Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

Science.gov (United States)

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.

Numerical simulations of helium flow through prismatic fuel elements of very high temperature reactors

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)

Numerical Simulation of the Pressure Distribution in the Reactor Vessel Downcomer Region Fluctuated by the Reactor Coolant Pump

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.

Numerical simulation and geometry optimization of hot-gas mixing in lower plenum of high temperature gas-cooled reactor

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)

Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

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.

General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics; Assemblee generale. Reunion technique: la simulation numerique et experimentale appliquee a la physique des reacteurs

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.)

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)

Nuclear reactor simulator

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)

Modelling and numerical simulation of the corrosion product transport in the pressurised water reactor primary circuit

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 analysis and scale experiment design of the hot water layer system of the Brazilian Multipurpose Reactor (RMB reactor)

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 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.

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.)

Numerical Simulation of Measurements during the Reactor Physical Startup at Unit 3 of Rostov NPP

Science.gov (United States)

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.

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

The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

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)

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

Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

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.

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)

Numerical Simulation of Particle Flow Motion in a Two-Dimensional Modular Pebble-Bed Reactor with Discrete Element Method

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.

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.)

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.

Evaluation method for core thermohydraulics during natural circulation in fast reactors numerical predictions of inter-wrapper flow

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)

Application of multicomponent medium model for numerical simulation of reactor element melting and melt relocation under severe accidents

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)

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)

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.)

Single-phase multi-dimensional thermohydraulics direct numerical simulation code DINUS-3. Input data description

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)

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)

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

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)

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.

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.

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 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.

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.)

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)

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

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)

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

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...

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)

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

Multi-physic simulations of irradiation experiments in a technological irradiation reactor

International Nuclear Information System (INIS)

Bonaccorsi, Th.

2007-09-01

A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

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.

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

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)

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

Numerical simulation of severe water ingress accidents in a modular high temperature gas cooled reactor

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

Numerical simulation of two phase flows in heat exchangers; Simulation numerique des ecoulements diphasiques dans les echangeurs

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)

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)

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

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

Development and successful operation of the enhanced-interlink system of experiment data and numerical simulation in LHD

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)

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.

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 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

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

Numerical simulation of ion transport membrane reactors: Oxygen permeation and transport and fuel conversion

KAUST Repository

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.

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

Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

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.

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

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)

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

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)

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)

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

The Simulator Development for RDE Reactor

Science.gov (United States)

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 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)

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

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.)

A numerical design and feasibility study of self-wastage experiment using simulant material in a sodium fast reactor

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

A Numerical Design and Feasibility Study of Self-Wastage Experiment Using Simulant Material in a Sodium Fast Reactor

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.

A numerical design and feasibility study of self-wastage experiment using simulant material in a sodium fast reactor

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.

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

A Multi-Physics simulation of the Reactor Core using CUPID/MASTER

International Nuclear Information System (INIS)

Lee, Jae Ryong; Cho, Hyoung Kyu; Yoon, Han Young; Cho, Jin Young; Jeong, Jae Jun

2011-01-01

KAERI has been developing a component-scale thermal hydraulics code, CUPID. The aim of the code is for multi-dimensional, multi-physics and multi-scale thermal hydraulics analysis. In our previous papers, the CUPID code has proved to be able to reproduce multidimensional thermal hydraulic analysis by validated with various conceptual problems and experimental data. For the numerical closure, it adopts a three dimensional, transient, two-phase and three-field model, and includes physical models and correlations of the interfacial mass, momentum, and energy transfer. For the multi-scale analysis, the CUPID is on progress to merge into system-scale thermal hydraulic code, MARS. In the present paper, a multi-physics simulation was performed by coupling the CUPID with three dimensional neutron kinetics code, MASTER. The MASTER is merged into the CUPID as a dynamic link library (DLL). The APR1400 reactor core during control rod drop/ejection accident was simulated as an example by adopting a porous media approach to employ fuel assembly. The following sections present the numerical modeling for the reactor core, coupling of the kinetics code, and the simulation results

Large-scale numerical simulations on two-phase flow behavior in a fuel bundle of RMWR with the earth simulator

International Nuclear Information System (INIS)

Kazuyuki, Takase; Hiroyuki, Yoshida; Hidesada, Tamai; Hajime, Akimoto; Yasuo, Ose

2003-01-01

Fluid flow characteristics in a fuel bundle of a reduced-moderation light water reactor (RMWR) with a tight-lattice core were analyzed numerically using a newly developed two-phase flow analysis code under the full bundle size condition. Conventional analysis methods such as sub-channel codes need composition equations based on the experimental data. In case that there are no experimental data regarding to the thermal-hydraulics in the tight-lattice core, therefore, it is difficult to obtain high prediction accuracy on the thermal design of the RMWR. Then the direct numerical simulations with the earth simulator were chosen. The axial velocity distribution in a fuel bundle changed sharply around a grid spacer and its quantitative evaluation was obtained from the present preliminary numerical study. The high prospect was acquired on the possibility of establishment of the thermal design procedure of the RMWR by large-scale direct simulations. (authors)

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)

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)

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)

RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

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

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.

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.

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)

Improved numerical algorithm and experimental validation of a system thermal-hydraulic/CFD coupling method for multi-scale transient simulations of pool-type reactors

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.

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

A Numerical Model for Trickle Bed Reactors

Science.gov (United States)

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.

On the theories, techniques, and computer codes used in numerical reactor criticality and burnup calculations

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

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)

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 Simulation to Phenomenon of Main Vessel Free Surface Flow Impact Coping for Fast Reactor by Moving Particle Semi-implicit Method

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)

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)

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)

Study on applicability of numerical simulation to evaluation of gas entrainment due to free surface vortex

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)

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

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

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)

RELAP/SCDAPSIM Reactor System Simulator Development and Training for University and Reactor Applications

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)

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.

Numerical simulation on single bubble rising behavior in liquid metal using moving particle semi-implicit method

International Nuclear Information System (INIS)

Zuo Juanli; Tian Wenxi; Qiu Suizheng; Chen Ronghua; Su Guanghui

2011-01-01

The gas-lift pump in liquid metal cooling fast reactor (LMFR) is an innovational conceptual design to enhance the natural circulation ability of reactor core. The two-phase flow character of gas-liquid metal makes significant improvement of the natural circulation capacity and reactor safety. In present basic study, the rising behavior of a single nitrogen bubble in five kinds of liquid metals (lead bismuth alloy, liquid kalium, sodium, potassium sodium alloy and lithium lead alloy) was numerically simulated using moving particle semi-implicit (MPS) method. The whole growing process of single nitrogen bubble in liquid metal was captured. The bubble shape and rising speed of single nitrogen bubble in each liquid metal were compared. The comparison between simulation results using MPS method and Grace graphical correlation shows a good agreement. (authors)

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)

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

Virtual environments simulation in research reactor

Science.gov (United States)

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.

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

A Transient Numerical Simulation of Perched Ground-Water Flow at the Test Reactor Area, Idaho National Engineering and Environmental Laboratory, Idaho, 1952-94

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)

Three dimensional numerical simulation of a full scale CANDU reactor moderator to study temperature fluctuations

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

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)

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.

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and MCT Facility

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

Comparison of CFD Simulations of Moderator Circulation Phenomena for a CANDU-6 Reactor and 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.

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

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

Dynamic simulation platform to verify the performance of the reactor regulating system for a research reactor

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)

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%

Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor

International Nuclear Information System (INIS)

Perez M, C.; Benitez R, J.S.; Lopez C, R.

2003-01-01

The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

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 simulation on self-regulating characteristics of a cold neutron source with a closed-thermosiphon

International Nuclear Information System (INIS)

Kawai, Takeshi; Utsuro, Masahiko; Okamoto, Sunao

1989-01-01

A cold neutron source (CNS) having a closed-thermosiphon cooling loop shows a characteristic of self-regulation to the heat load fluctuations if the moderator transfer tube fulfills certain conditions. A dynamical equation of the closed-thermosiphon type CNS having such a property has been presented on the basis of the non-equilibrium thermodynamics. Kyoto University Reactor (KUR) CNS is investigated by numerical simulation of this equation. The numerical predictions for the self-regulating characteristics are in agreement with available experimental data. (author)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

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)

Numerical verification/validation of the theory of coupled reactors for deuterium critical assembly, using MCNP5 and Serpent codes

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)

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

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

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.

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)

Modeling and simulation of CANDU reactor and its regulating system

Science.gov (United States)

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

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)

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)

Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

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

Numerical modeling of carrier gas flow in atomic layer deposition vacuum reactor: A comparative study of lattice Boltzmann models

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.

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)

The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

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.

The Numerical Nuclear Reactor for High-Fidelity Integrated Simulation of Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Phenomena

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

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.

Numerical Analysis of The Effect of Hydrodynamics and Operating Conditions on Biodiesel Synthesis in a Rotor-Stator Spinning Disk Reactor

Directory of Open Access Journals (Sweden)

Wen Zhuqing

2017-06-01

Full Text Available A rotor-stator spinning disk reactor for intensified biodiesel synthesis is described and numerically simulated in the present research. 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 injected into the reactor from centres 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 multicomponent transport model with the CFD software ANSYS©Fluent v. 13.0. Effect of operating conditions on TG conversion is particularly investigated. Simulation results indicate that there is occurrence of back flow close to the stator at the outlet zone. Small gap size and fast rotational speed generally help to intensify mixing among reagents, and consequently enhance TG conversion. However, increasing rotational speed of spinning disk leads to more backflow, which decreases TG conversion. Large flow rate of TG at inlet is not recommended as well because of the short mean residence time of reactants inside the reactor.

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.

Development of variable width ribbon heating elements for liquid metal and gas-cooled fast breeder reactor fuel rod simulators

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

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 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

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)

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

Mathematical Modeling and Simulation of the Dehydrogenation of Ethyl Benzene to Form Styrene Using Steady-State Fixed Bed Reactor

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.

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 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 analysis of temperature fluctuation in core outlet region of China experimental fast reactor

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 Personal Computer-Based Simulator for Nuclear-Heating Reactors

International Nuclear Information System (INIS)

Liu Jie; Zhang Zuoyi; Lu Dongsen; Shi Zhengang; Chen Xiaoming; Dong Yujie

2000-01-01

A personal computer (PC)-based simulator for nuclear-heating reactors (NHRs), PC-NHR, has been developed to provide an educational tool for understanding the design and operational characteristics of an NHR system. A general description of the reactor system as well as the technical basis for the design and operation of the heating reactor is provided. The basic models and equations for the NHR simulation are then given, which include models of the reactor core, the reactor coolant system, the containment, and the control system. The graphical user interface is described in detail to provide a manual for the user to operate the simulator properly. Steady state and several transients have been simulated. The results of PC-NHR are in good agreement with design data and the results of RETRAN-02. The real-time capability is also confirmed

Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

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)

A Friendly-Biological Reactor SIMulator (BioReSIM for studying biological processes in wastewater treatment processes

Directory of Open Access Journals (Sweden)

Raul Molina

2014-12-01

Full Text Available Biological processes for wastewater treatments are inherently dynamic systems because of the large variations in the influent wastewater flow rate, concentration composition and the adaptive behavior of the involved microorganisms. Moreover, the sludge retention time (SRT is a critical factor to understand the bioreactor performances when changes in the influent or in the operation conditions take place. Since SRT are usually in the range of 10-30 days, the performance of biological reactors needs a long time to be monitored in a regular laboratory demonstration, limiting the knowledge that can be obtained in the experimental lab practice. In order to overcome this lack, mathematical models and computer simulations are useful tools to describe biochemical processes and predict the overall performance of bioreactors under different working operation conditions and variations of the inlet wastewater composition. The mathematical solution of the model could be difficult as numerous biochemical processes can be considered. Additionally, biological reactors description (mass balance, etc. needs models represented by partial or/and ordinary differential equations associated to algebraic expressions, that require complex computational codes to obtain the numerical solutions. Different kind of software for mathematical modeling can be used, from large degree of freedom simulators capable of free models definition (as AQUASIM, to closed predefined model structure programs (as BIOWIN. The first ones usually require long learning curves, whereas the second ones could be excessively rigid for specific wastewater treatment systems. As alternative, we present Biological Reactor SIMulator (BioReSIM, a MATLAB code for the simulation of sequencing batch reactors (SBR and rotating biological contactors (RBC as biological systems of suspended and attached biomass for wastewater treatment, respectively. This BioReSIM allows the evaluation of simple and complex

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 model simulation of free surface behavior in spallation target of ADS

International Nuclear Information System (INIS)

Chai Xiang; Su Guanyu; Cheng Xu

2012-01-01

The spallation target in accelerator driven sub-critical system (ADS) couples the subcritical reactor core with accelerator. The design of a windowless target has to ensure the formation of a stable free surface with desirable shape, to avoid local over- heating of the heavy liquid metal (HLM). To investigate the free surface behavior of the spallation target, OpenFOAM, an opened CFD software platform, was used to simulate the formation and features of the free surface in the windowless target. VOF method was utilized as the interface-capturing methodology. The numerical results were compared to experimental data and numerical results obtained with FLUENT code. The effects of turbulence models were studied and recommendations were made related to application of turbulence models. (authors)

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.)

