核工程研究与设计最新文献

{"title":"Numerical Analysis of Single-Phase Thermal Hydraulic Parameters Along Nanostructured Coating Film","authors":"Omar S. Al-Yahia, Yacine Addad, Ho Joon Yoon, S. Cho","doi":"10.1115/icone2020-16316","DOIUrl":"https://doi.org/10.1115/icone2020-16316","url":null,"abstract":"\u0000 In typical pressurized water reactors, zirconium alloys are used as cladding material for the fuel. However, zircalloy is known to face problems with the high temperature steam, due to the chemical process of oxidation, the oxygen molecules will be separated from the water molecules of the coolant leading to hydrogen gas releases. Recently, a research team at KAIST, South Korea suggested a methodology to fabricate nanoporous oxide layer with the aim of preventing the zircalloy outer surface from reacting with the coolant. Although, this new proposal offers a better solution to prevent the potential hydrogen gas generation, it is still not well understood how the nanoporous-layer is going to affect the convective heat transfer rates between the coolant and the fuel. In fact, on one hand the low conductivity of the oxide layer is expected to reduce the conduction heat transfer within the cladding material; but on the other hand, the nanopores on the oxide layer might act as an effective surface roughness, hence affecting both the hydrodynamic and thermal fields within the coolant channels. In this study, a CFD analysis is carried out to investigate the influence of this nanoporous layer on the convective heat transfer rate and pressure drop coefficient. A detailed 2-D steady-state numerical analysis on single-phase model is performed using Star-CCM+ code. The study is conducted using pores with a diameter of 30 to 100 nm. The results obtained from these predictions are then compared with the ones obtained in the case of the smooth surface. Therefore, the main objectives of the present study are to examine the effect of this nanopourous layer on the thermal hydraulic parameters and to produce the corresponding correlations to be used in the system scale thermal-hydraulic codes.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"84820893","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}

{"title":"Computational Study on the Spherical Laminar Flame Speed of Hydrogen-Air Mixtures","authors":"Nuri Trianti, Kosuke Motegi, T. Sugiyama, Y. Maruyama","doi":"10.1115/icone2020-16841","DOIUrl":"https://doi.org/10.1115/icone2020-16841","url":null,"abstract":"\u0000 The computational fluid dynamics (CFD) have been developed to analyze the correlation equation for laminar flame speed of hydrogen-air mixtures. This analysis was carried out on the combustion of hydrogen-air mixtures performed at the spherical bomb experiment facility consists of a spherical vessel equipped (563 mm internal diameter). The facility has been designed and built at CNRS-ICARE laboratory. The simulation was carried out using the reactingFoam solver, one of a transient chemical reaction solver in OpenFOAM 5.0. The LaunderSharmaKE model was applied for turbulent flow. The interaction of the chemical reaction with the turbulent flow was taken into account using PaSR (Partial Stirred Reactor) model with 19 elementary reactions for the hydrogen combustion. The initial condition of spherical flame analysis was set so as to be consistent with those of the experiment. The position of the flame front was detected by the steep drop of hydrogen mass fraction in the spherical radii, and the flame propagation velocity was estimated from the time-position relationship. The analysis result showed the characteristic of spherical flame acceleration was qualitatively reproduced even though it has a discrepancy with the experiment. After validating the calculation of spherical experiments, a laminar burning velocity correlation is presented using the same boundary conditions with the variation of hydrogen concentration, temperature, and pressure. The calculation of laminar flame speed of hydrogen-air mixtures by reactingFoam use reference temperature Tref = 293 K and reference pressure Pref = 1 atm with validated in the range of hydrogen concentration 6–20%; range of temperature 293–493 K; and range of pressure 1–3 atm.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"86389752","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}

