Reactor numerical simulation and hydraulic test research
International Nuclear Information System (INIS)
Yang, L. S.
2009-01-01
In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device
Real-time numerical simulation with high efficiency for an experimental reactor system
International Nuclear Information System (INIS)
Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan
2006-01-01
The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)
Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)
International Nuclear Information System (INIS)
Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.
2013-01-01
Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)
Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1999-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases
Energy Technology Data Exchange (ETDEWEB)
Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)
Simplified numerical simulation of hot channel in sodium cooled reactor
International Nuclear Information System (INIS)
Fonseca, F. de A.S. da; Silva Filho, E.
1988-12-01
The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt
Numerical Simulation of Two-branch Hot Gas Mixing at Reactor Outlet of HTR-PM
International Nuclear Information System (INIS)
Hao Pengefei; Zhou Yangping; Li Fu; Shi Lei; He Heng
2014-01-01
A series of two-branch model experiment has been finished to investigate the thermal mixing efficiency of the HTR-PM reactor outlet. This paper introduces the numerical simulation on the design of thermal mixing structure of HTR-PM and the test facility with Fluent software. The profiles of temperature, pressure and velocity in the mixing structure design and the test facility are discussed by comparing with the model experiment results. The numerical simulation results of the test facility have good agreement to the experiment results. In addition, the thermal-fluid characters obtained by numerical simulation show the thermal mixing structure of HTR-PM has similarity with the test facility. Finally, it is concluded that the thermal mixing design at HTR-PM reactor outlet can fulfilled the requirements for high thermal mixing efficiency and appropriate pressure drop. (author)
International Nuclear Information System (INIS)
2001-10-01
The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)
Numerical simulation of flow field in the China advanced research reactor flow-guide tank
International Nuclear Information System (INIS)
Xu Changjiang
2002-01-01
The flow-guide tank in China advanced research reactor (CARR) acts as a reactor inlet coolant distributor and play an important role in reducing the flow-induced vibration of the internal components of the reactor core. Numerical simulations of the flow field in the flow-guide tank under different conceptual designing configurations are carried out using the PHOENICS3.2. It is seen that the inlet coolant is well distributed circumferentially into the flow-guide tank with the inlet buffer plate and the flow distributor barrel. The maximum cross-flow velocity within the flow-guide tank is reduced significantly, and the reduction of flow-induced vibration of reactor internals is expected
Numerical simulation of moderator flow and temperature distributions in a CANDU reactor vessel
International Nuclear Information System (INIS)
Carlucci, L.N.
1982-10-01
This paper describes numerical predictions of the two-dimensional flow and temperature fields of an internally-heated liquid in a typical CANDU reactor vessel. Turbulence momentum and energy transport are simulated using the k-epsilon model. Both steady-state and transient results are discussed. The finite control volume analogues of the conservation equations are solved using a modified version of the TEACH code
Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass
Energy Technology Data Exchange (ETDEWEB)
Miller, R.S.; Bellan, J. [California Inst. of Tech., Pasadena, CA (United States). Jet Propulsion Lab.
1998-08-01
A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.
Direct numerical simulation of reactor two-phase flows enabled by high-performance computing
Energy Technology Data Exchange (ETDEWEB)
Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.; Feng, Jinyong; Gouws, Andre; Li, Mengnan; Bolotnov, Igor A.
2018-04-01
Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent research progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.
Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP
Tereshonok, V. A.; Kryakvin, L. V.; Pitilimov, V. A.; Karpov, S. A.; Kulikov, V. I.; Zhylmaganbetov, N. M.; Kavun, O. Yu.; Popykin, A. I.; Shevchenko, R. A.; Shevchenko, S. A.; Semenova, T. V.
2017-12-01
The results of numerical calculations and measurements of some reactor parameters during the physical startup tests at unit 3 of Rostov NPP are presented. The following parameters are considered: the critical boron acid concentration and the currents from ionization chambers (IC) during the scram system efficiency evaluation. The scram system efficiency was determined using the inverse point kinetics equation with the measured and simulated IC currents. The results of steady-state calculations of relative power distribution and efficiency of the scram system and separate groups of control rods of the control and protection system are also presented. The calculations are performed using several codes, including precision ones.
CFD Numerical Simulation of Biodiesel Synthesis in a Spinning Disc Reactor
Directory of Open Access Journals (Sweden)
Wen Zhuqing
2015-03-01
Full Text Available In this paper a two-disc spinning disc reactor for intensified biodiesel synthesis is described and numerically simulated. The reactor consists of two flat discs, located coaxially and parallel to each other with a gap of 0.2 mm between the discs. The upper disc is located on a rotating shaft while the lower disc is stationary. The feed liquids, triglycerides (TG and methanol are introduced coaxially along the centre line of rotating disc and stationary disc. Fluid hydrodynamics in the reactor for synthesis of biodiesel from TG and methanol in the presence of a sodium hydroxide catalyst are simulated, using convection-diffusion-reaction species transport model by the CFD software ANSYS©Fluent v. 13.0. The effect of the upper disc’s spinning speed is evaluated. The results show that the rotational speed increase causes an increase of TG conversion despite the fact that the residence time decreases. Compared to data obtained from adequate experiments, the model shows a satisfactory agreement.
Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics
International Nuclear Information System (INIS)
Bernard, J.P.; Haapalehto, T.
1996-01-01
The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)
International Nuclear Information System (INIS)
Ribeiro, Felipe Lopes; Pinto, Joao Pedro C.T.A.
2013-01-01
The 4 th generation Very High Temperature Reactor (VHTR) most popular concept uses a graphite-moderated and helium cooled core with an outlet gas temperature of approximately 1000 deg C. The high output temperature allows the use of the process heat and the production of hydrogen through the thermochemical iodine-sulfur process as well as highly efficient electricity generation. There are two concepts of VHTR core: the prismatic block and the pebble bed core. The prismatic block core has two popular concepts for the fuel element: multihole and annular. In the multi-hole fuel element, prismatic graphite blocks contain cylindrical flow channels where the helium coolant flows removing heat from cylindrical fuel rods positioned in the graphite. In the other hand, the annular type fuel element has annular channels around the fuel. This paper shows the numerical evaluations of prismatic multi-hole and annular VHTR fuel elements and does a comparison between the results of these assembly reactors. In this study the analysis were performed using the CFD code ANSYS CFX 14.0. The simulations were made in 1/12 fuel element models. A numerical validation was performed through the energy balance, where the theoretical and the numerical generated heat were compared for each model. (author)
Energy Technology Data Exchange (ETDEWEB)
Jeong, Dong Hwa; Jung, Byung Ryul; Jang, Ho Cheol; Yune, Seok Jeong; Kim, Eun Kee [KEPCO EnC, Daejeon (Korea, Republic of)
2015-10-15
In this study the numerical simulation of the pressure distribution in the downcomer region resulting from the pressure pulsation by the Reactor Coolant Pump (RCP) is performed using the Finite Difference Method (FDM). Simulation is carried out for the cylindrical shaped 2-dimensional model equivalent to the outer surface of the Core Support Barrel (CSB) of APR1400 and a 1/2 model is adopted based on the bilateral symmetry by the inlet nozzle. The fluid temperature is 555 .deg. F and the forcing frequencies are 120Hz, 240Hz, 360Hz and 480Hz. Simulation results of the axial pressure distributions are provided as the Root Mean Square (RMS) values at the five locations of 0°, 45°, 90°, 135° and 180° in the circumferential direction from the inlet nozzle location. In the study, the numerical simulation of pressure distributions in the downcomer region induced by the RCP was performed using FDM and the results were reviewed. The interference of the waves returned from both boundaries in the axial direction and the source of the sinusoidal wave is shown on the inlet nozzle interface pressure point. It seems that the maximum pressures result from the superposition of the waves reflected from the seating surface and the waves newly arrived from the inlet nozzle interface pressure location.
Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model
Kazeminezhad, F.; Anghaie, S.
2008-01-01
Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.
International Nuclear Information System (INIS)
Marchetto, C.
2002-05-01
During operation of pressurised water reactor, corrosion of the primary circuit alloys leads to the release of metallic species such as iron, nickel and cobalt in the primary fluid. These corrosion products are implicated in different transport phenomena and are activated in the reactor core where they are submitted to neutron flux. The radioactive corrosion products are afterwards present in the out of flux parts of primary circuit where they generate a radiation field. The first part of this study deals with the modelling of the corrosion: product transport phenomena. In particular, considering the current state of the art, corrosion and release mechanisms are described empirically, which allows to take into account the material surface properties. New mass balance equations describing the corrosion product behaviour are thus obtained. The numerical resolution of these equations is implemented in the second part of this work. In order to obtain large time steps, we choose an implicit time scheme. The associated system is linearized from the Newton method and is solved by a preconditioned GMRES method. Moreover, a time step auto-adaptive management based on Newton iterations is performed. Consequently, an efficient resolution has been implemented, allowing to describe not only the quasi-steady evolutions but also the fast transients. In a last step, numerical simulations are carried out in order to validate the new corrosion product transport modelling and to illustrate the capabilities of this modelling. Notably, the numerical results obtained indicate that the code allows to restore the on-site observations underlining the influence of material surface properties on reactor contamination. (author)
Numerical simulation study on the air/water countercurrent flow limitation in nuclear reactors
Energy Technology Data Exchange (ETDEWEB)
Morghi, Youssef; Mesquita, Amir Z., E-mail: ssfmorghi@gmail.com, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Puente, Jesus, E-mail: jpuente720@gmail.com [Centro Federal de Educaçao Tecnologica Celso Suckowda Fonseca (CEFET), Angra dos Reis, RJ (Brazil); Baliza, Ana R., E-mail: baliza@eletronuclear.gov.br [Eletrobras Eletronuclear Angra dos Reis, RJ (Brazil)
2017-07-01
After a loss-of-coolant accident (LOCA) in a Pressurized Water Reactor (PWR), the temperature of the fuel elements cladding increases dramatically due to the heat produced by the fission products decay, which is not adequately removed by the vapor contained in the core. In order to avoid this sharp rise in temperature and consequent melting of the core, the Emergency Core Cooling System is activated. This system initially injects borated water from accumulator tanks of the reactor through the inlet pipe (cold leg) and the outlet pipe (hot leg), or through the cold leg only, depending on the plant manufacturer. Some manufacturers add to this, direct injection into the upper plenum of the reactor. The penetration of water into the reactor core is a complex thermo fluid dynamic process because it involves the mixing of water with the vapor contained in the reactor, added to that generated in the contact of the water with the still hot surfaces in various geometries. In some critical locations, the vapor flowing in the opposite direction of the water can control the penetration of this into the core. This phenomenon is known as Countercurrent Flow Limitation (CCFL) or Flooding, and it is characterized by the control that a gas exerts in the liquid flow in the opposite direction. This work presents a proposal to use a CFD to simulate the CCFL phenomenon. Numerical computing can provide important information and data that is difficult or expensive to measure or test experimentally. Given the importance of computational science today, it can be considered a third and independent branch of science on an equal footing with the theoretical and experimental sciences. (author)
Numerical simulation of progressive inlet orifices in boiling water reactor fuel
International Nuclear Information System (INIS)
Lundgren, Sara
2004-07-01
This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. Experimental work has been done with progressive orifices, the edge of which is half-oval in shape. The advantage of progressive orifices is the lower pressure losses with an increase of the Reynolds number, a similar phenomenon that appears in external flow around curved bodies. Since there are high costs associated with experimental generation of high- temperature and high-pressure data, it is of some interest to be able to reproduce and generate data using Computational Fluid Dynamics (CFD). This work deals with the possibility to use the CFD-code Fluent to do numerical simulations of the flow through progressive orifices. The following conclusions may be drawn from the numerical results: All simulations using Reynolds-Averaged Navier-Stokes (RANS) turbulence models, two-dimensional and three-dimensional, capture an abrupt decrease of the resistance coefficient at higher Reynolds numbers. Two-equation models seem to under-predict the critical Reynolds number. The five-equation Reynolds Stress Model (RSM) gives a critical Reynolds number of the same order of magnitude of that measured in experiments. No major differences have
Thermal-hydraulic numerical simulation of fuel sub-assembly for Sodium-cooled Fast Reactor
International Nuclear Information System (INIS)
Saxena, Aakanksha
2014-01-01
The thesis focuses on the numerical simulation of sodium flow in wire wrapped sub-assembly of Sodium-cooled Fast Reactor (SFR). First calculations were carried out by a time averaging approach called RANS (Reynolds- Averaged Navier-Stokes equations) using industrial code STAR-CCM+. This study gives a clear understanding of heat transfer between the fuel pin and sodium. The main variables of the macroscopic flow are in agreement with correlations used hitherto. However, to obtain a detailed description of temperature fluctuations around the spacer wire, more accurate approaches like LES (Large Eddy Simulation) and DNS (Direct Numerical Simulation) are clearly needed. For LES approach, the code TRIO U was used and for the DNS approach, a research code was used. These approaches require a considerable long calculation time which leads to the need of representative but simplified geometry. The DNS approach enables us to study the thermal hydraulics of sodium that has very low Prandtl number inducing a very different behavior of thermal field in comparison to the hydraulic field. The LES approach is used to study the local region of sub-assembly. This study shows that spacer wire generates the local hot spots (∼20 C) on the wake side of spacer wire with respect to the sodium flow at the region of contact with the fuel pin. Temperature fluctuations around the spacer wire are low (∼1 C-2 C). Under nominal operation, the spectral analysis shows the absence of any dominant peak for temperature oscillations at low frequency (2-10 Hz). The obtained spectra of temperature oscillations can be used as an input for further mechanical studies to determine its impact on the solid structures. (author) [fr
Hong, Jongsup
2012-07-01
Ion transport membrane (ITM) based reactors have been suggested as a novel technology for several applications including fuel reforming and oxy-fuel combustion, which integrates air separation and fuel conversion while reducing complexity and the associated energy penalty. To utilize this technology more effectively, it is necessary to develop a better understanding of the fundamental processes of oxygen transport and fuel conversion in the immediate vicinity of the membrane. In this paper, a numerical model that spatially resolves the gas flow, transport and reactions is presented. The model incorporates detailed gas phase chemistry and transport. The model is used to express the oxygen permeation flux in terms of the oxygen concentrations at the membrane surface given data on the bulk concentration, which is necessary for cases when mass transfer limitations on the permeate side are important and for reactive flow modeling. The simulation results show the dependence of oxygen transport and fuel conversion on the geometry and flow parameters including the membrane temperature, feed and sweep gas flow, oxygen concentration in the feed and fuel concentration in the sweep gas. © 2012 Elsevier B.V.
International Nuclear Information System (INIS)
Sarchami, Araz; Ashgriz, Nasser; Kwee, Marc
2014-01-01
Highlights: • 3D model of a Candu reactor is modeled to investigate flow distribution. • The results show the temperature distribution is not symmetrical. • Temperature contours show the hot regions at the top left-hand side of the tank. • Interactions of momentum flows and buoyancy flows create circulation zones. • The results indicate that the moderator tank operates in the buoyancy driven mode. -- Abstract: Three dimensional numerical simulations are conducted on a full scale CANDU Moderator and transient variations of the temperature and velocity distributions inside the tank are determined. The results show that the flow and temperature distributions inside the moderator tank are three dimensional and no symmetry plane can be identified. Competition between the upward moving buoyancy driven flows and the downward moving momentum driven flows in the center region of the tank, results in the formation of circulation zones. The moderator tank operates in the buoyancy driven mode and any small disturbances in the flow or temperature makes the system unstable and asymmetric. Different types of temperature fluctuations are noted inside the tank: (i) large amplitude are at the boundaries between the hot and cold; (ii) low amplitude are in the core of the tank; (iii) high frequency fluctuations are in the regions with high velocities and (iv) low frequency fluctuations are in the regions with lower velocities
Energy Technology Data Exchange (ETDEWEB)
Morghi, Youssef; Mesquita, Amir Zacarias; Santos, Andre Augusto Campagnole dos; Vasconcelos, Victor, E-mail: ymo@cdtn.br, E-mail: amir@cdtn.br, E-mail: aacs@cdtn.br, E-mail: vitors@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
For the experimental study on the air/water countercurrent flow limitation in Nuclear Reactors, were built at CDTN an acrylic test sections with the same geometric shape of 'hot leg' of a Pressurized Water Reactor (PWR). The hydraulic circuit is designed to be used with air and water at pressures near to atmospheric and ambient temperature. Due to the complexity of the CCFL experimental, the numerical simulation has been used. The aim of the numerical simulations is the validation of experimental data. It is a global trend, the use of computational fluid dynamics (CFD) modeling and prediction of physical phenomena related to heat transfer in nuclear reactors. The most used CFD codes are: FLUENT®, STAR- CD®, Open Foam® and CFX®. In CFD, closure models are required that must be validated, especially if they are to be applied to nuclear reactor safety. The Thermal- Hydraulics Laboratory of CDTN offers computing infrastructure and license to use commercial code CFX®. This article describes a review about CCFL and the use of CFD for numerical simulation of this phenomenal for Nuclear Rector. (author)
International Nuclear Information System (INIS)
Morghi, Youssef; Mesquita, Amir Zacarias; Santos, Andre Augusto Campagnole dos; Vasconcelos, Victor
2015-01-01
For the experimental study on the air/water countercurrent flow limitation in Nuclear Reactors, were built at CDTN an acrylic test sections with the same geometric shape of 'hot leg' of a Pressurized Water Reactor (PWR). The hydraulic circuit is designed to be used with air and water at pressures near to atmospheric and ambient temperature. Due to the complexity of the CCFL experimental, the numerical simulation has been used. The aim of the numerical simulations is the validation of experimental data. It is a global trend, the use of computational fluid dynamics (CFD) modeling and prediction of physical phenomena related to heat transfer in nuclear reactors. The most used CFD codes are: FLUENT®, STAR- CD®, Open Foam® and CFX®. In CFD, closure models are required that must be validated, especially if they are to be applied to nuclear reactor safety. The Thermal- Hydraulics Laboratory of CDTN offers computing infrastructure and license to use commercial code CFX®. This article describes a review about CCFL and the use of CFD for numerical simulation of this phenomenal for Nuclear Rector. (author)
Numerical simulation of scale-up effects of methanol-to-olefins fluidized bed reactors
DEFF Research Database (Denmark)
Lu, Bona; Zhang, Jingyuan; Luo, Hao
2017-01-01
)-based drag models are combined in simulations. The fluidization characteristics in terms of flow structures, velocity distribution, mass fractions of gaseous product and coke distribution are presented against available experimental data for different-sized reactors. It is found that typical hydrodynamic...
International Nuclear Information System (INIS)
Zhang Zuoyi; Scherer, W.
1996-01-01
This report analyzes reverse water ingress accidents in the SIEMENS 200 MW Modular Pebble-Bed High Temperature Gas Cooled Reactor (HTR-MODULE) under the assumption of no active safety protection systems in order to find the safety margins of the current HTR-MODULE design and to realize a catastrophe-free nuclear technology. A water, steam and helium multi-phase cavity model is developed and implemented in the DSNP simulation system. The DSNP system is then used to simulate the primary and secondary circuit of a HTR-MODULE power plant. Comparisons of the model with experiments and with TINTE calculations serve as validation of the simulation. The analysis of the primary circuit tries to answer the question how fast the water enters the reactor core. It was found that the maximum H 2 O concentration increase in the reactor core is smaller than 0.3 kg/(m 3 s). The liquid water vaporization in the steam generator and H 2 O transport from the steam generator to the reactor core reduce the ingress velocity of the H 2 O into the reactor core. In order to answer the question how much water enters the primary circuit, the full cavitation of the feed water pumps is analyzed. It is found that if the secondary circuit is depressurized enough, the feed water pumps will be inherently stopped by the full cavitation. This limits the water to be pumped from the deaerator to the steam generator. A comprehensive simulation of the MODUL-HTR power plant then shows that the H 2 O inventory in the primary circuit can be limited to about 3000 kg. The nuclear reactivity increase caused by the water ingress leads to a fast power excursion, which, however, is inherently counterbalanced by negative feedback effects. Concerning the integrity of the fuel elements, the safety relevant temperature limit of 1600 C was not reached in any case. (orig.) [de
International Nuclear Information System (INIS)
Wang Hang; Wang Jie; Laurien, E.
2010-01-01
The lower plenum in high temperature gas-cooled reactor was designed to mix the gas of different temperatures from the reactor core. Previous researches suggest the current geometry of the lower plenum to be improved for better mixing capability and lower pressure drop. In the presented work, a series of varied geometries were investigated with numerical simulation way. The choice of appropriate mesh type and size used in the geometry variation was discussed with the reference of experimental data. The original thin ribs in the current design were merged into thicker ones, and a junction located at the starting end of the outlet pipe was introduced. After comparing several potential optimization methods, an improved geometry was selected with the merged ribs increasing the pre-defined mixing coefficient and the junction reducing the pressure drop. Future work was discussed based on the simulation of real reactor case. The work shows a direction for design improvements of the lower plenum geometry. (authors)
International Nuclear Information System (INIS)
Benner, J.
1984-03-01
A method for the numerical simulation of the Pressurized Water Reactor (PWR) core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. All these models have been implemented into the code Flux-4. For the solution of the very complex, coupled equations of motions for fluid and fuel rods an efficient numerical solution technique has been developed. With the new code-version Flux-5 the PWR-blowdown is parametically investigated. The calculated core barrel loadings are compared with Flux-4 results, simulating the core's inertia by a mass ring of HDR type. (orig.) [de
International Nuclear Information System (INIS)
Barros, R.C. de.
1992-05-01
Presented here is a new numerical nodal method for the simulation of the axial power distribution within nuclear reactors using the one-dimensional one speed kinetics diffusion model with one group of delayed neutron precursors. Our method is based on a spectral analysis of the nodal kinetics equations. These equations are obtained by integrating the original kinetics equations separately over a time step and over a spatial node, and then considering flat approximations for the forward difference terms. These flat approximations are the only approximations that are considered in the method. As a result, the spectral nodal method for space - time reactor kinetics generates numerical solutions for space independent problems or for time independent problems that are completely free from truncation errors. We show numerical results to illustrate the method's accuracy for coarse mesh calculations. (author)
Energy Technology Data Exchange (ETDEWEB)
Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)
2015-12-31
A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.
Numerical simulation of large systems: application to a pressurized water reactor
International Nuclear Information System (INIS)
Tallec, Michele.
1981-10-01
This note describes the design of a pressurized water reactor power plant simulator using a minicomputer. It contains the description of the models used to simulate the dynamic behavior of the various components of the nuclear power station (i.e. the reactor core, two steam generators, the pressurizer and the control systems associated with them); the algorithms used to integrate the resulting system of algebraic differential equations; the solution of problems associated with the use of a mini-computer; the control deck outlay designed and the variables shown on it to the user; and finally the description of tests made to validate the models used and the results obtained for various transients using plant signal is presented. These results are compared to corresponding plant signals and outputs of other, already existing models [fr
Numerical simulation of stochastic point kinetic equation in the dynamical system of nuclear reactor
International Nuclear Information System (INIS)
Saha Ray, S.
2012-01-01
Highlights: ► In this paper stochastic neutron point kinetic equations have been analyzed. ► Euler–Maruyama method and Strong Taylor 1.5 order method have been discussed. ► These methods are applied for the solution of stochastic point kinetic equations. ► Comparison between the results of these methods and others are presented in tables. ► Graphs for neutron and precursor sample paths are also presented. -- Abstract: In the present paper, the numerical approximation methods, applied to efficiently calculate the solution for stochastic point kinetic equations () in nuclear reactor dynamics, are investigated. A system of Itô stochastic differential equations has been analyzed to model the neutron density and the delayed neutron precursors in a point nuclear reactor. The resulting system of Itô stochastic differential equations are solved over each time-step size. The methods are verified by considering different initial conditions, experimental data and over constant reactivities. The computational results indicate that the methods are simple and suitable for solving stochastic point kinetic equations. In this article, a numerical investigation is made in order to observe the random oscillations in neutron and precursor population dynamics in subcritical and critical reactors.
Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor
International Nuclear Information System (INIS)
In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae
2008-01-01
Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400
Application of 2DOF controller for reactor power control. Verification by numerical simulation
International Nuclear Information System (INIS)
Ishikawa, Nobuyuki; Suzuki, Katsuo
1996-09-01
In this report the usefulness of the two degree of freedom (2DOF) control is discussed to improve the reference response characteristics and robustness for reactor power control system. The 2DOF controller consists of feedforward and feedback elements. The feedforward element was designed by model matching method and the feedback element by solving the mixed sensitivity problem of H ∞ control. The 2DOF control gives good performance in both reference response and robustness to disturbance and plant perturbation. The simulation of reactor power control was performed by digitizing the 2DOF controller with the digital control periods of 10[msec]. It is found that the control period of 10[msec] is enough not to make degradation of the control performance by digitizing. (author)
Reactor Thermal Hydraulic Numerical Calculation And Modeling
International Nuclear Information System (INIS)
Duong Ngoc Hai; Dang The Ba
2008-01-01
In the paper the results of analysis of thermal hydraulic state models using the numerical codes such as COOLOD, EUREKA and RELAP5 for simulation of the reactor thermal hydraulic states are presented. The calculations, analyses of reactor thermal hydraulic state and safety were implemented using different codes. The received numerical results, which were compared each to other, to experiment measurement of Dalat (Vietnam) research reactor and published results, show their appropriateness and capacity for analyses of different appropriate cases. (author)
Directory of Open Access Journals (Sweden)
Guodong Liu
2013-01-01
Full Text Available Modular pebble-bed nuclear reactor (MPBNR technology is promising due to its attractive features such as high fuel performance and inherent safety. Particle motion of fuel and graphite pebbles is highly associated with the performance of pebbled-bed modular nuclear reactor. To understand the mechanism of pebble’s motion in the reactor, we numerically studied the influence of number ratio of fuel and graphite pebbles, funnel angle of the reactor, height of guide ring on the distribution of pebble position, and velocity by means of discrete element method (DEM in a two-dimensional MPBNR. Velocity distributions at different areas of the reactor as well as mixing characteristics of fuel and graphite pebbles were investigated. Both fuel and graphite pebbles moved downward, and a uniform motion was formed in the column zone, while pebbles motion in the cone zone was accelerated due to the decrease of the cross sectional flow area. The number ratio of fuel and graphite pebbles and the height of guide ring had a minor influence on the velocity distribution of pebbles, while the variation of funnel angle had an obvious impact on the velocity distribution. Simulated results agreed well with the work in the literature.
Numerical simulation for debris bed behavior in sodium cooled fast reactor
International Nuclear Information System (INIS)
Tagami, Hirotaka; Tobita, Yoshiharu
2014-01-01
For safety analysis of SFR, it is necessary to evaluate behavior along with coolability of debris bed in lower plenum which is formed in severe accident. In order to analyze debris behavior, model for dense sediment particles behavior was proposed and installed in SFR safety analysis code SIMMER. SIMMER code could adequately reproduce experimental results simulating the self-leveling phenomena with appropriate model parameters for bed stiffness. In reactor condition, the self-leveling experiment for prototypical debris bed has not been performed. Additionally, the prototypical debris bed consists of non-spherical particles and it is difficult to quantify model parameters. This situation brings sensitivity analysis to investigate effect of model parameters on the self-leveling phenomena of prototypical debris bed in present paper. As initial condition for sensitivity analysis, simple mound-like debris bed in sodium-filled lower plenum in reactor vessel is considered. The bed consists of the mixture of fuel debris of 3,300 kg and steel debris of 1,570 kg. Decay heat is given to this fuel debris. The model parameter is chosen as sensitivity parameter. Sensitivity analysis shows that the model parameters can effect on intensity of self-leveling phenomena and eventual flatness of bed. In all analyses, however, coolant and sodium vapor break the debris bed at mainly center part of bed and the debris is relocated to outside of bed. Through this process, the initial debris bed is almost planarized before re-melting of debris. This result shows that the model parameters affect the self-leveling phenomena, but its effect in the safety analysis of SFRs is limited. (author)
Energy Technology Data Exchange (ETDEWEB)
Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)
2016-11-15
Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.
International Nuclear Information System (INIS)
Vladimir Ya Kumaev
2005-01-01
Full text of publication follows: Numerical simulation of the melting processes is necessary in substantiating the safety of new generation reactors to determine the quantitative characteristics of the melt formed, destruction of reactor vessel and components, melt interaction processes in the melt localization systems (MLS), formation and transport of hydrogen, radioactive aerosols under severe accidents. The results of computations will be applied in developing the procedures for severe accident management and mitigation of its consequences and designing melt localization systems. The report is devoted to the development and application of the two-dimensional and three-dimensional versions of the DINCOR code intended for numerical simulation of the thermal hydraulic processes in a multicomponent medium with solid-liquid phase changes. The basic set of equations of multicomponent medium is presented. The numerical method to solve the governing equations is discussed. Some examples of two-dimensional code applications are presented. The experience of application of the code has shown that joint calculations of hydrodynamics, heat transfer, stratification and chemical interaction enable the process description accuracy to be significantly increased and the number of initial experimental data to be reduced. The multicomponent medium model can be used as the base for the development of a three-dimensional version of the code. At the same time, it was established that the models being used need be further developed. The most important problems are the following: -development of the local mathematical models of liquefaction and solidification of materials under front melting and melting due to the action of internal sources; -development of the model of incompressible components separation; -development of the models of dissolution and chemical interaction of multicomponent medium components. In conclusion possible verification of the computer code is discussed. (author)
On numerical simulation of fuel assembly bow in pressurized water reactors
Energy Technology Data Exchange (ETDEWEB)
Horváth, Ákos, E-mail: akoshorvath@t-online.hu [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany); Budapest University of Technology and Economics, Department of Aircraft and Ships, Stoczek Street 6, Building J, H-1111 Budapest (Hungary); Dressel, Bernd [AREVA, AREVA NP GmbH, Paul-Gossen-Str. 100, 91052 Erlangen (Germany)
2013-12-15
Highlights: • Simulation of fuel assembly bow by coupled CFD and finite element method. • Comparison of calculated and experimentally measured bow shapes. • Investigation of boundary condition effect on bow pattern of a fuel assembly row. • Highlighting importance of consideration of fluid–structure interaction. • Assessment of flow redistribution within the fuel assembly row model. - Abstract: Fuel assembly bow in pressurized water reactor cores is largely triggered by lateral hydraulic forces together with creep processes generated by neutron flux. A detailed understanding of the flow induced bow behaviour is, therefore, an important issue. The experimental feedbacks and laboratory tests on fuel assembly bow show that it is characterized to a high degree by fluid–structure interaction (FSI) effects, therefore, consideration of FSI is essential and indispensable in full comprehension of the bow mechanism. In the present study, coupled computational fluid dynamics (CFD) and finite element simulations are introduced, calculating fuel assembly deformation under different conditions as a quasi-stationary phenomenon. The aim has been, on the one hand, to develop such a simplified fuel assembly CFD model, which allows set up of fuel assembly rows without loosing its main hydraulic characteristic; on the other hand, to investigate the bow pattern of a given fuel assembly row under different boundary conditions. The former one has been achieved by comparing bow shapes obtained with different fuel assembly (spacer grid) modelling approaches and mesh resolutions with experimental data. In the second part of the paper a row model containing 7.5 fuel assemblies is introduced, investigating the effect of flow distribution at inlet and outlet boundary regions on fuel assembly bow behaviour. The post processing has been focused on the bow pattern, lateral hydraulic forces, and horizontal flow distribution. The results have revealed importance of consideration of
Hong, Jongsup; Kirchen, Patrick; Ghoniem, Ahmed F.
2012-01-01
Ion transport membrane (ITM) based reactors have been suggested as a novel technology for several applications including fuel reforming and oxy-fuel combustion, which integrates air separation and fuel conversion while reducing complexity
Numerical simulation of Venturi ejector reactor in yellow phosphorus purification system
Energy Technology Data Exchange (ETDEWEB)
Wang, Xiao-jing; Tang, Lei, E-mail: alanleyfly@gmail.com; Jiang, Zeng
2014-03-15
Highlights: • Venturi ejector reactor is used in yellow phosphorus purification system to obtain high purity phosphorus. • We study the changes of vacuum region and the performances of Venturi ejector reactor with different operating pressure. • The whole study is aim to investigate the operating conditions, rather than to find out the small details of the chemical reaction. - Abstract: A novel type of Venturi ejector reactor, which was used in a pilot plant test in a factory in Guizhou in China, was developed to overcome the insufficiency of chemical reaction in the stirred-tank reactor in yellow phosphorus purification system. The effects of different working medium, the changes of vacuum region, and the performances of the Venturi ejector reactor with different operating pressure were investigated by FLUENT. Results show that the absolute value of vacuum pressure of single-phase flow was smaller than two-phase flow at the same operating conditions, which meat two-phase flow has a higher suction capability. Reflow phenomena occurred near the exit of suction pipe and nozzle. The former reflow which leads to energy loss of vacuum region was undesirable, and the latter was beneficial to the dispersion of liquid yellow phosphorus. With a flow rate ratio below 0.45, the performance of the Venturi ejector reactor was effective. By adjusting the operating pressure, a proper flow rate ratio could be satisfied to meet the production needs in yellow phosphorus purification system.
International Nuclear Information System (INIS)
Baptista, Vinicius Damas
1996-01-01
The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)
Numerical simulation of a reinforced concrete shield around a nuclear reactor
International Nuclear Information System (INIS)
Mahama, Mumuni Salifu
1996-02-01
Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)
International Nuclear Information System (INIS)
Hermansky, P.; Krajcovic, M.
2012-01-01
The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)
International Nuclear Information System (INIS)
Jang, Sung Hyun; Takata, Takashi; Yamaguchi, Akira
2016-01-01
A sodium-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called 'self-wastage phenomenon.' In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment
Directory of Open Access Journals (Sweden)
Sunghyon Jang
2016-04-01
Full Text Available A sodium–water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium–water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called “self-wastage phenomenon.” In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.
Energy Technology Data Exchange (ETDEWEB)
Jang, Sung Hyun; Takata, Takashi [Graduate School of Engineering, Osaka University, Osaka (Japan); Yamaguchi, Akira [Nuclear Professional School, The University of Tokyo, Ibaraki (Japan)
2016-04-15
A sodium-water reaction takes place when high-pressured water vapor leaks into sodium through a tiny defect on the surface of the heat transfer tube in a steam generator of the sodium-cooled fast reactor. The sodium-water reaction brings deterioration of the mechanical strength of the heat transfer tube at the initial leakage site. As a result, it damages the crack itself, which may eventually enlarge into a larger opening. This self-enlargement is called 'self-wastage phenomenon.' In this study, a simulant experiment was proposed to reproduce the self-enlargement of a crack and to evaluate the mechanism of the self-wastage. The damage on the surface of the crack was simulated by making the neutralization reaction with hydrochloric acid solution and sodium hydroxide solution. A numerical investigation was carried out to validate the feasibility of the approach and to determine experimental conditions. From the computation results, it is observed that when 5M HCl is injected into 5M of NaOH with 0.05 m/s inlet velocity, the temperature at the surface near the crack increased over 319.26 K. The computational results show that the self-wastage phenomenon is capable of being reproduced by the simulant experiment.
Reactor simulator development. Workshop material
International Nuclear Information System (INIS)
2001-01-01
process, including numerical methods, and an introduction to the CASSIM TM development system. Techniques and practice are presented, with exercises, for development and application of reactor simulators
International Nuclear Information System (INIS)
Venter, A.M.
1973-08-01
A short discussion is given of the physics of a nuclear reactor and the parameters which are used in the study of neutron transport. The mathematical formulation and detailed derivation is given of the neutron diffusion and transport equations. A description is given of the computer programmes, FIRE-5 and PELSN, developed at Pelindaba for the evaluation of both thermal and fast reactor systems. It is indicated how these computer programmes have been applied in the study of the PELINDUNA-O and other known critical facilities. The application of Lie-series to the solution of the neutron diffusion equation is discussed in detail. The time dependence of the variables is removed by means of a Laplacetransformation and the semi-analytical solution is written in terms of a transfer matrix. A complete set of recursion formulae, applicable to both homogeneous and heterogeneous reactor systems, is derived. The method used in the evaluation of the effective multiplication factor, k-eff, and the alpha-eigen-value is described. A computer programme was written to solve the neutron diffusion equation in terms of the Lie-series. The results are compared with the TIMOC and PELSN computer programmes. A method is suggested in which the Lie-series are used to solve the neutron transport equation. The transfer matrix for this case, is derived. A complete discussion is given of the solution to the space and time dependent diffusion equation in the presence of a delta source [af
Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor
International Nuclear Information System (INIS)
Braga, C.V.M.; Carajilescov, P.
1981-01-01
Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt
International Nuclear Information System (INIS)
Zhukov, A.V.; Sorokin, A.P.
2000-01-01
The problems of numerical modeling of thermohydraulics in assembly of fuel elements of fast reactors with the partial blockage of cross-section under the coolant are considered. The information about existing codes constructed on use of subchannel technique and model of porous body are presented. The results of calculation obtained by these codes are presented. (author)
International Nuclear Information System (INIS)
Wang Haijun; He Huining; Luo Yushan; Wang Weishu
2011-01-01
In the present work, a numerical simulation was performed to study the flow characteristics of lean jet to cross flow in a main tube in the safety injection of reactor cooling system. The influence scope and mixing characteristics of the confined lean jet in cross-flow were studied. It can be concluded that three basic flow regimes are marked, namely the attached lean jet, lift-off lean jet and impinging lean jet. The velocity ratio V R is the key factor in the flow state. The depth and region of jet to main flow are enhanced with the increase of the velocity ratio. The jet flow penetrates through the main flow with the increase of the velocity ratio. At higher velocity ratio, the jet flow strikes the main flow bottom and circumfluence happens in upriver of main flow. The vortex flow characteristics dominate the flow near region of jet to cross-flow and the mixture of jet to cross-flow. At different velocity ratio V R , the vortex grows from the same displacement, but the vortex type and the vortex is different. At higher velocity ratio, the vortex develops fleetly, wears off sharp and dies out sharp. The study is very important to the heat transfer experiments of cross-flow jet and thermal stress analysis in the designs of nuclear engineering. (authors)
Energy Technology Data Exchange (ETDEWEB)
Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)
1995-09-01
This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.
Energy Technology Data Exchange (ETDEWEB)
Kim, K. S.; Ju, H. G.; Jeon, T. H. and others
2005-03-15
A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well.
International Nuclear Information System (INIS)
Kim, K. S.; Ju, H. G.; Jeon, T. H. and others
2005-03-01
A comprehensive high fidelity reactor core modeling capability has been developed for detailed analysis of current and advanced reactor designs as part of a US-ROK collaborative I-NERI project. High fidelity was accomplished by integrating highly refined solution modules for the coupled neutronic, thermal-hydraulic, and thermo-mechanical phenomena. Each solution module employs methods and models that are formulated faithfully to the first-principles governing the physics, real geometry, and constituents. Specifically, the critical analysis elements that are incorporated in the coupled code capability are whole-core neutron transport solution, ultra-fine-mesh computational fluid dynamics/heat transfer solution, and finite-element-based thermo-mechanics solution, all obtained with explicit (fuel pin cell level) heterogeneous representations of the components of the core. The vast computational problem resulting from such highly refined modeling is solved on massively parallel computers, and serves as the 'numerical nuclear reactor'. Relaxation of modeling parameters were also pursued to make problems run on clusters of workstations and PCs for smaller scale applications as well
International Nuclear Information System (INIS)
Stosic, Z.V.; Stevanovic, V.D.
2001-01-01
A methodology for the simulation and analysis of one-phase and two-phase coolant flows around one or a row of spacers is presented. It is based on the multidimensional two-fluid mass, momentum and energy balance equations and application of adequate turbulence models. Necessary closure laws for interfacial transfer processes are presented. The stated general approach enables simulation and analyses of reactor coolant flow around spacers on different scale levels of the rod bundle geometry: detailed modelling of coolant flow around spacers and investigation of the influence of spacer's geometry on the coolant thermal-hydraulics, as well as prediction of global thermal-hydraulic parameters within the whole rod bundle with the investigation of the influence of rows of spacers on the bulk thermal-hydraulic processes. Sample problems are included illustrating these different modelling approaches. (author)
Numerical simulation in astrophysics
International Nuclear Information System (INIS)
Miyama, Shoken
1985-01-01
There have been many numerical simulations of hydrodynamical problems in astrophysics, e.g. processes of star formation, supernova explosion and formation of neutron stars, and general relativistic collapse of star to form black hole. The codes are made to be suitable for computing such problems. In astrophysical hydrodynamical problems, there are the characteristics: problems of self-gravity or external gravity acting, objects of scales very large or very short, objects changing by short period or long time scale, problems of magnetic force and/or centrifugal force acting. In this paper, we present one of methods of numerical simulations which may satisfy these requirements, so-called smoothed particle methods. We then introduce the methods briefly. Then, we show one of the applications of the methods to astrophysical problem (fragmentation and collapse of rotating isothermal cloud). (Mori, K.)
Numerical simulation of plasmas
International Nuclear Information System (INIS)
Dnestrovskii, Y.N.; Kostomarov, D.P.
1986-01-01
This book contains a modern consistent and systematic presentation of numerical computer simulation of plasmas in controlled thermonuclear fusion. The authors focus on the Soviet research in mathematical modelling of Tokamak plasmas, and present kinetic hydrodynamic and transport models with special emphasis on the more recent hybrid models. Compared with the first edition (in Russian) this book has been greatly revised and updated. (orig./WL)
Energy Technology Data Exchange (ETDEWEB)
David, Dijo K.; Mangarjuna Rao, P., E-mail: pmr@igcar.gov.in; Nashine, B.K.; Selvaraj, P.; Chellapandi, P.
2015-09-15
Highlights: • PAHR in SFR under large core relocation to in-vessel core catcher is numerically analyzed. • A 1-D thermal conduction model and a 2-D axisymmetric CFD model are developed for turbulent natural convection phenomenon. • The side pool (cold pool) was found out to be instrumental in storing heat and dissipating it to the heat sink. • Single tray type in-vessel core catcher is found to be thermally effective under one-fourth core relocation. - Abstract: A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal
Comments on numerical simulations
International Nuclear Information System (INIS)
Sato, T.
1984-01-01
The author comments on a couple of things about numerical simulation. One is just about the philosophical discussion that is, spontaneous or driven. The other thing is the numerical or technical one. Frankly, the author didn't want to touch on the technical matter because this should be a common sense one for those who are working at numerical simulation. But since many people take numerical simulation results at their face value, he would like to remind you of the reality hidden behind them. First, he would point out that the meaning of ''driven'' in driven reconnection is different from that defined by Schindler or Akasofu. The author's definition is closer to Axford's definition. In the spontaneous case, for some unpredicted reason an excess energy of the system is suddenly released at a certain point. However, one does not answer how such an unstable state far beyond a stable limit is realized in the magnetotail. In the driven case, there is a definite energy buildup phase starting from a stable state; namely, energy in the black box increases from a stable level subject to an external source. When the state has reached a certain position, the energy is released suddenly. The difference between driven and spontaneous is whether the cause (plasma flow) to trigger reconnection is specified or reconnection is triggered unpredictably. Another difference is that in driven reconnection the reconnection rate is dependent on the speed of the external plasma flow, but in spontaneous reconnection the rate is dependent on the internal condition such as the resistivity
International Nuclear Information System (INIS)
Georgieva, V.; Bogaerts, A.
2005-01-01
A one-dimensional particle-in-cell/Monte Carlo model is used to investigate Ar/CF 4 /N 2 discharges sustained in capacitively coupled dual frequency reactors, with special emphasis on the influence of the reactor parameters such as applied voltage amplitudes and frequencies of the two voltage sources. The presented calculation results include plasma density, ion current, average sheath potential and width, electron and ion average energies and energy distributions, and ionization rates. The simulations were carried out for high frequencies (HFs) of 27, 40, 60, and 100 MHz and a low frequency (LF) of 1 or 2 MHz, varying the LF voltage and keeping the HF voltage constant and vice versa. It is observed that the decoupling of the two sources is possible by increasing the applied HF to very high values (above 60 MHz) and it is not defined by the frequency ratio. Both voltage sources have influence on the plasma characteristics at a HF of 27 MHz and to some extent at 40 MHz. At HFs of 60 and 100 MHz, the plasma density and ion flux are determined only by the HF voltage source. The ion energy increases and the ion energy distribution function (IEDF) becomes broader with HF or LF voltage amplitude, when the other voltage is kept constant. The IEDF is broader with the increase of HF or the decrease of LF
Energy Technology Data Exchange (ETDEWEB)
NONE
2001-10-01
The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)
International Nuclear Information System (INIS)
Li Ying; Zhou Wenxia; Zhang Jige; Wang Dezhong
2009-01-01
In order to achieve the level of self-design and domestic manufacture of the reactor coolant pump (nuclear main pump), the software FLUENT was used to simulate the three-dimensional flow through full passage of one nuclear main pump basing on RNG κ-ε turbulence model and SIMPLE algorithm. The distribution of pressure and velocity of the flow in the impeller's surface was analyzed in different working conditions. Moreover, the performance of the pump was predicted based on the simulation results. The results show that the distributions of pressure and velocity are reasonable in both the working and back face of the blade in the steady working condition. The pressure of the flow is increased from the inlet to the outlet of the pump, and shows the maximal value in the impeller region. Comparatively satisfactory efficiency and head value were obtained in the condition of the pump design. The shaft power of the nuclear main pump is gradually increased with the increase of the flow flux. These results are helpful in understanding the change of the internal flow field in the nuclear main pump, which is of some importance for the pre-exploration and theoretical research on the domestic manufacture of the nuclear main pump. (authors)
Confidence in Numerical Simulations
Energy Technology Data Exchange (ETDEWEB)
Hemez, Francois M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2015-02-23
This PowerPoint presentation offers a high-level discussion of uncertainty, confidence and credibility in scientific Modeling and Simulation (M&S). It begins by briefly evoking M&S trends in computational physics and engineering. The first thrust of the discussion is to emphasize that the role of M&S in decision-making is either to support reasoning by similarity or to “forecast,” that is, make predictions about the future or extrapolate to settings or environments that cannot be tested experimentally. The second thrust is to explain that M&S-aided decision-making is an exercise in uncertainty management. The three broad classes of uncertainty in computational physics and engineering are variability and randomness, numerical uncertainty and model-form uncertainty. The last part of the discussion addresses how scientists “think.” This thought process parallels the scientific method where by a hypothesis is formulated, often accompanied by simplifying assumptions, then, physical experiments and numerical simulations are performed to confirm or reject the hypothesis. “Confidence” derives, not just from the levels of training and experience of analysts, but also from the rigor with which these assessments are performed, documented and peer-reviewed.
Confidence in Numerical Simulations
International Nuclear Information System (INIS)
Hemez, Francois M.
2015-01-01
This PowerPoint presentation offers a high-level discussion of uncertainty, confidence and credibility in scientific Modeling and Simulation (M&S). It begins by briefly evoking M&S trends in computational physics and engineering. The first thrust of the discussion is to emphasize that the role of M&S in decision-making is either to support reasoning by similarity or to ''forecast,'' that is, make predictions about the future or extrapolate to settings or environments that cannot be tested experimentally. The second thrust is to explain that M&S-aided decision-making is an exercise in uncertainty management. The three broad classes of uncertainty in computational physics and engineering are variability and randomness, numerical uncertainty and model-form uncertainty. The last part of the discussion addresses how scientists ''think.'' This thought process parallels the scientific method where by a hypothesis is formulated, often accompanied by simplifying assumptions, then, physical experiments and numerical simulations are performed to confirm or reject the hypothesis. ''Confidence'' derives, not just from the levels of training and experience of analysts, but also from the rigor with which these assessments are performed, documented and peer-reviewed.
Reactor refueling machine simulator
International Nuclear Information System (INIS)
Rohosky, T.L.; Swidwa, K.J.
1987-01-01
This patent describes in combination: a nuclear reactor; a refueling machine having a bridge, trolley and hoist each driven by a separate motor having feedback means for generating a feedback signal indicative of movement thereof. The motors are operable to position the refueling machine over the nuclear reactor for refueling the same. The refueling machine also has a removable control console including means for selectively generating separate motor signals for operating the bridge, trolley and hoist motors and for processing the feedback signals to generate an indication of the positions thereof, separate output leads connecting each of the motor signals to the respective refueling machine motor, and separate input leads for connecting each of the feedback means to the console; and a portable simulator unit comprising: a single simulator motor; a single simulator feedback signal generator connected to the simulator motor for generating a simulator feedback signal in response to operation of the simulator motor; means for selectively connecting the output leads of the console to the simulator unit in place of the refueling machine motors, and for connecting the console input leads to the simulator unit in place of the refueling machine motor feedback means; and means for driving the single simulator motor in response to any of the bridge, trolley or hoist motor signals generated by the console and means for applying the simulator feedback signal to the console input lead associated with the motor signal being generated by the control console
Numerical aerodynamic simulation (NAS)
International Nuclear Information System (INIS)
Peterson, V.L.; Ballhaus, W.F. Jr.; Bailey, F.R.
