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Sample records for multi-physics reactor simulations

  1. Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Obabko, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Jain, Rajeev [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay [Argonne National Lab. (ANL), Argonne, IL (United States); Tautges, Timothy [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Ferencz, Robert Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Whitesides, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-12-21

    This report describes to employ SHARP to perform a first-of-a-kind analysis of the core radial expansion phenomenon in an SFR. This effort required significant advances in the framework Multi-Physics Demonstration Problem with the SHARP Reactor Simulation Toolkit used to drive the coupled simulations, manipulate the mesh in response to the deformation of the geometry, and generate the necessary modified mesh files. Furthermore, the model geometry is fairly complex, and consistent mesh generation for the three physics modules required significant effort. Fully-integrated simulations of a 7-assembly mini-core test problem have been performed, and the results are presented here. Physics models of a full-core model of the Advanced Burner Test Reactor have also been developed for each of the three physics modules. Standalone results of each of the three physics modules for the ABTR are presented here, which provides a demonstration of the feasibility of the fully-integrated simulation.

  2. Coupled multi-physics simulation frameworks for reactor simulation: A bottom-up approach

    International Nuclear Information System (INIS)

    Tautges, Timothy J.; Caceres, Alvaro; Jain, Rajeev; Kim, Hong-Jun; Kraftcheck, Jason A.; Smith, Brandon M.

    2011-01-01

    A 'bottom-up' approach to multi-physics frameworks is described, where first common interfaces to simulation data are developed, then existing physics modules are adapted to communicate through those interfaces. Physics modules read and write data through those common interfaces, which also provide access to common simulation services like parallel IO, mesh partitioning, etc.. Multi-physics codes are assembled as a combination of physics modules, services, interface implementations, and driver code which coordinates calling these various pieces. Examples of various physics modules and services connected to this framework are given. (author)

  3. 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)

  4. 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

  5. Nuclear reactor multi-physics simulations with coupled MCNP5 and STAR-CCM+

    International Nuclear Information System (INIS)

    Cardoni, Jeffrey Neil; Rizwan-uddin

    2011-01-01

    The MCNP5 Monte Carlo particle transport code has been coupled to the computational fluid dynamics code, STAR-CCM+, to provide a high fidelity multi-physics simulation tool for pressurized water nuclear reactors. The codes are executed separately and coupled externally through a Perl script. The Perl script automates the exchange of temperature, density, and volumetric heating information between the codes using ASCII text data files. Fortran90 and Java utility programs assist job automation with data post-processing and file management. The MCNP5 utility code, MAKXSF, pre-generates temperature dependent cross section libraries for the thermal feedback calculations. The MCNP5–STAR-CCM+ coupled simulation tool, dubbed MULTINUKE, was applied to a steady state, PWR cell model to demonstrate its usage and capabilities. The demonstration calculation showed reasonable results that agree with PWR values typically reported in literature. Temperature and fission reaction rate distributions were realistic and intuitive. Reactivity coefficients were also deemed reasonable in comparison to historically reported data. The demonstration problem consisted of 9,984 CFD cells and 7,489 neutronic cells. MCNP5 tallied fission energy deposition over 3,328 UO_2 cells. The coupled solution converged within eight hours and in three MULTINUKE iterations. The simulation was carried out on a 64 bit, quad core, Intel 2.8 GHz microprocessor with 1 GB RAM. The simulations on a quad core machine indicated that a massively parallelized implementation of MULTINUKE can be used to assess larger multi-million cell models. (author)

  6. Advanced multi-physics simulation capability for very high temperature reactors

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Tak, Nam Il; Jo Chang Keun; Noh, Jae Man; Cho, Bong Hyun; Cho, Jin Woung; Hong, Ser Gi

    2012-01-01

    The purpose of this research is to develop methodologies and computer code for high-fidelity multi-physics analysis of very high temperature gas-cooled reactors(VHTRs). The research project was performed through Korea-US I-NERI program. The main research topic was development of methodologies for high-fidelity 3-D whole core transport calculation, development of DeCART code for VHTR reactor physics analysis, generation of VHTR specific 190-group cross-section library for DeCART code, development of DeCART/CORONA coupled code system for neutronics/thermo-fluid multi-physics analysis, and benchmark analysis against various benchmark problems derived from PMR200 reactor. The methodologies and the code systems will be utilized a key technologies in the Nuclear Hydrogen Development and Demonstration program. Export of code system is expected in the near future and the code systems developed in this project are expected to contribute to development and export of nuclear hydrogen production system

  7. 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)

  8. A EU simulation platform for nuclear reactor safety: multi-scale and multi-physics calculations, sensitivity and uncertainty analysis (NURESIM project)

    International Nuclear Information System (INIS)

    Chauliac, Christian; Bestion, Dominique; Crouzet, Nicolas; Aragones, Jose-Maria; Cacuci, Dan Gabriel; Weiss, Frank-Peter; Zimmermann, Martin A.

    2010-01-01

    The NURESIM project, the numerical simulation platform, is developed in the frame of the NURISP European Collaborative Project (FP7), which includes 22 organizations from 14 European countries. NURESIM intends to be a reference platform providing high quality software tools, physical models, generic functions and assessment results. The NURESIM platform provides an accurate representation of the physical phenomena by promoting and incorporating the latest advances in core physics, two-phase thermal-hydraulics and fuel modelling. It includes multi-scale and multi-physics features, especially for coupling core physics and thermal-hydraulics models for reactor safety. Easy coupling of the different codes and solvers is provided through the use of a common data structure and generic functions (e.g., for interpolation between non-conforming meshes). More generally, the platform includes generic pre-processing, post-processing and supervision functions through the open-source SALOME software, in order to make the codes more user-friendly. The platform also provides the informatics environment for testing and comparing different codes. The contribution summarizes the achievements and ongoing developments of the simulation platform in core physics, thermal-hydraulics, multi-physics, uncertainties and code integration

  9. Multi-scale multi-physics computational chemistry simulation based on ultra-accelerated quantum chemical molecular dynamics method for structural materials in boiling water reactor

    International Nuclear Information System (INIS)

    Miyamoto, Akira; Sato, Etsuko; Sato, Ryo; Inaba, Kenji; Hatakeyama, Nozomu

    2014-01-01

    In collaboration with experimental experts we have reported in the present conference (Hatakeyama, N. et al., “Experiment-integrated multi-scale, multi-physics computational chemistry simulation applied to corrosion behaviour of BWR structural materials”) the results of multi-scale multi-physics computational chemistry simulations applied to the corrosion behaviour of BWR structural materials. In macro-scale, a macroscopic simulator of anode polarization curve was developed to solve the spatially one-dimensional electrochemical equations on the material surface in continuum level in order to understand the corrosion behaviour of typical BWR structural material, SUS304. The experimental anode polarization behaviours of each pure metal were reproduced by fitting all the rates of electrochemical reactions and then the anode polarization curve of SUS304 was calculated by using the same parameters and found to reproduce the experimental behaviour successfully. In meso-scale, a kinetic Monte Carlo (KMC) simulator was applied to an actual-time simulation of the morphological corrosion behaviour under the influence of an applied voltage. In micro-scale, an ultra-accelerated quantum chemical molecular dynamics (UA-QCMD) code was applied to various metallic oxide surfaces of Fe 2 O 3 , Fe 3 O 4 , Cr 2 O 3 modelled as same as water molecules and dissolved metallic ions on the surfaces, then the dissolution and segregation behaviours were successfully simulated dynamically by using UA-QCMD. In this paper we describe details of the multi-scale, multi-physics computational chemistry method especially the UA-QCMD method. This method is approximately 10,000,000 times faster than conventional first-principles molecular dynamics methods based on density-functional theory (DFT), and the accuracy was also validated for various metals and metal oxides compared with DFT results. To assure multi-scale multi-physics computational chemistry simulation based on the UA-QCMD method for

  10. 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)

  11. Multi-Physics Simulation of TREAT Kinetics using MAMMOTH

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark; Gleicher, Frederick; Ortensi, Javier; Alberti, Anthony; Palmer, Todd

    2015-11-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific fuels transient tests range from simple temperature transients to full fuel melt accidents. The current TREAT core is driven by highly enriched uranium (HEU) dispersed in a graphite matrix (1:10000 U-235/C atom ratio). At the center of the core, fuel is removed allowing for the insertion of an experimental test vehicle. TREAT’s design provides experimental flexibility and inherent safety during neutron pulsing. This safety stems from the graphite in the driver fuel having a strong negative temperature coefficient of reactivity resulting from a thermal Maxwellian shift with increased leakage, as well as graphite acting as a temperature sink. Air cooling is available, but is generally used post-transient for heat removal. DOE and INL have expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility, with an emphasis on effective and safe operation while minimizing experimental time and cost. At INL, the Multi-physics Object Oriented Simulation Environment (MOOSE) has been selected as the model development framework for this work. This paper describes the results of preliminary simulations of a TREAT fuel element under transient conditions using the MOOSE-based MAMMOTH reactor physics tool.

  12. Specification of the Advanced Burner Test Reactor Multi-Physics Coupling Demonstration Problem

    Energy Technology Data Exchange (ETDEWEB)

    Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Lee, C. H. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-12-21

    This document specifies the multi-physics nuclear reactor demonstration problem using the SHARP software package developed by NEAMS. The SHARP toolset simulates the key coupled physics phenomena inside a nuclear reactor. The PROTEUS neutronics code models the neutron transport within the system, the Nek5000 computational fluid dynamics code models the fluid flow and heat transfer, and the DIABLO structural mechanics code models structural and mechanical deformation. The three codes are coupled to the MOAB mesh framework which allows feedback from neutronics, fluid mechanics, and mechanical deformation in a compatible format.

  13. IMPETUS - Interactive MultiPhysics Environment for Unified Simulations.

    Science.gov (United States)

    Ha, Vi Q; Lykotrafitis, George

    2016-12-08

    We introduce IMPETUS - Interactive MultiPhysics Environment for Unified Simulations, an object oriented, easy-to-use, high performance, C++ program for three-dimensional simulations of complex physical systems that can benefit a large variety of research areas, especially in cell mechanics. The program implements cross-communication between locally interacting particles and continuum models residing in the same physical space while a network facilitates long-range particle interactions. Message Passing Interface is used for inter-processor communication for all simulations. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Advanced Mesh-Enabled Monte carlo capability for Multi-Physics Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, Paul; Evans, Thomas; Tautges, Tim

    2012-12-24

    This project will accumulate high-precision fluxes throughout reactor geometry on a non- orthogonal grid of cells to support multi-physics coupling, in order to more accurately calculate parameters such as reactivity coefficients and to generate multi-group cross sections. This work will be based upon recent developments to incorporate advanced geometry and mesh capability in a modular Monte Carlo toolkit with computational science technology that is in use in related reactor simulation software development. Coupling this capability with production-scale Monte Carlo radiation transport codes can provide advanced and extensible test-beds for these developments. Continuous energy Monte Carlo methods are generally considered to be the most accurate computational tool for simulating radiation transport in complex geometries, particularly neutron transport in reactors. Nevertheless, there are several limitations for their use in reactor analysis. Most significantly, there is a trade-off between the fidelity of results in phase space, statistical accuracy, and the amount of computer time required for simulation. Consequently, to achieve an acceptable level of statistical convergence in high-fidelity results required for modern coupled multi-physics analysis, the required computer time makes Monte Carlo methods prohibitive for design iterations and detailed whole-core analysis. More subtly, the statistical uncertainty is typically not uniform throughout the domain, and the simulation quality is limited by the regions with the largest statistical uncertainty. In addition, the formulation of neutron scattering laws in continuous energy Monte Carlo methods makes it difficult to calculate adjoint neutron fluxes required to properly determine important reactivity parameters. Finally, most Monte Carlo codes available for reactor analysis have relied on orthogonal hexahedral grids for tallies that do not conform to the geometric boundaries and are thus generally not well

  15. Development of a three dimension multi-physics code for molten salt fast reactor

    International Nuclear Information System (INIS)

    Cheng Maosong; Dai Zhimin

    2014-01-01

    Molten Salt Reactor (MSR) was selected as one of the six innovative nuclear reactors by the Generation IV International Forum (GIF). The circulating-fuel in the can-type molten salt fast reactor makes the neutronics and thermo-hydraulics of the reactor strongly coupled and different from that of traditional solid-fuel reactors. In the present paper: a new coupling model is presented that physically describes the inherent relations between the neutron flux, the delayed neutron precursor, the heat transfer and the turbulent flow. Based on the model, integrating nuclear data processing, CAD modeling, structured and unstructured mesh technology, data analysis and visualization application, a three dimension steady state simulation code system (MSR3DS) for the can-type molten salt fast reactor is developed and validated. In order to demonstrate the ability of the code, the three dimension distributions of the velocity, the neutron flux, the delayed neutron precursor and the temperature were obtained for the simplified MOlten Salt Advanced Reactor Transmuter (MOSART) using this code. The results indicate that the MSR3DS code can provide a feasible description of multi-physical coupling phenomena in can-type molten salt fast reactor. Furthermore, the code can well predict the flow effect of fuel salt and the transport effect of the turbulent diffusion. (authors)

  16. Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

    International Nuclear Information System (INIS)

    Ivanov, K.; Avramova, M.

    2007-01-01

    Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

  17. Progress and challenges in the development and qualification of multi-level multi-physics coupled methodologies for reactor analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, K.; Avramova, M. [Pennsylvania State Univ., University Park, PA (United States)

    2007-07-01

    Current trends in nuclear power generation and regulation as well as the design of next generation reactor concepts along with the continuing computer technology progress stimulate the development, qualification and application of multi-physics multi-scale coupled code systems. The efforts have been focused on extending the analysis capabilities by coupling models, which simulate different phenomena or system components, as well as on refining the scale and level of detail of the coupling. This paper reviews the progress made in this area and outlines the remaining challenges. The discussion is illustrated with examples based on neutronics/thermohydraulics coupling in the reactor core modeling. In both fields recent advances and developments are towards more physics-based high-fidelity simulations, which require implementation of improved and flexible coupling methodologies. First, the progresses in coupling of different physics codes along with the advances in multi-level techniques for coupled code simulations are discussed. Second, the issues related to the consistent qualification of coupled multi-physics and multi-scale code systems for design and safety evaluation are presented. The increased importance of uncertainty and sensitivity analysis are discussed along with approaches to propagate the uncertainty quantification between the codes. The incoming OECD LWR Uncertainty Analysis in Modeling (UAM) benchmark is the first international activity to address this issue and it is described in the paper. Finally, the remaining challenges with multi-physics coupling are outlined. (authors)

  18. Sensitivity and Uncertainty Analysis of Coupled Reactor Physics Problems : Method Development for Multi-Physics in Reactors

    NARCIS (Netherlands)

    Perkó, Z.