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 thermal stratification in cold legs by using openFOAM

International Nuclear Information System (INIS)

Cai, Jiejin; Watanabe, Tadashi

2010-01-01

During a small-break loss-of-coolant accident in pressurized water reactors (PWRs), emergency core cooling system (ECCS) is actuated and cold water is injected into cold legs. Insufficient mixing of injected cold water and hot primary coolant results in thermal stratification, which is a matter of concern for evaluation of pressurized thermal shock (PTS) in view of aging and life extension of nuclear power plants. In this study, an open source CFD software, OpenFOAM, is used to simulate mixing and thermal stratification in the cold leg of ROSA/LSTF, which is the largest thermal-hydraulic integral test facility simulating PWR. One of the cold-leg is numerically simulated from the outlet of primary coolant pump to the inlet of downcomer. ECCS water is injected from injection nozzle connected at the top of the cold leg into the steady-state natural circulation flow under high-pressure and high-temperature conditions. The temperature distribution in the cold leg is compared with experimental and FLUENT's results. Effects of turbulent flow models and secondary flow due to the elbow section of the cold leg are discussed for the case with the single-phase natural circulation. Injection into a two-phase stratified flow is also simulated and predictive and numerical capabilities of OpenFOAM are discussed. (author)

Numerical simulation of fuel assembly thermohydraulics of fast reactors with the partial blockage of cross section under the coolant

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)

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

Compilation of reactor physics data of the year 1984, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-12-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1984 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

Compilation of reactor physics data of the year 1983, AVR reactor

International Nuclear Information System (INIS)

Werner, H.; Bergerfurth, A.; Thomas, F.; Geskes, B.

1985-06-01

The 'AVR reactor physics data' is a documentation published once a year, the data presented being obtained by a simulation of reactor operation using the AVR-80 numerical model. This model is constantly updated and improved in response to new results and developments in the field of reactor theory and thermohydraulics, and in response to theoretical or practical modifications of reactor operation or in the computer system. The large variety of measured data available in the AVR reactor simulation system also makes it an ideal testing system for verification of the computing programs presented in the compilation. A survey of the history of operations in 1983 and a short explanation of the computerized simulation methods are followed by tables and graphs that serve as a source of topical data for readers interested in the physics of high-temperature pebble-bed reactors. (orig./HP) [de

WWER-1000 reactor simulator. Material for training courses and workshops. 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 publication: Training Course Series No.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, 2nd edition, Pressurized Water Reactor Simulator (2005) and Training Course Series No.23, 2nd edition, Boiling Water Reactor Simulator (2005). This report consists of course material for workshops using the WWER-1000 Reactor Department Simulator from the Moscow Engineering and Physics Institute, Russian Federation

Problem-Oriented Simulation Packages and Computational Infrastructure for Numerical Studies of Powerful Gyrotrons

International Nuclear Information System (INIS)

Damyanova, M; Sabchevski, S; Vasileva, E; Balabanova, E; Zhelyazkov, I; Dankov, P; Malinov, P

2016-01-01

Powerful gyrotrons are necessary as sources of strong microwaves for electron cyclotron resonance heating (ECRH) and electron cyclotron current drive (ECCD) of magnetically confined plasmas in various reactors (most notably ITER) for controlled thermonuclear fusion. Adequate physical models and efficient problem-oriented software packages are essential tools for numerical studies, analysis, optimization and computer-aided design (CAD) of such high-performance gyrotrons operating in a CW mode and delivering output power of the order of 1-2 MW. In this report we present the current status of our simulation tools (physical models, numerical codes, pre- and post-processing programs, etc.) as well as the computational infrastructure on which they are being developed, maintained and executed. (paper)

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)

Numerical modelling of heat and mass transfer in adsorption solar reactor of ammonia on active carbon

Science.gov (United States)

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.

Simulation of Safety and Transient Analysis of a Pressurized Water Reactor using the Personal Computer Transient Analyzer

Directory of Open Access Journals (Sweden)

Sunday J. IBRAHIM

2013-06-01

Full Text Available Safety and transient analyses of a pressurised water reactor (PWR using the Personal Computer Transient Analyzer (PCTRAN simulator was carried out. The analyses presented a synergistic integration of a numerical model; a full scope high fidelity simulation system which adopted point reactor neutron kinetics model and movable boundary two phase fluid models to simplify the calculation of the program, so it could achieve real-time simulation on a personal computer. Various scenarios of transients and accidents likely to occur at any nuclear power plant were simulated. The simulations investigated the change of signals and parameters vis a vis loss of coolant accident, scram, turbine trip, inadvertent control rod insertion and withdrawal, containment failure, fuel handling accident in auxiliary building and containment, moderator dilution as well as a combination of these parameters. Furthermore, statistical analyses of the PCTRAN results were carried out. PCTRAN results for the loss of coolant accident (LOCA caused a rapid drop in coolant pressure at the rate of 21.8KN/m2/sec triggering a shutdown of the reactor protection system (RPS, while the turbine trip accident showed a rapid drop in total plant power at the rate of 14.3 MWe/sec causing a downtime in the plant. Fuel handling accidents mimic results showed release of radioactive materials in unacceptable doses. This work shows the potential classes of nuclear accidents likely to occur during operation in proposed reactor sites. The simulations are very appropriate in the light of Nigeria’s plan to generate nuclear energy in the region of 1000 MWe from reactors by 2017.

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...

Numerical simulation of flood barriers

Science.gov (United States)

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 Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR

Directory of Open Access Journals (Sweden)

G. V. Durga Prasad

2008-01-01

Full Text Available Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR are very important in order to study its performance for different designs and various off-design conditions and for design optimization. In the present work, parametric studies of the primary heat transport loop of NCBWR have been performed using lumped parameter models and RELAP5/MOD3.4 code. The lumped parameter models are based on the drift flux model and homogeneous equilibrium mixture (HEM model of two-phase flow. Numerical simulations are performed with both models. Compared to the results obtained from the HEM model, those obtained from the drift flux model are closer to RELAP5. The variations of critical heat flux with various geometric parameters and operating conditions are thoroughly investigated. The material required to construct the primary heat transport (PHT loop of NCBWR has been minimized using sequential quadratic programming. The stability of NCBWR has also been verified at the optimum point.

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

An Open Source-based Approach to the Development of Research Reactor Simulator

International Nuclear Information System (INIS)

Joo, Sung Moon; Suh, Yong Suk; Park, Cheol Park

2016-01-01

In reactor design, operator training, safety analysis, or research using a reactor, it is essential to simulate time dependent reactor behaviors such as neutron population, fluid flow, and heat transfer. Furthermore, in order to use the simulator to train and educate operators, a mockup of the reactor user interface is required. There are commercial software tools available for reactor simulator development. However, it is costly to use those commercial software tools. Especially for research reactors, it is difficult to justify the high cost as regulations on research reactor simulators are not as strict as those for commercial Nuclear Power Plants(NPPs). An open source-based simulator for a research reactor is configured as a distributed control system based on EPICS framework. To demonstrate the use of the simulation framework proposed in this work, we consider a toy example. This example approximates a 1-second impulse reactivity insertion in a reactor, which represents the instantaneous removal and reinsertion of a control rod. The change in reactivity results in a slightly delayed change in power and corresponding increases in temperatures throughout the system. We proposed an approach for developing research reactor simulator using open source software tools, and showed preliminary results. The results demonstrate that the approach presented in this work can provide economical and viable way of developing research reactor simulators

Numerical and experimental study of hydraulic dashpot used in the shut-off rod drive mechanism of a nuclear reactor

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)

Numerical and experimental study of hydraulic dashpot used in the shut-off rod drive mechanism of a nuclear reactor

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)

On mathematical modeling and numerical simulation of chemical kinetics in turbulent lean premixed combustion

Energy Technology Data Exchange (ETDEWEB)

Lilleberg, Bjorn

2011-07-01

This thesis investigates turbulent reacting lean premixed flows with detailed treatment of the chemistry. First, the fundamental equations which govern laminar and turbulent reacting flows are presented. A perfectly stirred reactor numerical code is developed to investigate the role of unmixedness and chemical kinetics in driving combustion instabilities. This includes both global single-step and detailed chemical kinetic mechanisms. The single-step mechanisms predict to some degree a similar behavior as the detailed mechanisms. However, it is shown that simple mechanisms can by themselves introduce instabilities. Magnussens Eddy Dissipation Concept (EDC) for turbulent combustion is implemented in the open source CFD toolbox OpenFOAM R for treatment of both fast and detailed chemistry. RANS turbulence models account for the turbulent compressible flow. A database of pre-calculated chemical time scales, which contains the influence of chemical kinetics, is coupled to EDC with fast chemistry to account for local extinction in both diffusion and premixed flames. Results are compared to fast and detailed chemistry calculations. The inclusion of the database shows significantly better results than the fast chemistry calculations while having a comparably small computational cost. Numerical simulations of four piloted lean premixed jet flames falling into the 'well stirred reactor/broken reaction zones' regime, with strong finite-rate chemistry effects, are performed. Measured and predicted scalars compare well for the two jets with the lowest velocities. The two jets with the highest velocities experience extinction and reignition, and the simulations are able to capture the decrease and increase of the OH mass fractions, but the peak values are higher than in the experiments. Also numerical simulations of a lean premixed lifted jet flame with high sensitivity to turbulence modeling and chemical kinetics are performed. Limitations of the applied turbulence and

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

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.

Steady State Simulation of Two-Gas Phase Fluidized Bed Reactors in Series for Producing Linear Low Density Polyethylene

Directory of Open Access Journals (Sweden)

Ali Farhangiyan Kashani

2012-12-01

Full Text Available A linear low density polyethylene (LLDPE production process, including two- fuidized bed reactors in series (FBRS and other process equipment, was completely simulated by Aspen Polymer Plus software. Fluidized bed reactors were considered as continuous stirred tank reactors (CSTR consisted of polymer and gas phases. POLY-SRK and NRTL-RK equations of state were used to describe polymer and non-polymer streams, respectively. In this simulation, a kinetic model, based on a double active site heterogeneous Ziegler-Natta catalyst was used for simulation of LLDPE process consisting of two FBRS. Simulator using this model has the capability to  predict a number of  principal characteristics of LLDPE such as melt fow index (MFI, density, polydispersity index, numerical and weight average molecular weights (Mn,Mw and copolymer molar fraction (SFRAC. The results of the simulation were compared with industrial plant data and a good agreement was observed between the predicted model and plant data. The simulation results show the relative error of about 0.59% for prediction of polymer mass fow and 2.67% and 0.04% for prediction of product MFI and density, respectively.

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

Methodology of the On-Iine FoIIow Simulation of Pebble-bed High-temperature Reactors

International Nuclear Information System (INIS)

Xia Bing; Li Fu; Wei Chunlin; Zheng Yanhua; Chen Fubing; Zhang Jian; Guo Jiong

2014-01-01

The on-line fuel management is an essential feature of the pebble-bed high-temperature reactors (PB-HTRs), which is strongly coupled with the normal operation of the reactor. For the purpose of on-line analysis of the continuous shuffling scheme of numerous fuel pebbles, the follow simulation upon the real operation is necessary for the PB-HTRs. In this work, the on-line follow simulation methodology of the PB-HTRs’ operation is described, featured by the parallel treatments of both neutronics analysis and fuel cycling simulation. During the simulation, the operation history of the reactor is divided into a series of burn-up cycles according to the behavior of operation data, in which the steady-state neutron transport equations are solved and the diffusion theory is utilized to determine the physical features of the reactor core. The burn-up equations of heavy metals, fission products and neutron poisons including B-10, decoupled from the pebble flow term, are solved to analyze the burn-up process within a single burn-up cycle. The effect of pebble flow is simulated separately through a discrete fuel shuffling pattern confined by curved pebble flow channels, and the effect of multiple pass of the fuel is represented by logical batches within each spatial region of the core. The on-line thermal-hydraulics feedback is implemented for each bur-up cycle by using the real thermal-hydraulics data of the core operation. The treatment of control rods and absorber balls is carried out by utilizing a coupled neutron transport-diffusion calculation along with discontinuity factors. The physical models mentioned above are established mainly by using a revised version of the V.S.O.P program system. The real operation data of HTR-10 is utilized to verify the methodology presented in this work, which gives good agreement between simulation results and operation data. (author)

Development of variable-width ribbon heating elements for liquid-metal and gas-cooled fast breeder reactor fuel-pin simulators

International Nuclear Information System (INIS)

McCulloch, R.W.; Post, D.W.; Lovell, R.T.; Snyder, S.D.

1981-04-01

Variable-width ribbon heating elements that provide a chopped-cosine variable heat flux profile have been fabricated for fuel pin 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 are used to derive an analytical expression that precisely describes ribbon contour in terms of the major fabrication parameters. These parameters are used to generate numerical control tapes that control ribbon cutting and winding machines. Infrared scanning techniques are developed to determine the optimum transient thermal profile of the coils and relate this profile to that generated by the coils in completed fuel pin simulators

Development of numerical simulation technology for high resolution thermal hydraulic analysis

International Nuclear Information System (INIS)

Yoon, Han Young; Kim, K. D.; Kim, B. J.; Kim, J. T.; Park, I. K.; Bae, S. W.; Song, C. H.; Lee, S. W.; Lee, S. J.; Lee, J. R.; Chung, S. K.; Chung, B. D.; Cho, H. K.; Choi, S. K.; Ha, K. S.; Hwang, M. K.; Yun, B. J.; Jeong, J. J.; Sul, A. S.; Lee, H. D.; Kim, J. W.