1984-01-01
The Numerical Aerodynamic Simulation (NAS) Program is designed to provide a leading-edge computational capability to the aerospace community. It was recognized early in the program that, in addition to more advanced computers, the entire computational process ranging from problem formulation to publication of results needed to be improved to realize the full impact of computational aerodynamics. Therefore, the NAS Program has been structured to focus on the development of a complete system that can be upgraded periodically with minimum impact on the user and on the inventory of applications software. The implementation phase of the program is now under way. It is based upon nearly 8 yr of study and should culminate in an initial operational capability before 1986. The objective of this paper is fivefold: 1) to discuss the factors motivating the NAS program, 2) to provide a history of the activity, 3) to describe each of the elements of the processing-system network, 4) to outline the proposed allocation of time to users of the facility, and 5) to describe some of the candidate problems being considered for the first benchmark codes
International Nuclear Information System (INIS)
Toti, A.; Vierendeels, J.; Belloni, F.
2017-01-01
Highlights: • A system thermal-hydraulic/CFD coupling methodology is proposed for high-fidelity transient flow analyses. • The method is based on domain decomposition and implicit numerical scheme. • A novel interface Quasi-Newton algorithm is implemented to improve stability and convergence rate. • Preliminary validation analyses on the TALL-3D experiment. - Abstract: The paper describes the development and validation of a coupling methodology between the best-estimate system thermal-hydraulic code RELAP5-3D and the CFD code FLUENT, conceived for high fidelity plant-scale safety analyses of pool-type reactors. The computational tool is developed to assess the impact of three-dimensional phenomena occurring in accidental transients such as loss of flow (LOF) in the research reactor MYRRHA, currently in the design phase at the Belgian Nuclear Research Centre, SCK• CEN. A partitioned, implicit domain decomposition coupling algorithm is implemented, in which the coupled domains exchange thermal-hydraulics variables at coupling boundary interfaces. Numerical stability and interface convergence rates are improved by a novel interface Quasi-Newton algorithm, which is compared in this paper with previously tested numerical schemes. The developed computational method has been assessed for validation purposes against the experiment performed at the test facility TALL-3D, operated by the Royal Institute of Technology (KTH) in Sweden. This paper details the results of the simulation of a loss of forced convection test, showing the capability of the developed methodology to predict transients influenced by local three-dimensional phenomena.
International Nuclear Information System (INIS)
Orr, B. R.
1999-01-01
Studies of flow through the unsaturated zone and perched ground-water zones above the Snake River Plain aquifer are part of the overall assessment of ground-water flow and determination of the fate and transport of contaminants in the subsurface at the Idaho National Engineering and Environmental Laboratory (INEEL). These studies include definition of the hydrologic controls on the formation of perched ground-water zones and description of the transport and fate of wastewater constituents as they moved through the unsaturated zone. The definition of hydrologic controls requires stratigraphic correlation of basalt flows and sedimentary interbeds within the saturated zone, analysis of hydraulic properties of unsaturated-zone rocks, numerical modeling of the formation of perched ground-water zones, and batch and column experiments to determine rock-water geochemical processes. This report describes the development of a transient numerical simulation that was used to evaluate a conceptual model of flow through perched ground-water zones beneath wastewater infiltration ponds at the Test Reactor Area (TRA)
International Nuclear Information System (INIS)
Xue Xiuli; Yang Hongyi; Yang Fuchang
2008-01-01
Detailed 3D thermal-hydraulic numerical analyses to the fuel zone outlet are actualized with the STAR-CD CFD code. The performance of sodium mixing is studied and detailed velocity and temperature distribution are obtained in this region which will offer foundations and references to study the rationality of temperature monitoring-spot arrangement and to assess the effect of temperature fluctuations to control rod guide tubes in this region, and so on. (authors)
International Nuclear Information System (INIS)
Wei Yuanyuan; Lu Daogang
2009-01-01
There is the free surface in the main vessel of fast reactor, when long period earthquakes happen, the fluid will impact the coping of vessel and make the reactor dangerous. The flow of the fluid was simulated by moving particle semi-implicit method. The phenomenon on sloshing response of the free surface in the main vessel of fast reactor excited by 3 sine waves was simulated. The impact pressure from the research can provide important loadings for the integrality analysis of the main vessel. (authors)
Reactor core simulations in Canada
International Nuclear Information System (INIS)
Roy, R.; Koclas, J.; Shen, W.; Jenkins, D. A.; Altiparmakov, D.; Rouben, B.
2004-01-01
This review will address the current simulation flow-chart currently used for reactor-physics simulations in the Canadian industry. The neutron behaviour in heavy-water moderated power reactors is quite different from that in other power reactors, thus the core physics approximations are somewhat different Some codes used are particular to the context of heavy-water reactors, and the paper focuses on this aspect. The paper also shows simulations involving new design features of the Advanced Candu Reactor TM (ACR TM), and provides insight into future development, expected in the coming years. (authors)
International Nuclear Information System (INIS)
Zaza, Chady
2015-01-01
The numerical simulation of steam generators of pressurized water reactors is a complex problem, involving different flow regimes and a wide range of length and time scales. An accidental scenario may be associated with very fast variations of the flow with an important Mach number. In contrast in the nominal regime the flow may be stationary, at low Mach number. Moreover whatever the regime under consideration, the array of U-tubes is modelled by a porous medium in order to avoid taking into account the complex geometry of the steam generator, which entails the issue of the coupling conditions at the interface with the free-fluid. We propose a new pressure-correction scheme for cell-centered finite volumes for solving the compressible Navier-Stokes and Euler equations at all Mach number. The existence of a discrete solution, the consistency of the scheme in the Lax sense and the positivity of the internal energy were proved. Then the scheme was extended to the homogeneous two-phase flow models of the GENEPI code developed at CEA. Lastly a multigrid-AMR algorithm was adapted for using our pressure-correction scheme on adaptive grids. Regarding the second issue addressed in this work, the numerical simulation of a fluid flow over a porous bed involves very different length scales. Macroscopic interface models - such as Ochoa-Tapia-Whitaker or Beavers-Joseph law for a viscous flow - represent the transition region between the free-fluid and the porous region by an interface of discontinuity associated with specific transmission conditions. An extension to the Beavers-Joseph law was proposed for the convective regime. By introducing a jump in the kinetic energy at the interface, we recover an interface condition close to the Beavers-Joseph law but with a non-linear slip coefficient, which depends on the free-fluid velocity at the interface and on the Darcy velocity. The validity of this new transmission condition was assessed with direct numerical simulations at
Simulator for materials testing reactors
International Nuclear Information System (INIS)
Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo
2013-06-01
A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)
Simulation development for TRIGA reactor
International Nuclear Information System (INIS)
Handoyo, D.
1997-01-01
A simulator of the dynamic of TRIGA reactor has been made. this simulator is meant to study the reactor kinetic behavior and for operator training to more assure the safety and the reliability of the real operation of TRIGA reactor. the simulator consists of PC (Personal Computer) for processing the calculation of reactivity, neutron flux, period, ect and control panel for regulating the input data such as the change of power range, control rod position as well as cooling flow rate. the result will be displayed on screen monitor of personal computer as given in the real control room of TRIGA reactor. the output of simulator will be verified by comparing with measurement result in the real TRIGA MARK II reactor of Musashi institute of technology. for the change of reactivity of 0.3, 0.5 and 0.7 the reactor power and fuel temperature between the simulator and measurements are comparable
Numerical simulation of welding
DEFF Research Database (Denmark)
Hansen, Jan Langkjær; Thorborg, Jesper
Aim of project:To analyse and model the transient thermal field from arc welding (SMAW, V-shaped buttweld in 15mm plate) and to some extend the mechanical response due to the thermal field. - To implement this model in a general purpose finite element program such as ABAQUS.The simulation...... stress is also taken into account.Work carried out:With few means it is possible to define a thermal model which describes the thermal field from the welding process in reasonable agreement with reality. Identical results are found with ABAQUS and Rosenthal’s analytical solution of the governing heat...... transfer equation under same conditions. It is relative easy tointroduce boundary conditions such as convection and radiation where not surprisingly the radiation has the greatest influence especially from the high temperature regions in the weld pool and the heat affected zone.Due to the large temperature...
Numerical simulation of flood barriers
Srb, Pavel; Petrů, Michal; Kulhavý, Petr
This paper deals with testing and numerical simulating of flood barriers. The Czech Republic has been hit by several very devastating floods in past years. These floods caused several dozens of causalities and property damage reached billions of Euros. The development of flood measures is very important, especially for the reduction the number of casualties and the amount of property damage. The aim of flood control measures is the detention of water outside populated areas and drainage of water from populated areas as soon as possible. For new flood barrier design it is very important to know its behaviour in case of a real flood. During the development of the barrier several standardized tests have to be carried out. Based on the results from these tests numerical simulation was compiled using Abaqus software and some analyses were carried out. Based on these numerical simulations it will be possible to predict the behaviour of barriers and thus improve their design.
Numerical simulation of laser resonators
International Nuclear Information System (INIS)
Yoo, J. G.; Jeong, Y. U.; Lee, B. C.; Rhee, Y. J.; Cho, S. O.
2004-01-01
We developed numerical simulation packages for laser resonators on the bases of a pair of integral equations. Two numerical schemes, a matrix formalism and an iterative method, were programmed for finding numeric solutions to the pair of integral equations. The iterative method was tried by Fox and Li, but it was not applicable for high Fresnel numbers since the numerical errors involved propagate and accumulate uncontrollably. In this paper, we implement the matrix method to extend the computational limit further. A great number of case studies are carried out with various configurations of stable and unstable r;esonators to compute diffraction losses, phase shifts, intensity distributions and phases of the radiation fields on mirrors. Our results presented in this paper show not only a good agreement with the results previously obtained by Fox and Li, but also the legitimacy of our numerical procedures for high Fresnel numbers.
International Nuclear Information System (INIS)
Tauveron, N.
2006-02-01
The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge
Numerical simulation in plasma physics
International Nuclear Information System (INIS)
Samarskii, A.A.
1980-01-01
Plasma physics is not only a field for development of physical theories and mathematical models but also an object of application of the computational experiment comprising analytical and numerical methods adapted for computers. The author considers only MHD plasma physics problems. Examples treated are dissipative structures in plasma; MHD model of solar dynamo; supernova explosion simulation; and plasma compression by a liner. (Auth.)
Chernobyl reactor transient simulation study
International Nuclear Information System (INIS)
Gaber, F.A.; El Messiry, A.M.
1988-01-01
This paper deals with the Chernobyl nuclear power station transient simulation study. The Chernobyl (RBMK) reactor is a graphite moderated pressure tube type reactor. It is cooled by circulating light water that boils in the upper parts of vertical pressure tubes to produce steam. At equilibrium fuel irradiation, the RBMK reactor has a positive void reactivity coefficient. However, the fuel temperature coefficient is negative and the net effect of a power change depends upon the power level. Under normal operating conditions the net effect (power coefficient) is negative at full power and becomes positive under certain transient conditions. A series of dynamic performance transient analysis for RBMK reactor, pressurized water reactor (PWR) and fast breeder reactor (FBR) have been performed using digital simulator codes, the purpose of this transient study is to show that an accident of Chernobyl's severity does not occur in PWR or FBR nuclear power reactors. This appears from the study of the inherent, stability of RBMK, PWR and FBR under certain transient conditions. This inherent stability is related to the effect of the feed back reactivity. The power distribution stability in the graphite RBMK reactor is difficult to maintain throughout its entire life, so the reactor has an inherent instability. PWR has larger negative temperature coefficient of reactivity, therefore, the PWR by itself has a large amount of natural stability, so PWR is inherently safe. FBR has positive sodium expansion coefficient, therefore it has insufficient stability it has been concluded that PWR has safe operation than FBR and RBMK reactors
Numerical simulation of Higgs models
International Nuclear Information System (INIS)
Jaster, A.
1995-10-01
The SU(2) Higgs and the Schwinger model on the lattice were analysed. Numerical simulations of the SU(2) Higgs model were performed to study the finite temperature electroweak phase transition. With the help of the multicanonical method the distribution of an order parameter at the phase transition point was measured. This was used to obtain the order of the phase transition and the value of the interface tension with the histogram method. Numerical simulations were also performed at zero temperature to perform renormalization. The measured values for the Wilson loops were used to determine the static potential and from this the renormalized gauge coupling. The Schwinger model was simulated at different gauge couplings to analyse the properties of the Kaplan-Shamir fermions. The prediction that the mass parameter gets only multiplicative renormalization was tested and verified. (orig.)
A Numerical Model for Trickle Bed Reactors
Propp, Richard M.; Colella, Phillip; Crutchfield, William Y.; Day, Marcus S.
2000-12-01
Trickle bed reactors are governed by equations of flow in porous media such as Darcy's law and the conservation of mass. Our numerical method for solving these equations is based on a total-velocity splitting, sequential formulation which leads to an implicit pressure equation and a semi-implicit mass conservation equation. We use high-resolution finite-difference methods to discretize these equations. Our solution scheme extends previous work in modeling porous media flows in two ways. First, we incorporate physical effects due to capillary pressure, a nonlinear inlet boundary condition, spatial porosity variations, and inertial effects on phase mobilities. In particular, capillary forces introduce a parabolic component into the recast evolution equation, and the inertial effects give rise to hyperbolic nonconvexity. Second, we introduce a modification of the slope-limiting algorithm to prevent our numerical method from producing spurious shocks. We present a numerical algorithm for accommodating these difficulties, show the algorithm is second-order accurate, and demonstrate its performance on a number of simplified problems relevant to trickle bed reactor modeling.
A numerical technique for reactor subchannel analysis
International Nuclear Information System (INIS)
Fath, Hassan E.S.
1983-01-01
A numerical technique is developed for the solution of the transient boundary layer equations with a moving liquid-vapour interface boundary. The technique uses the finite difference method with the velocity components defined over an Eulerian mesh. A system of interface massless markers is defined where the markers move with the flow field according to a simple kinematic relation between the interface geometry and the fluid velocity. Different applications of nuclear engineering interest are reported with some available results. The present technique is capable of predicting the interface profile near the wall which is important in the reactor subchannel analysis
Combining Narrative and Numerical Simulation
DEFF Research Database (Denmark)
Hansen, Mette Sanne; Ladeby, Klaes Rohde; Rasmussen, Lauge Baungaard
2011-01-01
for decision makers to systematically test several different outputs of possible solutions in order to prepare for future consequences. The CSA can be a way to evaluate risks and address possible unforeseen problems in a more methodical way than either guessing or forecasting. This paper contributes...... to the decision making in operations and production management by providing new insights into modelling and simulation based on the combined narrative and numerical simulation approach as a tool for strategy making. The research question asks, “How can the CSA be applied in a practical context to support strategy...... making?” The paper uses a case study where interviews and observations were carried out in a Danish corporation. The CSA is a new way to address decision making and has both practical value and further expands the use of strategic simulation as a management tool....
Compact tokamak reactors part 2 (numerical results)
International Nuclear Information System (INIS)
Wiley, J.C.; Wootton, A.J.; Ross, D.W.
1996-01-01
The authors describe a numerical optimization scheme for fusion reactors. The particular application described is to find the smallest copper coil spherical tokamak, although the numerical scheme is sufficiently general to allow many other problems to be solved. The solution to the steady state energy balance is found by first selecting the fixed variables. The range of all remaining variables is then selected, except for the temperature. Within the specified ranges, the temperature which satisfies the power balance is then found. Tests are applied to determine that remaining constraints are satisfied, and the acceptable results then stored. Results are presented for a range of auxiliary current drive efficiencies and different scaling relationships; for the range of variables chosen the machine encompassing volume increases or remains approximately unchanged as the aspect ratio is reduced
Numerical simulation of fire vortex
Barannikova, D. D.; Borzykh, V. E.; Obukhov, A. G.
2018-05-01
The article considers the numerical simulation of the swirling flow of air around the smoothly heated vertical cylindrical domain in the conditions of gravity and Coriolis forces action. The solutions of the complete system of Navie-Stocks equations are numerically solved at constant viscosity and heat conductivity factors. Along with the proposed initial and boundary conditions, these solutions describe the complex non-stationary 3D flows of viscous compressible heat conducting gas. For various instants of time of the initial flow formation stage using the explicit finite-difference scheme the calculations of all gas dynamics parameters, that is density, temperature, pressure and three velocity components of gas particles, have been run. The current instant lines corresponding to the trajectories of the particles movement in the emerging flow have been constructed. A negative direction of the air flow swirling occurred in the vertical cylindrical domain heating has been defined.
Plasma modelling and numerical simulation
International Nuclear Information System (INIS)
Van Dijk, J; Kroesen, G M W; Bogaerts, A
2009-01-01
Plasma modelling is an exciting subject in which virtually all physical disciplines are represented. Plasma models combine the electromagnetic, statistical and fluid dynamical theories that have their roots in the 19th century with the modern insights concerning the structure of matter that were developed throughout the 20th century. The present cluster issue consists of 20 invited contributions, which are representative of the state of the art in plasma modelling and numerical simulation. These contributions provide an in-depth discussion of the major theories and modelling and simulation strategies, and their applications to contemporary plasma-based technologies. In this editorial review, we introduce and complement those papers by providing a bird's eye perspective on plasma modelling and discussing the historical context in which it has surfaced. (editorial review)
Water simulation of sodium reactors
International Nuclear Information System (INIS)
Grewal, S.S.; Gluekler, E.L.
1981-01-01
The thermal hydraulic simulation of a large sodium reactor by a scaled water model is examined. The Richardson Number, friction coefficient and the Peclet Number can be closely matched with the water system at full power and the similarity is retained for buoyancy driven flows. The simulation of thermal-hydraulic conditions in a reactor vessel provided by a scaled water experiment is better than that by a scaled sodium test. Results from a correctly scaled water test can be tentatively extrapolated to a full size sodium system
Development of Reactor Console Simulator for PUSPATI TRIGA Reactor
International Nuclear Information System (INIS)
Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat
2012-01-01
The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)
Numerical methods used in simulation
International Nuclear Information System (INIS)
Caseau, Paul; Perrin, Michel; Planchard, Jacques
1978-01-01
The fundamental numerical problem posed by simulation problems is the stability of the resolution diagram. The system of the most used equations is defined, since there is a family of models of increasing complexity with 3, 4 or 5 equations although only models with 3 and 4 equations have been used extensively. After defining what is meant by explicit or implicit, the best established stability results is given for one-dimension problems and then for two-dimension problems. It is shown that two types of discretisation may be defined: four and eight point diagrams (in one or two dimensions) and six and ten point diagrams (in one or two dimensions). To end, some results are given on problems that are not usually treated very much, i.e. non-asymptotic stability and the stability of diagrams based on finite elements [fr
Real time simulator for material testing reactor
Energy Technology Data Exchange (ETDEWEB)
Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)
2012-03-15
Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)
Real time simulator for material testing reactor
International Nuclear Information System (INIS)
Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo
2012-01-01
Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)
Final Stage Development of Reactor Console Simulator
International Nuclear Information System (INIS)
Mohamad Idris Taib; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Nurfarhana Ayuni Joha
2013-01-01
The Reactor Console Simulator PUSPATI TRIGA Reactor was developed since end of 2011 and now in the final stage of development. It is will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behavior and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of human system interface (HSI) is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate and estimated reactor console parameters. The capabilities in user interface, reactor physics and thermal-hydraulics can be expanded and explored to simulation as well as modeling for New Reactor Console, Research Reactor and Nuclear Power Plant. (author)
Numerical simulation of two phase flows in heat exchangers
International Nuclear Information System (INIS)
Grandotto Biettoli, M.
2006-04-01
The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)
Status report on high fidelity reactor simulation
International Nuclear Information System (INIS)
Palmiotti, G.; Smith, M.; Rabiti, C.; Lewis, E.; Yang, W.; Leclere, M.; Siegel, A.; Fischer, P.; Kaushik, D.; Ragusa, J.; Lottes, J.; Smith, B.
2006-01-01
This report presents the effort under way at Argonne National Laboratory toward a comprehensive, integrated computational tool intended mainly for the high-fidelity simulation of sodium-cooled fast reactors. The main activities carried out involved neutronics, thermal hydraulics, coupling strategies, software architecture, and high-performance computing. A new neutronics code, UNIC, is being developed. The first phase involves the application of a spherical harmonics method to a general, unstructured three-dimensional mesh. The method also has been interfaced with a method of characteristics. The spherical harmonics equations were implemented in a stand-alone code that was then used to solve several benchmark problems. For thermal hydraulics, a computational fluid dynamics code called Nek5000, developed in the Mathematics and Computer Science Division for coupled hydrodynamics and heat transfer, has been applied to a single-pin, periodic cell in the wire-wrap geometry typical of advanced burner reactors. Numerical strategies for multiphysics coupling have been considered and higher-accuracy efficient methods proposed to finely simulate coupled neutronic/thermal-hydraulic reactor transients. Initial steps have been taken in order to couple UNIC and Nek5000, and simplified problems have been defined and solved for testing. Furthermore, we have begun developing a lightweight computational framework, based in part on carefully selected open source tools, to nonobtrusively and efficiently integrate the individual physics modules into a unified simulation tool
Nuclear characteristic simulation device for reactor core
International Nuclear Information System (INIS)
Arakawa, Akio; Kobayashi, Yuji.
1994-01-01
In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)
Relativistic positioning systems: Numerical simulations
Puchades Colmenero, Neus
The position of users located on the Earth's surface or near it may be found with the classic positioning systems (CPS). Certain information broadcast by satellites of global navigation systems, as GPS and GALILEO, may be used for positioning. The CPS are based on the Newtonian formalism, although relativistic post-Newtonian corrections are done when they are necessary. This thesis contributes to the development of a different positioning approach, which is fully relativistic from the beginning. In the relativistic positioning systems (RPS), the space-time position of any user (ship, spacecraft, and so on) can be calculated with the help of four satellites, which broadcast their proper times by means of codified electromagnetic signals. In this thesis, we have simulated satellite 4-tuples of the GPS and GALILEO constellations. If a user receives the signals from four satellites simultaneously, the emission proper times read -after decoding- are the user "emission coordinates". In order to find the user "positioning coordinates", in an appropriate almost inertial reference system, there are two possibilities: (a) the explicit relation between positioning and emission coordinates (broadcast by the satellites) is analytically found or (b) numerical codes are designed to calculate the positioning coordinates from the emission ones. Method (a) is only viable in simple ideal cases, whereas (b) allows us to consider realistic situations. In this thesis, we have designed numerical codes with the essential aim of studying two appropriate RPS, which may be generalized. Sometimes, there are two real users placed in different positions, which receive the same proper times from the same satellites; then, we say that there is bifurcation, and additional data are needed to choose the real user position. In this thesis, bifurcation is studied in detail. We have analyzed in depth two RPS models; in both, it is considered that the satellites move in the Schwarzschild's space
Pressurized water reactor simulator. Workshop material
International Nuclear Information System (INIS)
2003-01-01
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator
Programming for a nuclear reactor instrument simulator
International Nuclear Information System (INIS)
Cohn, C.E.
1989-01-01
A new computerized control system for a transient test reactor incorporates a simulator for pre-operational testing of control programs. The part of the simulator pertinent to the discussion here consists of two microprocessors. An 8086/8087 reactor simulator calculates simulated reactor power by solving the reactor kinetics equations. An 8086 instrument simulator takes the most recent power value developed by the reactor simulator and simulates the appropriate reading on each of the eleven reactor instruments. Since the system is required to run on a one millisecond cycle, careful programming was required to take care of all eleven instruments in that short time. This note describes the special programming techniques used to attain the needed performance
Boiling water reactor simulator. Workshop material
International Nuclear Information System (INIS)
2003-01-01
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA
Global physical and numerical stability of a nuclear reactor core
International Nuclear Information System (INIS)
Morales-Sandoval, Jaime; Hernandez-Solis, Augusto
2005-01-01
Low order models are used to investigate the influence of integration methods on observed power oscillations of some nuclear reactor simulators. The zero-power point reactor kinetics with six-delayed neutron precursor groups are time discretized using explicit, implicit and Crank-Nicholson methods, and the stability limit of the time mesh spacing is exactly obtained by locating their characteristic poles in the z-transform plane. These poles are the s to z mappings of the inhour equation roots and, except for one of them, they show little or no dependence on the integration method. Conditions for stable power oscillations can be also obtained by tracking when steady state output signals resulting from reactivity oscillations in the s-Laplace plane cross the imaginary axis. The dynamics of a BWR core operating at power conditions is represented by a reduced order model obtained by adding three ordinary differential equations, which can model void and Doppler reactivity feedback effects on power, and collapsing all delayed neutron precursors in one group. Void dynamics are modeled as a second order system and fuel heat transfer as a first order system. This model shows rich characteristics in terms of indicating the relative importance of different core parameters and conditions on both numerical and physical oscillations observed by large computer code simulations. A brief discussion of the influence of actual core and coolant conditions on the reduced order model is presented
Simulated nuclear reactor fuel assembly
International Nuclear Information System (INIS)
Berta, V.T.
1993-01-01
An apparatus for electrically simulating a nuclear reactor fuel assembly. It includes a heater assembly having a top end and a bottom end and a plurality of concentric heater tubes having electrical circuitry connected to a power source, and radially spaced from each other. An outer target tube and an inner target tube is concentric with the heater tubes and with each other, and the outer target tube surrounds and is radially spaced from the heater tubes. The inner target tube is surrounded by and radially spaced from the heater tubes and outer target tube. The top of the assembly is generally open to allow for the electrical power connection to the heater tubes, and the bottom of the assembly includes means for completing the electrical circuitry in the heater tubes to provide electrical resistance heating to simulate the power profile in a nuclear reactor. The embedded conductor elements in each heater tube is split into two halves for a substantial portion of its length and provided with electrical isolation such that each half of the conductor is joined at one end and is not joined at the other end
WWER-1000 reactor simulator. Workshop material
International Nuclear Information System (INIS)
2003-01-01
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series 12, 'Reactor Simulator Development' (2001). Course material for workshops using a pressurized water reactor (PWR) Simulator developed for the IAEA by Cassiopeia Technologies Inc. of Canada is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003) and Training Course Series No. 23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation. N. V. Tikhonov and S. B. Vygovsky of the Moscow Engineering and Physics Institute prepared this report for the IAEA
Simulation of Molten Salt Reactor dynamics
International Nuclear Information System (INIS)
Krepel, J.; Rohde, U.; Grundmann, U.
2005-01-01
Dynamics of the Molten Salt Reactor - one of the 'Generation IV' concepts - was studied in this paper. The graphite-moderated channel type MSR was selected for the numerical simulation of the reactor with liquid fuel. The MSR dynamics is very specific because of two physical peculiarities of the liquid fueled reactor: the delayed neutrons precursors are drifted by the fuel flow and the fission energy is immediately released directly into the coolant. Presently, there are not many accessible numerical codes appropriate for the MSR simulation, therefore the DYN3D-MSR code was developed based on the FZR in-house code DYN3D. It allows calculating of full 3D transient neutronics in combination with parallel channel type thermal-hydraulics. By means of DYN3D-MSR, several transients typical for the liquid fuel system were analyzed. Those transients were initiated by reactivity insertion, by overcooling of fuel at the core inlet, by the fuel pump start-up or coast-down, or by the blockage of selected fuel channels. In these considered transients, the response of the MSR is characterized by the immediate change of the fuel temperature with changing power and fast negative temperature feedback to the power. The response through the graphite temperature is slower. Furthermore, for big MSR cores fueled with U233 the graphite feedback coefficient can be positive. In this case the addition of erbium to the graphite can ensure the inherent safety features. The DYN3D-MSR code has been shown to be an effective tool for MSR dynamics studies. (author)
Numerical simulation of muzzle blast
Tyler-Street, M.
2014-01-01
Structural design methods for naval ships include environmental, operational and military load cases. One of the operational loads acting on a typical naval vessel is the muzzle blast from a gun. Simulating the muzzle blast load acting on a ship structure with CFD and ALE methods leads to large
NUMERICAL SIMULATION AND OPTIMIZATION OF ...
African Journals Online (AJOL)
30 juin 2011 ... This article has as an aim the study and the simulation of the photovoltaic cells containing CdTe materials, contributing to the development of renewable energies, and able to feed from the houses, the shelters as well as ... and the output energy of conversion is 18.26%.Optimization is made according to the.
Numerical methods in simulation of resistance welding
DEFF Research Database (Denmark)
Nielsen, Chris Valentin; Martins, Paulo A.F.; Zhang, Wenqi
2015-01-01
Finite element simulation of resistance welding requires coupling betweenmechanical, thermal and electrical models. This paper presents the numerical models and theircouplings that are utilized in the computer program SORPAS. A mechanical model based onthe irreducible flow formulation is utilized...... a resistance welding point of view, the most essential coupling between the above mentioned models is the heat generation by electrical current due to Joule heating. The interaction between multiple objects is anothercritical feature of the numerical simulation of resistance welding because it influences...... thecontact area and the distribution of contact pressure. The numerical simulation of resistancewelding is illustrated by a spot welding example that includes subsequent tensile shear testing...
Coincidental match of numerical simulation and physics
Pierre, B.; Gudmundsson, J. S.
2010-08-01
Consequences of rapid pressure transients in pipelines range from increased fatigue to leakages and to complete ruptures of pipeline. Therefore, accurate predictions of rapid pressure transients in pipelines using numerical simulations are critical. State of the art modelling of pressure transient in general, and water hammer in particular include unsteady friction in addition to the steady frictional pressure drop, and numerical simulations rely on the method of characteristics. Comparison of rapid pressure transient calculations by the method of characteristics and a selected high resolution finite volume method highlights issues related to modelling of pressure waves and illustrates that matches between numerical simulations and physics are purely coincidental.
International Nuclear Information System (INIS)
Schweizer, Fernando Lage Araújo
2014-01-01
The Brazilian Multipurpose Reactor (RMB) consists in a 30 MW open pool research reactor and its design is currently in development. The RMB is intended to produce a neutron flux applied at material irradiation for radioisotope production and materials and nuclear fuel tests. The reactor is immersed in a deep water pool needed for radiation shielding and thermal protection. A heating and purifying system is applied in research reactors with high thermal power in order to create a Hot Water Layer (HWL) on the pool top preventing that contaminated water from the reactor core neighboring reaches its surface reducing the room radiation dose rate. This dissertation presents a study of the HWL behavior during the reactor operation first hours where perturbations due to the cooling system and pool heating induce a mixing flow in the HWL reducing its protection. Numerical simulations using the CFD code CFX 14.0 have been performed for theoretical dose rate estimation during reactor operation, for a 1/10 scaled down model using dimensional analysis and mesh testing as an initial verification of the commercial code application. Equipment and sensor needed for an experimental bench project were defined by the CFD numerical simulation. (author)
Numerical modeling of a nuclear production reactor cooling lake
International Nuclear Information System (INIS)
Hamm, L.L.; Pepper, D.W.
1987-01-01
A finite element model has been developed which predicts flow and temperature distributions within a nuclear reactor cooling lake at the Savannah River Plant near Aiken, South Carolina. Numerical results agree with values obtained from a 3-D EPA numerical lake model and actual measurements obtained from the lake. Because the effluent water from the reactor heat exchangers discharges directly into the lake, downstream temperatures at mid-lake could exceed the South Carolina DHEC guidelines for thermal exchanges during the summer months. Therefore, reactor power was reduced to maintain temperature compliance at mid-lake. Thermal mitigation measures were studied that included placing a 6.1 m deep fabric curtain across mid-lake and moving the reactor outfall upstream. These measurements were calculated to permit about an 8% improvement in reactor power during summer operation
Numerical simulation of edge plasma in tokamak
International Nuclear Information System (INIS)
Chen Yiping; Qiu Lijian
1996-02-01
The transport process and transport property of plasma in edge layer of Tokamak are simulated by solving numerically two-dimensional and multi-fluid plasma transport equations using suitable simulation code. The simulation results can show plasma parameter distribution characteristics in the area of edge layer, especially the characteristics near the first wall and divertor target plate. The simulation results play an important role in the design of divertor and first wall of Tokamak. (2 figs)
Simulation of a pool type research reactor
International Nuclear Information System (INIS)
Oliveira, Andre Felipe da Silva de; Moreira, Maria de Lourdes
2011-01-01
Computational fluid dynamic is used to simulate natural circulation condition after a research reactor shutdown. A benchmark problem was used to test the viability of usage such code to simulate the reactor model. A model which contains the core, the pool, the reflector tank, the circulation pipes and chimney was simulated. The reactor core contained in the full scale model was represented by a porous media. The parameters of porous media were obtained from a separate CFD analysis of the full core model. Results demonstrate that such studies can be carried out for research and test of reactors design. (author)
Visualization of numerically simulated aerodynamic flow fields
International Nuclear Information System (INIS)
Hian, Q.L.; Damodaran, M.
1991-01-01
The focus of this paper is to describe the development and the application of an interactive integrated software to visualize numerically simulated aerodynamic flow fields so as to enable the practitioner of computational fluid dynamics to diagnose the numerical simulation and to elucidate essential flow physics from the simulation. The input to the software is the numerical database crunched by a supercomputer and typically consists of flow variables and computational grid geometry. This flow visualization system (FVS), written in C language is targetted at the Personal IRIS Workstations. In order to demonstrate the various visualization modules, the paper also describes the application of this software to visualize two- and three-dimensional flow fields past aerodynamic configurations which have been numerically simulated on the NEC-SXIA Supercomputer. 6 refs
Numerical simulations of disordered superconductors
International Nuclear Information System (INIS)
Bedell, K.S.; Gubernatis, J.E.; Scalettar, R.T.; Zimanyi, G.T.
1997-01-01
This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at Los Alamos National Laboratory (LANL). The authors carried out Monte Carlo studies of the critical behavior of superfluid 4 He in aerogel. They found the superfluid density exponent increases in the presence of fractal disorder with a value roughly consistent with experimental results. They also addressed the localization of flux lines caused by splayed columnar pins. Using a Sine-Gordon-type of renormalization group study they obtained an analytic form for the critical temperature. They also determined the critical temperature from I-V characteristics obtained from a molecular dynamics simulation. The combined studies enabled one to construct the phase diagram as a function of interaction strength, temperature, and disorder. They also employed the recently developed mapping between boson world-lines and the flux motion to use quantum Monte Carlo simulations to analyze localization in the presence of disorder. From measurements of the transverse flux line wandering, they determined the critical ratio of columnar to point disorder strength needed to localize the bosons
Numerical simulation of HPT processing
International Nuclear Information System (INIS)
Verleysen, P; Van den Abeele, F; Degrieck, J
2014-01-01
The principle of achieving high strength and superior properties in metal alloys through the application of severe plastic deformation has been exploited in the metal processing industry for many decades. In this contribution finite element simulations are presented of the HPT process. As opposed to most studies in literature, in which rigid sample holders are considered, the real elasto-plastic behavior of the holders is modeled. The simulations show that during the compression stage, plastic deformation occurs in the holders: initially, at the outside boundary of the sample cavity and, at a later stage, underneath the centre of the sample. The latter region of plastic deformation is rapidly growing and has a non-negligible effect on the response of the sample. Major conclusion is that the sample holders, and more specific, their deformability is key for the conditions in the specimen. Indeed, it severely affects important parameters for both the microstructural changes in the sample material, such as the amplitude and distribution of the hydrostatic stress, and its final shape
Numerical simulation of hypersonic flight experiment vehicle
Yamamoto, Yukimitsu; Yoshioka, Minako; 山本 行光; 吉岡 美菜子
1994-01-01
Hypersonic aerodynamic characteristics of Hypersonic FLight EXperiment (HYFLEX vehicle were investigated by numerical simulations using Navier-Stokes CFD (Computational Fluid Dynamics) code of NAL. Numerical results were compared with experimental data obtained at Hypersonic Wind Tunnel at NAL. In order to investigate real flight aerodynamic characteristics. numerical calculations corresponding to the flight conditions suffering from maximum aero thermodynamic heating were also made and the d...
Numerical simulation of mechatronic sensors and actuators
Kaltenbacher, Manfred
2007-01-01
Focuses on the physical modeling of mechatronic sensors and actuators and their precise numerical simulation using the Finite Element Method (FEM). This book discusses the physical modeling as well as numerical computation. It also gives a comprehensive introduction to finite elements, including their computer implementation.
Simulation of a marine nuclear reactor
International Nuclear Information System (INIS)
Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki
1995-01-01
A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship's motions because of the ship's maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship's motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship's motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship's motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship's motions on the reactor behavior can be accurately simulated by NESSY
Direct Numerical Simulation of Driven Cavity Flows
Verstappen, R.; Wissink, J.G.; Veldman, A.E.P.
Direct numerical simulations of 2D driven cavity flows have been performed. The simulations exhibit that the flow converges to a periodically oscillating state at Re=11,000, and reveal that the dynamics is chaotic at Re=22,000. The dimension of the attractor and the Kolmogorov entropy have been
Numerical Simulation of Cyclic Thermodynamic Processes
DEFF Research Database (Denmark)
Andersen, Stig Kildegård
2006-01-01
This thesis is on numerical simulation of cyclic thermodynamic processes. A modelling approach and a method for finding periodic steady state solutions are described. Examples of applications are given in the form of four research papers. Stirling machines and pulse tube coolers are introduced...... and a brief overview of the current state of the art in methods for simulating such machines is presented. It was found that different simulation approaches, which model the machines with different levels of detail, currently coexist. Methods using many simplifications can be easy to use and can provide...... models flexible and easy to modify, and to make simulations fast. A high level of accuracy was achieved for integrations of a model created using the modelling approach; the accuracy depended on the settings for the numerical solvers in a very predictable way. Selection of fast numerical algorithms...
The Simulator Development for RDE Reactor
Subekti, Muhammad; Bakhri, Syaiful; Sunaryo, Geni Rina
2018-02-01
BATAN is proposing the construction of experimental power reactor (RDE reactor) for increasing the public acceptance on NPP development plan, proofing the safety level of the most advanced reactor by performing safety demonstration on the accidents such as Chernobyl and Fukushima, and owning the generation fourth (G4) reactor technology. For owning the reactor technology, the one of research activities is RDE’s simulator development that employing standard equation. The development utilizes standard point kinetic and thermal equation. The examination of the simulator carried out comparison in which the simulation’s calculation result has good agreement with assumed parameters and ChemCAD calculation results. The transient simulation describes the characteristic of the simulator to respond the variation of power increase of 1.5%/min, 2.5%/min, and 3.5%/min.
Numerical study of optimal equilibrium cycles for pressurized water reactors
International Nuclear Information System (INIS)
Mahlers, Y.P.
2003-01-01
An algorithm based on simulated annealing and successive linear programming is applied to solve equilibrium cycle optimization problems for pressurized water reactors. In these problems, the core reload scheme is represented by discrete variables, while the cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are treated as continuous variables. The enrichments are considered to be distinct in all feed fuel assemblies. The number of batches and their sizes are not fixed and also determined by the algorithm. An important feature of the algorithm is that all the parameters are determined by the solution of one optimization problem including both discrete and continuous variables. To search for the best reload scheme, simulated annealing is used. The optimum cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are determined for each reload pattern examined using successive linear programming. Numerical results of equilibrium cycle optimization for various values of the effective price of electricity and fuel reprocessing cost are studied
Analog reactor simulator RAS; Reaktorski analogni simulator RAS
Energy Technology Data Exchange (ETDEWEB)
Radanovic, Lj; Bingulac, S; Popovic, D [The Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Yugoslavia)
1961-07-01
Analog computer RAS was designed as a nuclear reactor simulator, but it can be simultaneously used for solving a number of other problems. This paper contains a brief description of the simulator parts and their principal characteristics.
Real time simulation method for fast breeder reactors dynamics
International Nuclear Information System (INIS)
Miki, Tetsushi; Mineo, Yoshiyuki; Ogino, Takamichi; Kishida, Koji; Furuichi, Kenji.
1985-01-01
The development of multi-purpose real time simulator models with suitable plant dynamics was made; these models can be used not only in training operators but also in designing control systems, operation sequences and many other items which must be studied for the development of new type reactors. The prototype fast breeder reactor ''Monju'' is taken as an example. Analysis is made on various factors affecting the accuracy and computer load of its dynamic simulation. A method is presented which determines the optimum number of nodes in distributed systems and time steps. The oscillations due to the numerical instability are observed in the dynamic simulation of evaporators with a small number of nodes, and a method to cancel these oscillations is proposed. It has been verified through the development of plant dynamics simulation codes that these methods can provide efficient real time dynamics models of fast breeder reactors. (author)
Practical integrated simulation systems for coupled numerical simulations in parallel
Energy Technology Data Exchange (ETDEWEB)
Osamu, Hazama; Zhihong, Guo [Japan Atomic Energy Research Inst., Centre for Promotion of Computational Science and Engineering, Tokyo (Japan)
2003-07-01
In order for the numerical simulations to reflect 'real-world' phenomena and occurrences, incorporation of multidisciplinary and multi-physics simulations considering various physical models and factors are becoming essential. However, there still exist many obstacles which inhibit such numerical simulations. For example, it is still difficult in many instances to develop satisfactory software packages which allow for such coupled simulations and such simulations will require more computational resources. A precise multi-physics simulation today will require parallel processing which again makes it a complicated process. Under the international cooperative efforts between CCSE/JAERI and Fraunhofer SCAI, a German institute, a library called the MpCCI, or Mesh-based Parallel Code Coupling Interface, has been implemented together with a library called STAMPI to couple two existing codes to develop an 'integrated numerical simulation system' intended for meta-computing environments. (authors)
Fessenheim simulator for OECD Halden Reactor Project
International Nuclear Information System (INIS)
Oudot, G.; Bonnissent, B.
1998-01-01
A full scope NPP simulator is presently under manufacture by THOMSON TRAINING and SIMULATION (TTandS) in Cergy (France) for the OECD HALDEN REACTOR PROJECT. The reference plant of this simulator is the Fessenheim CP0 PWR power plant operated by the French utility EDF, for which TTandS has delivered a full scope training simulator in mid 1997. The simulator for HALDEN Reactor Project is based on a software duplication of the Fessenheim simulator delivered to EDF, ported on the most recent computers and O.S. available. This paper outlines the main features of this new simulator generation which reaps benefit of the advanced technologies of the SIPA design simulator introduced inside a full scope simulator. This kind of simulator is in fact the synthesis between training and design simulators and offers therefore added technical capabilities well suited to HALDEN needs. (author)
Numerical computation of fluid flow in different nonferrous metallurgical reactors
International Nuclear Information System (INIS)
Lackner, A.
1996-10-01
Heat, mass and fluid flow phenomena in metallurgical reactor systems such as smelting cyclones or electrolytic cells are complex and intricately linked through the governing equations of fluid flow, chemical reaction kinetics and chemical thermodynamics. The challenges for the representation of flow phenomena in such reactors as well as the transfers of these concepts to non-specialist modelers (e.g. plant operators and management personnel) can be met through scientific flow visualization techniques. In the first example the fluid flow of the gas phase and of concentrate particles in a smelting cyclone for copper production are calculated three dimensionally. The effect of design parameters (length and diameter of reactor, concentrate feeding tangentially or from the top, ..) and operating conditions are investigated. Single particle traces show, how to increase particle retention time before the particles reach the liquid film flowing down the cyclone wall. Cyclone separators are widely used in the metallurgical and chemical industry for collection of large quantities of dust. Most of the empirical models, which today are applied for the design, are lacking in being valid in the high temperature region. Therefore the numerical prediction of the collection efficiency of dust particles is done. The particle behavior close to the wall is considered by applying a particle restitution model, which calculates individual particle restitution coefficients as functions of impact velocity and impact angle. The effect of design parameters and operating are studied. Moreover, the fluid flow inside a copper refining electrolysis cell is modeled. The simulation is based on density variations in the boundary layer at the electrode surface. Density and thickness of the boundary layer are compared to measurements in a parametric study. The actual inhibitor concentration in the cell is calculated, too. Moreover, a two-phase flow approach is developed to simulate the behavior of
Numerical simulation of sand jet in water
Energy Technology Data Exchange (ETDEWEB)
Azimi, A.H.; Zhu, D.; Rajaratnam, N. [Alberta Univ., Edmonton, AB (Canada). Dept. of Civil and Environmental Engineering
2008-07-01
A numerical simulation of sand jet in water was presented. The study involved a two-phase flow using two-phase turbulent jets. A literature review was also presented, including an experiment on particle laden air jet using laser doppler velocimetry (LDV); experiments on the effect of particle size and concentration on solid-gas jets; an experimental study of solid-liquid jets using particle image velocimetry (PIV) technique where mean velocity and fluctuations were measured; and an experimental study on solid-liquid jets using the laser doppler anemometry (LDA) technique measuring both water axial and radial velocities. Other literature review results included a photographic study of sand jets in water; a comparison of many two-phase turbulent flow; and direct numerical simulation and large-eddy simulation to study the effect of particle in gas jet flow. The mathematical model and experimental setup were also included in the presentation along with simulation results for sand jets, concentration, and kinetic energy. The presentation concluded with some proposed future studies including numerical simulation of slurry jets in water and numerical simulation of slurry jets in MFT. tabs., figs.
Numerical simulation of "an American haboob"
Vukovic, A.; Vujadinovic, M.; Pejanovic, G.; Andric, J.; Kumjian, M. R.; Djurdjevic, V.; Dacic, M.; Prasad, A. K.; El-Askary, H. M.; Paris, B. C.; Petkovic, S.; Nickovic, S.; Sprigg, W. A.
2014-01-01
A dust storm of fearful proportions hit Phoenix in the early evening hours of 5 July 2011. This storm, an American haboob, was predicted hours in advance because numerical, land–atmosphere modeling, computing power and remote sensing of dust events have improved greatly over the past decade. High-resolution numerical models are required for accurate simulation of the small scales of the haboob process, with high velocity surface winds produced by strong convection and severe...
Energy Technology Data Exchange (ETDEWEB)
Carasik, Lane B., E-mail: lcarasik@tamu.edu [Texas A& M University, Department of Nuclear Engineering, 3133 TAMU, College Station, TX 77843-3133 (United States); Sebilleau, Frédéric, E-mail: Frederic.sebilleau11@imperial.ac.uk [Imperial College London, Mechanical Engineering Department, London SW7 SBX (United Kingdom); Walker, Simon P., E-mail: s.p.walker@imperial.ac.uk [Imperial College London, Mechanical Engineering Department, London SW7 SBX (United Kingdom); Hassan, Yassin A., E-mail: y-hassan@tamu.edu [Texas A& M University, Department of Nuclear Engineering, 3133 TAMU, College Station, TX 77843-3133 (United States)
2017-02-15
Highlights: • Simulations of thermal stratification in large enclosures using different turbulence models. • The recent elliptic blending k–ε was implemented in this work. • Direct comparisons of experimental temperature measurements to CFD predictions. • Spurious prediction of jet stabilisation and diffuse stratification by both low-Re k–ε and SST k–ω. - Abstract: An ability to predict the behavior of buoyant jets entering a large body of relatively stationary fluid is important in analysis of a wide variety of nuclear accidents, including for example the use of large tanks of water as heat sinks, or the release of hot gases into the secondary containment. In particular, the degree to which temperature stratification occurs is important, as it can affect markedly the effectiveness of the body of fluid as a heat sink. In this paper, we report the results of measurements on an experimental facility designed to exhibit such behavior, and the results of attempts to predict this experiment using CFD. In particular, we here investigate the effectiveness of three alternative turbulence models for this analysis; low-Re k–e, elliptic-blended k–e and Shear Stress Transport k–ω models. Both the degree of thermal stratification and the stability of the jet that were predicted differed markedly between the three models. Two of the models, the low-Re k–e and the Shear Stress Transport k–ω, tend to predict, wrongly, significant turbulent intensity in regions where fluid velocities are essentially zero. This spurious high turbulent intensity in turn causes (i) a high turbulent viscosity to be applied, wrongly stabilizing the jet, and (ii) increased turbulent diffusion of heat, causing too deep and diffuse a stratification to be predicted.
REACTOR: a computer simulation for schools
International Nuclear Information System (INIS)
Squires, D.
1985-01-01
The paper concerns computer simulation of the operation of a nuclear reactor, for use in schools. The project was commissioned by UKAEA, and carried out by the Computers in the Curriculum Project, Chelsea College. The program, for an advanced gas cooled reactor, is briefly described. (U.K.)
Nodal methods in numerical reactor calculations
International Nuclear Information System (INIS)
Hennart, J.P.; Valle, E. del
2004-01-01
The present work describes the antecedents, developments and applications started in 1972 with Prof. Hennart who was invited to be part of the staff of the Nuclear Engineering Department at the School of Physics and Mathematics of the National Polytechnic Institute. Since that time and up to 1981, several master theses based on classical finite element methods were developed with applications in point kinetics and in the steady state as well as the time dependent multigroup diffusion equations. After this period the emphasis moved to nodal finite elements in 1, 2 and 3D cartesian geometries. All the thesis were devoted to the numerical solution of the neutron multigroup diffusion and transport equations, few of them including the time dependence, most of them related with steady state diffusion equations. The main contributions were as follows: high order nodal schemes for the primal and mixed forms of the diffusion equations, block-centered finite-differences methods, post-processing, composite nodal finite elements for hexagons, and weakly and strongly discontinuous schemes for the transport equation. Some of these are now being used by several researchers involved in nuclear fuel management. (Author)
Nodal methods in numerical reactor calculations
Energy Technology Data Exchange (ETDEWEB)
Hennart, J P [UNAM, IIMAS, A.P. 20-726, 01000 Mexico D.F. (Mexico); Valle, E del [National Polytechnic Institute, School of Physics and Mathematics, Department of Nuclear Engineering, Mexico, D.F. (Mexico)
2004-07-01
The present work describes the antecedents, developments and applications started in 1972 with Prof. Hennart who was invited to be part of the staff of the Nuclear Engineering Department at the School of Physics and Mathematics of the National Polytechnic Institute. Since that time and up to 1981, several master theses based on classical finite element methods were developed with applications in point kinetics and in the steady state as well as the time dependent multigroup diffusion equations. After this period the emphasis moved to nodal finite elements in 1, 2 and 3D cartesian geometries. All the thesis were devoted to the numerical solution of the neutron multigroup diffusion and transport equations, few of them including the time dependence, most of them related with steady state diffusion equations. The main contributions were as follows: high order nodal schemes for the primal and mixed forms of the diffusion equations, block-centered finite-differences methods, post-processing, composite nodal finite elements for hexagons, and weakly and strongly discontinuous schemes for the transport equation. Some of these are now being used by several researchers involved in nuclear fuel management. (Author)
Numerical simulation of radial compressor stage
Syka, T.; Luňáček, O.
