Thermal efficiency and safety of Generation-IV nuclear-power-reactor concept - Supercritical Water-cooled Reactor (SCWR) depend on solid knowledge of specifics of SCW thermophysical properties and heat transfer within these conditions. As a preliminary, but conservative approach to uncover these specifics is analysis of experimental data obtained in bare tubes including numerical investigation. This paper presents the numerical investigation, based on computational fluid dynamics, of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm). The flow and heat-transfer mechanism of SCW in the vertical tube under the influence of buoyancy and flow acceleration are analyzed. Results of numerical simulation predict the experimental data with reasonable accuracy. The results indicated that in the region of q/G > 0.4 kJ/kg, the wall temperature distribution tends to be non-linear, and heat transfer may deteriorate. When Tb < Tpc < Tw, internal wall temperature shows peaks, which corresponds to heat-transfer deterioration. Meanwhile the position, where the deterioration occurs is continuously moved forward to the inlet as the heat flux increases. Velocity changes near the wall show an M shape according to mass conservation for the density change.
{"title":"Numerical Investigation on Heat Transfer to Supercritical Water Flowing Upward in a 4-M Long Bare Vertical Circular Tube","authors":"Dong-Keun Yang, Qixian Wu, Lin Chen, I. Pioro","doi":"10.1115/icone2020-16456","DOIUrl":"https://doi.org/10.1115/icone2020-16456","url":null,"abstract":"\u0000 Thermal efficiency and safety of Generation-IV nuclear-power-reactor concept - Supercritical Water-cooled Reactor (SCWR) depend on solid knowledge of specifics of SCW thermophysical properties and heat transfer within these conditions. As a preliminary, but conservative approach to uncover these specifics is analysis of experimental data obtained in bare tubes including numerical investigation. This paper presents the numerical investigation, based on computational fluid dynamics, of the heat-transfer characteristics of SCW flow in a 4-m long circular tube (ID = 10 mm). The flow and heat-transfer mechanism of SCW in the vertical tube under the influence of buoyancy and flow acceleration are analyzed. Results of numerical simulation predict the experimental data with reasonable accuracy. The results indicated that in the region of q/G > 0.4 kJ/kg, the wall temperature distribution tends to be non-linear, and heat transfer may deteriorate. When Tb < Tpc < Tw, internal wall temperature shows peaks, which corresponds to heat-transfer deterioration. Meanwhile the position, where the deterioration occurs is continuously moved forward to the inlet as the heat flux increases. Velocity changes near the wall show an M shape according to mass conservation for the density change.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"1 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"116883735","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
A. Kraus, E. Merzari, T. Norddine, O. Marin, S. Benhamadouche
Rod bundle flows are commonplace in nuclear engineering, and are present in light water reactors (LWRs) as well as other more advanced concepts. Inhomogeneities in the bundle cross section can lead to complex flow phenomena, including varying local conditions of turbulence. Despite the decades of numerical and experimental investigations regarding this topic, and the importance of elucidating the physics of the flow field, to date there are few publicly available direct numerical simulations (DNS) of the flow in multiple-pin rod bundles. Thus a multiple-pin DNS study can provide significant value toward reaching a deeper understanding of the flow physics, as well as a reference simulation for development of various reduced-resolution analysis techniques. To this end, DNS of the flow in a square 5 × 5 rod bundle at Reynolds number of 19,000 has been performed using the highly-parallel spectral element code Nek5000. The geometrical dimensions were representative of typical LWR fuel designs. The DNS was designed using microscales estimated from an advanced Reynolds-Averaged Navier-Stokes (RANS) model. Characteristics of the velocity field, Reynolds stresses, and anisotropy are presented in detail for various regions of the bundle. The turbulent kinetic energy budget is also presented and discussed.
