Nuclear Engineering and Design最新文献

{"title":"Uncertainty and sensitivity analysis of the fission product behaviour in the Phébus FPT1 test with the system code AC2","authors":"L. Tiborcz ,&nbsp;S. Beck","doi":"10.1016/j.nucengdes.2024.113594","DOIUrl":"10.1016/j.nucengdes.2024.113594","url":null,"abstract":"<div><div>The Phèbus FP (Fission Product) research program, conducted by IRSN and CEA, aimed to improve the understanding of the phenomena governing core melt down, fission product release and transport, as well as the fission product behaviour in the containment. The test facility represents a 900 MW_e PWR at a 1/5000 scale. The second test, FPT1, was carried out in 1996 and utilised an Ag-In-Cd alloy control rod. There is a growing interest in Best Estimate Plus Uncertainty analysis both in the research, as well as in the licensing. It has a long-standing history in case of thermal–hydraulic system codes focusing on design basis accidents, but its application is limited on the severe accident domain. In recent times however an increasing attention has been paid to evaluating severe accident codes’ uncertainties. This study presents an uncertainty and sensitivity analysis (UaSA) of the FPT1 test, focusing on fission product behavior modeling within the severe accident code system AC². The detailed application of the coupled codes ATHLET-CD and COCOSYS in this context is novelty. Detailed description of the methodological and practical approach is provided. The uncertain input parameters chosen and quantified are directly related to the phenomena describing the fission product transport and retention phenomena both in the primary circuit, as well as in the containment, while thermal–hydraulic conditions within the primary circuit were kept as much as possible constant. Finally, the study has been complemented by a sensitivity analysis deriving the Spearman's rank correlation coefficient for each uncertain input parameter.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-10-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142428802","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Risk-informed design: An unavailability allocation approach","authors":"Orestes Castillo-Hernández ,&nbsp;David A. Quintanar Gago ,&nbsp;Pamela F. Nelson","doi":"10.1016/j.nucengdes.2024.113611","DOIUrl":"10.1016/j.nucengdes.2024.113611","url":null,"abstract":"<div><div>Due to the technological complexity of nuclear power plants and the large number of components involved in their design, the objective space for balancing nuclear safety, availability performance, and cost in an optimization model becomes extremely vast. Consequently, this poses challenges in achieving operational optimization at the plant level, complicating the assurance of compliance with regulatory requirements. One solution is to partition the problem and analyze it at a system level to reduce the objective space. In this work, different options are analyzed to optimize cost and unavailability for systems that provide the means for meeting previously proposed unavailability targets. This ensures that licensing basis events (LBE) comply with risk targets and regulatory criteria. Additionally, the role of system unavailability due to maintenance activities and reliability issues related to unplanned component failures (e.g., random failures, common cause failures) in the optimization problem is analyzed. The various proposed optimization methods are implemented in the design of a pressurized water reactor (PWR) system. Evolutionary algorithms are used as optimization methods, with Genetic Algorithms for single-objective problems and Non-dominated Sorting Genetic Algorithm III (NSGA-III) for multi-objective problems. The main finding suggests that traditional models, which aim to minimize costs and unavailability, face challenges in adapting to the systems unavailability target. Therefore, optimization occurs by shifting the focus towards minimizing costs and distance to the unavailability target. This approach ensures that the results closely align with the unavailability target, thereby creating more efficiency in operating and maintenance costs while also ensuring acceptable design and operational regulatory compliance.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359542","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":"Incorporation of uranium nitride fuel capability into the ENIGMA fuel performance code: Model development and validation","authors":"Aiden Peakman ,&nbsp;Glyn Rossiter","doi":"10.1016/j.nucengdes.2024.113604","DOIUrl":"10.1016/j.nucengdes.2024.113604","url":null,"abstract":"<div><div>Uranium dioxide (UO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span>) is the primary fuel form for nuclear reactors but its moderate uranium density and low thermal conductivity have prompted the exploration of alternative materials. Uranium nitride (UN) has emerged as a promising candidate for a variety of reactors, offering higher uranium density and thermal conductivity. This paper details