Nuclear Engineering and Design最新文献

{"title":"Monte Carlo multiphysics simulation on adaptive unstructured mesh geometry","authors":"","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":"","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":"","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":"","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}

{"title":"Pareto front of sodium void worth and breeding ratio in metal-fueled sodium fast reactor with axially heterogeneous core design by coupling neutronics and genetic algorithm","authors":"","doi":"10.1016/j.nucengdes.2024.113598","DOIUrl":"10.1016/j.nucengdes.2024.113598","url":null,"abstract":"<div><div>For sodium-cooled fast reactors (SFRs), achieving low sodium void worth (SVW) and flexible breeding ratios (BR) is crucial. This study establishes a method coupling neutronics simulations with genetic algorithms (GA) and applies it to a 3000 MWth metal-fueled SFR to optimize the SVW, BR, and the power distribution of axially heterogeneous cores. The Pareto front, i.e. optimal front, between SVW and BR is determined, quantitatively clarifying the trade-off relationship between these two parameters. Axially heterogeneous cores with equal inner and outer fissile heights can achieve SVW adjustments from −71 to 3148 pcm and BR adjustments from 1.26 to 1.62. Designs with unequal fissile heights slightly expand the feasible region, but the impact on the optimal front for equal-height designs is limited. Through the analysis of the spatial distribution of the sodium void effect, it is shown that the bond sodium model significantly influences the SVW in low sodium void cores, while its impact on high breeding cores is minimal. The results indicate that the power distribution can be optimized by adjusting the enrichment of subregions within the constraints of SVW and BR. For low sodium void cores, retaining a certain amount of lower fertile material is necessary to flatten the power distribution. While the criticality search during depletion calculations affects the parameters, its impact on identifying the optimal front is limited.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311661","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":"Numerical simulation of corium flow through rod bundle and/or debris bed geometries with a model based on Lattice Boltzmann method","authors":"","doi":"10.1016/j.nucengdes.2024.113603","DOIUrl":"10.1016/j.nucengdes.2024.113603","url":null,"abstract":"<div><div>A new model is proposed to investigate the relocation and the distribution of hot corium flows in different configurations (rod bundle, porous debris bed) representative of a severe accident in a Light Water Reactor (LWR). Our model relies on the coupling between a modified Lattice Boltzmann Method (LBM), called Free-Surface LBM, that solves hydrodynamics of unsaturated corium and a Finite Volume Method (FVM) that solves heat transfers. Corium solidification and melting are addressed by implementing a correlation between the temperature and the viscosity. Several simulations on representative elementary volumes were performed, varying configurations (debris bed, rod bundle with and without grid). From the results, it is possible to capture important details of the flow at a scale lower than the pore scale and, at the same time, it is possible to take into account the average effects at the scale of several pores. Presented as a proof of concept these preliminary studies show the interest of this kind of CFD approach to identify which parameters at microstructure scale can potentially govern the corium relocation kinetics at macroscopic scale. It will provide useful information that might improve core degradation models in severe accident codes, such as ASTEC.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-24","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142315933","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":"Computational Fluid Dynamics investigation of the impact of 6% crept pressure tubes on flow behaviour, fuel temperature, and pressure tube wall temperature of a single CANDU 37M fuel bundle","authors":"","doi":"10.1016/j.nucengdes.2024.113593","DOIUrl":"10.1016/j.nucengdes.2024.113593","url":null,"abstract":"<div><div>CANDU nuclear generating stations experience aging effects that affect the reactor operation, including pressure tube deformation (<em>i.e</em>., diametral expansion, sag, and elongation). The diametral expansion of the pressure tube will alter coolant flow behaviour, which will impact CANDU fuel and pressure tube temperatures, thereby directly affecting the reactor’s operational performance and safety margins. However, these impacts are not yet fully understood at this point. In this study, two Computational Fluid Dynamics simulations were conducted with STAR CCM+ on a single CANDU Modified 37-element (37M) fuel bundle placed in both non-crept and 6% crept pressure tubes under normal operating conditions. The predicted coolant flow behaviour, fuel temperatures, and pressure tube wall temperatures were compared between both cases to predict the impact of diametral expansion on these aspects. The results indicate that approximately 29% of the coolant flow bypasses the bundle in the 6% crept pressure tube, leading to a reduction of up to 25% in subchannel flow velocity and a maximum increase of 36.7 K in fuel maximum temperature. Both the non-crept and 6% crept pressure tube wall temperature profiles were found to be asymmetric with respect to the bundle’s horizontal axis. The temperature at the bottom of the pressure tube is relatively higher than at the top in the non-crept case, while the temperature difference is noticeably greater in the 6% crept case.