    2015-01-01

    This thesis presents novel adjoint and spectral methods for the sensitivity and uncertainty (S&U) analysis of multi-physics problems encountered in the field of reactor physics. The first part focuses on the steady state of reactors and extends the adjoint sensitivity analysis methods well

  19. Review of multi-physics temporal coupling methods for analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Zerkak, Omar; Kozlowski, Tomasz; Gajev, Ivan

    2015-01-01

    Highlights: • Review of the numerical methods used for the multi-physics temporal coupling. • Review of high-order improvements to the Operator Splitting coupling method. • Analysis of truncation error due to the temporal coupling. • Recommendations on best-practice approaches for multi-physics temporal coupling. - Abstract: The advanced numerical simulation of a realistic physical system typically involves multi-physics problem. For example, analysis of a LWR core involves the intricate simulation of neutron production and transport, heat transfer throughout the structures of the system and the flowing, possibly two-phase, coolant. Such analysis involves the dynamic coupling of multiple simulation codes, each one devoted to the solving of one of the coupled physics. Multiple temporal coupling methods exist, yet the accuracy of such coupling is generally driven by the least accurate numerical scheme. The goal of this paper is to review in detail the approaches and numerical methods that can be used for the multi-physics temporal coupling, including a comprehensive discussion of the issues associated with the temporal coupling, and define approaches that can be used to perform multi-physics analysis. The paper is not limited to any particular multi-physics process or situation, but is intended to provide a generic description of multi-physics temporal coupling schemes for any development stage of the individual (single-physics) tools and methods. This includes a wide spectrum of situation, where the individual (single-physics) solvers are based on pre-existing computation codes embedded as individual components, or a new development where the temporal coupling can be developed and implemented as a part of code development. The discussed coupling methods are demonstrated in the framework of LWR core analysis

  20. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    International Nuclear Information System (INIS)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W.

    2016-01-01

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  1. KIT multi-physics tools for the analysis of design and beyond design basis accidents of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez, Victor Hugo; Miassoedov, Alexei; Steinbrueck, M.; Tromm, W. [Karlsruhe Institute of Technology (KIT), Eggenstein-Leopoldshafen (Germany)

    2016-05-15

    This paper describes the KIT numerical simulation tools under extension and validation for the analysis of design and beyond design basis accidents (DBA) of Light Water Reactors (LWR). The description of the complex thermal hydraulic, neutron kinetics and chemo-physical phenomena going on during off-normal conditions requires the development of multi-physics and multi-scale simulations tools which are fostered by the rapid increase in computer power nowadays. The KIT numerical tools for DBA and beyond DBA are validated using experimental data of KIT or from abroad. The developments, extensions, coupling approaches and validation work performed at KIT are shortly outlined and discussed in this paper.

  2. Multi-physics design and analyses of long life reactors for lunar outposts

    Science.gov (United States)

    Schriener, Timothy M.

    event of a launch abort accident. Increasing the amount of fuel in the reactor core, and hence its operational life, would be possible by launching the reactor unfueled and fueling it on the Moon. Such a reactor would, thus, not be subject to launch criticality safety requirements. However, loading the reactor with fuel on the Moon presents a challenge, requiring special designs of the core and the fuel elements, which lend themselves to fueling on the lunar surface. This research investigates examples of both a solid core reactor that would be fueled at launch as well as an advanced concept which could be fueled on the Moon. Increasing the operational life of a reactor fueled at launch is exercised for the NaK-78 cooled Sectored Compact Reactor (SCoRe). A multi-physics design and analyses methodology is developed which iteratively couples together detailed Monte Carlo neutronics simulations with 3-D Computational Fluid Dynamics (CFD) and thermal-hydraulics analyses. Using this methodology the operational life of this compact, fast spectrum reactor is increased by reconfiguring the core geometry to reduce neutron leakage and parasitic absorption, for the same amount of HEU in the core, and meeting launch safety requirements. The multi-physics analyses determine the impacts of the various design changes on the reactor's neutronics and thermal-hydraulics performance. The option of increasing the operational life of a reactor by loading it on the Moon is exercised for the Pellet Bed Reactor (PeBR). The PeBR uses spherical fuel pellets and is cooled by He-Xe gas, allowing the reactor core to be loaded with fuel pellets and charged with working fluid on the lunar surface. The performed neutronics analyses ensure the PeBR design achieves a long operational life, and develops safe launch canister designs to transport the spherical fuel pellets to the lunar surface. The research also investigates loading the PeBR core with fuel pellets on the Moon using a transient Discrete

  3. A multi-physics analysis for the actuation of the SSS in opal reactor

    Directory of Open Access Journals (Sweden)

    Ferraro Diego

    2018-01-01

    models, available for OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

  4. A multi-physics analysis for the actuation of the SSS in opal reactor

    Science.gov (United States)

    Ferraro, Diego; Alberto, Patricio; Villarino, Eduardo; Doval, Alicia

    2018-05-01

    OPAL reactor, are coupled by means of a shared unstructured mesh geometry definition of relevant zones inside the Reflector Vessel. Several scenarios, both regarding coupled and uncoupled neutronic & thermohydraulic behavior, are presented and analyzed, showing the capabilities to develop and manage advanced modelling that allows to predict multi-physics variables observed when an in-depth performance analysis of a Research Reactor like OPAL is carried out.

  5. Three-dimensional multi-physics model of the European sodium fast reactor design applied to DBA analysis - 15293

    International Nuclear Information System (INIS)

    Lazaro, A.; Ordonez, J.; Martorell, S.; Przemyslaw, S.; Ammirabile, L.; Tsige-Tamirat, H.

    2015-01-01

    The sodium cooled fast reactor (SFR) is one of the reactor types selected by the Generation IV International Forum. SFR stand out due to its remarkable past operational experience in related projects and its potential to achieve the ambitious goals laid for the new generation of nuclear reactors. Regardless its operational experience, there is a need to apply computational tools able to simulate the system behaviour under conditions that may overtake the reactor safety limits from the early stages of the design process, including the three-dimensional phenomena that may arise in these transients. This paper presents the different steps followed towards the development of a multi-physics platform with capabilities to simulate complex phenomena using a coupled neutronic-thermal-hydraulic scheme. The development started with a one-dimensional thermal-hydraulic model of the European Sodium Fast Reactor (ESFR) design with point kinetic neutronic feedback benchmarked with its peers in the framework of the FP7-CP-ESFR project using the state-of-the-art thermal-hydraulic system code TRACE. The model was successively extended into a three-dimensional model coupled with the spatial kinetic neutronic code PARCS able to simulate three-dimensional multi-physic phenomena along with the comparison of the results for symmetric cases. The last part of the paper shows the application of the developed tool to the analysis of transients involving asymmetrical effects, such as the coast-down of a primary and secondary pump or the withdrawal of a peripheral control rod bank, demonstrating the unique capability of the code to simulate such transients and the capability of the design to withstand them under design basis

  6. The application of a multi-physics tool kit to spatial reactor dynamics

    International Nuclear Information System (INIS)

    Clifford, I.; Jasak, H.

    2009-01-01

    Traditionally coupled field nuclear reactor analysis has been carried out using several loosely coupled solvers, each having been developed independently from the others. In the field of multi-physics, the current generation of object-oriented tool kits provides robust close coupling of multiple fields on a single framework. This paper describes the initial results obtained as part of continuing research in the use of the OpenFOAM multi-physics tool kit for reactor dynamics application development. An unstructured, three-dimensional, time-dependent multi-group diffusion code Diffusion FOAM has been developed using the OpenFOAM multi-physics tool kit as a basis. The code is based on the finite-volume methodology and uses a newly developed block-coupled sparse matrix solver for the coupled solution of the multi-group diffusion equations. A description of this code is given with particular emphasis on the newly developed block-coupled solver, along with a selection of results obtained thus far. The code has performed well, indicating that the OpenFOAM tool kit is suited to reactor dynamics applications. This work has shown that the neutronics and simplified thermal-hydraulics of a reactor May be represented and solved for using a common calculation platform, and opens up the possibility for research into robust close-coupling of neutron diffusion and thermal-fluid calculations. This work has further opened up the possibility for research in a number of other areas, including research into three-dimensional unstructured meshes for reactor dynamics applications. (authors)

  7. Predictive modeling of coupled multi-physics systems: II. Illustrative application to reactor physics

    International Nuclear Information System (INIS)

    Cacuci, Dan Gabriel; Badea, Madalina Corina

    2014-01-01

    Highlights: • We applied the PMCMPS methodology to a paradigm neutron diffusion model. • We underscore the main steps in applying PMCMPS to treat very large coupled systems. • PMCMPS reduces the uncertainties in the optimally predicted responses and model parameters. • PMCMPS is for sequentially treating coupled systems that cannot be treated simultaneously. - Abstract: This work presents paradigm applications to reactor physics of the innovative mathematical methodology for “predictive modeling of coupled multi-physics systems (PMCMPS)” developed by Cacuci (2014). This methodology enables the assimilation of experimental and computational information and computes optimally predicted responses and model parameters with reduced predicted uncertainties, taking fully into account the coupling terms between the multi-physics systems, but using only the computational resources that would be needed to perform predictive modeling on each system separately. The paradigm examples presented in this work are based on a simple neutron diffusion model, chosen so as to enable closed-form solutions with clear physical interpretations. These paradigm examples also illustrate the computational efficiency of the PMCMPS, which enables the assimilation of additional experimental information, with a minimal increase in computational resources, to reduce the uncertainties in predicted responses and best-estimate values for uncertain model parameters, thus illustrating how very large systems can be treated without loss of information in a sequential rather than simultaneous manner

  8. Advanced graphical user interface for multi-physics simulations using AMST

    Science.gov (United States)

    Hoffmann, Florian; Vogel, Frank

    2017-07-01

    Numerical modelling of particulate matter has gained much popularity in recent decades. Advanced Multi-physics Simulation Technology (AMST) is a state-of-the-art three dimensional numerical modelling technique combining the eX-tended Discrete Element Method (XDEM) with Computational Fluid Dynamics (CFD) and Finite Element Analysis (FEA) [1]. One major limitation of this code is the lack of a graphical user interface (GUI) meaning that all pre-processing has to be made directly in a HDF5-file. This contribution presents the first graphical pre-processor developed for AMST.

  9. Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Jun [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Buechler, Cynthia Eileen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-07-17

    The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operating scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi-physics

  10. Heat transfer in solids using infrared photothermal radiometry and simulation by Com sol multi physics

    International Nuclear Information System (INIS)

    Suarez, V.; Hernandez W, J.; Calderon, A.; Rojas T, J. B.; Juarez, A. G.; Marin, E.; Castaneda, A.

    2012-10-01

    We investigate the heat transfer through a homogeneous and isotropic solid exited by periodic light beam on its front surface. For this, we use the infrared photothermal radiometry in order to obtain the evolution of the temperature difference on the rear surface of the silicon sample as a function of the exposure time. Also, we solved the heat conduction equation for this problem with the boundary conditions congruent with the physical situation, by means of application the Com sol multi physics software and the heat transfer module. Our results show a good agree between the experimental and simulated results, which demonstrate the utility of this methodology in the study of the thermal response in solids. (Author)

  11. The Cea multi-scale and multi-physics simulation project for nuclear applications

    International Nuclear Information System (INIS)

    Ledermann, P.; Chauliac, C.; Thomas, J.B.

    2005-01-01

    Full text of publication follows. Today numerical modelling is everywhere recognized as an essential tool of capitalization, integration and share of knowledge. For this reason, it becomes the central tool of research. Until now, the Cea developed a set of scientific software allowing to model, in each situation, the operation of whole or part of a nuclear installation and these codes are largely used in nuclear industry. However, for the future, it is essential to aim for a better accuracy, a better control of uncertainties and better performance in computing times. The objective is to obtain validated models allowing accurate predictive calculations for actual complex nuclear problems such as fuel behaviour in accidental situation. This demands to master a large and interactive set of phenomena ranging from nuclear reaction to heat transfer. To this end, Cea, with industrial partners (EDF, Framatome-ANP, ANDRA) has designed an integrated platform of calculation, devoted to the study of nuclear systems, and intended at the same time for industries and scientists. The development of this platform is under way with the start in 2005 of the integrated project NURESIM, with 18 European partners. Improvement is coming not only through a multi-scale description of all phenomena but also through an innovative design approach requiring deep functional analysis which is upstream from the development of the simulation platform itself. In addition, the studies of future nuclear systems are increasingly multidisciplinary (simultaneous modelling of core physics, thermal-hydraulics and fuel behaviour). These multi-physics and multi-scale aspects make mandatory to pay very careful attention to software architecture issues. A global platform is thus developed integrating dedicated specialized platforms: DESCARTES for core physics, NEPTUNE for thermal-hydraulics, PLEIADES for fuel behaviour, SINERGY for materials behaviour under irradiation, ALLIANCES for the performance

  12. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  13. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  14. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany); Schuetze, Jochen [ANSYS Germany GmbH, Darmstadt (Germany); Frank, Thomas [ANSYS Germany GmbH, Otterfing (Germany); Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo [Karlsruher Institut fuer Technologie (KIT), Eggenstein-Leopoldshafen (Germany)

    2011-07-15

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  15. Development of multi-physics code systems based on the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Kliem, Soeren; Gommlich, Andre; Grahn, Alexander; Rohde, Ulrich; Schuetze, Jochen; Frank, Thomas; Gomez Torres, Armando M.; Sanchez Espinoza, Victor Hugo

    2011-01-01

    The reactor dynamics code DYN3D has been coupled with the CFD code ANSYS CFX and the 3D thermal hydraulic core model FLICA4. In the coupling with ANSYS CFX, DYN3D calculates the neutron kinetics and the fuel behavior including the heat transfer to the coolant. The physical data interface between the codes is the volumetric heat release rate into the coolant. In the coupling with FLICA4 only the neutron kinetics module of DYN3D is used. Fluid dynamics and related transport phenomena in the reactor's coolant and fuel behavior is calculated by FLICA4. The correctness of the coupling of DYN3D with both thermal hydraulic codes was verified by the calculation of different test problems. These test problems were set-up in such a way that comparison with the DYN3D stand-alone code was possible. This included steady-state and transient calculations of a mini-core consisting of nine real-size PWR fuel assemblies with ANSYS CFX/DYN3D as well as mini-core and a full core steady-state calculation using FLICA4/DYN3D. (orig.)

  16. Computational simulation of biomolecules transport with multi-physics near microchannel surface for development of biomolecules-detection devices.