the development and implementation of UN fuel capabilities within the ENIGMA fuel performance code for Light Water Reactor (LWR) applications. The new UN capability in ENIGMA includes correlations for theoretical density at room temperature, thermal conductivity, specific heat capacity, enthalpy, thermal expansion strain, Young’s modulus, Poisson’s ratio, thermal creep strain rate, irradiation creep strain rate and emissivity. Additionally, it incorporates models for densification, solid fission product swelling, fission gas bubble swelling, and fission gas release, along with a modified RADAR model for determining the pellet radial power profile and helium generation. Validation of the UN model was conducted using data from the L414 pin irradiation in the JOYO fast reactor in Japan. Further validation efforts are planned using datasets from JOYO and the Siloé thermal reactor in France. The paper also outlines areas of future work to address experimental data gaps and enhance model accuracy to cover a broader range of cladding materials, manufacturing parameters (including porosity volume fraction) and operating conditions (including fuel temperatures and burnups). Although focused on LWR applications, the work outlined supports the use of UN fuel across various reactor systems.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-30","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359541","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Two-phase modelling for fission gas sweeping in restructuring nuclear oxide fuel","authors":"G. Zullo ,&nbsp;A. Scolaro ,&nbsp;T. Barani ,&nbsp;D. Pizzocri","doi":"10.1016/j.nucengdes.2024.113602","DOIUrl":"10.1016/j.nucengdes.2024.113602","url":null,"abstract":"<div><div>In this work, we propose a modelling approach for the intra-granular fission gas behaviour in UO₂ under restructuring process. Leveraging the definition of restructured volume fraction, we consider the fuel matrix transition from the non-restructured to the restructured phase, together with the evolution of the corresponding fission gas concentrations retained in the fuel matrix. Firstly, we derive a sweeping term that exchanges fission gas atoms from the non-restructured to the restructured fuel region. The sweeping term is then included in the conventional intra-granular fission gas diffusion problem. Secondly, the spectral diffusion algorithm is employed to solve two spatially-dimensionless problems, properly representing the non-restructured region with micrometric grains and the restructured region with sub-micrometric grains. The model developed is implemented in SCIANTIX, a 0D meso-scale code for physics-based modelling of fission gas behaviour in nuclear oxide fuel and compared with experimental data and semi-empirical models.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-28","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142359540","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":"Review on crack growth driving force at the tip of stress corrosion cracking in the safe end dissimilar metal welded joint","authors":"Zheng Wang ,&nbsp;Yuxuan Xue ,&nbsp;Rongxin Wang ,&nbsp;Jun Wu ,&nbsp;Yubiao Zhang ,&nbsp;He Xue","doi":"10.1016/j.nucengdes.2024.113609","DOIUrl":"10.1016/j.nucengdes.2024.113609","url":null,"abstract":"<div><div>The welded structural materials of nuclear power plants (NPPs) are susceptible to environmentally-assisted cracking (EAC), represented by stress corrosion cracking (SCC), in prolonged high-temperature and high-pressure water environments, posing a significant threat to plant safety. This study aims to provide a critical review for the crack growth driving force at the tip of SCC in the safe end dissimilar metal welded joint (DMWJ) of NPPs. Firstly, SCC’s background, importance, and current research status are introduced. Secondly, a review and analysis are conducted on SCC’s initiation and growth stages, focusing on experimental methods, predictive models of crack growth rate, crack tip mechanical states, and influencing factors, clarifying the main achievements and challenges in current experimental and theoretical research. Finally, a method to mitigate crack tip driving force is proposed, followed by an in-depth analysis from a mechanical perspective on the relationship between crack growth driving force and crack growth resistance, highlighting future research trends. This review provides theoretical references and technical support for addressing the issue of SCC in welded structural materials of NPP primary circuit.