusivities of Zr⁴⁺ and Ru⁴⁺ relative to U self-diffusion differ between UO₂ and UO_2+x systems. This discrepancy is attributed to Ru⁴⁺ preferentially forming metallic precipitates in UO₂, while remaining dissolved in the matrix in UO_2+x, with Zr⁴⁺ exhibiting similar behavior. The differences in the diffusivities of fission products compared to U self-diffusion are primarily influenced by the chemical state of the fission products and their ionic potential differences relative to U⁴⁺. These phenomena indicate that a more stable chemical state of a fission product leads to greater solubility, a smaller ionic potential difference from U⁴⁺, and diffusivity that more closely aligns with U self-diffusion.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114341"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738979","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 and evaluation of prestress loss of containment based on field monitoring data","authors":"Kaixing Liao ,&nbsp;Ying Huang ,&nbsp;Jiapei Xu ,&nbsp;Zhijie Tang ,&nbsp;Weiping Zhang ,&nbsp;Yong Zhou","doi":"10.1016/j.nucengdes.2025.114344","DOIUrl":"10.1016/j.nucengdes.2025.114344","url":null,"abstract":"<div><div>Reliable prediction of prestress loss in nuclear containment structures is essential for ensuring long-term structural integrity and is a critical component of the Time-Limited Aging Analysis (TLAA) required for Operating License Extension (OLE) of nuclear power plants. This study presents a comprehensive method that integrates field monitoring data with numerical modeling to predict the prestress behavior of a CPR1000 prestressed concrete containment over a 60-year service life. Firstly, a refined three-dimensional finite element (FE) model is developed and calibrated using short-term strain data obtained during the Containment Tightness Test (CTT), allowing accurate identification of the elastic modulus and Poisson’s ratio of concrete and tendons through sensitivity analysis. Subsequently, long-term concrete strain data obtained from 30 years of continuous monitoring are used to select appropriate creep and shrinkage models, enabling the development of a time-dependent prestress prediction method. Compared with conventional approaches based on laboratory tests or small-scale mock-ups, this method is validated using full-scale, in-service data, offering enhanced accuracy and practical applicability. The predicted prestress levels remain above the Minimum Required Value (MRV) throughout the 60-year period, satisfying regulatory criteria. The FE model is further employed to assess the structural responses of the containment under various internal pressure scenarios, accounting for the time-dependent loss of prestress. The results confirm that the structural integrity of the containment is preserved throughout the extended service life. This study provides a validated, monitoring-based framework for long-term prestress evaluation, offering a technically robust tool to support safety assessment and life extension of nuclear power plant containment structures.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114344"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738976","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 QUENCH-15 and QUENCH-19 tests using MELCOR 2.2 code","authors":"Tereza Abrman Marková ,&nbsp;Guido Mazzini ,&nbsp;Martin Ševeček","doi":"10.1016/j.nucengdes.2025.114351","DOIUrl":"10.1016/j.nucengdes.2025.114351","url":null,"abstract":"<div><div>This study presents a benchmarking analysis of the behavior of traditional Zr-based alloy (ZIRLO^TM) and accident tolerant fuel cladding candidate FeCrAl based on the QUENCH-15 and QUENCH-19 bundle tests, which simulate severe accident conditions. A sensitivity analysis is conducted on parameters influencing oxidation equations for the tested nuclear fuel materials. A comparative analysis of the simulations performed in three MELCOR 2.2 versions is presented, highlighting differences in hydrogen generation and release and maximum temperature predictions. The results demonstrate that the generic oxidation model is more sensitive to user effect compared to the Arrhenius model for Zr-based alloys. The generic oxidation model sensitivity coefficients do not significantly alter predicted hydrogen production results. Furthermore, it was found that the sensitivity of the results to radiation factors, material used, parameters of the oxidation equations, and other conditions can be utilized in future benchmarking analyses and in modeling for commercial nuclear power plants.