{"title":"Stress Analysis of the Lower Head of Central Measuring Shroud Under Thermal Striping and Thermal Shock Conditions","authors":"Shu Zheng, D. Lu, Q. Cao","doi":"10.1115/icone2020-16668","DOIUrl":"https://doi.org/10.1115/icone2020-16668","url":null,"abstract":"\u0000 The central measuring shroud, as an important in-vessel component, provides guidance and protection for control rods and measuring equipment in a sodium-cooled fast reactor (SFR). The lower head of central measuring shroud (LHCMS), which is located above the core outlet, is only 500mm away from the core outlet. Therefore, the LHCMS is affected by the liquid sodium from core outlet for a long period, especially the temperature effects of the following two types. On the one hand, under the operating condition of the SFRs, the uneven distribution of the core power causes the phenomenon of thermal striping, which may cause high cycle fatigue and even initial crack. On the other hand, under the scram condition, the coolant temperature at the core outlet is sharply reduced due to the decrease of the core power, inducing the phenomenon of thermal shock that may cause large thermal stress and low cycle fatigue. Therefore, stress and fatigue analyses of the LHCMS under the thermal striping and thermal shock conditions are very necessary. In the paper, finite element model of the LHCMS was first established, and then according to the temperature curves under thermal striping conditions and thermal shock conditions, the thermal stress of the LHCMS was simulated. The results showed that although the temperature fluctuation outside the LHCMS is severe, the stress caused by thermal striping only slightly fluctuates at 123MPa level, the maximum stress range is 11MPa. Besides, at 20s, there exists the maximum stress difference between thermal striping and thermal shock conditions, the maximum stress caused by thermal shock is about 3 time larger than that caused by thermal striping. According to high cycle and low cycle fatigue analyses, the fatigue damage factor of thermal striping is only 0.0078, while the fatigue damage factor of thermal shock is 3.416, which should provide a reference for the design of the LHCMS.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88437491","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}

{"title":"Hybrid Nodal Integral/Finite Element Method (NI-FEM) for Time-Dependent Convection Diffusion Equation","authors":"Sundar Namala, R. Uddin","doi":"10.1115/icone2020-16703","DOIUrl":"https://doi.org/10.1115/icone2020-16703","url":null,"abstract":"\u0000 Nodal integral methods (NIM) are a class of efficient coarse mesh method that use transverse averaging to reduce the governing partial differential equation(s) (PDE) into a set of ordinary differential equations (ODE), and these ODEs or their approximations are analytically solved. Since this method depends on transverse averaging, the standard application of this approach gets restricted to domains that have boundaries that are parallel to one of the coordinate axes (2D) or coordinate planes (3D). The hybrid nodal-integral/finite-element method (NI-FEM) has been developed to extend the application of NIM to arbitrary domains. NI-FEM is based on the idea that the interior region and the regions with boundaries parallel to the coordinate axes (2D) or coordinate planes (3D) can be solved using NIM and the rest of the domain can be solved using FEM. The crux of the hybrid NI-FEM is in developing interfacial conditions at the common interfaces between the regions solved by the NIM and the FEM. Since the discrete variables in the two numerical approaches are different, this requires special treatment of the discrete quantities on the interface between the two different types of discretized elements. We here report the development of hybrid NI-FEM in a parallel framework in Fortran using PETSc for the time-dependent convection-diffusion equation (CDE) in arbitrary domains. Numerical solutions are compared with exact solutions, and the scheme is shown to be second order accurate in both space and time. The order of approximations used for the development of the scheme are also shown to be second order. The hybrid method is efficient compared to standalone conventional numerical schemes like FEM.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"88483734","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}

{"title":"Analytical Study on Dynamic Response of Reinforced Concrete Structure With Internal Equipment Subjected to Projectile Impact","authors":"Y. Okuda, Zuoyi Kang, A. Nishida, H. Tsubota, Yinsheng Li","doi":"10.1115/icone2020-16849","DOIUrl":"https://doi.org/10.1115/icone2020-16849","url":null,"abstract":"\u0000 In case of a projectile impact on a reactor building of a nuclear power plant, stress waves propagate from the impacted wall to the structure’s interior. It is important to assess the effect of dynamic responses generated by the projectile’s impact on internal equipment, because stress waves are likely to excite high-frequency vibrations of internal equipment. The OECD (Organization for Economic Co-operation and Development) / NEA (Nuclear Energy Agency) launched the IRIS (Improving Robustness Assessment Methodologies for Structures Impacted by Projectiles) benchmark project in order to assess the dynamic response of a nuclear facility to projectile impact, and the third phase of IRIS (IRIS 3) [1] contributes to the investigation of the dynamic responses of reinforced concrete (RC) structures that house internal equipment. We have participated in IRIS 3 and have performed calibration analyses of projectile impact tests on a structure that models a reactor building that houses internal equipment. Specifically, we have developed and validated a numerical approach to investigation of impact responses of an RC structure that houses internal equipment through calibration correction. This paper presents partial simulation results of the dynamic responses of this structure and discusses the effects of support conditions of the internal equipment and stress wave propagation.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"78976725","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}