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142326753","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Prediction of dK/da effects on stress corrosion crack growth rate of irradiated stainless steels based on slip oxidation mechanism","authors":"Masato Koshiishi","doi":"10.1016/j.nucengdes.2024.113610","DOIUrl":"10.1016/j.nucengdes.2024.113610","url":null,"abstract":"<div><div>This study developed a predicting method for the crack growth rate (CGR) of stress corrosion cracking (SCC) under varying stress intensity factor (K) conditions for irradiated stainless steel (SS). First, optimization of the Hashimoto-Koshiishi model equation for calculating CGR under varying dK/da conditions was carried out using the Rice, Dugan, and Sham (RDS) equation for the crack tip strain rate. Second, the model input coefficients were set to incorporate the effect of dK/da on SCC CGR to fit the experimental data. Finally, the effect of dK/da on the CGR for 3 dpa irradiated SS of a core shroud was evaluated taking into account the weld residual stress distribution. The model prediction showed that there was little effect on the CGR for the increasing and decreasing K regions in the plate thickness direction of the core shroud under boiling water reactor (BWR) normal water chemistry, but the effect was not negligible for the decreasing K region under BWR hydrogen water chemistry.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-27","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142326752","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":"Monte Carlo multiphysics simulation on adaptive unstructured mesh geometry","authors":"A.J. Novak ,&nbsp;H. Brooks ,&nbsp;P. Shriwise ,&nbsp;A. Davis","doi":"10.1016/j.nucengdes.2024.113589","DOIUrl":"10.1016/j.nucengdes.2024.113589","url":null,"abstract":"<div><div>Monte Carlo simulation based on Constructive Solid Geometry (CSG) brings unique challenges for multiphysics simulation, including establishing field transfers with mesh-based physics codes, the combination of stochastic and deterministic solvers, and high computational expense. In this work, an adaptive, on-the-fly mesh-based Monte Carlo geometry algorithm is implemented in Cardinal to reduce the barrier-to-entry for high-fidelity multiphysics by (i) eliminating ambiguity in defining CSG cells for temperature and density feedback, (ii) enabling simple mesh convergence studies, and (iii) more closely integrating Computer Aided Design (CAD) workflows with Monte Carlo methods. During Picard iterations, an OpenMC mesh geometry is adaptively refined or coarsened by contouring temperature and/or density fields from a thermal-fluid solver. This algorithm is applied to a full-core Molten Salt Fast Reactor (MSFR) geometry with NekRS Large Eddy Simulation (LES) coupled to OpenMC neutron transport. A performance study indicates a net speedup of 2.3<span><math><mo>×</mo></math></span> in the OpenMC solver when using an adaptive geometry for cell sizes chosen intermediate to the as-built CAD geometry versus 1:1 element tracking, which points to future algorithmic research in accelerated Monte Carlo mesh tracking.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142322812","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 novel method of estimating earthquake durations for the analysis of floor vibrations of nuclear power plants","authors":"Vilho Jussila ,&nbsp;Ludovic Fülöp ,&nbsp;Päivi Mäntyniemi ,&nbsp;Jari Puttonen","doi":"10.1016/j.nucengdes.2024.113606","DOIUrl":"10.1016/j.nucengdes.2024.113606","url":null,"abstract":"<div><div>Many low-seismicity countries such as Finland have adopted IAEA requirements and recommendations for seismic design of new and existing nuclear power plants (NPPs). In low seismic regions, the structural seismic design is associated with floor vibration of NPPs. The floor vibration analysis is usually conducted in the time domain for which maximum amplitudes are retrieved from design spectra while the duration of ground motion is estimated as an interval between 5% and 75% of accumulation of the Arias intensity. As this method was developed for active seismic regions, it often overestimates the duration for the regions with low seismicity. The present article introduces a new twofold method for estimating the duration. First, the Arias intensity is calculated for a complete and consecutively reduced accelerograms resulting in a deviation curve. Second, this curve is simplified by a piecewise linear regression fitting. The simplified deviation curve has a linear time frame that includes the most significant part of the Arias intensity. The length of the time frame defines the effective duration of a specific ground motion. This implies that the effective duration depends directly on the ground motion instead of predefined percentiles of the Aries intensity. In this study, the method was applied to a set of ground accelerations adopted from eastern Canada, which is geologically similar to the Fennoscandian Shield where appropriate recordings are absent. The results showed that the durations depend on distance, but they were insensitive of magnitude for short rupture distances. This indicates that smaller events can also be useful for estimating the durations even though they do not meet the requirement of design basis earthquake in terms of the peak ground acceleration. The durations obtained with the proposed method were typically shorter than those based on the 5%–75% criterion. The durations received can be used to generate the acceleration time histories compliant with the design response spectra. We also propose durations with different rupture distances for the seismic design of the structures, systems, and components of nuclear facilities in Finland. In a feasibility study, we calculated floor vibrations of a generic reactor building using 3D finite element analysis. The results show that floor accelerations are very similar, when the base accelerogram is complete or shortened to the length proposed in this study.