    Science.gov (United States)

    Suzuki, Yuma; Shimizu, Tetsuhide; Yang, Ming

    2017-01-01

    The quantitative evaluation of the biomolecules transport with multi-physics in nano/micro scale is demanded in order to optimize the design of microfluidics device for the biomolecules detection with high detection sensitivity and rapid diagnosis. This paper aimed to investigate the effectivity of the computational simulation using the numerical model of the biomolecules transport with multi-physics near a microchannel surface on the development of biomolecules-detection devices. The biomolecules transport with fluid drag force, electric double layer (EDL) force, and van der Waals force was modeled by Newtonian Equation of motion. The model validity was verified in the influence of ion strength and flow velocity on biomolecules distribution near the surface compared with experimental results of previous studies. The influence of acting forces on its distribution near the surface was investigated by the simulation. The trend of its distribution to ion strength and flow velocity was agreement with the experimental result by the combination of all acting forces. Furthermore, EDL force dominantly influenced its distribution near its surface compared with fluid drag force except for the case of high velocity and low ion strength. The knowledges from the simulation might be useful for the design of biomolecules-detection devices and the simulation can be expected to be applied on its development as the design tool for high detection sensitivity and rapid diagnosis in the future.

  17. A capacitive CMOS-MEMS sensor designed by multi-physics simulation for integrated CMOS-MEMS technology

    Science.gov (United States)

    Konishi, Toshifumi; Yamane, Daisuke; Matsushima, Takaaki; Masu, Kazuya; Machida, Katsuyuki; Toshiyoshi, Hiroshi

    2014-01-01

    This paper reports the design and evaluation results of a capacitive CMOS-MEMS sensor that consists of the proposed sensor circuit and a capacitive MEMS device implemented on the circuit. To design a capacitive CMOS-MEMS sensor, a multi-physics simulation of the electromechanical behavior of both the MEMS structure and the sensing LSI was carried out simultaneously. In order to verify the validity of the design, we applied the capacitive CMOS-MEMS sensor to a MEMS accelerometer implemented by the post-CMOS process onto a 0.35-µm CMOS circuit. The experimental results of the CMOS-MEMS accelerometer exhibited good agreement with the simulation results within the input acceleration range between 0.5 and 6 G (1 G = 9.8 m/s2), corresponding to the output voltages between 908.6 and 915.4 mV, respectively. Therefore, we have confirmed that our capacitive CMOS-MEMS sensor and the multi-physics simulation will be beneficial method to realize integrated CMOS-MEMS technology.

  18. Multi-Physics Modelling of Fault Mechanics Using REDBACK: A Parallel Open-Source Simulator for Tightly Coupled Problems

    Science.gov (United States)

    Poulet, Thomas; Paesold, Martin; Veveakis, Manolis

    2017-03-01

    Faults play a major role in many economically and environmentally important geological systems, ranging from impermeable seals in petroleum reservoirs to fluid pathways in ore-forming hydrothermal systems. Their behavior is therefore widely studied and fault mechanics is particularly focused on the mechanisms explaining their transient evolution. Single faults can change in time from seals to open channels as they become seismically active and various models have recently been presented to explain the driving forces responsible for such transitions. A model of particular interest is the multi-physics oscillator of Alevizos et al. (J Geophys Res Solid Earth 119(6), 4558-4582, 2014) which extends the traditional rate and state friction approach to rate and temperature-dependent ductile rocks, and has been successfully applied to explain spatial features of exposed thrusts as well as temporal evolutions of current subduction zones. In this contribution we implement that model in REDBACK, a parallel open-source multi-physics simulator developed to solve such geological instabilities in three dimensions. The resolution of the underlying system of equations in a tightly coupled manner allows REDBACK to capture appropriately the various theoretical regimes of the system, including the periodic and non-periodic instabilities. REDBACK can then be used to simulate the drastic permeability evolution in time of such systems, where nominally impermeable faults can sporadically become fluid pathways, with permeability increases of several orders of magnitude.

  19. Multi-scale and multi-physics simulations using the multi-fluid plasma model

    Science.gov (United States)

    2017-04-25

    small The simulation uses 512 second-order elements Bz = 1.0, Te = Ti = 0.01, ui = ue = 0 ne = ni = 1.0 + e−10(x−6) 2 Baboolal, Math . and Comp. Sim. 55...DISTRIBUTION Clearance No. 17211 23 / 31 SUMMARY The blended finite element method (BFEM) is presented DG spatial discretization with explicit Runge...Kutta (i+, n) CG spatial discretization with implicit Crank-Nicolson (e−, fileds) DG captures shocks and discontinuities CG is efficient and robust for

  20. Modeling and simulation of multi-physics multi-scale transport phenomenain bio-medical applications

    International Nuclear Information System (INIS)

    Kenjereš, Saša

    2014-01-01

    We present a short overview of some of our most recent work that combines the mathematical modeling, advanced computer simulations and state-of-the-art experimental techniques of physical transport phenomena in various bio-medical applications. In the first example, we tackle predictions of complex blood flow patterns in the patient-specific vascular system (carotid artery bifurcation) and transfer of the so-called 'bad' cholesterol (low-density lipoprotein, LDL) within the multi-layered artery wall. This two-way coupling between the blood flow and corresponding mass transfer of LDL within the artery wall is essential for predictions of regions where atherosclerosis can develop. It is demonstrated that a recently developed mathematical model, which takes into account the complex multi-layer arterial-wall structure, produced LDL profiles within the artery wall in good agreement with in-vivo experiments in rabbits, and it can be used for predictions of locations where the initial stage of development of atherosclerosis may take place. The second example includes a combination of pulsating blood flow and medical drug delivery and deposition controlled by external magnetic field gradients in the patient specific carotid artery bifurcation. The results of numerical simulations are compared with own PIV (Particle Image Velocimetry) and MRI (Magnetic Resonance Imaging) in the PDMS (silicon-based organic polymer) phantom. A very good agreement between simulations and experiments is obtained for different stages of the pulsating cycle. Application of the magnetic drug targeting resulted in an increase of up to ten fold in the efficiency of local deposition of the medical drug at desired locations. Finally, the LES (Large Eddy Simulation) of the aerosol distribution within the human respiratory system that includes up to eight bronchial generations is performed. A very good agreement between simulations and MRV (Magnetic Resonance Velocimetry) measurements is

  1. Modeling and simulation of multi-physics multi-scale transport phenomenain bio-medical applications

    Science.gov (United States)

    Kenjereš, Saša

    2014-08-01

    We present a short overview of some of our most recent work that combines the mathematical modeling, advanced computer simulations and state-of-the-art experimental techniques of physical transport phenomena in various bio-medical applications. In the first example, we tackle predictions of complex blood flow patterns in the patient-specific vascular system (carotid artery bifurcation) and transfer of the so-called "bad" cholesterol (low-density lipoprotein, LDL) within the multi-layered artery wall. This two-way coupling between the blood flow and corresponding mass transfer of LDL within the artery wall is essential for predictions of regions where atherosclerosis can develop. It is demonstrated that a recently developed mathematical model, which takes into account the complex multi-layer arterial-wall structure, produced LDL profiles within the artery wall in good agreement with in-vivo experiments in rabbits, and it can be used for predictions of locations where the initial stage of development of atherosclerosis may take place. The second example includes a combination of pulsating blood flow and medical drug delivery and deposition controlled by external magnetic field gradients in the patient specific carotid artery bifurcation. The results of numerical simulations are compared with own PIV (Particle Image Velocimetry) and MRI (Magnetic Resonance Imaging) in the PDMS (silicon-based organic polymer) phantom. A very good agreement between simulations and experiments is obtained for different stages of the pulsating cycle. Application of the magnetic drug targeting resulted in an increase of up to ten fold in the efficiency of local deposition of the medical drug at desired locations. Finally, the LES (Large Eddy Simulation) of the aerosol distribution within the human respiratory system that includes up to eight bronchial generations is performed. A very good agreement between simulations and MRV (Magnetic Resonance Velocimetry) measurements is obtained

  2. Application of the new GeN-Foam multi-physics solver to the European sodium fast reactor and verification against available codes - 15226

    International Nuclear Information System (INIS)

    Fiorina, C.; Mikityuk, K.

    2015-01-01

    A new multi-physics solver for nuclear reactor analysis, named GeN-Foam (Generalized Nuclear Foam), has been developed by the FAST group at the Paul Scherrer Institut. It is based on OpenFOAM and has been developed for the multi-physics transient analyses of pin-based (e.g., liquid metal Fast Reactors, Light Water Reactors) or homogeneous (e.g., fast spectrum Molten Salt Reactors) nuclear reactors. It includes solutions of coarse or fine mesh thermal-hydraulics, thermal-mechanics and neutron diffusion. In particular, thermal-hydraulics solution can combine on the same mesh both a traditional RANS model and a porous medium model, depending on the desired degree of approximation for each region. In case the active reactor core is modeled as a porous medium, a simple sub-solver computes the sub-scale radial temperature profiles in fuel and cladding. The mesh used for neutronics calculations is deformed according to the displacement field predicted by the thermal-mechanics solver, thus allowing for a direct prediction of expansion-related feedback effects in Fast Reactors. To limit computational requirements, GeN-Foam permits the use of three different unstructured meshes for thermal-hydraulics, thermal-mechanics and neutron diffusion. For the same reason, an adaptive time step is employed. The different equations can be solved altogether or selectively included. In this work, GeN-Foam is applied to the analysis of the European Sodium Fast Reactor (ESFR). In particular, a 3-D model of the ESFR core is set up employing a coarse-mesh porous-medium approach for the thermal-hydraulics. The reactor steady-state and different accidental transients are investigated to offer an overview of GeN-Foam use and capabilities, as well as to preliminarily investigate the impact of a relatively accurate thermal-mechanic treatment on the predicted ESFR behavior. A code-to-code benchmark against the TRACE system code is performed to verify the adequacy of the results provided by the new

  3. Methodology Development for Passive Component Reliability Modeling in a Multi-Physics Simulation Environment

    Energy Technology Data Exchange (ETDEWEB)

    Aldemir, Tunc [The Ohio State Univ., Columbus, OH (United States); Denning, Richard [The Ohio State Univ., Columbus, OH (United States); Catalyurek, Umit [The Ohio State Univ., Columbus, OH (United States); Unwin, Stephen [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2015-01-23

    Reduction in safety margin can be expected as passive structures and components undergo degradation with time. Limitations in the traditional probabilistic risk assessment (PRA) methodology constrain its value as an effective tool to address the impact of aging effects on risk and for quantifying the impact of aging management strategies in maintaining safety margins. A methodology has been developed to address multiple aging mechanisms involving large numbers of components (with possibly statistically dependent failures) within the PRA framework in a computationally feasible manner when the sequencing of events is conditioned on the physical conditions predicted in a simulation environment, such as the New Generation System Code (NGSC) concept. Both epistemic and aleatory uncertainties can be accounted for within the same phenomenological framework and maintenance can be accounted for in a coherent fashion. The framework accommodates the prospective impacts of various intervention strategies such as testing, maintenance, and refurbishment. The methodology is illustrated with several examples.

  4. Methodology Development for Passive Component Reliability Modeling in a Multi-Physics Simulation Environment

    International Nuclear Information System (INIS)

    Aldemir, Tunc; Denning, Richard; Catalyurek, Umit; Unwin, Stephen

    2015-01-01

    Reduction in safety margin can be expected as passive structures and components undergo degradation with time. Limitations in the traditional probabilistic risk assessment (PRA) methodology constrain its value as an effective tool to address the impact of aging effects on risk and for quantifying the impact of aging management strategies in maintaining safety margins. A methodology has been developed to address multiple aging mechanisms involving large numbers of components (with possibly statistically dependent failures) within the PRA framework in a computationally feasible manner when the sequencing of events is conditioned on the physical conditions predicted in a simulation environment, such as the New Generation System Code (NGSC) concept. Both epistemic and aleatory uncertainties can be accounted for within the same phenomenological framework and maintenance can be accounted for in a coherent fashion. The framework accommodates the prospective impacts of various intervention strategies such as testing, maintenance, and refurbishment. The methodology is illustrated with several examples.

  5. Multi-physics modeling and numerical simulation of weld pool in GTA welding

    International Nuclear Information System (INIS)

    Nguyen, Minh-Chien

    2015-01-01

    In this work, we develop a 3D physical and numerical model of the GTA (Gas Tungsten Arc) welding process in order to predict, for given welding parameters, useful quantities for the designer of welded assembly: weld bead shape, fluid flow in the weld pool as well as thermal distribution in the work piece. The model is developed in the Cast3M (http://www-cast3m.cea.fr/) finite element software and takes into account the main physical phenomena acting in the work piece and particularly in the weld pool, subject to source terms modeling the arc part of the welding process. A steady solution of this model is thought for and involves the coupling of the nonlinear thermohydraulics and electromagnetic equations together with the displacement of the deformable free surface of the weld pool. A first step in the development consisted in modeling the electromagnetic phenomena with two different numerical methods, in comparing the numerical results obtained with those of the literature and in quantifying the importance of the Lorentz force and the Joule effect compared to the other mechanical and thermal sources by computing power balances. Then, in order to assess the predictive capability of the model, simulations of various welding configurations are performed: variation in the chemical composition of the material, of the welding speed, of the prescribed arc pressure and of the welding positions, which is a focus of this work, are studied. A good agreement is obtained between the results of our model and other experimental and numerical results of the literature. Eventually, a model accounting for metal filling is proposed and its results are discussed. Thus, our complete model can be seen as a solid foundation towards future totally-coupled 3D welding models including the arc and it will be included in WPROCESS the in-house CEA software dedicated to the numerical simulation of welding. (author) [fr

  6. Extreme scale multi-physics simulations of the tsunamigenic 2004 Sumatra megathrust earthquake

    Science.gov (United States)

    Ulrich, T.; Gabriel, A. A.; Madden, E. H.; Wollherr, S.; Uphoff, C.; Rettenberger, S.; Bader, M.

    2017-12-01

    SeisSol (www.seissol.org) is an open-source software package based on an arbitrary high-order derivative Discontinuous Galerkin method (ADER-DG). It solves spontaneous dynamic rupture propagation on pre-existing fault interfaces according to non-linear friction laws, coupled to seismic wave propagation with high-order accuracy in space and time (minimal dispersion errors). SeisSol exploits unstructured meshes to account for complex geometries, e.g. high resolution topography and bathymetry, 3D subsurface structure, and fault networks. We present the up-to-date largest (1500 km of faults) and longest (500 s) dynamic rupture simulation modeling the 2004 Sumatra-Andaman earthquake. We demonstrate the need for end-to-end-optimization and petascale performance of scientific software to realize realistic simulations on the extreme scales of subduction zone earthquakes: Considering the full complexity of subduction zone geometries leads inevitably to huge differences in element sizes. The main code improvements include a cache-aware wave propagation scheme and optimizations of the dynamic rupture kernels using code generation. In addition, a novel clustered local-time-stepping scheme for dynamic rupture has been established. Finally, asynchronous output has been implemented to overlap I/O and compute time. We resolve the frictional sliding process on the curved mega-thrust and a system of splay faults, as well as the seismic wave field and seafloor displacement with frequency content up to 2.2 Hz. We validate the scenario by geodetic, seismological and tsunami observations. The resulting rupture dynamics shed new light on the activation and importance of splay faults.