{"title":"Modeling Axial Relocation of Fragmented Fuel During Loss of Coolant Conditions by Using ABAQUS","authors":"Zehua Ma, K. Shirvan, Wei Li, Yingwei Wu","doi":"10.1115/icone2020-16291","DOIUrl":"https://doi.org/10.1115/icone2020-16291","url":null,"abstract":"\u0000 In a light-water reactor, during normal operating condition, the UO2 nuclear fuel pellets undergo fragmentation primarily due to presence of thermal stresses, fission gas development and pellet-clad mechanical interaction. Under Loss of Coolant Accident (LOCA) conditions, a portion of fuel fragments can freely move downwards to the ballooning region due to the significant cladding deformation. The fuel relocation can localize the heat load and in turn accelerate the cladding balloon and burst process. Cladding burst is of great concern because of the potential for fuel dispersal into coolant and clad structural stability. In our work, we built up a finite element model considering cladding balloon, fuel relocation and its resultant thermal feedback during LOCA condition with ABAQUS. The clad balloon model includes phase transformation, swelling, thermal and irradiation creep, irradiation hardening and annealing and other important thermal-mechanical properties. The mass of relocation model was verified against the analytical cases of single balloon and twin balloons. The cladding balloon model combined with fuel thermal conductivity degradation was verified against fuel performance code, FRAPTRAN. Finally, with the evolution of pellet-cladding gap, the fuel mass relocation was calculated and compared against the IFA-650.4 transient test from the Halden reactor.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"76441449","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}

{"title":"Research on Hydraulic Model Test of Pumping Station Forebay","authors":"Jiale Jian, Fang Wang, Rong Zhang, Benjing Tang","doi":"10.1115/icone2020-16782","DOIUrl":"https://doi.org/10.1115/icone2020-16782","url":null,"abstract":"\u0000 A physical model with a scale of 1:20 was used to study the hydraulic characteristics of the flow in the forebay area of the pump station in Tian Wan Nuclear Power Plant Project Unit 5&6. The original layout scheme of the pump station forebay is limited by space, the turning radius of the intake gallery is small and the straight line section is short, the uniformity of water distribution is poor after seawater enters the forebay, the flow in the forebay cannot fully diffuse, the mainstream is concentrated, the flow distribution in the inlet section of each channel of the pump station is uneven, and the flow pattern is poor. Firstly, the distribution of water flow and uniformity of water distribution in the pump station forebay under different turning radius of tunnel are compared, and the turning radius of intake corridor is determined by calculating the head loss in different parts of tunnel. A variety of rectification facilities are arranged in the pumping station forebay, including modifying the grid type, setting diversion wall, diversion plate, energy dissipation beam, rectification bottom sill, etc. The recommended scheme is determined by hydraulic model test. The scheme can satisfy the requirement of uniformity of water distribution in the forebay, and the flow pattern along each section of the channel is also good.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"79146948","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}