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-26","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142322813","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Strategic fuel management via implementation of a combined reload-reshuffle scheme in small modular reactors","authors":"Abidur Rahman Ishraq ,&nbsp;Anton Evgenievich Kruglikov ,&nbsp;H. Rainad Khan Rohan","doi":"10.1016/j.nucengdes.2024.113605","DOIUrl":"10.1016/j.nucengdes.2024.113605","url":null,"abstract":"<div><div>This study aims to implement a coupled fuel reload-reshuffle scheme for a PWR-based SMR. Considering 540 EFPD as the cycle length, a heterogenous poison-free core based on the design of ACP-100 with 57 fuel assemblies (FAs) utilizing three different enrichments (3.0 wt.%, 4.0 wt.%, and 4.45 wt.%) was modeled in SERPENT. Initially, the core achieved a k_eff of 1.31492 and a radial PPF of 1.77, which decreased to 1.10914 and 1.19 respectively at the end of the first cycle. Reloading 12 fresh FAs and shuffling 32 irradiated FAs within the core at this point increased the k_eff to 1.1584, sustaining criticality for an additional 540 EFPDs (the second cycle). Two more burnup cycles were simulated with the refueling patterns being established by evaluating the assembly discharge burnup and core power profile. Through a hybrid combination of in-out and out-in loading approaches, a high cumulative average discharge burnup exceeding 30 MWD/kg (over 40 MWD/kg for some assemblies) was achieved at the end of the fourth cycle (2160 EFPDs). Although the employed refueling patterns raised the power peaking factors (PPFs) at the beginning of each cycle, the core power distribution in general became more uniform and the PPF decreased as burnup progressed. Other than the beginning of the fourth cycle, the obtained PPF values were less than or around 2.00 even without the use of any control systems. Both the fuel and moderator temperature coefficients remained sufficiently negative throughout the burnup cycles. Further iterations of the implemented refueling schemes can be carried out depending on plant operational requirements.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142320408","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":"Vibrations of PWR fuel assembly under axial coolant flow and oblique impingement of jet cross-flow from LOCA holes","authors":"Ibrahim Gad-el-Hak ,&nbsp;Njuki Mureithi ,&nbsp;Kostas Karazis","doi":"10.1016/j.nucengdes.2024.113586","DOIUrl":"10.1016/j.nucengdes.2024.113586","url":null,"abstract":"<div><div>Nuclear fuel bundles are exposed to localized normal impinging jet cross-flow at certain locations along the fuel rod span in a specific design of pressurized water reactors (PWRs) with a fail-safe feature for a loss-of-coolant accident (LOCA). The combined axial flow and jet cross-flow from LOCA holes can induce extensive fuel rod vibration, leading to fretting wear, particularly in rods near the LOCA holes. The dynamics of the fuel assembly depend strongly on the jet impingement angle (<span><math><mi>θ</mi></math></span>) where the jet flow impinges the fuel rods. This paper investigates experimentally the effect of oblique impingement of a circular jet flow in axial flow on the vibration of a reduced scale model array of the fuel assembly. The mock-up array is tested for three jet inclination angles relative to the rod axis. In addition, the effect of axial flow on the jet-induced dynamics in the array is investigated. The tests are done for the jet centerline located symmetrically or eccentrically relative to the rod inter-column gap. The jet eccentricity is found to have a significant effect on rod bundle stability. The results show that the axial flow has a stabilizing effect on the jet-induced instability. The stability threshold of the array is significantly affected by the jet injection angle. The array becomes significantly more unstable when the jet flow is injected at <span><math><mrow><mi>θ</mi><mo>=</mo><mn>70</mn><mo>°</mo></mrow></math></span> compared to the normal jet impingement case. For a jet impingement angle larger than 90 degrees, the stability behavior is more complex. While the rod bundle undergoes instability, further increasing the jet velocity did not exacerbate the vibration response, thus suggesting an apparently self-limiting instability for the non-eccentric jet.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-25","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142320409","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}