  7. Three-dimensional multi-physics coupled simulation of ignition transient in a dual pulse solid rocket motor

    Science.gov (United States)

    Li, Yingkun; Chen, Xiong; Xu, Jinsheng; Zhou, Changsheng; Musa, Omer

    2018-05-01

    In this paper, numerical investigation of ignition transient in a dual pulse solid rocket motor has been conducted. An in-house code has been developed in order to solve multi-physics governing equations, including unsteady compressible flow, heat conduction and structural dynamic. The simplified numerical models for solid propellant ignition and combustion have been added. The conventional serial staggered algorithm is adopted to simulate the fluid structure interaction problems in a loosely-coupled manner. The accuracy of the coupling procedure is validated by the behavior of a cantilever panel subjected to a shock wave. Then, the detailed flow field development, flame propagation characteristics, pressure evolution in the combustion chamber, and the structural response of metal diaphragm are analyzed carefully. The burst-time and burst-pressure of the metal diaphragm are also obtained. The individual effects of the igniter's mass flow rate, metal diaphragm thickness and diameter on the ignition transient have been systemically compared. The numerical results show that the evolution of the flow field in the combustion chamber, the temperature distribution on the propellant surface and the pressure loading on the metal diaphragm surface present a strong three-dimensional behavior during the initial ignition stage. The rupture of metal diaphragm is not only related to the magnitude of pressure loading on the diaphragm surface, but also to the history of pressure loading. The metal diaphragm thickness and diameter have a significant effect on the burst-time and burst-pressure of metal diaphragm.

  8. GeN-Foam: a novel OpenFOAM{sup ®} based multi-physics solver for 2D/3D transient analysis of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Fiorina, Carlo, E-mail: carlo.fiorina@psi.ch [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Clifford, Ivor [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland); Aufiero, Manuele [LPSC-IN2P3-CNRS/UJF/Grenoble INP, 53 avenue des Martyrs, 38026 Grenoble Cedex (France); Mikityuk, Konstantin [Paul Scherrer Institut, Nuclear Energy and Safety Department, Laboratory for Reactor Physics and Systems Behaviour – PSI, Villigen 5232 (Switzerland)

    2015-12-01

    Highlights: • Development of a new multi-physics solver based on OpenFOAM{sup ®}. • Tight coupling of thermal-hydraulics, thermal-mechanics and neutronics. • Combined use of traditional RANS and porous-medium models. • Mesh for neutronics deformed according to the predicted displacement field. • Use of three unstructured meshes, adaptive time step, parallel computing. - Abstract: The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM{sup ®}, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

  9. GeN-Foam: a novel OpenFOAM"® based multi-physics solver for 2D/3D transient analysis of nuclear reactors

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Clifford, Ivor; Aufiero, Manuele; Mikityuk, Konstantin

    2015-01-01

    Highlights: • Development of a new multi-physics solver based on OpenFOAM"®. • Tight coupling of thermal-hydraulics, thermal-mechanics and neutronics. • Combined use of traditional RANS and porous-medium models. • Mesh for neutronics deformed according to the predicted displacement field. • Use of three unstructured meshes, adaptive time step, parallel computing. - Abstract: The FAST group at the Paul Scherrer Institut has been developing a code system for reactor analysis for many years. For transient analysis, this code system is currently based on a state-of-the-art coupled TRACE-PARCS routine. This work presents an attempt to supplement the FAST code system with a novel solver characterized by tight coupling between the different equations, parallel computing capabilities, adaptive time-stepping and more accurate treatment of some of the phenomena involved in a reactor transient. The new solver is based on OpenFOAM"®, an open-source C++ library for the solution of partial differential equations using finite-volume discretization. It couples together a multi-scale fine/coarse mesh sub-solver for thermal-hydraulics, a multi-group diffusion sub-solver for neutronics, a displacement-based sub-solver for thermal-mechanics and a finite-difference model for the temperature field in the fuel. It is targeted toward the analysis of pin-based reactors (e.g., liquid metal fast reactors or light water reactors) or homogeneous reactors (e.g., fast-spectrum molten salt reactors). This paper presents each “single-physics” sub-solver and the overall coupling strategy, using the sodium-cooled fast reactor as a test case, and essential code verification tests are described.

  10. 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

  11. 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)

  12. Coupled Multi-physical Simulations for the Assessment of Nuclear Waste Repository Concepts: Modeling, Software Development and Simulation

    Science.gov (United States)

    Massmann, J.; Nagel, T.; Bilke, L.; Böttcher, N.; Heusermann, S.; Fischer, T.; Kumar, V.; Schäfers, A.; Shao, H.; Vogel, P.; Wang, W.; Watanabe, N.; Ziefle, G.; Kolditz, O.

    2016-12-01

    As part of the German site selection process for a high-level nuclear waste repository, different repository concepts in the geological candidate formations rock salt, clay stone and crystalline rock are being discussed. An open assessment of these concepts using numerical simulations requires physical models capturing the individual particularities of each rock type and associated geotechnical barrier concept to a comparable level of sophistication. In a joint work group of the Helmholtz Centre for Environmental Research (UFZ) and the German Federal Institute for Geosciences and Natural Resources (BGR), scientists of the UFZ are developing and implementing multiphysical process models while BGR scientists apply them to large scale analyses. The advances in simulation methods for waste repositories are incorporated into the open-source code OpenGeoSys. Here, recent application-driven progress in this context is highlighted. A robust implementation of visco-plasticity with temperature-dependent properties into a framework for the thermo-mechanical analysis of rock salt will be shown. The model enables the simulation of heat transport along with its consequences on the elastic response as well as on primary and secondary creep or the occurrence of dilatancy in the repository near field. Transverse isotropy, non-isothermal hydraulic processes and their coupling to mechanical stresses are taken into account for the analysis of repositories in clay stone. These processes are also considered in the near field analyses of engineered barrier systems, including the swelling/shrinkage of the bentonite material. The temperature-dependent saturation evolution around the heat-emitting waste container is described by different multiphase flow formulations. For all mentioned applications, we illustrate the workflow from model development and implementation, over verification and validation, to repository-scale application simulations using methods of high performance computing.

  13. Reactor simulator development. Workshop material

    International Nuclear Information System (INIS)

    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 reactor 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. This publication consists of course material for workshops on development of such reactor simulators. Participants in the workshops are provided with instruction and practice in the development of reactor simulation computer codes using a model development system that assembles integrated codes from a selection of pre-programmed and tested sub-components. This provides insight and understanding into the construction and assumptions of the codes that model the design and operational characteristics of various power reactor systems. The main objective is to demonstrate simple nuclear reactor dynamics with hands-on simulation experience. Using one of the modular development systems, CASSIM tm , a simple point kinetic reactor model is developed, followed by a model that simulates the Xenon/Iodine concentration on changes in reactor power. Lastly, an absorber and adjuster control rod, and a liquid zone model are developed to control reactivity. The built model is used to demonstrate reactor behavior in sub-critical, critical and supercritical states, and to observe the impact of malfunctions of various reactivity control mechanisms on reactor dynamics. Using a PHWR simulator, participants practice typical procedures for a reactor startup and approach to criticality. This workshop material consists of an introduction to systems used for developing reactor simulators, an overview of the dynamic simulation

  14. 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)

  15. 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)

  16. 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

  17. Discrete multi-physics simulations of diffusive and convective mass transfer in boundary layers containing motile cilia in lungs.

    Science.gov (United States)

    Ariane, Mostapha; Kassinos, Stavros; Velaga, Sitaram; Alexiadis, Alessio

    2018-04-01

    In this paper, the mass transfer coefficient (permeability) of boundary layers containing motile cilia is investigated by means of discrete multi-physics. The idea is to understand the main mechanisms of mass transport occurring in a ciliated-layer; one specific application being inhaled drugs in the respiratory epithelium. The effect of drug diffusivity, cilia beat frequency and cilia flexibility is studied. Our results show the existence of three mass transfer regimes. A low frequency regime, which we called shielding regime, where the presence of the cilia hinders mass transport; an intermediate frequency regime, which we have called diffusive regime, where diffusion is the controlling mechanism; and a high frequency regime, which we have called convective regime, where the degree of bending of the cilia seems to be the most important factor controlling mass transfer in the ciliated-layer. Since the flexibility of the cilia and the frequency of the beat changes with age and health conditions, the knowledge of these three regimes allows prediction of how mass transfer varies with these factors. Copyright © 2018 Elsevier Ltd. All rights reserved.

  18. 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.

  19. 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.

  20. 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

  1. An Efficient Upscaling Process Based on a Unified Fine-scale Multi-Physics Model for Flow Simulation in Naturally Fracture Carbonate Karst Reservoirs

    KAUST Repository

    Bi, Linfeng

    2009-01-01

    The main challenges in modeling fluid flow through naturally-fractured carbonate karst reservoirs are how to address various flow physics in complex geological architectures due to the presence of vugs and caves which are connected via fracture networks at multiple scales. In this paper, we present a unified multi-physics model that adapts to the complex flow regime through naturally-fractured carbonate karst reservoirs. This approach generalizes Stokes-Brinkman model (Popov et al. 2007). The fracture networks provide the essential connection between the caves in carbonate karst reservoirs. It is thus very important to resolve the flow in fracture network and the interaction between fractures and caves to better understand the complex flow behavior. The idea is to use Stokes-Brinkman model to represent flow through rock matrix, void caves as well as intermediate flows in very high permeability regions and to use an idea similar to discrete fracture network model to represent flow in fracture network. Consequently, various numerical solution strategies can be efficiently applied to greatly improve the computational efficiency in flow simulations. We have applied this unified multi-physics model as a fine-scale flow solver in scale-up computations. Both local and global scale-up are considered. It is found that global scale-up has much more accurate than local scale-up. Global scale-up requires the solution of global flow problems on fine grid, which generally is computationally expensive. The proposed model has the ability to deal with large number of fractures and caves, which facilitate the application of Stokes-Brinkman model in global scale-up computation. The proposed model flexibly adapts to the different flow physics in naturally-fractured carbonate karst reservoirs in a simple and effective way. It certainly extends modeling and predicting capability in efficient development of this important type of reservoir.

  2. 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

  3. Development of Multi-Physics Dynamics Models for High-Frequency Large-Amplitude Structural Response Simulation

    Science.gov (United States)

    Derkevorkian, Armen; Peterson, Lee; Kolaini, Ali R.; Hendricks, Terry J.; Nesmith, Bill J.

    2016-01-01

    An analytic approach is demonstrated to reveal potential pyroshock -driven dynamic effects causing power losses in the Thermo -Electric (TE) module bars of the Mars Science Laboratory (MSL) Multi -Mission Radioisotope Thermoelectric Generator (MMRTG). This study utilizes high- fidelity finite element analysis with SIERRA/PRESTO codes to estimate wave propagation effects due to large -amplitude suddenly -applied pyro shock loads in the MMRTG. A high fidelity model of the TE module bar was created with approximately 30 million degrees -of-freedom (DOF). First, a quasi -static preload was applied on top of the TE module bar, then transient tri- axial acceleration inputs were simultaneously applied on the preloaded module. The applied input acceleration signals were measured during MMRTG shock qualification tests performed at the Jet Propulsion Laboratory. An explicit finite element solver in the SIERRA/PRESTO computational environment, along with a 3000 processor parallel super -computing framework at NASA -AMES, was used for the simulation. The simulation results were investigated both qualitatively and quantitatively. The predicted shock wave propagation results provide detailed structural responses throughout the TE module bar, and key insights into the dynamic response (i.e., loads, displacements, accelerations) of critical internal spring/piston compression systems, TE materials, and internal component interfaces in the MMRTG TE module bar. They also provide confidence on the viability of this high -fidelity modeling scheme to accurately predict shock wave propagation patterns within complex structures. This analytic approach is envisioned for modeling shock sensitive hardware susceptible to intense shock environments positioned near shock separation devices in modern space vehicles and systems.

  4. 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)

  5. Coupling biomechanics to a cellular level model: an approach to patient-specific image driven multi-scale and multi-physics tumor simulation.

    Science.gov (United States)

    May, Christian P; Kolokotroni, Eleni; Stamatakos, Georgios S; Büchler, Philippe

    2011-10-01

    Modeling of tumor growth has been performed according to various approaches addressing different biocomplexity levels and spatiotemporal scales. Mathematical treatments range from partial differential equation based diffusion models to rule-based cellular level simulators, aiming at both improving our quantitative understanding of the underlying biological processes and, in the mid- and long term, constructing reliable multi-scale predictive platforms to support patient-individualized treatment planning and optimization. The aim of this paper is to establish a multi-scale and multi-physics approach to tumor modeling taking into account both the cellular and the macroscopic mechanical level. Therefore, an already developed biomodel of clinical tumor growth and response to treatment is self-consistently coupled with a biomechanical model. Results are presented for the free growth case of the imageable component of an initially point-like glioblastoma multiforme tumor. The composite model leads to significant tumor shape corrections that are achieved through the utilization of environmental pressure information and the application of biomechanical principles. Using the ratio of smallest to largest moment of inertia of the tumor material to quantify the effect of our coupled approach, we have found a tumor shape correction of 20% by coupling biomechanics to the cellular simulator as compared to a cellular simulation without preferred growth directions. We conclude that the integration of the two models provides additional morphological insight into realistic tumor growth behavior. Therefore, it might be used for the development of an advanced oncosimulator focusing on tumor types for which morphology plays an important role in surgical and/or radio-therapeutic treatment planning. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. 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)

  7. 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)

  8. 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)

  9. Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Mustafa Sacit; none,; Flanagan, George F. [ORNL; Poore III, Willis P. [ORNL; Muhlheim, Michael David [ORNL

    2014-07-30

    An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two types of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link library (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.

  10. Science based integrated approach to advanced nuclear fuel development - integrated multi-scale multi-physics hierarchical modeling and simulation framework Part III: cladding

    International Nuclear Information System (INIS)

    Tome, Carlos N.; Caro, J.A.; Lebensohn, R.A.; Unal, Cetin; Arsenlis, A.; Marian, J.; Pasamehmetoglu, K.

    2010-01-01

    Advancing the performance of Light Water Reactors, Advanced Nuclear Fuel Cycles, and Advanced Reactors, such as the Next Generation Nuclear Power Plants, requires enhancing our fundamental understanding of fuel and materials behavior under irradiation. The capability to accurately model the nuclear fuel systems to develop predictive tools is critical. Not only are fabrication and performance models needed to understand specific aspects of the nuclear fuel, fully coupled fuel simulation codes are required to achieve licensing of specific nuclear fuel designs for operation. The backbone of these codes, models, and simulations is a fundamental understanding and predictive capability for simulating the phase and microstructural behavior of the nuclear fuel system materials and matrices. In this paper we review the current status of the advanced modeling and simulation of nuclear reactor cladding, with emphasis on what is available and what is to be developed in each scale of the project, how we propose to pass information from one scale to the next, and what experimental information is required for benchmarking and advancing the modeling at each scale level.