2012-04-01

A realistic simulation of two phase flows is essential for the advanced design and safe operation of a nuclear reactor system. The need for a multi dimensional analysis of thermal hydraulics in nuclear reactor components is further increasing with advanced design features, such as a direct vessel injection system, a gravity driven safety injection system, and a passive secondary cooling system. These features require more detailed analysis with enhanced accuracy. In this regard, KAERI has developed a three dimensional thermal hydraulics code, CUPID, for the analysis of transient, multi dimensional, two phase flows in nuclear reactor components. The code was designed for use as a component scale code, and/or a three dimensional component, which can be coupled with a system code. This report presents an overview of the CUPID code development and preliminary assessment, mainly focusing on the numerical solution method and its verification and validation. It was shown that the CUPID code was successfully verified. The results of the validation calculations show that the CUPID code is very promising, but a systematic approach for the validation and improvement of the physical models is still needed

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)

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.)

An assessment of coupling algorithms for nuclear reactor core physics simulations

Energy Technology Data Exchange (ETDEWEB)

Hamilton, Steven, E-mail: hamiltonsp@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Berrill, Mark, E-mail: berrillma@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Clarno, Kevin, E-mail: clarnokt@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Pawlowski, Roger, E-mail: rppawlo@sandia.gov [Sandia National Laboratories, MS 0316, P.O. Box 5800, Albuquerque, NM 87185 (United States); Toth, Alex, E-mail: artoth@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Kelley, C.T., E-mail: tim_kelley@ncsu.edu [North Carolina State University, Department of Mathematics, Box 8205, Raleigh, NC 27695 (United States); Evans, Thomas, E-mail: evanstm@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States); Philip, Bobby, E-mail: philipb@ornl.gov [Oak Ridge National Laboratory, 1 Bethel Valley Rd., Oak Ridge, TN 37831 (United States)

2016-04-15

This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss–Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton–Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

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

Numerical evaluation of fluid mixing phenomena in boiling water reactor using advanced interface tracking method

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)

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

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

International Nuclear Information System (INIS)

Lee, Y. G.; Kim, J. W.; Yoon, S. J.; Park, G. C.

2010-10-01

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

Development of system analysis code for thermal-hydraulic simulation of integral reactor, Rex-10

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-10-15

Rex-10 is an environment-friendly and economical small-scale nuclear reactor to provide the energy for district heating as well as the electric power in micro-grid. This integral reactor comprises several innovative concepts supported by advanced primary circuit components, low coolant parameters and natural circulation cooling. To evaluate the system performance and thermal-hydraulic behavior of the reactor, a system analysis code is being developed so that the new designs and technologies adopted in Rex-10 can be reflected. The research efforts are absorbed in programming the simple and fast-running thermal-hydraulic analysis software. The details of hydrodynamic governing equations component models and numerical solution scheme used in this code are presented in this paper. On the basis of one-dimensional momentum integral model, the models of point reactor neutron kinetics for thorium-fueled core, physical processes in the steam-gas pressurizer, and heat transfers in helically coiled steam generator are implemented to the system code. Implicit numerical scheme is employed to momentum and energy equations to assure the numerical stability. The accuracy of simulation is validated by applying the solution method to the Rex-10 test facility. Calculated natural circulation flow rate and coolant temperature at steady-state are compared to the experimental data. The validation is also carried out for the transients in which the sudden reduction in the core power or the feedwater flow takes place. The code's capability to predict the steady-state flow by natural convection and the qualitative behaviour of the primary system in the transients is confirmed. (Author)

Numerical simulation of thermal stratification in cold legs by using OpenFOAM

International Nuclear Information System (INIS)

Cai, Jiejin; Watanabe, Tadashi

2011-01-01

During a small-break loss-of-coolant accident in pressurized water reactors (PWRs), emergency core cooling system (ECCS) is actuated and cold water is injected into cold legs. Insufficient mixing of injected cold water and hot primary coolant results in thermal stratification, which is a matter of concern for evaluation of pressurized thermal shock (PTS) in view of aging and life extension of nuclear power plants. In this study, an open source CFD software, OpenFOAM, is used to simulate mixing and thermal stratification in the cold leg of ROSA/LSTF, which is the largest thermal-hydraulic integral test facility simulating PWR. One of the cold-leg is numerically simulated from the outlet of primary coolant pump to the inlet of downcomer. ECCS water is injected from injection nozzle connected at the top of the cold leg into the steady-state natural circulation flow under high-pressure and high-temperature conditions. The temperature distribution in the cold leg is compared with experimental and FLUENT's results. Effects of turbulent flow models and secondary flow due to the elbow section of the cold leg are discussed for the case with the single-phase natural circulation. Injection into a two-phase stratified flow is also simulated and predictive and numerical capabilities of OpenFOAM are discussed. (author)

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)

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 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.

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

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

Development of sodium droplet combustion analysis methodology using direct numerical simulation in 3-dimensional coordinate (COMET)

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)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

Energy Technology Data Exchange (ETDEWEB)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-15

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

V.S.O.P. (99/09) computer code system for reactor physics and fuel cycle simulation. Version 2009

International Nuclear Information System (INIS)

Ruetten, H.J.; Haas, K.A.; Brockmann, H.; Ohlig, U.; Pohl, C.; Scherer, W.

2010-07-01

V.S.O.P. (99/ 09) represents the further development of V.S.O.P. (99/ 05). Compared to its precursor, the code system has been improved again in many details. The main motivation for this new code version was to update the basic nuclear libraries used by the code system. Thus, all cross section libraries involved in the code have now been based on ENDF/B-VII. V.S.O.P. is a computer code system for the comprehensive numerical simulation of the physics of thermal reactors. It implies the setup of the reactor and of the fuel element, processing of cross sections, neutron spectrum evaluation, neutron diffusion calculation in two or three dimensions, fuel burnup, fuel shuffling, reactor control, thermal hydraulics and fuel cycle costs. The thermal hydraulics part (steady state and time-dependent) is restricted to gas-cooled reactors and to two spatial dimensions. The code can simulate the reactor operation from the initial core towards the equilibrium core. This latest code version was developed and tested under the WINDOWS-XP - operating system. (orig.)

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 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)

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)

Implementation of a pressurized water reactor simulator for teaching on a mini-computer

International Nuclear Information System (INIS)

Tallec, Michele.

1982-06-01

This paper presents the design of a pressurized water reactor power plant simulator using a mini-computer. This simulator is oriented towards teaching. It operates real-time simulations and many parameters can be changed by the student during execution of the digital code. First, a state variable model of the dynamic behavior of the plant is derived from the physical laws. The second part presents the problems associated with the use of a mini-computer for the resolution of a large differential system, notably the problems of memory-space availability, execution time and numerical integration. Finally, it contains the description of the control deck outlay used to interfer with the digital code, and of the the conditions that can be changed during an excution [fr

Modeling and simulation of a packed bed reactor for hydrogen by methanol steam reforming

International Nuclear Information System (INIS)

Aboudheir, A.; Idem, R.

2004-01-01

'Full text:' The performance of a catalytic packed bed tubular reactor for hydrogen production depends on mass transport characteristics and temperature distribution in the reactor. To accurately predict this performance, a rigorous numerical model has been developed based on coupled mass, energy, and momentum balance equations in cylindrical coordinates. This comprehensive model takes into account the variations of the concentration and temperature in both the axial and radial directions as well as the pressure drop along the packed reactor. Also, experimental measurements for hydrogen production were collected using a manganese-promoted co-precipitated Cu-Al catalyst for methanol-steam reforming in a micro-reactor having 10 mm i.d. and 460 mm overall length. The operating temperature ranged from 443 to 523 K and the space-time ranged from 0.1 to 2.5 kg cat h/kmol CH3OH. The simulation results were found to be in close agreement with the experimental data over the various operating conditions. This confirms the validity of both the numerical model of this work and our previous published kinetics models for this reaction system. In addition, the model formulation is applicable to handle reactions, not only for the microreactor presented in this work, but also, for other laboratory size and industrial scale processes for hydrogen production by hydrocarbon reformation. (author)

Dynamic simulation of a sodium-cooled fast reactor power plant

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant.

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-08-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is the subject of this dissertation. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. A simulator referred to as the CRBRP-DCHT simulator was constructed for studying transients due to unexpected accidents followed by reactor scram. In this simulator emphasis was placed on simulating the auxiliary heat removal system, in order to determine its capability to remove the after-shut down fission and decay heat. The transients studied using the two versions of the ATWS simulator include step and ramp reactivity perturbations, and an electrical load perturbation in the controlled plant. An uncontrolled control rod withdrawal followed by reactor scram was studied using the DCHT simulator, although the duration of this transient was restricted to 20 sec. because of computer limitations. The results agree very well with the expected physical behavior of the plant

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

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 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

Science.gov (United States)

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.

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.)

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.

Physical modeling and numerical simulation of subcooled boiling in one- and three-dimensional representation of bundle geometry

International Nuclear Information System (INIS)

Bottoni, M.; Lyczkowski, R.; Ahuja, S.

1995-01-01

Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models

Numerical models for the analysis of thermal behavior and coolability of a particulate debris bed in reactor lower head

Energy Technology Data Exchange (ETDEWEB)

Ahn, Kwang Il; Kim, Sang Baik; Kim, Byung Seok [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-04-01

This report provides three distinctive, but closely related numerical models developed for the analysis of thermal behavior and coolability of a particulate debris bed that is may be formed inside the reactor lower head during severe accident late phases. The first numerical module presented in the report, MELTPRO-DRY, is used to analyze numerically heat-up and melting process of the dry particle bed, downward- and sideward-relocation of the liquid melt under gravity force and capillary force acting among porous particles, and solidification of the liquid melt relocated into colder region. The second module, MELTPROG-WET, is used to simulate numerically the cooling process of the particulate debris bed under the existence of water, which is subjected to two types of numerical models. The first type of WET module utilizes distinctive models that parametrically simulate the water cooling process, that is, quenching region, dryout region, and transition region. The choice of each parametric model depends on temperature gradient between the cooling water and the debris particles. The second type of WET module utilizes two-phase flow model that mechanically simulates the cooling process of the debris bed. For a consistent simulation from the water cooling to the dryout debris bed, on the other hand, the aforementioned two modules, MELTPROG-DRY and MELTPROG-WET, were integrated into a single computer program DBCOOL. Each of computational models was verified through limited applications to a heat-generating particulate bed contained in the rectangular cavity. 22 refs., 5 figs., 2 tabs. (Author)

Development and Experimental Verification of the Numerical Simulation Method for the Quasi-Steady SWR Phenomena in an LMR Steam Generator

International Nuclear Information System (INIS)

Eoh, Jae-Hyuk; Jeong, Ji-Young; Kim, Seong-O; Hahn, Dohee; Park, Nam-Cook

2005-01-01

A quasi-steady system analysis of the sodium-water reaction (SWR) phenomena in a liquid-metal reactor (LMR) was performed using the Sodium-water reaction Event Later Phase System Transient Analyzer (SELPSTA) computer simulation code. The code has been formulated by implementing various physical assumptions to simplify the complex SWR phenomena, and it adopts the long-term mass and energy transfer (LMET) model developed in the present study. The LMET model is based on the hypothesis that the system transient can be described by the pressure and temperature transient of the cover gas space, and it can be applied only to the reaction period characterized by bulk motion. To evaluate the feasibility of the physical model and its assumptions, a scale-down mock-up test was carried out, and it was demonstrated that the numerical simulation using the LMET model adequately replicates the overall phenomena of the experiment with reasonable understanding. Based on the findings, as a numerical example, the long-term system transient responses during the SWR event of the Korea Advanced LIquid MEtal Reactor (KALIMER) were investigated, and it was found that the long-term dynamic responses are strongly dependent on the design parameters and operational strategies. As a result, the numerical simulation method developed in the present study is practicable; furthermore, the SELPSTA code is useful to resolve the risk for the SWR event

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.)

Tightly Coupled Multiphysics Algorithm for Pebble Bed Reactors

International Nuclear Information System (INIS)

Park, HyeongKae; Knoll, Dana; Gaston, Derek; Martineau, Richard

2010-01-01

We have developed a tightly coupled multiphysics simulation tool for the pebble-bed reactor (PBR) concept, a type of Very High-Temperature gas-cooled Reactor (VHTR). The simulation tool, PRONGHORN, takes advantages of the Multiphysics Object-Oriented Simulation Environment library, and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN's algorithmic capability is demonstrated on a number of PBR test cases.

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 simulation of flow characteristics of lean jet to cross-flow in safety injection of reactor cooling system

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)

Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor; Desarrollo de un simulador para diseno y prueba de controladores de potencia en un reactor TRIGA Mark III

Energy Technology Data Exchange (ETDEWEB)

Perez M, C.; Benitez R, J.S.; Lopez C, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

2003-07-01

The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

Science.gov (United States)

Garber, Anne E.; Dickens, Ricky E.

2011-01-01

The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

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.)

