Nuclear Engineering and Design最新文献_第8页

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 通过耦合中子学和遗传算法计算金属燃料钠快堆轴向异质堆芯设计中钠空隙值和孕育比的帕累托前沿

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-24 DOI: 10.1016/j.nucengdes.2024.113598

Hui Guo , Shengtao Han , Xin Jin , Yiwei Wu , Qufei Song , Yao Xiao , Hanyang Gu

{"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":"Hui Guo , Shengtao Han , Xin Jin , Yiwei Wu , Qufei Song , Yao Xiao , Hanyang Gu","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}

引用次数: 0

Numerical simulation of corium flow through rod bundle and/or debris bed geometries with a model based on Lattice Boltzmann method 利用基于晶格玻尔兹曼法的模型，对流经棒束和/或碎片床几何形状的铈流进行数值模拟

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-24 DOI: 10.1016/j.nucengdes.2024.113603

J. Garcia Sarmiento , F. Fichot , V. Topin , P. Sagaut

{"title":"Numerical simulation of corium flow through rod bundle and/or debris bed geometries with a model based on Lattice Boltzmann method","authors":"J. Garcia Sarmiento , F. Fichot , V. Topin , P. Sagaut","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}

引用次数: 0

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 计算流体动力学研究 6% 爬升压力管对单个 CANDU 37M 燃料束的流动特性、燃料温度和压力管壁温度的影响

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-23 DOI: 10.1016/j.nucengdes.2024.113593

Z. Lu , M.H.A. Piro

{"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":"Z. Lu , M.H.A. Piro","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}

引用次数: 0

Incorporation of organic liquid waste in alkali activated mixed fly ash/blast furnace slag/metakaolin-based geopolymers 在碱活化混合粉煤灰/高炉矿渣/高岭土基土工聚合物中掺入有机液体废物

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-22 DOI: 10.1016/j.nucengdes.2024.113608

Sergey Sayenko , Yevhenii Svitlychnyi , Volodymyr Shkuropatenko , Federica Pancotti , Simona Sandalova , Arnaud Poulesquen , Isabelle Giboire , Abdelaziz Hasnaoui , Davide Cori , Gabriele Magugliani , Eros Mossini

{"title":"Incorporation of organic liquid waste in alkali activated mixed fly ash/blast furnace slag/metakaolin-based geopolymers","authors":"Sergey Sayenko , Yevhenii Svitlychnyi , Volodymyr Shkuropatenko , Federica Pancotti , Simona Sandalova , Arnaud Poulesquen , Isabelle Giboire , Abdelaziz Hasnaoui , Davide Cori , Gabriele Magugliani , Eros Mossini","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}

引用次数: 0

Evaluation of a hybrid in-vessel retention strategy with ex-vessel cooling for APR1400 under extended station blackout conditions 评估在长期停电条件下为 APR1400 采用的混合舱内保留和舱外冷却策略

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-21 DOI: 10.1016/j.nucengdes.2024.113600

Saja Rababah , Aya Diab

{"title":"Evaluation of a hybrid in-vessel retention strategy with ex-vessel cooling for APR1400 under extended station blackout conditions","authors":"Saja Rababah , Aya Diab","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}

引用次数: 0

Evaluation framework for molten salt reactors and other new nuclear power reactor systems 熔盐反应堆和其他新型核能反应堆系统的评估框架

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-21 DOI: 10.1016/j.nucengdes.2024.113588

Elliott J.T. Berg, Adriaan Buijs

{"title":"Evaluation framework for molten salt reactors and other new nuclear power reactor systems","authors":"Elliott J.T. Berg, Adriaan Buijs","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}

引用次数: 0

Numerical appraisal of the role of heat transfer regimes on transient response of carbon dioxide based supercritical natural circulation loop during power upsurge 基于二氧化碳的超临界自然循环回路在电力高峰期的瞬态响应中传热机制作用的数值评估

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-21 DOI: 10.1016/j.nucengdes.2024.113601