  11. 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.)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    International Nuclear Information System (INIS)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y.

    2011-01-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  18. A multi-physics code system based on ANC9, VIPRE-W and BOA for CIPS evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Sung, Y.; Secker, J.; Beard, C.; Hilton, P.; Wang, G.; Oelrich, R.; Karoutas, Z.; Sung, Y. [Westinghouse Electric Company LLC, Pittsburgh (United States)

    2011-07-01

    This paper summarizes the development of a multi-physics code system for evaluation of Crud Induced Power Shift (CIPS) phenomenon experienced in some Pressurized Water Reactors (PWR). CIPS is an unexpected change in reactor core axial power distribution, caused by boron compounds in crud deposited in the high power fuel assemblies undergoing subcooled boiling. As part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) sponsored by the US Department of Energy (DOE), this paper describes the initial linkage and application of a multi-physics code system ANC9/VIPRE-W/BOA for evaluating changes in core power distributions due to boron deposited in crud. The initial linkage of the code system along with the application results will be the base for the future CASL development. (author)

  19. Modular ORIGEN-S for multi-physics code systems

    International Nuclear Information System (INIS)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C.; Galloway, Jack

    2011-01-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  20. Modular ORIGEN-S for multi-physics code systems

    Energy Technology Data Exchange (ETDEWEB)

    Yesilyurt, Gokhan; Clarno, Kevin T.; Gauld, Ian C., E-mail: yesilyurtg@ornl.gov, E-mail: clarnokt@ornl.gov, E-mail: gauldi@ornl.gov [Oak Ridge National Laboratory, TN (United States); Galloway, Jack, E-mail: jack@galloways.net [Los Alamos National Laboratory, Los Alamos, NM (United States)

    2011-07-01

    The ORIGEN-S code in the SCALE 6.0 nuclear analysis code suite is a well-validated tool to calculate the time-dependent concentrations of nuclides due to isotopic depletion, decay, and transmutation for many systems in a wide range of time scales. Application areas include nuclear reactor and spent fuel storage analyses, burnup credit evaluations, decay heat calculations, and environmental assessments. Although simple to use within the SCALE 6.0 code system, especially with the ORIGEN-ARP graphical user interface, it is generally complex to use as a component within an externally developed code suite because of its tight coupling within the infrastructure of the larger SCALE 6.0 system. The ORIGEN2 code, which has been widely integrated within other simulation suites, is no longer maintained by Oak Ridge National Laboratory (ORNL), has obsolete data, and has a relatively small validation database. Therefore, a modular version of the SCALE/ORIGEN-S code was developed to simplify its integration with other software packages to allow multi-physics nuclear code systems to easily incorporate the well-validated isotopic depletion, decay, and transmutation capability to perform realistic nuclear reactor and fuel simulations. SCALE/ORIGEN-S was extensively restructured to develop a modular version that allows direct access to the matrix solvers embedded in the code. Problem initialization and the solver were segregated to provide a simple application program interface and fewer input/output operations for the multi-physics nuclear code systems. Furthermore, new interfaces were implemented to access and modify the ORIGEN-S input variables and nuclear cross-section data through external drivers. Three example drivers were implemented, in the C, C++, and Fortran 90 programming languages, to demonstrate the modular use of the new capability. This modular version of SCALE/ORIGEN-S has been embedded within several multi-physics software development projects at ORNL, including

  1. 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)

  2. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Science.gov (United States)

    Turinsky, Paul J.; Kothe, Douglas B.

    2016-05-01

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics ;core simulator; based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL's evolving M

  3. PUMA Development through a Multi physics Approach

    International Nuclear Information System (INIS)

    Cheon, Jinsik; Kim, Junehyung; Lee, Byoungoon; Lee, Chanbock

    2013-01-01

    Meanwhile advances of numerical methods make it possible for the multi physics problem to be solved in a fully coupled way. In addition to a multidimensional, multi physical approach, a nuclear fuel performance analysis code, which is 1D code, should be improved by accommodating the state-of-the-art in the numerical analysis to support current fuel design and performance analysis. In particular, the coupling between the mechanical equilibrium equation and a set of numerically stiff kinetics equations for fission gas release is of great importance for a multi physics simulation of nuclear fuel. Instead, coupling between temperature and fuel constituent was found to be made with a relative ease by employing an ordinary differential equations solver. As an effort for a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code), the deformation of U-Zr fuel for SFR in connection with a fission gas release model is analyzed. A finite element analyses for purely mechanical problems are performed using a backward differentiation formula, and are subjected to scrupulous verification with Abaqus. Then mechanical equilibrium equation and the equations for fission gas release are coupled with the same differential-algebraic equations (DAE) solver

  4. 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

  5. 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.

  6. 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.

  7. 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)

  8. Internal multi-scale multi-physics coupled system for high fidelity simulation of light water reactors

    International Nuclear Information System (INIS)

    Ivanov, A.; Sanchez, V.; Stieglitz, R.; Ivanov, K.

    2014-01-01

    Highlights: • The current paper focuses on a optimized method for performing coupled Monte-Carlo/thermal–hydraulics calculations. • Innovative on-the-fly method for supplying the temperature and density distributions is presented. • Convergence acceleration method is presented. It is proven applicable by generalizing the Robbins-Monro theorem. • Tallying is optimized by using collision probability estimator for the power profile estimation. - Abstract: In order to increase the accuracy and the degree of spatial and energy resolution of core design studies, coupled 3D neutronic (multi-group deterministic and continuous energy Monte-Carlo) and 3D thermal–hydraulic (CFD and subchannel) codes are being developed worldwide. At KIT, both deterministic and Monte-Carlo codes were coupled with subchannel codes and applied to predict the safety-related design parameters such as minimal critical power ratio (MCPR), maximal cladding and fuel temperature, departure from nuclide boiling ratio (DNBR). These coupling approaches were revised and considerably improved. Innovative method of internal on-the-fly thermal feedback interchange between the codes was implemented. It no longer relies on explicit material definitions and allows the modeling of temperature and density distributions based on the cell coordinates. In contrast to all existing coupled schemes, this method uses only standard MCNP geometry input and requires only proper definition of the geometrical dimensions. The initial material definition is arbitrary and is determined on-the-fly during the neutron transport by the thermal–hydraulic feedback. Another key issue addressed is the optimal application of parallel computing and the implementation of less time consuming tally estimators. Using multi-processor computer architectures and implementing collision density flux estimator, it is possible to reduce the Monte-Carlo running time and obtain converged results within reasonable time limit. The coupled calculation was accelerated further, by implementing stochastic approximation-based relaxation technique. Further, it is shown that large fuel assemblies can be analyzed on subchannel level

  9. 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.)

  10. Modeling and simulation challenges pursued by the Consortium for Advanced Simulation of Light Water Reactors (CASL)

    Energy Technology Data Exchange (ETDEWEB)

    Turinsky, Paul J., E-mail: turinsky@ncsu.edu [North Carolina State University, PO Box 7926, Raleigh, NC 27695-7926 (United States); Kothe, Douglas B., E-mail: kothe@ornl.gov [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6164 (United States)

    2016-05-15

    The Consortium for the Advanced Simulation of Light Water Reactors (CASL), the first Energy Innovation Hub of the Department of Energy, was established in 2010 with the goal of providing modeling and simulation (M&S) capabilities that support and accelerate the improvement of nuclear energy's economic competitiveness and the reduction of spent nuclear fuel volume per unit energy, and all while assuring nuclear safety. To accomplish this requires advances in M&S capabilities in radiation transport, thermal-hydraulics, fuel performance and corrosion chemistry. To focus CASL's R&D, industry challenge problems have been defined, which equate with long standing issues of the nuclear power industry that M&S can assist in addressing. To date CASL has developed a multi-physics “core simulator” based upon pin-resolved radiation transport and subchannel (within fuel assembly) thermal-hydraulics, capitalizing on the capabilities of high performance computing. CASL's fuel performance M&S capability can also be optionally integrated into the core simulator, yielding a coupled multi-physics capability with untapped predictive potential. Material models have been developed to enhance predictive capabilities of fuel clad creep and growth, along with deeper understanding of zirconium alloy clad oxidation and hydrogen pickup. Understanding of corrosion chemistry (e.g., CRUD formation) has evolved at all scales: micro, meso and macro. CFD R&D has focused on improvement in closure models for subcooled boiling and bubbly flow, and the formulation of robust numerical solution algorithms. For multiphysics integration, several iterative acceleration methods have been assessed, illuminating areas where further research is needed. Finally, uncertainty quantification and data assimilation techniques, based upon sampling approaches, have been made more feasible for practicing nuclear engineers via R&D on dimensional reduction and biased sampling. Industry adoption of CASL

  11. 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.

  12. GNES-R: Global nuclear energy simulator for reactors task 1: High-fidelity neutron transport

    International Nuclear Information System (INIS)

    Clarno, K.; De Almeida, V.; D'Azevedo, E.; De Oliveira, C.; Hamilton, S.

    2006-01-01

    A multi-laboratory, multi-university collaboration has formed to advance the state-of-the-art in high-fidelity, coupled-physics simulation of nuclear energy systems. We are embarking on the first-phase in the development of a new suite of simulation tools dedicated to the advancement of nuclear science and engineering technologies. We seek to develop and demonstrate a new generation of multi-physics simulation tools that will explore the scientific phenomena of tightly coupled physics parameters within nuclear systems, support the design and licensing of advanced nuclear reactors, and provide benchmark quality solutions for code validation. In this paper, we have presented the general scope of the collaborative project and discuss the specific challenges of high-fidelity neutronics for nuclear reactor simulation and the inroads we have made along this path. The high-performance computing neutronics code system utilizes the latest version of SCALE to generate accurate, problem-dependent cross sections, which are used in NEWTRNX - a new 3-D, general-geometry, discrete-ordinates solver based on the Slice-Balance Approach. The Global Nuclear Energy Simulator for Reactors (GNES-R) team is embarking on a long-term simulation development project that encompasses multiple laboratories and universities for the expansion of high-fidelity coupled-physics simulation of nuclear energy systems. (authors)

  13. 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)

  14. 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.

  15. 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

  16. 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

  17. Simulation for nuclear reactor technology

    International Nuclear Information System (INIS)

    Walton, D.G.

    1985-01-01

    Mathematical modelling forms the cornerstone of both the physical sciences and the engineering sciences. Computer modelling represents an extension of mathematical modelling into areas which are either not sufficiently accurate or not conveniently modelled by analytical techniques and where highly complex models can be investigated on a sufficiently short timescale. Simulation represents the application of these modelling techniques to real systems, thus enabling information on plant characteristics to be gained without either constructing or operating the full-scale plant or system under consideration. Developments in computer systems modelling techniques, safety and operating philosophy, and human psychology studies lead to improvements in simulation practice. New concepts in simulation also lead to a demand for new types of computers, dedicated computer systems are increasingly being developed for plant analysis, and the advent and development of microprocessors has also led to new techniques both in simulation and simulators. Progress in these areas is presented in this book

  18. Design and multi-physics optimization of rotary MRF brakes

    Science.gov (United States)

    Topcu, Okan; Taşcıoğlu, Yiğit; Konukseven, Erhan İlhan

    2018-03-01

    Particle swarm optimization (PSO) is a popular method to solve the optimization problems. However, calculations for each particle will be excessive when the number of particles and complexity of the problem increases. As a result, the execution speed will be too slow to achieve the optimized solution. Thus, this paper proposes an automated design and optimization method for rotary MRF brakes and similar multi-physics problems. A modified PSO algorithm is developed for solving multi-physics engineering optimization problems. The difference between the proposed method and the conventional PSO is to split up the original single population into several subpopulations according to the division of labor. The distribution of tasks and the transfer of information to the next party have been inspired by behaviors of a hunting party. Simulation results show that the proposed modified PSO algorithm can overcome the problem of heavy computational burden of multi-physics problems while improving the accuracy. Wire type, MR fluid type, magnetic core material, and ideal current inputs have been determined by the optimization process. To the best of the authors' knowledge, this multi-physics approach is novel for optimizing rotary MRF brakes and the developed PSO algorithm is capable of solving other multi-physics engineering optimization problems. The proposed method has showed both better performance compared to the conventional PSO and also has provided small, lightweight, high impedance rotary MRF brake designs.

  19. 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

  20. A theory manual for multi-physics code coupling in LIME.

    Energy Technology Data Exchange (ETDEWEB)

    Belcourt, Noel; Bartlett, Roscoe Ainsworth; Pawlowski, Roger Patrick; Schmidt, Rodney Cannon; Hooper, Russell Warren

    2011-03-01

    The Lightweight Integrating Multi-physics Environment (LIME) is a software package for creating multi-physics simulation codes. Its primary application space is when computer codes are currently available to solve different parts of a multi-physics problem and now need to be coupled with other such codes. In this report we define a common domain language for discussing multi-physics coupling and describe the basic theory associated with multiphysics coupling algorithms that are to be supported in LIME. We provide an assessment of coupling techniques for both steady-state and time dependent coupled systems. Example couplings are also demonstrated.

  1. 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

  2. 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

  3. Multi physical simulations in the nuclear sector using ANSYS: uses and benefits; Simulaciones multifisicas en el sector nuclear usando ANYS: Aplicaciones y beneficios

    Energy Technology Data Exchange (ETDEWEB)

    Morales, M.

    2014-07-01

    The numerical simulation as a tool of design and calculation for engineering has been widely used in the nuclear sector since the 1980s. The possibility of solving problems of great complexity, together with the rapid evolution of the computing power of computers has located these methods in the Centre of the activities of design and analysis.

  4. 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.)