棒束流在核工程中很常见，并且存在于轻水反应堆(LWRs)以及其他更先进的概念中。束截面的不均匀性会导致复杂的流动现象，包括不同的局部湍流条件。尽管数十年来关于该主题的数值和实验研究以及阐明流场物理特性的重要性，但迄今为止，很少有公开可用的多针杆束流动的直接数值模拟(DNS)。因此，多引脚DNS研究可以为更深入地理解流体物理提供重要价值，并为开发各种低分辨率分析技术提供参考模拟。为此，利用高度平行谱元代码Nek5000，对雷诺数为19000的方形5 × 5杆束内的流动进行了DNS分析。其几何尺寸具有典型轻水堆燃料设计的代表性。DNS采用先进的reynolds - average Navier-Stokes (RANS)模型估算的微观尺度进行设计。详细介绍了束束各区域的速度场、雷诺应力和各向异性特征。本文还提出并讨论了紊流动能收支。
{"title":"Direct Numerical Simulation of Fluid Flow in a 5x5 Square Rod Bundle Using Nek5000","authors":"A. Kraus, E. Merzari, T. Norddine, O. Marin, S. Benhamadouche","doi":"10.1115/icone2020-16643","DOIUrl":"https://doi.org/10.1115/icone2020-16643","url":null,"abstract":"\u0000 Rod bundle flows are commonplace in nuclear engineering, and are present in light water reactors (LWRs) as well as other more advanced concepts. Inhomogeneities in the bundle cross section can lead to complex flow phenomena, including varying local conditions of turbulence. Despite the decades of numerical and experimental investigations regarding this topic, and the importance of elucidating the physics of the flow field, to date there are few publicly available direct numerical simulations (DNS) of the flow in multiple-pin rod bundles. Thus a multiple-pin DNS study can provide significant value toward reaching a deeper understanding of the flow physics, as well as a reference simulation for development of various reduced-resolution analysis techniques. To this end, DNS of the flow in a square 5 × 5 rod bundle at Reynolds number of 19,000 has been performed using the highly-parallel spectral element code Nek5000. The geometrical dimensions were representative of typical LWR fuel designs. The DNS was designed using microscales estimated from an advanced Reynolds-Averaged Navier-Stokes (RANS) model. Characteristics of the velocity field, Reynolds stresses, and anisotropy are presented in detail for various regions of the bundle. The turbulent kinetic energy budget is also presented and discussed.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"72 2 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2020-07-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"132435486","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
Recent advances in horizontal cask designs for commercial spent nuclear fuel have significantly increased maximum thermal loading. This is due in part to greater efficiency in internal conduction pathways. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating thermal-hydraulic models of these storage cask designs. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system. The purpose of this investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to calculate cladding temperatures and induced cooling air flows in modern, horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and complement the existing knowledge base. Transverse and axial temperature profiles along with induced-cooling air flow are measured using various backfills of gases for a wide range of decay powers and canister pressures. The data from the HDCS tests will be used to host a blind model validation effort.
{"title":"Thermal-Hydraulic Investigations of a Horizontal Dry Cask Simulator","authors":"S. Durbin, E. Lindgren, Ramon Pulido, A. Salazar","doi":"10.2172/1607735","DOIUrl":"https://doi.org/10.2172/1607735","url":null,"abstract":"\u0000 Recent advances in horizontal cask designs for commercial spent nuclear fuel have significantly increased maximum thermal loading. This is due in part to greater efficiency in internal conduction pathways. Carefully measured data sets generated from testing of full-sized casks or smaller cask analogs are widely recognized as vital for validating thermal-hydraulic models of these storage cask designs. While several testing programs have been previously conducted, these earlier validation studies did not integrate all the physics or components important in a modern, horizontal dry cask system.\u0000 The purpose of this investigation is to produce data sets that can be used to benchmark the codes and best practices presently used to calculate cladding temperatures and induced cooling air flows in modern, horizontal dry storage systems. The horizontal dry cask simulator (HDCS) has been designed to generate this benchmark data and complement the existing knowledge base.\u0000 Transverse and axial temperature profiles along with induced-cooling air flow are measured using various backfills of gases for a wide range of decay powers and canister pressures. The data from the HDCS tests will be used to host a blind model validation effort.","PeriodicalId":414088,"journal":{"name":"Volume 3: Student Paper Competition; Thermal-Hydraulics; Verification and Validation","volume":"17 1","pages":"0"},"PeriodicalIF":0.0,"publicationDate":"2019-09-01","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"128720932","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":0,"RegionCategory":"","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}
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