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114351"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738978","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 one-class explainable AI framework for identification of non-stationary concurrent false data injections in nuclear reactor signals","authors":"Zachery Dahm,&nbsp;Vasileios Theos,&nbsp;Konstantinos Vasili,&nbsp;William Richards,&nbsp;Konstantinos Gkouliaras,&nbsp;Stylianos Chatzidakis","doi":"10.1016/j.nucengdes.2025.114359","DOIUrl":"10.1016/j.nucengdes.2025.114359","url":null,"abstract":"<div><div>The transition of next generation advanced nuclear reactor systems from analog to fully digital instrumentation and control will necessitate robust mechanisms to safeguard against potential data integrity threats. One challenge is the real-time characterization of false data injections, which can mask sensor signals and potentially disrupt reactor control systems. While significant progress has been made in anomaly detection within reactor systems, potential false data injections have been shown to bypass conventional linear time-invariant state estimators and failure detectors based on statistical thresholds. The dynamic, nonlinear, multi-variate nature of sensor signals, combined with inherent noise and limited availability of real-world training data, makes the characterization of such threats and more importantly their differentiation from anticipated process anomalies particularly challenging. In this paper, we present an eXplainable AI (XAI) framework for identifying non-stationary concurrent replay attacks in nuclear reactor signals with minimal training data. The proposed framework leverages progress on recurrent neural networks and residual analysis coupled with a modified SHAP algorithm and rule-based correlations. The recurrent neural networks are trained only on normal operational data while for residual analysis we introduce an adaptive windowing technique to improve detection accuracy. We successfully benchmarked this framework on a real-world dataset from Purdue’s nuclear reactor (PUR-1). We were able to detect false data injections with accuracy higher than 93% and less than 1% false positives, differentiate from expected process anomalies, and to identify the origin of the falsified signals.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114359"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750539","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":"Multi-cycle reload analysis of a long cycle gas-cooled fast modular reactor","authors":"Dylan J.L. Garofalo,&nbsp;Ben Lindley","doi":"10.1016/j.nucengdes.2025.114362","DOIUrl":"10.1016/j.nucengdes.2025.114362","url":null,"abstract":"<div><div>There is currently significant interest in deploying HALEU-fueled fast reactors, including the General Atomics (GA) Fast Modular Reactor (FMR). Such reactors can achieve very long fuel cycles, but with multi-batch loading will take decades to reach equilibrium. This motivates design and analysis of both the initial core and multi-cycle reload, which is typically performed using fast-running, deterministic fast reactor codes such as the Argonne Reactor Computation (ARC) codes. In this paper, multicycle reload of the GA FMR is analyzed using the ARC codes. The GA FMR utilizes 19.75 % enriched fuel in a 16 year cycle with a three-batch strategy, with twice-burned fuel placed on the core periphery. The GA FMR has a softened neutron spectrum due to reflecting elements in the core, so the neutronic solution is first benchmarked against the OpenMC Monte Carlo code. Discrepancy on <span><math><mrow><msub><mi>k</mi><mrow><mi>eff</mi></mrow></msub></mrow></math></span> is 400–600 pcm, likely due to the softened neutron spectrum, heterogeneous fuel assembly design and central reflector. However, the rms discrepancy on the assembly power distribution is only 0.6 %, despite the presence of the central reflector. A reload strategy is devised for the first three cycles of such a reactor, ultimately spanning the first 45–48 years of its operation. The fresh core uses 19.75 %, 19.25 % and 16.75 % enriched fuel in place of fresh, once-burned and twice-burned and is then subsequently refueled with only 19.75 % enriched fuel. The cycle length is varied over 3 cycles of operation to balance fuel utilization and reactor availability, specifically with use of an extended 18-year Cycle 1, followed by a shortened 11-year Cycle 2. Cycle 3 is close to the target 16-year length. Finally, placing twice burned assemblies next to the GA FMR central reflector can reduce power peaking by 3 %, at the expense of slightly reducing the cycle length.