Wave propagation simulation in the upper core of sodium-cooled fast reactors using a spectral-element method for heterogeneous media

Science.gov (United States)

Nagaso, Masaru; Komatitsch, Dimitri; Moysan, Joseph; Lhuillier, Christian

2018-01-01

ASTRID project, French sodium cooled nuclear reactor of 4th generation, is under development at the moment by Alternative Energies and Atomic Energy Commission (CEA). In this project, development of monitoring techniques for a nuclear reactor during operation are identified as a measure issue for enlarging the plant safety. Use of ultrasonic measurement techniques (e.g. thermometry, visualization of internal objects) are regarded as powerful inspection tools of sodium cooled fast reactors (SFR) including ASTRID due to opacity of liquid sodium. In side of a sodium cooling circuit, heterogeneity of medium occurs because of complex flow state especially in its operation and then the effects of this heterogeneity on an acoustic propagation is not negligible. Thus, it is necessary to carry out verification experiments for developments of component technologies, while such kind of experiments using liquid sodium may be relatively large-scale experiments. This is why numerical simulation methods are essential for preceding real experiments or filling up the limited number of experimental results. Though various numerical methods have been applied for a wave propagation in liquid sodium, we still do not have a method for verifying on three-dimensional heterogeneity. Moreover, in side of a reactor core being a complex acousto-elastic coupled region, it has also been difficult to simulate such problems with conventional methods. The objective of this study is to solve these 2 points by applying three-dimensional spectral element method. In this paper, our initial results on three-dimensional simulation study on heterogeneous medium (the first point) are shown. For heterogeneity of liquid sodium to be considered, four-dimensional temperature field (three spatial and one temporal dimension) calculated by computational fluid dynamics (CFD) with Large-Eddy Simulation was applied instead of using conventional method (i.e. Gaussian Random field). This three-dimensional numerical

Development of a numerical tool for safety assessment and emergency management of experimental reactors

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

Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

International Nuclear Information System (INIS)

Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won

2014-01-01

We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute 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 a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface

Development of Research Reactor Simulator and Its Application to Dynamic Test-bed

Energy Technology Data Exchange (ETDEWEB)

Kwon, Kee Choon; Park, Jae Chang; Lee, Seung Wook; Bang, Dane; Bae, Sung Won [KAERI, Daejeon (Korea, Republic of)

2014-08-15

We developed HANARO and the Jordan Research and Training Reactor (JRTR) real-time simulator for operating staff training. The main purpose of this simulator is operator training, but we modified this simulator as a dynamic test-bed to test the reactor regulating system in HANARO or JRTR before installation. The simulator configuration is divided into hardware and software. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The simulator software is divided into three major parts: a mathematical modeling module, which executes the plant dynamic modeling program in real-time, an instructor station module that manages user instructions, and a human machine interface (HMI) module. The developed research reactors are installed in the Korea Atomic Energy Research Institute 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 a hardware controller and the simulator and target controller were interfaced with a hard-wired and network-based interface.

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

Contribution to numerical methods for all Mach flow regimes and to fluid-porous coupling for the simulation of homogeneous two-phase flows in nuclear reactors

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

Pebble bed reactors simulation using MCNP: The Chinese HTR-10 reactor

Directory of Open Access Journals (Sweden)

SA Hosseini

2013-09-01

Full Text Available   Given the role of Gas-Graphite reactors as the fourth generation reactors and their recently renewed importance, in 2002 the IAEA proposed a set of Benchmarking problems. In this work, we propose a model both efficient in time and resources and exact to simulate the HTR-10 reactor using MCNP-4C code. During the present work, all of the pressing factors in PBM reactor design such as the inter-pebble leakage, fuel particle distribution and fuel pebble packing fraction effects have been taken into account to obtain an exact and easy to run model. Finally, the comparison between the results of the present work and other calculations made at INEEL proves the exactness of the proposed model.

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)

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

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

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

CFD-PBM Coupled Simulation of an Airlift Reactor with Non-Newtonian Fluid

Directory of Open Access Journals (Sweden)

Han Mei

2017-09-01

Full Text Available Hydrodynamics of an AirLift Reactor (ALR with tap water and non-Newtonian fluid was studied experimentally and by numerical simulations. The Population Balance Model (PBM with multiple breakup and coalescence mechanisms was used to describe bubble size characteristics in the ALR. The interphase forces for closing the two-fluid model were formulated by considering the effect of Bubble Size Distribution (BSD. The BSD in the ALR obtained from the coupled Computational Fluid Dynamics (CFD-PBM model was validated against results from digital imaging measurements. The simulated velocity fields of both the gas and liquid phases were compared to measured fields obtained with Particle Image Velocimetry (PIV. The simulated results show different velocity field profile features at the top of the ALR between tap water and non-Newtonian fluid, which are in agreement with experiments. In addition, good agreement between simulations and experiments was obtained in terms of overall gas holdup and bubble Sauter mean diameter.

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.

TRANPZ - A computer code for the simulation of reactors with axial dependence

International Nuclear Information System (INIS)

Sampaio, L.C.M.

1980-12-01

A computer code was developed to simulate a PWR reactor in steady state and during transients. The solution of one speed diffusion equation in the axial direction is obtained numerically dividing the core in various axial segments and the axial power distribution is obtained there from. A method was developed to determine the transient solution. The external reactivity effects are caused by the motion of the control rods, starting from the steady condition with the control rods in any position. The heat conduction equation in the fuel is numerically solved in the radial direction. Various tests were performed in steady state and transient conditions and the validity of the present model was verified. Results were compared in steady state condition with the code CITATION and a reasonable agreement was found. (E.G.) [pt

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

Computer simulation of the NASA water vapor electrolysis reactor

Science.gov (United States)

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.

Numerical study of radial stepwise fuel load reshuffling traveling wave reactor

International Nuclear Information System (INIS)

Zhang Dalin; Zheng Meiyin; Tian Wenxi; Qiu Suizheng; Su Guanghui

2015-01-01

Traveling wave reactor is a new conceptual fast breeder reactor, which can adopt natural uranium, depleted uranium and thorium directly to realize the self sustainable breeding and burning to achieve very high fuel utilization fraction. Based on the mechanism of traveling wave reactor, a concept of radial stepwise fuel load reshuffling traveling wave reactor was proposed for realistic application. It was combined with the typical design of sodium-cooled fast reactors, with which the asymptotic characteristics of the inwards stepwise fuel load reshuffling were studied numerically in two-dimension. The calculated results show that the asymptotic k_e_f_f parabolically varies with the reshuffling cycle length, while the burnup increases linearly. The highest burnup satisfying the reactor critical condition is 38%. The power peak shifts from the fuel discharging zone (core centre) to the fuel uploading zone (core periphery) and correspondingly the power peaking factor decreases along with the reshuffling cycle length. In addition, at the high burnup case the axial power distribution close to the core centre displays the M-shaped deformation. (authors)

Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

Energy Technology Data Exchange (ETDEWEB)

Bonaccorsi, Th

2007-09-15

A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

Development of research reactor simulator and its application to dynamic test-bed

International Nuclear Information System (INIS)

Kwon, Kee-Choon; Baang, Dane; Park, Jae-Chang; Lee, Seung-Wook; Bae, Sung Won

2014-01-01

We developed a real-time simulator for 'High-flux Advanced Neutron Application ReactOr (HANARO), and the Jordan Research and Training Reactor (JRTR). The main purpose of this simulator is operator training, but we modified this simulator into a dynamic test-bed (DTB) to test the functions and dynamic control performance of reactor regulating system (RRS) in HANARO or JRTR before installation. The simulator hardware consists of a host computer, 6 operator stations, a network switch, and a large display panel. The software includes a mathematical model that implements plant dynamics in real-time, an instructor station module that manages user instructions, and a human machine interface module. The developed research reactor simulators are installed in the Korea Atomic Energy Research Institute 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, 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, 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. The test result shows that the developed DTB and actual RRS cabinet works together simultaneously resulting in quite good dynamic control performances. (author)

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)

Numerical study of methanol–steam reforming and methanol–air catalytic combustion in annulus reactors for hydrogen production

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.

Efficient approach for simulating response of multi-body structure in reactor core subjected to seismic loading

International Nuclear Information System (INIS)

Zhang Hongkun; Cen Song; Wang Haitao; Cheng Huanyu

2012-01-01

An efficient 3D approach is proposed for simulating the complicated responses of the multi-body structure in reactor core under seismic loading. By utilizing the rigid-body and connector functions of the software Abaqus, the multi-body structure of the reactor core is simplified as a mass-point system interlinked by spring-dashpot connectors. And reasonable schemes are used for determining various connector coefficients. Furthermore, a scripting program is also complied for the 3D parametric modeling. Numerical examples show that, the proposed method can not only produce the results which satisfy the engineering requirements, but also improve the computational efficiency more than 100 times. (authors)

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

User's guide of DETRAS system-3. Description of the simulated reactor plant

International Nuclear Information System (INIS)

Yamaguchi, Yukichi

2006-12-01

DETRAS system is a PWR reactor simulator system for operation trainings whose distinguished feature is that it can be operated from the remote place of the simulator site. The document which is the third one of a series of three volumes of the user's guide of DETRAS, describes firstly an outline of the simulated reactor system then a user's interface needed for operation of the simulator of interest and finally a series of procedure for startup of the simulated reactor and shutdown of it from its rated operation state. (author)

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 Thermal-hydraulic Process in Reactor of HTR-PM

International Nuclear Information System (INIS)

Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

2014-01-01

This paper provides the physical process in the reactor of High Temperature Gas-cooled Reactor Pebble-bed Module (HTR-PM) and introduces the standard operation conditions. The FORTRAN code developed for the thermal hydraulic module of Full-Scale Simulator (FSS) of HTR-PM is used to simulate two typical operation transients including cold startup process and cold shutdown process. And the results were compared to the safety analysis code, namely TINTE. The good agreement indicates that the code is applicable for simulating the thermal-hydraulic process in reactor of HTR-PM. And for long time transient process, the code shows good stability and convergence. (author)

A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors - 202

International Nuclear Information System (INIS)

Recktenwald, G.D.; Bronk, L.A.; Deinert, M.R.

2010-01-01

Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks. (authors)

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

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

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.

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 and experimental research for the natural convection in an annular space in LMFBR

International Nuclear Information System (INIS)

Wang Zhou; Luo Rui; Yang Xianyong; Liang Taofeng

1999-01-01

In a pool fast reactor, the roof structure is penetrated by a number of pumps and heat exchanger units to form some annular spaces with various sizes. The natural convection of argon gas happens in the pool sky and the small annular gaps between those components and the roof containment due to thermosiphonic effects. The natural convection is studied experimentally and numerically to predict the temperature distributions inside the annular space and its surrounding structure. Numerical simulation is performed by using LVEL turbulence model and extending computational domain to the entire pool sky. The predicted results are in fair agreement with the experimental data. In comparison with commonly used k-ε model, LVEL model has better accuracy for the turbulent flow in a gap space

Simulation and calculation of three-reactor system of catalytic reforming

International Nuclear Information System (INIS)

Rikalovska, Tatjana; Markovska, Liljana; Meshko, Vera; Poposka, Filimena

1999-01-01

The process of catalytic reforming has been operated for quite a long time, one can not always find real data for the kinetics and thermodynamics of the reactions that take place during the catalytic reforming process in order to facilitate the designing of reactor system or its simulation in a wide:ran e of process parameters. Kinetic and thermodynamic data have been collected for the reactions that take place during the catalytic reforming process. The stress has been pointed on four major reactions: dehydrogenation of naphthenes (aromatization), dehydrocyclization of paraffins and hydrocracking of naphthenes and paraffins. On the base of such a kinetic model, the reforming process has been described with a system of differential equations. For the purpose of solving these equations computer programs for simulation of a three-reactor system for adiabatic operation of the reactors. The computer simulation of the mathematical model of this three-reactor system has been accomplished by use of the ISIM-dynamic simulator. The results obtained out of the simulation agree very good with the data of the real process of catalytic reforming in OKTA Crude Oil Refinery in Skopje, Macedonia. (Author)

A small-scale experimental reactor combined with a simulator for training purposes

International Nuclear Information System (INIS)

Destot, M.; Hagendorf, M.; Vanhumbeeck, D.; Lecocq-Bernard, J.

1981-01-01

The authors discuss how a small-scale reactor combined to a training simulator can be a valuable aid in all forms of training. They describe the CEN-based SILOETTE reactor in Grenoble and its combined simulator. They also take a look at prospects for the future of the system in the light of experience acquired with the ARIANE reactor and the trends for the development of simulators for training purposes [fr

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

Numerical simulation of TIG welding with filler of steel pieces of high thickness

Energy Technology Data Exchange (ETDEWEB)

Carmignani, B.; Toselli, G. [ENEA Centro Ricerche Ezio Clementel, Bologna (Italy). Dipt. Innovazione

1999-07-01

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. [Italian] Con il codice di analisi termo-strutturale non lineare agli EF, ABAQUS/S, viene affrontato per la prima volta il problema della simulazione numerica di un processo di saldatura con materiale d'apporto, in particolare il processo di saldatura TIG (tungsten inert gas) a filo freddo. Si e' fatto riferimento ad alcuni modelli sperimentali studiati e preparati ad hoc per conoscere meglio e valutare le metodologie di calcolo adottate ed i corrispondenti risultati ottenuti. Questo tipo di simulazione e' stato richiesto per fornire alcune conoscenze preliminari, che potessero essere di aiuto e guida nelle scelte di saldatura da eseguire per la fabbricazione e la quality assurance delle casse che dovranno contenere le bobine per la creazione del campo magnetico toroidale nella macchina ITER (international thermonuclear experiemental reactor) per la fusione nucleare. In questo lavoro, sono presentati, analizzati, discussi e confrontati con le corrispondenti misure sperimentali i primi risultati ottenuti. Sono descritte e discusse anche le difficolta' incontrate, le approssimazioni fatte e la ricerca di procedure di calcolo piu' semplificate.

Simulation of the behaviour of small and medium nuclear reactors on PCs

International Nuclear Information System (INIS)

Lyon, R.B.

1999-01-01

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. One of the simulation programs distributed by the IAEA is the the Advanced Reactor Simulator which simulates the behaviour of BWR, PWR and HWR reactor types. For this package, the modeling approach and assumptions are broadly described, together with a general description of the operation of the computer program. (author)

Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor

International Nuclear Information System (INIS)

Hernandez G, L.; Salazar S, E.

