Nuclear Engineering and Design最新文献

{"title":"Seismic response of vertical dry storage casks under three-dimensional earthquake motions","authors":"Tonmoy Das ,&nbsp;Elnaz Seylabi ,&nbsp;Maryam Tabbakhhah ,&nbsp;David McCallen","doi":"10.1016/j.nucengdes.2025.114339","DOIUrl":"10.1016/j.nucengdes.2025.114339","url":null,"abstract":"<div><div>Ensuring the long-term seismic safety of dry storage casks (DSCs) is becoming increasingly critical as these systems evolve from temporary to de facto permanent repositories for spent nuclear fuels. Traditional seismic soil–structure interaction (SSI) assessment methods use one-dimensional deconvolution or simplified boundary conditions to model incident waves. Although computationally appealing, simplifying assumptions may alter the seismic risk by neglecting the full complexity of three-dimensional (3D) wave propagation effects. To address this challenge, this paper introduces a novel high-fidelity computational framework that leverages the Domain Reduction Method (DRM) with perfectly matched layers (PML) to accurately transfer complex, 3D seismic wavefields from regional-scale fault-rupture simulations into local-scale finite element models of DSCs. Using broadband, physics-based ground motions from a generic <span><math><msub><mrow><mi>M</mi></mrow><mrow><mi>w</mi></mrow></msub></math></span>7.0 strike-slip event, both single-cask and multi-cask configurations were investigated under near- and far-field conditions. Emphasis is placed on capturing complex SSI, spatial variability in the ground motion, and nonlinear phenomena such as cask rocking and sliding. Numerical results demonstrate that near-field conditions, where forward directivity and fling-step effects dominate, lead to significantly higher DSC rocking and sliding. Far-field cases, by contrast, generally exhibit modest responses. Incorporating SSI tends to amplify or alter DSC response spectra and introduce response variability, which underscores the need for site-specific evaluations and robust modeling approaches to ensure the seismic integrity of DSCs in interim spent fuel storage installations.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114339"},"PeriodicalIF":2.1,"publicationDate":"2025-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144763981","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":"Consequence analyses of sabotage-induced radiological releases in sodium-cooled fast microreactors","authors":"M.D. Shah,&nbsp;D. Hartanto","doi":"10.1016/j.nucengdes.2025.114360","DOIUrl":"10.1016/j.nucengdes.2025.114360","url":null,"abstract":"<div><div>Analysis of three sodium-cooled fast microreactors (SFMs) with thermal powers of 10, 30, and 50<!--> <!-->MWt showed that smaller reactors result in lower radiological consequences during a postulated sabotage-induced event because of their reduced core inventory. All SFMs used U-10Zr metal fuel enriched to 15 wt% high-assay low-enriched uranium and operated until their respective effective multiplication factor (<em>k_eff</em>) reduced to less than 1 or until the end of their operational lifespan. Sabotage scenarios were simulated at this point, when the fuel inventory within the core contains the highest-level of radioactivity. Radionuclide core inventories were calculated using the SCALE code at shutdown and 3 <!--> <!-->days post-shutdown. Dose consequence analyses were performed for three sabotage scenarios using the RASCAL tool. As microreactor developers plan for minimal on-site or complete off-site emergency response, it remains essential to evaluate their physical protection needs and potential hazards, including assessing postulated sabotage-induced events that could become more relevant. SFM licensees should identify a credible worst-case, major accident, estimate release source terms, and perform dose consequence analyses to evaluate site-specific physical protection measures. This recommendation supports a risk-informed, performance-based approach, aligning with applicable regulatory requirements, i.e., 10 CFR Parts 100 and 53 rulemaking in the United States.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114360"},"PeriodicalIF":2.1,"publicationDate":"2025-08-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144757251","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":"Investigation of the passive heat removal from the SCW-SMR reactor core in the event of a Long-Term Station Blackout","authors":"Tamás Varju,&nbsp;Attila Aszódi","doi":"10.1016/j.nucengdes.2025.114349","DOIUrl":"10.1016/j.nucengdes.2025.114349","url":null,"abstract":"<div><div>In the frame of the EU ECC-SMART project, the pre-conceptual design of a water-cooled small modular reactor operating at supercritical pressure (SCW-SMR) with seven heat-up stages is under development and assessment. BME is contributing to this project as a consortium member and is developing several models and coupled code systems to investigate the features and characteristics of the proposed new concept. Following the steady-state analyses conducted previously, the Apros system code model was extended to encompass models of the SCW-SMR primary circuit, safety systems and I&amp;C systems. A series of calculations were carried out to investigate the possibility of natural circulation in the proposed core layout. Another objective was to check the feasibility of passive heat removal in the event of a long-term SBO, which was challenging in the predecessor HPLWR design. A detailed comparative assessment and a summary of the lessons learned are provided.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114349"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144750538","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":"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":"Experimental investigation on turbulent structures in a blocked narrow rectangular channel by TR-PIV","authors":"Wenhai Qu ,&nbsp;Hao Xie ,&nbsp;Hanyu Wang ,&nbsp;Jinbiao Xiong","doi":"10.1016/j.nucengdes.2025.114347","DOIUrl":"10.1016/j.nucengdes.2025.114347","url":null,"abstract":"<div><div>The plate-type fuel assembly has advantage of high-power density. However, the narrow rectangular flow channels between fuel plates can be blocked by debris and blisters, which may result in decreased flow rate and heat transfer deterioration in blocked flow channel. The turbulent flow in a blocked narrow rectangular channel should be studied to reveal the blockage effect on flow. In this study, turbulent flow structures downstream of central-located blockages, including cylinder, spherical crown or cylinder-plate, in a narrow rectangular channel (177.5 mm in width, 4.5 mm in thickness and 900 mm in length) were studied by time-resolved particle image velocimetry (TR-PIV). Based on experimental data, time-averaged flow structures and coherent structures around different blockages were analyzed. The vortex shedding from blockages and strong shear flow improve Reynolds stresses downstream of blockages. The frequency characteristics and length scales of coherent structures were identified based on power spectral density analysis and length scale analysis. The coherent structures were extracted by proper orthogonal decomposition.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114347"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738975","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":"Validation of a coupled STH and low-fidelity CFD approach using experimental data from NACIE-UP test facility","authors":"Pietro Stefanini ,&nbsp;Andrea Pucciarelli ,&nbsp;Ossama Halim ,&nbsp;Nicola Forgione ,&nbsp;Ivan Di Piazza","doi":"10.1016/j.nucengdes.2025.114353","DOIUrl":"10.1016/j.nucengdes.2025.114353","url":null,"abstract":"<div><div>The IAEA CRP “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” aims to support the development of numerical models suitable for the investigation of thermal–hydraulic analysis of liquid metal cooled reactors. In particular, this work presents a multiscale approach developed to assess cases where the addressed facility experiences a simulated Loss Of Flow Accident. A domain decomposition approach was used together with an explicit scheme, adopting a novel application philosophy for CFD modelling. Indeed, the CFD model was built to simplify as much as possible the wrapped wires design, since a detailed description of such components increases the cell count exponentially. The “simplified” model allows for a shorter running time of the simulation with an equivalent number of CPUs used, making this approach appealing for preliminary estimations of the main thermal-hydraulics phenomena involved. The results show that suitable predictions were usually achieved both for the initial and final steady-state conditions; improvements can be obtained in the transient phase working on the considered boundary conditions.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":"444 ","pages":"Article 114353"},"PeriodicalIF":2.1,"publicationDate":"2025-07-31","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"144738977","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}