  5. 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)

  6. 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

  7. 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

  8. Towards an Industrial Application of Statistical Uncertainty Analysis Methods to Multi-physical Modelling and Safety Analyses

    International Nuclear Information System (INIS)

    Zhang, Jinzhao; Segurado, Jacobo; Schneidesch, Christophe

    2013-01-01

    Since 1980's, Tractebel Engineering (TE) has being developed and applied a multi-physical modelling and safety analyses capability, based on a code package consisting of the best estimate 3D neutronic (PANTHER), system thermal hydraulic (RELAP5), core sub-channel thermal hydraulic (COBRA-3C), and fuel thermal mechanic (FRAPCON/FRAPTRAN) codes. A series of methodologies have been developed to perform and to license the reactor safety analysis and core reload design, based on the deterministic bounding approach. Following the recent trends in research and development as well as in industrial applications, TE has been working since 2010 towards the application of the statistical sensitivity and uncertainty analysis methods to the multi-physical modelling and licensing safety analyses. In this paper, the TE multi-physical modelling and safety analyses capability is first described, followed by the proposed TE best estimate plus statistical uncertainty analysis method (BESUAM). The chosen statistical sensitivity and uncertainty analysis methods (non-parametric order statistic method or bootstrap) and tool (DAKOTA) are then presented, followed by some preliminary results of their applications to FRAPCON/FRAPTRAN simulation of OECD RIA fuel rod codes benchmark and RELAP5/MOD3.3 simulation of THTF tests. (authors)

  9. Development of CFD software for the simulation of thermal hydraulics in advanced nuclear reactors. Final report

    International Nuclear Information System (INIS)

    Bachar, Abdelaziz; Haslinger, Wolfgang; Scheuerer, Georg; Theodoridis, Georgios

    2015-01-01

    The objectives of the project were: Improvement of the simulation accuracy for nuclear reactor thermo-hydraulics by coupling system codes with three-dimensional CFD software; Extension of CFD software to predict thermo-hydraulics in advanced reactor concepts; Validation of the CFD software by simulation different UPTF TRAM-C test cases and development of best practice guidelines. The CFD module was based on the ANSYS CFD software and the system code ATHLET of GRS. All three objectives were met: The coupled ATHLET-ANSYS CFD software is in use at GRS and TU Muenchen. Besides the test cases described in the report, it has been used for other applications, for instance the TALL-3D experiment of KTH Stockholm. The CFD software was extended with material properties for liquid metals, and validated using existing data. Several new concepts were tested when applying the CFD software to the UPTF test cases: Simulations with Conjugate Heat Transfer (CHT) were performed for the first time. This led to better agreement between predictions and data and reduced uncertainties when applying temperature boundary conditions. The meshes for the CHT simulation were also used for a coupled fluid-structure-thermal analysis which was another novelty. The results of the multi-physics analysis showed plausible results for the mechanical and thermal stresses. The workflow developed as part of the current project can be directly used for industrial nuclear reactor simulations. Finally, simulations for two-phase flows with and without interfacial mass transfer were performed. These showed good agreement with data. However, a persisting problem for the simulation of multi-phase flows are the long simulation times which make use for industrial applications difficult.

  10. 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

  11. Reactor numerical simulation and hydraulic test research

    International Nuclear Information System (INIS)

    Yang, L. S.

    2009-01-01

    In recent years, the computer hardware was improved on the numerical simulation on flow field in the reactor. In our laboratory, we usually use the Pro/e or UG commercial software. After completed topology geometry, ICEM-CFD is used to get mesh for computation. Exact geometrical similarity is maintained between the main flow paths of the model and the prototype, with the exception of the core simulation design of the fuel assemblies. The drive line system is composed of drive mechanism, guide bush assembly, fuel assembly and control rod assembly, and fitted with the rod level indicator and drive mechanism power device

  12. Predictive modeling of coupled multi-physics systems: I. Theory

    International Nuclear Information System (INIS)

    Cacuci, Dan Gabriel

    2014-01-01

    Highlights: • We developed “predictive modeling of coupled multi-physics systems (PMCMPS)”. • PMCMPS reduces predicted uncertainties in predicted model responses and parameters. • PMCMPS treats efficiently very large coupled systems. - Abstract: This work presents an innovative mathematical methodology for “predictive modeling of coupled multi-physics systems (PMCMPS).” This methodology takes into account fully the coupling terms between the systems but requires only the computational resources that would be needed to perform predictive modeling on each system separately. The PMCMPS methodology uses the maximum entropy principle to construct an optimal approximation of the unknown a priori distribution based on a priori known mean values and uncertainties characterizing the parameters and responses for both multi-physics models. This “maximum entropy”-approximate a priori distribution is combined, using Bayes’ theorem, with the “likelihood” provided by the multi-physics simulation models. Subsequently, the posterior distribution thus obtained is evaluated using the saddle-point method to obtain analytical expressions for the optimally predicted values for the multi-physics models parameters and responses along with corresponding reduced uncertainties. Noteworthy, the predictive modeling methodology for the coupled systems is constructed such that the systems can be considered sequentially rather than simultaneously, while preserving exactly the same results as if the systems were treated simultaneously. Consequently, very large coupled systems, which could perhaps exceed available computational resources if treated simultaneously, can be treated with the PMCMPS methodology presented in this work sequentially and without any loss of generality or information, requiring just the resources that would be needed if the systems were treated sequentially

  13. Specification of requirements for the virtual environment for reactor applications simulation environment

    International Nuclear Information System (INIS)

    Hess, S. M.; Pytel, M.

    2012-01-01

    In 2010, the United States Dept. of Energy initiated a research and development effort to develop modern modeling and simulation methods that could utilize high performance computing capabilities to address issues important to nuclear power plant operation, safety and sustainability. To respond to this need, a consortium of national laboratories, academic institutions and industry partners (the Consortium for Advanced Simulation of Light Water Reactors - CASL) was formed to develop an integrated Virtual Environment for Reactor Applications (VERA) modeling and simulation capability. A critical element for the success of the CASL research and development effort was the development of an integrated set of overarching requirements that provides guidance in the planning, development, and management of the VERA modeling and simulation software. These requirements also provide a mechanism from which the needs of a broad array of external CASL stakeholders (e.g. reactor / fuel vendors, plant owner / operators, regulatory personnel, etc.) can be identified and integrated into the VERA development plans. This paper presents an overview of the initial set of requirements contained within the VERA Requirements Document (VRD) that currently is being used to govern development of the VERA software within the CASL program. The complex interdisciplinary nature of these requirements together with a multi-physics coupling approach to realize a core simulator capability pose a challenge to how the VRD should be derived and subsequently revised to accommodate the needs of different stakeholders. Thus, the VRD is viewed as an evolving document that will be updated periodically to reflect the changing needs of identified CASL stakeholders and lessons learned during the progress of the CASL modeling and simulation program. (authors)

  14. FRESCO: fusion reactor simulation code for tokamaks

    International Nuclear Information System (INIS)

    Mantsinen, M.J.

    1995-03-01

    The study of the dynamics of tokamak fusion reactors, a zero-dimensional particle and power balance code FRESCO (Fusion Reactor Simulation Code) has been developed at the Department of Technical Physics of Helsinki University of Technology. The FRESCO code is based on zero-dimensional particle and power balance equations averaged over prescribed plasma profiles. In the report the data structure of the FRESCO code is described, including the description of the COMMON statements, program input, and program output. The general structure of the code is described, including the description of subprograms and functions. The physical model used and examples of the code performance are also included in the report. (121 tabs.) (author)

  15. Lithium-Ion Battery Safety Study Using Multi-Physics Internal Short-Circuit Model (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G-.H.; Smith, K.; Pesaran, A.

    2009-06-01

    This presentation outlines NREL's multi-physics simulation study to characterize an internal short by linking and integrating electrochemical cell, electro-thermal, and abuse reaction kinetics models.

  16. Two-Step Multi-Physics Analysis of an Annular Linear Induction Pump for Fission Power Systems

    Science.gov (United States)

    Geng, Steven M.; Reid, Terry V.

    2016-01-01

    One of the key technologies associated with fission power systems (FPS) is the annular linear induction pump (ALIP). ALIPs are used to circulate liquid-metal fluid for transporting thermal energy from the nuclear reactor to the power conversion device. ALIPs designed and built to date for FPS project applications have not performed up to expectations. A unique, two-step approach was taken toward the multi-physics examination of an ALIP using ANSYS Maxwell 3D and Fluent. This multi-physics approach was developed so that engineers could investigate design variations that might improve pump performance. Of interest was to determine if simple geometric modifications could be made to the ALIP components with the goal of increasing the Lorentz forces acting on the liquid-metal fluid, which in turn would increase pumping capacity. The multi-physics model first calculates the Lorentz forces acting on the liquid metal fluid in the ALIP annulus. These forces are then used in a computational fluid dynamics simulation as (a) internal boundary conditions and (b) source functions in the momentum equations within the Navier-Stokes equations. The end result of the two-step analysis is a predicted pump pressure rise that can be compared with experimental data.

  17. 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

  18. 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)

  19. 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)

  20. Research nuclear reactor start-up simulator

    International Nuclear Information System (INIS)

    Sofo Haro, M.; Cantero, P.

    2009-01-01

    This work presents the design and FPGA implementation of a research nuclear reactor start-up simulator. Its aim is to generate a set of signals that allow replacing the neutron detector for stimulated signals, to feed the measurement electronic of the start-up channels, to check its operation, together with the start-up security logic. The simulator presented can be configured on three independent channels and adjust the shape of the output pulses. Furthermore, each channel can be configured in 'rate' mode, where you can specify the growth rate of the pulse frequency in %/s. Result and details of the implementation on FPGA of the different functional blocks are given. (author)

  1. 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)

  2. Developing a multi-physics solver in APOLLO3 and applications to cross section homogenization

    International Nuclear Information System (INIS)

    Dugan, Kevin-James

    2016-01-01

    Multi-physics coupling is becoming of large interest in the nuclear engineering and computational science fields. The ability to obtain accurate solutions to realistic models is important to the design and licensing of novel reactor designs, especially in design basis accident situations. The physical models involved in calculating accident behavior in nuclear reactors includes: neutron transport, thermal conduction/convection, thermo-mechanics in fuel and support structure, fuel stoichiometry, among others. However, this thesis focuses on the coupling between two models, neutron transport and thermal conduction/convection.The goal of this thesis is to develop a multi-physics solver for simulating accidents in nuclear reactors. The focus is both on the simulation environment and the data treatment used in such simulations.This work discusses the development of a multi-physics framework based around the Jacobian-Free Newton-Krylov (JFNK) method. The framework includes linear and nonlinear solvers, along with interfaces to existing numerical codes that solve neutron transport and thermal hydraulics models (APOLLO3 and MCTH respectively) through the computation of residuals. a new formulation for the neutron transport residual is explored, which reduces the solution size and search space by a large factor; instead of the residual being based on the angular flux, it is based on the fission source.The question of whether using a fundamental mode distribution of the neutron flux for cross section homogenization is sufficiently accurate during fast transients is also explored. It is shown that in an infinite homogeneous medium, using homogenized cross sections produced with a fundamental mode flux differ significantly from a reference solution. The error is remedied by using an alternative weighting flux taken from a time dependent calculation; either a time-integrated flux or an asymptotic solution. The time-integrated flux comes from the multi-physics solution of the

  3. Nuclear reactor core modelling in multifunctional simulators

    International Nuclear Information System (INIS)

    Puska, E.K.

    1999-01-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  4. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  5. Updates to the Generation of Physics Data Inputs for MAMMOTH Simulations of the Transient Reactor Test Facility - FY2016

    Energy Technology Data Exchange (ETDEWEB)

    Ortensi, Javier [Idaho National Lab. (INL), Idaho Falls, ID (United States); Baker, Benjamin Allen [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schunert, Sebastian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Wang, Yaqi [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gleicher, Frederick Nathan [Idaho National Lab. (INL), Idaho Falls, ID (United States); DeHart, Mark David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-06-01

    The INL is currently evolving the modeling and simulation (M&S) capability that will enable improved core operation as well as design and analysis of TREAT experiments. This M&S capability primarily uses MAMMOTH, a reactor physics application being developed under Multi-physics Object Oriented Simulation Environment (MOOSE) framework. MAMMOTH allows the coupling of a number of other MOOSE-based applications. This second year of work has been devoted to the generation of a deterministic reference solution for the full core, the preparation of anisotropic diffusion coefficients, the testing of the SPH equivalence method, and the improvement of the control rod modeling. In addition, this report includes the progress made in the modeling of the M8 core configuration and experiment vehicle since January of this year.

  6. CFD simulation on reactor flow mixing phenomena

    International Nuclear Information System (INIS)

    Kwon, T.S.; Kim, K.H.

    2016-01-01

    A pre-test calculation for multi-dimensional flow mixing in a reactor core and downcomer has been studied using a CFD code. To study the effects of Reactor Coolant Pump (RCP) and core zone on the boron mixing behaviors in a lower downcomer and core inlet, a 1/5-scale CFD model of flow mixing test facility for the APR+ reference plant was simulated. The flow paths of the 1/5-scale model were scaled down by the linear scaling method. The aspect ratio (L/D) of all flow paths was preserved to 1. To preserve a dynamic similarity, the ratio of Euler number was also preserved to 1. A single phase water flow at low pressure and temperature conditions was considered in this calculation. The calculation shows that the asymmetric effect driven by RCPs shifted the high velocity field to the failed pump's flow zone. The borated water flow zone at the core inlet was also shifted to the failed RCP side. (author)

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. 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)

  12. 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.

  13. 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)

  14. 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.

  15. 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.

  16. 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)

  17. 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)

  18. Integral Full Core Multi-Physics PWR Benchmark with Measured Data

    Energy Technology Data Exchange (ETDEWEB)

    Forget, Benoit; Smith, Kord; Kumar, Shikhar; Rathbun, Miriam; Liang, Jingang

    2018-04-11

    In recent years, the importance of modeling and simulation has been highlighted extensively in the DOE research portfolio with concrete examples in nuclear engineering with the CASL and NEAMS programs. These research efforts and similar efforts worldwide aim at the development of high-fidelity multi-physics analysis tools for the simulation of current and next-generation nuclear power reactors. Like all analysis tools, verification and validation is essential to guarantee proper functioning of the software and methods employed. The current approach relies mainly on the validation of single physic phenomena (e.g. critical experiment, flow loops, etc.) and there is a lack of relevant multiphysics benchmark measurements that are necessary to validate high-fidelity methods being developed today. This work introduces a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading and re-loading patterns. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from 58 instrumented assemblies. The benchmark description is now available online and has been used by many groups. However, much work remains to be done on the quantification of uncertainties and modeling sensitivities. This work aims to address these deficiencies and make this benchmark a true non-proprietary international benchmark for the validation of high-fidelity tools. This report details the BEAVRS uncertainty quantification for the first two cycle of operations and serves as the final report of the project.

  19. RANS-based CFD simulations of sodium fast reactor wire-wrapped pin bundles

    International Nuclear Information System (INIS)

    Pointer, W. D.; Thomas, J.; Fanning, T.; Fischer, P.; Siegel, A.; Smith, J.; Tokuhiro, A.