2013-04-01
Article describes numerical simulations of air flow in radial compressor stage in NUMECA CFD software. In simulations geometry variants with and without seals are used. During tasks evaluating was observed seals influence on flow field and performance parameters of compressor stage. Also is described CFDresults comparison with results from design software based on experimental measurements and monitoring of influence of seals construction on compressor stage efficiency.
Numerical simulation of radial compressor stage
Luňáček O.; Syka T.
2013-01-01
Article describes numerical simulations of air flow in radial compressor stage in NUMECA CFD software. In simulations geometry variants with and without seals are used. During tasks evaluating was observed seals influence on flow field and performance parameters of compressor stage. Also is described CFDresults comparison with results from design software based on experimental measurements and monitoring of influence of seals construction on compressor stage efficiency.
Numerical simulation of radial compressor stage
Directory of Open Access Journals (Sweden)
Luňáček O.
2013-04-01
Full Text Available Article describes numerical simulations of air flow in radial compressor stage in NUMECA CFD software. In simulations geometry variants with and without seals are used. During tasks evaluating was observed seals influence on flow field and performance parameters of compressor stage. Also is described CFDresults comparison with results from design software based on experimental measurements and monitoring of influence of seals construction on compressor stage efficiency.
Numerical Simulation of Steady Supercavitating Flows
Ali Jafarian; Ahmad-Reza Pishevar
2016-01-01
In this research, the Supercavitation phenomenon in compressible liquid flows is simulated. The one-fluid method based on a new exact two-phase Riemann solver is used for modeling. The cavitation is considered as an isothermal process and a consistent equation of state with the physical behavior of the water is used. High speed flow of water over a cylinder and a projectile are simulated and the results are compared with the previous numerical and experimental results. The cavitation bubble p...
Numerical Simulation Of Silicon-Ribbon Growth
Woda, Ben K.; Kuo, Chin-Po; Utku, Senol; Ray, Sujit Kumar
1987-01-01
Mathematical model includes nonlinear effects. In development simulates growth of silicon ribbon from melt. Takes account of entire temperature and stress history of ribbon. Numerical simulations performed with new model helps in search for temperature distribution, pulling speed, and other conditions favoring growth of wide, flat, relatively defect-free silicon ribbons for solar photovoltaic cells at economically attractive, high production rates. Also applicable to materials other than silicon.
Numerical Simulation on Natural Convection Cooling of a FM Target
Energy Technology Data Exchange (ETDEWEB)
Park, Jong Pil; Park, Su Ki [KAERI, Daejeon (Korea, Republic of)
2016-05-15
The irradiated FM(Fission-Molly) target is unloaded from the irradiation hole during normal operation, and then cooled down in the reactor pool for a certain period of time. Therefore, it is necessary to identify the minimum decay time needed to cool down FM target sufficiently by natural convection. In the present work, numerical simulations are performed to predict cooling capability of a FM target cooled by natural convection using commercial computational fluid dynamics (CFD) code, CFX. The present study is carried out using CFD code to investigate cooling capability of a FM target cooled by natural convection. The steady state simulation as well as transient simulation is performed in the present work. Based on the transient simulation (T1), the minimum decay time that the maximum fuel temperature does not reach the design limit temperature (TONB-3 .deg. C) is around 15.60 seconds.
Spectral Methods in Numerical Plasma Simulation
DEFF Research Database (Denmark)
Coutsias, E.A.; Hansen, F.R.; Huld, T.
1989-01-01
An introduction is given to the use of spectral methods in numerical plasma simulation. As examples of the use of spectral methods, solutions to the two-dimensional Euler equations in both a simple, doubly periodic region, and on an annulus will be shown. In the first case, the solution is expanded...
Simple Numerical Simulation of Strain Measurement
Tai, H.
2002-01-01
By adopting the basic principle of the reflection (and transmission) of a plane polarized electromagnetic wave incident normal to a stack of films of alternating refractive index, a simple numerical code was written to simulate the maximum reflectivity (transmittivity) of a fiber optic Bragg grating corresponding to various non-uniform strain conditions including photo-elastic effect in certain cases.
Study on the numerical analysis of nuclear reactor kinetics equations
International Nuclear Information System (INIS)
Yang, J.C.
1980-01-01
A two-step alternating direction explict method is proposed for the solution of the space-and time-dependent diffusion theory reactor kinetics equations in two space dimensions as a special case of the general class of alternating direction implicit method and the truncation error of this method is estimated. To test the validity of this method it is applied to the Pressurized Water Reactor and CANDU-PHW reactor which have been operating and underconstructing in Korea. The time dependent neutron flux of the PWR reactor during control rod insertion and time dependent neutronic power of CANDU-PHW reactor in the case of postulated loss of coolant accident are obtained from the numerical calculation results. The results of the PWR reactor problem are shown the close agreement between implicit-difference method used in the TWIGL program and this method, and the results of the CANDU-PHW reactor are compared with the results of improved quasistic method and modal method. (Author)
Numerical simulation of large deformation polycrystalline plasticity
International Nuclear Information System (INIS)
Inal, K.; Neale, K.W.; Wu, P.D.; MacEwen, S.R.
2000-01-01
A finite element model based on crystal plasticity has been developed to simulate the stress-strain response of sheet metal specimens in uniaxial tension. Each material point in the sheet is considered to be a polycrystalline aggregate of FCC grains. The Taylor theory of crystal plasticity is assumed. The numerical analysis incorporates parallel computing features enabling simulations of realistic models with large number of grains. Simulations have been carried out for the AA3004-H19 aluminium alloy and the results are compared with experimental data. (author)
Fluid dynamics theory, computation, and numerical simulation
Pozrikidis, C
2001-01-01
Fluid Dynamics Theory, Computation, and Numerical Simulation is the only available book that extends the classical field of fluid dynamics into the realm of scientific computing in a way that is both comprehensive and accessible to the beginner The theory of fluid dynamics, and the implementation of solution procedures into numerical algorithms, are discussed hand-in-hand and with reference to computer programming This book is an accessible introduction to theoretical and computational fluid dynamics (CFD), written from a modern perspective that unifies theory and numerical practice There are several additions and subject expansions in the Second Edition of Fluid Dynamics, including new Matlab and FORTRAN codes Two distinguishing features of the discourse are solution procedures and algorithms are developed immediately after problem formulations are presented, and numerical methods are introduced on a need-to-know basis and in increasing order of difficulty Matlab codes are presented and discussed for a broad...
Fluid Dynamics Theory, Computation, and Numerical Simulation
Pozrikidis, Constantine
2009-01-01
Fluid Dynamics: Theory, Computation, and Numerical Simulation is the only available book that extends the classical field of fluid dynamics into the realm of scientific computing in a way that is both comprehensive and accessible to the beginner. The theory of fluid dynamics, and the implementation of solution procedures into numerical algorithms, are discussed hand-in-hand and with reference to computer programming. This book is an accessible introduction to theoretical and computational fluid dynamics (CFD), written from a modern perspective that unifies theory and numerical practice. There are several additions and subject expansions in the Second Edition of Fluid Dynamics, including new Matlab and FORTRAN codes. Two distinguishing features of the discourse are: solution procedures and algorithms are developed immediately after problem formulations are presented, and numerical methods are introduced on a need-to-know basis and in increasing order of difficulty. Matlab codes are presented and discussed for ...
International Nuclear Information System (INIS)
Okano, Yasushi; Yamano, Hidemasa
2015-01-01
As a part of a development of the risk assessment methodologies against external hazards, a new methodology to assess forest fire hazards is being developed. Frequency and consequence of the forest fire are analyzed to obtain the hazard intensity curve and then Level 1 probabilistic safety assessment is performed to obtain the conditional core damage probability due to the challenges by the forest fire. 'Heat', 'flame', 'smoke' and 'flying object' are the challenges to a nuclear power plant. For a sodium-cooled fast reactor, a decay heat removal under accident conditions is operated with an ultimate heat sink of air, then, the challenge by 'smoke' will potentially be on the air filter of the system. In this paper, numerical simulations of forest fire propagation and smoke transport were performed with sensibility studies to weather conditions, and the effect by the smoke on the air filter was quantitatively evaluated. Forest fire propagation simulations were performed using FARSITE code. A temporal increase of a forest fire spread area and a position of the frontal fireline are obtained by the simulation, and 'reaction intensity' and 'frontal fireline intensity' as the indexes of 'heat' are obtained as well. The boundary of the fire spread area is shaped like an ellipse on the terrain, and the boundary length is increased with time and fire spread. The sensibility analyses on weather conditions of wind, temperature, and humidity were performed, and it was summarized that 'forest fire spread rate' and 'frontal fireline intensity' depend much on wind speed and humidity. Smoke transport simulations were performed by ALOFT-FT code where three-dimensional spatial distribution of smoke density, especially of particle matters of PM2.5 and PM10, are evaluated. The snapshot outputs, namely 'reaction intensity' and 'position of frontal fireline', from the sensibility studies of the FARSITE were directly utilized as the input data for ALOFT-FT, whereas it is assumed that the
Virtual environments simulation in research reactor
Muhamad, Shalina Bt. Sheik; Bahrin, Muhammad Hannan Bin
2017-01-01
Virtual reality based simulations are interactive and engaging. It has the useful potential in improving safety training. Virtual reality technology can be used to train workers who are unfamiliar with the physical layout of an area. In this study, a simulation program based on the virtual environment at research reactor was developed. The platform used for virtual simulation is 3DVia software for which it's rendering capabilities, physics for movement and collision and interactive navigation features have been taken advantage of. A real research reactor was virtually modelled and simulated with the model of avatars adopted to simulate walking. Collision detection algorithms were developed for various parts of the 3D building and avatars to restrain the avatars to certain regions of the virtual environment. A user can control the avatar to move around inside the virtual environment. Thus, this work can assist in the training of personnel, as in evaluating the radiological safety of the research reactor facility.
Experiments and Numerical Simulations of Electrodynamic Tether
Iki, Kentaro; Kawamoto, Satomi; Takahashi, Ayaka; Ishimoto, Tomori; Yanagida, Atsushi; Toda, Susumu
As an effective means of suppressing space debris growth, the Aerospace Research and Development Directorate of the Japan Aerospace Exploration Agency (JAXA) has been investigating an active space debris removal system that employs highly efficient electrodynamic tether (EDT) technology for orbital transfer. This study investigates tether deployment dynamics by means of on-ground experiments and numerical simulations of an electrodynamic tether system. Some key parameters used in the numerical simulations, such as the elastic modulus and damping ratio of the tether, the spring constant of the coiling of the tether, and deployment friction, must be estimated, and various experiments are conducted to determine these values. As a result, the following values were obtained: The elastic modulus of the tether was 40 GPa, and the damping ratio of the tether was 0.02. The spring constant and the damping ratio of the tether coiling were 10-4 N/m and 0.025 respectively. The deployment friction was 0.038ν + 0.005 N. In numerical simulations using a multiple mass tether model, tethers with lengths of several kilometers are deployed and the attitude dynamics of satellites attached to the end of the tether and tether libration are calculated. As a result, the simulations confirmed successful deployment of the tether with a length of 500 m using the electrodynamic tether system.
Numerical Simulation of a Tornado Generating Supercell
Proctor, Fred H.; Ahmad, Nashat N.; LimonDuparcmeur, Fanny M.
2012-01-01
The development of tornadoes from a tornado generating supercell is investigated with a large eddy simulation weather model. Numerical simulations are initialized with a sounding representing the environment of a tornado producing supercell that affected North Carolina and Virginia during the Spring of 2011. The structure of the simulated storm was very similar to that of a classic supercell, and compared favorably to the storm that affected the vicinity of Raleigh, North Carolina. The presence of mid-level moisture was found to be important in determining whether a supercell would generate tornadoes. The simulations generated multiple tornadoes, including cyclonic-anticyclonic pairs. The structure and the evolution of these tornadoes are examined during their lifecycle.
Contributions to reinforced concrete structures numerical simulations
International Nuclear Information System (INIS)
Badel, P.B.
2001-07-01
In order to be able to carry out simulations of reinforced concrete structures, it is necessary to know two aspects: the behaviour laws have to reflect the complex behaviour of concrete and a numerical environment has to be developed in order to avoid to the user difficulties due to the softening nature of the behaviour. This work deals with these two subjects. After an accurate estimation of two behaviour models (micro-plan and mesoscopic models), two damage models (the first one using a scalar variable, the other one a tensorial damage of the 2 order) are proposed. These two models belong to the framework of generalized standard materials, which renders their numerical integration easy and efficient. A method of load control is developed in order to make easier the convergence of the calculations. At last, simulations of industrial structures illustrate the efficiency of the method. (O.M.)
Formulae for thermal feedback of group constants in digital reactor simulation
International Nuclear Information System (INIS)
Perneczky, L.; Toth, I.; Vigassy, J.
1976-01-01
The problem, how the feedback of the thermohydraulic field to the neutron density in a reactor can be calculated is analysed. After a brief survey of the digital models in reactor simulation the applied model based on the time-dependent two-group diffusion equations is described. Using the reactor physical code system THERESA numerical results for the VVER-440 reactor are presented. (Sz.Z.)
Hamiltonian circuited simulations in reactor physics
International Nuclear Information System (INIS)
Rio Hirowati Shariffudin
2002-01-01
In the assessment of suitability of reactor designs and in the investigations into reactor safety, the steady state of a nuclear reactor has to be studied carefully. The analysis can be done through mockup designs but this approach costs a lot of money and consumes a lot of time. A less expensive approach is via simulations where the reactor and its neutron interactions are modelled mathematically. Finite difference discretization of the diffusion operator has been used to approximate the steady state multigroup neutron diffusion equations. The steps include the outer scheme which estimates the resulting right hand side of the matrix equation, the group scheme which calculates the upscatter problem and the inner scheme which solves for the flux for a particular group. The Hamiltonian circuited simulations for the inner iterations of the said neutron diffusion equation enable the effective use of parallel computing, especially where the solutions of multigroup neutron diffusion equations involving two or more space dimensions are required. (Author)
Numerical simulation of electrostatic waves in plasmas
International Nuclear Information System (INIS)
Erz, U.
1981-08-01
In this paper the propagation of electrostatic waves in plasmas and the non-linear interactions, which occur in the case of large wave amplitudes, are studied using a new numerical method for plasma simulation. This mathematical description is based on the Vlasov-model. Changes in the distribution-function are taken into account and thus plasma kinetic effects can be treated. (orig./HT) [de
Numerical simulations on ion acoustic double layers
International Nuclear Information System (INIS)
Sato, T.; Okuda, H.
1980-07-01
A comprehensive numerical study of ion acoustic double layers has been performed for both periodic as well as for nonperiodic systems by means of one-dimensional particle simulations. For a nonperiodic system, an external battery and a resistance are used to model the magnetospheric convection potential and the ionospheric Pedersen resistance. It is found that the number of double layers and the associated potential buildup across the system increases with the system length
Numerical Simulations of Hyperfine Transitions of Antihydrogen
Kolbinger, B.; Diermaier, M.; Lehner, S.; Malbrunot, C.; Massiczek, O.; Sauerzopf, C.; Simon, M.C.; Widmann, E.
2015-02-04
One of the ASACUSA (Atomic Spectroscopy And Collisions Using Slow Antiprotons) collaboration's goals is the measurement of the ground state hyperfine transition frequency in antihydrogen, the antimatter counterpart of one of the best known systems in physics. This high precision experiment yields a sensitive test of the fundamental symmetry of CPT. Numerical simulations of hyperfine transitions of antihydrogen atoms have been performed providing information on the required antihydrogen events and the achievable precision.
Numerical simulations of hyperfine transitions of antihydrogen
Energy Technology Data Exchange (ETDEWEB)
Kolbinger, B., E-mail: bernadette.kolbinger@oeaw.ac.at; Capon, A.; Diermaier, M.; Lehner, S. [Stefan Meyer Institute for Subatomic Physics, Austrian Academy of Sciences (Austria); Malbrunot, C. [CERN (Switzerland); Massiczek, O.; Sauerzopf, C.; Simon, M. C.; Widmann, E. [Stefan Meyer Institute for Subatomic Physics, Austrian Academy of Sciences (Austria)
2015-08-15
One of the ASACUSA (Atomic Spectroscopy And Collisions Using Slow Antiprotons) collaboration’s goals is the measurement of the ground state hyperfine transition frequency in antihydrogen, the antimatter counterpart of one of the best known systems in physics. This high precision experiment yields a sensitive test of the fundamental symmetry of CPT. Numerical simulations of hyperfine transitions of antihydrogen atoms have been performed providing information on the required antihydrogen events and the achievable precision.
Application of finite element numerical technique to nuclear reactor geometries
Energy Technology Data Exchange (ETDEWEB)
Rouai, N M [Nuclear engineering department faculty of engineering Al-fateh universty, Tripoli (Libyan Arab Jamahiriya)
1995-10-01
Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs.
Application of finite element numerical technique to nuclear reactor geometries
International Nuclear Information System (INIS)
Rouai, N. M.
1995-01-01
Determination of the temperature distribution in nuclear elements is of utmost importance to ensure that the temperature stays within safe limits during reactor operation. This paper discusses the use of Finite element numerical technique (FE) for the solution of the two dimensional heat conduction equation in geometries related to nuclear reactor cores. The FE solution stats with variational calculus which considers transforming the heat conduction equation into an integral equation I(O) and seeks a function that minimizes this integral and hence gives the solution to the heat conduction equation. In this paper FE theory as applied to heat conduction is briefly outlined and a 2-D program is used to apply the theory to simple shapes and to two gas cooled reactor fuel elements. Good results are obtained for both cases with reasonable number of elements. 7 figs
PWR control rod ejection analysis with the numerical nuclear reactor
International Nuclear Information System (INIS)
Hursin, M.; Kochunas, B.; Downar, T. J.
2008-01-01
During the past several years, a comprehensive high fidelity reactor LWR core modeling capability has been developed and is referred to as the Numerical Nuclear Reactor (NNR). The NNR achieves high fidelity by integrating whole core neutron transport solution and ultra fine mesh computational fluid dynamics/heat transfer solution. The work described in this paper is a preliminary demonstration of the ability of NNR to provide a detailed intra pin power distribution during a control rod ejection accident. The motivation of the work is to quantify the impact on the fuel performance calculation of a more physically accurate representation of the power distribution within the fuel rod during the transient. The paper addresses first, the validation of the transient capability of the neutronic module of the NNR code system, DeCART. For this purpose, a 'mini core' problem consisting of a 3x3 array of typical PWR fuel assemblies is considered. The initial state of the 'mini core' is hot zero power with a control rod partially inserted into the central assembly which is fresh fuel and is adjacent to once and twice burned fuel representative of a realistic PWR arrangement. The thermal hydraulic feedbacks are provided by a simplified fluids and heat conduction solver consistent for both PARCS and DeCART. The control rod is ejected from the central assembly and the transient calculation is performed with DeCART and compared with the results of the U.S. NRC core simulation code PARCS. Because the pin power reconstruction in PARCS is based on steady state intra assembly pin power distributions which do not account for thermal feedback during the transient and which do not take into account neutron leakage from neighboring assemblies during the transient, there are some small differences in the PARCS and DeCART pin power prediction. Intra pin power density information obtained with DeCART represents new information not available with previous generation of methods. The paper then
Integral Pressurized Water Reactor Simulator Manual
International Nuclear Information System (INIS)
2017-01-01
This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.
Programming for a nuclear reactor instrument simulation
International Nuclear Information System (INIS)
Cohn, C.
1988-01-01
This note discusses 8086/8087 machine-language programming for simulation of nuclear reactor instrument current inputs by means of a digital-analog converter (DAC) feeding a bank of series input resistors. It also shows FORTRAN programming for generating the parameter tales used in the simulation. These techniques would be generally useful for high-speed simulation of quantities varying over many orders of magnitude
Simulation of decreasing reactor power level with BWR simulator
International Nuclear Information System (INIS)
Suwoto; Zuhair; Rivai, Abu Khalid
2002-01-01
Study on characteristic of BWR using Desktop PC Based Simulator Program was analysed. This simulator is more efficient and cheaper for analyzing of characteristic and dynamic respond than full scope simulator for decreasing power level of BW. Dynamic responses of BWR reactor was investigated during the power level reduction from 100% FP (Full Power) which is 3926 MWth to 0% FP with 25% steps and 1 % FP/sec rate. The overall results for core flow rate, reactor steam flow, feed-water flow and turbine-generator power show tendency proportional to reduction of reactor power. This results show that reactor power control in BWR could be done by control of re-circulation flow that alter the density of water used as coolant and moderator. Decreasing the re-circulation flow rate will decrease void density which has negative reactivity and also affect the position of control rods
NUMERICAL MODEL APPLICATION IN ROWING SIMULATOR DESIGN
Directory of Open Access Journals (Sweden)
Petr Chmátal
2016-04-01
Full Text Available The aim of the research was to carry out a hydraulic design of rowing/sculling and paddling simulator. Nowadays there are two main approaches in the simulator design. The first one includes a static water with no artificial movement and counts on specially cut oars to provide the same resistance in the water. The second approach, on the other hand uses pumps or similar devices to force the water to circulate but both of the designs share many problems. Such problems are affecting already built facilities and can be summarized as unrealistic feeling, unwanted turbulent flow and bad velocity profile. Therefore, the goal was to design a new rowing simulator that would provide nature-like conditions for the racers and provide an unmatched experience. In order to accomplish this challenge, it was decided to use in-depth numerical modeling to solve the hydraulic problems. The general measures for the design were taken in accordance with space availability of the simulator ́s housing. The entire research was coordinated with other stages of the construction using BIM. The detailed geometry was designed using a numerical model in Ansys Fluent and parametric auto-optimization tools which led to minimum negative hydraulic phenomena and decreased investment and operational costs due to the decreased hydraulic losses in the system.
Kinetic calculations for miniature neutron source reactor using analytical and numerical techniques
International Nuclear Information System (INIS)
Ampomah-Amoako, E.
2008-06-01
The analytical methods, step change in reactivity and ramp change in reactivity as well as numerical methods, fixed point iteration and Runge Kutta-gill were used to simulate the initial build up of neutrons in a miniature neutron source reactor with and without temperature feedback effect. The methods were modified to include photo neutron concentration. PARET 7.3 was used to simulate the transients behaviour of Ghana Research Reactor-1. The PARET code was capable of simulating the transients for 2.1 mk and 4 mk insertions of reactivity with peak powers of 49.87 kW and 92.34 kW, respectively. PARET code however failed to simulate 6.71 mk of reactivity which was predicted by Akaho et al through TEMPFED. (au)
Mathematical models and numerical simulation in electromagnetism
Bermúdez, Alfredo; Salgado, Pilar
2014-01-01
The book represents a basic support for a master course in electromagnetism oriented to numerical simulation. The main goal of the book is that the reader knows the boundary-value problems of partial differential equations that should be solved in order to perform computer simulation of electromagnetic processes. Moreover it includes a part devoted to electric circuit theory based on ordinary differential equations. The book is mainly oriented to electric engineering applications, going from the general to the specific, namely, from the full Maxwell’s equations to the particular cases of electrostatics, direct current, magnetostatics and eddy currents models. Apart from standard exercises related to analytical calculus, the book includes some others oriented to real-life applications solved with MaxFEM free simulation software.
Three-dimensional two-fluid numerical treatment of a reactor vessel in TRAC
International Nuclear Information System (INIS)
Liles, D.R.
1979-01-01
A three-dimensional two-fluid finite difference model has been used in TRAC (Transient Reactor Analysis Code) to represent a pressurized water reactor vessel. Mesh cells may be blocked off completely to represent large flow obstructions such as downcomer walls. The hydrodynamic volumes and flow areas may also be reduced in order to provide a porous matrix simulation of smaller scale strucuture. The finite difference equations are semi-implicit so that stability time scales are associated with material movement and not wave propagation. The block matrix structure is reduced during the implicit pass to a single element seven stripe system which is easily solved iteratively. This procedure has successfully performed numerous simulations of both full sized reactor accidents and smaller scale experments. It has proven to be a useful feature of the TRAC effort
Numerical analysis of hydrodynamics in a rotor-stator reactor for biodiesel synthesis
Energy Technology Data Exchange (ETDEWEB)
Wen, Zhuqing; Petera, Jerzy [Faculty of Process and Environmental Engineering, Lodz University of Technology, Lodz (Poland)
2016-06-08
A rotor-stator spinning disk reactor for intensified biodiesel synthesis is described and numerically simulated. The reactor consists of two flat disks, located coaxially and parallel to each other with a gap ranging from 0.1 mm to 0.2 mm between the disks. The upper disk is located on a rotating shaft while the lower disk is stationary. The feed liquids, triglycerides (TG) and methanol are introduced coaxially along the center line of rotating disk and stationary disk, respectively. Fluid hydrodynamics in the reactor for synthesis of biodiesel from TG and methanol in the presence of a sodium hydroxide catalyst are simulated, using convection-diffusion-reaction species transport model by the CFD software ANSYS©Fluent v. 13.0. The effects of upper disk’s spinning speed, gap size and flow rates at inlets are evaluated.
Numerical Simulation of a Seaway with Breaking
Dommermuth, Douglas; O'Shea, Thomas; Brucker, Kyle; Wyatt, Donald
2012-11-01
The focus of this presentation is to describe the recent efforts to simulate a fully non-linear seaway with breaking by using a high-order spectral (HOS) solution of the free-surface boundary value problem to drive a three-dimensional Volume of Fluid (VOF) solution. Historically, the two main types of simulations to simulate free-surface flows are the boundary integral equations method (BIEM) and high-order spectral (HOS) methods. BIEM calculations fail at the point at which the surface impacts upon itself, if not sooner, and HOS methods can only simulate a single valued free-surface. Both also employ a single-phase approximation in which the effects of the air on the water are neglected. Due to these limitations they are unable to simulate breaking waves and air entrainment. The Volume of Fluid (VOF) method on the other hand is suitable for modeling breaking waves and air entrainment. However it is computationally intractable to generate a realistic non-linear sea-state. Here, we use the HOS solution to quickly drive, or nudge, the VOF solution into a non-linear state. The computational strategies, mathematical formulation, and numerical implementation will be discussed. The results of the VOF simulation of a seaway with breaking will also be presented, and compared to the single phase, single valued HOS results.
Numerical simulation methods of fires in nuclear power plants
International Nuclear Information System (INIS)
Keski-Rahkonen, O.; Bjoerkman, J.; Heikkilae, L.
1992-01-01
Fire is a significant hazard to the safety of nuclear power plants (NPP). Fire may be serious accident as such, but even small fire at a critical point in a NPP may cause an accident much more serious than fire itself. According to risk assessments a fire may be an initial cause or a contributing factor in a large part of reactor accidents. At the Fire Technology and the the Nuclear Engineering Laboratory of the Technical Research Centre of Finland (VTT) fire safety research for NPPs has been carried out in a large extent since 1985. During years 1988-92 a project Advanced Numerical Modelling in Nuclear Power Plants (PALOME) was carried out. In the project the level of numerical modelling for fire research in Finland was improved by acquiring, preparing for use and developing numerical fire simulation programs. Large scale test data of the German experimental program (PHDR Sicherheitsprogramm in Kernforschungscentral Karlsruhe) has been as reference. The large scale tests were simulated by numerical codes and results were compared to calculations carried out by others. Scientific interaction with outstanding foreign laboratories and scientists has been an important part of the project. This report describes the work of PALOME-project carried out at the Fire Technology Laboratory only. A report on the work at the Nuclear Engineering Laboratory will be published separatively. (au)
Physical models and numerical methods of the reactor dynamic computer program RETRAN
International Nuclear Information System (INIS)
Kamelander, G.; Woloch, F.; Sdouz, G.; Koinig, H.
1984-03-01
This report describes the physical models and the numerical methods of the reactor dynamic code RETRAN simulating reactivity transients in Light-Water-Reactors. The neutron-physical part of RETRAN bases on the two-group-diffusion equations which are solved by discretization similar to the TWIGL-method. An exponential transformation is applied and the inner iterations are accelerated by a coarse-mesh-rebalancing procedure. The thermo-hydraulic model approximates the equation of state by a built-in steam-water-table and disposes of options for the calculation of heat-conduction coefficients and heat transfer coefficients. (Author) [de
Simulation for nuclear reactor technology
International Nuclear Information System (INIS)
Walton, D.G.
1985-01-01
Mathematical modelling forms the cornerstone of both the physical sciences and the engineering sciences. Computer modelling represents an extension of mathematical modelling into areas which are either not sufficiently accurate or not conveniently modelled by analytical techniques and where highly complex models can be investigated on a sufficiently short timescale. Simulation represents the application of these modelling techniques to real systems, thus enabling information on plant characteristics to be gained without either constructing or operating the full-scale plant or system under consideration. Developments in computer systems modelling techniques, safety and operating philosophy, and human psychology studies lead to improvements in simulation practice. New concepts in simulation also lead to a demand for new types of computers, dedicated computer systems are increasingly being developed for plant analysis, and the advent and development of microprocessors has also led to new techniques both in simulation and simulators. Progress in these areas is presented in this book
Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results
International Nuclear Information System (INIS)
Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young
2007-12-01
A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor
Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results
Energy Technology Data Exchange (ETDEWEB)
Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young
2007-12-15
A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.
The Consortium for Advanced Simulation of Light Water Reactors
International Nuclear Information System (INIS)
Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul
2011-01-01
The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.
Nuclear Power Reactor simulator - based training program
International Nuclear Information System (INIS)
Abdelwahab, S.A.S.
2009-01-01
nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude
Numerical simulation of real-world flows
Energy Technology Data Exchange (ETDEWEB)
Hayase, Toshiyuki, E-mail: hayase@ifs.tohoku.ac.jp [Institute of Fluid Science, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, 980-8577 (Japan)
2015-10-15
Obtaining real flow information is important in various fields, but is a difficult issue because measurement data are usually limited in time and space, and computational results usually do not represent the exact state of real flows. Problems inherent in the realization of numerical simulation of real-world flows include the difficulty in representing exact initial and boundary conditions and the difficulty in representing unstable flow characteristics. This article reviews studies dealing with these problems. First, an overview of basic flow measurement methodologies and measurement data interpolation/approximation techniques is presented. Then, studies on methods of integrating numerical simulation and measurement, namely, four-dimensional variational data assimilation (4D-Var), Kalman filters (KFs), state observers, etc are discussed. The first problem is properly solved by these integration methodologies. The second problem can be partially solved with 4D-Var in which only initial and boundary conditions are control parameters. If an appropriate control parameter capable of modifying the dynamical structure of the model is included in the formulation of 4D-Var, unstable modes are properly suppressed and the second problem is solved. The state observer and KFs also solve the second problem by modifying mathematical models to stabilize the unstable modes of the original dynamical system by applying feedback signals. These integration methodologies are now applied in simulation of real-world flows in a wide variety of research fields. Examples are presented for basic fluid dynamics and applications in meteorology, aerospace, medicine, etc. (topical review)
Numerical model simulation of atmospheric coolant plumes
International Nuclear Information System (INIS)
Gaillard, P.
1980-01-01
The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr
Lagrangian numerical methods for ocean biogeochemical simulations
Paparella, Francesco; Popolizio, Marina
2018-05-01
We propose two closely-related Lagrangian numerical methods for the simulation of physical processes involving advection, reaction and diffusion. The methods are intended to be used in settings where the flow is nearly incompressible and the Péclet numbers are so high that resolving all the scales of motion is unfeasible. This is commonplace in ocean flows. Our methods consist in augmenting the method of characteristics, which is suitable for advection-reaction problems, with couplings among nearby particles, producing fluxes that mimic diffusion, or unresolved small-scale transport. The methods conserve mass, obey the maximum principle, and allow to tune the strength of the diffusive terms down to zero, while avoiding unwanted numerical dissipation effects.
Efficient Numerical Simulation of Aerothermoelastic Hypersonic Vehicles
Klock, Ryan J.
Hypersonic vehicles operate in a high-energy flight environment characterized by high dynamic pressures, high thermal loads, and non-equilibrium flow dynamics. This environment induces strong fluid, thermal, and structural dynamics interactions that are unique to this flight regime. If these vehicles are to be effectively designed and controlled, then a robust and intuitive understanding of each of these disciplines must be developed not only in isolation, but also when coupled. Limitations on scaling and the availability of adequate test facilities mean that physical investigation is infeasible. Ever growing computational power offers the ability to perform elaborate numerical simulations, but also has its own limitations. The state of the art in numerical simulation is either to create ever more high-fidelity physics models that do not couple well and require too much processing power to consider more than a few seconds of flight, or to use low-fidelity analytical models that can be tightly coupled and processed quickly, but do not represent realistic systems due to their simplifying assumptions. Reduced-order models offer a middle ground by distilling the dominant trends of high-fidelity training solutions into a form that can be quickly processed and more tightly coupled. This thesis presents a variably coupled, variable-fidelity, aerothermoelastic framework for the simulation and analysis of high-speed vehicle systems using analytical, reduced-order, and surrogate modeling techniques. Full launch-to-landing flights of complete vehicles are considered and used to define flight envelopes with aeroelastic, aerothermal, and thermoelastic limits, tune in-the-loop flight controllers, and inform future design considerations. A partitioned approach to vehicle simulation is considered in which regions dominated by particular combinations of processes are made separate from the overall solution and simulated by a specialized set of models to improve overall processing
Spectral methods in numerical plasma simulation
International Nuclear Information System (INIS)
Coutsias, E.A.; Hansen, F.R.; Huld, T.; Knorr, G.; Lynov, J.P.
1989-01-01
An introduction is given to the use of spectral methods in numerical plasma simulation. As examples of the use of spectral methods, solutions to the two-dimensional Euler equations in both a simple, doubly periodic region, and on an annulus will be shown. In the first case, the solution is expanded in a two-dimensional Fourier series, while a Chebyshev-Fourier expansion is employed in the second case. A new, efficient algorithm for the solution of Poisson's equation on an annulus is introduced. Problems connected to aliasing and to short wavelength noise generated by gradient steepening are discussed. (orig.)
Numerical simulation of the cavitation's hydrodynamic excitement
International Nuclear Information System (INIS)
Hassis, H.; Dueymes, E.; Lauro, J.F.
1993-01-01
First, we study the motion, the velocity, the phases plane and the acoustic sources associated to a spherical bubble in a compressible or incompressible medium. The bubble can be excited by periodic or random excitements. We study the parameters which influence their behaviour: periodicity or not of motion, implosion and explosion or oscillation of bubble. We take into account this behaviour in a model of cavitation: it is a numerical simulation using population of bubbles which are with positions (in the cavitation volume) and sizes are random. These bubbles are excited by a random excitement: a model of turbulent flow or implosion and explosion of bubble. (author)
Numerical Simulations Of Flagellated Micro-Swimmers
Rorai, Cecilia; Markesteijn, Anton; Zaitstev, Mihail; Karabasov, Sergey
2017-11-01
We study flagellated microswimmers locomotion by representing the entire swimmer body. We discuss and contrast the accuracy and computational cost of different numerical approaches including the Resistive Force Theory, the Regularized Stokeslet Method and the Finite Element Method. We focus on how the accuracy of the methods in reproducing the swimming trajectories, velocities and flow field, compares to the sensitivity of these quantities to certain physical parameters, such as the body shape and the location of the center of mass. We discuss the opportunity and physical relevance of retaining inertia in our models. Finally, we present some preliminary results toward collective motion simulations. Marie Skodowska-Curie Individual Fellowship.
Numerical analysis of reactor internals under hydrodynamic loads
Energy Technology Data Exchange (ETDEWEB)
Kim, Da Hye; Chang, Yoon Suk [Kyung Hee Univ., Yongin (Korea, Republic of); Jhung, Myung Jo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)
2013-10-15
In the present study, six kinds of major equipment of a typical reactor internals were identified by incorporating recent research trend. Based on this, detailed numerical models were developed and used for establishment of optimum analysis methodology subjected to hydrodynamic loads. As a result, stress values of the major equipment were calculated through the acoustic-structure analysis under periodic hydrodynamic load and the turbulence-structure analysis under random hydrodynamic load. The numerical analysis scheme can be used for development of preventive action plan and management procedures of the reactor internals. Reactor internals installed in a pressure vessel have been exposed to harsh environment such as high neutron irradiation and temperature with complex fluid flow. As the increase of operational years of NPPs(Nuclear Power Plants), possibility of functional loss of the reactor internals is increased due to degradation caused by radiation embrittlement, thermal aging, fatigue, corrosion and FIV(Flow-Induced Vibration) etc. In practice, defects were detected at core support structure as well as upper and lower parts of structural assembly in European and United States NPPs. Recently, in a GALL(Generic Aging Lessons Learned) report, US NRC(Nuclear Regulatory Commission) identified reactor internals as a high priority component and addressed relevant management programs. In Korea, similar activities have been conducted for long-term operation beyond design lifetime but most of them were limited to qualitative evaluation based on examination and maintenance programs. Therefore, not only to reduce repair and replacement efforts but also to secure the stability of NPPs, necessity for development of quantitative evaluation technique as well as establishment of preventive action plan and management procedures is on the rise. The FIV represents the structural vibration phenomenon induced by liquid flow and generally occurs at contact surfaces. In the present
A multi-group neutron noise simulator for fast reactors
International Nuclear Information System (INIS)
Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe
2013-01-01
Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring
Simplified simulation of an experimental fast reactor plant
International Nuclear Information System (INIS)
Fujii, Masaaki; Fujita, Minoru.
1978-01-01
Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)
The numerical simulation of accelerator components
International Nuclear Information System (INIS)
Herrmannsfeldt, W.B.; Hanerfeld, H.
1987-05-01
The techniques of the numerical simulation of plasmas can be readily applied to problems in accelerator physics. Because the problems usually involve a single component ''plasma,'' and times that are at most, a few plasma oscillation periods, it is frequently possible to make very good simulations with relatively modest computation resources. We will discuss the methods and illustrate them with several examples. One of the more powerful techniques of understanding the motion of charged particles is to view computer-generated motion pictures. We will show several little movie strips to illustrate the discussions. The examples will be drawn from the application areas of Heavy Ion Fusion, electron-positron linear colliders and injectors for free-electron lasers. 13 refs., 10 figs., 2 tabs
Numerical simulation of human biped locomotion
International Nuclear Information System (INIS)
Ishiguro, Misako; Fujisaki, Masahide
1988-04-01
This report describes the numerical simulation of the motion of human-like robot which is one of the research theme of human acts simulation program (HASP) begun at the Computing Center of JAERI in 1987. The purpose of the theme is to model the human motion using robotics kinematic/kinetic equations and to get the joint angles as the solution. As the first trial, we treat the biped locomotion (walking) which is the most fundamental human motion. We implemented a computer program on FACOM M-780 computer, where the program is originated from the book of M. Vukobratovic in Yugoslavia, and made a graphic program to draw a walking shot sequence. Mainly described here are the mathematical model of the biped locomotion, implementation method of the computer program, input data for basic walking pattern, computed results and its validation, and graphic representation of human walking image. Literature survey on robotics equation and biped locomotion is also included. (author)
Direct numerical simulation of annular flows
Batchvarov, Assen; Kahouadji, Lyes; Chergui, Jalel; Juric, Damir; Shin, Seungwon; Craster, Richard V.; Matar, Omar K.
2017-11-01
Vertical counter-current two-phase flows are investigated using direct numerical simulations. The computations are carried out using Blue, a front-tracking-based CFD solver. Preliminary results show good qualitative agreement with experimental observations in terms of interfacial phenomena; these include three-dimensional, large-amplitude wave formation, the development of long ligaments, and droplet entrainment. The flooding phenomena in these counter current systems are closely investigated. The onset of flooding in our simulations is compared to existing empirical correlations such as Kutateladze-type and Wallis-type. The effect of varying tube diameter and fluid properties on the flooding phenomena is also investigated in this work. EPSRC, UK, MEMPHIS program Grant (EP/K003976/1), RAEng Research Chair (OKM).
Numerical Simulation of Duplex Steel Multipass Welding
Directory of Open Access Journals (Sweden)
Giętka T.
2016-12-01
Full Text Available Analyses based on FEM calculations have significantly changed the possibilities of determining welding strains and stresses at early stages of product design and welding technology development. Such an approach to design enables obtaining significant savings in production preparation and post-weld deformation corrections and is also important for utility properties of welded joints obtained. As a result, it is possible to make changes to a simulated process before introducing them into real production as well as to test various variants of a given solution. Numerical simulations require the combination of problems of thermal, mechanical and metallurgical analysis. The study presented involved the SYSWELD software-based analysis of GMA welded multipass butt joints made of duplex steel sheets. The analysis of the distribution of stresses and displacements were carried out for typical welding procedure as during real welding tests.
PC-Reactor-core transient simulation code
International Nuclear Information System (INIS)
Nakata, H.
1989-10-01
PC-REATOR, a reactor core transient simulation code has been developed for the real-time operator training on a IBM-PC microcomputer. The program presents capabilities for on-line exchange of the operating parameters during the transient simulation, by friendly keyboard instructions. The model is based on the point-kinetics approximation, with 2 delayed neutron percursors and up to 11 decay power generating groups. (author) [pt
Numerical simulation of a sour gas flare
Energy Technology Data Exchange (ETDEWEB)
Chambers, A. [Alberta Research Council, Devon, AB (Canada)
2008-07-01
Due to the limited amount of information in the literature on sour gas flares and the cost of conducting wind tunnel and field experiments on sour flares, this presentation presented a modelling project that predicted the effect of operating conditions on flare performance and emissions. The objectives of the project were to adapt an existing numerical model suitable for flare simulation, incorporate sulfur chemistry, and run simulations for a range of conditions typical of sour flares in Alberta. The study involved the use of modelling expertise at the University of Utah, and employed large eddy simulation (LES) methods to model open flames. The existing model included the prediction of turbulent flow field; hydrocarbon reaction chemistry; soot formation; and radiation heat transfer. The presentation addressed the unique features of the model and discussed whether LES could predict the flow field. Other topics that were presented included the results from a University of Utah comparison; challenges of the LES model; an example of a run time issue; predicting the impact of operating conditions; and the results of simulations. Last, several next steps were identified and preliminary results were provided. Future work will focus on reducing computation time and increasing information reporting. figs.
Numerical solution for identification of feedback coefficients in nuclear reactors
International Nuclear Information System (INIS)
Ebizuka, Yoshie; Sakai, Hideo
1975-01-01
Quasilinearization technique was studied to determine the Kinetic parameters of nuclear reactors. The method of solution was generalized to the determination of the parameters contained in a nonlinear system with nonlinear boundary conditions. A computer program, SNR-3, was developed to solve the resulting nonlinear two-point boundary value equations with generalized boundary conditions. In this paper, the problem formulation and the method of solution are explained for a general type of time dependent problem. A flow chart shows the procedure of numerical solution. The method was then applied to the determination of the critical factor and the reactivity feedback coefficients of reactors to investigate the accuracy and the applicability of the present method. The results showed that the present method was considerably successful, but that the random observation error effected the results of the identification. (Aoki, K.)
Numerical simulator of the CANDU fueling machine driving desk
International Nuclear Information System (INIS)
Doca, Cezar
2008-01-01
As a national and European premiere, in the 2003 - 2005 period, at the Institute for Nuclear Research Pitesti two CANDU fueling machine heads, no.4 and no.5, for the Nuclear Power Plant Cernavoda - Unit 2 were successfully tested. To perform the tests of these machines, a special CANDU fueling machine testing rig was built and was (and is) available for this goal. The design of the CANDU fueling machine test rig from the Institute for Nuclear Research Pitesti is a replica of the similar equipment operating in CANDU 6 type nuclear power plants. High technical level of the CANDU fueling machine tests required the using of an efficient data acquisition and processing Computer Control System. The challenging goal was to build a computer system (hardware and software) designed and engineered to control the test and calibration process of these fuel handling machines. The design takes care both of the functionality required to correctly control the CANDU fueling machine and of the additional functionality required to assist the testing process. Both the fueling machine testing rig and staff had successfully assessed by the AECL representatives during two missions. At same the time, at the Institute for Nuclear Research Pitesti was/is developed a numerical simulator for the CANDU fueling machine operators training. The paper presents the numerical simulator - a special PC program (software) which simulates the graphics and the functions and the operations at the main desk of the computer control system. The simulator permits 'to drive' a CANDU fueling machine in two manners: manual or automatic. The numerical simulator is dedicated to the training of operators who operate the CANDU fueling machine in a nuclear power plant with CANDU reactor. (author)
Visualization techniques in plasma numerical simulations
International Nuclear Information System (INIS)
Kulhanek, P.; Smetana, M.
2004-01-01
Numerical simulations of plasma processes usually yield a huge amount of raw numerical data. Information about electric and magnetic fields and particle positions and velocities can be typically obtained. There are two major ways of elaborating these data. First of them is called plasma diagnostics. We can calculate average values, variances, correlations of variables, etc. These results may be directly comparable with experiments and serve as the typical quantitative output of plasma simulations. The second possibility is the plasma visualization. The results are qualitative only, but serve as vivid display of phenomena in the plasma followed-up. An experience with visualizing electric and magnetic fields via Line Integral Convolution method is described in the first part of the paper. The LIC method serves for visualization of vector fields in two dimensional section of the three dimensional plasma. The field values can be known only in grid points of three-dimensional grid. The second part of the paper is devoted to the visualization techniques of the charged particle motion. The colour tint can be used for particle temperature representation. The motion can be visualized by a trace fading away with the distance from the particle. In this manner the impressive animations of the particle motion can be achieved. (author)
Reactor training simulator for the Replacement Research Reactor (RRR)
International Nuclear Information System (INIS)
Etchepareborda, A; Flury, C.A; Lema, F; Maciel, F; Alegrechi, D; Damico, M; Ibarra, G; Muguiro, M; Gimenez, M; Schlamp, M; Vertullo, A
2004-01-01
The main features of the ANSTO Replacement Research Reactor (RRR) Reactor Training Simulator (RTS) are presented.The RTS is a full-scope and partial replica simulator.Its scope includes a complete set of plant normal evolutions and malfunctions obtained from the plant design basis accidents list.All the systems necessary to implement the operating procedures associated to these transients are included.Within these systems both the variables connected to the plant SCADA and the local variables are modelled, leading to several thousands input-output variables in the plant mathematical model (PMM).The trainee interacts with the same plant SCADA, a Foxboro I/A Series system.Control room hardware is emulated through graphical displays with touch-screen.The main system models were tested against RELAP outputs.The RTS includes several modules: a model manager (MM) that encapsulates the plant mathematical model; a simulator human machine interface, where the trainee interacts with the plant SCADA; and an instructor console (IC), where the instructor commands the simulation.The PMM is built using Matlab-Simulink with specific libraries of components designed to facilitate the development of the nuclear, hydraulic, ventilation and electrical plant systems models [es
Reactor fuel element heat conduction via numerical Laplace transform inversion
International Nuclear Information System (INIS)
Ganapol, Barry D.; Furfaro, Roberto
2001-01-01
A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)
Reactor fuel element heat conduction via numerical Laplace transform inversion
Energy Technology Data Exchange (ETDEWEB)
Ganapol, Barry D.; Furfaro, Roberto [University of Arizona, Tucson, AZ (United States). Dept. of Aerospace and Mechanical Engineering], e-mail: ganapol@cowboy.ame.arizona.edu
2001-07-01
A newly developed numerical Laplace transform inversion (NLTI) will be presented to determine the transient temperature distribution within a nuclear reactor fuel element. The NLTI considered in this presentation has evolved to its present state over the past 10 years of application. The methodology adopted is one that relies on acceleration of the convergence of an infinite series towards its limit. The inversion will be applied to the prediction of the transient temperature distribution within an MTR type nuclear fuel element through a novel formulation of the solution to the transformed heat conduction equation. (author)
Direct numerical simulation of turbulent reacting flows
Energy Technology Data Exchange (ETDEWEB)
Chen, J.H. [Sandia National Laboratories, Livermore, CA (United States)
1993-12-01
The development of turbulent combustion models that reflect some of the most important characteristics of turbulent reacting flows requires knowledge about the behavior of key quantities in well defined combustion regimes. In turbulent flames, the coupling between the turbulence and the chemistry is so strong in certain regimes that is is very difficult to isolate the role played by one individual phenomenon. Direct numerical simulation (DNS) is an extremely useful tool to study in detail the turbulence-chemistry interactions in certain well defined regimes. Globally, non-premixed flames are controlled by two limiting cases: the fast chemistry limit, where the turbulent fluctuations. In between these two limits, finite-rate chemical effects are important and the turbulence interacts strongly with the chemical processes. This regime is important because industrial burners operate in regimes in which, locally the flame undergoes extinction, or is at least in some nonequilibrium condition. Furthermore, these nonequilibrium conditions strongly influence the production of pollutants. To quantify the finite-rate chemistry effect, direct numerical simulations are performed to study the interaction between an initially laminar non-premixed flame and a three-dimensional field of homogeneous isotropic decaying turbulence. Emphasis is placed on the dynamics of extinction and on transient effects on the fine scale mixing process. Differential molecular diffusion among species is also examined with this approach, both for nonreacting and reacting situations. To address the problem of large-scale mixing and to examine the effects of mean shear, efforts are underway to perform large eddy simulations of round three-dimensional jets.