2014-10-01

This work is planned to simulate a scenario in which the same conditions that caused the accident at the Fukushima Daichi nuclear power plant are present, using a simulator of a nuclear power plant with VVER-1000 reactor, a different type of technology to the NPP where the accident occurred, which used BWR reactors. The software where it will take place the simulation was created and distributed by the IAEA for academic purposes, which contains the essential systems that characterize this type of NPP. The simulator has tools for the analysis of the characteristic phenomena of a VVER-1000 reactor in the different systems together and planned training tasks. This makes possible to identify the function of each component and how connects to other systems, thus facilitating the visualization of possible failures and the consequences that they have on the general behavior of the reactor. To program the conditions in the simulator, is necessary to know and synthesize a series of events occurred in Fukushima in 2011 and the realized maneuvers to reduce the effects of the system failures. Being different technologies interpretation of the changes that would suffer the VVER systems in the scenario in question will be developed. The Fukushima accident was characterized by the power loss of regular supply and emergency of the cooling systems which resulted in an increase in reactor temperature and subsequent fusion of their nuclei. Is interesting to reproduce this type of failure in a VVER, and extrapolate the lack of power supply in the systems that comprise, as well as pumping systems for cooling, has a pressure regulating system which involves more variables in the balance of the system. (Author)

A syncopated leap-frog algorithm for orbit consistent plasma simulation of materials processing reactors

International Nuclear Information System (INIS)

Cobb, J.W.; Leboeuf, J.N.

1994-01-01

The authors present a particle algorithm to extend simulation capabilities for plasma based materials processing reactors. The orbit integrator uses a syncopated leap-frog algorithm in cylindrical coordinates, which maintains second order accuracy, and minimizes computational complexity. Plasma source terms are accumulated orbit consistently directly in the frequency and azimuthal mode domains. Finally they discuss the numerical analysis of this algorithm. Orbit consistency greatly reduces the computational cost for a given level of precision. The computational cost is independent of the degree of time scale separation

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

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

Petascale algorithms for reactor hydrodynamics

International Nuclear Information System (INIS)

Fischer, P.; Lottes, J.; Pointer, W.D.; Siegel, A.

2008-01-01

We describe recent algorithmic developments that have enabled large eddy simulations of reactor flows on up to P = 65, 000 processors on the IBM BG/P at the Argonne Leadership Computing Facility. Petascale computing is expected to play a pivotal role in the design and analysis of next-generation nuclear reactors. Argonne's SHARP project is focused on advanced reactor simulation, with a current emphasis on modeling coupled neutronics and thermal-hydraulics (TH). The TH modeling comprises a hierarchy of computational fluid dynamics approaches ranging from detailed turbulence computations, using DNS (direct numerical simulation) and LES (large eddy simulation), to full core analysis based on RANS (Reynolds-averaged Navier-Stokes) and subchannel models. Our initial study is focused on LES of sodium-cooled fast reactor cores. The aim is to leverage petascale platforms at DOE's Leadership Computing Facilities (LCFs) to provide detailed information about heat transfer within the core and to provide baseline data for less expensive RANS and subchannel models.

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.

Effect of in-core instrumentation mounting location on external reactor vessel cooling

International Nuclear Information System (INIS)

Suh, Jungsoo; Ha, Huiun

2017-01-01

Highlights: • Numerical simulations were conducted for the evaluation of an IVR-ERVC application. • The ULPU-V experiment was simulated for the validation of numerical method. • The effect of ICI mounting location on an IVR-ERVC application was investigated. • TM-ICI is founded to be superior to BM-ICI for successful application of IVR-ERVC. - Abstract: The effect of in-core instrumentation (ICI) mounting location on the application of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC), used to mitigate severe accidents in which the nuclear fuel inside the reactor vessel becomes molten, was investigated. Numerical simulations of the subcooled boiling flow within an advanced pressurized-water reactor (PWR) in IVR-ERVC applications were conducted for the cases of top-mounted ICI (TM-ICI) and bottom-mounted ICI (BM-ICI), using the commercially available computational fluid dynamics (CFD) software ANSYS-CFX. Shear stress transport (SST) and the RPI model were used for turbulence closure and subcooled flow boiling, respectively. To validate the numerical method for IVR applications, numerical simulations of ULPU-V experiments were also conducted. The BM-ICI reactor vessel was modeled using a simplified design of an advanced PWR with BM-ICI; the TM-ICI counterpart was modeled by removing the ICI parts from the original geometry. It was found that TM-ICI was superior to BM-ICI for successful application of IVR-ERVC. For the BM-ICI case, the flow field was complicated because of the existence of ICIs and a significant temperature gradient was observed near the ICI nozzles on the lower part of the reactor vessel, where the ICIs were attached. These observations suggest that the existence of ICI below the reactor vessel hinders reactor vessel cooling.

Quasi-direct numerical simulation of a pebble bed configuration, Part-II: Temperature field analysis

International Nuclear Information System (INIS)

Shams, A.; Roelofs, F.; Komen, E.M.J.; Baglietto, E.

2013-01-01

Highlights: ► Quasi direct numerical simulations (q-DNSs) of a pebble bed configuration have been performed. ► This q-DNS database may serve as a reference for the validation of different turbulence modeling approaches. ► A wide range of qualitative and quantitative data throughout the computational domain has been generated. ► Results for mean, RMS of temperature and respective turbulent heat fluxes are extensively reported in this paper. -- Abstract: Good prediction of the flow and heat transfer phenomena in the pebble bed core of a high temperature reactor (HTR) is a challenge for available turbulence models, which still require to be validated. While experimental data are generally desirable in this validation process, due to the complex geometric configuration and measurement difficulties, a very limited amount of data is currently available. On the other hand, direct numerical simulation (DNS) is considered an accurate simulation technique, which may serve as an alternative for validating turbulence models. In the framework of the present study, quasi-direct numerical simulation (q-DNS) of a single face cubic centered pebble bed is performed, which will serve as a reference for the validation of different turbulence modeling approaches in order to perform calculations for a randomly arranged pebble bed. These simulations were performed at a Reynolds number of 3088, based on pebble diameter, with a porosity level of 0.42. Results related to flow field (mean, RMS and covariance of velocity) have been presented in Part-I, whereas, in the present article, we focus our attention to the analysis of the temperature field. A wide range of qualitative and quantitative data for the thermal field (mean, RMS and turbulent heat flux) has been generated

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)

Mitigation of numerical noise for beam loss simulations

CERN Document Server

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.

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

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

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

A domain-specific analysis system for examining nuclear reactor simulation data for light-water and sodium-cooled fast reactors

International Nuclear Information System (INIS)

Billings, Jay Jay; Deyton, Jordan H.; Forest Hull, S.; Lingerfelt, Eric J.; Wojtowicz, Anna

2015-01-01

Highlights: • Data analysis for high-performance simulations of reactors will be a problem that we address with a new management system. • We describe new input-output libraries for nuclear reactor simulations. • We describe a new user interface for visualizing and analyzing simulation results. • We show the utility of these systems with a 17 × 17 fuel assembly example simulation. • The availability of the code and avenues for collaboration are presented. - Abstract: Building a new generation of fission reactors in the United States presents many technical and regulatory challenges. One important challenge is the need to share and present results from new high-fidelity, high-performance simulations in an easily usable way. Since modern multiscale, multi-physics simulations can generate petabytes of data, they will require the development of new techniques and methods to reduce the data to familiar quantities of interest (e.g., pin powers, temperatures) with a more reasonable resolution and size. Furthermore, some of the results from these simulations may be new quantities for which visualization and analysis techniques are not immediately available in the community and need to be developed. This paper describes a new system for managing high-performance simulation results in a domain-specific way that naturally exposes quantities of interest for light water and sodium-cooled fast reactors. It describes requirements to build such a system and the technical challenges faced in its development at all levels (simulation, user interface, etc.). An example comparing results from two different simulation suites for a single assembly in a light-water reactor is presented, along with a detailed discussion of the system’s requirements and design

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.

Numerical simulation of hypersonic flight experiment vehicle

OpenAIRE

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

CERN Document Server

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.

Behavior of Type 316 stainless steel under simulated fusion reactor irradiation

International Nuclear Information System (INIS)

Wiffen, F.W.; Maziasz, P.J.; Bloom, E.E.; Stiegler, J.O.; Grossbeck, M.L.

1978-05-01

Fusion reactor irradiation response in alloys containing nickel can be simulated in thermal-spectrum fission reactors, where displacement damage is produced by the high-energy neutrons and helium is produced by the capture of two thermal neutrons in the reactions: 58 Ni + n → 59 Ni + γ; 59 Ni + n → 56 Fe + α. Examination of type 316 stainless steel specimens irradiated in HFIR has shown that swelling due to cavity formation and degradation of mechanical properties are more severe than can be predicted from fast reactor irradiations, where the helium contents produced are far too low to simulate fusion reactor service. Swelling values are greater and the temperature dependence of swelling is different than in the fast reactor case

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)

Multi-physic simulations of irradiation experiments in a technological irradiation reactor; Modelisation pluridisciplinaire d'experiences d'irradiation dans un reacteur d'irradiation technologique

Energy Technology Data Exchange (ETDEWEB)

Bonaccorsi, Th

2007-09-15

A Material Testing Reactor (MTR) makes it possible to irradiate material samples under intense neutron and photonic fluxes. These experiments are carried out in experimental devices localised in the reactor core or in periphery (reflector). Available physics simulation tools only treat, most of the time, one physics field in a very precise way. Multi-physic simulations of irradiation experiments therefore require a sequential use of several calculation codes and data exchanges between these codes: this corresponds to problems coupling. In order to facilitate multi-physic simulations, this thesis sets up a data model based on data-processing objects, called Technological Entities. This data model is common to all of the physics fields. It permits defining the geometry of an irradiation device in a parametric way and to associate information about materials to it. Numerical simulations are encapsulated into interfaces providing the ability to call specific functionalities with the same command (to initialize data, to launch calculations, to post-treat, to get results,... ). Thus, once encapsulated, numerical simulations can be re-used for various studies. This data model is developed in a SALOME platform component. The first application case made it possible to perform neutronic simulations (OSIRIS reactor and RJH) coupled with fuel behavior simulations. In a next step, thermal hydraulics could also be taken into account. In addition to the improvement of the calculation accuracy due to the physical phenomena coupling, the time spent in the development phase of the simulation is largely reduced and the possibilities of uncertainty treatment are under consideration. (author)

Nuclear ship engineering simulator

International Nuclear Information System (INIS)

Itoh, Yasuyoshi; Kusunoki, Tsuyoshi; Hashidate, Koji

1991-01-01

The nuclear ship engineering simulator, which analyzes overall system response of nuclear ship numerically, is now being developed by JAERI as an advanced design tool with the latest computer technology in software and hardware. The development of the nuclear ship engineering simulator aims at grasping characteristics of a reactor plant under the situation generated by the combination of ocean, a ship hull and a reactor. The data from various tests with the nuclear ship 'MUTSU' will be used for this simulator to modulate and verify its functions of reproducing realistic response of nuclear ship, and then the simulator will be utilized for the research and development of advanced marine reactors. (author)

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

Coupling of kinetic Monte Carlo simulations of surface reactions to transport in a fluid for heterogeneous catalytic reactor modeling

International Nuclear Information System (INIS)

Schaefer, C.; Jansen, A. P. J.

2013-01-01

We have developed a method to couple kinetic Monte Carlo simulations of surface reactions at a molecular scale to transport equations at a macroscopic scale. This method is applicable to steady state reactors. We use a finite difference upwinding scheme and a gap-tooth scheme to efficiently use a limited amount of kinetic Monte Carlo simulations. In general the stochastic kinetic Monte Carlo results do not obey mass conservation so that unphysical accumulation of mass could occur in the reactor. We have developed a method to perform mass balance corrections that is based on a stoichiometry matrix and a least-squares problem that is reduced to a non-singular set of linear equations that is applicable to any surface catalyzed reaction. The implementation of these methods is validated by comparing numerical results of a reactor simulation with a unimolecular reaction to an analytical solution. Furthermore, the method is applied to two reaction mechanisms. The first is the ZGB model for CO oxidation in which inevitable poisoning of the catalyst limits the performance of the reactor. The second is a model for the oxidation of NO on a Pt(111) surface, which becomes active due to lateral interaction at high coverages of oxygen. This reaction model is based on ab initio density functional theory calculations from literature.

Modeling and Simulation of the Multi-module High Temperature Gas-cooled Reactor

International Nuclear Information System (INIS)

Liu Dan; Sun Jun; Sui Zhe; Xu Xiaolin; Ma Yuanle; Sun Yuliang

2014-01-01

The modular high temperature gas-cooled reactor (MHTGR) is characterized with the inherent safety. To enhance its economic benefit, the capital cost of MHTGR can be decreased by combining more reactor modules into one unit and realize the batch constructions in the concept of modularization. In the research and design of the multi-module reactors, one difficulty is to clarify the coupling effects of different modules in operating the reactors due to the shared feed water and main steam systems in the secondary loop. In the advantages of real-time simulation and coupling calculations of different modules and sub-systems, the operation of multi-module reactors can be studied and analyzed to understand the range and extent of the coupling effects. In the current paper; the engineering simulator for the multi-module reactors was realized and able to run in high performance computers, based on the research experience of the HTR-PM engineering simulator. The models were detailed introduced including the primary and secondary loops. The steady state of full power operation was demonstrated to show the good performance of six-module reactors. Typical dynamic processes, such as adjusting feed water flow rates and shutting down one reactor; were also tested to study the coupling effects in multi-module reactors. (author)

Coupled CFD - system-code simulation of a gas cooled reactor

International Nuclear Information System (INIS)

Yan, Yizhou; Rizwan-uddin

2011-01-01

A generic coupled CFD - system-code thermal hydraulic simulation approach was developed based on FLUENT and RELAP-3D, and applied to LWRs. The flexibility of the coupling methodology enables its application to advanced nuclear energy systems. Gas Turbine - Modular Helium Reactor (GT-MHR) is a Gen IV reactor design which can benefit from this innovative coupled simulation approach. Mixing in the lower plenum of the GT-MHR is investigated here using the CFD - system-code coupled simulation tool. Results of coupled simulations are presented and discussed. The potential of the coupled CFD - system-code approach for next generation of nuclear power plants is demonstrated. (author)

Mathematical modelling of plant transients in the PWR for simulator purposes

International Nuclear Information System (INIS)

Hartel, K.