{"title":"Study on the Effect of Different Factors of Displacement Cascades in Alpha-Fe by Molecular Dynamics Simulations","authors":"P. Lin, J. Nie, Meidan Liu","doi":"10.1115/icone2020-16161","DOIUrl":"https://doi.org/10.1115/icone2020-16161","url":null,"abstract":"\u0000 As the key component of RPV steel, α-Fe is under neutron irradiation during its long-term service, and lattice atoms of α-Fe are knocked by neutrons, which leads to irradiation damage. In this paper, molecular dynamics method is conducted to investigate the effect of temperature, vacancy concentration and tensile strain on irradiation-induced damage by displacement cascade simulations in α-Fe. The simulations are performed with primary knock-on atom energies ranging from 0.1 to 5 keV, temperature ranging from 100 to 500K, vacancy concentration ranging from 0% to 1% and applied tensile strain ranging from 0 to 5%. The time evolution of defects produced during displacement cascade can be obtained where the surviving number of Frenkel pairs increases rapidly at first, then decrease and comes to stability finally. The influence of these factors on defect production can be concluded as following: The increase of PKA energy, vacancy concentration and applied tensile strain can lead to the increase of defect numbers. In contrast, the increase of temperature decreases the defect numbers. Vacancies and interstitials cluster size distributions are varied in different case. The results are meaningful to describe some microcosmic mechanisms of RPV steels in nuclear system.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"73578244","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}

{"title":"Implementation and Validation of an Aerosol Collection Model by a Spray in a CFD Code: Application to the Scavenging of Aerosols Released During Laser Cutting Operations of Fuel Debris for the Dismantling of the Damaged Reactors of Fukushima Dai-ichi","authors":"T. Gelain, E. Porcheron, Yohan Leblois, I. Doyen, C. Chagnot, C. Journeau, D. Roulet","doi":"10.1115/icone2020-16141","DOIUrl":"https://doi.org/10.1115/icone2020-16141","url":null,"abstract":"\u0000 The general context of this article is related to the dismantling of the damaged reactors of Fukushima Dai-ichi and, more specifically, to the implementation of the laser cutting technique for the fuel debris retrieval. IRSN is involved in a project led by ONET Technologies and in partnership with CEA, to bring relevant elements in order to analyze the risks induced by the dispersion of aerosols released by the dismantling operations.\u0000 During the laser cutting operations in air or underwater conditions, particles will be produced, involving a potential risk of dispersion into the environment. Hence, in order to prevent this situation, their collection is one of the safety key issues in the in-situ dismantling actions. For that, IRSN performed CFD simulations of aerosol scavenging by a spray to evaluate the collection efficiency by this technique.\u0000 These simulations, conducted with the ANSYS CFX code, use an Eulerian method for the continuous phase, and a Lagrangian method for the spray for which a collection model detailed by Plumecocq [1] or Marchand [2] was implemented. Aerosols are modelled by a DQMOM population balance implemented by Gelain et al. [3] (already used for recent simulations in the same context), and enriched with a deposition model developed by Nerisson et al. [4].\u0000 At first, CFD simulations were performed with the geometry of the IRSN TOSQAN facility [5], comparatively to experimental results presented in a previous paper [6]. This step enables the validation of the collection model implementation and to study the sensitivity to the aerosol size.\u0000 Then, CFD simulations were conducted with the geometry of the pedestal of Fukushima Dai-ichi reactors, to be more representative of a realistic case. For this configuration, sensitivity studies are described, highlighting both the influence of a multispray and of thermal-hydraulic conditions (temperature) on aerosol scavenging efficiency.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"85269583","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}

{"title":"Proportional-Integral Disturbance Observer of Nuclear Reactors","authors":"Z. Dong","doi":"10.1115/icone2020-16207","DOIUrl":"https://doi.org/10.1115/icone2020-16207","url":null,"abstract":"\u0000 A proportional-integral disturbance observer (PI-DO) for monitoring nuclear reactors is newly proposed, which is driven by the measurements of neutron flux and coolant temperature at reactor inlet as well as their integrations. This PI-DO provides a globally asymptotic estimation with a bounded steady-state error for the reactor key process variables as well as the total disturbances in channels of the neutron kinetics and primary coolant thermal-hydraulics. Moreover, the PI-DO is applied to reconstruct the unmeasurable state variables and total disturbances of a nuclear heating reactor (NHR). Numerical simulation results not only verify the theoretic analysis but also show both the satisfactory performance and the influence of observer parameters.","PeriodicalId":63646,"journal":{"name":"核工程研究与设计","volume":null,"pages":null},"PeriodicalIF":0.0,"publicationDate":"2020-08-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"77515954","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}