Numerical modeling of disperse material evaporation in axisymmetric thermal plasma reactor
Directory of Open Access Journals (Sweden)
Stefanović Predrag Lj.
2003-01-01
Full Text Available A numerical 3D Euler-Lagrangian stochastic-deterministic (LSD model of two-phase flow laden with solid particles was developed. The model includes the relevant physical effects, namely phase interaction, panicle dispersion by turbulence, lift forces, particle-particle collisions, particle-wall collisions, heat and mass transfer between phases, melting and evaporation of particles, vapour diffusion in the gas flow. It was applied to simulate the processes in thermal plasma reactors, designed for the production of the ceramic powders. Paper presents results of extensive numerical simulation provided (a to determine critical mechanism of interphase heat and mass transfer in plasma flows, (b to show relative influence of some plasma reactor parameters on solid precursor evaporation efficiency: 1 - inlet plasma temperature, 2 - inlet plasma velocity, 3 - particle initial diameter, 4 - particle injection angle a, and 5 - reactor wall temperature, (c to analyze the possibilities for high evaporation efficiency of different starting solid precursors (Si, Al, Ti, and B2O3 powder, and (d to compare different plasma reactor configurations in conjunction with disperse material evaporation efficiency.
Apparatus for simulating a reactor core
International Nuclear Information System (INIS)
Yokomizo, Osamu; Kiguchi, Takashi; Motoda, Hiroshi; Takeda, Renzo.
1975-01-01
Object: To facilitate searching of input and output of information and to efficiently perform trial-and-error in a short time. Structure: Kinds of necessary input information are stored in an input information converter and are displayed by an image display through an image control. An operator operates an information input device to input information. This input information is converted by the input information converter into a form used in a reactor core simulation counter. The reactor core simulation counter simulates a state of the core to the input information converted, and outputs it as an output information. An output information converter converts output information into a form that may be displayed as an image and feeds it to the image control. The operator may correct the input information while viewing the output information displayed on the image display to immediately perform succeeding calculation. (Kamimura, M.)
Study and simulation of a parallel numerical processing machine
International Nuclear Information System (INIS)
Bel Hadj, Slaheddine
1981-12-01
This study has been carried out in the perspective of the implementation on a minicomputer of the NEPTUNIX package (software for the resolution of very large algebra-differential equation systems). Aiming at increasing the system performance, a previous research work has shown the necessity of reducing the execution time of certain numerical computation tasks, which are of frequent use. It has also demonstrated the feasibility of handling these tasks with efficient algorithms of parallel type. The present work deals with the study and simulation of a parallel architecture processor adapted to the fast execution of these algorithms. A minicomputer fitted with a connection to such a parallel processor, has a greatly extended computing power. Then the architecture of a parallel numerical processor, based on the use of VLSI microprocessors and co-processors, is described. Its design aims at the best cost / performance ratio. The last part deals with the simulation processor with the 'CHAMBOR' program. Results show an increasing factor of 30 in speed, in comparison with the execution on a MITRA 15 minicomputer. Moreover the conflicts importance, mainly at the level of access to a shared resource is evaluated. Although this implementation has been designed having in mind a dedicated application, other uses could be envisaged, particularly for the simulation of nuclear reactors: operator guiding system, the behavioural study under accidental circumstances, etc. (author) [fr
Numerical simulation on coolant flow and heat transfer in core
International Nuclear Information System (INIS)
Yao Zhaohui; Wang Xuefang; Shen Mengyu
1997-01-01
To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis
Advanced nuclear reactors and their simulators
International Nuclear Information System (INIS)
Chaushevski, Anton; Boshevski, Tome
2003-01-01
Population growth, economy development and improvement life standard impact on continually energy needs as well as electricity. Fossil fuels have limited reserves, instability market prices and destroying environmental impacts. The hydro energy capacities highly depend on geographic and climate conditions. The nuclear fission is significant factor for covering electricity needs in this century. Reasonable capital costs, low fuel and operating expenses, environmental acceptable are some of the facts that makes the nuclear energy an attractive option especially for the developing countries. The simulators for nuclear reactors are an additional software tool in order to understand, study research and analyze the processes in nuclear reactors. (Original)
Interatomic potentials for fusion reactor material simulations
International Nuclear Information System (INIS)
Bjoerkas, C.
2009-01-01
In this thesis, the behaviour of a material situated in a fusion reactor was studied using molecular dynamics simulations. Simulations of processes in the next generation fusion reactor ITER include the reactor materials beryllium, carbon and tungsten as well as the plasma hydrogen isotopes. This means that interaction models, i.e. interatomic potentials, for this complicated quaternary system are needed. The task of finding such potentials is nonetheless nearly at its end, since models for the beryllium-carbon-hydrogen interactions were constructed in this thesis and as a continuation of that work, a beryllium-tungsten model is under development. These potentials are combinable with the earlier tungsten-carbon-hydrogen ones. The potentials were used to explain the chemical sputtering of beryllium due to deuterium plasma exposure. During experiments, a large fraction of the sputtered beryllium atoms were observed to be released as BeD molecules, and the simulations identified the swift chemical sputtering mechanism, previously not believed to be important in metals, as the underlying mechanism. Radiation damage in the reactor structural materials vanadium, iron and iron chromium, as well as in the wall material tungsten and the mixed alloy tungsten carbide, was also studied in this thesis. Interatomic potentials for vanadium, tungsten and iron were modified to be better suited for simulating collision cascades that are formed during particle irradiation, and the potential features affecting the resulting primary damage were identified. Including the often neglected electronic effects in the simulations was also shown to have an impact on the damage. With proper tuning of the electronphonon interaction strength, experimentally measured quantities related to ion-beam mixing in iron could be reproduced. The damage in tungsten carbide alloys showed elemental asymmetry, as the major part of the damage consisted of carbon defects. On the other hand, modelling the damage
Numerical simulation of premixed turbulent methane combustion
International Nuclear Information System (INIS)
Bell, John B.; Day, Marcus S.; Grcar, Joseph F.
2001-01-01
In this paper we study the behavior of a premixed turbulent methane flame in three dimensions using numerical simulation. The simulations are performed using an adaptive time-dependent low Mach number combustion algorithm based on a second-order projection formulation that conserves both species mass and total enthalpy. The species and enthalpy equations are treated using an operator-split approach that incorporates stiff integration techniques for modeling detailed chemical kinetics. The methodology also incorporates a mixture model for differential diffusion. For the simulations presented here, methane chemistry and transport are modeled using the DRM-19 (19-species, 84-reaction) mechanism derived from the GRIMech-1.2 mechanism along with its associated thermodynamics and transport databases. We consider a lean flame with equivalence ratio 0.8 for two different levels of turbulent intensity. For each case we examine the basic structure of the flame including turbulent flame speed and flame surface area. The results indicate that flame wrinkling is the dominant factor leading to the increased turbulent flame speed. Joint probability distributions are computed to establish a correlation between heat release and curvature. We also investigate the effect of turbulent flame interaction on the flame chemistry. We identify specific flame intermediates that are sensitive to turbulence and explore various correlations between these species and local flame curvature. We identify different mechanisms by which turbulence modulates the chemistry of the flame
Intensification of transesterification via sonication numerical simulation and sensitivity study
International Nuclear Information System (INIS)
Janajreh, Isam; ElSamad, Tala; Noorul Hussain, Mohammed
2017-01-01
Highlights: • 3D numerical simulation of transesterification is accomplished. • A non-isothermal, reactive Navier–stokes was carried out. • Conventional and sonicated process was compared as far as reaction kinetics and yield. • Higher kinetic rates are achieved at lower molar ratios in sonicated process. • It validates feasibility of numerical simulation for transesterification assessment. - Abstract: Transesterification is known as slow reaction that can take over several hours to complete. The process involves two immiscible reactants to produce the biodiesel and the byproduct glycerol. Biodiesel commercialization has always been hindered by the long process times of the transesterification reaction. Catalyzing the process and increasing the agitation rate is the mode of intensifying the process additional to the increase of the molar ratio, temperature, circulation that all penalize the overall process metrics. Finding shorter path by reducing the reaction into a few minutes and ensures high quality biodiesel, in economically viable way is coming along with sonication. This drastic reduction moves the technology from the slow batch process into the high throughput continuous process. In a practical sense this means a huge optimization for the biodiesel production process which opens pathways for faster, voluminous and cheaper production. The mechanism of sonication assisted reaction is explained by the creation of microbubbles which increases the interfacial surface reaction areas and the presence of high localized temperature and turbulence as these microbubbles implode. As a result the reaction kinetics of sonicated transesterification as inferred by several authors is much faster. The aim of this work is to implement the inferred rates in a high fidelity numerical reactive flow simulation model while considering the reactor geometry. It is based on Navier–Stokes equations coupled with energy equation for non-isothermal flow and the transport
Directory of Open Access Journals (Sweden)
Ferschneider G.
2006-11-01
Full Text Available Fixed bed reactors with a single fluid phase are widely used in the refining or petrochemical industries for reaction processes catalysed by a solid phase. The design criteria for industrial reactors are relatively well known. However, they rely on a one-dimensional writing and on the separate resolution of the equation of conservation of mass and energy, and of momentum. Thus, with complex geometries, the influence of hydrodynamics on the effectiveness of the catalyst bed cannot be taken into account. The calculation method proposed is based on the multi-dimensional writing and the simultaneous resolution of the local conservation equations. The example discussed concerns fixed-bed catalytic reactors. These reactors are distinguished by their annular geometry and the radial circulation of the feedstock. The flow is assumed to be axisymmetric. The reaction process is reflected by a simplified kinetic mechanism involving ten chemical species. Calculation of the hydrodynamic (mean velocities, pressure, thermal and mass fields (concentration of each species serves to identify the influence of internal components in two industrial reactor geometries. The map of the quantity of coke formed and deposited on the catalyst, calculated by the model, reveals potential areas of poor operation. Les réacteurs à lit fixe avec une seule phase fluide sont largement utilisés dans l'industrie du raffinage et de la pétrochimie, pour mettre en oeuvre un processus réactionnel catalysé par une phase solide. Les règles de conception des réacteurs industriels sont relativement bien connues. Cependant, elles reposent sur l'écriture monodimensionnelle et la résolution séparée, d'une part, des équations de conservation de la masse et de l'énergie et d'autre part, de la quantité de mouvement. Ainsi dans le cas de géométries complexes, l'influence de l'hydrodynamique sur l'efficacité du lit catalytique ne peut être prise en compte. La méthode de calcul
International Nuclear Information System (INIS)
Kamide, H.; Kimura, N.; Miyakoshi, H.; Nagasawa, K.
2001-01-01
Decay heat removal using natural circulation is one of the important functions for the safety of fast reactors. As a decay heat removal system, direct reactor auxiliary cooling system has been selected in current designs of fast reactors. In this design, dumped heat exchanger provides cold sodium and it covers the reactor core outlet. The cold sodium can penetrate into the gap region between the subassemblies. This gap flow is referred as inter-wrapper flow (IWF). A numerical estimation method for such natural circulation phenomena in a reactor core has been developed, which models each subassembly as a rectangular duct with gap region between the subassemblies and also the upper plenum in a reactor vessel. This numerical simulation method was verified based on experimental data of a sodium test using 7- subassembly core model and also a water test which simulates IWF using the 1/12 sector model of a reactor core. We applied the estimation method to the natural circulation in a 600 MW class fast reactor. The temperature in the core strongly depended on IWF, flow redistribution in the core, and inter-subassembly heat transfer. It is desired for prediction methods on the natural circulation to simulate these phenomena. (author)
Numerical simulation of heterogeneous phase transformations
International Nuclear Information System (INIS)
Combeau, H.; Lacaze, J.
1993-01-01
A numerical model is presented for the simulation of diffusion controlled phase transformations in multicomponent alloys. A closed system is considered, with simple geometric shape, either planar, cylindrical or spherical. The temperature inside this microscopic volume is homogeneous, but can vary according to any specified monoteneous law. Particular care has been given to the description of the solute profiles where the concentration gradients are the steepest, i.e. near the interface between the parent and the resultant phases. Solute redistribution at the interface is described by means of an original method which ensures that the overall solute balance is satisfied. A non linear system is obtained which includes the diffusion equations in both phases and the boundary conditions. The solution of this system makes use of a special algorithm which has been devised for a quick convergence. An example is presented which deals with microsegregation build-up during solidification of a multi-component nickel base alloy. (orig.)
Numerical simulations of coupled problems in engineering
2014-01-01
This book presents and discusses mathematical models, numerical methods and computational techniques used for solving coupled problems in science and engineering. It takes a step forward in the formulation and solution of real-life problems with a multidisciplinary vision, accounting for all of the complex couplings involved in the physical description. Simulation of multifaceted physics problems is a common task in applied research and industry. Often a suitable solver is built by connecting together several single-aspect solvers into a network. In this book, research in various fields was selected for consideration: adaptive methodology for multi-physics solvers, multi-physics phenomena and coupled-field solutions, leading to computationally intensive structural analysis. The strategies which are used to keep these problems computationally affordable are of special interest, and make this an essential book.
Numerical simulation of distorted crystal Darwin width
International Nuclear Information System (INIS)
Wang Li; Xu Zhongmin; Wang Naxiu
2012-01-01
A new numerical simulation method according to distorted crystal optical theory was used to predict the direct-cooling crystal monochromator optical properties(crystal Darwin width) in this study. The finite element analysis software was used to calculate the deformed displacements of DCM crystal and to get the local reciprocal lattice vector of distorted crystal. The broadening of direct-cooling crystal Darwin width in meridional direction was estimated at 4.12 μrad. The result agrees well with the experimental data of 5 μrad, while it was 3.89 μrad by traditional calculation method of root mean square (RMS) of the slope error in the center line of footprint. The new method provides important theoretical support for designing and processing of monochromator crystal for synchrotron radiation beamline. (authors)
Numerical simulation of magnetic heat pumps
International Nuclear Information System (INIS)
Smaili, A.; Masson, C.
2002-01-01
This article presents a numerical method for performance predictions of magnetic heat pump (MHP) devices. Such devices consist primarily of a magnetic regenerator (solid refrigerant media) and circulating fluid. Unlike conventional gas-cycles, MHP devices function according to thermomagnetic cycles which do not require neither compressor nor expander. In this paper, the flow field throughout the regenerator is described by continuity and unsteady incompressible Navier-Stokes equations. The heat transfer between fluid and solid is introduced by considering the corresponding energy equations. The proposed mathematical model has been solved using a control volume finite element method. The fully implicit scheme is used for time discretization. Simulation results including heating capacity and coefficient of performance are presented for a given MHP cycle. Mainly, the effects of cycle frequency, mass flow rate and the magnetic regenerator mass are investigated. (author)
Numerical simulations of convectively excited gravity waves
International Nuclear Information System (INIS)
Glatzmaier, G.A.
1983-01-01
Magneto-convection and gravity waves are numerically simulated with a nonlinear, three-dimensional, time-dependent model of a stratified, rotating, spherical fluid shell heated from below. A Solar-like reference state is specified while global velocity, magnetic field, and thermodynamic perturbations are computed from the anelastic magnetohydrodynamic equations. Convective overshooting from the upper (superadiabatic) part of the shell excites gravity waves in the lower (subadiabatic) part. Due to differential rotation and Coriolis forces, convective cell patterns propagate eastward with a latitudinally dependent phase velocity. The structure of the excited wave motions in the stable region is more time-dependent than that of the convective motions above. The magnetic field tends to be concentrated over giant-cell downdrafts in the convective zone but is affected very little by the wave motion in the stable region
The Beam Break-Up Numerical Simulator
International Nuclear Information System (INIS)
Travish, G.A.
1989-11-01
Beam Break-Up (BBU) is a severe constraint in accelerator design, limiting beam current and quality. The control of BBU has become the focus of much research in the design of the next generation collider, recirculating and linear induction accelerators and advanced accelerators. Determining the effect on BBU of modifications to cavities, the focusing elements or the beam is frequently beyond the ability of current analytic models. A computer code was written to address this problem. The Beam Break-Up Numerical Simulator (BBUNS) was designed to numerically solve for beam break-up (BBU) due to an arbitrary transverse wakefield. BBUNS was developed to be as user friendly as possible on the Cray computer series. The user is able to control all aspects of input and output by using a single command file. In addition, the wakefield is specified by the user and read in as a table. The program can model energy variations along and within the beam, focusing magnetic field profiles can be specified, and the graphical output can be tailored. In this note we discuss BBUNS, its structure and application. Included are detailed instructions, examples and a sample session of BBUNS. This program is available for distribution. 50 refs., 18 figs., 5 tabs
Numerical simulation of installation of skirt foundations
Energy Technology Data Exchange (ETDEWEB)
Vangelsten, Bjoern Vidar
1997-12-31
Skirt foundation has been increasingly used for permanent offshore oil installations and anchors for drilling ships. Suction is commonly used in skirt foundation installing. If a large amount of suction is applied, the soil around the foundation may fail and the foundation become useless. This thesis studies failure due to high seepage gradients, aiming to provide a basis for reducing the risk of such failures. Skirt penetration model testing has shown that to solve the problem one must understand what is going on at the skirt tip during suction installation. A numerical model based on micro mechanics was developed as continuum hypothesis was seen to be unsuitable to describe the processes in the critical phases of the failure. The numerical model combines two-dimensional elliptical particles with the finite difference method for flow to model water flow in a granular material. The key idea is to formulate the permeability as a function of the porosity of the grain assembly and so obtain an interaction between the finite difference method on flow and the particle movement. A computer program, DYNELL, was developed and used to simulate: (1) weight penetration of a skirt wall, (2) combined suction and weight penetration of a skirt wall, and (3) critical gradient tests around a skirt wall to study failure mechanisms. The model calculations agree well with laboratory experiments. 16 refs., 124 figs., 21 tabs.
Coupled numerical simulation of fire in tunnel
Pesavento, F.; Pachera, M.; Schrefler, B. A.; Gawin, D.; Witek, A.
2018-01-01
In this work, a coupling strategy for the analysis of a tunnel under fire is presented. This strategy consists in a "one-way" coupling between a tool considering the computational fluid dynamics and radiation with a model treating concrete as a multiphase porous material exposed to high temperature. This global approach allows for taking into account in a realistic manner the behavior of the "system tunnel", composed of the fluid and the solid domain (i.e. the concrete structures), from the fire onset, its development and propagation to the response of the structure. The thermal loads as well as the moisture exchange between the structure surface and the environment are calculated by means of computational fluid dynamics. These set of data are passed in an automatic way to the numerical tool implementing a model based on Multiphase Porous Media Mechanics. Thanks to this strategy the structural verification is no longer based on the standard fire curves commonly used in the engineering practice, but it is directly related to a realistic fire scenario. To show the capability of this strategy some numerical simulations of a fire in the Brenner Base Tunnel, under construction between Italy and Austria, is presented. The numerical simulations show the effects of a more realistic distribution of the thermal loads with respect to the ones obtained by using the standard fire curves. Moreover, it is possible to highlight how the localized thermal load generates a non-uniform pressure rise in the material, which results in an increase of the structure stress state and of the spalling risk. Spalling is likely the most dangerous collapse mechanism for a concrete structure. This coupling approach still represents a "one way" strategy, i.e. realized without considering explicitly the mass and energy exchange from the structure to the fluid through the interface. This results in an approximation, but from physical point of view the current form of the solid-fluid coupling is
Numerical simulations of capillary barrier field tests
International Nuclear Information System (INIS)
Morris, C.E.; Stormont, J.C.
1997-01-01
Numerical simulations of two capillary barrier systems tested in the field were conducted to determine if an unsaturated flow model could accurately represent the observed results. The field data was collected from two 7-m long, 1.2-m thick capillary barriers built on a 10% grade that were being tested to investigate their ability to laterally divert water downslope. One system had a homogeneous fine layer, while the fine soil of the second barrier was layered to increase its ability to laterally divert infiltrating moisture. The barriers were subjected first to constant infiltration while minimizing evaporative losses and then were exposed to ambient conditions. The continuous infiltration period of the field tests for the two barrier systems was modelled to determine the ability of an existing code to accurately represent capillary barrier behavior embodied in these two designs. Differences between the field test and the model data were found, but in general the simulations appeared to adequately reproduce the response of the test systems. Accounting for moisture retention hysteresis in the layered system will potentially lead to more accurate modelling results and is likely to be important when developing reasonable predictions of capillary barrier behavior
Numerical simulation for nuclear pumped laser
Energy Technology Data Exchange (ETDEWEB)
Sakasai, Kaoru [Japan Atomic Energy Research Inst., Tokyo (Japan)
1998-07-01
To apply nuclear pumped laser of {sup 3}He-Ne-Ar gas to detect neutron, the optimum gas mixture was investigated by numerical simulation. When {sup 3}He-Ne-Ar mixture gas are irradiated by neutron, proton and triton with high velocity are produced by {sup 3}He(np)T and two charge particles ionized {sup 3}He, Ne and Ar which reacted each other and attained to 3p`(1/2){sub 0}-3S`(1/2). The calculation method is constructed by defining the rate equations of each ion and exited atom and the electron energy balance equation and by time integrating the simultaneous differential equations of the above two equations and the law of conservation of charge. Penning ionization and energy transport by elastic collision of neutral atom were considered in the transport process of electron energy direct ionization by secondary charge particle. Calculation time was 1 msec. The optimum component was shown 3 atm He, 24 Torr He and 8 Torr Ar by simulation. Laser oscilation was generated under the conditions 3.3 x 10{sup 14} (N/cm{sup 2}/5) thermal neutron flux at 50 cm laser cell length and 99% coefficient of reflection of mirror. After laser oscilation, laser output was proportional to neutron flux. These results showed nuclear pumped laser of {sup 3}He-Ne-Ar was able to detect optically neutron. (S.Y)
Collisionless microinstabilities in stellarators. II. Numerical simulations
International Nuclear Information System (INIS)
Proll, J. H. E.; Xanthopoulos, P.; Helander, P.
2013-01-01
Microinstabilities exhibit a rich variety of behavior in stellarators due to the many degrees of freedom in the magnetic geometry. It has recently been found that certain stellarators (quasi-isodynamic ones with maximum-J geometry) are partly resilient to trapped-particle instabilities, because fast-bouncing particles tend to extract energy from these modes near marginal stability. In reality, stellarators are never perfectly quasi-isodynamic, and the question thus arises whether they still benefit from enhanced stability. Here, the stability properties of Wendelstein 7-X and a more quasi-isodynamic configuration, QIPC, are investigated numerically and compared with the National Compact Stellarator Experiment and the DIII-D tokamak. In gyrokinetic simulations, performed with the gyrokinetic code GENE in the electrostatic and collisionless approximation, ion-temperature-gradient modes, trapped-electron modes, and mixed-type instabilities are studied. Wendelstein 7-X and QIPC exhibit significantly reduced growth rates for all simulations that include kinetic electrons, and the latter are indeed found to be stabilizing in the energy budget. These results suggest that imperfectly optimized stellarators can retain most of the stabilizing properties predicted for perfect maximum-J configurations
Numerical simulation of the Polywell device
International Nuclear Information System (INIS)
Simmons, K.H.; Santarius, J.F.
1995-01-01
Recent ideas concerning inertial-electrostatic confinement (IEC) of fusion plasmas coupled with recent experimental results have motivated looking at the problem of confinement of these plasmas in both the gridded (pure electrostatic) and magnetically assisted (via confinement of high beta plasmas in a magnetic cusp) configuration. Questions exist as to the nature of the potential well structure and the confinement properties of high beta plasmas in magnetic cusp configurations. This work focuses on the magnetically assisted concept known as the Polywell trademark. Results are reported on the numerical simulation of IEC plasmas aimed at answering some of these questions. In particular the authors focus on two aspects of the Polywell, namely the structure of the magnetic cusp field in the Polywell configuration and the nature of the confinement of a high beta plasma in a magnetic cusp field. The existence of line cusps in the Polywell is still in dispute. A computer code for modeling the magnetic field structure and mod-B surface has been written and results are presented for the Polywell. Another source of controversy is the nature of the confinement of a high beta plasma in a magnetic cusp, and in particular in the polywell. Results from 2-D Particle In Cell (PIC) simulations aimed at answering some of these questions are presented
Direct numerical simulation of human phonation
Bodony, Daniel; Saurabh, Shakti
2017-11-01
The generation and propagation of the human voice in three-dimensions is studied using direct numerical simulation. A full body domain is employed for the purpose of directly computing the sound in the region past the speaker's mouth. The air in the vocal tract is modeled as a compressible and viscous fluid interacting with the elastic vocal folds. The vocal fold tissue material properties are multi-layered, with varying stiffness, and a linear elastic transversely isotropic model is utilized and implemented in a quadratic finite element code. The fluid-solid domains are coupled through a boundary-fitted interface and utilize a Poisson equation-based mesh deformation method. A kinematic constraint based on a specified minimum gap between the vocal folds is applied to prevent collision during glottal closure. Both near VF flow dynamics and far-field acoustics have been studied. A comparison is drawn to current two-dimensional simulations as well as to data from the literature. Near field vocal fold dynamics and glottal flow results are studied and in good agreement with previous three-dimensional phonation studies. Far-field acoustic characteristics, when compared to their two-dimensional counterpart, are shown to be sensitive to the dimensionality. Supported by the National Science Foundation (CAREER Award Number 1150439).
Energy Technology Data Exchange (ETDEWEB)
Grandotto Biettoli, M
2006-04-15
The report presents globally the works done by the author in the thermohydraulic applied to nuclear reactors flows. It presents the studies done to the numerical simulation of the two phase flows in the steam generators and a finite element method to compute these flows. (author)
Numerical Simulations of Hypersonic Boundary Layer Transition
Bartkowicz, Matthew David
Numerical schemes for supersonic flows tend to use large amounts of artificial viscosity for stability. This tends to damp out the small scale structures in the flow. Recently some low-dissipation methods have been proposed which selectively eliminate the artificial viscosity in regions which do not require it. This work builds upon the low-dissipation method of Subbareddy and Candler which uses the flux vector splitting method of Steger and Warming but identifies the dissipation portion to eliminate it. Computing accurate fluxes typically relies on large grid stencils or coupled linear systems that become computationally expensive to solve. Unstructured grids allow for CFD solutions to be obtained on complex geometries, unfortunately, it then becomes difficult to create a large stencil or the coupled linear system. Accurate solutions require grids that quickly become too large to be feasible. In this thesis a method is proposed to obtain more accurate solutions using relatively local data, making it suitable for unstructured grids composed of hexahedral elements. Fluxes are reconstructed using local gradients to extend the range of data used. The method is then validated on several test problems. Simulations of boundary layer transition are then performed. An elliptic cone at Mach 8 is simulated based on an experiment at the Princeton Gasdynamics Laboratory. A simulated acoustic noise boundary condition is imposed to model the noisy conditions of the wind tunnel and the transitioning boundary layer observed. A computation of an isolated roughness element is done based on an experiment in Purdue's Mach 6 quiet wind tunnel. The mechanism for transition is identified as an instability in the upstream separation region and a comparison is made to experimental data. In the CFD a fully turbulent boundary layer is observed downstream.
Numerical simulations of the mantle lithosphere delamination
Morency, C.; Doin, M.-P.
2004-03-01
Sudden uplift, extension, and increased igneous activity are often explained by rapid mechanical thinning of the lithospheric mantle. Two main thinning mechanisms have been proposed, convective removal of a thickened lithospheric root and delamination of the mantle lithosphere along the Moho. In the latter case, the whole mantle lithosphere peels away from the crust by the propagation of a localized shear zone and sinks into the mantle. To study this mechanism, we perform two-dimensional (2-D) numerical simulations of convection using a viscoplastic rheology with an effective viscosity depending strongly on temperature, depth, composition (crust/mantle), and stress. The simulations develop in four steps. (1) We first obtain "classical" sublithospheric convection for a long time period (˜300 Myr), yielding a slightly heterogeneous lithospheric temperature structure. (2) At some time, in some simulations, a strong thinning of the mantle occurs progressively in a small area (˜100 km wide). This process puts the asthenosphere in direct contact with the lower crust. (3) Large pieces of mantle lithosphere then quickly sink into the mantle by the horizontal propagation of a detachment level away from the "asthenospheric conduit" or by progressive erosion on the flanks of the delaminated area. (4) Delamination pauses or stops when the lithospheric mantle part detaches or when small-scale convection on the flanks of the delaminated area is counterbalanced by heat diffusion. We determine the parameters (crustal thicknesses, activation energies, and friction coefficients) leading to delamination initiation (step 2). We find that delamination initiates where the Moho temperature is the highest, as soon as the crust and mantle viscosities are sufficiently low. Delamination should occur on Earth when the Moho temperature exceeds ˜800°C. This condition can be reached by thermal relaxation in a thickened crust in orogenic setting or by corner flow lithospheric erosion in the
A research reactor simulator for operators training and teaching
International Nuclear Information System (INIS)
De Carvalho, R. P.; Maiorino, J. R.
2006-01-01
This work describes a training simulator of Research Reactors (RR). The simulator is an interactive tool for teaching and operator training of the bases of the RR operation, reactor physics and thermal hydraulics. The Brazilian IEA-R1 RR was taken as the reference (default configuration). The implementation of the simulator consists of the modeling of the process and system (neutronics, thermal hydraulics), its numerical solution, and the implementation of the man-machine interface through visual interactive screens. The point kinetics model was used for the nuclear process and the heat and mass conservation models were used for the thermal hydraulic feed back in the average core channel. The heat exchanger and cooling tower were also modeled. The main systems were: the reactivity control system, including the automatic control, and the primary and secondary coolant systems. The Visual C++ was used to codes and graphics lay-outs. The simulator is to be used in a PC with Windows XP system. The simulator allows simulation in real time of start up, power maneuver, and shut down. (authors)
Numerical simulation of a semi-indirect evaporative cooler
Energy Technology Data Exchange (ETDEWEB)
Martin, R. Herrero [Departamento de Ingenieria Termica y de Fluidos, Universidad Politecnica de Cartagena, C/Dr. Fleming, s/n (Campus Muralla), 30202 Cartagena, Murcia (Spain)
2009-11-15
This paper presents the experimental study and numerical simulation of a semi-indirect evaporative cooler (SIEC), which acts as an energy recovery device in air conditioning systems. The numerical simulation was conducted by applying the CFD software FLUENT implementing a UDF to model evaporation/condensation. The numerical model was validated by comparing the simulation results with experimental data. Experimental data and numerical results agree for the lower relative humidity series but not for higher relative humidity values. (author)
International Nuclear Information System (INIS)
Hussein, M.S; Lewis, B.J.; Bonin, H.W.
2013-01-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)
2013-07-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)
Numerical simulation of "an American haboob"
Vukovic, A.; Vujadinovic, M.; Pejanovic, G.; Andric, J.; Kumjian, M. R.; Djurdjevic, V.; Dacic, M.; Prasad, A. K.; El-Askary, H. M.; Paris, B. C.; Petkovic, S.; Nickovic, S.; Sprigg, W. A.
2014-04-01
A dust storm of fearful proportions hit Phoenix in the early evening hours of 5 July 2011. This storm, an American haboob, was predicted hours in advance because numerical, land-atmosphere modeling, computing power and remote sensing of dust events have improved greatly over the past decade. High-resolution numerical models are required for accurate simulation of the small scales of the haboob process, with high velocity surface winds produced by strong convection and severe downbursts. Dust productive areas in this region consist mainly of agricultural fields, with soil surfaces disturbed by plowing and tracks of land in the high Sonoran Desert laid barren by ongoing draught. Model simulation of the 5 July 2011 dust storm uses the coupled atmospheric-dust model NMME-DREAM (Non-hydrostatic Mesoscale Model on E grid, Janjic et al., 2001; Dust REgional Atmospheric Model, Nickovic et al., 2001; Pérez et al., 2006) with 4 km horizontal resolution. A mask of the potentially dust productive regions is obtained from the land cover and the normalized difference vegetation index (NDVI) data from the Moderate Resolution Imaging Spectroradiometer (MODIS). The scope of this paper is validation of the dust model performance, and not use of the model as a tool to investigate mechanisms related to the storm. Results demonstrate the potential technical capacity and availability of the relevant data to build an operational system for dust storm forecasting as a part of a warning system. Model results are compared with radar and other satellite-based images and surface meteorological and PM10 observations. The atmospheric model successfully hindcasted the position of the front in space and time, with about 1 h late arrival in Phoenix. The dust model predicted the rapid uptake of dust and high values of dust concentration in the ensuing storm. South of Phoenix, over the closest source regions (~25 km), the model PM10 surface dust concentration reached ~2500 μg m-3, but
Simulating the temperature noise in fast reactor fuel assemblies
International Nuclear Information System (INIS)
Kebadze, B.V.; Pykhtina, T.V.; Tarasko, M.Z.
1987-01-01
Characteristics of temperature noise at various modes of coolant flow in fast reactor fuel assemblies (FA) and for different points of sensor installation are investigated. Stationary mode of coolant flow and mode with a partial overlapping of FA through cross section, resulting in local temperature increase and sodium boiling, are considered. Numerical simulation permits to evaluate time characteristicsof temperature noise and to formulate requirements for dynamic characteristics of the sensors, and also to clarify the dependence of coolant distribution parameters on the sensor location and peculiarities of stationary temperature profile
Aroudam, El. H.
In this paper, we present a modelling of the performance of a reactor of a solar cooling machine based carbon-ammonia activated bed. Hence, for a solar radiation, measured in the Energetic Laboratory of the Faculty of Sciences in Tetouan (northern Morocco), the proposed model computes the temperature distribution, the pressure and the ammonia concentration within the activated carbon bed. The Dubinin-Radushkevich formula is used to compute the ammonia concentration distribution and the daily cycled mass necessary to produce a cooling effect for an ideal machine. The reactor is heated at a maximum temperature during the day and cool at the night. A numerical simulation is carried out employing the recorded solar radiation data measured locally and the daily ambient temperature for the typical clear days. Initially the reactor is at ambient temperature, evaporating pressure; Pev=Pst(Tev=0 ∘C) and maintained at uniform concentration. It is heated successively until the threshold temperature corresponding to the condensing pressure; Pcond=Pst(Tam) (saturation pressure at ambient temperature; in the condenser) and until a maximum temperature at a constant pressure; Pcond. The cooling of the reactor is characterised by a fall of temperature to the minimal values at night corresponding to the end of a daily cycle. We use the mass balance equations as well as energy equation to describe heat and mass transfer inside the medium of three phases. A numerical solution of the obtained non linear equations system based on the implicit finite difference method allows to know all parameters characteristic of the thermodynamic cycle and consider principally the daily evolution of temperature, ammonia concentration for divers positions inside the reactor. The tube diameter of the reactor shows the dependence of the optimum value on meteorological parameters for 1 m2 of collector surface.
Modeling and simulation of CANDU reactor and its regulating system
Javidnia, Hooman
Analytical computer codes are indispensable tools in design, optimization, and control of nuclear power plants. Numerous codes have been developed to perform different types of analyses related to the nuclear power plants. A large number of these codes are designed to perform safety analyses. In the context of safety analyses, the control system is often neglected. Although there are good reasons for such a decision, that does not mean that the study of control systems in the nuclear power plants should be neglected altogether. In this thesis, a proof of concept code is developed as a tool that can be used in the design. optimization. and operation stages of the control system. The main objective in the design of this computer code is providing a tool that is easy to use by its target audience and is capable of producing high fidelity results that can be trusted to design the control system and optimize its performance. Since the overall plant control system covers a very wide range of processes, in this thesis the focus has been on one particular module of the the overall plant control system, namely, the reactor regulating system. The center of the reactor regulating system is the CANDU reactor. A nodal model for the reactor is used to represent the spatial neutronic kinetics of the core. The nodal model produces better results compared to the point kinetics model which is often used in the design and analysis of control system for nuclear reactors. The model can capture the spatial effects to some extent. although it is not as detailed as the finite difference methods. The criteria for choosing a nodal model of the core are: (1) the model should provide more detail than point kinetics and capture spatial effects, (2) it should not be too complex or overly detailed to slow down the simulation and provide details that are extraneous or unnecessary for a control engineer. Other than the reactor itself, there are auxiliary models that describe dynamics of different
Computer simulation of two-phase flow in nuclear reactors
International Nuclear Information System (INIS)
Wulff, W.
1993-01-01
Two-phase flow models dominate the requirements of economic resources for the development and use of computer codes which serve to analyze thermohydraulic transients in nuclear power plants. An attempt is made to reduce the effort of analyzing reactor transients by combining purpose-oriented modelling with advanced computing techniques. Six principles are presented on mathematical modeling and the selection of numerical methods, along with suggestions on programming and machine selection, all aimed at reducing the cost of analysis. Computer simulation is contrasted with traditional computer calculation. The advantages of run-time interactive access operation in a simulation environment are demonstrated. It is explained that the drift-flux model is better suited than the two-fluid model for the analysis of two-phase flow in nuclear reactors, because of the latter's closure problems. The advantage of analytical over numerical integration is demonstrated. Modeling and programming techniques are presented which minimize the number of needed arithmetical and logical operations and thereby increase the simulation speed, while decreasing the cost. (orig.)
International Nuclear Information System (INIS)
El-Osery, I.A.
1981-01-01
The purpose of this paper is to discuss the theories, techniques and computer codes that are frequently used in numerical reactor criticality and burnup calculations. It is a part of an integrated nuclear reactor calculation scheme conducted by the Reactors Department, Inshas Nuclear Research Centre. The crude part in numerical reactor criticality and burnup calculations includes the determination of neutron flux distribution which can be obtained in principle as a solution of Boltzmann transport equation. Numerical methods used for solving transport equations are discussed. Emphasis are made on numerical techniques based on multigroup diffusion theory. These numerical techniques include nodal, modal, and finite difference ones. The most commonly known computer codes utilizing these techniques are reviewed. Some of the main computer codes that have been already developed at the Reactors Department and related to numerical reactor criticality and burnup calculations have been presented
Energy Technology Data Exchange (ETDEWEB)
Muramatsu, Toshiharu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center
1998-08-01
This report explains the numerical methods and the set-up method of input data for a single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3 (Direct Numerical Simulation using a 3rd-order upwind scheme). The code was developed to simulate non-stationary temperature fluctuation phenomena related to thermal striping phenomena, developed at Power Reactor and Nuclear Fuel Development Corporation (PNC). The DINUS-3 code was characterized by the use of a third-order upwind scheme for convection terms in instantaneous Navier-Stokes and energy equations, and an adaptive control system based on the Fuzzy theory to control time step sizes. Author expect this report is very useful to utilize the DINUS-3 code for the evaluation of various non-stationary thermohydraulic phenomena in reactor applications. (author)
Developments in numerical simulation of IFE target and chamber physics
International Nuclear Information System (INIS)
Velarde, G.; Minguez, E.; Alonso, E.; Gil, J.M.; Malerba, L.; Marian, J.; Martel, P.; Martinez-Val, J.M.; Munoz, R.; Ogando, F.; Perlado, J.M.; Piera, M.; Reyes, S.; Rubiano, J.G.; Sanz, J.; Sauvan, P.; Velarde, M.; Velarde, P.
2000-01-01
The work presented outlines the global frame given at the Institute of Nuclear Fusion (DENIM) for having an integral perspective of the different research areas with the development of Inertial Fusion for energy generation. The coupling of a new radiation transport (RT) solver with an existing multi-material fluid dynamics code using Adaptive Mesh Refinement (ARM) is presented in Section 2, including improvements and additional information about the solver precision. In Section 3, new developments in the atomic physics codes under target conditions, to determine populations, opacity data and emissivities have been performed. Exotic and innovative ideas about Inertial Fusion Energy (IFE), as catalytic fuels and Z-pinches have been explored, and they are explained in Section 4. Numerical simulations demonstrate important reductions in the tritium inventory. Section 5 is devoted to safety and environment of the IFE. Uncertainties analysis in activation calculations have been included in the ACAB activation code, and also calculations on pulse activation in IFE reactors and on the activation of target debris in NIF are presented. A comparison of the accidental releases of tritium from some IFE reactors computed using MACCS2 code is explained. Finally, Section 6 contains the research on the basic mechanisms of neutron damage in SiC (low-activation material) and FeCu alloy using the DENIM/LLNL molecular dynamics code MDCASK. (authors)
Transonic aeroelastic numerical simulation in aeronautical engineering
International Nuclear Information System (INIS)
Yang, G.
2005-01-01
An LU-SGS (lower-upper symmetric Gauss-Seidel) subiteration scheme is constructed for time-marching of the fluid equations. The HLLEW (Harten-Lax-van Leer-Einfeldt-Wada) scheme is used for the spatial discretization. The same subiteration formulation is applied directly to the structural equations of motion in generalized coordinates. Through subiteration between the fluid and structural equations, a fully implicit aeroelastic solver is obtained for the numerical simulation of fluid/structure interaction. To improve the ability for application to complex configurations, a multiblock grid is used for the flow field calculation and Transfinite Interpolation (TFI) is employed for the adaptive moving grid deformation. The infinite plate spline (IPS) and the principal of virtual work are utilized for the data transformation between the fluid and structure. The developed code was first validated through the comparison of experimental and computational results for the AGARD 445.6 standard aeroelastic wing. Then the flutter character of a tail wing with control surface was analyzed. Finally, flutter boundaries of a complex aircraft configuration were predicted. (author)
Proton decay: Numerical simulations confront grand unification
International Nuclear Information System (INIS)
Brower, R.C.; Maturana, G.; Giles, R.C.; Moriarty, K.J.M.; Samuel, S.
1985-01-01
The Grand Unified Theories of the electromagnetic, weak and strong interactions constitute a far reaching attempt to synthesize our knowledge of theoretical particle physics into a consistent and compelling whole. Unfortunately, many quantitative predictions of such unified theories are sensitive to the analytically intractible effects of the strong subnuclear theory (Quantum Chromodynamics or QCD). The consequence is that even ambitious experimental programs exploring weak and super-weak interaction effects often fail to give definitive theoretical tests. This paper describes large-scale calculations on a supercomputer which can help to overcome this gap between theoretical predictions and experimental results. Our focus here is on proton decay, though the methods described are useful for many weak processes. The basic algorithms for the numerical simulation of QCD are well known. We will discuss the advantages and challenges of applying these methods to weak transitions. The algorithms require a very large data base with regular data flow and are natural candidates for vectorization. Also, 32-bit floating point arithmetic is adequate. Thus they are most naturally approached using a supercomputer alone or in combination with a dedicated special purpose processor. (orig.)
FRESCO: fusion reactor simulation code for tokamaks
International Nuclear Information System (INIS)
Mantsinen, M.J.
1995-03-01
The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)
Pressurized water reactor simulator. Workshop material. 2. ed
International Nuclear Information System (INIS)
2005-01-01
The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator
Numerical investigation of heat transfer in high-temperature gas-cooled reactors
Energy Technology Data Exchange (ETDEWEB)
Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)
1995-09-01
This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.
International Nuclear Information System (INIS)
Zhu Huanjun; Xu Yijun
2014-01-01
The temperature fluctuation in core outlet region of China Experimental Fast Reactor (CEFR) was numerically simulated by the CFD software Star CCM+. With the core outlet temperatures, flows etc. under rated conditions given as boundary conditions, a 1/4 region model of the reactor core outlet region was established and calculated using LES method for this problem. The analysis results show that while CEFR operates under rated conditions, the temperature fluctuation in lower part of core outlet region is mainly concentrated in area over the edge components (steel components, control rod assembly), and one in upper part is remarkable in area above all the components. The largest fluctuation amplitude is 19 K and the remarkable frequency is below 5 Hz, and it belongs to typically low frequency fluctuation. The conclusion is useful for further experimental work. (authors)
A Numerical Simulation for a Deterministic Compartmental ...
African Journals Online (AJOL)
In this work, an earlier deterministic mathematical model of HIV/AIDS is revisited and numerical solutions obtained using Eulers numerical method. Using hypothetical values for the parameters, a program was written in VISUAL BASIC programming language to generate series for the system of difference equations from the ...
Numerical simulation of pulse-tube refrigerators
Lyulina, I.A.; Mattheij, R.M.M.; Tijsseling, A.S.; Waele, de A.T.A.M.
2004-01-01
A new numerical model has been introduced to study steady oscillatory heat and mass transfer in the tube section of a pulse-tube refrigerator. Conservation equations describing compressible gas flow in the tube are solved numerically, using high resolution schemes. The equation of conservation of
Large-scale numerical simulations of plasmas
International Nuclear Information System (INIS)
Hamaguchi, Satoshi
2004-01-01
The recent trend of large scales simulations of fusion plasma and processing plasmas is briefly summarized. Many advanced simulation techniques have been developed for fusion plasmas and some of these techniques are now applied to analyses of processing plasmas. (author)
Research nuclear reactor start-up simulator
International Nuclear Information System (INIS)
Sofo Haro, M.; Cantero, P.
2009-01-01
This work presents the design and FPGA implementation of a research nuclear reactor start-up simulator. Its aim is to generate a set of signals that allow replacing the neutron detector for stimulated signals, to feed the measurement electronic of the start-up channels, to check its operation, together with the start-up security logic. The simulator presented can be configured on three independent channels and adjust the shape of the output pulses. Furthermore, each channel can be configured in 'rate' mode, where you can specify the growth rate of the pulse frequency in %/s. Result and details of the implementation on FPGA of the different functional blocks are given. (author)
Thermohydraulic modeling and simulation of breeder reactors
International Nuclear Information System (INIS)
Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.
1982-01-01
This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed
Neutronic/Thermal-hydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations
International Nuclear Information System (INIS)
Ragusa, Jean; Siegel, Andrew; Ruggieri, Jean-Michel
2010-01-01
The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.
Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations
Energy Technology Data Exchange (ETDEWEB)
Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri
2010-09-28
The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.
Computer code for simulating pressurized water reactor core
International Nuclear Information System (INIS)
Serrano, A.M.B.
1978-01-01
A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)
Development of Pelton turbine using numerical simulation
Energy Technology Data Exchange (ETDEWEB)
Patel, K; Patel, B; Yadav, M [Hydraulic Engineer, ALSTOM Hydro R and D India Ltd., GIDC Maneja, Vadodara - 390 013, Gujarat (India); Foggia, T, E-mail: patel@power.alstom.co [Hydraulic Engineer, Alstom Hydro France, Etablissement de Grenoble, 82, avenue Leon Blum BP 75, 38041 Grenoble Cedex (France)
2010-08-15
This paper describes recent research and development activities in the field of Pelton turbine design. Flow inside Pelton turbine is most complex due to multiphase (mixture of air and water) and free surface in nature. Numerical calculation is useful to understand flow physics as well as effect of geometry on flow. The optimized design is obtained using in-house special optimization loop. Either single phase or two phase unsteady numerical calculation could be performed. Numerical results are used to visualize the flow pattern in the water passage and to predict performance of Pelton turbine at full load as well as at part load. Model tests are conducted to determine performance of turbine and it shows good agreement with numerically predicted performance.
Development of Pelton turbine using numerical simulation
Patel, K.; Patel, B.; Yadav, M.; Foggia, T.
2010-08-01
This paper describes recent research and development activities in the field of Pelton turbine design. Flow inside Pelton turbine is most complex due to multiphase (mixture of air and water) and free surface in nature. Numerical calculation is useful to understand flow physics as well as effect of geometry on flow. The optimized design is obtained using in-house special optimization loop. Either single phase or two phase unsteady numerical calculation could be performed. Numerical results are used to visualize the flow pattern in the water passage and to predict performance of Pelton turbine at full load as well as at part load. Model tests are conducted to determine performance of turbine and it shows good agreement with numerically predicted performance.
Coherent Structures in Numerically Simulated Plasma Turbulence
DEFF Research Database (Denmark)
Kofoed-Hansen, O.; Pécseli, H.L.; Trulsen, J.
1989-01-01
Low level electrostatic ion acoustic turbulence generated by the ion-ion beam instability was investigated numerically. The fluctuations in potential were investigated by a conditional statistical analysis revealing propagating coherent structures having the form of negative potential wells which...
Development of Pelton turbine using numerical simulation
International Nuclear Information System (INIS)
Patel, K; Patel, B; Yadav, M; Foggia, T
2010-01-01
This paper describes recent research and development activities in the field of Pelton turbine design. Flow inside Pelton turbine is most complex due to multiphase (mixture of air and water) and free surface in nature. Numerical calculation is useful to understand flow physics as well as effect of geometry on flow. The optimized design is obtained using in-house special optimization loop. Either single phase or two phase unsteady numerical calculation could be performed. Numerical results are used to visualize the flow pattern in the water passage and to predict performance of Pelton turbine at full load as well as at part load. Model tests are conducted to determine performance of turbine and it shows good agreement with numerically predicted performance.
Numerical simulation of single bubble boiling behavior
Directory of Open Access Journals (Sweden)
Junjie Liu
2017-06-01
Full Text Available The phenomena of a single bubble boiling process are studied with numerical modeling. The mass, momentum, energy and level set equations are solved using COMSOL multi-physics software. The bubble boiling dynamics, the transient pressure field, velocity field and temperature field in time are analyzed, and reasonable results are obtained. The numeral model is validated by the empirical equation of Fritz and could be used for various applications.
Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles
Energy Technology Data Exchange (ETDEWEB)
Shang Zhi, E-mail: shangzhi@tsinghua.org.c [Faculty of Engineering, Kingston University, London SW15 3DW (United Kingdom); Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.co [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)
2010-04-15
The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90 deg. the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.
Modular numerical tool for gas turbine simulation
Sampedro Casis, Rodrigo
2015-01-01
In this work a free tool for the simulation of turboprops was implemented, capable of simulating the various components of a jet engine, separately or in conjunction, with different degrees of thermodynamic modelling or complexity, in order to simulate an entire jet engine. The main characteristics of this software includes its compatibility, open code and GNU license, non-existing in today's market. Furthermore, the tool was designed with a greater flexibility and a more adapted work environ...
International Nuclear Information System (INIS)
Maas, L.; Beuter, A.; Seropian, C.
2010-01-01
The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. Among its duties, one important item is to provide help for emergency situations management in case of an accident occurring in a French nuclear facility. In this framework, IRSN develops and applies numerical tools dealing with containment management issues. Up to now IRSN has not got any specific tool for experimental reactors. Accordingly, it has been then decided to extend the ASTEC code, devoted to severe accident scenarios for Pressurized Water Reactors, to this kind of reactors. This lumped-parameter code, co-developed by IRSN and GRS (Germany), covers the entire phenomenology from the initiating event up to fission products release outside the reactor containment, except for the steam explosion and the mechanical integrity of the containment. A first application to experimental reactors was carried out to assess the High Flux Reactor (HFR) operator's improvement proposal concerning the containment management during accidental situations. This reactor, located in Grenoble (France), is composed of a double wall containment with a pressurized containment annulus preventing any direct leakage into the environment. Until now, in case of severe accidents (mainly core melting in pool, explosive reactivity accident called BORAX), the HFR emergency management consisted in isolating the containment building in the early stage of the accident, to prevent any radioactive products release to the environment. The operator decided to improve this containment management during accidental situations by using an air filtering venting system able to maintain a slight sub-atmospheric pressure in the reactor building. The operator's demonstration of the efficiency of this new system is mainly based on containment pressure evaluations during accidental transients. IRSN assessed these calculations through ASTEC calculations. Finally, a global agreement was
Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment
Energy Technology Data Exchange (ETDEWEB)
Prabhudharwadkar, Deoras M. [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Iyer, Kannan N., E-mail: kiyer@iitb.ac.i [Department of Mechanical Engineering, Indian Institute of Technology, Mumbai (India); Mohan, Nalini; Bajaj, Satinder S. [Nuclear Power Corporation of India Ltd., Mumbai (India); Markandeya, Suhas G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India)
2011-03-15
Research highlights: This work addresses hydrogen dispersion in commercial nuclear reactor containment. The numerical tool used for simulation is first benchmarked with experimental data. Parametric results are then carried out for different release configurations. Results lead to the conclusion that the dispersal is buoyancy dominated. Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most
Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment
International Nuclear Information System (INIS)
Prabhudharwadkar, Deoras M.; Iyer, Kannan N.; Mohan, Nalini; Bajaj, Satinder S.; Markandeya, Suhas G.
2011-01-01
Research highlights: → This work addresses hydrogen dispersion in commercial nuclear reactor containment. → The numerical tool used for simulation is first benchmarked with experimental data. → Parametric results are then carried out for different release configurations. → Results lead to the conclusion that the dispersal is buoyancy dominated. → Also, the hydrogen concentration is high enough to demand mitigation devices. - Abstract: The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is
Numerical simulation of TIG welding with filler of steel pieces of high thickness
International Nuclear Information System (INIS)
Carmignani, B.; Toselli, G.
1999-01-01
The problem of the numerical simulation of welding process with filler, in particular TIG (tungsten inert gas) with cold filler, has been approached with ABAQUS/S code. Reference has been made to some experimental models studied and prepared ad hoc in order to better know the physical phenomena involved in the TIG welding technique and to validate the computation methodologies and results obtained. This numerical simulation has been required in order to assist the fabrication development and QA for TF (toroidal field) coil case, an important component of ITER (international thermonuclear experimental reactor) machine [it
Applicability of numerical simulation code TPFIT to two-phase flow in Venturi scrubber
International Nuclear Information System (INIS)
Horiguchi, Naoki; Kanagawa, Tetsuya; Kaneko, Akiko; Abe, Yutaka; Yoshida, Hiroyuki
2015-01-01
As one of the filtered venting devices for light water reactor, Venturi scrubber can operate with effective decontamination efficiency because dispersed flow is formed in the Venturi scrubber by pressure difference between inside and outside of holes for liquid suction. Droplet diameter and its distribution in cross-section area are important for the decontamination. However, they are changed by hydraulic behavior of suctioned liquid until atomized, and kinds of atomization phenomena. In this report, to understand the hydraulic behavior of the liquid in detail for the filtered venting, we performed visualized observation experimentally and numerical simulation by TPFIT. Then the numerical simulation result was validated by the experimental data. (author)
Numerical simulation of turbulent combustion: Scientific challenges
Ren, ZhuYin; Lu, Zhen; Hou, LingYun; Lu, LiuYan
2014-08-01
Predictive simulation of engine combustion is key to understanding the underlying complicated physicochemical processes, improving engine performance, and reducing pollutant emissions. Critical issues as turbulence modeling, turbulence-chemistry interaction, and accommodation of detailed chemical kinetics in complex flows remain challenging and essential for high-fidelity combustion simulation. This paper reviews the current status of the state-of-the-art large eddy simulation (LES)/prob-ability density function (PDF)/detailed chemistry approach that can address the three challenging modelling issues. PDF as a subgrid model for LES is formulated and the hybrid mesh-particle method for LES/PDF simulations is described. Then the development need in micro-mixing models for the PDF simulations of turbulent premixed combustion is identified. Finally the different acceleration methods for detailed chemistry are reviewed and a combined strategy is proposed for further development.
Efficient algorithms for flow simulation related to nuclear reactor safety
International Nuclear Information System (INIS)
Gornak, Tatiana
2013-01-01
Safety analysis is of ultimate importance for operating Nuclear Power Plants (NPP). The overall modeling and simulation of physical and chemical processes occuring in the course of an accident is an interdisciplinary problem and has origins in fluid dynamics, numerical analysis, reactor technology and computer programming. The aim of the study is therefore to create the foundations of a multi-dimensional non-isothermal fluid model for a NPP containment and software tool based on it. The numerical simulations allow to analyze and predict the behavior of NPP systems under different working and accident conditions, and to develop proper action plans for minimizing the risks of accidents, and/or minimizing the consequences of possible accidents. A very large number of scenarios have to be simulated, and at the same time acceptable accuracy for the critical parameters, such as radioactive pollution, temperature, etc., have to be achieved. The existing software tools are either too slow, or not accurate enough. This thesis deals with developing customized algorithm and software tools for simulation of isothermal and non-isothermal flows in a containment pool of NPP. Requirements to such a software are formulated, and proper algorithms are presented. The goal of the work is to achieve a balance between accuracy and speed of calculation, and to develop customized algorithm for this special case. Different discretization and solution approaches are studied and those which correspond best to the formulated goal are selected, adjusted, and when possible, analysed. Fast directional splitting algorithm for Navier-Stokes equations in complicated geometries, in presence of solid and porous obstacles, is in the core of the algorithm. Developing suitable pre-processor and customized domain decomposition algorithms are essential part of the overall algorithm amd software. Results from numerical simulations in test geometries and in real geometries are presented and discussed.
Nuclear reactor core modelling in multifunctional simulators
International Nuclear Information System (INIS)
Puska, E.K.
1999-01-01
The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been
Nuclear reactor core modelling in multifunctional simulators
Energy Technology Data Exchange (ETDEWEB)
Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)
1999-06-01
The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been
CFD simulation on reactor flow mixing phenomena
International Nuclear Information System (INIS)
Kwon, T.S.; Kim, K.H.
2016-01-01
A pre-test calculation for multi-dimensional flow mixing in a reactor core and downcomer has been studied using a CFD code. To study the effects of Reactor Coolant Pump (RCP) and core zone on the boron mixing behaviors in a lower downcomer and core inlet, a 1/5-scale CFD model of flow mixing test facility for the APR+ reference plant was simulated. The flow paths of the 1/5-scale model were scaled down by the linear scaling method. The aspect ratio (L/D) of all flow paths was preserved to 1. To preserve a dynamic similarity, the ratio of Euler number was also preserved to 1. A single phase water flow at low pressure and temperature conditions was considered in this calculation. The calculation shows that the asymmetric effect driven by RCPs shifted the high velocity field to the failed pump's flow zone. The borated water flow zone at the core inlet was also shifted to the failed RCP side. (author)
Detailed numerical simulations of laser cooling processes
Ramirez-Serrano, J.; Kohel, J.; Thompson, R.; Yu, N.
2001-01-01
We developed a detailed semiclassical numerical code of the forces applied on atoms in optical and magnetic fields to increase the understanding of the different roles that light, atomic collisions, background pressure, and number of particles play in experiments with laser cooled and trapped atoms.
International Nuclear Information System (INIS)
Yoshida, Hiroyuki; Takase, Kazuyuki
2008-01-01
Thermal-hydraulic design of the current boiling water reactor (BWR) is performed with the subchannel analysis codes which incorporated the correlations based on empirical results including actual-size tests. Then, for the Innovative Water Reactor for Flexible Fuel Cycle (FLWR) core, an actual size test of an embodiment of its design is required to confirm or modify such correlations. In this situation, development of a method that enables the thermal-hydraulic design of nuclear reactors without these actual size tests is desired, because these tests take a long time and entail great cost. For this reason, we developed an advanced thermal-hydraulic design method for FLWRs using innovative two-phase flow simulation technology. In this study, a detailed Two-Phase Flow simulation code using advanced Interface Tracking method: TPFIT is developed to calculate the detailed information of the two-phase flow. In this paper, firstly, we tried to verify the TPFIT code by comparing it with the existing 2-channel air-water mixing experimental results. Secondary, the TPFIT code was applied to simulation of steam-water two-phase flow in a model of two subchannels of a current BWRs and FLWRs rod bundle. The fluid mixing was observed at a gap between the subchannels. The existing two-phase flow correlation for fluid mixing is evaluated using detailed numerical simulation data. This data indicates that pressure difference between fluid channels is responsible for the fluid mixing, and thus the effects of the time average pressure difference and fluctuations must be incorporated in the two-phase flow correlation for fluid mixing. When inlet quality ratio of subchannels is relatively large, it is understood that evaluation precision of the existing two-phase flow correlations for fluid mixing are relatively low. (author)
NUMERICAL SIMULATION AND MODELING OF UNSTEADY FLOW ...
African Journals Online (AJOL)
2014-06-30
Jun 30, 2014 ... objective of this study is to control the simulation of unsteady flows around structures. ... Aerospace, our results were in good agreement with experimental .... Two-Equation Eddy-Viscosity Turbulence Models for Engineering.
Numerical simulation of ion-surface interactions
International Nuclear Information System (INIS)
Hou, M.
1994-01-01
This paper, based on examples from the author's contribution, aims to illustrate the role of ballistic simulations of the interaction between an ion beam and a surface in the characterization of surface properties. Several aspects of the ion-surface interaction have been modelled to various levels of sophistication by computer simulation. Particular emphasis is given to the ion scattering in the impact mode, in the multiple scattering regime and at grazing incidence, as well as to the Auger emission resulting from electronic excitation. Some examples are then given in order to illustrate the use of the combination between simulation and experiment to study the ion-surface interaction and surface properties. Ion-induced Auger emission, the determination of potentials and of overlay structures are discusse. The possibility to tackle dynamical surface properties by menas of a combination between molecular dynamics, ballistic simulations and ion scattering measurements in then briefly discussed. (orig.)
A numerical simulation of a contrail
Energy Technology Data Exchange (ETDEWEB)
Levkov, L.; Boin, M.; Meinert, D. [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)
1997-12-31
The formation of a contrail from an aircraft flying near the tropopause is simulated using a three-dimensional mesoscale atmospheric model including a very complex scheme of parameterized cloud microphysical processes. The model predicted ice concentrations are in very good agreement with data measured during the International Cirrus Experiment (ICE), 1989. Sensitivity simulations were run to determine humidity forcing on the life time of contrails. (author) 4 refs.
A numerical simulation of a contrail
Energy Technology Data Exchange (ETDEWEB)
Levkov, L; Boin, M; Meinert, D [GKSS-Forschungszentrum Geesthacht GmbH, Geesthacht (Germany)
1998-12-31
The formation of a contrail from an aircraft flying near the tropopause is simulated using a three-dimensional mesoscale atmospheric model including a very complex scheme of parameterized cloud microphysical processes. The model predicted ice concentrations are in very good agreement with data measured during the International Cirrus Experiment (ICE), 1989. Sensitivity simulations were run to determine humidity forcing on the life time of contrails. (author) 4 refs.
Numerical simulation of liquid film flow on revolution surfaces with momentum integral method
International Nuclear Information System (INIS)
Bottoni Maurizio
2005-01-01
The momentum integral method is applied in the frame of safety analysis of pressure water reactors under hypothetical loss of coolant accident (LOCA) conditions to simulate numerically film condensation, rewetting and vaporization on the inner surface of pressure water reactor containment. From the conservation equations of mass and momentum of a liquid film arising from condensation of steam upon the inner of the containment during a LOCA in a pressure water reactor plant, an integro-differential equation is derived, referring to an arbitrary axisymmetric surface of revolution. This equation describes the velocity distribution of the liquid film along a meridian of a surface of revolution. From the integro-differential equation and ordinary differential equation of first order for the film velocity is derived and integrated numerically. From the velocity distribution the film thickness distribution is obtained. The solution of the enthalpy equation for the liquid film yields the temperature distribution on the inner surface of the containment. (authors)
Numerical simulations of rarefied gas flows in thin film processes
Dorsman, R.
2007-01-01
Many processes exist in which a thin film is deposited from the gas phase, e.g. Chemical Vapor Deposition (CVD). These processes are operated at ever decreasing reactor operating pressures and with ever decreasing wafer feature dimensions, reaching into the rarefied flow regime. As numerical
Conjugate heat transfer simulations of advanced research reactor fuel
Energy Technology Data Exchange (ETDEWEB)
Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.
2014-07-01
Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with
Numerical Simulation of the Kinetic Critical Nucleus
Sanada, Masaaki; Nishioka, Kazumi; Okada, Masahumi; Maksimov, Igor, L.
1997-01-01
Our main interest is to see whether the number density indicates a peak at the kinetically stable critical nucleus due to its kinetical stability. We have numerically calculated the time evolution of the number densities of clusters in the case of water vapor nucleation. We employ the condition in which the difference between the size of the thermodynamic crtitical nucleus and that of the kinetic one is appreciable. The results show that the peak does not appear in the number densities of clu...
Numerical simulation of hemorrhage in human injury
Chong, Kwitae; Jiang, Chenfanfu; Santhanam, Anand; Benharash, Peyman; Teran, Joseph; Eldredge, Jeff
2015-11-01
Smoothed Particle Hydrodynamics (SPH) is adapted to simulate hemorrhage in the injured human body. As a Lagrangian fluid simulation, SPH uses fluid particles as computational elements and thus mass conservation is trivially satisfied. In order to ensure anatomical fidelity, a three-dimensional reconstruction of a portion of the human body -here, demonstrated on the lower leg- is sampled as skin, bone and internal tissue particles from the CT scan image of an actual patient. The injured geometry is then generated by simulation of ballistic projectiles passing through the anatomical model with the Material Point Method (MPM) and injured vessel segments are identified. From each such injured segment, SPH is used to simulate bleeding, with inflow boundary condition obtained from a coupled 1-d vascular tree model. Blood particles interact with impermeable bone and skin particles through the Navier-Stokes equations and with permeable internal tissue particles through the Brinkman equations. The SPH results are rendered in post-processing for improved visual fidelity. The overall simulation strategy is demonstrated on several injury scenarios in the lower leg.
Numerical characteristics of quantum computer simulation
Chernyavskiy, A.; Khamitov, K.; Teplov, A.; Voevodin, V.; Voevodin, Vl.
2016-12-01
The simulation of quantum circuits is significantly important for the implementation of quantum information technologies. The main difficulty of such modeling is the exponential growth of dimensionality, thus the usage of modern high-performance parallel computations is relevant. As it is well known, arbitrary quantum computation in circuit model can be done by only single- and two-qubit gates, and we analyze the computational structure and properties of the simulation of such gates. We investigate the fact that the unique properties of quantum nature lead to the computational properties of the considered algorithms: the quantum parallelism make the simulation of quantum gates highly parallel, and on the other hand, quantum entanglement leads to the problem of computational locality during simulation. We use the methodology of the AlgoWiki project (algowiki-project.org) to analyze the algorithm. This methodology consists of theoretical (sequential and parallel complexity, macro structure, and visual informational graph) and experimental (locality and memory access, scalability and more specific dynamic characteristics) parts. Experimental part was made by using the petascale Lomonosov supercomputer (Moscow State University, Russia). We show that the simulation of quantum gates is a good base for the research and testing of the development methods for data intense parallel software, and considered methodology of the analysis can be successfully used for the improvement of the algorithms in quantum information science.
Numerical simulation of turbulent forced convection in liquid metals
International Nuclear Information System (INIS)
Vodret, S; Di Maio, D Vitale; Caruso, G
2014-01-01
In the frame of the future generation of nuclear reactors, liquid metals are foreseen to be used as a primary coolant. Liquid metals are characterized by a very low Prandtl number due to their very high heat diffusivity. As such, they do not meet the so-called Reynolds analogy which assumes a complete similarity between the momentum and the thermal boundary layers via the use of the turbulent Prandtl number. Particularly, in the case of industrial fluid-dynamic calculations where a resolved computation near walls could be extremely time consuming and could need very large computational resources, the use of the classical wall function approach could lead to an inaccurate description of the temperature profile close to the wall. The first aim of the present study is to investigate the ability of a well- established commercial code (ANSYS FLUENT v.14) to deal with this issue, validating a suitable expression for the turbulent Prandtl number. Moreover, a thermal wall-function developed at Universite Catholique de Louvain has been implemented in FLUENT and validated, overcoming the limits of the solver to define it directly. Both the resolved and unresolved approaches have been carried out for a channel flow case and assessed against available direct numerical and large eddy simulations. A comparison between the numerically evaluated Nusselt number and the main correlations available in the literature has been also carried out. Finally, an application of the proposed methodology to a typical sub-channel case has been performed, comparing the results with literature correlations for tube banks
Numerical simulation of baseflow modification due to effects of ...
African Journals Online (AJOL)
Numerical simulation of baseflow modification due to effects of sediment yield. ... Physically-based mathematical modelling affords the opportunity to look at this kind of interaction, which should be simulated by deterministic responses of both water and fluvial processes. In addition to simulating the streamflow and ...
Computer simulation of the NASA water vapor electrolysis reactor
Bloom, A. M.
1974-01-01
The water vapor electrolysis (WVE) reactor is a spacecraft waste reclamation system for extended-mission manned spacecraft. The WVE reactor's raw material is water, its product oxygen. A computer simulation of the WVE operational processes provided the data required for an optimal design of the WVE unit. The simulation process was implemented with the aid of a FORTRAN IV routine.
Computer simulation system of neural PID control on nuclear reactor
International Nuclear Information System (INIS)
Chen Yuzhong; Yang Kaijun; Shen Yongping
2001-01-01
Neural network proportional integral differential (PID) controller on nuclear reactor is designed, and the control process is simulated by computer. The simulation result show that neutral network PID controller can automatically adjust its parameter to ideal state, and good control result can be gotten in reactor control process
International Nuclear Information System (INIS)
Ito, Kei; Kunugi, Tomoaki; Ohshima, Hiroyuki
2008-01-01
An onset condition of gas entrainment (GE) due to free surface vortex has been studied to establish a design of sodium-cooled fast reactor with a higher coolant velocity than conventional designs. Numerous investigations have been conducted experimentally and theoretically; however, the universal onset condition of the GE has not been determined yet due to the nonlinear characteristics of the GE. Recently, we have been studying numerical simulation methods as a promising method to evaluate GE, instead of the reliable but costly real-scale tests. In this paper, the applicability of the numerical simulation methods to the evaluation of the GE is discussed. For the purpose, a quasi-steady vortex in a cylindrical tank and a wake vortex (unsteady vortex) in a rectangular channel were numerically simulated using the volume-of-fluid type two-phase flow calculation method. The simulated velocity distributions and free surface shapes of the quasi-steady vortex showed good (not perfect, however) agreements with experimental results when a fine mesh subdivision and a high-order discretization scheme were employed. The unsteady behavior of the wake vortex was also simulated with high accuracy. Although the onset condition of the GE was slightly underestimated in the simulation results, the applicability of the numerical simulation methods to the GE evaluation was confirmed. (author)
Numerical simulations of progressive hardening by using ABAQUS FEA software
Directory of Open Access Journals (Sweden)
Domański Tomasz
2018-01-01
Full Text Available The paper concerns numerical simulations of progressive hardening include phase transformations in solid state of steel. Abaqus FEA software is used for numerical analysis of temperature field and phase transformations. Numerical subroutines, written in fortran programming language are used in computer simulations where models of the distribution of movable heat source, kinetics of phase transformations in solid state as well as thermal and structural strain are implemented. Model for evaluation of fractions of phases and their kinetics is based on continuous heating diagram and continuous cooling diagram. The numerical analysis of thermal fields, phase fractions and strain associated progressive hardening of elements made of steel were done.
Numerical simulations of nanostructured gold films
DEFF Research Database (Denmark)
Repän, Taavi; Frydendahl, Christian; Novikov, Sergey M.
2017-01-01
We present an approach to analyse near-field effects on nanostructured gold films by finite element simulations. The studied samples are formed by fabricating gold films near the percolation threshold and then applying laser damage. Resulting samples have complicated structures, which...
IAEA activities in nuclear reactor simulation for educational purposes
International Nuclear Information System (INIS)
Lyon, R.B.
2001-01-01
The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Two simulation programs are presented at this workshop: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)
IAEA activities in nuclear reactors simulation for educational purposes
International Nuclear Information System (INIS)
Lyon, R.B.
1998-01-01
The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has two simulation programs: the Classroom-based Advanced Reactor Demonstrators package, and the Advanced Reactor Simulator. Both packages simulate the behaviour of BWR, PWR and HWR reactor types. For each package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer programs. (author)
Numerical simulation of cross field amplifiers
International Nuclear Information System (INIS)
Eppley, K.
1990-01-01
Cross field amplifiers (CFA) have been used in many applications where high power, high frequency microwaves are needed. Although these tubes have been manufactured for decades, theoretical analysis of their properties is not as highly developed as for other microwave devices such as klystrons. One feature distinguishing cross field amplifiers is that the operating current is produced by secondary emission from a cold cathode. This removes the need for a heater and enables the device to act as a switch tube, drawing no power until the rf drive is applied. However, this method of generating the current does complicate the simulation. We are developing a simulation model of cross field amplifiers using the PIC code CONDOR. We simulate an interaction region, one traveling wavelength long, with periodic boundary conditions. An electric field with the appropriate phase velocity is imposed on the upper boundary of the problem. Evaluation of the integral of E·J gives the power interchanged between the wave and the beam. Given the impedance of the structure, we then calculate the change in the traveling wave field. Thus we simulate the growth of the wave through the device. The main advance of our model over previous CFA simulations is the realistic tracking of absorption and secondary emission. The code uses experimental curves to calculate secondary production as a function of absorbed energy, with a theoretical expression for the angular dependence. We have used this code to model the 100 MW X-band CFA under construction at SLAC, as designed by Joseph Feinstein and Terry Lee. We are examining several questions of practical interest, such as the power and spectrum of absorbed electrons, the minimum traveling wave field needed to initiate spoke formation, and the variation of output power with dc voltage, anode-cathode gap, and magnetic field. 5 refs., 8 figs
Numerical simulation of avascular tumor growth
Energy Technology Data Exchange (ETDEWEB)
Slezak, D Fernandez; Suarez, C; Soba, A; Risk, M; Marshall, G [Laboratorio de Sistemas Complejos, Departamento de Computacion, Facultad de Ciencias Exactas y Naturales, Universidad de Buenos Aires (C1428EGA) Buenos Aires (Argentina)
2007-11-15
A mathematical and numerical model for the description of different aspects of microtumor development is presented. The model is based in the solution of a system of partial differential equations describing an avascular tumor growth. A detailed second-order numeric algorithm for solving this system is described. Parameters are swiped to cover a range of feasible physiological values. While previous published works used a single set of parameters values, here we present a wide range of feasible solutions for tumor growth, covering a more realistic scenario. The model is validated by experimental data obtained with a multicellular spheroid model, a specific type of in vitro biological model which is at present considered to be optimum for the study of complex aspects of avascular microtumor physiology. Moreover, a dynamical analysis and local behaviour of the system is presented, showing chaotic situations for particular sets of parameter values at some fixed points. Further biological experiments related to those specific points may give potentially interesting results.
Numerical simulation of two-phase flow behavior in Venturi scrubber by interface tracking method
Energy Technology Data Exchange (ETDEWEB)
Horiguchi, Naoki, E-mail: s1430215@u.tsukuba.ac.jp [Japan Atomic Energy Agency, 2-4, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); University of Tsukuba, 1-1-1, Tennodai, Tsukuba, Ibaraki, 305-8577 (Japan); Yoshida, Hiroyuki [Japan Atomic Energy Agency, 2-4, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Abe, Yutaka [University of Tsukuba, 1-1-1, Tennodai, Tsukuba, Ibaraki, 305-8577 (Japan)
2016-12-15
Highlights: • Self-priming occur because of pressure balance between inside and outside of throat is confirmed. • VS has similar flow with a Venturi tube except of disturbance and burble flow is considered. • Some of atomization simulated are validated qualitatively by comparison with previous studies. - Abstract: From the viewpoint of protecting a containment vessel of light water reactor and suppressing the diffusion of radioactive materials from a light water reactor, it is important to develop the device which allows a filtered venting of contaminated high pressure gas. In the filtered venting system that used in European reactors, so called Multi Venturi scrubbers System is used to realize filtered venting without any power supply. This system is able to define to be composed of Venturi scrubbers (VS) and a bubble column. In the VS, scrubbing of contaminated gas is promoted by both gas releases through the submerged VS and gas-liquid contact with splay flow formed by liquid suctioned through a hole provided by the pressure difference between inner and outer regions of a throat part of the VS. However, the scrubbing mechanism of the self-priming VS including effects of gas mass flow rate and shape of the VS are understood insufficiently in the previous studies. Therefore, we started numerical and experimental study to understand the detailed two-phase flow behavior in the VS. In this paper, to understand the VS operation characteristics for the filtered venting, we performed numerical simulations of two-phase flow behavior in the VS. In the first step of this study, we perform numerical simulations of supersonic flow by the TPFIT to validate the applicability of the TPFIT for high velocity flow like flow in the VS. In the second step, numerical simulation of two-phase flow behavior in the VS including self-priming phenomena. As the results, dispersed flow in the VS was reproduced in the numerical simulation, as same as the visualization experiments.
Numerical simulation of two-phase flow behavior in Venturi scrubber by interface tracking method
International Nuclear Information System (INIS)
Horiguchi, Naoki; Yoshida, Hiroyuki; Abe, Yutaka
2016-01-01
Highlights: • Self-priming occur because of pressure balance between inside and outside of throat is confirmed. • VS has similar flow with a Venturi tube except of disturbance and burble flow is considered. • Some of atomization simulated are validated qualitatively by comparison with previous studies. - Abstract: From the viewpoint of protecting a containment vessel of light water reactor and suppressing the diffusion of radioactive materials from a light water reactor, it is important to develop the device which allows a filtered venting of contaminated high pressure gas. In the filtered venting system that used in European reactors, so called Multi Venturi scrubbers System is used to realize filtered venting without any power supply. This system is able to define to be composed of Venturi scrubbers (VS) and a bubble column. In the VS, scrubbing of contaminated gas is promoted by both gas releases through the submerged VS and gas-liquid contact with splay flow formed by liquid suctioned through a hole provided by the pressure difference between inner and outer regions of a throat part of the VS. However, the scrubbing mechanism of the self-priming VS including effects of gas mass flow rate and shape of the VS are understood insufficiently in the previous studies. Therefore, we started numerical and experimental study to understand the detailed two-phase flow behavior in the VS. In this paper, to understand the VS operation characteristics for the filtered venting, we performed numerical simulations of two-phase flow behavior in the VS. In the first step of this study, we perform numerical simulations of supersonic flow by the TPFIT to validate the applicability of the TPFIT for high velocity flow like flow in the VS. In the second step, numerical simulation of two-phase flow behavior in the VS including self-priming phenomena. As the results, dispersed flow in the VS was reproduced in the numerical simulation, as same as the visualization experiments.
Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core
Energy Technology Data Exchange (ETDEWEB)
Yamaji, Bogdan; Aszodi, Attila [Budapest Univ. of Technology and Economics (Hungary). Inst. of Nuclear Techniques
2017-09-15
In the paper measurement results of experimental modelling of a molten salt fast reactor concept will be presented and compared with three-dimensional computational fluid dynamics (CFD) simulation results. Purpose of this article is twofold, on one hand to introduce a geometry modification in order to avoid the disadvantages of the original geometry and discuss new measurement results. On the other hand to present an analysis in order to suggest a method of proper numerical modelling of the problem based on the comparison of calculation results and measurement data for the new, modified geometry. The investigated concept has a homogeneous cylindrical core without any internal structures. Previous measurements on the scaled and segmented plexiglas model of the concept core and simulation results have shown that this core geometry could be optimized for better thermal-hydraulics characteristics. In case of the original geometry strong undesired flow separation could develop, that could negatively affect the characteristics of the core from neutronics point of view as well. An internal flow distributor plate was designed and installed with the purpose of optimizing the flow field in the core by enhancing its uniformity. Particle image velocimetry (PIV) measurement results of the modified experimental model will be presented and compared to numerical simulation results with the purpose of CFD model validation.
Numerical Simulation of 3-D Wave Crests
Institute of Scientific and Technical Information of China (English)
YU Dingyong; ZHANG Hanyuan
2003-01-01
A clear definition of 3-D wave crest and a description of the procedures to detect the boundary of wave crest are presented in the paper. By using random wave theory and directional wave spectrum, a MATLAB-platformed program is designed to simulate random wave crests for various directional spectral conditions in deep water. Statistics of wave crests with different directional spreading parameters and different directional functions are obtained and discussed.
Numerical evaluation of flow through a prismatic very high temperature gas-cooled reactor
International Nuclear Information System (INIS)
Barros Filho, Jose A.; Santos, Andre A.C.; Navarro, Moyses A.; Ribeiro, Felipe Lopes
2011-01-01
The High-temperature Gas-cooled reactor (HTGR) is a Next Generation Nuclear System that has a good chance to be used as energy generation source in the near future owing to its potential capacity to supply hydrogen without greenhouse gas emission for the future humanity. Recently, improvements in the HTGR design led to the Very High Temperature Reactor (VHTR) concept in which the outlet temperature of the coolant gas reaches to 1000 deg C increasing the efficiency of the hydrogen and electricity generation. Among the core concepts emerging in the VHTR development stands out the prismatic block which uses coated fuel microspheres named TRISO pressed into cylinders and assembled in hexagonal graphite blocks staked to form columns. The graphite blocks contain flow channels around the fuel cylinders for the helium coolant. In this study an analysis is performed using the CFD code CFX 13.0 on a prismatic fuel assembly in order to investigate its thermo-fluid dynamic performance. The simulations were made in a 1/12 fuel element model of the GT-MHR design which was developed by General Atomics. A numerical mesh verification process based on the Grid Convergence Index (GCI) was performed using five progressively refined meshes to assess the numerical uncertainty of the simulation and determine adequate mesh parameters. An analysis was also performed to evaluate different methods to define the inlet and outlet boundary conditions. In this study simulations of models with and without inlet and outlet plena were compared, showing that the presence of the plena offers a more realistic flow distribution. (author)
IAEA activities in nuclear reactor simulation for educational purposes
International Nuclear Information System (INIS)
Badulescu, A.; Lyon, R.
2001-01-01
The International Atomic Energy Agency (IAEA) has established a programme in nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the supply or development of simulation programs and training material, sponsors training courses and workshops, and distributes documentation and computer programs. Currently, the IAEA has simulation programs available for distribution that simulate the behaviour of BWR, PWR and HWR reactor types. (authors)
Comparison of simulated and measured quantities of a duplex reactor
Energy Technology Data Exchange (ETDEWEB)
Koskela, M.; Kajava, M. [ABB Marine, Helsinki (Finland)
1997-12-31
The purpose of this article is to illustrate the use of an analog simulator as a design tool when designing new power electric equipment. The purpose of simulation is to predict the functionality of electrical equipment to be constructed. Duplex reactor is an electromagnetic device designed to reduce voltage harmonics and short circuit currents in the ship electrical network. In this report a comparison between simulated and measured electrical quantities of a duplex reactor has been made. The purpose of the measurements was to show the correct functioning of the reactor. The simulation results and the measured waveforms corresponds well to each other. (orig.) 4 refs.
Numerical analysis of magnetoelastic coupled buckling of fusion reactor components
International Nuclear Information System (INIS)
Demachi, K.; Yoshida, Y.; Miya, K.
1994-01-01
For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated
Numerical simulation of distributed parameter processes
Colosi, Tiberiu; Unguresan, Mihaela-Ligia; Muresan, Vlad
2013-01-01
The present monograph defines, interprets and uses the matrix of partial derivatives of the state vector with applications for the study of some common categories of engineering. The book covers broad categories of processes that are formed by systems of partial derivative equations (PDEs), including systems of ordinary differential equations (ODEs). The work includes numerous applications specific to Systems Theory based on Mpdx, such as parallel, serial as well as feed-back connections for the processes defined by PDEs. For similar, more complex processes based on Mpdx with PDEs and ODEs as components, we have developed control schemes with PID effects for the propagation phenomena, in continuous media (spaces) or discontinuous ones (chemistry, power system, thermo-energetic) or in electro-mechanics (railway – traction) and so on. The monograph has a purely engineering focus and is intended for a target audience working in extremely diverse fields of application (propagation phenomena, diffusion, hydrodyn...
Fluid dynamics theory, computation, and numerical simulation
Pozrikidis, C
2017-01-01
This book provides an accessible introduction to the basic theory of fluid mechanics and computational fluid dynamics (CFD) from a modern perspective that unifies theory and numerical computation. Methods of scientific computing are introduced alongside with theoretical analysis and MATLAB® codes are presented and discussed for a broad range of topics: from interfacial shapes in hydrostatics, to vortex dynamics, to viscous flow, to turbulent flow, to panel methods for flow past airfoils. The third edition includes new topics, additional examples, solved and unsolved problems, and revised images. It adds more computational algorithms and MATLAB programs. It also incorporates discussion of the latest version of the fluid dynamics software library FDLIB, which is freely available online. FDLIB offers an extensive range of computer codes that demonstrate the implementation of elementary and advanced algorithms and provide an invaluable resource for research, teaching, classroom instruction, and self-study. This ...
Partial Differential Equations Modeling and Numerical Simulation
Glowinski, Roland
2008-01-01
This book is dedicated to Olivier Pironneau. For more than 250 years partial differential equations have been clearly the most important tool available to mankind in order to understand a large variety of phenomena, natural at first and then those originating from human activity and technological development. Mechanics, physics and their engineering applications were the first to benefit from the impact of partial differential equations on modeling and design, but a little less than a century ago the Schrödinger equation was the key opening the door to the application of partial differential equations to quantum chemistry, for small atomic and molecular systems at first, but then for systems of fast growing complexity. Mathematical modeling methods based on partial differential equations form an important part of contemporary science and are widely used in engineering and scientific applications. In this book several experts in this field present their latest results and discuss trends in the numerical analy...
High accuracy mantle convection simulation through modern numerical methods
Kronbichler, Martin; Heister, Timo; Bangerth, Wolfgang
2012-01-01
Numerical simulation of the processes in the Earth's mantle is a key piece in understanding its dynamics, composition, history and interaction with the lithosphere and the Earth's core. However, doing so presents many practical difficulties related
Direct Numerical Simulations of Statistically Stationary Turbulent Premixed Flames
Im, Hong G.; Arias, Paul G.; Chaudhuri, Swetaprovo; Uranakara, Harshavardhana A.
2016-01-01
Direct numerical simulations (DNS) of turbulent combustion have evolved tremendously in the past decades, thanks to the rapid advances in high performance computing technology. Today’s DNS is capable of incorporating detailed reaction mechanisms
Experimental and numerical investigation of bubble column reactors
Bai, W.
2010-01-01
Due to various advantages, such as simple geometry, ease of operation, low operating and maintenance costs, excellent heat and mass transfer characteristics, bubble column reactors are frequently used in chemical, petrochemical, biochemical, pharmaceutical, metallurgical industries for a variety of
Direct Numerical Simulations of Turbulent Autoigniting Hydrogen Jets
Asaithambi, Rajapandiyan
Autoignition is an important phenomenon and a tool in the design of combustion engines. To study autoignition in a canonical form a direct numerical simulation of a turbulent autoigniting hydrogen jet in vitiated coflow conditions at a jet Reynolds number of 10,000 is performed. A detailed chemical mechanism for hydrogen-air combustion and non-unity Lewis numbers for species transport is used. Realistic inlet conditions are prescribed by obtaining the velocity eld from a fully developed turbulent pipe flow simulation. To perform this simulation a scalable modular density based method for direct numerical simulation (DNS) and large eddy simulation (LES) of compressible reacting flows is developed. The algorithm performs explicit time advancement of transport variables on structured grids. An iterative semi-implicit time advancement is developed for the chemical source terms to alleviate the chemical stiffness of detailed mechanisms. The algorithm is also extended from a Cartesian grid to a cylindrical coordinate system which introduces a singularity at the pole r = 0 where terms with a factor 1/r can be ill-defined. There are several approaches to eliminate this pole singularity and finite volume methods can bypass this issue by not storing or computing data at the pole. All methods however face a very restrictive time step when using a explicit time advancement scheme in the azimuthal direction (theta) where the cell sizes are of the order DelrDeltheta. We use a conservative finite volume based approach to remove the severe time step restriction imposed by the CFL condition by merging cells in the azimuthal direction. In addition, fluxes in the radial direction are computed with an implicit scheme to allow cells to be clustered along the jet's shear layer. This method is validated and used to perform the large scale turbulent reacting simulation. The resulting flame structure is found to be similar to a turbulent diusion flame but stabilized by autoignition at the
Conceptual design for simulator of irradiation test reactors
International Nuclear Information System (INIS)
Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko
2012-03-01
A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)
Numerical investigation of the flow at the pebble bed of the high temperature gas cooled reactors
International Nuclear Information System (INIS)
Costa, Franklin C.; Navarro, Moyses A.; Santos, Andre A.C.
2011-01-01
This paper presents a numerical investigation of the thermal and fluid dynamics among the fuel spheres and the cooling fluid, appearing in the core of pebble bed reactor (PBR-Peeble Bed Reactor) using the CFD-Computational Fluid Dynamics CFX 13.0. The paper presents the two analysis results. In the first phase it was considered two heat transfer models for the fuel spheres. In a model it was established volumetric load generation, with thermal conduction for both the fuel and coating. The other model prescribes a heat flux at the sphere surfaces. In this analysis, it was proceed two simulation in the two sphere arrangements, one considering the spheres in contact, and the other with 2 mm spacing between them. At the second analysis it was evaluated the sphere arrangement influence on the thermal and fluid dynamic behavior of the bed. The four simulations present differences in the flow and in the surface and maximum temperature profiles of the coating.(author)
NUMERICAL METHODS FOR THE SIMULATION OF HIGH INTENSITY HADRON SYNCHROTRONS.
Energy Technology Data Exchange (ETDEWEB)
LUCCIO, A.; D' IMPERIO, N.; MALITSKY, N.
2005-09-12
Numerical algorithms for PIC simulation of beam dynamics in a high intensity synchrotron on a parallel computer are presented. We introduce numerical solvers of the Laplace-Poisson equation in the presence of walls, and algorithms to compute tunes and twiss functions in the presence of space charge forces. The working code for the simulation here presented is SIMBAD, that can be run as stand alone or as part of the UAL (Unified Accelerator Libraries) package.
Numerical simulation and physical aspects of supersonic vortex breakdown
Liu, C. H.; Kandil, O. A.; Kandil, H. A.
1993-01-01
Existing numerical simulations and physical aspects of subsonic and supersonic vortex-breakdown modes are reviewed. The solution to the problem of supersonic vortex breakdown is emphasized in this paper and carried out with the full Navier-Stokes equations for compressible flows. Numerical simulations of vortex-breakdown modes are presented in bounded and unbounded domains. The effects of different types of downstream-exit boundary conditions are studied and discussed.
Numerical simulation of exploding pusher targets
Atzeni, S.; Rosenberg, M. J.; Gatu Johnson, M.; Petrasso, R. D.
2017-10-01
Exploding pusher targets, i.e. gas-filled large aspect-ratio glass or plastic shells, driven by a strong laser-generated shock, are widely used as pulsed sources of neutrons and fast charged particles. Recent experiments on exploding pushers provided evidence for the transition from a purely fluid behavior to a kinetic one. Indeed, fluid models largely overpredict yield and temperature as the Knudsen number Kn (ratio of ion mean-free path to compressed gas radius) is comparable or larger than one. At Kn = 0.3 - 1, fluid codes reasonably estimate integral quantities as yield and neutron-averaged temperatures, but do not reproduce burn radii, burn profiles and DD/DHe3 yield ratio. This motivated a detailed simulation study of intermediate-Kn exploding pushers. We will show how simulation results depend on models for laser-interaction, electron conductivity (flux-limited local vs nonlocal), viscosity (physical vs artificial), and ion mixing. Work partially supported by Sapienza Project C26A15YTMA, Sapienza 2016 (n. 257584), and Eurofusion Project AWP17-ENR-IFE-CEA-01.
Development of an educational nuclear research reactor simulator
International Nuclear Information System (INIS)
Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim; Ashoub, Nagieb
2014-01-01
This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.
Development of an educational nuclear research reactor simulator
Energy Technology Data Exchange (ETDEWEB)
Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.
2014-12-15
This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.
Energy Technology Data Exchange (ETDEWEB)
NONE
2015-07-01
Digital instrumentation and controls system technique is being introduced in new constructed research reactor or life extension of older research reactor. Digital systems are easy to change and optimize but the validated process for them is required. Also, to reduce project risk or cost, we have to make it sure that configuration and control functions is right before the commissioning phase on research reactor. For this purpose, simulators have been widely used in developing control systems in automotive and aerospace industries. In these literatures, however, very few of these can be found regarding test on the control system of research reactor with simulator. Therefore, this paper proposes a simulation platform to verify the performance of RRS (Reactor Regulating System) for research reactor. This simulation platform consists of the reactor simulation model and the interface module. This simulation platform is applied to I and C upgrade project of TRIGA reactor, and many problems of RRS configuration were found and solved. And it proved that the dynamic performance testing based on simulator enables significant time saving and improves economics and quality for RRS in the system test phase. (authors)
Direct Numerical Simulations of turbulent flow in a driven cavity
Verstappen, R.; Wissink, J.G.; Cazemier, W.; Veldman, A.E.P.
Direct numerical simulations (DNS) of 2 and 3D turbulent flows in a lid-driven cavity have been performed. DNS are numerical solutions of the unsteady (here: incompressible) Navier-Stokes equations that compute the evolution of all dynamically significant scales of motion. In view of the large
Energy Technology Data Exchange (ETDEWEB)
Tauveron, N
2006-02-15
The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge
Energy Technology Data Exchange (ETDEWEB)
Tauveron, N
2006-02-15
The subject of the present work was to develop models able to simulate axial instabilities occurrence and development in multistage turbomachines. The construction of a 1D unsteady axisymmetric model of internal flow in a turbomachine (at the scale of the row) has followed different steps: generation of steady correlations, adapted to different regimes (off-design conditions, low mass flowrate, negative mass flow rate); building of a model able to describe transient behaviour; use of implicit time schemes adapted to long transients; validation of the model in comparison of experimental investigations, measurements and numerical results from the bibliography. This model is integrated in a numerical tool, which has the capacity to describe the gas dynamics in a complete circuit containing different elements (ducts, valves, plenums). Thus, the complete model can represent the coupling between local and global phenomena, which is a very important mechanism in axial instability occurrence and development. An elementary theory has also been developed, based on a generalisation of Greitzer's model. These models, which were validated on various configurations, have provided complementary elements for the validation of the complete model. They have also allowed a more comprehensive description of physical phenomena at stake in instability occurrence and development by quantifying various effects (inertia, compressibility, performance levels) and underlying the main phenomena (in particular the collapse and recovery kinetics of the plenum), which were the only retained in the final elementary theory. The models were first applied to academic configurations (compression system), and then to an innovative industrial project: a helium cooled fast nuclear reactor with a Brayton cycle. The use of the models have brought comprehensive elements to surge occurrence due to a break event. It has been shown that surge occurrence is highly dependent of break location and that surge
Numerical simulation of a precessing vortex breakdown
International Nuclear Information System (INIS)
Jochmann, P.; Sinigersky, A.; Hehle, M.; Schaefer, O.; Koch, R.; Bauer, H.-J.
2006-01-01
The objective of this work is to present the results of time-dependent numerical predictions of a turbulent symmetry breaking vortex breakdown in a realistic gas turbine combustor. The unsteady Reynolds-averaged Navier-Stokes (URANS) equations are solved by using the k-ε two-equation model as well as by a full second-order closure using the Reynolds stress model of Speziale, Sarkar and Gatski (SSG). The results for a Reynolds number of 5.2 x 10 4 , a swirl number of 0.52 and an expansion ratio of 5 show that the flow is emerging from the swirler as a spiral gyrating around a zone of strong recirculation which is also asymmetric and precessing. These flow structures which are typical for the spiral type (S-type) vortex breakdown have been confirmed by PIV and local LDA measurements in a corresponding experimental setup. Provided that high resolution meshes are employed the calculations with both turbulence models are capable to reproduce the spatial and temporal dynamics of the flow
Numerical simulation of superconducting accelerator magnets
Kurz, Stefan
2002-01-01
Modeling and simulation are key elements in assuring the fast and successful design of superconducting magnets. After a general introduction the paper focuses on electromagnetic field computations, which are an indipensable tool in the design process. A technique which is especially well suited for the accurate computation of magnetic fields in superconducting magnets is presented. This method couples Boundary Elements (BEM) which discretize the surface of the iron yoke and Finite Elements (FEM) for the modeling of the non linear interior of the yoke. The formulation is based on a total magnetic scalar potential throughout the whole problem domain. The results for a short dipole model are presented and compared to previous results, which have been obtained from a similar BEM-FEM coupled vector potential formulation. 10 Refs. --- 25 --- AN
Numerical simulation of aeolian sand ripples
International Nuclear Information System (INIS)
Kang Liqiang; Guo Liejin
2004-01-01
With a new horizontal saltation displacement vector, a model is implemented to simulate the initiation and evolution of aeolian sand ripples. In the model, saltation distance considers the effects of surface height and slope. A linear stability analysis is also carried out for formation of sand ripples. The results show that, the model can be able to successfully reproduce sand ripples which can increase in scale by merging of small ripples. The linear stability analysis indicates that sand ripples appear when the relaxation rate parameter is below a threshold value and wind strength parameter is larger than a critical value. The results also verified that the formation of sand ripples is a self-organization process
A numerical relativity scheme for cosmological simulations
Daverio, David; Dirian, Yves; Mitsou, Ermis
2017-12-01
Cosmological simulations involving the fully covariant gravitational dynamics may prove relevant in understanding relativistic/non-linear features and, therefore, in taking better advantage of the upcoming large scale structure survey data. We propose a new 3 + 1 integration scheme for general relativity in the case where the matter sector contains a minimally-coupled perfect fluid field. The original feature is that we completely eliminate the fluid components through the constraint equations, thus remaining with a set of unconstrained evolution equations for the rest of the fields. This procedure does not constrain the lapse function and shift vector, so it holds in arbitrary gauge and also works for arbitrary equation of state. An important advantage of this scheme is that it allows one to define and pass an adaptation of the robustness test to the cosmological context, at least in the case of pressureless perfect fluid matter, which is the relevant one for late-time cosmology.