    2009-01-01

    In response to recent renewed interest in the development of advanced fast reactors, an effort is underway to develop a high-performance computational multi-physics simulation suite for the design and safety analysis of sodium cooled fast reactors. Within the multi-resolution thermal-hydraulics simulation component of this framework, high-resolution spectral large eddy simulation methods are used to improve turbulence models from coarser resolution Reynolds-averaged Navier-Stokes methods, and in turn, that data is used to improve or extend correlations used in traditional sub-channel tools. These ongoing studies provide the foundation for the development of the intermediate RANS-based resolution level. Prior work has focused on the benchmarking of flow field predictions on in 7-pin, 19-pin, and 37-pin fuel assemblies. The present work extends these studies to 217-pin assemblies in support of initial efforts to benchmark heat transfer predictions using the RANS models against conventional sub-channel models. In an effort to reduce the number of computational cells required to describe a 217-pin geometry, the effects of simplification of the geometric description of the contact point between the wire and the pin are investigated. The advantages of using polyhedral-based meshing methods rather than trimmed cell meshing methods have been demonstrated, and the effects of changes in axial mesh resolution in these meshes have been investigated. Results show that the geometric simplification has little impact on predicted flow fields, as does the use of a polyhedral mesh of comparable mesh density in place of the original trimmed cell mesh. While reducing axial mesh density has a notable impact on the velocity field, reducing predicted exchange velocities between adjacent subchannels by as much 25%, the impact on predicted temperature fields is negligible. (authors)

  20. 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)

  1. 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)

  2. 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)

  3. Multi-physics and multi-objective design of heterogeneous SFR core: development of an optimization method under uncertainty

    International Nuclear Information System (INIS)

    Ammar, Karim

    2014-01-01

    Since Phenix shutting down in 2010, CEA does not have Sodium Fast Reactor (SFR) in operating condition. According to global energetic challenge and fast reactor abilities, CEA launched a program of industrial demonstrator called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), a reactor with electric power capacity equal to 600 MW. Objective of the prototype is, in first to be a response to environmental constraints, in second demonstrates the industrial viability of SFR reactor. The goal is to have a safety level at least equal to 3. generation reactors. ASTRID design integrates Fukushima feedback; Waste reprocessing (with minor actinide transmutation) and it linked industry. Installation safety is the priority. In all cases, no radionuclide should be released into environment. To achieve this objective, it is imperative to predict the impact of uncertainty sources on reactor behaviour. In this context, this thesis aims to develop new optimization methods for SFR cores. The goal is to improve the robustness and reliability of reactors in response to existing uncertainties. We will use ASTRID core as reference to estimate interest of new methods and tools developed. The impact of multi-Physics uncertainties in the calculation of the core performance and the use of optimization methods introduce new problems: How to optimize 'complex' cores (i.e. associated with design spaces of high dimensions with more than 20 variable parameters), taking into account the uncertainties? What is uncertainties behaviour for optimization core compare to reference core? Taking into account uncertainties, optimization core are they still competitive? Optimizations improvements are higher than uncertainty margins? The thesis helps to develop and implement methods necessary to take into account uncertainties in the new generation of simulation tools. Statistical methods to ensure consistency of complex multi-Physics simulation results are also

  4. 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

  5. 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

  6. Code development for nuclear reactor simulation

    International Nuclear Information System (INIS)

    Chauliac, C.; Verwaerde, D.; Pavageau, O.

    2006-01-01

    Full text of publication follows: Since several years, CEA, EDF and FANP have developed several numerical codes which are currently used for nuclear industry applications and will be remain in use for the coming years. Complementary to this set of codes and in order to better meet the present and future needs, a new system is being developed through a joint venture between CEA, EDF and FANP, with a ten year prospect and strong intermediate milestones. The focus is put on a multi-scale and multi-physics approach enabling to take into account phenomena from microscopic to macroscopic scale, and to describe interactions between various physical fields such as neutronics (DESCARTES), thermal-hydraulics (NEPTUNE) and fuel behaviour (PLEIADES). This approach is based on a more rational design of the softwares and uses a common integration platform providing pre-processing, supervision of computation and post-processing. This paper will describe the overall system under development and present the first results obtained. (authors)

  7. 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

  8. 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)

  9. 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%

  10. 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.

  11. 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

  12. Liquid metal cooled experimental fast reactor simulator

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Braz Filho, Francisco; Borges, Eduardo M.; Rosa, Mauricio A.P.; Rocamora, Francisco; Hirdes, Viviane R.

    1997-01-01

    This paper is a continuation of the work that has been done in the area of fast reactor component dynamic analysis, as part of the REARA project at the IEAv/CTA-Brazil. A couple of preceding papers, presented in other meetings, introduced major concept design components of the REARA reactor. The components are set together in order to represent a full model of the power plant. Full model transient results will be presented, together with several parameters to help us to better establish the REARA experimental plant concept. (author). 8 refs., 6 figs., 3 tabs

  13. NEPTUNIX, a general program of simulation applied to nuclear reactors

    International Nuclear Information System (INIS)

    Bonnemay, A.; Dansac Bon, V.

    1978-01-01

    Most simulation languages admit an incremental description and involve explicit integration algorithms. NEPTUNIX is a simulation language directly admitting algebraic differential equations under an implicit form, and it involves a very efficient implicit integration method with variable step and order. NEPTUNIX is a tool used for building large systems models in the field of nuclear reactors [fr

  14. 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)

  15. 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)

  16. 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)

  17. 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.)

  18. 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)

  19. 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)

  20. Nonlinear punctual dynamic applied to simulation of PWR type reactors

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-01-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model to the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and C S M P using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (author)

  1. A study on naphtha catalytic reforming reactor simulation and analysis.

    Science.gov (United States)

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-06-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation unit data.

  2. A study on naphtha catalytic reforming reactor simulation and analysis

    OpenAIRE

    Liang, Ke-min; Guo, Hai-yan; Pan, Shi-wei

    2005-01-01

    A naphtha catalytic reforming unit with four reactors in series is analyzed. A physical model is proposed to describe the catalytic reforming radial flow reactor. Kinetics and thermodynamics equations are selected to describe the naphtha catalytic reforming reactions characteristics based on idealizing the complex naphtha mixture by representing the paraffin, naphthene, and aromatic groups by single compounds. The simulation results based above models agree very well with actual operation uni...

  3. Non-linear punctual kinetics applied to PWR reactors simulation

    International Nuclear Information System (INIS)

    Cysne, F.S.

    1978-11-01

    In order to study some kinds of nuclear reactor accidents, a simulation is made using the punctual kinetics model for the reactor core. The following integration methods are used: Hansen's method in which a linearization is made and CSMP using a variable interval fourth-order Runge Kutta method. The results were good and were compared with those obtained by the code Dinamica I which uses a finite difference integration method of backward kind. (Author) [pt

  4. Acoustic emission from fuel pellets in a simulated reactor environment

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Kennedy, C.R.; Reimann, K.J.

    1977-01-01

    Thermal-shock damage of nuclear reactor fuel pellets in a simulated reactor environment has been correlated with acoustic-emission data obtained from sensors placed on extensions of the electrical feedthroughs. Ringdown counts, rms output data, and event-location data has been acquired for experiments carried out with single pellets as well as multiple pellet stacks. These tests have shown that acoustic-emission monitoring can provide information indicating the onset and the extent of cracking

  5. 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)

  6. Simulation of Reforming Reactor Tube: Quantifying Catalyst Pellet's Effectiveness Factor

    OpenAIRE

    Da Cruz, Flavio Eduardo

    2016-01-01

    In this work, a consistent mathematical model to simulate a spherical catalytic pellet and a Packed-Bed Reactor (PBR) is develop. The Dusty Gas Model (DGM) is applied to the calculation of the diffusive fluxes in the porous media. Simulations are executed considering hydrogen production from steam methane reforming. Species’ diffusivities are calculated using data from literature as well as the values for tortuosity and porosity. The pellet simulation is performed considering mass, species, m...

  7. Validation of a 3D multi-physics model for unidirectional silicon solidification

    NARCIS (Netherlands)

    Simons, P.; Lankhorst, A.M.; Habraken, A.; Faber, A.J.; Tiuleanu, D.; Pingel, R.

    2012-01-01

    A model for transient movements of solidification fronts has been added to X-stream, an existing multi-physics simulation program for high temperature processes with flow and chemical reactions. The implementation uses an enthalpy formulation and works on fixed grids. First we show the results of a

  8. 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)

  9. 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.

  10. 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

  11. A Global Sensitivity Analysis Methodology for Multi-physics Applications

    Energy Technology Data Exchange (ETDEWEB)

    Tong, C H; Graziani, F R

    2007-02-02

    Experiments are conducted to draw inferences about an entire ensemble based on a selected number of observations. This applies to both physical experiments as well as computer experiments, the latter of which are performed by running the simulation models at different input configurations and analyzing the output responses. Computer experiments are instrumental in enabling model analyses such as uncertainty quantification and sensitivity analysis. This report focuses on a global sensitivity analysis methodology that relies on a divide-and-conquer strategy and uses intelligent computer experiments. The objective is to assess qualitatively and/or quantitatively how the variabilities of simulation output responses can be accounted for by input variabilities. We address global sensitivity analysis in three aspects: methodology, sampling/analysis strategies, and an implementation framework. The methodology consists of three major steps: (1) construct credible input ranges; (2) perform a parameter screening study; and (3) perform a quantitative sensitivity analysis on a reduced set of parameters. Once identified, research effort should be directed to the most sensitive parameters to reduce their uncertainty bounds. This process is repeated with tightened uncertainty bounds for the sensitive parameters until the output uncertainties become acceptable. To accommodate the needs of multi-physics application, this methodology should be recursively applied to individual physics modules. The methodology is also distinguished by an efficient technique for computing parameter interactions. Details for each step will be given using simple examples. Numerical results on large scale multi-physics applications will be available in another report. Computational techniques targeted for this methodology have been implemented in a software package called PSUADE.

  12. 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

  13. 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

  14. DESIGN of MICRO CANTILEVER BEAM for VAPOUR DETECTION USING COMSOL MULTI PHYSICS SOFTWARE

    OpenAIRE

    Sivacoumar R; Parvathy JM; Pratishtha Deep

    2015-01-01

    This paper gives an overview of micro cantilever beam of various shapes and materials for vapour detection. The design of micro cantilever beam, analysis and simulation is done for each shape. The simulation is done using COMSOL Multi physics software using structural mechanics and chemical module. The simulation results of applied force and resulting Eigen frequencies will be analyzed for different beam structures. The vapour analysis is done using flow cell that consists of chemical pill...

  15. 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.

  16. Engineering simulator applications to emergency preparedness at DOE reactor sites

    International Nuclear Information System (INIS)

    Beelman, R.J.

    1990-01-01

    This paper reports that since 1984 the Idaho National Engineering Laboratory (INEL) has conducted twenty-seven comprehensive emergency preparedness exercises at the U.S. Nuclear Regulatory Commission's (NRC) Headquarters Operations Center and Regional Incident Response Centers using the NRC's Nuclear Plant Analyzer (NPA), developed at the INEL, as an engineering simulator. The objective of these exercises has been to assist the NRC in upgrading its preparedness to provide technical support backup and oversight to U.S. commercial nuclear plant licensees during emergencies. With the current focus on Department of Energy (DOE) reactor operational safety and emergency preparedness, this capability is envisioned as a means of upgrading emergency preparedness at DOE production and test reactor sites such as the K-Reactor at Savannah River Laboratory (SRL) and the Advanced Test Reactor (ATR) at INEL

  17. Simulator platform for fast reactor operation and safety technology demonstration

    International Nuclear Information System (INIS)

    Vilim, R.B.; Park, Y.S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J.

    2012-01-01

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  18. Simulator platform for fast reactor operation and safety technology demonstration

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Park, Y. S.; Grandy, C.; Belch, H.; Dworzanski, P.; Misterka, J. (Nuclear Engineering Division)

    2012-07-30

    A simulator platform for visualization and demonstration of innovative concepts in fast reactor technology is described. The objective is to make more accessible the workings of fast reactor technology innovations and to do so in a human factors environment that uses state-of-the art visualization technologies. In this work the computer codes in use at Argonne National Laboratory (ANL) for the design of fast reactor systems are being integrated to run on this platform. This includes linking reactor systems codes with mechanical structures codes and using advanced graphics to depict the thermo-hydraulic-structure interactions that give rise to an inherently safe response to upsets. It also includes visualization of mechanical systems operation including advanced concepts that make use of robotics for operations, in-service inspection, and maintenance.

  19. Simulation of a porous ceramic membrane reactor for hydrogen production

    Energy Technology Data Exchange (ETDEWEB)

    Yu, W.; Ohmori, T.; Yamamoto, T.; Endo, A.; Nakaiwa, M.; Hayakawa, T. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Itoh, N. [National Inst. of Advanced Industrial Science and Technology, Tsukuba (Japan); Utsunomiya Univ. (Japan). Dept. of Applied Chemistry

    2005-08-01

    A systematic simulation study was performed to investigate the performance of a porous ceramic membrane reactor for hydrogen production by means of methane steam reforming. The results show that the methane conversions much higher than the corresponding equilibrium values can be achieved in the membrane reactor due to the selective removal of products from the reaction zone. The comparison of isothermal and non-isothermal model predictions was made. It was found that the isothermal assumption overestimates the reactor performance and the deviation of calculation results between the two models is subject to the operating conditions. The effects of various process parameters such as the reaction temperature, the reaction side pressure, the feed flow rate and the steam to methane molar feed ratio as well as the sweep gas flow rate and the operation modes, on the behavior of membrane reactor were analyzed and discussed. (author)

  20. Primary loop simulation of the SP-100 space nuclear reactor

    International Nuclear Information System (INIS)

    Borges, Eduardo M.; Braz Filho, Francisco A.; Guimaraes, Lamartine N.F.

    2011-01-01

    Between 1983 and 1992 the SP-100 space nuclear reactor development project for electric power generation in a range of 100 to 1000 kWh was conducted in the USA. Several configurations were studied to satisfy different mission objectives and power systems. In this reactor the heat is generated in a compact core and refrigerated by liquid lithium, the primary loops flow are controlled by thermoelectric electromagnetic pumps (EMTE), and thermoelectric converters produce direct current energy. To define the system operation point for an operating nominal power, it is necessary the simulation of the thermal-hydraulic components of the space nuclear reactor. In this paper the BEMTE-3 computer code is used to EMTE pump design performance evaluation to a thermalhydraulic primary loop configuration, and comparison of the system operation points of SP-100 reactor to two thermal powers, with satisfactory results. (author)

  1. 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

  2. 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.

  3. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Waintraub, M.

    1986-01-01

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  4. 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.

  5. Real-time numerical simulation with high efficiency for an experimental reactor system

    International Nuclear Information System (INIS)

    Ding Shuling; Li Fu; Li Sifeng; Chu Xinyuan

    2006-01-01

    The paper presents a systematic and efficient method for numerical real-time simulation of an experimental reactor. The reactor models were built based on the physical characteristics of the experimental reactor, and several real-time simulation approaches were discussed and compared in the paper. How to implement the real-time reactor simulation system in Windows platform for the sake of hardware-in-loop experiment for the reactor power control system was discussed. (authors)

  6. Thermohydraulic modeling and simulation of breeder reactors

    International Nuclear Information System (INIS)

    Agrawal, A.K.; Khatib-Rahbar, M.; Curtis, R.T.; Hetrick, D.L.; Girijashankar, P.V.