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114362"},"PeriodicalIF":2.1,"publicationDate":"2025-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144723469","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":"Scale and model effects analysis on the radial jet thermal–hydraulic behavior in sodium fast reactor","authors":"Benjamin Jourdy ,&nbsp;David Guenadou ,&nbsp;Nathalie Seiler ,&nbsp;Alexandre Labergue ,&nbsp;Michel Gradeck","doi":"10.1016/j.nucengdes.2025.114352","DOIUrl":"10.1016/j.nucengdes.2025.114352","url":null,"abstract":"<div><div>Safety studies for sodium-cooled fast reactors (SFRs) rely on validated numerical tools, often using experimental data from mock-ups. A key challenge is ensuring these mock-ups accurately represent the full-scale reactor. In this context, a 1:6 downscaled mock-up of the upper plenum of the French SFR ASTRID reactor design was used to carry out experiments to study the core flow field. The present study focuses on a specific instability of the flow which concerns the behavior of the radial jet at the exit of the core. Since results have been obtained on the downscaled water mock-up, a further question lies in upscaling these results to real reactor scale considering sodium as operating fluid. To address the issue of this transposition to the scale and fluid of the ASTRID reactor, a dedicated methodology has been implemented based on the use of two additional mock-ups. It is shown that the stability of the radial jet is primarily governed by a balance of forces involving buoyancy, inertia and Coanda effects. So, a dimensionless number (namely <span><math><mrow><mi>L</mi></mrow></math></span>) has been introduced to characterize the competition between these phenomena. Experimental results obtained at different scales show that the conservation of this number ensures a quite correct transposition to a different scale. Finally, as small-scale mock-ups are operating with water instead of sodium, the transposition related to the operating fluid (namely model effect) is studied based on the Péclet number evolution. It is shown that distortions from sodium to water transposition are balanced by the scale effects from the reactor to the mock-ups. This study highlights that small-scale water mock-ups are representative of the thermal behavior of full-scale sodium reactors regarding the issue of the radial hot jet behavior.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114352"},"PeriodicalIF":2.1,"publicationDate":"2025-07-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144723470","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Experimental investigations into the rocking response of graphite-block assemblies in a HTGR core under horizontal earthquake shaking","authors":"Sai Sharath Parsi ,&nbsp;Andrew S. Whittaker ,&nbsp;Mettupalayam V. Sivaselvan ,&nbsp;Enrique Velez-Lopez ,&nbsp;W. Robb Stewart ,&nbsp;Koroush Shirvan","doi":"10.1016/j.nucengdes.2025.114300","DOIUrl":"10.1016/j.nucengdes.2025.114300","url":null,"abstract":"<div><div>This paper presents earthquake simulator experiments of graphite block assemblies that are representative of the core of a horizontal compact high temperature gas reactor (HC-HTGR). The reactor core consists of vertically stacked prismatic graphite blocks, connected via shear keys. The columns of blocks are separated by finite gaps in both the transverse and longitudinal directions to accommodate fabrication and installation tolerances, and swelling of graphite. The dynamic response of the graphite core is governed by a combination of factors, including rigid body rocking of blocks, friction and clearances within shear keys, kinematic constraints, design and manufacturing tolerances, uplift and disengagement, and impacts between adjacent columns of blocks. Characterizing these effects involves: (a) understanding the rocking dynamics of standalone columns of blocks without considerations of impact, which is the focus of this paper, and (b) evaluating the impact dynamics between closely spaced columns. This paper presents results of and observations from a series of vibration and seismic experiments conducted at the University at Buffalo. The tests include quasi-static loading to evaluate rocking initiation and progression across various block interfaces, harmonic excitation to assess frequency-dependent rocking behavior of standalone columns of blocks, free vibration measurements to investigate oscillatory and damping characteristics across varying column geometries (e.g., stack height, rocking axis), and seismic simulations of individual columns and as well as 2D wall assemblies using representative ground motions. Rigid-body responses of the blocks were measured using a high-precision optical tracking system, generating data that provide valuable insights into their dynamic (and rocking) characteristics, and a robust technical basis to support the development and validation of numerical models.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"443 ","pages":"Article 114300"},"PeriodicalIF":2.1,"publicationDate":"2025-07-29","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144722927","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":"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}