Glucose isomerization in simulated moving bed reactor by Glucose isomerase
Directory of Open Access Journals (Sweden)
Eduardo Alberto Borges da Silva
2006-05-01
Full Text Available Studies were carried out on the production of high-fructose syrup by Simulated Moving Bed (SMB technology. A mathematical model and numerical methodology were used to predict the behavior and performance of the simulated moving bed reactors and to verify some important aspects for application of this technology in the isomerization process. The developed algorithm used the strategy that considered equivalences between simulated moving bed reactors and true moving bed reactors. The kinetic parameters of the enzymatic reaction were obtained experimentally using discontinuous reactors by the Lineweaver-Burk technique. Mass transfer effects in the reaction conversion using the immobilized enzyme glucose isomerase were investigated. In the SMB reactive system, the operational variable flow rate of feed stream was evaluated to determine its influence on system performance. Results showed that there were some flow rate values at which greater purities could be obtained.Neste trabalho a tecnologia de Leito Móvel Simulado (LMS reativo é aplicada no processo de isomerização da glicose visando à produção de xarope concentrado de frutose. É apresentada a modelagem matemática e uma metodologia numérica para predizer o comportamento e o desempenho de unidades reativas de leito móvel simulado para verificar alguns aspectos importantes para o emprego desta tecnologia no processo de isomerização. O algoritmo desenvolvido utiliza a abordagem que considera as equivalências entre as unidades reativas de leito móvel simulado e leito móvel verdadeiro. Parâmetros cinéticos da reação enzimática são obtidos experimentalmente usando reatores em batelada pela técnica Lineweaver-Burk. Efeitos da transferência de massa na conversão de reação usando a enzima imobilizada glicose isomerase são verificados. No sistema reativo de LMS, a variável operacional vazão da corrente de alimentação é avaliada para conhecer o efeito de sua influência no
International Nuclear Information System (INIS)
Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han
2013-01-01
The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility
Energy Technology Data Exchange (ETDEWEB)
Kim, Hyoung Tae; Cha, Jae Eun Cha; Seo, Han [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2013-10-15
The Korea Atomic Energy Research Institute is constructing a Moderator Circulation Test (MCT) facility to simulate thermal-hydraulic phenomena in a 1/4 scale-down moderator tank similar to that in a prototype power plant during steady state operation and accident conditions. In the present study, two numerical CFD simulations for the prototype and scaled-down moderator tanks were carried out to check whether the moderator flow and temperature patterns of both the prototype reactor and scaled-down facility are identical. Two different sets of simulations of the moderator circulation phenomena were performed for a CANDU-6 reactor and MCT facility. The results of both simulations were compared to study the effects of scaling on the moderator flow and temperature patterns. There is no significant difference in the results between the prototype and scaled-down model. It was concluded that the present scaling method is properly employed to model the real reactor in the MCT facility.
International Nuclear Information System (INIS)
Emoto, M.; Suzuki, C.; Suzuki, Y.; Yokoyama, M.; Seki, R.; Ida, K.
2014-10-01
The enhanced-interlink system of experiment data and numerical simulation has been developed, and successfully operated routinely in the Large Helical Device (LHD). This system consists of analyzed diagnostic data, real-time coordinate mapping, and automatic data processing. It has enabled automated data handling/transferring between experiment and numerical simulation, to extensively perform experiment analyses. It can be considered as one of the prototypes for a seamless data-centric approach for integrating experiment data and numerical simulation/modellings in fusion experiments. Utilizing this system, experimental analyses by numerical simulations have extensively progressed. The authors believe this data-centric approach for integrating experiment data and numerical simulation/modellings will contribute to not only the LHD but to other plasma fusion projects including DEMO reactor in the future. (author)
Simulation tools and new developments of the molten salt fast reactor
International Nuclear Information System (INIS)
Merle-Lucotte, E.; Doligez, X.; Heuer, D.; Allibert, M.; Ghetta, V.
2010-01-01
Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, we have performed parametric studies in terms of safety coefficients, reprocessing requirements and breeding capabilities. In the frame of this major re-evaluation of the molten salt reactor (MSR), we have developed a new concept called Molten Salt Fast Reactor or MSFR, based on the Thorium fuel cycle and a fast neutron spectrum. This concept has been selected for further studies by the MSR steering committee of the Generation IV International Forum in 2009. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman's equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR's fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing capacities and the fertile or fissile alimentation. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system (reactor and reprocessing unit) and radioprotection issues. (authors)
Batman-cracks. Observations and numerical simulations
Selvadurai, A. P. S.; Busschen, A. Ten; Ernst, L. J.
1991-05-01
To ensure mechanical strength of fiber reinforced plastics (FRP), good adhesion between fibers and the matrix is considered to be an essential requirement. An efficient test of fiber-matrix interface characterization is the fragmentation test which provides information about the interface slip mechanism. This test consists of the longitudinal loading of a single fiber which is embedded in a matrix specimen. At critical loads the fiber experiences fragmentation. This fragmentation will terminate depending upon the shear-slip strength of the fiber-matrix adhesion, which is inversely proportional to average fragment lengths. Depending upon interface strength characteristics either bond or slip matrix fracture can occur at the onset of fiber fracture. Certain particular features of matrix fracture are observed at the locations of fiber fracture in situations where there is sufficient interface bond strength. These refer to the development of fractures with a complex surface topography. The experimental procedure involved in the fragmentation tests is discussed and the boundary element technique to examine the development of multiple matrix fractures at the fiber fracture locations is examined. The mechanics of matrix fracture is examined. When bond integrity is maintained, a fiber fracture results in a matrix fracture. The matrix fracture topography in a fragmentation test is complex; however, simplified conoidal fracture patterns can be used to investigate the crack extension phenomena. Via a mixed-mode fracture criterion, the generation of a conoidal fracture pattern in the matrix is investigated. The numerical results compare favorably with observed experimental data derived from tests conducted on fragmentation test specimens consisting of a single glass fiber which is embedded in a polyester matrix.
Reactor Simulations for Safeguards with the MCNP Utility for Reactor Evolution Code
International Nuclear Information System (INIS)
Shiba, T.; Fallot, M.
2015-01-01
To tackle nuclear material proliferation, we conducted several proliferation scenarios using the MURE (MCNP Utility for Reactor Evolution) code. The MURE code, developed by CNRS laboratories, is a precision, open-source code written in C++ that automates the preparation and computation of successive MCNP (Monte Carlo N-Particle) calculations and solves the Bateman equations in between, for burnup or thermal-hydraulics purposes. In addition, MURE has been completed recently with a module for the CHaracterization of Radioactive Sources, called CHARS, which computes the emitted gamma, beta and alpha rays associated to any fuel composition. Reactor simulations could allow knowing how plutonium or other material generation evolves inside reactors in terms of time and amount. The MURE code is appropriate for this purpose and can also provide knowledge on associated particle emissions. Using MURE, we have both developed a cell simulation of a typical CANDU reactor and a detailed model of light water PWR core, which could be used to analyze the composition of fuel assemblies as a function of time or burnup. MURE is also able to provide, thanks to its extension MURE-CHARTS, the emitted gamma rays from fuel assemblies unloaded from the core at any burnup. Diversion cases of Generation IV reactors have been also developed; a design of Very High Temperature Reactor (a Pebble Bed Reactor (PBR), loaded with UOx, PuOx and ThUOx fuels), and a Na-cooled Fast Breeder Reactor (FBR) (with depleted Uranium or Minor Actinides in the blanket). The loading of Protected Plutonium Production (P3) in the FBR was simulated. The simulations of various reactor designs taking into account reactor physics constraints may bring valuable information to inspectors. At this symposium, we propose to show the results of these reactor simulations as examples of the potentiality of reactor simulations for safeguards. (author)
International Nuclear Information System (INIS)
Okano, Yasushi; Ohira, Hiroaki
1998-08-01
In the early stage of sodium leak event of liquid metal fast breeder reactor, LMFBR, liquid sodium flows out from a piping, and ignition and combustion of liquid sodium droplet might occur under certain environmental condition. Compressible forced air flow, diffusion of chemical species, liquid sodium droplet behavior, chemical reactions and thermodynamic properties should be evaluated with considering physical dependence and numerical connection among them for analyzing combustion of sodium liquid droplet. A direct numerical simulation code was developed for numerical analysis of sodium liquid droplet in forced convection air flow. The numerical code named COMET, 'Sodium Droplet COmbustion Analysis METhodology using Direct Numerical Simulation in 3-Dimensional Coordinate'. The extended MAC method was used to calculate compressible forced air flow. Counter diffusion among chemical species is also calculated. Transport models of mass and energy between droplet and surrounding atmospheric air were developed. Equation-solving methods were used for computing multiphase equilibrium between sodium and air. Thermodynamic properties of chemical species were evaluated using dynamic theory of gases. Combustion of single sphere liquid sodium droplet in forced convection, constant velocity, uniform air flow was numerically simulated using COMET. Change of droplet diameter with time was closely agree with d 2 -law of droplet combustion theory. Spatial distributions of combustion rate and heat generation and formation, decomposition and movement of chemical species were analyzed. Quantitative calculations of heat generation and chemical species formation in spray combustion are enabled for various kinds of environmental condition by simulating liquid sodium droplet combustion using COMET. (author)
Numerical simulation of anisotropic polymeric foams
Directory of Open Access Journals (Sweden)
Volnei Tita
Full Text Available This paper shows in detail the modelling of anisotropic polymeric foam under compression and tension loadings, including discussions on isotropic material models and the entire procedure to calibrate the parameters involved. First, specimens of poly(vinyl chloride (PVC foam were investigated through experimental analyses in order to understand the mechanical behavior of this anisotropic material. Then, isotropic material models available in the commercial software AbaqusTM were investigated in order to verify their ability to model anisotropic foams and how the parameters involved can influence the results. Due to anisotropy, it is possible to obtain different values for the same parameter in the calibration process. The obtained set of parameters are used to calibrate the model according to the application of the structure. The models investigated showed minor and major limitations to simulate the mechanical behavior of anisotropic PVC foams under compression, tension and multi-axial loadings. Results show that the calibration process and the choice of the material model applied to the polymeric foam can provide good quantitative results and save project time. Results also indicate what kind and order of error one will get if certain choices are made throughout the modelling process. Finally, even though the developed calibration procedure is applied to specific PVC foam, it still outlines a very broad drill to analyze other anisotropic cellular materials.
Parallel Numerical Simulations of Water Reservoirs
Torres, Pedro; Mangiavacchi, Norberto
2010-11-01
The study of the water flow and scalar transport in water reservoirs is important for the determination of the water quality during the initial stages of the reservoir filling and during the life of the reservoir. For this scope, a parallel 2D finite element code for solving the incompressible Navier-Stokes equations coupled with scalar transport was implemented using the message-passing programming model, in order to perform simulations of hidropower water reservoirs in a computer cluster environment. The spatial discretization is based on the MINI element that satisfies the Babuska-Brezzi (BB) condition, which provides sufficient conditions for a stable mixed formulation. All the distributed data structures needed in the different stages of the code, such as preprocessing, solving and post processing, were implemented using the PETSc library. The resulting linear systems for the velocity and the pressure fields were solved using the projection method, implemented by an approximate block LU factorization. In order to increase the parallel performance in the solution of the linear systems, we employ the static condensation method for solving the intermediate velocity at vertex and centroid nodes separately. We compare performance results of the static condensation method with the approach of solving the complete system. In our tests the static condensation method shows better performance for large problems, at the cost of an increased memory usage. Performance results for other intensive parts of the code in a computer cluster are also presented.
NUMERICAL SIMULATION OF ICE ACCRETION ON AIRFOIL
Directory of Open Access Journals (Sweden)
Nicusor ALEXANDRESCU
2009-09-01
Full Text Available This work consists in the simulation of the ice accretion in the leading edge of aerodynamic profiles and our proposed model encompasses: geometry generation, calculation of the potential flow around the body, boundary layer thickness computation, water droplet trajectory computation, heat and mass balances and the consequent modification of the geometry by the ice growth. The flow calculation is realized with panel methods, using only segments defined over the body contour. The viscous effects are considered using the Karman-Pohlhausen method for the laminar boundary layer. The local heat transfer coefficient is obtained by applying the Smith-Spalding method for the thermal boundary layer. The ice accretion limits and the collection efficiency are determined by computing water droplet trajectories impinging the surface. The heat transfer process is analyzed with an energy and a mass balance in each segment defining the body. Finally, the geometry is modified by the addition of the computed ice thickness to the respective panel. The process by repeating all the steps. The model validation is done using a selection of problems with experimental solution, CIRA (the CESAR project. Hereinafter, results are obtained for different aerodynamic profiles, angles of attack and meteorological parameters
International Nuclear Information System (INIS)
Hohorst, J.K.; Allison, C.M.
2010-01-01
The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)
Energy Technology Data Exchange (ETDEWEB)
Hohorst, J.K.; Allison, C.M. [Innovative Systems Software, 1242 South Woodruff Avenue, Idaho Falls, Idaho 83404 (United States)
2010-07-01
The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of an international nuclear technology development program called SDTP (SCDAP Development and Training Program). SDTP involves more than 60 organizations in 28 countries. One of the important applications of the code is for simulator training of university faculty and students, reactor analysts, and reactor operations and technical support staff. Examples of RELAP/SCDAPSIM-based system thermal hydraulic and severe accident simulator packages include the SAFSIM simulator developed by NECSA for the SAFARI research reactor in South Africa, university-developed simulators at the University of Mexico and Shanghai Jiao Tong University in China, and commercial VISA and RELSIM packages used for analyst and reactor operations staff training. This paper will briefly describe the different packages/facilities. (authors)
Tests of numerical simulation algorithms for the Kubo oscillator
International Nuclear Information System (INIS)
Fox, R.F.; Roy, R.; Yu, A.W.
1987-01-01
Numerical simulation algorithms for multiplicative noise (white or colored) are tested for accuracy against closed-form expressions for the Kubo oscillator. Direct white noise simulations lead to spurious decay of the modulus of the oscillator amplitude. A straightforward colored noise algorithm greatly reduces this decay and also provides highly accurate results in the white noise limit
Numerical solutions to critical problem of reflected cylindrical reactor
International Nuclear Information System (INIS)
Horie, Junnosuke
1977-01-01
The multi-region critical problem can be transformed into an eigenvalue problem in the classical sense by using the method of Kuscer and Corngold and of Wing. This transformation is applied to derive a variational formulation for a reflected reactor. An approximate critical value of the multiplying factor is determined by maximizing the Rayleigh quotient for radially and totally reflected cylindrical reactors. It is shown that this approximate critical value is an upper bound of the true critical value. From the facts that the operator is self-adjoint and the eigenfunction is positive, an expression is derived for the upper and lower bounds of the true eigenvalue, by making use of the approximate distribution. The difference of the upper and lower bounds is an uncertainty of the presumption of the true critical value. It is found that we can compute the bounds to any required precision. The narrow bounds are calculated for two radially and one totally reflected cylindrical reactors. (auth.)
Level tracking in detailed reactor simulations
Energy Technology Data Exchange (ETDEWEB)
Aktas, B.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)
1995-09-01
We introduce a useful test problem for judging the performance of reactor safety codes in situations where moving two-phase mixture levels are present. The test problem tracks a two-phase liquid level as it rises and then falls back to its original position. Pure air exists above the level, and a low void air-water mixture is below the level. Conditions are subcooled and isothermal to remove complications resulting from failures of interfacial heat transfer packages to properly account for the level. Comparisons are made between the performance of current versions of CATHARE, RELAP5, TRAC-BF1, and TRAC-PF1. These system codes are based on finite-difference methods with a fixed, Eulerian staggered grid in space. When a partially filled cell with a mixture level discontinuity becomes the donor cell, the sharp changes in fluid properties across the interface results in numerical oscillations of various terms. Furthermore, the cell-to-cell convection of mass, momentum and energy are inaccurately predicted nearby a mixture level. To adequately model moving mixture levels, an efficient discontinuity tracking method for the finite-difference Eulerian approximations is described. This model had been implemented in the TRAC-BWR code for the two-phase mixture level tracking since the TRAC-BD1 Version (released April 1984). The result of the test problem run by the current version of TRAC-BF1/MOD1 with the mixture level tracking model shows some peculiar behavior of the variables such as velocities, pressures and interfacial terms. A systematic approach to improving performance of the tracking method is described. Implementing this approach in TRAC-BF1/MOD1 has shown a major improvement in the results.
Simulating the Behaviour of the Fast Reactor Joyo (Draft)
International Nuclear Information System (INIS)
Juutilainen, Pauli
2008-01-01
Motivated by the development of fast reactors the behaviour of the Japanese experimental fast reactor Joyo is simulated with two Monte Carlo codes: Monte Carlo NParticle (MCNP) and Probabilistic Scattering Game (PSG). The simulations are based on the benchmark study 'Japan's Experimental Fast Reactor Joyo MKI core: Sodium-Cooled Uranium-Plutonium Mixed Oxide Fueled Fast Core Surrounded by UO 2 Blanket'. The study is focused on the criticality of the reactor, control rod worth, sodium void reactivity and isothermal temperature coefficient of the reactor. These features are calculated by applying both homogeneous and heterogeneous reactor core models that are built according to the benchmark instructions. The results of the two models obtained by the two codes are compared with each other and especially with the experimental results presented in the benchmark. (author)
Three-Dimensional Numerical Simulation to Mud Turbine for LWD
Yao, Xiaojiang; Dong, Jingxin; Shang, Jie; Zhang, Guanqi
Hydraulic performance analysis was discussed for a type of turbine on generator used for LWD. The simulation models were built by CFD analysis software FINE/Turbo, and full three-dimensional numerical simulation was carried out for impeller group. The hydraulic parameter such as power, speed and pressure drop, were calculated in two kinds of medium water and mud. Experiment was built in water environment. The error of numerical simulation was less than 6%, verified by experiment. Based on this rationalization proposals would be given to choice appropriate impellers, and the rationalization of methods would be explored.
Mitigation of numerical noise for beam loss simulations
Kesting, Frederik
2017-01-01
Numerical noise emerges in self-consistent simulations of charged particles, and its mitigation is investigated since the first numerical studies in plasma physics. In accelerator physics, recent studies find an artificial diffusion of the particle beam due to numerical noise in particle-in-cell tracking, which is of particular importance for high intensity machines with a long storage time, as the SIS100 at FAIR or in context of the LIU upgrade at CERN. In beam loss simulations for these projects artificial effects must be distinguished from physical beam loss. Therefore, it is important to relate artificial diffusion to artificial beam loss, and to choose simulation parameters such that physical beam loss is well resolved. As a practical tool, we therefore suggest a scaling law to find optimal simulation parameters for a given maximum percentage of acceptable artificial beam loss.
Numerical simulation of random stresses on an annular turbulent flow
International Nuclear Information System (INIS)
Marti-Moreno, Marta
2000-01-01
The flow along a circular cylinder may induce structural vibrations. For the predictive analysis of such vibrations, the turbulent forcing spectrum needs to be characterized. The aim of this work is to study the turbulent fluid forces acting on a single tube in axial flow. More precisely we have performed numerical simulations of an annular flow. These simulations were carried out on a cylindrical staggered mesh by a finite difference method. We consider turbulent flow with Reynolds number up to 10 6 . The Large Eddy Simulation Method has been used. A survey of existent experiments showed that hydraulic diameter acts as an important parameter. We first showed the accuracy of the numerical code by reproducing the experiments of Mulcahy. The agreement between pressure spectra from computations and from experiments is good. Then, we applied this code to simulate new numerical experiments varying the hydraulic diameter and the flow velocity. (author) [fr
Study for discharge coefficient of flow nozzles. Prediction by using numerical simulation
International Nuclear Information System (INIS)
Ikeda, Hiroshi; Sakai, Norio; Yamamoto, Yasushi; Arai, Kenji; Matsumoto, Masaaki
2008-01-01
In nuclear plant, as water feeding into reactor have much effect on thermal power of plant, it is important to measure accurately the flow rate of water. Flow nozzle is on of typical differential pressure type flow meters and the discharge coefficient is used to calculate the flow rate. This coefficient is given by actual experiment and theory. We studied the theoretical assumption of the discharge coefficient curve using numerical simulation and evaluated the effect of flow nozzle configuration on the coefficient numerically and experimentally. As the result, numerical simulation can predict the discharge coefficient of theoretical curve within 0.3%. And we found that the throat length and throat tapping location of flow nozzle have much effect on the coefficient. (author)
Practical considerations in developing numerical simulators for thermal recovery
Energy Technology Data Exchange (ETDEWEB)
Abou-Kassem, J.H. [Chemical and Petroleum Engineering Department, UAE University, Al-Ain (United Arab Emirates)
1996-08-15
Numerical simulation of steam injection and in-situ combustion-based oil recovery processes is of great importance in project design. Development of such numerical simulators is an on-going process, with improvements made as the process description becomes more complete, and also as better methods are devised to resolve certain numerical difficulties. This paper addresses some of the latter, and based on the author`s experience gives useful guidelines for developing more efficient numerical simulators of steam injection and in-situ combustion. The paper takes up a series of questions related to simulating thermal processes. Included are: the elimination of constraint equations at the matrix level, phase change, steam injection rate, alternative treatments of heat loss, relative permeabilities and importance of hysteresis effects, improved solutions to the grid orientation problem and other simulation problems such as potential inversion, grid block size, time-step size control and induced fractures. The points discussed in the paper should be of use to both simulator developers and users alike, and will lead to a better understanding of simulation results
Simulation of the TREAT-Upgrade Automatic Reactor Control System
International Nuclear Information System (INIS)
Lipinski, W.C.; Kirsch, L.W.; Valente, A.D.
1984-01-01
This paper describes the design of the Automatic Reactor Control System (ARCS) for the Transient Reactor Test Facility (TREAT) Upgrade. A simulation was used to facilitate the ARCS design and to completely test and verify its operation before installation at the TREAT facility
Development of a full scope reactor engineering simulator
International Nuclear Information System (INIS)
Venhuizen, J.R.; Laats, E.T.
1988-01-01
An engineering laboratory is pursuing the development of an engineering simulator for use by several agencies of the U.S. Government. According to the authors, this simulator will provide the highest fidelity simulation with initial objectives for studying augmented nuclear reactor operator training, and later for advanced concepts testing as applicable to control room accident diagnosis and management
Experimental and numerical analysis of fluid - structure interaction effects in a fast reactor core
International Nuclear Information System (INIS)
Martelli, A.; Forni, M.; Melloni, R.; Paoluzzi, R.; Bonacina, G.; Castoldi, A.; Zola, M.
1990-01-01
Dynamic experiments in air and water (simulating liquid sodium) were performed by ISMES, on behalf of ENEA, on various core element groups of the Italian PEC fast reactor. Bundles of one, seven and nineteen mock-ups reproducing fuel, reflecting and neutron shield elements in full scale were analysed on shaking tables. Tests concerned both groups of equal elements and mixed configurations which corresponded to real core parts. The effects of PEC core-restraint ring were also studied. Seismic excitations of up to 2.5 g were applied to core diagrid. Test results were analysed by use of the one-dimensional program CORALIE and the two-dimensional program CLASH. The study allowed the fluid effects in the PEC core to be evaluated; it also contributed to validation of the above mentioned programs for their general use for fast reactor core analysis. This paper presents the main features of the experimental and the numerical studies and reports comparisons between calculations and measurements. (author)
Numerical Modelling of Wood Gasification in Thermal Plasma Reactor
Czech Academy of Sciences Publication Activity Database
Hirka, Ivan; Živný, Oldřich; Hrabovský, Milan
2017-01-01
Roč. 37, č. 4 (2017), s. 947-965 ISSN 0272-4324 Institutional support: RVO:61389021 Keywords : Plasma modelling * CFD * Thermal plasma reactor * Biomass * Gasification * Syngas Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.355, year: 2016 https://link.springer.com/article/10.1007/s11090-017-9812-z
International Nuclear Information System (INIS)
Jumel, Stephanie; Van-Duysen, Jean Claude
2005-01-01
Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program
Comparison of GPU-Based Numerous Particles Simulation and Experiment
International Nuclear Information System (INIS)
Park, Sang Wook; Jun, Chul Woong; Sohn, Jeong Hyun; Lee, Jae Wook
2014-01-01
The dynamic behavior of numerous grains interacting with each other can be easily observed. In this study, this dynamic behavior was analyzed based on the contact between numerous grains. The discrete element method was used for analyzing the dynamic behavior of each particle and the neighboring-cell algorithm was employed for detecting their contact. The Hertzian and tangential sliding friction contact models were used for calculating the contact force acting between the particles. A GPU-based parallel program was developed for conducting the computer simulation and calculating the numerous contacts. The dam break experiment was performed to verify the simulation results. The reliability of the program was verified by comparing the results of the simulation with those of the experiment
Numerical simulation of gasket behaviour during severe accidents (ATHERMIP project)
International Nuclear Information System (INIS)
Castro Lopez, Fernando; Orden Martinez, Alfredo
1998-01-01
This paper summarises the work carried out to numerically simulate the thermo-mechanical behaviour of sealing gasket in large containment penetrations during a severe accident. The gasket material is an elastomeric material and the thermo-mechanical characterization was based on experimentation. The difficulty of numerical simulation lies in the high non-linearity of the analysis, due on one hand, to the high strain levels reached, and on the other, to stiffness changes introduced by contact/takeoff indicators. Also, the stiffness parameters of the gasket material are not constant, but are subject to changes, both regarding the strain level and the environmental conditions (temperature, radiation). The results obtained allow presenting a calculation model capable of simulating and explaining the behaviour of the sealing gasket during a severe accident. Also, the failure hypothesis numerically obtained was environmentally validated. (author)
Numerical Simulation of Anisotropic Preheating Ablative Rayleigh–Taylor Instability
International Nuclear Information System (INIS)
Li-Feng, Wang; Wen-Hua, Ye; Ying-Jun, Li
2010-01-01
The linear growth rate of the anisotropic preheating ablative Rayleigh–Taylor instability (ARTI) is studied by numerical simulations. The preheating model κ(T) = κ SH [1 + f(T)] is applied, where f(T) is the preheating function interpreting the preheating tongue effect in the cold plasma ahead of the ablative front. An arbitrary coefficient D is introduced in the energy equation to study the influence of transverse thermal conductivity on the growth of the ARTI. We find that enhancing diffusion in a plane transverse to the mean longitudinal flow can strongly reduce the growth of the instability. Numerical simulations exhibit a significant stabilization of the ablation front by improving the transverse thermal conduction. Our results are in general agreement with the theory analysis and numerical simulations by Masse [Phys. Rev. Lett. 98 (2007) 245001]. (physics of gases, plasmas, and electric discharges)
Numerical simulation of anisotropic preheating ablative Rayleigh-Taylor instability
International Nuclear Information System (INIS)
Wang Lifeng; Ye Wenhua; Li Yingjun
2010-01-01
The linear growth rate of the anisotropic preheating ablative Rayleigh-Taylor instability (ARTI) is studied by numerical simulations. The preheating model κ(T)=κ SH [1+f(T)] is applied, where f(T) is the preheating function interpreting the preheating tongue effect in the cold plasma ahead of the ablative front. An arbitrary coefficient D is introduced in the energy equation to study the influence of transverse thermal conductivity on the growth of the ARTI. We find that enhancing diffusion in a plane transverse to the mean longitudinal flow can strongly reduce the growth of the instability. Numerical simulations exhibit a significant stabilization of the ablation front by improving the transverse thermal conduction. Our results are in general agreement with the theory analysis and numerical simulations by Masse. (authors)
Investigations into radiation damages of reactor materials by computer simulation
International Nuclear Information System (INIS)
Bronnikov, V.A.
2004-01-01
Data on the state of works in European countries in the field of computerized simulation of radiation damages of reactor materials under the context of the international projects ITEM (European Database for Multiscale Modelling) and SIRENA (Simulation of Radiation Effects in Zr-Nb alloys) - computerized simulation of stress corrosion when contact of Zr-Nb alloys with iodine are presented. Computer codes for the simulation of radiation effects in reactor materials were developed. European Database for Multiscale Modelling (EDAM) was organized using the results of the investigations provided in the ITEM project [ru
Vortex locking in direct numerical simulations of quantum turbulence.
Morris, Karla; Koplik, Joel; Rouson, Damian W I
2008-07-04
Direct numerical simulations are used to examine the locking of quantized superfluid vortices and normal fluid vorticity in evolving turbulent flows. The superfluid is driven by the normal fluid, which undergoes either a decaying Taylor-Green flow or a linearly forced homogeneous isotropic turbulent flow, although the back reaction of the superfluid on the normal fluid flow is omitted. Using correlation functions and wavelet transforms, we present numerical and visual evidence for vortex locking on length scales above the intervortex spacing.
Numerical simulation on quantum turbulence created by an oscillating object
Energy Technology Data Exchange (ETDEWEB)
Fujiyama, S; Tsubota, M [Department of Physics, Osaka City University, 3-3-138 Sugimoto, Sumiyoshi-ku, Osaka City, Osaka (Japan)], E-mail: fujiyama@sci.osaka-cu.ac.jp
2009-02-01
We have conducted a numerical simulation of vortex dynamics in superfluid {sup 4}He in the presence of an oscillating sphere. The experiment on a vibrating wire that measured the transition from laminar to turbulent flow is modelled in our simulations. The simulation exhibits the details of vortex growth by the oscillating sphere. Our result also shows that a more realistic modelling may change the destiny of the vortex rings detached from the sphere. We have evaluated the force driven by the sphere in the simulation and have confirmed the onset of the quantum turbulence.
Recent developments in numerical simulation techniques of thermal recovery processes
Energy Technology Data Exchange (ETDEWEB)
Tamim, M. [Bangladesh University of Engineering and Technology, Bangladesh (Bangladesh); Abou-Kassem, J.H. [Chemical and Petroleum Engineering Department, UAE University, Al-Ain 17555 (United Arab Emirates); Farouq Ali, S.M. [University of Alberta, Alberta (Canada)
2000-05-01
Numerical simulation of thermal processes (steam flooding, steam stimulation, SAGD, in-situ combustion, electrical heating, etc.) is an integral part of a thermal project design. The general tendency in the last 10 years has been to use commercial simulators. During the last decade, only a few new models have been reported in the literature. More work has been done to modify and refine solutions to existing problems to improve the efficiency of simulators. The paper discusses some of the recent developments in simulation techniques of thermal processes such as grid refinement, grid orientation, effect of temperature on relative permeability, mathematical models, and solution methods. The various aspects of simulation discussed here promote better understanding of the problems encountered in the simulation of thermal processes and will be of value to both simulator users and developers.
Development of Reactor TRIGA PUSPATI Simulator for Education and Training
International Nuclear Information System (INIS)
Mohd Sabri Minhat; Zarina Masood; Muhammad Rawi Mohamed Zin
2016-01-01
The real-time simulator for Reactor TRIGA PUSPATI (RTP) which is under development. The main purpose of this simulator is operator training and a dynamic test bed (DTB) to test and validate the control logics in reactor regulating system (RRS) of RTP. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, operator station, a network switch, control rod drive mechanism and a large display panel. The RTP hardwired panel is replicated similar to real console. The software includes a mathematical model includes reactor kinetics and thermal-hydraulics that implements plant dynamics in real-time using LabVIEW, an instructor station module work as host computer that manages user instructions, and a human machine interface module as a graphical user interface which is used in the real RTP plant. The developed TRIGA reactor simulators are installed in the Malaysian Nuclear Agency nuclear training center for reactor operator training. To use the simulator as a dynamic test-bed, the reactor regulating system modeling software of the simulator was replaced by actual RRS cabinet which is consist of Programmable Logic Controller (PLC) S7-1500, and was interfaced using a hard-wired and network-based interface. RRS cabinet generates control signals for reactor power control based on the various feedback signals from DTB such as neutron detector signal and control rod positions, and the DTB runs plant dynamics based on the RRS control signals. Thus the Hardware-In-the-Loop Simulation between RRS and the emulated plant (DTB) has been implemented and tested in this configuration. Normal and abnormal case test have been emulated for this project. In conclusion, the functions and the control performance of the developed RTP dynamic test bed simulator have been tested showing reasonable and acceptable results. (author)
Hybrid simulation of reactor kinetics in CANDU reactors using a modal approach
International Nuclear Information System (INIS)
Monaghan, B.M.; McDonnell, F.N.; Hinds, H.W.T.; m.
1980-01-01
A hybrid computer model for simulating the behaviour of large CANDU (Canada Deuterium Uranium) reactor cores is presented. The main dynamic variables are expressed in terms of weighted sums of a base set of spatial natural-mode functions with time-varying co-efficients. This technique, known as the modal or synthesis approach, permits good three-dimensional representation of reactor dynamics and is well suited to hybrid simulation. The hybrid model provides improved man-machine interaction and real-time capability. The model was used in two applications. The first studies the transient that follows a loss of primary coolant and reactor shutdown; the second is a simulation of the dynamics of xenon, a fission product which has a high absorption cross-section for neutrons and thus has an important effect on reactor behaviour. Comparison of the results of the hybrid computer simulation with those of an all-digital one is good, within 1% to 2%
The development of fast simulation program for marine reactor parameters
International Nuclear Information System (INIS)
Chen Zhiyun; Hao Jianli; Chen Wenzhen
2012-01-01
Highlights: ► The simplified physical and mathematical models are proposed for a marine reactor system. ► A program is developed with Simulink module and Matlab file. ► The program developed has the merit of easy input preparation, output processing and fast running. ► The program can be used for the fast simulation of marine reactor parameters on the operating field. - Abstract: The fast simulation program for marine reactor parameters is developed based on the Simulink simulating software according to the characteristics of marine reactor with requirement of maneuverability and acute and fast response. The simplified core physical and thermal model, pressurizer model, steam generator model, control rod model, reactivity model and the corresponding Simulink modules are established. The whole program is developed by coupling all the Simulink modules. Two typical transient processes of marine reactor with fast load increase at low power level and load rejection at high power level are adopted to verify the program. The results are compared with those of Relap5/Mod3.2 with good consistency, and the program runs very fast. It is shown that the program is correct and suitable for the fast and accurate simulation of marine reactor parameters on the operating field, which is significant to the marine reactor safe operation.
Direct numerical simulation of noninvasive channel healing in electrical field
Wang, Yi
2017-11-25
Noninvasive channel healing is a new idea to repair the broken pipe wall, using external electric fields to drive iron particles to the destination. The repair can be done in the normal operation of the pipe flow without any shutdown of the pipeline so that this method can be a potentially efficient and safe technology of pipe healing. However, the real application needs full knowledge of healing details. Numerical simulation is an effective method. Thus, in this research, we first established a numerical model for noninvasive channel healing technology to represent fluid–particle interaction. The iron particles can be attached to a cracking area by external electrostatic forces or can also be detached by mechanical forces from the fluid. When enough particles are permanently attached on the cracking area, the pipe wall can be healed. The numerical criterion of the permanent attachment is discussed. A fully three-dimensional finite difference framework of direct numerical simulation is established and applied to different cases to simulate the full process of channel healing. The impact of Reynolds number and particle concentration on the healing process is discussed. This numerical investigation provides valuable reference and tools for further simulation of real pipe healing in engineering.
On the elimination of numerical Cerenkov radiation in PIC simulations
International Nuclear Information System (INIS)
Greenwood, Andrew D.; Cartwright, Keith L.; Luginsland, John W.; Baca, Ernest A.
2004-01-01
Particle-in-cell (PIC) simulations are a useful tool in modeling plasma in physical devices. The Yee finite difference time domain (FDTD) method is commonly used in PIC simulations to model the electromagnetic fields. However, in the Yee FDTD method, poorly resolved waves at frequencies near the cut off frequency of the grid travel slower than the physical speed of light. These slowly traveling, poorly resolved waves are not a problem in many simulations because the physics of interest are at much lower frequencies. However, when high energy particles are present, the particles may travel faster than the numerical speed of their own radiation, leading to non-physical, numerical Cerenkov radiation. Due to non-linear interaction between the particles and the fields, the numerical Cerenkov radiation couples into the frequency band of physical interest and corrupts the PIC simulation. There are two methods of mitigating the effects of the numerical Cerenkov radiation. The computational stencil used to approximate the curl operator can be altered to improve the high frequency physics, or a filtering scheme can be introduced to attenuate the waves that cause the numerical Cerenkov radiation. Altering the computational stencil is more physically accurate but is difficult to implement while maintaining charge conservation in the code. Thus, filtering is more commonly used. Two previously published filters by Godfrey and Friedman are analyzed and compared to ideally desired filter properties
Numerical Simulation of Antennae by Discrete Exterior Calculus
International Nuclear Information System (INIS)
Xie Zheng; Ye Zheng; Ma Yujie
2009-01-01
Numerical simulation of antennae is a topic in computational electromagnetism, which is concerned with the numerical study of Maxwell equations. By discrete exterior calculus and the lattice gauge theory with coefficient R, we obtain the Bianchi identity on prism lattice. By defining an inner product of discrete differential forms, we derive the source equation and continuity equation. Those equations compose the discrete Maxwell equations in vacuum case on discrete manifold, which are implemented on Java development platform to simulate the Gaussian pulse radiation on antennaes. (electromagnetism, optics, acoustics, heat transfer, classical mechanics, and fluid dynamics)
On the numerical simulation of tracer flows in porous media
International Nuclear Information System (INIS)
Aquino, J.; Pereira, F.; Amaral Souto, H.P.; Francisco, A.S.
2007-01-01
We discuss in detail a new Lagrangian, locally conservative procedure which has been proposed for the numerical solution of linear transport problems in porous media. The new scheme is computationally efficient, virtually free of numerical diffusion, and can be applied to investigate numerically the time evolution of radionuclide contaminant plumes. Results of two-dimensional simulations of tracer flows will be presented to show the influence on the computed solutions of distinct interpolation functions for evaluating the velocity field at any position of the physical domain, as required by the Lagrangian scheme. (author)
Numerical simulation of explosive magnetic cumulative generator EMG-720
Energy Technology Data Exchange (ETDEWEB)
Deryugin, Yu N; Zelenskij, D K; Kazakova, I F; Kargin, V I; Mironychev, P V; Pikar, A S; Popkov, N F; Ryaslov, E A; Ryzhatskova, E G [All-Russian Research Inst. of Experimental Physics, Sarov (Russian Federation)
1997-12-31
The paper discusses the methods and results of numerical simulations used in the development of a helical-coaxial explosive magnetic cumulative generator (EMG) with the stator up to 720 mm in diameter. In the process of designing, separate units were numerically modeled, as was the generator operation with a constant inductive-ohmic load. The 2-D processes of the armature acceleration by the explosion products were modeled as well as those of the formation of the sliding high-current contact between the armature and stator`s insulated turns. The problem of the armature integrity in the region of the detonation waves collision was numerically analyzed. 8 figs., 2 refs.
NUMERICAL SIMULATION OF SHOCK WAVE REFRACTION ON INCLINED CONTACT DISCONTINUITY
Directory of Open Access Journals (Sweden)
P. V. Bulat
2016-05-01
Full Text Available We consider numerical simulation of shock wave refraction on plane contact discontinuity, separating two gases with different density. Discretization of Euler equations is based on finite volume method and WENO finite difference schemes, implemented on unstructured meshes. Integration over time is performed with the use of the third-order Runge–Kutta stepping procedure. The procedure of identification and classification of gas dynamic discontinuities based on conditions of dynamic consistency and image processing methods is applied to visualize and interpret the results of numerical calculations. The flow structure and its quantitative characteristics are defined. The results of numerical and experimental visualization (shadowgraphs, schlieren images, and interferograms are compared.
Processing biobased polymers using plasticizers: Numerical simulations versus experiments
Desplentere, Frederik; Cardon, Ludwig; Six, Wim; Erkoç, Mustafa
2016-03-01
In polymer processing, the use of biobased products shows lots of possibilities. Considering biobased materials, biodegradability is in most cases the most important issue. Next to this, bio based materials aimed at durable applications, are gaining interest. Within this research, the influence of plasticizers on the processing of the bio based material is investigated. This work is done for an extrusion grade of PLA, Natureworks PLA 2003D. Extrusion through a slit die equipped with pressure sensors is used to compare the experimental pressure values to numerical simulation results. Additional experimental data (temperature and pressure data along the extrusion screw and die are recorded) is generated on a dr. Collin Lab extruder producing a 25mm diameter tube. All these experimental data is used to indicate the appropriate functioning of the numerical simulation tool Virtual Extrusion Laboratory 6.7 for the simulation of both the industrial available extrusion grade PLA and the compound in which 15% of plasticizer is added. Adding the applied plasticizer, resulted in a 40% lower pressure drop over the extrusion die. The combination of different experiments allowed to fit the numerical simulation results closely to the experimental values. Based on this experience, it is shown that numerical simulations also can be used for modified bio based materials if appropriate material and process data are taken into account.
Fine numerical modelling of thermohydraulic phenomena in EDF PWR reactors
International Nuclear Information System (INIS)
Boulot, F.
1993-01-01
Over the last 20 years, EDF has developed a family of 2D and 3D industrial thermohydraulics software to solve problems encountered in existing PWR power plants and to design new reactors for the future. The equations used in the models are the averaged Navier-Stokes and energy equations. A brief description is given of the four main codes developed for single-phase and two-phase water-steam flows, some of which use finite differences or finite volumes methods, while others make use of finite elements methods. An example of application is given for each code. (author). 4 figs., 4 refs
Dynamic simulation of the 2 MWt slowpoke heating reactor
International Nuclear Information System (INIS)
Tseng, C.M.; Lepp, R.M.
1982-04-01
A 2 MWt SLOWPOKE reactor, intended for commercial space heating, is being developed at the Chalk River Nuclear Laboratories. A small-signal dynamic simulation of this reactor, without closed-loop control, was developed. Basic equations were used to describe the physical phenomena in each kf the eight reactor subsystems. These equations were then linearized about the normal operation conditions and rearranged in a dimensionless form for implementation. The overall simulation is non-linear. Slow transient responses (minutes to days) of the simulation to both reactivity and temperature perturbations were measured at full power. In all cases the system reached a new steady state in times varying from 12 h to 250 h. These results illustrate the benefits of the inherent negative reactivity feedback of this reactor concept. The addition of closed-loop control using core outlet temperature as the controlled variable to move a beryllium reflector is also examined
Numerical simulations and mathematical models of flows in complex geometries
DEFF Research Database (Denmark)
Hernandez Garcia, Anier
The research work of the present thesis was mainly aimed at exploiting one of the strengths of the Lattice Boltzmann methods, namely, the ability to handle complicated geometries to accurately simulate flows in complex geometries. In this thesis, we perform a very detailed theoretical analysis...... and through the Chapman-Enskog multi-scale expansion technique the dependence of the kinetic viscosity on each scheme is investigated. Seeking for optimal numerical schemes to eciently simulate a wide range of complex flows a variant of the finite element, off-lattice Boltzmann method [5], which uses...... the characteristic based integration is also implemented. Using the latter scheme, numerical simulations are conducted in flows of different complexities: flow in a (real) porous network and turbulent flows in ducts with wall irregularities. From the simulations of flows in porous media driven by pressure gradients...
Numerical simulation of airfoil trailing edge serration noise
DEFF Research Database (Denmark)
Zhu, Wei Jun; Shen, Wen Zhong
In the present work, numerical simulations are carried out for a low noise airfoil with and without serrated Trailing Edge. The Ffowcs Williams-Hawkings acoustic analogy is implemented into the in-house incompressible flow solver EllipSys3D. The instantaneous hydrodynamic pressure and velocity...... field are obtained using Large Eddy Simulation. To obtain the time history data of sound pressure, the flow quantities are integrated around the airfoil surface through the FW-H approach. The extended length of the serration is about 16.7% of the airfoil chord and the geometric angle of the serration...... is 28 degrees. The chord based Reynolds number is around 1.5x106. Simulations are compared with existing wind tunnel experiments at various angles of attack. Even though the airfoil under investigation is already optimized for low noise emission, numerical simulations and wind tunnel experiments show...
Numerical simulations of comets - predictions for Comet Giacobini-Zinner
International Nuclear Information System (INIS)
Fedder, J.A.; Lyon, J.G.; Giuliani, J.L. Jr.
1986-01-01
Simulations of Comet Giacobini-Zinner's interaction with solar wind are described and results are presented. The simulations are carried out via the numerical solution of the ideal MHD equations as an initial value problem in a uniform solar wind. The calculations are performed on a Cartesian mesh centered at the comet. Results reveal that the first significant modifications of the solar wind along the ISEE/ICE trajectory will occur 100,000 km from the solar wind comet axis. 6 references
3D numerical simulation of transient processes in hydraulic turbines
International Nuclear Information System (INIS)
Cherny, S; Chirkov, D; Lapin, V; Eshkunova, I; Bannikov, D; Avdushenko, A; Skorospelov, V
2010-01-01
An approach for numerical simulation of 3D hydraulic turbine flows in transient operating regimes is presented. The method is based on a coupled solution of incompressible RANS equations, runner rotation equation, and water hammer equations. The issue of setting appropriate boundary conditions is considered in detail. As an illustration, the simulation results for runaway process are presented. The evolution of vortex structure and its effect on computed runaway traces are analyzed.
3D numerical simulation of transient processes in hydraulic turbines
Cherny, S.; Chirkov, D.; Bannikov, D.; Lapin, V.; Skorospelov, V.; Eshkunova, I.; Avdushenko, A.
2010-08-01
An approach for numerical simulation of 3D hydraulic turbine flows in transient operating regimes is presented. The method is based on a coupled solution of incompressible RANS equations, runner rotation equation, and water hammer equations. The issue of setting appropriate boundary conditions is considered in detail. As an illustration, the simulation results for runaway process are presented. The evolution of vortex structure and its effect on computed runaway traces are analyzed.
International Nuclear Information System (INIS)
Chein, Reiyu; Chen, Yen-Cho; Chung, J.N.
2013-01-01
Highlights: ► Performance of mini-scale integrated annulus reactors for hydrogen production. ► Flow rates fed to combustor and reformer control the reactor performance. ► Optimum performance is found from balance of flow rates to combustor and reformer. ► Better performance can be found when shell side is designed as combustor. -- Abstract: This study presents the numerical simulation on the performance of mini-scale reactors for hydrogen production coupled with liquid methanol/water vaporizer, methanol/steam reformer, and methanol/air catalytic combustor. These reactors are designed similar to tube-and-shell heat exchangers. The combustor for heat supply is arranged as the tube or shell side. Based on the obtained results, the methanol/air flow rate through the combustor (in terms of gas hourly space velocity of combustor, GHSV-C) and the methanol/water feed rate to the reformer (in terms of gas hourly space velocity of reformer, GHSV-R) control the reactor performance. With higher GHSV-C and lower GHSV-R, higher methanol conversion can be achieved because of higher reaction temperature. However, hydrogen yield is reduced and the carbon monoxide concentration is increased due to the reversed water gas shift reaction. Optimum reactor performance is found using the balance between GHSV-C and GHSV-R. Because of more effective heat transfer characteristics in the vaporizer, it is found that the reactor with combustor arranged as the shell side has better performance compared with the reactor design having the combustor as the tube side under the same operating conditions.
Direct Numerical Simulation and Visualization of Subcooled Pool Boiling
Directory of Open Access Journals (Sweden)
Tomoaki Kunugi
2014-01-01
Full Text Available A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors. On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated.
Numerical simulation and experimental validation of coiled adiabatic capillary tubes
Energy Technology Data Exchange (ETDEWEB)
Garcia-Valladares, O. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico (UNAM), Apdo. Postal 34, 62580 Temixco, Morelos (Mexico)
2007-04-15
The objective of this study is to extend and validate the model developed and presented in previous works [O. Garcia-Valladares, C.D. Perez-Segarra, A. Oliva, Numerical simulation of capillary tube expansion devices behaviour with pure and mixed refrigerants considering metastable region. Part I: mathematical formulation and numerical model, Applied Thermal Engineering 22 (2) (2002) 173-182; O. Garcia-Valladares, C.D. Perez-Segarra, A. Oliva, Numerical simulation of capillary tube expansion devices behaviour with pure and mixed refrigerants considering metastable region. Part II: experimental validation and parametric studies, Applied Thermal Engineering 22 (4) (2002) 379-391] to coiled adiabatic capillary tube expansion devices working with pure and mixed refrigerants. The discretized governing equations are coupled using an implicit step by step method. A special treatment has been implemented in order to consider transitions (subcooled liquid region, metastable liquid region, metastable two-phase region and equilibrium two-phase region). All the flow variables (enthalpies, temperatures, pressures, vapor qualities, velocities, heat fluxes, etc.) together with the thermophysical properties are evaluated at each point of the grid in which the domain is discretized. The numerical model allows analysis of aspects such as geometry, type of fluid (pure substances and mixtures), critical or non-critical flow conditions, metastable regions, and transient aspects. Comparison of the numerical simulation with a wide range of experimental data presented in the technical literature will be shown in the present article in order to validate the model developed. (author)
Can numerical simulations accurately predict hydrodynamic instabilities in liquid films?
Denner, Fabian; Charogiannis, Alexandros; Pradas, Marc; van Wachem, Berend G. M.; Markides, Christos N.; Kalliadasis, Serafim
2014-11-01
Understanding the dynamics of hydrodynamic instabilities in liquid film flows is an active field of research in fluid dynamics and non-linear science in general. Numerical simulations offer a powerful tool to study hydrodynamic instabilities in film flows and can provide deep insights into the underlying physical phenomena. However, the direct comparison of numerical results and experimental results is often hampered by several reasons. For instance, in numerical simulations the interface representation is problematic and the governing equations and boundary conditions may be oversimplified, whereas in experiments it is often difficult to extract accurate information on the fluid and its behavior, e.g. determine the fluid properties when the liquid contains particles for PIV measurements. In this contribution we present the latest results of our on-going, extensive study on hydrodynamic instabilities in liquid film flows, which includes direct numerical simulations, low-dimensional modelling as well as experiments. The major focus is on wave regimes, wave height and wave celerity as a function of Reynolds number and forcing frequency of a falling liquid film. Specific attention is paid to the differences in numerical and experimental results and the reasons for these differences. The authors are grateful to the EPSRC for their financial support (Grant EP/K008595/1).
Liquid metal cooled experimental fast reactor simulator
International Nuclear Information System (INIS)
Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.
