Nuclear Engineering and Design最新文献_第9页

Assessment of loading scheme and moderation impact on minor actinide transmutation in VVER-1000 fuel assembly 评估 VVER-1000 燃料组件中的加载方案和对小锕系元素嬗变的影响

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-18 DOI: 10.1016/j.nucengdes.2024.113591

Md. Nobir Hosen, Md. Hossain Sahadath, H. Rainad Khan Rohan

{"title":"Assessment of loading scheme and moderation impact on minor actinide transmutation in VVER-1000 fuel assembly","authors":"Md. Nobir Hosen, Md. Hossain Sahadath, H. Rainad Khan Rohan","doi":"10.1016/j.nucengdes.2024.113591","DOIUrl":"10.1016/j.nucengdes.2024.113591","url":null,"abstract":"<div><p>This study aims to investigate the impact of loading scheme and moderation on minor actinide (MA) transmutation in a thermal-spectrum VVER-1000 reactor. Six different assembly models, consisting of inner-coated, outer-coated and homogeneously mixed MA-containing fuel elements were simulated in OpenMC for a period of 600 EFPDs. Three of the models employed fuel elements cooled by light-water from the outside only, while the other three were dual-cooled. A uniform loading of minor actinides in each assembly was ensured. All models demonstrated successful transmutation of Np and Am isotopes, while the concentration of Cm increased with burnup. Model 5 with homogeneously mixed MAs exhibited the highest overall transmutation rate (TR) of approximately 23.139%/y, followed by Model 2 (21.405%/y) and Model 1 (21.202%/y) utilizing inner-coated MAs. The loading of MAs as outer coatings around fuel elements proved to be the least effective approach for transmutation. Additional moderation via dual cooling resulted in a slight decrease of TR, except for Model 2 with inner-coating of MAs directly in contact with the coolant flowing inside. MA transmutation was also accompanied by a penalty in fuel cycle performance, with Model 5 suffering a 27.85% reduction in discharge burnup and a 27.16% decrease in cycle length compared to the reference. However, both the fuel and moderator temperature coefficients of reactivity became more negative with minor actinide loading and increased moderation. The introduction of MAs also affected the power distribution in the assembly, with homogenous loadings raising the power peaking factor the most.</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":"142243432","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 passive residual heat removal capacity of supercritical carbon dioxide 超临界二氧化碳被动余热去除能力研究

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-16 DOI: 10.1016/j.nucengdes.2024.113592

Gonghao Lu , Guangxu Zhang , Chao Jin , Jiajun Tang , Rongshun Xie , Gang Hong , Yaoli Zhang

{"title":"Research on the passive residual heat removal capacity of supercritical carbon dioxide","authors":"Gonghao Lu , Guangxu Zhang , Chao Jin , Jiajun Tang , Rongshun Xie , Gang Hong , Yaoli Zhang","doi":"10.1016/j.nucengdes.2024.113592","DOIUrl":"10.1016/j.nucengdes.2024.113592","url":null,"abstract":"<div><p>To investigate the passive residual heat removal capacity of supercritical carbon dioxide (S-CO<sub>2</sub>), this study takes a 12MWe lead–bismuth fast reactor as an example and analyzes two indirect S-CO<sub>2</sub> passive residual heat removal system (PRHRS) designs (design 1 based on steam generator (SG) and design 2 based on independent heat exchanger (IHEX)). One-dimensional modeling was conducted using the Modelica language, and detailed analysis was performed on important components. The passive residual heat removal capacity of S-CO<sub>2</sub> was evaluated through simulation calculations. The research results show that the maximum temperature of relevant design based on SG and IHEX does not exceed the pipeline design basis. Both design can reasonably and effectively remove the residual heat from the reactor core. The original equipment of the S-CO<sub>2</sub> Brayton recompression cycle can be directly used to remove residual heat without compromising the relatively simple arrangement. The reasons for the occurrence of peak flow rate are discussed. Meanwhile, an analysis of the outlet temperature of the loop under different pressures in the SG design concludes that the operating pressure of the scheme should exceed 8.5 MPa.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142243428","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

Valorization of waste feather fiber: One uranium resource recycling material 废弃羽毛纤维的价值评估：一种铀资源回收材料

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-16 DOI: 10.1016/j.nucengdes.2024.113596

Xuchai Shan , Xinyu Shi , Xiaohong Tan , Yadong Pu , Taotao Qiang

引用次数: 0

Experimental and numerical examination of low cycle fatigue behaviour on AISI304L steel AISI304L 钢低循环疲劳行为的实验和数值检验

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-16 DOI: 10.1016/j.nucengdes.2024.113599

Sushant Bhalchandra Pate , Gintautas Dundulis , Sigitas Kilikevičius , Albertas Grybenas