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-23","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324006939/pdfft?md5=41da0822104687cb7b38045c14ca96d9&pid=1-s2.0-S0029549324006939-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311660","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":"Incorporation of organic liquid waste in alkali activated mixed fly ash/blast furnace slag/metakaolin-based geopolymers","authors":"","doi":"10.1016/j.nucengdes.2024.113608","DOIUrl":"10.1016/j.nucengdes.2024.113608","url":null,"abstract":"<div><div>The solidification of liquid oils (Shell Spirax and Nevastane EP 100) used as simulants of radioactive liquid organic waste (RLOW) in a specifically developed mix fly ash, blast furnace slag and metakaolin based geopolymer was studied in the present work. The process consists of obtaining the geopolymer paste slurry, produced by dispersing the solid precursors in the aqueous alkaline solution, and then adding RLOW via direct incorporation into the slurry under mixing to create an emulsion, before the geopolymer hardens. Geopolymer/oil composites have been prepared with various oil content (10, 20, 30 and 40 %v.), and subsequently characterized to verify their compliance with basic waste acceptance criteria. The positive role of the addition of a superplasticizer, to improve the fluidity of the paste, the density, and the homogeneity of the structure of geopolymer hardened materials was also demonstrated. The mechanical and engineering properties of the pastes and of solidified materials have been verified via rheological measurements and compressive strength tests. The optimized reference formulation loaded with 30 %v. oil waste has been tested in terms of raw materials variability and mixing proportion as part of a robustness study. Finally, the possibility to incorporate in the developed formulation other surrogated RLOW (tributyl phosphate/dodecane (30/70) and Liquid Scintillation Cocktail) has been studied with promising results.</div></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-22","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324007088/pdfft?md5=e164e843ac9e76abe7764dc921851ec6&pid=1-s2.0-S0029549324007088-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142311658","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":"Evaluation of a hybrid in-vessel retention strategy with ex-vessel cooling for APR1400 under extended station blackout conditions","authors":"","doi":"10.1016/j.nucengdes.2024.113600","DOIUrl":"10.1016/j.nucengdes.2024.113600","url":null,"abstract":"<div><p>The purpose of this study is to examine the success window of a hybrid in-vessel retention (IVR) strategy coupled with ex-vessel cooling (ERVC) under an extended Station Blackout (SBO). The high-power-density reactor, APR-1400, is selected and modelled using the computer code ASYST, to examine the thermal–hydraulic response and evaluate the efficacy of a hybrid IVR-ERVC strategy as the accident progresses. Specifically, the hybrid IVR-ERVC strategy refers to combining in-vessel injection as well as ex-vessel cooling to maintain the vessel integrity. Naturally, depressurization of the pressure vessel, which is a precursor to the in-vessel injection, is also applied. The hybrid IVR-ERVC strategy is meant to mitigate the accident and prevent a vessel breach using a set of operator actions within the framework of severe accident management guidelines (SAMG), capitalizing on the portable equipment of the Diverse and Flexible (FLEX) strategy. Three high level candidate actions (HLCAs), namely primary-side depressurization and in-vessel injection along with ex-vessel cooling via cavity flooding are systematically implemented to assess their effectiveness in maintaining the vessel’s integrity for a mission time of 72 h. By combining those high level actions, the corium can be cooled both internally as well as externally to avoid the critical heat flux bottleneck.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142272949","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":"Evaluation framework for molten salt reactors and other new nuclear power reactor systems","authors":"","doi":"10.1016/j.nucengdes.2024.113588","DOIUrl":"10.1016/j.nucengdes.2024.113588","url":null,"abstract":"<div><p>In recent years there has been a serious effort throughout many nations to advance new nuclear power reactor designs for commercial deployment. There are many competing technologies classes and specific designs among the technologies. The primary objective of this study is to provide a framework to evaluate and ultimately optimize reactor designs, on a cost basis.</p><p>Although the framework is generally technology independent, it is presented as it pertains to one particular reactor type, Molten Salt Reactors (MSRs). For MSRs the framework provides the basis from which to optimize both the salt composition and key geometric parameters. It is broad in scope and is therefore divided into several metrics of performance, direct cost, waste, safety, proliferation, modularity and feasibility (technical difficulty). This novel framework relates reactor design/construction conditions as well as specific configuration parameters to cost, thereby enriching understanding of the costs and trade-offs associated with numerous design characteristics.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-21","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"https://www.sciencedirect.com/science/article/pii/S0029549324006885/pdfft?md5=afbe2fcce90675f2d180e43febb562f1&pid=1-s2.0-S0029549324006885-main.pdf","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142272948","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}