1984-01-01

This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

International Nuclear Information System (INIS)

Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim; Ashoub, Nagieb

2015-01-01

This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

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). Nuclear Research Center

2015-11-15

This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

Optimal Homogenization of Perfusion Flows in Microfluidic Bio-Reactors: A Numerical Study

DEFF Research Database (Denmark)

Okkels, Fridolin; Dufva, Martin; Bruus, Henrik

2011-01-01

In recent years, the interest in small-scale bio-reactors has increased dramatically. To ensure homogeneous conditions within the complete area of perfused microfluidic bio-reactors, we develop a general design of a continually feed bio-reactor with uniform perfusion flow. This is achieved...... by introducing a specific type of perfusion inlet to the reaction area. The geometry of these inlets are found using the methods of topology optimization and shape optimization. The results are compared with two different analytic models, from which a general parametric description of the design is obtained...... and tested numerically. Such a parametric description will generally be beneficial for the design of a broad range of microfluidic bioreactors used for, e. g., cell culturing and analysis and in feeding bio-arrays....

Model tests and numerical analysis on restoring force characteristics of reactor buildings

International Nuclear Information System (INIS)

Uchiyama, Y.; Suzuki, S.; Akino, K.

1987-01-01

Seismic shear walls of nuclear reactor buildings are composed of cylindrical, truncated cone-shape, box-shape, irregular polygonal walls or its combination and they are generally heavily reinforced concrete (RC) walls. So the elasto-plastic behaviors of those RC structures in ultimate regions have many unsolved and may be considered as especially important factors for explaining nonlinear response of nuclear reactor buildings. Following these research demands, the authors have prepared a nonlinear F.E.M. code called ''SANREF'' and made an extensive study for the restoring force characteristics of the inner concrete structures (I/C) of a PWR-type containment vessel and the principal seismic shear walls of a BWR-type reactor building by some series of reduced model tests and simulation analysis for the tests results. The detailed objectives of this study can be summarized as follows: (1) Examine the effectiveness of the configurations of shear walls, reinforcement ratios, shear span ratios (M/Qd) and vertical axial stress by ''partial model test'' which simulates some independent shear walls of the PWR-type and BWR-type reactor buildings. (2) Obtain fundamental data of restoring force characteristics of the complex shaped RC structures by ''composite model test'' which models are composed of the partial model test specimens. (3) Verify the applicability of analytical methods and constitutive modelings in SANREF code for complex shaped RC structures through nonlinear simulation analysis for the composite model test

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

Startup transient simulation for natural circulation boiling water reactors in PUMA facility

International Nuclear Information System (INIS)

Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

2006-01-01

In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

Fusion core start-up, ignition, and burn simulations of reversed-field pinch (RFP) reactors

International Nuclear Information System (INIS)

Chu, Y.Y.

1988-01-01

A transient reactor simulation model is developed to investigate and simulate the start-up, ignition, and burn of a reversed-field pinch reactor. The simulation is based upon a spatially averaged plasma balance model with field profiles obtained from MHD quasi-equilibrium analysis. Alpha particle heating is estimated from Fokker-Planck calculations. The instantaneous plasma current is derived from a self-consistent circuit analysis for plasma/coil/eddy current interactions. The simulation code is applied to the TITAN RFP reactor design which features a compact, high-power-density reversed-field pinch fusion system. A contour analysis is performed using the steady-state global plasma balance. The results are presented with contours of constant plasma current. A saddle point is identified in the contour plot which determined the minimum value of plasma current required to achieve ignition. In the simulations of the TITAN RFP reactor, the OH-driven super-conducting EF coils are found to deviate from the required equilibrium values as the induced plasma current increases. A set of basic results from the simulation of TITAN RFP reactor yield a picture of RFP plasma operation in a reactor. Investigations of eddy currents are also presented and have very important in reactor design

Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

International Nuclear Information System (INIS)

Yamada, Tomonori

2010-01-01

The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (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)

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)

Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment

International Nuclear Information System (INIS)

Perez M, C.

2004-01-01

The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

Numerical simulation of radial compressor stage

Science.gov (United States)

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

OpenAIRE

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.

CFD simulation analysis and validation for CPR1000 pressurized water reactor

International Nuclear Information System (INIS)

Zhang Mingqian; Ran Xiaobing; Liu Yanwu; Yu Xiaolei; Zhu Mingli

2013-01-01

Background: With the rapid growth in the non-nuclear area for industrial use of Computational fluid dynamics (CFD) which has been accompanied by dramatically enhanced computing power, the application of CFD methods to problems relating to Nuclear Reactor Safety (NRS) is rapidly accelerating. Existing research data have shown that CFD methods could predict accurately the pressure field and the flow repartition in reactor lower plenum. But simulations for the full domain of the reactor have not been reported so far. Purpose: The aim is to determine the capabilities of the codes to model accurately the physical phenomena which occur in the full reactor vessel. Methods: The flow field of the CPR1000 reactor which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX. The pressure loss in reactor pressure vessel, the hydraulic loads of guide tubes and support columns, and the bypass flow of head dome were obtained by calculations for the full domain of the reactor. The results were validated by comparing with the determined reference value of the operating nuclear plant (LingAo nuclear plant), and the transient simulation was conducted in order to better understand the flow in reactor pressure vessel. Results: It was shown that the predicted pressure loss with CFD code was slightly different with the determined value (10% relative deviation for the total pressure loss), the hydraulic loads were less than the determined value with maximum relative deviation 50%, and bypass flow of head dome was approximately the same with determined value. Conclusion: This analysis practice predicts accurately the physical phenomena which occur in the full reactor vessel, and can be taken as a guidance for the nuclear plant design development and improve our understanding of reactor flow phenomena. (authors)

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)

Numerical simulation of "an American haboob"

OpenAIRE

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...

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.

CFD-DEM simulation of a conceptual gas-cooled fluidized bed nuclear reactor

International Nuclear Information System (INIS)

Almeida, Lucilla C.; Su, Jian

2015-01-01

Several conceptual designs of the fluidized-bed nuclear reactor have been proposed due to its many advantages over conventional nuclear reactors such as PWRs and BWRs. Amongst their characteristics, the enhanced heat transfer and mixing enables a more uniform temperature distribution, reducing the risk of hot-spot and excessive fuel temperature, in addition to resulting in a higher burnup of the fuel. Furthermore, the relationship between the bed height and reactor neutronics turns the coolant flow rate control into a power production mechanism. Moreover, the possibility of removing the fuel by gravity from the movable core in case of a loss-of-cooling accident increases its safety. High-accuracy modeling of particles and coolant flow in fluidized bed reactors is needed to evaluate reliably the thermal-hydraulic efficiency and safety margin. The two-way coupling between solid and fluid can account for high-fidelity solid-solid interaction and reasonable accuracy in fluid calculation and fluid-solid interaction. In the CFD-DEM model, the particles are modeled as a discrete phase, following the DEM approach, whereas the fluid flow is treated as a continuous phase, described by the averaged Navier-Stokes equations on a computational cell scale. In this work, the coupling methodology between Fluent and Rocky is described. The numerical approach was applied to the simulation of a bubbling fluidized bed and the results were compared to experimental data and showed good agreement. (author)

G4-STORK: A Geant4-based Monte Carlo reactor kinetics simulation code

International Nuclear Information System (INIS)

Russell, Liam; Buijs, Adriaan; Jonkmans, Guy

2014-01-01

Highlights: • G4-STORK is a new, time-dependent, Monte Carlo code for reactor physics applications. • G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. • G4-STORK was designed to simulate short-term fluctuations in reactor cores. • G4-STORK is well suited for simulating sub- and supercritical assemblies. • G4-STORK was verified through comparisons with DRAGON and MCNP. - Abstract: In this paper we introduce G4-STORK (Geant4 STOchastic Reactor Kinetics), a new, time-dependent, Monte Carlo particle tracking code for reactor physics applications. G4-STORK was built by adapting and expanding on the Geant4 Monte Carlo toolkit. The toolkit provides the fundamental physics models and particle tracking algorithms that track each particle in space and time. It is a framework for further development (e.g. for projects such as G4-STORK). G4-STORK derives reactor physics parameters (e.g. k eff ) from the continuous evolution of a population of neutrons in space and time in the given simulation geometry. In this paper we detail the major additions to the Geant4 toolkit that were necessary to create G4-STORK. These include a renormalization process that maintains a manageable number of neutrons in the simulation even in very sub- or supercritical systems, scoring processes (e.g. recording fission locations, total neutrons produced and lost, etc.) that allow G4-STORK to calculate the reactor physics parameters, and dynamic simulation geometries that can change over the course of simulation to illicit reactor kinetics responses (e.g. fuel temperature reactivity feedback). The additions are verified through simple simulations and code-to-code comparisons with established reactor physics codes such as DRAGON and MCNP. Additionally, G4-STORK was developed to run a single simulation in parallel over many processors using MPI (Message Passing Interface) pipes

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.

Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

International Nuclear Information System (INIS)

EL Khatib, H.H.A.

2013-01-01

The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

Dynamic simulation of a sodium-cooled fast reactor power plant

International Nuclear Information System (INIS)

Shinaishin, M.A.M.

1976-01-01

Simulation of the dynamic behavior of the Clinch River Breeder Reactor Plant (CRBRP) is dealt with. The range of transients under consideration extends from a moderate transient, of the type referred to as Anticipated Transient Without Scram (ATWS), to a transient initiated by an unexpected accident followed by reactor scram. The moderate range of transients can be simulated by a digital simulator referred to as the CRBRP ATWS simulator. Two versions of this simulator were prepared; in one, the plant controllers were not included, whereas, in the other, the controllers were incorporated. In addition to the usual assumption of lumped parameters, uniform heat transfer and point kinetics (prompt jump) have been the main approximations in this and other simulators (see below). Two different transport-delay models have also been installed in all simulators. The simulators were constructed using the DARE-P System, developed by the Electrical Engineering Department at the University of Arizona

Coupled high fidelity thermal hydraulics and neutronics for reactor safety simulations

International Nuclear Information System (INIS)

Vincent A. Mousseau; Hongbin Zhang; Haihua Zhao

2008-01-01

This work is a continuation of previous work on the importance of accuracy in the simulation of nuclear reactor safety transients. This work is qualitative in nature and future work will be more quantitative. The focus of this work will be on a simplified single phase nuclear reactor primary. The transient of interest investigates the importance of accuracy related to passive (inherent) safety systems. The transient run here will be an Unprotected Loss of Flow (ULOF) transient. Here the coolant pump is turned off and the un-SCRAM-ed reactor transitions from forced to free convection (Natural circulation). Results will be presented that show the difference that the first order in time truncation physics makes on the transient. The purpose of this document is to illuminate a possible problem in traditional reactor simulation approaches. Detailed studies need to be done on each simulation code for each transient analyzed to determine if the first order truncation physics plays an important role

Simulation of Water Gas Shift Zeolite Membrane Reactor

Science.gov (United States)

Makertiharta, I. G. B. N.; Rizki, Z.; Zunita, Megawati; Dharmawijaya, P. T.

2017-07-01

The search of alternative energy sources keeps growing from time to time. Various alternatives have been introduced to reduce the use of fossil fuel, including hydrogen. Many pathways can be used to produce hydrogen. Among all of those, the Water Gas Shift (WGS) reaction is the most common pathway to produce high purity hydrogen. The WGS technique faces a downstream processing challenge due to the removal hydrogen from the product stream itself since it contains a mixture of hydrogen, carbon dioxide and also the excess reactants. An integrated process using zeolite membrane reactor has been introduced to improve the performance of the process by selectively separate the hydrogen whilst boosting the conversion. Furthermore, the zeolite membrane reactor can be further improved via optimizing the process condition. This paper discusses the simulation of Zeolite Membrane Water Gas Shift Reactor (ZMWGSR) with variation of process condition to achieve an optimum performance. The simulation can be simulated into two consecutive mechanisms, the reaction prior to the permeation of gases through the zeolite membrane. This paper is focused on the optimization of the process parameters (e.g. temperature, initial concentration) and also membrane properties (e.g. pore size) to achieve an optimum product specification (concentration, purity).

Numerical simulation of heat transfer in metal foams

Science.gov (United States)

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.

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Numerical solution of the point reactor kinetics equations with fuel burn-up and temperature feedback

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.

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 Silicon-Ribbon Growth

Science.gov (United States)

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.