1997-01-01
This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs
Development of numerical methods for thermohydraulic problems in reactor safety
International Nuclear Information System (INIS)
Chabrillac, M.; Kavenoky, A.; Le Coq, G.; L'Heriteau, J.P.; Stewart, B.; Rousseau, J.C.
1976-01-01
Numerical methods are being developed for the LOCA calculation; the first part is devoted to the BERTHA model and the associated characteristic treatment for the first seconds of the blowdown, the second part presents the problems encountered for accounting for velocity difference between phases. The FLIRA treatment of the reflooding is presented in the last part: this treatment allows the calculation of the quenching front velocity
Seasonal cycle of Martian climate : Experimental data and numerical simulation
Rodin, A. V.; Willson, R. J.
2006-01-01
The most adequate theoretical method of investigating the present-day Martian climate is numerical simulation based on a model of general circulation of the atmosphere. First and foremost, such models encounter the greatest difficulties in description of aerosols and clouds, which in turn
Decoupled numerical simulation of a solid fuel fired retort boiler
International Nuclear Information System (INIS)
Ryfa, Arkadiusz; Buczynski, Rafal; Chabinski, Michal; Szlek, Andrzej; Bialecki, Ryszard A.
2014-01-01
The paper deals with numerical simulation of the retort boiler fired with solid fuel. Such constructions are very popular for heating systems and their development is mostly based on the designer experience. The simulations have been done in ANSYS/Fluent package and involved two numerical models. The former deals with a fixed-bed combustion of the solid fuel and free-board gas combustion. Solid fuel combustion is based on the coal kinetic parameters. This model encompasses chemical reactions, radiative heat transfer and turbulence. Coal properties have been defined with user defined functions. The latter model describes flow of water inside a water jacked that surrounds the combustion chamber and flue gas ducts. The novelty of the proposed approach is separating of the combustion simulation from the water flow. Such approach allows for reducing the number of degrees of freedom and thus lowering the necessary numerical effort. Decoupling combustion from water flow requires defining interface boundary condition. As this boundary condition is unknown it is adjusted iteratively. The results of the numerical simulation have been successfully validated against measurement data. - Highlights: • New decoupled modelling of small scale boiler is proposed. • Fixed-bed combustion model based on kinetic parameters is introduced. • Decoupling reduced the complexity of the model and computational time. • Simple and computationally inexpensive coupling algorithm is proposed. • Model is successfully validated against measurements
A review of numerical simulation of hydrothermal systems.
Mercer, J.W.; Faust, C.R.
1979-01-01
Many advances in simulating single and two-phase fluid flow and heat transport in porous media have recently been made in conjunction with geothermal energy research. These numerical models reproduce system thermal and pressure behaviour and can be used for other heat-transport problems, such as high-level radioactive waste disposal and heat-storage projects. -Authors
Application of HPCN to direct numerical simulation of turbulent flow
Verstappen, RWCP; Veldman, AEP; van Waveren, GM; Hertzberger, B; Sloot, P
1997-01-01
This poster shows how HPCN can be used as a path-finding tool for turbulence research. The parallelization of direct numerical simulation of turbulent flow using the data-parallel model and Fortran 95 constructs is treated, both on a shared memory and a distributed memory computer.
Numerical simulation of thermal fracture in functionally graded
Indian Academy of Sciences (India)
Numerical simulation of thermal fracture in functionally graded materials using element-free ... Initially, the temperature distribution over the domain is obtained by solving the heat transfer problem. ... Department of Mechanical Engineering, National Institute of Technology, Hamirpur 177005, India ... Contact | Site index.
Numerical simulations of the metallicity distribution in dwarf spheroidal galaxies
Ripamonti, E.; Tolstoy, E.; Helmi, A.; Battaglia, G.; Abel, T.
2006-01-01
Abstract: Recent observations show that the number of stars with very low metallicities in the dwarf spheroidal satellites of the Milky Way is low, despite the low average metallicities of stars in these systems. We undertake numerical simulations of star formation and metal enrichment of dwarf
Numerical convergence improvements for porflow unsaturated flow simulations
Energy Technology Data Exchange (ETDEWEB)
Flach, Greg [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)
2017-08-14
Section 3.6 of SRNL (2016) discusses various PORFLOW code improvements to increase modeling efficiency, in preparation for the next E-Area Performance Assessment (WSRC 2008) revision. This memorandum documents interaction with Analytic & Computational Research, Inc. (http://www.acricfd.com/default.htm) to improve numerical convergence efficiency using PORFLOW version 6.42 for unsaturated flow simulations.
Direct numerical simulation of particulate flow with heat transfer
Tavassoli Estahbanati, H; Kriebitzsch, S.H.L.; Hoef, van der M.A.; Peters, E.A.J.F.; Kuipers, J.A.M.
2013-01-01
The Immersed Boundary (IB) method proposed by Uhlmann for Direct Numerical Simulation (DNS) of fluid flow through dense fluid-particle systems is extended to systems with interphase heat transport. A fixed Eulerian grid is employed to solve the momentum and energy equations by traditional
Experimental and numerical simulation of carbon manganese steel ...
African Journals Online (AJOL)
Experimental and numerical simulation of carbon manganese steel for cyclic plastic behaviour. J Shit, S Dhar, S Acharyya. Abstract. The paper deals with finite element modeling of saturated low cycle fatigue and the cyclic hardening phenomena of the materials Sa333 grade 6 carbon steel and SS316 stainless steel.
Numerical simulation of the drying of inkjet-printed droplets
Siregar, D.P.; Kuerten, J.G.M.; Geld, van der C.W.M.
2013-01-01
In this paper we study the behavior of an inkjet-printed droplet of a solute dissolved in a solvent on a solid horizontal surface by numerical simulation. An extended model for drying of a droplet and the final distribution of the solute on an impermeable substrate is proposed. The model extends the
Direct Numerical Simulation Sediment Transport in Horizontal Channel
International Nuclear Information System (INIS)
Uhlmann, M.
2006-01-01
We numerically simulate turbulent flow in a horizontal plane channel over a bed of mobile particles. All scales of fluid motion are resolved without modeling and the phase interface is accurately represented. Our results indicate a possible scenario for the onset of erosion through collective motion induced by buffer-layer streaks. (Author) 27 refs
Numerical simulations of time-resolved quantum electronics
International Nuclear Information System (INIS)
Gaury, Benoit; Weston, Joseph; Santin, Matthieu; Houzet, Manuel; Groth, Christoph; Waintal, Xavier
2014-01-01
Numerical simulation has become a major tool in quantum electronics both for fundamental and applied purposes. While for a long time those simulations focused on stationary properties (e.g. DC currents), the recent experimental trend toward GHz frequencies and beyond has triggered a new interest for handling time-dependent perturbations. As the experimental frequencies get higher, it becomes possible to conceive experiments which are both time-resolved and fast enough to probe the internal quantum dynamics of the system. This paper discusses the technical aspects–mathematical and numerical–associated with the numerical simulations of such a setup in the time domain (i.e. beyond the single-frequency AC limit). After a short review of the state of the art, we develop a theoretical framework for the calculation of time-resolved observables in a general multiterminal system subject to an arbitrary time-dependent perturbation (oscillating electrostatic gates, voltage pulses, time-varying magnetic fields, etc.) The approach is mathematically equivalent to (i) the time-dependent scattering formalism, (ii) the time-resolved non-equilibrium Green’s function (NEGF) formalism and (iii) the partition-free approach. The central object of our theory is a wave function that obeys a simple Schrödinger equation with an additional source term that accounts for the electrons injected from the electrodes. The time-resolved observables (current, density, etc.) and the (inelastic) scattering matrix are simply expressed in terms of this wave function. We use our approach to develop a numerical technique for simulating time-resolved quantum transport. We find that the use of this wave function is advantageous for numerical simulations resulting in a speed up of many orders of magnitude with respect to the direct integration of NEGF equations. Our technique allows one to simulate realistic situations beyond simple models, a subject that was until now beyond the simulation
Mathematical modeling and numerical simulation of Czochralski Crystal Growth
Energy Technology Data Exchange (ETDEWEB)
Jaervinen, J.; Nieminen, R. [Center for Scientific Computing, Espoo (Finland)
1996-12-31
A detailed mathematical model and numerical simulation tools based on the SUPG Finite Element Method for the Czochralski crystal growth has been developed. In this presentation the mathematical modeling and numerical simulation of the melt flow and the temperature distribution in a rotationally symmetric crystal growth environment is investigated. The temperature distribution and the position of the free boundary between the solid and liquid phases are solved by using the Enthalpy method. Heat inside of the Czochralski furnace is transferred by radiation, conduction and convection. The melt flow is governed by the incompressible Navier-Stokes equations coupled with the enthalpy equation. The melt flow is numerically demonstrated and the temperature distribution in the whole Czochralski furnace. (author)
Behavioral modeling of SRIM tables for numerical simulation
Energy Technology Data Exchange (ETDEWEB)
Martinie, S., E-mail: sebastien.martinie@cea.fr; Saad-Saoud, T.; Moindjie, S.; Munteanu, D.; Autran, J.L., E-mail: jean-luc.autran@univ-amu.fr
2014-03-01
Highlights: • Behavioral modeling of SRIM data is performed on the basis of power polynomial fitting functions. • Fast and continuous numerical functions are proposed for the stopping power and projected range. • Functions have been successfully tested for a wide variety of ions and targets. • Typical accuracies below the percent have been obtained in the range 1 keV–1 GeV. - Abstract: This work describes a simple way to implement SRIM stopping power and range tabulated data in the form of fast and continuous numerical functions for intensive simulation. We provide here the methodology of this behavioral modeling as well as the details of the implementation and some numerical examples for ions in silicon target. Developed functions have been successfully tested and used for the simulation of soft errors in microelectronics circuits.
Mathematical modeling and numerical simulation of Czochralski Crystal Growth
Energy Technology Data Exchange (ETDEWEB)
Jaervinen, J; Nieminen, R [Center for Scientific Computing, Espoo (Finland)
1997-12-31
A detailed mathematical model and numerical simulation tools based on the SUPG Finite Element Method for the Czochralski crystal growth has been developed. In this presentation the mathematical modeling and numerical simulation of the melt flow and the temperature distribution in a rotationally symmetric crystal growth environment is investigated. The temperature distribution and the position of the free boundary between the solid and liquid phases are solved by using the Enthalpy method. Heat inside of the Czochralski furnace is transferred by radiation, conduction and convection. The melt flow is governed by the incompressible Navier-Stokes equations coupled with the enthalpy equation. The melt flow is numerically demonstrated and the temperature distribution in the whole Czochralski furnace. (author)
On the characteristics of a numerical fluid dynamics simulator
International Nuclear Information System (INIS)
Winkler, K.H.A.; Norman, M.L.; Norton, J.L.
1986-01-01
John von Neumann envisioned scientists and mathematicians analyzing and controlling their numerical experiments on nonlinear dynamic systems interactively. The authors describe their concept of a real-time Numerical Fluid Dynamics Simulator NFDS. The authors envision the NFDS to be composed of simulation processors, data storage devices, and image processing devices of extremely high power and capacity, interconnected by very high throughput communication channels. They present individual component performance requirements for both real-time and playback operating modes of the NFDS, using problems of current interest in fluid dynamics as examples. Scaling relations are derived showing the dependence of system requirements on the dimensionality and complexity of the numerical model. The authors conclude by extending their analysis to the system requirements posed in modeling the more involved physics of radiation hydrodynamics
Agglomeration processes in carbonaceous dusty plasmas, experiments and numerical simulations
International Nuclear Information System (INIS)
Dap, S; Hugon, R; De Poucques, L; Bougdira, J; Lacroix, D; Patisson, F
2010-01-01
This paper deals with carbon dust agglomeration in radio frequency acetylene/argon plasma. Two studies, an experimental and a numerical one, were carried out to model dust formation mechanisms. Firstly, in situ transmission spectroscopy of dust clouds in the visible range was performed in order to observe the main features of the agglomeration process of the produced carbonaceous dust. Secondly, numerical simulation tools dedicated to understanding the achieved experiments were developed. A first model was used for the discretization of the continuous population balance equations that characterize the dust agglomeration process. The second model is based on a Monte Carlo ray-tracing code coupled to a Mie theory calculation of dust absorption and scattering parameters. These two simulation tools were used together in order to numerically predict the light transmissivity through a dusty plasma and make comparisons with experiments.
Behavioral modeling of SRIM tables for numerical simulation
International Nuclear Information System (INIS)
Martinie, S.; Saad-Saoud, T.; Moindjie, S.; Munteanu, D.; Autran, J.L.
2014-01-01
Highlights: • Behavioral modeling of SRIM data is performed on the basis of power polynomial fitting functions. • Fast and continuous numerical functions are proposed for the stopping power and projected range. • Functions have been successfully tested for a wide variety of ions and targets. • Typical accuracies below the percent have been obtained in the range 1 keV–1 GeV. - Abstract: This work describes a simple way to implement SRIM stopping power and range tabulated data in the form of fast and continuous numerical functions for intensive simulation. We provide here the methodology of this behavioral modeling as well as the details of the implementation and some numerical examples for ions in silicon target. Developed functions have been successfully tested and used for the simulation of soft errors in microelectronics circuits
NEPTUNIX, a general program of simulation applied to nuclear reactors
International Nuclear Information System (INIS)
Bonnemay, A.; Dansac Bon, V.
1978-01-01
Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr
Scouting the feasibility of Monte Carlo reactor dynamics simulations
International Nuclear Information System (INIS)
Legrady, David; Hoogenboom, J. Eduard
2008-01-01
In this paper we present an overview of the methodological questions related to Monte Carlo simulation of time dependent power transients in nuclear reactors. Investigations using a small fictional 3D reactor with isotropic scattering and a single energy group we have performed direct Monte Carlo transient calculations with simulation of delayed neutrons and with and without thermal feedback. Using biased delayed neutron sampling and population control at time step boundaries calculation times were kept reasonably low. We have identified the initial source determination and the prompt chain simulations as key issues that require most attention. (authors)
Scouting the feasibility of Monte Carlo reactor dynamics simulations
Energy Technology Data Exchange (ETDEWEB)
Legrady, David [Forschungszentrum Dresden-Rossendorf, Dresden (Germany); Hoogenboom, J. Eduard [Delft University of Technology, Delft (Netherlands)
2008-07-01
In this paper we present an overview of the methodological questions related to Monte Carlo simulation of time dependent power transients in nuclear reactors. Investigations using a small fictional 3D reactor with isotropic scattering and a single energy group we have performed direct Monte Carlo transient calculations with simulation of delayed neutrons and with and without thermal feedback. Using biased delayed neutron sampling and population control at time step boundaries calculation times were kept reasonably low. We have identified the initial source determination and the prompt chain simulations as key issues that require most attention. (authors)
Numerical simulation of a possible counterexample to cosmic censorship
International Nuclear Information System (INIS)
Garfinkle, David
2004-01-01
A numerical simulation is presented here of the evolution of initial data of the kind that was conjectured by Hertog, Horowitz, and Maeda to be a violation of cosmic censorship. Those initial data are essentially a thick domain wall connecting two regions of anti-de Sitter space. The initial data have a free parameter that is the initial size of the wall. The simulation shows no violation of cosmic censorship, but rather the formation of a small black hole. The simulation described here is for a moderate wall size and leaves open the possibility that cosmic censorship might be violated for larger walls
3D numerical simulations of multiphase continental rifting
Naliboff, J.; Glerum, A.; Brune, S.
2017-12-01
Observations of rifted margin architecture suggest continental breakup occurs through multiple phases of extension with distinct styles of deformation. The initial rifting stages are often characterized by slow extension rates and distributed normal faulting in the upper crust decoupled from deformation in the lower crust and mantle lithosphere. Further rifting marks a transition to higher extension rates and coupling between the crust and mantle lithosphere, with deformation typically focused along large-scale detachment faults. Significantly, recent detailed reconstructions and high-resolution 2D numerical simulations suggest that rather than remaining focused on a single long-lived detachment fault, deformation in this phase may progress toward lithospheric breakup through a complex process of fault interaction and development. The numerical simulations also suggest that an initial phase of distributed normal faulting can play a key role in the development of these complex fault networks and the resulting finite deformation patterns. Motivated by these findings, we will present 3D numerical simulations of continental rifting that examine the role of temporal increases in extension velocity on rifted margin structure. The numerical simulations are developed with the massively parallel finite-element code ASPECT. While originally designed to model mantle convection using advanced solvers and adaptive mesh refinement techniques, ASPECT has been extended to model visco-plastic deformation that combines a Drucker Prager yield criterion with non-linear dislocation and diffusion creep. To promote deformation localization, the internal friction angle and cohesion weaken as a function of accumulated plastic strain. Rather than prescribing a single zone of weakness to initiate deformation, an initial random perturbation of the plastic strain field combined with rapid strain weakening produces distributed normal faulting at relatively slow rates of extension in both 2D and
Development of numerical models for Monte Carlo simulations of Th-Pb fuel assembly
Directory of Open Access Journals (Sweden)
Oettingen Mikołaj
2017-01-01
Full Text Available The thorium-uranium fuel cycle is a promising alternative against uranium-plutonium fuel cycle, but it demands many advanced research before starting its industrial application in commercial nuclear reactors. The paper presents the development of the thorium-lead (Th-Pb fuel assembly numerical models for the integral irradiation experiments. The Th-Pb assembly consists of a hexagonal array of ThO2 fuel rods and metallic Pb rods. The design of the assembly allows different combinations of rods for various types of irradiations and experimental measurements. The numerical model of the Th-Pb assembly was designed for the numerical simulations with the continuous energy Monte Carlo Burnup code (MCB implemented on the supercomputer Prometheus of the Academic Computer Centre Cyfronet AGH.
Understanding casing flow in Pelton turbines by numerical simulation
Rentschler, M.; Neuhauser, M.; Marongiu, J. C.; Parkinson, E.
2016-11-01
For rehabilitation projects of Pelton turbines, the flow in the casing may have an important influence on the overall performance of the machine. Water sheets returning on the jets or on the runner significantly reduce efficiency, and run-away speed depends on the flow in the casing. CFD simulations can provide a detailed insight into this type of flow, but these simulations are computationally intensive. As in general the volume of water in a Pelton turbine is small compared to the complete volume of the turbine housing, a single phase simulation greatly reduces the complexity of the simulation. In the present work a numerical tool based on the SPH-ALE meshless method is used to simulate the casing flow in a Pelton turbine. Using improved order schemes reduces the numerical viscosity. This is necessary to resolve the flow in the jet and on the casing wall, where the velocity differs by two orders of magnitude. The results are compared to flow visualizations and measurement in a hydraulic laboratory. Several rehabilitation projects proved the added value of understanding the flow in the Pelton casing. The flow simulation helps designing casing insert, not only to see their influence on the flow, but also to calculate the stress in the inserts. In some projects, the casing simulation leads to the understanding of unexpected behavior of the flow. One such example is presented where the backsplash of a deflector hit the runner, creating a reversed rotation of the runner.
Numerical simulations for impact damage detection in composites using vibrothermography
International Nuclear Information System (INIS)
Pieczonka, L J; Uhl, T; Szwedo, M; Staszewski, W J; Aymerich, F
2010-01-01
Composite materials are widely used in many engineering applications due to their high strength-to-weight ratios. However, it is well known that composites are susceptible to impact damage. Detection of impact damage is an important issue in maintenance of composite structures. Various non-destructive image-based techniques have been developed for damage detection in composite materials. These include vibrothermography that detects surface temperature changes due to heating associated with frictional energy dissipation by damage. In the present paper numerical simulations are used to investigate heat generation in a composite plate with impact damage in order to support damage detection analysis with vibrothermography. Explicit finite elements are used to model ultrasonic wave propagation in the damaged plate. Simulated delamination and cracks induce frictional heating in the plate. Coupled thermo-mechanical simulations are performed in high frequencies using commercial LS-Dyna finite element code. Very good qualitative agreement between measurements and simulations has been obtained. The area of increased temperature corresponds very well with the damaged area in both experiments and simulations. Numerical model has to be further refined in order to quantitatively match the experiments. The main issues of concern are frictional and thermal properties of composites. The final goal of these research efforts is to predict damage detection sensitivity of vibrothermography in real engineering applications based on numerical models.
Analysis of core and core barrel heat-up under conditions simulating severe reactor accidents
International Nuclear Information System (INIS)
Chellaiah, S.; Viskanta, R.; Ranganathan, P.; Anand, N.K.
1987-01-01
This paper reports on the development of a model for estimating the temperature distributions in the reactor core, core barrel, thermal shield and reactor pressure vessel of a PWR during an undercooling transient. A number of numerical calculations simulating the core uncovering of the TMI-2 reactor and the subsequent heat-up of the core have been performed. The results of the calculations show that the exothermic heat release due to Zircaloy oxidation contributes to the sharp heat-up of the core. However, the core barrel temperature rise which is driven by the temperature increase of the edge of the core (e.g., the core baffle) is very modest. The maximum temperature of the core barrel never exceeded 610 K (at a system pressure of 68 bar) after a 75 minute simulation following the start of core uncovering
Numerical Simulation of Hydrogen Combustion: Global Reaction Model and Validation
Energy Technology Data Exchange (ETDEWEB)
Zhang, Yun [School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an (China); Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY (United States); Liu, Yinhe, E-mail: yinheliu@mail.xjtu.edu.cn [School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an (China)
2017-11-20
Due to the complexity of modeling the combustion process in nuclear power plants, the global mechanisms are preferred for numerical simulation. To quickly perform the highly resolved simulations with limited processing resources of large-scale hydrogen combustion, a method based on thermal theory was developed to obtain kinetic parameters of global reaction mechanism of hydrogen–air combustion in a wide range. The calculated kinetic parameters at lower hydrogen concentration (C{sub hydrogen} < 20%) were validated against the results obtained from experimental measurements in a container and combustion test facility. In addition, the numerical data by the global mechanism (C{sub hydrogen} > 20%) were compared with the results by detailed mechanism. Good agreement between the model prediction and the experimental data was achieved, and the comparison between simulation results by the detailed mechanism and the global reaction mechanism show that the present calculated global mechanism has excellent predictable capabilities for a wide range of hydrogen–air mixtures.
Numerical simulation of particle settling and cohesion in liquid
Energy Technology Data Exchange (ETDEWEB)
Johno, Y; Nakashima, K; Shigematsu, T; Ono, B [SASEBO National College of Technology, 1-1 Okishin, Sasebo, Nagasaki, 857-1193 (Japan); Satomi, M, E-mail: yjohno@post.cc.sasebo.ac.j [Sony Semiconductor Kyushu Corporation, Kikuchigun, Kumamoto (Japan)
2009-02-01
In this study, the motions of particles and particle clusters in liquid were numerically simulated. The particles of two sizes (Dp=40mum and 20mum) settle while repeating cohesion and dispersion, and finally the sediment of particles are formed at the bottom of a hexahedron container which is filled up with pure water. The flow field was solved with the Navier-Stokes equations and the particle motions were solved with the Lagrangian-type motion equations, where the interaction between fluid and particles due to drag forces were taken into account. The collision among particles was calculated using Distinct Element Method (DEM), and the effects of cohesive forces by van der Waals force acting on particle contact points were taken into account. Numerical simulations were performed under conditions in still flow and in shear flow. It was found that the simulation results enable us to know the state of the particle settling and the particle condensation.
Numerical simulation of manual operation at MID stand control room
International Nuclear Information System (INIS)
Doca, C.; Dobre, A.; Predescu, D.; Mielcioiu, A.
2003-01-01
Since 2000 at INR Pitesti a package of software products devoted to numerical simulation of manual operations at fueling machine control room was developed. So far, specified, designed, worked out and implemented was the PUPITRU code. The following issues were solved: graphical aspects of specific computer - human operator interface; functional and graphical simulation of the whole associated equipment of the control desk components; implementation of the main notation as used in the automated schemes of the control desk in view of the fast identification of the switches, lamps, instrumentation, etc.; implementation within PUPITRU code of the entire data base used in the frame of MID tests; implementation of a number of about 1000 numerical simulation equations describing specific operational MID testing situations
Numerical simulation of small scale soft impact tests
International Nuclear Information System (INIS)
Varpasuo, Pentti
2008-01-01
This paper describes the small scale soft missile impact tests. The purpose of the test program is to provide data for the calibration of the numerical simulation models for impact simulation. In the experiments, both dry and fluid filled missiles are used. The tests with fluid filled missiles investigate the release speed and the droplet size of the fluid release. This data is important in quantifying the fire hazard of flammable liquid after the release. The spray release velocity and droplet size are also input data for analytical and numerical simulation of the liquid spread in the impact. The behaviour of the impact target is the second investigative goal of the test program. The response of reinforced and pre-stressed concrete walls is studied with the aid of displacement and strain monitoring. (authors)
Numerical Simulation of Hydrogen Combustion: Global Reaction Model and Validation
International Nuclear Information System (INIS)
Zhang, Yun; Liu, Yinhe
2017-01-01
Due to the complexity of modeling the combustion process in nuclear power plants, the global mechanisms are preferred for numerical simulation. To quickly perform the highly resolved simulations with limited processing resources of large-scale hydrogen combustion, a method based on thermal theory was developed to obtain kinetic parameters of global reaction mechanism of hydrogen–air combustion in a wide range. The calculated kinetic parameters at lower hydrogen concentration (C hydrogen < 20%) were validated against the results obtained from experimental measurements in a container and combustion test facility. In addition, the numerical data by the global mechanism (C hydrogen > 20%) were compared with the results by detailed mechanism. Good agreement between the model prediction and the experimental data was achieved, and the comparison between simulation results by the detailed mechanism and the global reaction mechanism show that the present calculated global mechanism has excellent predictable capabilities for a wide range of hydrogen–air mixtures.
Direct numerical simulations of turbulent lean premixed combustion
International Nuclear Information System (INIS)
Sankaran, Ramanan; Hawkes, Evatt R; Chen, Jacqueline H; Lu Tianfeng; Law, Chung K
2006-01-01
In recent years, due to the advent of high-performance computers and advanced numerical algorithms, direct numerical simulation (DNS) of combustion has emerged as a valuable computational research tool, in concert with experimentation. The role of DNS in delivering new Scientific insight into turbulent combustion is illustrated using results from a recent 3D turbulent premixed flame simulation. To understand the influence of turbulence on the flame structure, a 3D fully-resolved DNS of a spatially-developing lean methane-air turbulent Bunsen flame was performed in the thin reaction zones regime. A reduced chemical model for methane-air chemistry consisting of 13 resolved species, 4 quasi-steady state species and 73 elementary reactions was developed specifically for the current simulation. The data is analyzed to study possible influences of turbulence on the flame thickness. The results show that the average flame thickness increases, in qualitative agreement with several experimental results
Simulation tools and new developments of the molten salt fast reactor
International Nuclear Information System (INIS)
Heuer, D.; Merle-Lucotte, E.; Allibert, M.; Doligez, X.; Ghetta, V.
2010-01-01
In the MSFR (Molten Salt Fast Reactor), the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this design characteristic, the MSFR can thus operate with a widely varying fuel composition. Our reactor's studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for Bateman's equations computing the population of any nucleus inside any part of the reactor at any moment. The classical Bateman's equations have been modified by adding 2 terms representing the reprocessing capacities and an online addition. We have thus coupled neutronic and reprocessing simulation codes in a numerical tool used to calculate the extraction efficiencies of fission products, their location in the whole system and radioprotection issues. The very preliminary results show the potential of the neutronic-reprocessing coupling we have developed. We also show that these studies are limited by the uncertainties on the design and the knowledge of the chemical reprocessing processes. (A.C.)
Configuration Management File Manager Developed for Numerical Propulsion System Simulation
Follen, Gregory J.
1997-01-01
One of the objectives of the High Performance Computing and Communication Project's (HPCCP) Numerical Propulsion System Simulation (NPSS) is to provide a common and consistent way to manage applications, data, and engine simulations. The NPSS Configuration Management (CM) File Manager integrated with the Common Desktop Environment (CDE) window management system provides a common look and feel for the configuration management of data, applications, and engine simulations for U.S. engine companies. In addition, CM File Manager provides tools to manage a simulation. Features include managing input files, output files, textual notes, and any other material normally associated with simulation. The CM File Manager includes a generic configuration management Application Program Interface (API) that can be adapted for the configuration management repositories of any U.S. engine company.
International Nuclear Information System (INIS)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J.
2013-01-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)
Energy Technology Data Exchange (ETDEWEB)
Hussein, M.S.; Bonin, H.W.; Lewis, B.J., E-mail: mohamed.hussein@rmc.ca, E-mail: bonin-h@rmc.ca, E-mail: lewis-b@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)
2013-07-01
The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as Deuterium Critical Assembly, (DCA). The variations of the criticality factors and the coupling coefficients were investigated by changing of the water levels in the inner and outer cores. The numerical results of the model developed with MCNP5 code were validated and verified against published results and the mathematical model based on coupled reactor theory. (author)
International Nuclear Information System (INIS)
Tashakor, S.; Jahanfarnia, G.; Hashemi-Tilehnoee, M.
2010-01-01
Point reactor kinetics equations are solved numerically using one group of delayed neutrons and with fuel burn-up and temperature feedback included. To calculate the fraction of one-group delayed neutrons, a group of differential equations are solved by an implicit time method. Using point reactor kinetics equations, changes in mean neutrons density, temperature, and reactivity are calculated in different times during the reactor operation. The variation of reactivity, temperature, and maximum power with time are compared with the predictions by other methods.
International Nuclear Information System (INIS)
Pan, Dongqing; Chien Jen, Tien; Li, Tao; Yuan, Chris
2014-01-01
This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired
Energy Technology Data Exchange (ETDEWEB)
Pan, Dongqing; Chien Jen, Tien [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, Milwaukee, Wisconsin 53201 (United States); Li, Tao [School of Mechanical Engineering, Dalian University of Technology, Dalian 116024 (China); Yuan, Chris, E-mail: cyuan@uwm.edu [Department of Mechanical Engineering, University of Wisconsin-Milwaukee, 3200 North Cramer Street, Milwaukee, Wisconsin 53211 (United States)
2014-01-15
This paper characterizes the carrier gas flow in the atomic layer deposition (ALD) vacuum reactor by introducing Lattice Boltzmann Method (LBM) to the ALD simulation through a comparative study of two LBM models. Numerical models of gas flow are constructed and implemented in two-dimensional geometry based on lattice Bhatnagar–Gross–Krook (LBGK)-D2Q9 model and two-relaxation-time (TRT) model. Both incompressible and compressible scenarios are simulated and the two models are compared in the aspects of flow features, stability, and efficiency. Our simulation outcome reveals that, for our specific ALD vacuum reactor, TRT model generates better steady laminar flow features all over the domain with better stability and reliability than LBGK-D2Q9 model especially when considering the compressible effects of the gas flow. The LBM-TRT is verified indirectly by comparing the numerical result with conventional continuum-based computational fluid dynamics solvers, and it shows very good agreement with these conventional methods. The velocity field of carrier gas flow through ALD vacuum reactor was characterized by LBM-TRT model finally. The flow in ALD is in a laminar steady state with velocity concentrated at the corners and around the wafer. The effects of flow fields on precursor distributions, surface absorptions, and surface reactions are discussed in detail. Steady and evenly distributed velocity field contribute to higher precursor concentration near the wafer and relatively lower particle velocities help to achieve better surface adsorption and deposition. The ALD reactor geometry needs to be considered carefully if a steady and laminar flow field around the wafer and better surface deposition are desired.
Education and training by utilizing irradiation test reactor simulator
International Nuclear Information System (INIS)
Eguchi, Shohei; Koike, Sumio; Takemoto, Noriyuki; Tanimoto, Masataka; Kusunoki, Tsuyoshi
2016-01-01
The Japan Atomic Energy Agency, at its Japan Materials Testing Reactor (JMTR), completed an irradiation test reactor simulator in May 2012. This simulator simulates the operation, irradiation test, abnormal transient change during operation, and accident progress events, etc., and is able to perform operation training on reactor and irradiation equipment corresponding to the above simulations. This simulator is composed of a reactor control panel, process control panel, irradiation equipment control panel, instructor control panel, large display panel, and compute server. The completed simulator has been utilized in the education and training of JMTR operators for the purpose of the safe and stable operation of JMTR and the achievement of high operation rate after resuming operation. For the education and training, an education and training curriculum has been prepared for use in not only operation procedures at the time of normal operation, but also learning of fast and accurate response in case of accident events. In addition, this simulator is also being used in operation training for the purpose of contributing to the cultivation of human resources for atomic power in and out of Japan. (A.O.)
Development of a research nuclear reactor simulator using LABVIEW®
Energy Technology Data Exchange (ETDEWEB)
Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)
2015-07-01
The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)
Development of a research nuclear reactor simulator using LABVIEW®
International Nuclear Information System (INIS)
Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade
2015-01-01
The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)
International Nuclear Information System (INIS)
McCulloch, R.W.; Lovell, R.T.; Post, D.W.; Snyder, S.D.
1980-01-01
Variable width ribbon heating elements have been fabricated which provide a chopped cosine, variable heat flux profile for fuel rod simulators used in test loops by the Breeder Reactor Program Thermal Hydraulic Out-of-Reactor Safety test facility and the Gas-Cooled Fast Breeder Reactor Core Flow Test Loop. Thermal, mechanical, and electrical design considerations result in the derivation of an analytical expression for the ribbon contours. From this, the ribbons are machined and wound on numerically controlled equipment. Postprocessing and inspection results in a wound, variable width ribbon with the precise dimensional, electrical, and mechanical properties needed for use in fuel pin simulators
Nonlinear punctual dynamic applied to simulation of PWR type reactors
International Nuclear Information System (INIS)
Cysne, F.S.
1978-01-01
In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)
A study on naphtha catalytic reforming reactor simulation and analysis.
Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei
2005-06-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.
A study on naphtha catalytic reforming reactor simulation and analysis
Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei
2005-01-01
A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation uni...
Non-linear punctual kinetics applied to PWR reactors simulation
International Nuclear Information System (INIS)
Cysne, F.S.
1978-11-01
In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model for the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and CSMP using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (Author) [pt
Acoustic emission from fuel pellets in a simulated reactor environment
International Nuclear Information System (INIS)
Kupperman, D.S.; Kennedy, C.R.; Reimann, K.J.
1977-01-01
Thermal-shock damage of nuclear reactor fuel pellets in a simulated reactor environment has been correlated with acoustic-emission data obtained from sensors placed on extensions of the electrical feedthroughs. Ringdown counts, rms output data, and event-location data has been acquired for experiments carried out with single pellets as well as multiple pellet stacks. These tests have shown that acoustic-emission monitoring can provide information indicating the onset and the extent of cracking
Energy Technology Data Exchange (ETDEWEB)
Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)
2000-03-01
JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)
Numerical simulation investigation on centrifugal compressor performance of turbocharger
International Nuclear Information System (INIS)
Li, Jie; Yin, Yuting; Li, Shuqi; Zhang, Jizhong
2013-01-01
In this paper, the mathematical model of the flow filed in centrifugal compressor of turbocharger was studied. Based on the theory of computational fluid dynamics (CFD), performance curves and parameter distributions of the compressor were obtained from the 3-D numerical simulation by using CFX. Meanwhile, the influences of grid number and distribution on compressor performance were investigated, and numerical calculation method was analyzed and validated, through combining with test data. The results obtained show the increase of the grid number has little influence on compressor performance while the grid number of single-passage is above 300,000. The results also show that the numerical calculation mass flow rate of compressor choke situation has a good consistent with test results, and the maximum difference of the diffuser exit pressure between simulation and experiment decrease to 3.5% with the assumption of 6 kPa additional total pressure loss at compressor inlet. The numerical simulation method in this paper can be used to predict compressor performance, and the difference of total pressure ratio between calculation and test is less than 7%, and the total-to-total efficiency also have a good consistent with test.
Numerical simulation investigation on centrifugal compressor performance of turbocharger
Energy Technology Data Exchange (ETDEWEB)
Li, Jie [China Iron and Steel Research Institute Group, Beijing (China); Yin, Yuting [China North Engine Research Institute, Datong (China); Li, Shuqi; Zhang, Jizhong [Science and Technology Diesel Engine Turbocharging Laboratory, Datong (China)
2013-06-15
In this paper, the mathematical model of the flow filed in centrifugal compressor of turbocharger was studied. Based on the theory of computational fluid dynamics (CFD), performance curves and parameter distributions of the compressor were obtained from the 3-D numerical simulation by using CFX. Meanwhile, the influences of grid number and distribution on compressor performance were investigated, and numerical calculation method was analyzed and validated, through combining with test data. The results obtained show the increase of the grid number has little influence on compressor performance while the grid number of single-passage is above 300,000. The results also show that the numerical calculation mass flow rate of compressor choke situation has a good consistent with test results, and the maximum difference of the diffuser exit pressure between simulation and experiment decrease to 3.5% with the assumption of 6 kPa additional total pressure loss at compressor inlet. The numerical simulation method in this paper can be used to predict compressor performance, and the difference of total pressure ratio between calculation and test is less than 7%, and the total-to-total efficiency also have a good consistent with test.
Simulation of Reforming Reactor Tube: Quantifying Catalyst Pellet's Effectiveness Factor
Da Cruz, Flavio Eduardo
2016-01-01
In this work, a consistent mathematical model to simulate a spherical catalytic pellet and a Packed-Bed Reactor (PBR) is develop. The Dusty Gas Model (DGM) is applied to the calculation of the diffusive fluxes in the porous media. Simulations are executed considering hydrogen production from steam methane reforming. Species’ diffusivities are calculated using data from literature as well as the values for tortuosity and porosity. The pellet simulation is performed considering mass, species, m...
Numerical simulation in material science: principles and applications
International Nuclear Information System (INIS)
Ruste, Jacky
2006-06-01
The objective is here to describe the main simulation techniques currently used in material science. After a presentation of the concepts of modelling and simulation, of their objectives and uses, of the issue of simulation scale, and of means of numeric simulation, the author addresses simulations performed at a nano-scopic scale: 'ab-initio' methods, molecular dynamics, examples of applications of ab-initio methods to energy issues or to the study of surface properties of nano-materials. The next chapter addresses various Monte Carlo methods (Metropolis, atomic kinetics, objects kinetics, transport with the simulation of particle trajectories, generation of random numbers). The next parts address simulations performed at a mesoscopic scale (simulation and microstructure, phase field methods, dynamics of discrete dislocations, homogeneous chemical kinetics) and at a macroscopic scale (medium discretization with the notion of mesh, simulation of structure mechanics and of fluid behaviour). The issues of code coupling and scale coupling are then discussed. The last part proposes an overview of virtual metallurgy and modelling of industrial processes (welding, vacuum arc re-fusion, rolling, forming)
Numerical simulation of heat transfer in metal foams
Gangapatnam, Priyatham; Kurian, Renju; Venkateshan, S. P.
2018-02-01
This paper reports a numerical study of forced convection heat transfer in high porosity aluminum foams. Numerical modeling is done considering both local thermal equilibrium and non local thermal equilibrium conditions in ANSYS-Fluent. The results of the numerical model were validated with experimental results, where air was forced through aluminum foams in a vertical duct at different heat fluxes and velocities. It is observed that while the LTE model highly under predicts the heat transfer in these foams, LTNE model predicts the Nusselt number accurately. The novelty of this study is that once hydrodynamic experiments are conducted the permeability and porosity values obtained experimentally can be used to numerically simulate heat transfer in metal foams. The simulation of heat transfer in foams is further extended to find the effect of foam thickness on heat transfer in metal foams. The numerical results indicate that though larger foam thicknesses resulted in higher heat transfer coefficient, this effect weakens with thickness and is negligible in thick foams.
Numerical simulation of gas metal arc welding parametrical study
International Nuclear Information System (INIS)
Szanto, M.; Gilad, I.; Shai, I.; Quinn, T.P.
2002-01-01
The Gas Metal Arc Welding (GMAW) is a widely used welding process in the industry. The process variables are usually determined through extensive experiments. Numerical simulation, reduce the cost and extends the understanding of the process. In the present work, a versatile model for numerical simulation of GMAW is presented. The model provides the basis for fundamental understanding of the process. The model solves the magneto-hydrodynamic equations for the flow and temperature fields of the molten electrode and the plasma simultaneously, to form a fully coupled model. A commercial CFD code was extended to include the effects of radiation, Lorentz forces, Joule heating and thermoelectric effects. The geometry of the numerical model assembled to fit an experimental apparatus. To demonstrate the method, an aluminum electrode was modeled in a pure argon arc. Material properties and welding parameters are the input variables in the numerical model. In a typical process, the temperature distribution of the plasma is over 15000 K, resulting high non-linearity of the material properties. Moreover, there is high uncertainty in the available property data, at that range of temperatures. Therefore, correction factors were derived for the material properties to adjust between the numerical and the experimental results. Using the compensated properties, parametric study was performed. The effects of the welding parameters on the process, such the working voltage, electrode feed rate and shielding gas flow, were derived. The principal result of the present work is the ability to predict, by numerical simulation, the mode, size and frequency of the metal transferred from the electrode, which is the main material and energy source for the welding pool in GMAW
Numerical simulation support to the ESA/THOR mission
Valentini, F.; Servidio, S.; Perri, S.; Perrone, D.; De Marco, R.; Marcucci, M. F.; Daniele, B.; Bruno, R.; Camporeale, E.
2016-12-01
THOR is a spacecraft concept currently undergoing study phase as acandidate for the next ESA medium size mission M4. THOR has been designedto solve the longstanding physical problems of particle heating andenergization in turbulent plasmas. It will provide high resolutionmeasurements of electromagnetic fields and particle distribution functionswith unprecedented resolution, with the aim of exploring the so-calledkinetic scales. We present the numerical simulation framework which is supporting the THOR mission during the study phase. The THOR teamincludes many scientists developing and running different simulation codes(Eulerian-Vlasov, Particle-In-Cell, Gyrokinetics, Two-fluid, MHD, etc.),addressing the physics of plasma turbulence, shocks, magnetic reconnectionand so on.These numerical codes are being used during the study phase, mainly withthe aim of addressing the following points:(i) to simulate the response of real particle instruments on board THOR, byemploying an electrostatic analyser simulator which mimics the response ofthe CSW, IMS and TEA instruments to the particle velocity distributions ofprotons, alpha particle and electrons, as obtained from kinetic numericalsimulations of plasma turbulence.(ii) to compare multi-spacecraft with single-spacecraft configurations inmeasuring current density, by making use of both numerical models ofsynthetic turbulence and real data from MMS spacecraft.(iii) to investigate the validity of the Taylor hypothesis indifferent configurations of plasma turbulence
Graphics interfaces and numerical simulations: Mexican Virtual Solar Observatory
Hernández, L.; González, A.; Salas, G.; Santillán, A.
2007-08-01
Preliminary results associated to the computational development and creation of the Mexican Virtual Solar Observatory (MVSO) are presented. Basically, the MVSO prototype consists of two parts: the first, related to observations that have been made during the past ten years at the Solar Observation Station (EOS) and at the Carl Sagan Observatory (OCS) of the Universidad de Sonora in Mexico. The second part is associated to the creation and manipulation of a database produced by numerical simulations related to solar phenomena, we are using the MHD ZEUS-3D code. The development of this prototype was made using mysql, apache, java and VSO 1.2. based GNU and `open source philosophy'. A graphic user interface (GUI) was created in order to make web-based, remote numerical simulations. For this purpose, Mono was used, because it is provides the necessary software to develop and run .NET client and server applications on Linux. Although this project is still under development, we hope to have access, by means of this portal, to other virtual solar observatories and to be able to count on a database created through numerical simulations or, given the case, perform simulations associated to solar phenomena.
Direct numerical simulations of nucleate boiling flows of binary mixtures
International Nuclear Information System (INIS)
Didier Jamet; Celia Fouillet
2005-01-01
Full text of publication follows: Better understand the origin and characteristics of boiling crisis is still a scientific challenge despite many years of valuable studies. One of the reasons why boiling crisis is so difficult to understand is that local and coupled physical phenomena are believed to play a key role in the trigger of instabilities which lead to the dry out of large portions of the heated solid phase. Nucleate boiling of a single bubble is fairly well understood compared to boiling crisis. Therefore, the numerical simulation of a single bubble growth during nucleate boiling is a good candidate to evaluate the capabilities of a numerical method to deal with complex liquid-vapor phenomena with phase-change and eventually to tackle the boiling crisis problem. In this paper, we present results of direct numerical simulations of nucleate boiling. The numerical method used is the second gradient method, which is a diffuse interface method dedicated to liquid vapor flows with phase-change. This study is not intended to provide quantitative results, partly because all the simulations are two-dimensional. However, particular attention is paid to the influence of some parameters on the main features of nucleate boiling, i.e. the radius of departure and the frequency of detachment of bubbles. In particular, we show that, as the contact angle increases, the radius of departure increases whereas the frequency of detachment decreases. Moreover, the influence of the existence of quasi non-condensable gas is studied. Numerical results show an important decrease of the heat exchange coefficient when a small amount of a quasi non-condensable gas is added to the pure liquid-vapor water system. This result is in agreement with experimental observations. Beyond these qualitative results, this numerical study allows to get insight into some important physical phenomena and to confirm that during nucleate boiling, large scale quantities are influenced by small scale
Numerical simulation of the RISOe1-airfoil dynamic stall
Energy Technology Data Exchange (ETDEWEB)
Bertagnolio, F.; Soerensen, N. [Risoe National Lab., Wind Energy and Atmospheric Physics Dept., Roskilde (Denmark)
1997-12-31
In this paper we are concerned with the numerical computation of the dynamic stall that occur in the viscous flowfield over an airfoil. These results are compared to experimental data that were obtained with the new designed RISOe1-airfoil, both for a motionless airfoil and for a pitching motion. Moreover, we present some numerical computations of the plunging and lead-lag motions. We also investigate the possibility of using the pitching motion to simulate the plunging and lead-lag situations. (au)
Modeling and numerical simulations of the influenced Sznajd model
Karan, Farshad Salimi Naneh; Srinivasan, Aravinda Ramakrishnan; Chakraborty, Subhadeep
2017-08-01
This paper investigates the effects of independent nonconformists or influencers on the behavioral dynamic of a population of agents interacting with each other based on the Sznajd model. The system is modeled on a complete graph using the master equation. The acquired equation has been numerically solved. Accuracy of the mathematical model and its corresponding assumptions have been validated by numerical simulations. Regions of initial magnetization have been found from where the system converges to one of two unique steady-state PDFs, depending on the distribution of influencers. The scaling property and entropy of the stationary system in presence of varying level of influence have been presented and discussed.
Numerical simulation of water quality in Yangtze Estuary
Directory of Open Access Journals (Sweden)
Xi Li
2009-12-01
Full Text Available In order to monitor water quality in the Yangtze Estuary, water samples were collected and field observation of current and velocity stratification was carried out using a shipboard acoustic Doppler current profiler (ADCP. Results of two representative variables, the temporal and spatial variation of new point source sewage discharge as manifested by chemical oxygen demand (COD and the initial water quality distribution as manifested by dissolved oxygen (DO, were obtained by application of the Environmental Fluid Dynamics Code (EFDC with solutions for hydrodynamics during tides. The numerical results were compared with field data, and the field data provided verification of numerical application: this numerical model is an effective tool for water quality simulation. For point source discharge, COD concentration was simulated with an initial value in the river of zero. The simulated increments and distribution of COD in the water show acceptable agreement with field data. The concentration of DO is much higher in the North Branch than in the South Branch due to consumption of oxygen in the South Branch resulting from discharge of sewage from Shanghai. The DO concentration is greater in the surface layer than in the bottom layer. The DO concentration is low in areas with a depth of less than 20 m, and high in areas between the 20-m and 30-m isobaths. It is concluded that the numerical model is valuable in simulation of water quality in the case of specific point source pollutant discharge. The EFDC model is also of satisfactory accuracy in water quality simulation of the Yangtze Estuary.
Energy Technology Data Exchange (ETDEWEB)
Singh, Narendra K., E-mail: nksingh_chikki@yahoo.com [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Badodkar, Deepak N. [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India); Homi Bhabha National Institute, Anushaktinagar, Mumbai 400094 (India); Singh, Manjit [Division of Remote Handling and Robotics, Bhabha Atomic Research Centre, Mumbai 400085 (India)
2014-07-01
Highlights: • Hydraulic dashpot performance is studied numerically as well as experimentally. • Instantaneous pressure built-up in the dashpot is mainly contributing for damping of freely falling shut-off rod at the end of its travel. • At elevated temperature, dashpot pressure does not reduce in proportion to the reduction in viscosity. • ‘C’ grove in the dashpot shaft flattens the pressure peak and shifts it toward the end of operation. - Abstract: Hydraulic dashpot design for shut-off rod drive mechanism application in a nuclear reactor has been analyzed both numerically and experimentally in this paper. Finite element commercial code COMSOL Multiphysics 4.3 has been used for numerical analysis. Experimental validation has been done at two different cases. Experimental test set-ups and hydraulic dashpot constructions have been described in detail. Various combinations of dashpot oil viscosity and clearance thickness have been analyzed. Important experimental results are also presented and discussed. Pressure distributions in the dashpot chambers obtained from COMSOL are given for both the set-ups. Numerical and experimental results are compared. Dashpot designs have been qualified after detailed analysis and testing on full-scale test stations simulating actual reactor conditions (except radiation)
International Nuclear Information System (INIS)
Singh, Narendra K.; Badodkar, Deepak N.; Singh, Manjit
2014-01-01
Highlights: • Hydraulic dashpot performance is studied numerically as well as experimentally. • Instantaneous pressure built-up in the dashpot is mainly contributing for damping of freely falling shut-off rod at the end of its travel. • At elevated temperature, dashpot pressure does not reduce in proportion to the reduction in viscosity. • ‘C’ grove in the dashpot shaft flattens the pressure peak and shifts it toward the end of operation. - Abstract: Hydraulic dashpot design for shut-off rod drive mechanism application in a nuclear reactor has been analyzed both numerically and experimentally in this paper. Finite element commercial code COMSOL Multiphysics 4.3 has been used for numerical analysis. Experimental validation has been done at two different cases. Experimental test set-ups and hydraulic dashpot constructions have been described in detail. Various combinations of dashpot oil viscosity and clearance thickness have been analyzed. Important experimental results are also presented and discussed. Pressure distributions in the dashpot chambers obtained from COMSOL are given for both the set-ups. Numerical and experimental results are compared. Dashpot designs have been qualified after detailed analysis and testing on full-scale test stations simulating actual reactor conditions (except radiation)
Application of a Russian nuclear reactor simulator VVER-1000
International Nuclear Information System (INIS)
Lopez-Peniche S, A.; Salazar S, E.