Tanuj Srivastava , Ashok Kumar Gond , Dipankar N. Basu

{"title":"Numerical appraisal of the role of heat transfer regimes on transient response of carbon dioxide based supercritical natural circulation loop during power upsurge","authors":"Tanuj Srivastava , Ashok Kumar Gond , Dipankar N. Basu","doi":"10.1016/j.nucengdes.2024.113601","DOIUrl":"10.1016/j.nucengdes.2024.113601","url":null,"abstract":"<div><div>Appearance of steep property gradients with change in temperature is a fascinating feature of any supercritical fluid, which can instigate intricate dynamics in supercritical natural circulation loops by modulating the effective forces. While most of the relevant literature focuses on stability evaluation, anticipation regarding the transient response of the system during power transition is of utmost significance, especially in high-power applications. Present study aims at furnishing insight on the same by developing a one-dimensional numerical model of a rectangular loop with supercritical CO<span><math><msub><mrow></mrow><mrow><mn>2</mn></mrow></msub></math></span> as the working medium, and characterizing the temporal trends over a wide range of heating power. Two different profiles of power upsurge have been tested for different regimes of heat transfer, unearthing intriguing characteristics. The combination of initial and final regimes during any transformation is found to be the most crucial factor. Single-step rise in power, in general, is the most vulnerable one, specifically during large-scale change of the order of 1000 W, and better be employed only at low-power regime. Even single-step change <span><math><mo>∼</mo></math></span> 25 W can inflict instability and flow transition within the transition regime. Power transformation following linear ramp profile with transition periods of 5, 10 and 20 s is identified to be the most suitable one across all the regimes. It can successfully mitigate instability even in the later parts of the transition regime, albeit at the expense of greater time requirement to attain the final stable state and possibly a greater period of transformation. Change through multiple small steps (about 250 W for large change in low power regime and 15 W within transition regime) can also be a feasible option for avoiding the growth of unstable oscillations at higher powers.</div></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":"142311659","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}

引用次数: 0

Research on the corrosion resistance of porous coatings for heat transfer enhancement on the head of nuclear reactor pressure vessels 用于增强核反应堆压力容器头部传热的多孔涂层的耐腐蚀性研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-20 DOI: 10.1016/j.nucengdes.2024.113597

Ming Jiao , Li Zhang , Ping Cheng , Mingguang Zheng , Xiaoqiang Liu

{"title":"Research on the corrosion resistance of porous coatings for heat transfer enhancement on the head of nuclear reactor pressure vessels","authors":"Ming Jiao , Li Zhang , Ping Cheng , Mingguang Zheng , Xiaoqiang Liu","doi":"10.1016/j.nucengdes.2024.113597","DOIUrl":"10.1016/j.nucengdes.2024.113597","url":null,"abstract":"<div><p>The preparation of a porous coating on the outer surface of nuclear power reactor pressure vessel’s lower head is considered an effective measure to enhance heat transfer and ensure safety under the condition of In-Vessel Retention (IVR) during severe accidents. However, oxidation and corrosion of the porous coating are inevitable, as it will be exposed to the marine atmosphere during the vessel’s construction and operation phases. Therefore, it is necessary to strengthen its corrosion resistance under the premise of high critical heat flux (CHF). In this paper, based on the successful development of high CHF porous coatings, a composite coating consisting of a Ni-Cr bottom layer and a 316L stainless steel porous coating was designed and prepared by flame spraying on SA-508 Gr.3 steel. The corrosion resistance of the composite coating was evaluated by Neutral Salt Spray Test (NSS) and the corrosion mechanism of the coating was analyzed. The results showed that the Ni-Cr bottom coating exhibited no obvious corrosion after 240 h of the NSS test, but the corrosion resistance of the porous coating was decreased due to surface oxidation, chromium deficiency in the particle melting zone and the presence of ferrite. By incorporating corrosion-resistant alloying elements into 316L powders, the corrosion resistance of the coating can be significantly improved while preserving its porosity. The investigation of porous composite coatings for reactor pressure vessels is a critical endeavor with the potential to significantly enhance the performance, safety, and longevity of nuclear reactors.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-20","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142261453","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}