University Reactor Matching Grants Program

International Nuclear Information System (INIS)

John Valentine; Farzad Rahnema; Said Abdel-Khalik

2003-01-01

During the 2002 Fiscal year, funds from the DOE matching grant program, along with matching funds from the industrial sponsors, have been used to support research in the area of thermal-hydraulics. Both experimental and numerical research projects have been performed. Experimental research focused on two areas: (1) Identification of the root cause mechanism for axial offset anomaly in pressurized water reactors under prototypical reactor conditions, and (2) Fluid dynamic aspects of thin liquid film protection schemes for inertial fusion reactor chambers. Numerical research focused on two areas: (1) Multi-fluid modeling of both two-phase and two-component flows for steam conditioning and mist cooling applications, and (2) Modeling of bounded Rayleigh-Taylor instability with interfacial mass transfer and fluid injection through a porous wall simulating the ''wetted wall'' protection scheme in inertial fusion reactor chambers. Details of activities in these areas are given

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Experiments and Numerical Simulations of Electrodynamic Tether

Science.gov (United States)

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.

Development of a nuclear reactor control system simulator using virtual instruments

International Nuclear Information System (INIS)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

2011-01-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. 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 nuclear reactor control system simulator using virtual instruments

Energy Technology Data Exchange (ETDEWEB)

Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Takase, Kazuyuki

1994-11-01

The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

Direct Numerical Simulation of Driven Cavity Flows

NARCIS (Netherlands)

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 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.

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

3-Dimensional numerical simulation of sodium spray fire accidents in LMFBRs

International Nuclear Information System (INIS)

Zhang Bin; Zhu Jizhou; Han Lang

2005-01-01

In order to estimate and foresee the sequence of sodium spray fires that may occur in the liquid metal cooled fast breeder reactors (LMFBRs), this paper develops a program to analyze such sodium fire accidents. The present study gives a 3-dimensional numerical analysis code for sodium spray fires. The spatial distributions of gas temperature and chemical species concentrations in the cell that sodium spray fires happened are given. This paper gives detailed explanation of combustion models and heat transfer models that applied in the program. And the calculation procedure and method in solving the fluid field are narrated in detail. Good agreements of an overall transient behavior are obtained in a sodium spray combustion test analysis. The comparison between the analytical and experimental results shows that the program presented in this paper is creditable and reasonable for simulating the sodium spray fires. (author)

A PC-based high temperature gas reactor simulator for Indonesian conceptual HTR reactor basic training

Science.gov (United States)

Syarip; Po, L. C. C.

2018-05-01

In planning for nuclear power plant construction in Indonesia, helium cooled high temperature reactor (HTR) is favorable for not relying upon water supply that might be interrupted by earthquake. In order to train its personnel, BATAN has cooperated with Micro-Simulation Technology of USA to develop a 200 MWt PC-based simulation model PCTRAN/HTR. It operates in Win10 environment with graphic user interface (GUI). Normal operation of startup, power maneuvering, shutdown and accidents including pipe breaks and complete loss of AC power have been conducted. A sample case of safety analysis simulation to demonstrate the inherent safety features of HTR was done for helium pipe break malfunction scenario. The analysis was done for the variation of primary coolant pipe break i.e. from 0,1% - 0,5 % and 1% - 10 % helium gas leakages, while the reactor was operated at the maximum constant power of 10 MWt. The result shows that the highest temperature of HTR fuel centerline and coolant were 1150 °C and 1296 °C respectively. With 10 kg/s of helium flow in the reactor core, the thermal power will back to the startup position after 1287 s of helium pipe break malfunction.

Application of assembly module to high-temperature gas-cooled reactor full-scope simulation system

International Nuclear Information System (INIS)

Li Sifeng; Li Fu; Ma Yuanle; Shi Lei

2007-01-01

According to the circumstances that exist in the reactor full-scope simulators development as long development cycle, very difficult upgrade and narrow range of applicability, a kind of new model was developed based on assembly module which root in Linux kernel and successfully applied to the design of high-temperature gas-cooled reactor full-scope simulator system. The simulation results are coincident with the experimental ones, and it indicates that the new model based on assembly module is feasible to design of high-temperature gas cooled reactor simulation system. (authors)

Computational Flame Diagnostics for Direct Numerical Simulations with Detailed Chemistry of Transportation Fuels

Energy Technology Data Exchange (ETDEWEB)

Lu, Tianfeng [Univ. of Connecticut, Storrs, CT (United States)

2017-02-16

The goal of the proposed research is to create computational flame diagnostics (CFLD) that are rigorous numerical algorithms for systematic detection of critical flame features, such as ignition, extinction, and premixed and non-premixed flamelets, and to understand the underlying physicochemical processes controlling limit flame phenomena, flame stabilization, turbulence-chemistry interactions and pollutant emissions etc. The goal has been accomplished through an integrated effort on mechanism reduction, direct numerical simulations (DNS) of flames at engine conditions and a variety of turbulent flames with transport fuels, computational diagnostics, turbulence modeling, and DNS data mining and data reduction. The computational diagnostics are primarily based on the chemical explosive mode analysis (CEMA) and a recently developed bifurcation analysis using datasets from first-principle simulations of 0-D reactors, 1-D laminar flames, and 2-D and 3-D DNS (collaboration with J.H. Chen and S. Som at Argonne, and C.S. Yoo at UNIST). Non-stiff reduced mechanisms for transportation fuels amenable for 3-D DNS are developed through graph-based methods and timescale analysis. The flame structures, stabilization mechanisms, local ignition and extinction etc., and the rate controlling chemical processes are unambiguously identified through CFLD. CEMA is further employed to segment complex turbulent flames based on the critical flame features, such as premixed reaction fronts, and to enable zone-adaptive turbulent combustion modeling.

CFD Simulation of Dimethyl Ether Synthesis from Methanol in an Adiabatic Fixed-bed Reactor

Directory of Open Access Journals (Sweden)

Mohammad Golshadi

2013-04-01

Full Text Available A computational fluid dynamic (CFD study of methanol (MeOH to dimethyl ether (DME process in an adiabatic fixed-bed reactor is presented. One of the methods of industrial DME production is the catalytic dehydration of MeOH. Kinetic model was derived based on Bercic rate. The parameters of this equation for a specific catalyst were tuned by solving a one-dimensional homogenous model using MATLAB optimization module. A two-dimensional CFD simulation of the reaction is demonstrated and considered as numerical experiments. A sensitivity analysis was run in order to find the effect of temperature, pressure, and WHSV on the reactor performance. Good agreement was achieved between bench experimental data and the model. The results show that the maximum conversion of reaction (about 85.03% is obtained at WHSV=10 h-1 and T=563.15 K, whereas the inlet temperature has a greater effect on methanol conversion. Moreover, the effect of water in inlet feed on methanol conversion is quantitatively studied. It was concluded that the results obtained from CFD analysis give precise guidelines for further studies on the optimization of reactor performance.

Numerical simulations of rarefied gas flows in thin film processes

NARCIS (Netherlands)

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

Simulation and control of the site-dependent neutron density in a nuclear reactor

International Nuclear Information System (INIS)

Stark, K.

1974-01-01

The present work deals with the simulation and control of a pressurized-water reactor such as is used in nuclear power plants today. In the first part of the work, the mathematical model equations of the reactor are set up. They take into consideration the local distribution of the various reactor parameters as far as seems necessary for further investigations. Taking the given approximations, the mathematical model is locally one-dimensional; it is valid for the period of time in which a power control of the reactor must work. The model equations set up are calculated on an analog/hybride computer according to the modal simulation method in true time. The method is distinguished in the present problem here through good convergence and enables the observation of the simulation results as a stationary picture on an oscillograph screen. For this reason, a simulation of this type seems particularly suitable for the training of operational personnel. The aim of the second part of the work is the development of a simple control concept which enables the control of the total power of the reactor as well as of the distribution of the power density in the reactor core. The fundamentals of the control design are the non-linear system equations of the nuclear reactor. The developed control is based on the controlling of eigenfunctions; it controls the total power of the reactor as well as the distribution of the power density in the reactor core where a uniform burn-up of the nuclear fuel is seen to. Part-absorbing control rods amongst others are used as actuators like they are already used in that type of reactors. (orig./LH) [de

Requirements for advanced simulation of nuclear reactor and chemicalseparation plants.

Energy Technology Data Exchange (ETDEWEB)

Palmiotti, G.; Cahalan, J.; Pfeiffer, P.; Sofu, T.; Taiwo, T.; Wei,T.; Yacout, A.; Yang, W.; Siegel, A.; Insepov, Z.; Anitescu, M.; Hovland,P.; Pereira, C.; Regalbuto, M.; Copple, J.; Willamson, M.

2006-12-11

This report presents requirements for advanced simulation of nuclear reactor and chemical processing plants that are of interest to the Global Nuclear Energy Partnership (GNEP) initiative. Justification for advanced simulation and some examples of grand challenges that will benefit from it are provided. An integrated software tool that has its main components, whenever possible based on first principles, is proposed as possible future approach for dealing with the complex problems linked to the simulation of nuclear reactor and chemical processing plants. The main benefits that are associated with a better integrated simulation have been identified as: a reduction of design margins, a decrease of the number of experiments in support of the design process, a shortening of the developmental design cycle, and a better understanding of the physical phenomena and the related underlying fundamental processes. For each component of the proposed integrated software tool, background information, functional requirements, current tools and approach, and proposed future approaches have been provided. Whenever possible, current uncertainties have been quoted and existing limitations have been presented. Desired target accuracies with associated benefits to the different aspects of the nuclear reactor and chemical processing plants were also given. In many cases the possible gains associated with a better simulation have been identified, quantified, and translated into economical benefits.

Relap5 simulation for severe accident analysis of RSG-GAS Reactor

International Nuclear Information System (INIS)

Andi Sofrany Ekariansyah; Endiah P-Hastuti; Sudarmono

2018-01-01

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational characteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element. (author)

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 simulation of AP1000 LBLOCA with SCDAP/RELAP 4.0 code

International Nuclear Information System (INIS)

Xie Heng

2017-01-01

The risk of large-break loss of coolant accident (LBLOCA) is that core will be exposed once the accident occurs, and may cause core damages. New phenomena may occur in LBLOCA due to passive safety injection adopted by AP1000. This paper used SCDAP/RELAP5 4.0 to build the numerical model of AP1000 and double-end guillotine of cold leg is simulated. Reactor coolant system and passive core cooling system were modeled by RELAP5 modular. HEAT STRUCTURE component of RELAP5 was used to simulate the fuel rod. The reflood option in RELAP5 was chosen to be activated or not to study the effect of axial heat conduction. Results show that the axial heat conduction plays an important role in the reflooding phase and can effectively shorten reflood process. An alternative core model is built by SCDAP modular. It is found that the SCDAP model predicts higher maximum peak cladding temperature and longer reflood process than RELAP5 model. Analysis shows that clad oxidation heat plays a key role in the reflood. From the simulation results, it can be concluded that the cladding will keep intact and fission product will not be released from fuel to coolant in LBLOCA. (author)

Application of a Russian nuclear reactor simulator VVER-1000; Aplicacion de un simulador de reactor nuclear ruso VVER-1000

Energy Technology Data Exchange (ETDEWEB)

Lopez-Peniche S, A. [UNAM, Facultad de Ingenieria, Circuito Interior, Ciudad Universitaria, 04360 Mexico D. F. (Mexico); Salazar S, E., E-mail: alpsordo@hotmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

2012-10-15

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)

Simulators IV

International Nuclear Information System (INIS)

Fairchild, B.T.

1987-01-01

These proceedings contain papers on simulators with artificial intelligence, and the human decision making process; visuals for simulators: human factors, training, and psycho-physical impacts; the role of institutional structure on simulation projects; maintenance trainers for economic value and safety; biomedical simulators for understanding nature, for medical benefits, and the physiological effects of simulators; the mathematical models and numerical techniques that drive today's simulators; and the demography of simulators, with census papers identifying the population of real-time simulator training devices; nuclear reactors

A numerical approach to the simulation of one-phase and two phase reactor coolant flow around nuclear fuel spacers

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 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...

Processing biobased polymers using plasticizers: Numerical simulations versus experiments

Science.gov (United States)

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.