2012-10-01
The objective of the present work is to give to know the most important characteristics in the Russian nuclear reactor of pressurized light water VVER-1000, doing emphasis in the differences that has with the western equivalent the reactor PWR in the design and the safety systems. Therefore, a description of the computerized simulation of the reactor VVER-1000 developed by the company Eniko TSO that the International Atomic of Energy Agency distributes to the states members with academic purposes will take place. The simulator includes mathematical models that represent to the essential systems in the real nuclear power plant, for what is possible to reproduce common faults and transitory characteristic of the nuclear industry with a behavior sufficiently attached to the reality. In this work is analyzed the response of the system before a turbine shot. After the accident in the nuclear power plant of Three Mile Island (US) they have been carried out improvements in the design of the reactor PWR and their safety systems. To know the reach and the limitations of the program, the events that gave place to this accident will be reproduced in the simulator VVER-1000. With base to the results of the simulation we will conclude that so reliable is the response of the safety system of this reactor. (Author)
Expert System Architecture for Rocket Engine Numerical Simulators: A Vision
Mitra, D.; Babu, U.; Earla, A. K.; Hemminger, Joseph A.
1998-01-01
Simulation of any complex physical system like rocket engines involves modeling the behavior of their different components using mostly numerical equations. Typically a simulation package would contain a set of subroutines for these modeling purposes and some other ones for supporting jobs. A user would create an input file configuring a system (part or whole of a rocket engine to be simulated) in appropriate format understandable by the package and run it to create an executable module corresponding to the simulated system. This module would then be run on a given set of input parameters in another file. Simulation jobs are mostly done for performance measurements of a designed system, but could be utilized for failure analysis or a design job such as inverse problems. In order to use any such package the user needs to understand and learn a lot about the software architecture of the package, apart from being knowledgeable in the target domain. We are currently involved in a project in designing an intelligent executive module for the rocket engine simulation packages, which would free any user from this burden of acquiring knowledge on a particular software system. The extended abstract presented here will describe the vision, methodology and the problems encountered in the project. We are employing object-oriented technology in designing the executive module. The problem is connected to the areas like the reverse engineering of any simulation software, and the intelligent systems for simulation.
Parallelization and automatic data distribution for nuclear reactor simulations
Energy Technology Data Exchange (ETDEWEB)
Liebrock, L.M. [Liebrock-Hicks Research, Calumet, MI (United States)
1997-07-01
Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed.
Parallelization and automatic data distribution for nuclear reactor simulations
International Nuclear Information System (INIS)
Liebrock, L.M.
1997-01-01
Detailed attempts at realistic nuclear reactor simulations currently take many times real time to execute on high performance workstations. Even the fastest sequential machine can not run these simulations fast enough to ensure that the best corrective measure is used during a nuclear accident to prevent a minor malfunction from becoming a major catastrophe. Since sequential computers have nearly reached the speed of light barrier, these simulations will have to be run in parallel to make significant improvements in speed. In physical reactor plants, parallelism abounds. Fluids flow, controls change, and reactions occur in parallel with only adjacent components directly affecting each other. These do not occur in the sequentialized manner, with global instantaneous effects, that is often used in simulators. Development of parallel algorithms that more closely approximate the real-world operation of a reactor may, in addition to speeding up the simulations, actually improve the accuracy and reliability of the predictions generated. Three types of parallel architecture (shared memory machines, distributed memory multicomputers, and distributed networks) are briefly reviewed as targets for parallelization of nuclear reactor simulation. Various parallelization models (loop-based model, shared memory model, functional model, data parallel model, and a combined functional and data parallel model) are discussed along with their advantages and disadvantages for nuclear reactor simulation. A variety of tools are introduced for each of the models. Emphasis is placed on the data parallel model as the primary focus for two-phase flow simulation. Tools to support data parallel programming for multiple component applications and special parallelization considerations are also discussed
Numerical simulation and optimization of nickel-hydrogen batteries
Yu, Li-Jun; Qin, Ming-Jun; Zhu, Peng; Yang, Li
2008-05-01
A three-dimensional, transient numerical model of an individual pressure vessel (IPV) nickel-hydrogen battery has been developed based on energy conservation law, mechanisms of heat and mass transfer, and electrochemical reactions in the battery. The model, containing all components of a battery including the battery shell, was utilized to simulate the transient temperature of the battery, using computational fluid dynamics (CFD) technology. The comparison of the model prediction and experimental data shows a good agreement, which means that the present model can be used for the engineering design and parameter optimization of nickel-hydrogen batteries in aerospace power systems. Two kinds of optimization schemes were provided and evaluated by the simulated temperature field. Based on the model, the temperature simulation during five successive periods in a designed space battery was conducted and the simulation results meet the requirement of safe operation.
Numerical simulation of nuclear power plants on multiprocessor systems
International Nuclear Information System (INIS)
Boeer, R.; Finnemann, H.; Mueller, R.; Krapf, J.
1990-04-01
The present final report gives an account of work performed and results obtained at Siemens/KWU within the frame of SUPRENUM project. The coupled neutron kinetics - thermal hydraulics program system 3D-SIM for the calculation of steady-state and transient conditions in the core and in the coolant loops of light water reactors has been developed. Making efficient use of the parallel computing architecture required the implementation of novel, purpose-built software concepts which can be characterized by multigrid algorithms and parallelization by domain decomposition. The modules treating the reactor core are 3D-NEUT for neutron kinetics and 3D-THERM for thermal hydraulics, the remaining system and control components being represented by the module NLOOP. This modular structure is a main feature of 3D-SIM which allows a flexible choice of applications, employing modules individually or in combinations to fulfil a specific calculation task. In particular, 3D-SIM reactor core representation is fully 3-dimensional for both neutron diffusion and thermal-hydraulic equations, allowing coupled calculations and evaluation of crossflow between fuel elements or subchannels. Including also the loops and plant models, 3D-SIM is a powerful analytical tool for simulating pressurized plant response to a wide spectrum of events. (orig.) With 25 refs., 11 figs [de
GPU based numerical simulation of core shooting process
Directory of Open Access Journals (Sweden)
Yi-zhong Zhang
2017-11-01
Full Text Available Core shooting process is the most widely used technique to make sand cores and it plays an important role in the quality of sand cores. Although numerical simulation can hopefully optimize the core shooting process, research on numerical simulation of the core shooting process is very limited. Based on a two-fluid model (TFM and a kinetic-friction constitutive correlation, a program for 3D numerical simulation of the core shooting process has been developed and achieved good agreements with in-situ experiments. To match the needs of engineering applications, a graphics processing unit (GPU has also been used to improve the calculation efficiency. The parallel algorithm based on the Compute Unified Device Architecture (CUDA platform can significantly decrease computing time by multi-threaded GPU. In this work, the program accelerated by CUDA parallelization method was developed and the accuracy of the calculations was ensured by comparing with in-situ experimental results photographed by a high-speed camera. The design and optimization of the parallel algorithm were discussed. The simulation result of a sand core test-piece indicated the improvement of the calculation efficiency by GPU. The developed program has also been validated by in-situ experiments with a transparent core-box, a high-speed camera, and a pressure measuring system. The computing time of the parallel program was reduced by nearly 95% while the simulation result was still quite consistent with experimental data. The GPU parallelization method can successfully solve the problem of low computational efficiency of the 3D sand shooting simulation program, and thus the developed GPU program is appropriate for engineering applications.
Computational fluid dynamics simulations of light water reactor flows
International Nuclear Information System (INIS)
Tzanos, C.P.; Weber, D.P.
1999-01-01
Advances in computational fluid dynamics (CFD), turbulence simulation, and parallel computing have made feasible the development of three-dimensional (3-D) single-phase and two-phase flow CFD codes that can simulate fluid flow and heat transfer in realistic reactor geometries with significantly reduced reliance, especially in single phase, on empirical correlations. The objective of this work was to assess the predictive power and computational efficiency of a CFD code in the analysis of a challenging single-phase light water reactor problem, as well as to identify areas where further improvements are needed
Simulated annealing algorithm for reactor in-core design optimizations
International Nuclear Information System (INIS)
Zhong Wenfa; Zhou Quan; Zhong Zhaopeng
2001-01-01
A nuclear reactor must be optimized for in core fuel management to make full use of the fuel, to reduce the operation cost and to flatten the power distribution reasonably. The author presents a simulated annealing algorithm. The optimized objective function and the punishment function were provided for optimizing the reactor physics design. The punishment function was used to practice the simulated annealing algorithm. The practical design of the NHR-200 was calculated. The results show that the K eff can be increased by 2.5% and the power distribution can be flattened
Direct numerical simulations of gas-liquid multiphase flows
Tryggvason, Grétar; Zaleski, Stéphane
2011-01-01
Accurately predicting the behaviour of multiphase flows is a problem of immense industrial and scientific interest. Modern computers can now study the dynamics in great detail and these simulations yield unprecedented insight. This book provides a comprehensive introduction to direct numerical simulations of multiphase flows for researchers and graduate students. After a brief overview of the context and history the authors review the governing equations. A particular emphasis is placed on the 'one-fluid' formulation where a single set of equations is used to describe the entire flow field and
Numerical simulation of flow behavior in tight lattice rod bundle
International Nuclear Information System (INIS)
Yu Yiqi; Yang Yanhua; Gu Hanyang; Cheng Xu; Song Xiaoming; Wang Xiaojun
2009-01-01
The Numerical investigation is performed on the air turbulent flow in triangular rod bundle array. Based on the experimental data, the eddy viscosity turbulent model and the Reynold stress turbulent model are evaluated to simulate the flow behavior in the tight lattice. The results show that SSG Reynolds Stress Model has shown superior predictive performance than other Reynolds-stress models, which indicates that the simulation of the anisotropy of the turbulence is significant in the tight lattice. The result with different Reynolds number and geometry shows that the magnitude of the secondary flow is almost independent of the Reynolds number, but it increases with the decrease of the P/D. (authors)
Numerical simulation of tornado-borne missile impact
International Nuclear Information System (INIS)
Tu, D.K.; Murray, R.C.
1977-01-01
The feasibility of using a finite element procedure to examine the impact phenomenon of a tornado-borne missile impinging on a reinforced concrete barrier was assessed. The major emphasis of this study was to simulate the impact of a nondeformable missile. Several series of simulations were run, using an 8-in.-dia steel slug as the impacting missile. The numerical results were then compared with experimental field tests and empirical formulas. The work is in support of tornado design practices for fuel reprocessing and fuel fabrication plants
Numerical Simulation of Cast Distortion in Gas Turbine Engine Components
International Nuclear Information System (INIS)
Inozemtsev, A A; Dubrovskaya, A S; Dongauser, K A; Trufanov, N A
2015-01-01
In this paper the process of multiple airfoilvanes manufacturing through investment casting is considered. The mathematical model of the full contact problem is built to determine stress strain state in a cast during the process of solidification. Studies are carried out in viscoelastoplastic statement. Numerical simulation of the explored process is implemented with ProCASTsoftware package. The results of simulation are compared with the real production process. By means of computer analysis the optimization of technical process parameters is done in order to eliminate the defect of cast walls thickness variation. (paper)
Numerical simulation of internal reconnection event in spherical tokamak
International Nuclear Information System (INIS)
Hayashi, Takaya; Mizuguchi, Naoki; Sato, Tetsuya
1999-07-01
Three-dimensional magnetohydrodynamic simulations are executed in a full toroidal geometry to clarify the physical mechanisms of the Internal Reconnection Event (IRE), which is observed in the spherical tokamak experiments. The simulation results reproduce several main properties of IRE. Comparison between the numerical results and experimental observation indicates fairly good agreements regarding nonlinear behavior, such as appearance of localized helical distortion, appearance of characteristic conical shape in the pressure profile during thermal quench, and subsequent appearance of the m=2/n=1 type helical distortion of the torus. (author)
Numerical simulation of void growth under dynamic loading
International Nuclear Information System (INIS)
Iqbal, A.
1996-01-01
Following a brief general review of developments in material behavior under high strain rates, a cylindrical cell surrounding a spherical void in OFHC copper is numerically simulated by Zerri-Armstrong model. This simulation results show that the plastic deformation tends to be concentrated in the vicinity of voids either in the axial or transverse direction depending upon the stress state. This event is associated with the accelerated void through accompanying coalescence causing ductile fracture. A3-node triangular mesh generation code used as input for finite element code is developed by a 'Central Generation' technique. (author)
Numerical simulation of low Mach number reacting flows
International Nuclear Information System (INIS)
Bell, J B; Aspden, A J; Day, M S; Lijewski, M J
2007-01-01
Using examples from active research areas in combustion and astrophysics, we demonstrate a computationally efficient numerical approach for simulating multiscale low Mach number reacting flows. The method enables simulations that incorporate an unprecedented range of temporal and spatial scales, while at the same time, allows an extremely high degree of reaction fidelity. Sample applications demonstrate the efficiency of the approach with respect to a traditional time-explicit integration method, and the utility of the methodology for studying the interaction of turbulence with terrestrial and astrophysical flame structures
Numerical simulation of the accident of Three Mile Island
International Nuclear Information System (INIS)
Perrin, M.H.; Kastelanski, P.
1981-01-01
The chief object of the present study was to assess the ability of our numerical code for the dynamic behavior of power plants, SICLE, to handle the simulation of small accidents in PWRs. In the first part of the paper the authors introduce the main principles, equations and numerical methods of the code. In the second part those of the elements of Three Mile Island Power Plant which were simulated, the different phases of the accident and the results obtained with the code are described. These results are compared to the values recorded in the plant and generally a good agreement is found (for instance the primary pressure). As a conclusion SICLE is the minimum code for representing accidents such as Three Mile Island; its main advantage lies in its ability to take into account all the elements of the plant which are important in the study
Numerical simulation of draft tube flow of a bulb turbine
Energy Technology Data Exchange (ETDEWEB)
Coelho, J.G. [Federal University of Triangulo Mineiro, Institute of Technological and Exact Sciences, Avenida Doutor Randolfo Borges Junior, 1250 – Uberaba – MG (Brazil); Brasil, A.C.P. Jr. [University of Brasilia, Department of Mechanical Engineering, Campus Darcy Ribeiro, Brasilia – DF (Brazil)
2013-07-01
In this work a numerical study of draft tube of a bulb hydraulic turbine is presented, where a new geometry is proposed. This new proposal of draft tube has the unaffected ratio area, a great reduction in his length and approximately the same efficiency of the draft tube conventionally used. The numerical simulations were obtained in commercial software of calculation of flow (CFX-14), using the turbulence model SST, that allows a description of the field fluid dynamic near to the wall. The simulation strategy has an intention of identifying the stall of the boundary layer precisely limits near to the wall and recirculations in the central part, once those are the great causes of the decrease of efficiency of a draft tube. Finally, it is obtained qualitative and quantitative results about the flow in draft tubes.
Numerical simulation of the circulation of the atmosphere of Titan
Hourdin, F.; Levan, P.; Talagrand, O.; Courtin, Regis; Gautier, Daniel; Mckay, Christopher P.
1992-01-01
A three dimensional General Circulation Model (GCM) of Titan's atmosphere is described. Initial results obtained with an economical two dimensional (2D) axisymmetric version of the model presented a strong superrotation in the upper stratosphere. Because of this result, a more general numerical study of superrotation was started with a somewhat different version of the GCM. It appears that for a slowly rotating planet which strongly absorbs solar radiation, circulation is dominated by global equator to pole Hadley circulation and strong superrotation. The theoretical study of this superrotation is discussed. It is also shown that 2D simulations systemically lead to instabilities which make 2D models poorly adapted to numerical simulation of Titan's (or Venus) atmosphere.
Three-dimensional numerical simulation during laser processing of CFRP
Ohkubo, Tomomasa; Sato, Yuji; Matsunaga, Ei-ichi; Tsukamoto, Masahiro
2017-09-01
We performed three-dimensional numerical simulation about laser processing of carbon-fiber-reinforced plastic (CFRP) using OpenFOAM as libraries of finite volume method (FVM). Although a little theoretical or numerical studies about heat affected zone (HAZ) formation were performed, there is no research discussing how HAZ is generated considering time development about removal of each material. It is important to understand difference of removal speed of carbon fiber and resin in order to improve quality of cut surface of CFRP. We demonstrated how the carbon fiber and resin are removed by heat of ablation plume by our simulation. We found that carbon fiber is removed faster than resin at first stage because of the difference of thermal conductivity, and after that, the resin is removed faster because of its low combustion temperature. This result suggests the existence of optimal contacting time of the laser ablation and kerf of the target.
3D numerical simulation and analysis of railgun gouging mechanism
Directory of Open Access Journals (Sweden)
Jin-guo Wu
2016-04-01
Full Text Available A gouging phenomenon with a hypervelocity sliding electrical contact in railgun not only shortens the rail lifetime but also affects the interior ballistic performance. In this paper, a 3-D numerical model was introduced to simulate and analyze the generation mechanism and evolution of the rail gouging phenomenon. The results show that a rail surface bulge is an important factor to induce gouging. High density and high pressure material flow on the contact surface, obliquely extruded into the rail when accelerating the armature to a high velocity, can produce gouging. Both controlling the bulge size to a certain range and selecting suitable materials for rail surface coating will suppress the formation of gouging. The numerical simulation had a good agreement with experiments, which validated the computing model and methodology are reliable.
Numerical Relativity Simulations for Black Hole Merger Astrophysics
Baker, John G.
2010-01-01
Massive black hole mergers are perhaps the most energetic astronomical events, establishing their importance as gravitational wave sources for LISA, and also possibly leading to observable influences on their local environments. Advances in numerical relativity over the last five years have fueled the development of a rich physical understanding of general relativity's predictions for these events. Z will overview the understanding of these event emerging from numerical simulation studies. These simulations elucidate the pre-merger dynamics of the black hole binaries, the consequent gravitational waveform signatures ' and the resulting state, including its kick velocity, for the final black hole produced by the merger. Scenarios are now being considered for observing each of these aspects of the merger, involving both gravitational-wave and electromagnetic astronomy.
Numerical simulations on self-leveling behaviors with cylindrical debris bed
Energy Technology Data Exchange (ETDEWEB)
Guo, Liancheng, E-mail: Liancheng.guo@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Morita, Koji, E-mail: morita@nucl.kyushu-u.ac.jp [Faculty of Engineering, Kyushu University, 2-3-7, 744 Motooka, Nishi-ku, Fukuoka 819-0395 (Japan); Tobita, Yoshiharu, E-mail: tobita.yoshiharu@jaea.go.jp [Fast Reactor Safety Technology Development Department, Japan Atomic Energy Agency, 4002 Narita, O-arai, Ibaraki 311-1393 (Japan)
2017-04-15
Highlights: • A 3D coupled method was developed by combining DEM with the multi-fluid model of SIMMER-IV code. • The method was validated by performing numerical simulations on a series of experiments with cylindrical particle bed. • Reasonable agreement can demonstrate the applicability of the method in reproducing the self-leveling behavior. • Sensitivity analysis on some model parameters was performed to assess their impacts. - Abstract: The postulated core disruptive accidents (CDAs) are regarded as particular difficulties in the safety analysis of liquid-metal fast reactors (LMFRs). In the CDAs, core debris may settle on the core-support structure and form conic bed mounds. Then debris bed can be levelled by the heat convection and vaporization of surrounding coolant sodium, which is named “self-leveling behavior”. The self-leveling behavior is a crucial issue in the safety analysis, due to its significant effect on the relocation of molten core and heat-removal capability of the debris bed. Considering its complicate multiphase mechanism, a comprehensive computational tool is needed to reasonably simulate transient particle behavior as well as thermal-hydraulic phenomenon of surrounding fluid phases. The SIMMER program is a successful computer code initially developed as an advanced tool for CDA analysis of LMFRs. It is a multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Until now, the code has been successfully applied in numerical simulations for reproducing key thermal-hydraulic phenomena involved in CDAs as well as performing reactor safety assessment. However, strong interactions between massive solid particles as well as particle characteristics in multiphase flows were not taken into consideration in its fluid-dynamics models. To solve this problem, a new method is developed by combining the discrete element method (DEM
Experimentation and numerical simulation of steel fibre reinforced concrete pipes
International Nuclear Information System (INIS)
Fuente, A. de la; Domingues de Figueiredo, A.; Aguado, A.; Molins, C.; Chama Neto, P. J.
2011-01-01
The results concerning on an experimental and a numerical study related to SFRCP are presented. Eighteen pipes with an internal diameter of 600 mm and fibre dosages of 10, 20 and 40 kg/m3 were manufactured and tested. Some technological aspects were concluded. Likewise, a numerical parameterized model was implemented. With this model, the simulation of the resistant behaviour of SFRCP can be performed. In this sense, the results experimentally obtained were contrasted with those suggested by means MAP reaching very satisfactory correlations. Taking it into account, it could be said that the numerical model is a useful tool for the optimal design of the SFRCP fibre dosages, avoiding the need of the systematic employment of the test as an indirect design method. Consequently, the use of this model would reduce the overall cost of the pipes and would give fibres a boost as a solution for this structural typology. (Author) 27 refs.
Numerical Simulation of Polynomial-Speed Convergence Phenomenon
Li, Yao; Xu, Hui
2017-11-01
We provide a hybrid method that captures the polynomial speed of convergence and polynomial speed of mixing for Markov processes. The hybrid method that we introduce is based on the coupling technique and renewal theory. We propose to replace some estimates in classical results about the ergodicity of Markov processes by numerical simulations when the corresponding analytical proof is difficult. After that, all remaining conclusions can be derived from rigorous analysis. Then we apply our results to seek numerical justification for the ergodicity of two 1D microscopic heat conduction models. The mixing rate of these two models are expected to be polynomial but very difficult to prove. In both examples, our numerical results match the expected polynomial mixing rate well.
Numerical simulation of droplet evaporation between two circular plates
International Nuclear Information System (INIS)
Bam, Hang Jin; Son, Gi Hun
2015-01-01
Numerical simulation is performed for droplet evaporation between two circular plates. The flow and thermal characteristics of the droplet evaporation are numerically investigated by solving the conservation equations of mass, momentum, energy and mass fraction in the liquid and gas phases. The liquid-gas interface is tracked by a sharp-interface level-set method which is modified to include the effects of evaporation at the liquid-gas interface and contact angle hysteresis at the liquid-gas-solid contact line. An analytical model to predict the droplet evaporation is also developed by simplifying the mass and vapor fraction equations in the gas phase. The numerical results demonstrate that the 1-D analytical prediction is not applicable to the high rate evaporation process. The effects of plate gap and receding contact angle on the droplet evaporation are also quantified.
Determination of adsorption parameters in numerical simulation for polymer flooding
Bao, Pengyu; Li, Aifen; Luo, Shuai; Dang, Xu
2018-02-01
A study on the determination of adsorption parameters for polymer flooding simulation was carried out. The study mainly includes polymer static adsorption and dynamic adsorption. The law of adsorption amount changing with polymer concentration and core permeability was presented, and the one-dimensional numerical model of CMG was established under the support of a large number of experimental data. The adsorption laws of adsorption experiments were applied to the one-dimensional numerical model to compare the influence of two adsorption laws on the historical matching results. The results show that the static adsorption and dynamic adsorption abide by different rules, and differ greatly in adsorption. If the static adsorption results were directly applied to the numerical model, the difficulty of the historical matching will increase. Therefore, dynamic adsorption tests in the porous medium are necessary before the process of parameter adjustment in order to achieve the ideal history matching result.
MHD turbulent dynamo in astrophysics: Theory and numerical simulation
Chou, Hongsong
2001-10-01
This thesis treats the physics of dynamo effects through theoretical modeling of magnetohydrodynamic (MHD) systems and direct numerical simulations of MHD turbulence. After a brief introduction to astrophysical dynamo research in Chapter 1, the following issues in developing dynamic models of dynamo theory are addressed: In Chapter 2, nonlinearity that arises from the back reaction of magnetic field on velocity field is considered in a new model for the dynamo α-effect. The dependence of α-coefficient on magnetic Reynolds number, kinetic Reynolds number, magnetic Prandtl number and statistical properties of MHD turbulence is studied. In Chapter 3, the time-dependence of magnetic helicity dynamics and its influence on dynamo effects are studied with a theoretical model and 3D direct numerical simulations. The applicability of and the connection between different dynamo models are also discussed. In Chapter 4, processes of magnetic field amplification by turbulence are numerically simulated with a 3D Fourier spectral method. The initial seed magnetic field can be a large-scale field, a small-scale magnetic impulse, and a combination of these two. Other issues, such as dynamo processes due to helical Alfvénic waves and the implication and validity of the Zeldovich relation, are also addressed in Appendix B and Chapters 4 & 5, respectively. Main conclusions and future work are presented in Chapter 5. Applications of these studies are intended for astrophysical magnetic field generation through turbulent dynamo processes, especially when nonlinearity plays central role. In studying the physics of MHD turbulent dynamo processes, the following tools are developed: (1)A double Fourier transform in both space and time for the linearized MHD equations (Chapter 2 and Appendices A & B). (2)A Fourier spectral numerical method for direct simulation of 3D incompressible MHD equations (Appendix C).
Numerical Simulation of Nanofluid Suspensions in a Geothermal Heat Exchanger
Xiao-Hui Sun; Hongbin Yan; Mehrdad Massoudi; Zhi-Hua Chen; Wei-Tao Wu
2018-01-01
It has been shown that using nanofluids as heat carrier fluids enhances the conductive and convective heat transfer of geothermal heat exchangers. In this paper, we study the stability of nanofluids in a geothermal exchanger by numerically simulating nanoparticle sedimentation during a shut-down process. The nanofluid suspension is modeled as a non-linear complex fluid; the nanoparticle migration is modeled by a particle flux model, which includes the effects of Brownian motion, gravity, turb...
Numerical simulations of the decay of primordial magnetic turbulence
International Nuclear Information System (INIS)
Kahniashvili, Tina; Brandenburg, Axel; Tevzadze, Alexander G.; Ratra, Bharat
2010-01-01
We perform direct numerical simulations of forced and freely decaying 3D magnetohydrodynamic turbulence in order to model magnetic field evolution during cosmological phase transitions in the early Universe. Our approach assumes the existence of a magnetic field generated either by a process during inflation or shortly thereafter, or by bubble collisions during a phase transition. We show that the final configuration of the magnetic field depends on the initial conditions, while the velocity field is nearly independent of initial conditions.
Numerical simulation methods for wave propagation through optical waveguides
International Nuclear Information System (INIS)
Sharma, A.
1993-01-01
The simulation of the field propagation through waveguides requires numerical solutions of the Helmholtz equation. For this purpose a method based on the principle of orthogonal collocation was recently developed. The method is also applicable to nonlinear pulse propagation through optical fibers. Some of the salient features of this method and its application to both linear and nonlinear wave propagation through optical waveguides are discussed in this report. 51 refs, 8 figs, 2 tabs
Numerical Simulation on Zonal Disintegration in Deep Surrounding Rock Mass
Xuguang Chen; Yuan Wang; Yu Mei; Xin Zhang
2014-01-01
Zonal disintegration have been discovered in many underground tunnels with the increasing of embedded depth. The formation mechanism of such phenomenon is difficult to explain under the framework of traditional rock mechanics, and the fractured shape and forming conditions are unclear. The numerical simulation was carried out to research the generating condition and forming process of zonal disintegration. Via comparing the results with the geomechanical model test, the zonal disintegration p...
Numerical simulation of vertical infiltration for leaching fluid in situ
International Nuclear Information System (INIS)
Li Jinxuan; Shi Weijun; Zhang Weimin
1998-01-01
Based on the analysis of movement law of leaching fluid in breaking and leaching experiment in situ, the movement of leaching fluid can be divided into two main stages in the leaching process in situ: Vertical Infiltration in unsaturation zone and horizontal runoff in saturation zone. The corresponding mathematics models are sep up, and the process of vertical infiltration of leaching fluid is numerically simulated
EXTENDED SCALING LAWS IN NUMERICAL SIMULATIONS OF MAGNETOHYDRODYNAMIC TURBULENCE
International Nuclear Information System (INIS)
Mason, Joanne; Cattaneo, Fausto; Perez, Jean Carlos; Boldyrev, Stanislav
2011-01-01
Magnetized turbulence is ubiquitous in astrophysical systems, where it notoriously spans a broad range of spatial scales. Phenomenological theories of MHD turbulence describe the self-similar dynamics of turbulent fluctuations in the inertial range of scales. Numerical simulations serve to guide and test these theories. However, the computational power that is currently available restricts the simulations to Reynolds numbers that are significantly smaller than those in astrophysical settings. In order to increase computational efficiency and, therefore, probe a larger range of scales, one often takes into account the fundamental anisotropy of field-guided MHD turbulence, with gradients being much slower in the field-parallel direction. The simulations are then optimized by employing the reduced MHD equations and relaxing the field-parallel numerical resolution. In this work we explore a different possibility. We propose that there exist certain quantities that are remarkably stable with respect to the Reynolds number. As an illustration, we study the alignment angle between the magnetic and velocity fluctuations in MHD turbulence, measured as the ratio of two specially constructed structure functions. We find that the scaling of this ratio can be extended surprisingly well into the regime of relatively low Reynolds number. However, the extended scaling easily becomes spoiled when the dissipation range in the simulations is underresolved. Thus, taking the numerical optimization methods too far can lead to spurious numerical effects and erroneous representation of the physics of MHD turbulence, which in turn can affect our ability to identify correctly the physical mechanisms that are operating in astrophysical systems.
Numerical Simulations of Settlement of Jet Grouting Columns
Directory of Open Access Journals (Sweden)
Juzwa Anna
2016-03-01
Full Text Available The paper presents the comparison of results of numerical analyses of interaction between group of jet grouting columns and subsoil. The analyses were conducted for single column and groups of three, seven and nine columns. The simulations are based on experimental research in real scale which were carried out by authors. The final goal for the research is an estimation of an influence of interaction between columns working in a group.
Transient productivity index for numerical well test simulations
Energy Technology Data Exchange (ETDEWEB)
Blanc, G.; Ding, D.Y.; Ene, A. [Institut Francais du Petrole, Pau (France)] [and others
1997-08-01
The most difficult aspect of numerical simulation of well tests is the treatment of the Bottom Hole Flowing (BHF) Pressure. In full field simulations, this pressure is derived from the Well-block Pressure (WBP) using a numerical productivity index which accounts for the grid size and permeability, and for the well completion. This productivity index is calculated assuming a pseudo-steady state flow regime in the vicinity of the well and is therefore constant during the well production period. Such a pseudo-steady state assumption is no longer valid for the early time of a well test simulation as long as the pressure perturbation has not reached several grid-blocks around the well. This paper offers two different solutions to this problem: (1) The first one is based on the derivation of a Numerical Transient Productivity Index (NTPI) to be applied to Cartesian grids; (2) The second one is based on the use of a Corrected Transmissibility and Accumulation Term (CTAT) in the flow equation. The representation of the pressure behavior given by both solutions is far more accurate than the conventional one as shown by several validation examples which are presented in the following pages.
Automated numerical simulation of cracked plates, pipes and elbows
International Nuclear Information System (INIS)
Reddy, Babu; Sreehari Kumar, B.; Bhate, S.R.; Kushwaha, H.S.
2008-01-01
In the nuclear industry, piping components are one of the key elements participating in its operation. Integrity of structural tubes and pipes plays a major role in nuclear power plants. The ideal procedure to ensure this aspect would be to conduct experimental studies on pilot/test specimens. However, it may not always be feasible to carry out the experimental investigation, as it requires pre-requisite infrastructure which may not be economically viable. This makes it imperative to conduct numerical simulations of the same particularly in the study of presence of cracks in the critical components. While performing the effect of cracks, the quality of the finite element mesh nearer to the crack tip plays a critical role while estimating J-integral value. The designer is often familiar with design methodology only and he obviously requires a convenient and reliable numerical tool to model and perform the analysis. In this context, an effort has been made in NISA, the general purpose finite element software, to automate the generation of FE meshes for a set of pre-defined components with different crack configurations. To simplify the procedure of FE mesh generation, analysis, and post processing, a graphical user interface (GUI) has been developed accordingly. This paper discusses the automated numerical simulation of plates and pipes with different crack configurations. This simulation software is also designed to help parametric study of cracked pipes. (author)
Numerical solutions of the aerosol general dynamic equation for nuclear reactor safety studies
International Nuclear Information System (INIS)
Park, J.W.
1988-01-01
Methods and approximations inherent in modeling of aerosol dynamics and evolution for nuclear reactor source term estimation have been investigated. Several aerosol evolution problems are considered to assess numerical methods of solving the aerosol dynamic equation. A new condensational growth model is constructed by generalizing Mason's formula to arbitrary particle sizes, and arbitrary accommodation of the condensing vapor and background gas at particle surface. Analytical solution is developed for the aerosol growth equation employing the new condensation model. The space-dependent aerosol dynamic equation is solved to assess implications of spatial homogenization of aerosol distributions. The results of our findings are as follows. The sectional method solving the aerosol dynamic equation is quite efficient in modeling of coagulation problems, but should be improved for simulation of strong condensation problems. The J-space transform method is accurate in modeling of condensation problems, but is very slow. For the situation considered, the new condensation model predicts slower aerosol growth than the corresponding isothermal model as well as Mason's model, the effect of partial accommodation is considerable on the particle evolution, and the effect of the energy accommodation coefficient is more pronounced than that of the mass accommodation coefficient. For the initial conditions considered, the space-dependent aerosol dynamics leads to results that are substantially different from those based on the spatially homogeneous aerosol dynamic equation
VALIDATION OF NUMERICAL METHODS TO CALCULATE BYPASS FLOW IN A PRISMATIC GAS-COOLED REACTOR CORE
Directory of Open Access Journals (Sweden)
NAM-IL TAK
2013-11-01
Full Text Available For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR, intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI and the AGREE code of the University of Michigan (U of M. One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.
Numerical simulation of residual stress in piping components at Framatome-ANP
International Nuclear Information System (INIS)
Gilies, P.; Franco, C.; Cipiere, M.-F.; Ould, P.
2005-01-01
Numerous manufacturing processes induce residual stresses and distortions in piping components and associated welds: quenching of cast pipings, machining and welding. In Pressurized Water Reactors, most of the components have a large thickness for sustaining pressure and distortions are a minor source of concern. This is not the case for residual stresses which may have a strong influence on several type of damage such as fatigue, corrosion, brittle fracture. In low toughness components, residual stress fields may contribute to ductile tearing initiation. These potential damages are mitigated after welding by stress relief heat treatment, which is applied in a systematic manner to ferritic components of the primary system in nuclear reactors. This treatment is not applied on austenitic piping for which the heat treatment temperature is limited due to the risk of sensitization and residual stresses are difficult to eliminate completely. Since on site measurements are costly and difficult to perform, numerical simulation appears to be an attractive tool for estimating residual stress distributions. Framatome-ANP is working on modelling manufacturing processes with that purpose in mind. This paper presents three kinds of applications illustrating efforts on welding, quenching and machining simulation. First a comparison is shown between computations and measurements of residual stress induced by welding of a dissimilar weld metal junction. Then numerical simulations of quenching of a cast stainless steel nozzle are presented. Finally quenching followed by machining and grinding of this cast component are considered in a full simulation of the manufacturing process. Computed distortions and residual stresses are compared with experimental measurements at different stages of the manufacturing process. (authors)
Efficient numerical simulation of heat storage in subsurface georeservoirs
Boockmeyer, A.; Bauer, S.
2015-12-01
The transition of the German energy market towards renewable energy sources, e.g. wind or solar power, requires energy storage technologies to compensate for their fluctuating production. Large amounts of energy could be stored in georeservoirs such as porous formations in the subsurface. One possibility here is to store heat with high temperatures of up to 90°C through borehole heat exchangers (BHEs) since more than 80 % of the total energy consumption in German households are used for heating and hot water supply. Within the ANGUS+ project potential environmental impacts of such heat storages are assessed and quantified. Numerical simulations are performed to predict storage capacities, storage cycle times, and induced effects. For simulation of these highly dynamic storage sites, detailed high-resolution models are required. We set up a model that accounts for all components of the BHE and verified it using experimental data. The model ensures accurate simulation results but also leads to large numerical meshes and thus high simulation times. In this work, we therefore present a numerical model for each type of BHE (single U, double U and coaxial) that reduces the number of elements and the simulation time significantly for use in larger scale simulations. The numerical model includes all BHE components and represents the temporal and spatial temperature distribution with an accuracy of less than 2% deviation from the fully discretized model. By changing the BHE geometry and using equivalent parameters, the simulation time is reduced by a factor of ~10 for single U-tube BHEs, ~20 for double U-tube BHEs and ~150 for coaxial BHEs. Results of a sensitivity study that quantify the effects of different design and storage formation parameters on temperature distribution and storage efficiency for heat storage using multiple BHEs are then shown. It is found that storage efficiency strongly depends on the number of BHEs composing the storage site, their distance and
Numerical Simulation of Liquid Sloshing Problem under Resonant Excitation
Directory of Open Access Journals (Sweden)
Fu-kun Gui
2014-04-01
Full Text Available Numerical simulations were conducted to investigate the fluid resonance in partially filled rectangular tank based on the OpenFOAM package of viscous fluid model. The numerical model was validated by the available theoretical, numerical, and experimental data. The study was mainly focused on the large amplitude sloshing motion and the corresponding impact force around the resonant condition. It was found that, for the 2D situation, the double pressure peaks happened near to the side walls around the still water level. And they were corresponding to the local free surface rising up and set-down, respectively. The impulsive loads on the tank corner with extreme magnitudes were observed as the free surface impacted the ceiling. The 3D numerical results showed that the free surface amplitudes along the side walls varied diversely, depending on the direction and frequency of the external excitation. The characteristics of the pressure around the still water level and tank ceiling were also presented. According to the computational results, it was found that the 2D numerical model can predict the impact loads near the still water level as accurately as 3D model. However, the impulsive pressure near the tank ceiling corner was remarkably underestimated.
Engineering simulator applications to emergency preparedness at DOE reactor sites
International Nuclear Information System (INIS)
Beelman, R.J.
1990-01-01
This paper reports that since 1984 the Idaho National Engineering Laboratory (INEL) has conducted twenty-seven comprehensive emergency preparedness exercises at the U.S. Nuclear Regulatory Commission's (NRC) Headquarters Operations Center and Regional Incident Response Centers using the NRC's Nuclear Plant Analyzer (NPA), developed at the INEL, as an engineering simulator. The objective of these exercises has been to assist the NRC in upgrading its preparedness to provide technical support backup and oversight to U.S. commercial nuclear plant licensees during emergencies. With the current focus on Department of Energy (DOE) reactor operational safety and emergency preparedness, this capability is envisioned as a means of upgrading emergency preparedness at DOE production and test reactor sites such as the K-Reactor at Savannah River Laboratory (SRL) and the Advanced Test Reactor (ATR) at INEL
Simulator platform for fast reactor operation and safety technology demonstration
International Nuclear Information System (INIS)
Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.
2012-01-01
A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.
Simulator platform for fast reactor operation and safety technology demonstration
Energy Technology Data Exchange (ETDEWEB)
Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)
2012-07-30
A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.
Simulation of a porous ceramic membrane reactor for hydrogen production
Energy Technology Data Exchange (ETDEWEB)
Yu, W.; Ohmori, T.; Yamamoto, T.; Endo, A.; Nakaiwa, M.; Hayakawa, T. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Itoh, N. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Utsunomiya Univ. (Japan). Dept. of Applied Chemistry
2005-08-01
A systematic simulation study was performed to investigate the performance of a porous ceramic membrane reactor for hydrogen production by means of methane steam reforming. The results show that the methane conversions much higher than the corresponding equilibrium values can be achieved in the membrane reactor due to the selective removal of products from the reaction zone. The comparison of isothermal and non-isothermal model predictions was made. It was found that the isothermal assumption overestimates the reactor performance and the deviation of calculation results between the two models is subject to the operating conditions. The effects of various process parameters such as the reaction temperature, the reaction side pressure, the feed flow rate and the steam to methane molar feed ratio as well as the sweep gas flow rate and the operation modes, on the behavior of membrane reactor were analyzed and discussed. (author)
Primary loop simulation of the SP-100 space nuclear reactor
International Nuclear Information System (INIS)
Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.
2011-01-01
Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)
Directory of Open Access Journals (Sweden)
Zaidon M. Shakoor
2013-05-01
Full Text Available In this research, two models are developed to simulate the steady state fixed bed reactor used for styrene production by ethylbenzene dehydrogenation. The first is one-dimensional model, considered axial gradient only while the second is two-dimensional model considered axial and radial gradients for same variables.The developed mathematical models consisted of nonlinear simultaneous equations in multiple dependent variables. A complete description of the reactor bed involves partial, ordinary differential and algebraic equations (PDEs, ODEs and AEs describing the temperatures, concentrations and pressure drop across the reactor was given. The model equations are solved by finite differences method. The reactor models were coded with Mat lab 6.5 program and various numerical techniques were used to obtain the desired solution.The simulation data for both models were validated with industrial reactor results with a very good concordance.
Numerical simulation of flow-induced vibrations in tube bundles
International Nuclear Information System (INIS)
Elisabeth Longatte; Zaky Bendjeddou; Mhamed Souli
2005-01-01
Full text of publication follows: In many industrial components mechanical structures like rod cluster control assembly, fuel assembly and heat exchanger tube bundles are submitted to complex flows causing possible vibrations and damage. Fluid forces are usually split into two parts: structure motion independent forces and fluid-elastic forces coupled with tube motion and responsible for possible dynamic instability development leading to possible short term failures through high amplitude vibrations. Most classical fluid force identification methods rely on structure response experimental measurements associated with convenient data processes. Owing to recent improvements in Computational Fluid Dynamics (C.F.D.), numerical fluid force identification is now practicable in the presence of industrial configurations. The present paper is devoted to numerical simulation of flow-induced vibrations of tube bundles submitted to single-phase cross flows by using C.F.D. codes. Direct Numerical Simulation (D.N.S.), Arbitrary Lagrange Euler formulation (A.L.E.) and code coupling process are involved to predict fluid forces responsible for tube bundle vibrations in the presence of fluid structure and fluid-elastic coupling effects. In the presence of strong multi-physics coupling, simulation of flow-induced vibrations requires a fluid structure code coupling process. The methodology consists in solving in the same time thermohydraulics and mechanics problems by using an A.L.E. formulation for the fluid computation. The purpose is to take into account coupling between flow and structure motions in order to be able to capture coupling effects. From a numerical point of view, there are three steps in the computation: the fluid problem is solved on the computational domain; fluid forces acting on the moving tube are estimated; finally they are introduced in the structure solver providing the tube displacement that is used to actualize the fluid computational domain. Specific
Directory of Open Access Journals (Sweden)
Zhenkun Sang
2018-04-01
Full Text Available Ultra-low calorific value gas (ULCVG not only poses a problem for environmental pollution, but also createsa waste of energy resources if not utilized. A novel reactor, a rotary regenerator-type catalytic combustion reactor (RRCCR, which integrates the functions of a regenerator and combustor into one component, is proposed for the elimination and utilization of ULCVG. Compared to reversal-flow reactor, the operation of the RRCCR is achieved by incremental rotation rather than by valve control, and it has many outstanding characteristics, such as a compact structure, flexible application, and limited energy for circulation. Due to the effects of the variation of the gas flow and concentration on the performance of the reactor, different inlet velocities and concentrations are analyzed by numerical investigations. The results reveal that the two factors have a major impact on the performance of the reactor. The performance of the reactor is more sensitive to the increase of velocity and the decrease of methane concentration. When the inlet concentration (2%vol. is reduced by 50%, to maintain the methane conversion over 90%, the inlet velocity can be reduced by more than three times. Finally, the highly-efficient and stable operating envelope of the reactor is drawn.
NUMERICAL MULTIGROUP TRANSIENT ANALYSIS OF SLAB NUCLEAR REACTOR WITH THERMAL FEEDBACK
Directory of Open Access Journals (Sweden)
Filip Osuský
2016-12-01
Full Text Available The paper describes a new numerical code for multigroup transient analyses with thermal feedback. The code is developed at Institute of Nuclear and Physical Engineering. It is necessary to carefully investigate transient states of fast neutron reactors, due to recriticality issues after accident scenarios. The code solves numerical diffusion equation for 1D problem with possible neutron source incorporation. Crank-Nicholson numerical method is used for the transient states. The investigated cases are describing behavior of PWR fuel assembly inside of spent fuel pool and with the incorporated neutron source for better illustration of thermal feedback.
Direct Numerical Simulations of Rayleigh-Taylor instability
International Nuclear Information System (INIS)
Livescu, D; Wei, T; Petersen, M R
2011-01-01
The development of the Rayleigh-Taylor mixing layer is studied using data from an extensive new set of Direct Numerical Simulations (DNS), performed on the 0.5 Petaflops, 150k compute cores BG/L Dawn supercomputer at Lawrence Livermore National Laboratory. This includes a suite of simulations with grid size of 1024 2 × 4608 and Atwood number ranging from 0.04 to 0.9, in order to examine small departures from the Boussinesq approximation as well as large Atwood number effects, and a high resolution simulation of grid size 4096 2 × 4032 and Atwood number of 0.75. After the layer width had developed substantially, additional branched simulations have been run under reversed and zero gravity conditions. While the bulk of the results will be published elsewhere, here we present preliminary results on: 1) the long-standing open question regarding the discrepancy between the numerically and experimentally measured mixing layer growth rates and 2) mixing characteristics.
Numerical simulation of plasma vertical position stabilization in ITER
International Nuclear Information System (INIS)
Astapkovich, A.M.; Sadakov, S.N.
1992-01-01
The paper deals with numerical simulation of plasma vertical position stabilization in ITER. The calculations are performed using EDDY C-2 code by the method of direct numerical simulation of transient electromagnetic processes taking into account the evolution of plasma position, cross-section shape and full plasma current. When simulating free vertical plasma drift in ITER with twin passive stabilization loops, it was shown that account of the effects of cross-section deformation and plasma current alternations results in almost two fold degradation of passive stabilization parameters as compared to the calculations for 'rigid displacement' model. In terms of methodology, the account of the effects of cross section deformation and plasma current alternations requires clarification of the definitions for reverse increment of vertical instability and for stability margin coefficient. The simulation of plasma pinch return to equilibrium position after the closure of control coils allows to assess the required parameters of active control system and demonstrate the effect of screen current reverse in twin loops. The obtained results were used to develop the ITER conceptual design and affected the choice of the concept of twin passive loops and new positron of control coils as the basis approaches. 11 refs.; 12 figs.; 1 tab
Numerical simulations of rubber bearing tests and shaking table tests
International Nuclear Information System (INIS)
Hirata, K.; Matsuda, A.; Yabana, S.
2002-01-01
Test data concerning rubber bearing tests and shaking table tests of base-isolated model conducted by CRIEPI are provided to the participants of Coordinated Research Program (CRP) on 'Intercomparison of Analysis Methods for predicting the behaviour of Seismically Isolated Nuclear Structure', which is organized by International Atomic Energy Agency (IAEA), for the comparison study of numerical simulation of base-isolated structure. In this paper outlines of the test data provided and the numerical simulations of bearing tests and shaking table tests are described. Using computer code ABAQUS, numerical simulations of rubber bearing tests are conducted for NRBs, LRBs (data provided by CRIEPI) and for HDRs (data provided by ENEA/ENEL and KAERI). Several strain energy functions are specified according to the rubber material test corresponding to each rubber bearing. As for lead plug material in LRB, mechanical characteristics are reevaluated and are made use of. Simulation results for these rubber bearings show satisfactory agreement with the test results. Shaking table test conducted by CRIEPI is of a base isolated rigid mass supported by LRB. Acceleration time histories, displacement time histories of the isolators as well as cyclic loading test data of the LRB used for the shaking table test are provided to the participants of the CRP. Simulations of shaking table tests are conducted for this rigid mass, and also for the steel frame model which is conducted by ENEL/ENEA. In the simulation of the rigid mass model test, where LRBs are used, isolators are modeled either by bilinear model or polylinear model. In both cases of modeling of isolators, simulation results show good agreement with the test results. In the case of the steel frame model, where HDRs are used as isolators, bilinear model and polylinear model are also used for modeling isolators. The response of the model is simulated comparatively well in the low frequency range of the floor response, however, in