引用次数: 0

Various additives for improving the performances of solidified forms of radioactive wastes by cementation technology: Recent advances and perspectives 通过胶结技术改善放射性废物固化形式性能的各种添加剂：最新进展和前景

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-19 DOI: 10.1016/j.nucengdes.2024.113595

Lei Chen , Junfeng Li , Jianlong Wang

{"title":"Various additives for improving the performances of solidified forms of radioactive wastes by cementation technology: Recent advances and perspectives","authors":"Lei Chen , Junfeng Li , Jianlong Wang","doi":"10.1016/j.nucengdes.2024.113595","DOIUrl":"10.1016/j.nucengdes.2024.113595","url":null,"abstract":"<div><p>The additives are crucial for improving the performance of the solidified forms of radioactive wastes by cementation technology. This paper reviewed the recent research advances in improving the cement and waste solidification performances using various additives, including the traditional pozzolanic active substances, biochar, fibers, nanomaterials and water reducer. The enhanced effect of various additives on the compressive strength, flexural strength, and nuclide leaching resistance, which belong to the performance requirements of the radioactive waste cement solidified forms, was summarized and analyzed. Silica fume, slag and fly ash can provide higher compressive strength, while zeolite, metakaolin, bentonite and activated carbon can reduce nuclide leaching. Fibers can significantly increase the flexural strength. Nanomaterials can provide superior mechanical properties, but currently there is little research and application in the waste cementation field. The water reducer can provide the necessary fluidity for the above multiple cement-based mixed materials. This review will deepen the understanding of solidification of radioactive waste and provide references for specific cementation formulations.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142243434","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}

引用次数: 0

Intelligent optimization of power distribution for fast reactor NCLFR-Oil based on SPN method 基于 SPN 方法的快堆 NCLFR-Oil 功率分配智能优化

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-18 DOI: 10.1016/j.nucengdes.2024.113580

Shaoning Shen, Wenshun Duan, Weixiang Wang, Aoguang Wu, Kefan Zhang, Hongli Chen

{"title":"Intelligent optimization of power distribution for fast reactor NCLFR-Oil based on SPN method","authors":"Shaoning Shen, Wenshun Duan, Weixiang Wang, Aoguang Wu, Kefan Zhang, Hongli Chen","doi":"10.1016/j.nucengdes.2024.113580","DOIUrl":"10.1016/j.nucengdes.2024.113580","url":null,"abstract":"<div><p>A custom-developed neutron transport module based on the COMSOL finite element solver was created to enable efficient optimization and parameter evaluation in core design, and it can be integrated with other built-in modules for enhanced capabilities. This work began by establishing a practical foundation for a multi-dimensional SP<sub>N</sub> method using the PDE solver, capable of simulating both steady-state (k-eigenvalue) and time-dependent transport problems. The steady-state solver showed good agreement with 3D TAKEDA and 2D C5G7 benchmarks, while the transient solver was well-validated with TWIGL and LMW benchmarks. For modeling the self-designed fast reactor NCLFR-Oil, OpenMC was used to generate few-group constants, which were then imported into COMSOL’s SP<sub>3</sub> neutron transport module as equation coefficients. The SP<sub>3</sub> model’s capability to simulate the core’s physical field was validated by testing eigenvalues, control rod worth, the power and neutron flux distribution. Sensitivity analysis was performed using COMSOL’s uncertainty quantification module to assess the impact of control rod positions on core eigenvalues and power distribution, refining the parameter space for optimization and enhancing efficiency. To further improve optimization efficiency, a surrogate model based on “Polynomial Chaos Expansion” was employed to approximate the core’s physical model, predicting relationships between input parameters and optimization objectives. This model proved more efficient than the gradient-free “Coordinate Search” method, reducing computational resource consumption. The optimization results showed a significant reduction in the custom power flattening factor, bringing more power factors closer to the target value of 1.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142243433","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}

引用次数: 0