{"title":"Experimental and numerical examination of low cycle fatigue behaviour on AISI304L steel","authors":"Sushant Bhalchandra Pate , Gintautas Dundulis , Sigitas Kilikevičius , Albertas Grybenas","doi":"10.1016/j.nucengdes.2024.113599","DOIUrl":"10.1016/j.nucengdes.2024.113599","url":null,"abstract":"<div><p>The integrity of the components of nuclear power plants must be guaranteed under operating and emergency conditions. It is necessary to evaluate all possible degradation mechanisms that could impact the structural integrity of nuclear power plant structures such as the pipelines and other components of the cooling systems. Environmental fatigue significantly influences the degradation mechanisms of steel components operating in water, which eventually affects the operational lifetime of the components. Exploring non-codified methods for more precise assessment of fatigue phenomena can pave the way for developing safer nuclear power plants with longer operational lifetimes. This is very important as the global demand for cleaner energy is increasing. This research study deals with a numerical investigation of the low-cycle fatigue behaviour of steel used for those nuclear reactor components where environmental fatigue plays a major role. The numerical simulation methodology is proposed to study the dependence of the kinematic hardening parameter on the strain amplitude to investigate the low-cycle fatigue behaviour of AISI 304L steel for the prediction of the fatigue curves that can be used for the estimation of nuclear power plant safety and its lifetime.</p><p>The experimental data was used to estimate the parameters required to define the material model for the numerical simulation and to validate the results of the numerical simulation of the low cycle fatigue behaviour. The presented methodology can be used for fully reversed constant-amplitude strain loading low cycle fatigue simulations for various strain ranges to predict the low-cycle fatigue behaviour of steel under repetitive loading.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142243431","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

Deep learning driven inverse solving method for neutron diffusion equations and three-dimensional core power reconstruction technology 深度学习驱动的中子扩散方程逆求解方法与三维核心动力重构技术

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-16 DOI: 10.1016/j.nucengdes.2024.113590

Dong Liu , Bin Zhang , Yong Jiang , Ping An , Zhang Chen

{"title":"Deep learning driven inverse solving method for neutron diffusion equations and three-dimensional core power reconstruction technology","authors":"Dong Liu , Bin Zhang , Yong Jiang , Ping An , Zhang Chen","doi":"10.1016/j.nucengdes.2024.113590","DOIUrl":"10.1016/j.nucengdes.2024.113590","url":null,"abstract":"<div><p>Online monitoring of nuclear reactor core plays a significant role in safe-operation and economic improvement of nuclear power plant. In the process of reactor online monitoring, limited amount of the timing measured data inside and outside the reactor will be used to solve the core power distribution. The traditional methods such as interpolation and harmonic-based methods still have room for improvement in power reconstruction accuracy and robustness. This paper introduces the basic principle of solving neutron diffusion equation and the general framework of power reconstruction driven by deep learning techniques. This method has good performances in online monitoring, even under the conditions of limited measurement data, missing boundary conditions, and partial detector failure. The key techniques of multi-source data fusion, inverse solution of diffusion equations, and detector failure correction with the actual boundary condition missing are proposed in the work. We conducted several standard benchmarks to confirm the accuracy of the solution to neutron diffusion equations based on deep learning method. Additionally, we validated the new technique for power reconstruction, demonstrating its accuracy and effectiveness through an engineering problem simulation. Hence, a new technical approach for reactor core power monitoring is explored in this work.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-16","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142243430","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

Uncertainty analysis of FLiNaK thermophysical properties on convective heat transfer characteristics in circular tube FLiNaK 热物理性质对圆管中对流传热特性的不确定性分析

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-13 DOI: 10.1016/j.nucengdes.2024.113585

Haoyang Li , Qiunan Sun , Ming Ding , Zehua Guo

{"title":"Uncertainty analysis of FLiNaK thermophysical properties on convective heat transfer characteristics in circular tube","authors":"Haoyang Li , Qiunan Sun , Ming Ding , Zehua Guo","doi":"10.1016/j.nucengdes.2024.113585","DOIUrl":"10.1016/j.nucengdes.2024.113585","url":null,"abstract":"<div><p>Molten salts have attracted a spate of quantity of interests in energy and chemistry fields due to their large volumetric capacity, and thermodynamic stability at elevated temperature. However, there are significant uncertainties in the physical properties of molten salts, such as density, specific heat capacity, dynamic viscosity, and thermal conductivity, due to inexperienced measurement techniques and technology. These uncertainties could have certain impact on the evaluation of molten salts heat transfer characteristics. Therefore, an uncertainty analysis is performed regarding FLiNaK thermophysical properties on convective heat transfer characteristics in circular tube. Then, a polynomial chaos expansion (PCE) method is performed to study the uncertainty affecting the heat transfer characteristic. In the process of uncertainty analysis, tensor product quadrature nodes are used to calculate representative sample points and reduce computational costs. Three primary physical properties of molten salt are taken into account as the input parameters. The Sobol composition method is also used to analyze the contributions of each parameter to heat transfer. The results of the uncertainty analysis suggest that the uncertainties in dynamic viscosity, thermal conductivity, and specific heat capacity have a significant impact on the heat transfer of molten salt, contributing to 80 %, 14 %, and 6 % of the total variability, respectively. It also suggests that the polynomial chaos expansion methodology is both novel and reliable when applied to uncertainty analysis of molten salt.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142171874","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

Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO2 针对掺杂铬的二氧化铀释放裂变气体的单独效应辐照试验的建模和设计

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-13 DOI: 10.1016/j.nucengdes.2024.113571

Jacob P. Gorton , Annabelle G. Le Coq , Zane G. Wallen , Christian M. Petrie , Joshua T. White , John T. Dunwoody , Shane Mann , Nathan A. Capps , Andrew T. Nelson

{"title":"Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO2","authors":"Jacob P. Gorton , Annabelle G. Le Coq , Zane G. Wallen , Christian M. Petrie , Joshua T. White , John T. Dunwoody , Shane Mann , Nathan A. Capps , Andrew T. Nelson","doi":"10.1016/j.nucengdes.2024.113571","DOIUrl":"10.1016/j.nucengdes.2024.113571","url":null,"abstract":"<div><p>Fission gas release (FGR) from nuclear fuel during operation can diminish heat transfer properties across the pellet-cladding gap and increase the fuel rod internal pressure, thereby posing a concern to fuel reliability and safety during an accident. Enlarging the fuel grain size, which has been shown to improve fission gas retention, can be achieved by doping the fuel feedstock prior to sintering. In this work, the BISON fuel performance code was used to predict FGR from undoped and chromia-doped UO<sub>2</sub> (referred to as <em>Cr-doped UO<sub>2</sub></em>) fuel specimens with different grain sizes and across various temperatures. The BISON models identified the irradiation conditions for which FGR is most significant, and a separate effects irradiation experiment in the High Flux Isotope Reactor (HFIR) was then developed targeting those conditions. The experiment leveraged the MiniFuel irradiation capability at Oak Ridge National Laboratory and consisted of 12 fuel specimens of varying grain size and Cr content. A coupling scheme between BISON FGR results and the ANSYS finite element thermal model used for experiment design was formulated to predict cumulative FGR from each fuel specimen based on expected irradiation temperature histories. The fuel samples were fabricated and characterized as a part of this work, and the fuel compositions modeled in BISON were representative of the specimens used in the experiment. This combined modeling and experimental effort aims to study the effect of fuel grain size and Cr content on FGR and to provide simulated BISON FGR results that can be used for future model validation activities.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142171873","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

Enhancing accuracy of prediction of critical heat flux in Circular channels by ensemble of deep sparse autoencoders and deep neural Networks 利用深度稀疏自动编码器和深度神经网络的集合提高圆形通道临界热通量预测的准确性

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-13 DOI: 10.1016/j.nucengdes.2024.113587

Rehan Zubair Khalid , Ibrahim Ahmed , Atta Ullah , Enrico Zio , Asifullah Khan

{"title":"Enhancing accuracy of prediction of critical heat flux in Circular channels by ensemble of deep sparse autoencoders and deep neural Networks","authors":"Rehan Zubair Khalid , Ibrahim Ahmed , Atta Ullah , Enrico Zio , Asifullah Khan","doi":"10.1016/j.nucengdes.2024.113587","DOIUrl":"10.1016/j.nucengdes.2024.113587","url":null,"abstract":"<div><p>Accurate prediction of Critical Heat Flux (CHF) is essential for ensuring safety and economic efficiency of water-cooled reactors and two-phase flow boiling heat transfer systems. However, the lack of a deterministic theory for CHF prediction remains a significant challenge in the thermal engineering domain. This has led to the development of numerous prediction models based on various CHF experimental data, with no single universally acceptable model covering the wide range of flow conditions encountered in practice. In this paper, we explore the use of a comprehensive CHF experimental dataset in conjunction with artificial intelligence techniques to predict CHF in vertical tubes, contributing to the ongoing efforts to address this critical issue. The proposed method stands on the collection of comprehensive CHF experimental data from various sources, covering a wide range of operating conditions (pressure of 100 – 21,197 kPa, hydraulic diameters of 1 – 44.7 mm, mass fluxes of 10 – 20,910 kg/m<sup>2</sup>s, inlet-subcooling of 0.6 – 3,555 kJ/kg, heated lengths of 9 – 6,000 mm and critical qualities of −0.494 – 0.981), and is based on a new prediction model for the prediction of the CHF. Specifically, the prediction model consists of an ensemble of deep sparse autoencoders (AEs) used as a base-learner to extract robust features from the input data and a deep neural network (DNN) built on top of the ensemble of deep sparse AEs for use as a <em>meta</em>-learner to predict the CHF. The proposed method is validated on the collected CHF data and the obtained results show a substantial improvement in CHF prediction accuracy, outperforming standalone and other state of-the-art machine learning models. This innovative approach offers a notable improvement in CHF prediction, potentially contributing to the development of more reliable and efficient nuclear reactors.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142230024","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

Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark 利用 BEAVRS 基准验证 DDC-3D 代码系统的中子和热液压耦合分析功能

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-12 DOI: 10.1016/j.nucengdes.2024.113583

Binhang Zhang , Zenghao Liu , Xianbao Yuan , Yonghong Zhang , Jianjun Zhou , HaiBo Tang , Yunlong Xiao

{"title":"Validation of DDC-3D code system for neutronics and thermal-hydraulics coupling analysis using BEAVRS benchmark","authors":"Binhang Zhang , Zenghao Liu , Xianbao Yuan , Yonghong Zhang , Jianjun Zhou , HaiBo Tang , Yunlong Xiao","doi":"10.1016/j.nucengdes.2024.113583","DOIUrl":"10.1016/j.nucengdes.2024.113583","url":null,"abstract":"<div><p>The direct whole-core calculations can provide accurate results and insights to the physics phenomena of the reactor. It can also capture the local effects of temperature and density fields on fuel depletion. However, the computational cost of the direct whole-core calculations is expensive. To compromise between computational cost and accuracy, the DDC-3D code system has been developed to perform neutronics and thermal-hydraulics coupling analysis. The DDC-3D code system couples the open-source codes DRAGON & DONJON based on two-step method and subchannel code COBRA-EN. The Picard iteration method is applied to ensure the stability of numerical calculation. Then the BEAVRS benchmark is used to validate the computational capabilities of DDC-3D code system. The critical boron concentrations, control rod worths and fission rate distributions are calculated in HZP condition. The results show a good agreement with measured data. The results demonstrate that the two-step method is applicable and valid for multiphysics simulations. For the result of HFP condition for cycle 1, the results also agree well with measured data, including the trend of the critical boron concentrations and power distributions throughout the cycle 1. Although the detailed thermal–hydraulic experimental values are not available, the thermal-hydraulics analysis of the hot fuel assemblies indicates that the calculation results are reasonable. In general, the results demonstrate the feasibility and accuracy of DDC-3D code system for neutronics and thermal-hydraulics coupling calculations and life cycle simulation of PWRs.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142171871","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

Highly-detailed neutronic and thermal-hydraulic coupled calculations for OPAL reactor using diverse codes and approaches 使用不同的代码和方法对 OPAL 反应堆进行高度详细的中子和热液耦合计算

IF 1.9 3区工程技术

Nuclear Engineering and Design Pub Date : 2024-09-12 DOI: 10.1016/j.nucengdes.2024.113579

Diego Ferraro , Ignacio Ferrari , Alicia Doval , Eduardo Villarino , Basar Ozar

{"title":"Highly-detailed neutronic and thermal-hydraulic coupled calculations for OPAL reactor using diverse codes and approaches","authors":"Diego Ferraro , Ignacio Ferrari , Alicia Doval , Eduardo Villarino , Basar Ozar","doi":"10.1016/j.nucengdes.2024.113579","DOIUrl":"10.1016/j.nucengdes.2024.113579","url":null,"abstract":"<div><p>The industry-standard approach for designing and operating research reactors cores relies on well-established methodologies that consider uncoupled neutronic calculations and a subchannel analysis of the Thermal-Hydraulic (TH) associated problem. Advancements in computing power and codes allow detailed multiphysics approaches to be implemented, thereby reducing conservatism. In this study, a comparative analysis of results from diverse detailed neutronic-TH coupled core approaches is developed. To address a realistic application case, the comparison is made for a reported critical configuration from the Open Pool Australian Lightwater research reactor (OPAL) at Hot Full Power (HFP) and low burnup. Both cell-core and stochastic methodologies for neutronics are evaluated, whereas two different subchannel codes are considered for TH. The convergence of the coupled schemes, and the consistency of the main parameters are discussed, showing the compatibility of the alternative methods and their ability to offer critical insights not captured by standard practices.</p></div>","PeriodicalId":19170,"journal":{"name":"Nuclear Engineering and Design","volume":null,"pages":null},"PeriodicalIF":1.9,"publicationDate":"2024-09-12","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"142171872","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