Numerical reactor evaluation

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 code for combustion of sodium liquid droplet and its verification

International Nuclear Information System (INIS)

Okano, Yasushi

1997-11-01

The computer programs for sodium leak and burning phenomena had been developed based on mechanistic approach. Direct numerical simulation code for sodium liquid droplet burning had been developed for numerical analysis of droplet combustion in forced convection air flow. Distributions of heat generation and temperature and reaction rate of chemical productions, such as sodium oxide and hydroxide, are calculated and evaluated with using this numerical code. Extended MAC method coupled with a higher-order upwind scheme had been used for combustion simulation of methane-air mixture. In the numerical simulation code for combustion of sodium liquid droplet, chemical reaction model of sodium was connected with the extended MAC method. Combustion of single sodium liquid droplet was simulated in this report for the verification of developed numerical simulation code. The changes of burning rate and reaction product with droplet diameter and inlet wind velocity were investigated. These calculation results were qualitatively and quantitatively conformed to the experimental and calculation observations in combustion engineering. It was confirmed that the numerical simulation code was available for the calculation of sodium liquid droplet burning. (author)

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)

Development of a training simulator to operators of the IEA-R1 research reactor

International Nuclear Information System (INIS)

Carvalho, Ricardo Pinto de

2006-01-01

This work reports the development of a Simulator for the IEA-R1 Research Reactor. The Simulator was developed with Visual C++ in two stages: construction of the mathematics models and development and configuration of graphics interfaces in a Windows XP executable. A simplified modeling was used for main physics phenomena, using a point kinetics model for the nuclear process and the energy and mass conservation laws in the average channel of the reactor for the thermal hydraulic process. The dynamics differential equations were solved by using finite differences through the 4th order Runge- Kutta method. The reactivity control, reactor cooling, and reactor protection systems were also modeled. The process variables are stored in ASCII files. The Simulator allows navigating by screens of the systems and monitoring tendencies of the operational transients, being an interactive tool for teaching and training of IEA-R1 operators. It also can be used by students, professors, and researchers in teaching activities in reactor and thermal hydraulics theory. The Simulator allows simulations of operations of start up, power maneuver, and shut down. (author)

Computational simulation of Argonauta/IEN nuclear reactor using MCNPX code

International Nuclear Information System (INIS)

Cunha, Victor Lusis Lassance; Silva Junior, Wilson F. Rebello da

2011-01-01

The study consisted of developing a computer simulation of a nuclear research reactor using the MCNPX. The reactor modeled is the Argonauta located at IEN (Rio de Janeiro) designed by Argonne National Laboratory (USA), which is primarily used for non-destructive testing with neutron beam and teaching purposes. It was entirely modeled with geometric fidelity, including detailed material description, shielding and irradiation channels. When available, the model was based on the as-built drawings. Four different simulations were made, the first set of two for criticality calculations and the other set for flux measurement. The first simulation set consisted of estimating the reactors reactivity. The second set consisted of placing detectors on specific places where the reactor is monitored and on the fuel axis covering the multiplicative and non-multiplicative media. Based on this data, the thermal neutron flux profile was plotted. All the outputs were compared with experimental data. Since it is a stochastic method, the statistical convergence was successfully checked for all simulations. The results were in good agreement with the experimental values. For the criticality calculations, the relative error was smaller then 1%. The flux measurements were also very well reproduced. The values were normalized for a reference point and the proportionality between the different spots was respected. The neutron flux profile along the core had the expected shape and values. Based on the good results, it can be said that the model is validated. (author)

Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

International Nuclear Information System (INIS)

Dahl, Pamela M.; Su, Jian

2017-01-01

Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

Simulation in CFD of a Pebble Bed: Advanced high temperature reactor core using OpenFOAM

Energy Technology Data Exchange (ETDEWEB)

Dahl, Pamela M.; Su, Jian, E-mail: sujian@nuclear.ufrj.br [Coordenacao de Pos-Graduacao e Pesquisa de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

2017-07-01

Numerical simulations of a Pebble Bed nuclear reactor core are presented using the multi-physics tool-kit OpenFOAM. The HTR-PM is modeled using the porous media approach, accounting both for viscous and inertial effects through the Darcy and Forchheimer model. Initially, cylindrical 2D and 3D simulations are compared, in order to evaluate their differences and decide if the 2D simulations carry enough of the sought information, considering the savings in computational costs. The porous medium is considered to be isotropic, with the whole length of the packed bed occupied homogeneously with the spherical fuel elements. Steady-state simulations for normal equilibrium operation are performed, using a semi sine function of the power density along the vertical axis as the source term for the energy balance equation.Total pressure drop is calculated and compared with that obtained from literature for a similar case. At a second stage, transient simulations are performed, where relevant parameters are calculated and compared to those of the literature. (author)

Numerical analysis on ingress-of-coolant events in fusion reactors with TRAC-PF1 code

International Nuclear Information System (INIS)

Ose, Yasuo; Takase, Kazuyuki; Akimoto, Hajime

2000-01-01

As for accident events related with thermal-hydraulics, in a fusion experimental reactor an ingress-of-coolant event (ICE) and a loss-of-vacuum-accident event (LOVA) should be considered. An integrated ICE/LOVA test apparatus is under planning in order to estimate quantitatively heat transfer and fluid flow characteristics under ICE and LOVA events. This study was carried out to predict numerically the thermal-hydraulic characteristics in fusion reactors at the ICE events before construction of the integrated ICE/LOVA test apparatus. The TRAC-PF1 code, which was originally developed for the thermal-hydraulic safety analysis in light water reactors, was used. The numerical analyses were performed for two kinds of system configuration with/without a pressure-suppression tank:the former for is investigation of the pressure rise characteristics and two-phase flow behavior; the latter for estimation of an effect of the pressure reduction due to the pressure-suppression tank. From the present analytical results, effects of the ingress water flow rate and vessel temperatures on the pressure rise ware clarified quantitatively. Furthermore, the pressure-rise suppression effect due to the vapor condensation in the pressure-suppression tank was predicted numerically. In addition, the useful information regarding to the design of the integrated ICE/LOVA test apparatus and the knowledge with respect to the effective usage of the TRAC-PF1 code were obtained through the present numerical study. (author)

Simulation test of PIUS-type reactor with large scale experimental apparatus

International Nuclear Information System (INIS)

Tamaki, M.; Tsuji, Y.; Ito, T.; Tasaka, K.; Kukita, Yutaka

1995-01-01

A large scale experimental apparatus for simulating the PIUS-type reactor has been constructed keeping the volumetric scaling ratio to the realistic reactor model. Fundamental experiments such as a steady state operation and a pump trip simulation were performed. Experimental results were compared with those obtained by the small scale apparatus in JAERI. We have already reported the effectiveness of the feedback control for the primary loop pump speed (PI control) for the stable operation. In this paper this feedback system is modified and the PID control is introduced. This new system worked well for the operation of the PIUS-type reactor even in a rapid transient condition. (author)

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)

Building a dynamic code to simulate new reactor concepts

International Nuclear Information System (INIS)

Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.

2012-01-01

Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.

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.)

3D numerical simulations of multiphase continental rifting

Science.gov (United States)

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

Numerical simulation and physical aspects of supersonic vortex breakdown

Science.gov (United States)

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.

Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

Science.gov (United States)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.

Experimental and Numerical Evaluation of the By-Pass Flow in a Catalytic Plate Reactor for Hydrogen Production

DEFF Research Database (Denmark)

Sigurdsson, Haftor Örn; Kær, Søren Knudsen

2011-01-01

Numerical and experimental study is performed to evaluate the reactant by-pass flow in a catalytic plate reactor with a coated wire mesh catalyst for steam reforming of methane for hydrogen generation. By-pass of unconverted methane is evaluated under different wire mesh catalyst width to reactor...

Numerical simulation of dual frequency etching reactors: Influence of the external process parameters on the plasma characteristics

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

Design of virtual SCADA simulation system for pressurized water reactor

International Nuclear Information System (INIS)

Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

2016-01-01

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

Design of virtual SCADA simulation system for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

2016-02-08

The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

Numerical Simulation of Steady Supercavitating Flows

OpenAIRE

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...

Simulation of the operational monitoring of a BWR with Simulate-3; Simulacion del seguimiento operacional de un reactor BWR con Simulate-3

Energy Technology Data Exchange (ETDEWEB)

Jimenez F, J. O.; Martin del Campo M, C.; Fuentes M, L.; Francois L, J. L., E-mail: ace.jo.cu@gmail.com [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico)

2015-09-15

This work was developed in order to describe the methodology for calculating the fuel burned of nuclear power reactors throughout the duration of their operating cycle and for each fuel reload. In other words, simulate and give monitoring to the main operation parameters of sequential way along its operation cycles. For this particular case, the operational monitoring of five consecutive cycles of a reactor was realized using the information reported by their processes computer. The simulation was performed with the Simulate-3 software and the results were compared with those of the process computer. The goal is to get the fuel burned, cycle after cycle for obtain the state conditions of the reactor needed for the fuel reload analyses, stability studies and transients analysis, and the development of a methodology that allows to manage and resolve similar cases for future fuel cycles of the nuclear power plant and explore the various options offered by the simulator. (Author)

A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

Energy Technology Data Exchange (ETDEWEB)

Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1998-12-31

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

A simple dynamic model and transient simulation of the nuclear power reactor on microcomputers

Energy Technology Data Exchange (ETDEWEB)

Han, Yang Gee; Park, Cheol [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-12-31

A simple dynamic model is developed for the transient simulation of the nuclear power reactor. The dynamic model includes the normalized neutron kinetics model with reactivity feedback effects and the core thermal-hydraulics model. The main objective of this paper demonstrates the capability of the developed dynamic model to simulate various important variables of interest for a nuclear power reactor transient. Some representative results of transient simulations show the expected trends in all cases, even though no available data for comparison. In this work transient simulations are performed on a microcomputer using the DESIRE/N96T continuous system simulation language which is applicable to nuclear power reactor transient analysis. 3 refs., 9 figs. (Author)

Optimal design of a composite space shield based on numerical simulations

International Nuclear Information System (INIS)

Son, Byung Jin; Yoo, Jeong Hoon; Lee, Min Hyung

2015-01-01

In this study, optimal design of a stuffed Whipple shield is proposed by using numerical simulations and new penetration criterion. The target model was selected based on the shield model used in the Columbus module of the international space station. Because experimental results can be obtained only in the low velocity region below 7 km/s, it is required to derive the Ballistic limit curve (BLC) in the high velocity region above 7 km/s by numerical simulation. AUTODYN-2D, the commercial hydro-code package, was used to simulate the nonlinear transient analysis for the hypervelocity impact. The Smoothed particle hydrodynamics (SPH) method was applied to projectile and bumper modeling to represent the debris cloud generated after the impact. Numerical simulation model and selected material properties were validated through a quantitative comparison between numerical and experimental results. A new criterion to determine whether the penetration occurs or not is proposed from kinetic energy analysis by numerical simulation in the velocity region over 7 km/s. The parameter optimization process was performed to improve the protection ability at a specific condition through the Design of experiment (DOE) method and the Response surface methodology (RSM). The performance of the proposed optimal design was numerically verified.

Numerical simulation of single bubbles rising through subchannels with interface tracking method

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Takuji Nagayoshi; Hidesada Tamai; Tazuyuki Takase; Hajime Akimoto

2005-01-01

Full text of publication follows: Although the sub-channel codes are used for the thermal-hydraulic analysis of fuel bundles in nuclear reactors from the former, many compositions and empirical equations based on experimental results are needed to predict the two-phase flow behavior in details. When there are no experimental data such as the reduced-moderation light water reactor (RMWR) which is studied by the Japan Atomic Energy Research Institute (JAERI), therefore, it is very difficult to obtain highly precise predictions. The RMWR core has remarkably narrow gap spacing between fuel rods (i.e., around 1 mm) which are arranged at a triangular tight-lattice configuration. To evaluate the feasibility and to optimize the thermal design of the RMWR core, a full-scale bundle test is required. However, several systematic full-scale tests are difficult to perform during an initial design phase from economic and temporal reason. Thus, we made a plan to develop a mechanistic BT model to evaluate the effects of the geometry configuration by a two-phase flow numerical simulation. In the plan of the mechanistic BT model development, three dimensional two-phase flow simulation codes with the interface tracking method, the moving particle semi-implicit method and the advanced two-fluid model are developed. In this study, as a part of this model development, detailed two-phase flow simulation code using interface tracking method (named TPFIT) is developed. In this paper, the results of TPFIT code with the advanced interface tracking method applied to single bubbles behavior through subchannels) to verify TPFIT code performance in complicated flow channel as rod bundles. In the simulation, the flow channel is composed of a square duct and four tubes with outside diameters D = 12 mm. The width and height of the duct are 27.2 mm and 192 mm, respectively. In the flow channel, the tubes are used to simulate fuel rods. One center subchannel and four periphery subchannels exist in the

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.

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

Forced vibration tests and simulation analyses of a nuclear reactor building. Part 2: simulation analyses

International Nuclear Information System (INIS)

Kuno, M.; Nakagawa, S.; Momma, T.; Naito, Y.; Niwa, M.; Motohashi, S.

1995-01-01

Forced vibration tests of a BWR-type reactor building. Hamaoka Unit 4, were performed. Valuable data on the dynamic characteristics of the soil-structure interaction system were obtained through the tests. Simulation analyses of the fundamental dynamic characteristics of the soil-structure system were conducted, using a basic lumped mass soil-structure model (lattice model), and strong correlation with the measured data was obtained. Furthermore, detailed simulation models were employed to investigate the effects of simultaneously induced vertical response and response of the adjacent turbine building on the lateral response of the reactor building. (author). 4 refs., 11 figs

Systematic simulation of a tubular recycle reactor on the basis of pilot plant experiments

Energy Technology Data Exchange (ETDEWEB)

Paar, H; Narodoslawsky, M; Moser, A [Technische Univ., Graz (Austria). Inst. fuer Biotechnologie, Mikrobiologie und Abfalltechnologie

1990-10-10

Systematic simulatiom may decisively help in development and optimization of bioprocesses. By applying simulation techniques, optimal use can be made of experimental data, decreasing development costs and increasing the accuracy in predicting the behavior of an industrial scale plant. The procedure of the dialogue between simulation and experimental efforts will be exemplified in a case study. Alcoholic fermentation of glucose by zymomonas mobilis bacteria in a gasified turbular recycle reactor was studied first by systematic simulation, using a computer model based solely on literature data. On the base of the results of this simulation, a 0.013 m{sup 3} pilot plant reactor was constructed. The pilot plant experiments, too, were based on the results of the systematic simulation. Simulated and experimental data were well in agreement. The pilot plant experiments reiterated the trends and limits of the process as shown by the simulation results. Data from the pilot plant runs were then used to improve the simulation model. This improved model was subsequently used to simulate the performances of an industrial scale plant. The results of this simulation are presented. They show that the alcohol fermentation in a tubular recycle reactor is potentially advantageous to other reactor configurations, especially to continuous stirred tanks. (orig.).

Numerical Simulation of Anisotropic Preheating Ablative Rayleigh–Taylor Instability