    1982-01-01

    This paper deals with the modeling and simulation of system-wide transients in LMFBRs. Unprotected events (i.e., the presumption of failure of the plant protection system) leading to core-melt are not considered in this paper. The existing computational capabilities in the area of protected transients in the US are noted. Various physical and numerical approximations that are made in these codes are discussed. Finally, the future direction in the area of model verification and improvements is discussed

  7. 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

  8. Level tracking in detailed reactor simulations

    Energy Technology Data Exchange (ETDEWEB)

    Aktas, B.; Mahaffy, J.H. [Pennsylvania State Univ., University Park, PA (United States)

    1995-09-01

    We introduce a useful test problem for judging the performance of reactor safety codes in situations where moving two-phase mixture levels are present. The test problem tracks a two-phase liquid level as it rises and then falls back to its original position. Pure air exists above the level, and a low void air-water mixture is below the level. Conditions are subcooled and isothermal to remove complications resulting from failures of interfacial heat transfer packages to properly account for the level. Comparisons are made between the performance of current versions of CATHARE, RELAP5, TRAC-BF1, and TRAC-PF1. These system codes are based on finite-difference methods with a fixed, Eulerian staggered grid in space. When a partially filled cell with a mixture level discontinuity becomes the donor cell, the sharp changes in fluid properties across the interface results in numerical oscillations of various terms. Furthermore, the cell-to-cell convection of mass, momentum and energy are inaccurately predicted nearby a mixture level. To adequately model moving mixture levels, an efficient discontinuity tracking method for the finite-difference Eulerian approximations is described. This model had been implemented in the TRAC-BWR code for the two-phase mixture level tracking since the TRAC-BD1 Version (released April 1984). The result of the test problem run by the current version of TRAC-BF1/MOD1 with the mixture level tracking model shows some peculiar behavior of the variables such as velocities, pressures and interfacial terms. A systematic approach to improving performance of the tracking method is described. Implementing this approach in TRAC-BF1/MOD1 has shown a major improvement in the results.

  9. 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.

  10. Computer codes for simulation of Angra 1 reactor steam generator

    International Nuclear Information System (INIS)

    Pinto, A.C.

    1978-01-01

    A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

  11. Tritium behavior in the Caisson, a simulated fusion reactor room

    International Nuclear Information System (INIS)

    Hayashi, Takumi; Kobayashi, Kazuhiro; Iwai, Yasunori; Yamada, Masayuki; Suzuki, Takumi; O'hira, Shigeru; Nakamura, Hirofumi; Shu, Weimin; Yamanishi, Toshihiko; Kawamura, Yoshinori; Isobe, Kanetsugu; Konishi, Satoshi; Nishi, Masataka

    2000-01-01

    In order to confirm tritium confinement ability in the deuterium-tritium (DT) fusion reactor, intentional tritium release experiments have been started in a specially fabricated test stand called 'Caisson', at Tritium Process Laboratory in Japan Atomic Energy Research Institute. The Caisson is a stainless steel leak-tight vessel of 12 m 3 , simulating a reactor room or a tritium handling room. In the first stage experiments, about 260 MBq of pure tritium was put into the Caisson under simulated constant ventilation of four times air exchanges per h. The tritium mixing and migration in the Caisson was investigated with tritium contamination measurement and detritiation behavior measurement. The experimental tritium migration and removal behavior was almost perfectly reproduced and could almost be simulated by a three-dimensional flow analysis code

  12. 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.

  13. 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

  14. A JAVA applet to simulate a CANDU reactor

    International Nuclear Information System (INIS)

    Varin, E.; Desarmenien, J.

    2004-01-01

    Here we present a CANDU nuclear power plant simulator, directly available on a web page. The developed applet has two mains objectives: to expose the CANDU technology to a large public on the internet; and to construct a realistic simulator to be used as a pedagogical tool for nuclear introduction to high school or under-graduate students. The neutronic behavior and control algorithms of the reactor are simulated. Java programming language enables a very flexible environment for public information and user interaction with the plant. Examples of shutdown and power maneuver are explained. (author)

  15. Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; J. Michael Doster; Alan Rominger

    2012-09-01

    Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

  16. Simulation of pulsed accidental energy release in a reactor core

    International Nuclear Information System (INIS)

    Ryshanskii, V.A.; Ivanov, A.G.; Uskov, A.A.

    1995-01-01

    At the present time the strength of the load-bearing members of VVER and fast reactors during a hypothetical accident is ordinarily investigated in model experiments [1]. A power burst during an accident is simulated by a nonnuclear exothermal reaction in water, which simulates the coolant and fills the model. The problem is to make the correct choice of the simulator of the accidental energy burst as an effective (i.e., sufficiently high working capacity) source of dangerous loads, corresponding to the conditions of an accident. What factors and parameters determine the energy release? The answers to these questions are contradictory

  17. Computation of Thermodynamic Equilibria Pertinent to Nuclear Materials in Multi-Physics Codes

    Science.gov (United States)

    Piro, Markus Hans Alexander

    Nuclear energy plays a vital role in supporting electrical needs and fulfilling commitments to reduce greenhouse gas emissions. Research is a continuing necessity to improve the predictive capabilities of fuel behaviour in order to reduce costs and to meet increasingly stringent safety requirements by the regulator. Moreover, a renewed interest in nuclear energy has given rise to a "nuclear renaissance" and the necessity to design the next generation of reactors. In support of this goal, significant research efforts have been dedicated to the advancement of numerical modelling and computational tools in simulating various physical and chemical phenomena associated with nuclear fuel behaviour. This undertaking in effect is collecting the experience and observations of a past generation of nuclear engineers and scientists in a meaningful way for future design purposes. There is an increasing desire to integrate thermodynamic computations directly into multi-physics nuclear fuel performance and safety codes. A new equilibrium thermodynamic solver is being developed with this matter as a primary objective. This solver is intended to provide thermodynamic material properties and boundary conditions for continuum transport calculations. There are several concerns with the use of existing commercial thermodynamic codes: computational performance; limited capabilities in handling large multi-component systems of interest to the nuclear industry; convenient incorporation into other codes with quality assurance considerations; and, licensing entanglements associated with code distribution. The development of this software in this research is aimed at addressing all of these concerns. The approach taken in this work exploits fundamental principles of equilibrium thermodynamics to simplify the numerical optimization equations. In brief, the chemical potentials of all species and phases in the system are constrained by estimates of the chemical potentials of the system

  18. Modelling of thermalhydraulics and reactor physics in simulators

    International Nuclear Information System (INIS)

    Miettinen, J.

    1994-01-01

    The evolution of thermalhydraulic analysis methods for analysis and simulator purposes has brought closer the thermohydraulic models in both application areas. In large analysis codes like RELAP5, TRAC, CATHARE and ATHLET the accuracy for calculating complicated phenomena has been emphasized, but in spite of large development efforts many generic problems remain unsolved. For simulator purposes fast running codes have been developed and these include only limited assessment efforts. But these codes have more simulator friendly features than large codes, like portability and modular code structure. In this respect the simulator experiences with SMABRE code are discussed. Both large analysis codes and special simulator codes have their advances in simulator applications. The evolution of reactor physical calculation methods in simulator applications has started from simple point kinetic models. For analysis purposes accurate 1-D and 3-D codes have been developed being capable for fast and complicated transients. For simulator purposes capability for simulation of instruments has been emphasized, but the dynamic simulation capability has been less significant. The approaches for 3-dimensionality in simulators requires still quite much development, before the analysis accuracy is reached. (orig.) (8 refs., 2 figs., 2 tabs.)

  19. 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.

  20. 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

  1. 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.)

  2. 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)

  3. 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)

  4. Simulation of MILD combustion using Perfectly Stirred Reactor model

    KAUST Repository

    Chen, Z.

    2016-07-06

    A simple model based on a Perfectly Stirred Reactor (PSR) is proposed for moderate or intense low-oxygen dilution (MILD) combustion. The PSR calculation is performed covering the entire flammability range and the tabulated chemistry approach is used with a presumed joint probability density function (PDF). The jet, in hot and diluted coflow experimental set-up under MILD conditions, is simulated using this reactor model for two oxygen dilution levels. The computed results for mean temperature, major and minor species mass fractions are compared with the experimental data and simulation results obtained recently using a multi-environment transported PDF approach. Overall, a good agreement is observed at three different axial locations for these comparisons despite the over-predicted peak value of CO formation. This suggests that MILD combustion can be effectively modelled by the proposed PSR model with lower computational cost.

  5. SIMODIS - a software package for simulating nuclear reactor components

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Borges, Eduardo M.

    2000-01-01

    In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C ++ . From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

  6. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    Lautard, J.J.; Magnaud, C.; Moreau, F.; Baudron, A.M.

    1999-01-01

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  7. Component and system simulation models for High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

  8. 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).

  9. Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

    Energy Technology Data Exchange (ETDEWEB)

    Le Pallec, J. C.; Crouzet, N.; Bergeaud, V.; Delavaud, C. [CEA/DEN/DM2S, CEA/Saclay, 91191 Gif sur Yvette Cedex (France)

    2012-07-01

    The control of uncertainties in the field of reactor physics and their propagation in best-estimate modeling are a major issue in safety analysis. In this framework, the CEA develops a methodology to perform multi-physics simulations including uncertainties analysis. The present paper aims to present and apply this methodology for the analysis of an accidental situation such as REA (Rod Ejection Accident). This accident is characterized by a strong interaction between the different areas of the reactor physics (neutronic, fuel thermal and thermal hydraulic). The modeling is performed with CRONOS2 code. The uncertainties analysis has been conducted with the URANIE platform developed by the CEA: For each identified response from the modeling (output) and considering a set of key parameters with their uncertainties (input), a surrogate model in the form of a neural network has been produced. The set of neural networks is then used to carry out a sensitivity analysis which consists on a global variance analysis with the determination of the Sobol indices for all responses. The sensitivity indices are obtained for the input parameters by an approach based on the use of polynomial chaos. The present exercise helped to develop a methodological flow scheme, to consolidate the use of URANIE tool in the framework of parallel calculations. Finally, the use of polynomial chaos allowed computing high order sensitivity indices and thus highlighting and classifying the influence of identified uncertainties on each response of the analysis (single and interaction effects). (authors)

  10. Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

    International Nuclear Information System (INIS)

    Salko, Robert K.; Schmidt, Rodney C.; Avramova, Maria N.

    2015-01-01

    Highlights: • COBRA-TF was adopted by the Consortium for Advanced Simulation of LWRs. • We have improved code performance to support running large-scale LWR simulations. • Code optimization has led to reductions in execution time and memory usage. • An MPI parallelization has reduced full-core simulation time from days to minutes. - Abstract: This paper describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis. A set of serial code optimizations—including fixing computational inefficiencies, optimizing the numerical approach, and making smarter data storage choices—are first described and shown to reduce both execution time and memory usage by about a factor of ten. Next, a “single program multiple data” parallelization strategy targeting distributed memory “multiple instruction multiple data” platforms utilizing domain decomposition is presented. In this approach, data communication between processors is accomplished by inserting standard Message-Passing Interface (MPI) calls at strategic points in the code. The domain decomposition approach implemented assigns one MPI process to each fuel assembly, with each domain being represented by its own CTF input file. The creation of CTF input files, both for serial and parallel runs, is also fully automated through use of a pressurized water reactor (PWR) pre-processor utility that uses a greatly simplified set of user input compared with the traditional CTF input. To run CTF in

  11. 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.

  12. 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

  13. Modeling and simulation of pressurized water reactor power plant

    International Nuclear Information System (INIS)

    Wang, S.J.

    1983-01-01

    Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

  14. 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)

  15. Tightly coupled simulation of nuclear reactor transients with artificial intelligence

    International Nuclear Information System (INIS)

    Makowitz, H.; Ragheb, M.; Laats, E.T.; Bray, M.A.

    1985-01-01

    The authors' current efforts are directed toward exploring new avenues of research in simulation of nuclear reactor kinetics transients with artificial intelligence (AI). Being examined are advanced graphics systems such as the Nuclear Plant Analyzer designed to run in parallel with the RELAP5 code, faster than real-time best-estimate simulations, the utilization of the multi-CPU super computers, and simulation as knowledge by attempting to develop new assessment methodologies for artificial intelligence systems and their associated interfaces. This new and fertile area of research should be viewed by the educational and university community as an indication of the future possibilities for AI developments in a number of academic and engineering disciplines

  16. 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

  17. 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

  18. Computer simulation of multi-elemental fusion reactor materials

    International Nuclear Information System (INIS)

    Voertler, K.

    2011-01-01

    Thermonuclear fusion is a sustainable energy solution, in which energy is produced using similar processes as in the sun. In this technology hydrogen isotopes are fused to gain energy and consequently to produce electricity. In a fusion reactor hydrogen isotopes are confined by magnetic fields as ionized gas, the plasma. Since the core plasma is millions of degrees hot, there are special needs for the plasma-facing materials. Moreover, in the plasma the fusion of hydrogen isotopes leads to the production of high energetic neutrons which sets demanding abilities for the structural materials of the reactor. This thesis investigates the irradiation response of materials to be used in future fusion reactors. Interactions of the plasma with the reactor wall leads to the removal of surface atoms, migration of them, and formation of co-deposited layers such as tungsten carbide. Sputtering of tungsten carbide and deuterium trapping in tungsten carbide was investigated in this thesis. As the second topic the primary interaction of the neutrons in the structural material steel was examined. As model materials for steel iron chromium and iron nickel were used. This study was performed theoretically by the means of computer simulations on the atomic level. In contrast to previous studies in the field, in which simulations were limited to pure elements, in this work more complex materials were used, i.e. they were multi-elemental including two or more atom species. The results of this thesis are in the microscale. One of the results is a catalogue of atom species, which were removed from tungsten carbide by the plasma. Another result is e.g. the atomic distributions of defects in iron chromium caused by the energetic neutrons. These microscopic results are used in data bases for multiscale modelling of fusion reactor materials, which has the aim to explain the macroscopic degradation in the materials. This thesis is therefore a relevant contribution to investigate the

  19. 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.)

  20. A simulator-based nuclear reactor emergency response training exercise.

    Science.gov (United States)

    Waller, Edward; Bereznai, George; Shaw, John; Chaput, Joseph; Lafortune, Jean-Francois

    Training offsite emergency response personnel basic awareness of onsite control room operations during nuclear power plant emergency conditions was the primary objective of a week-long workshop conducted on a CANDU® virtual nuclear reactor simulator available at the University of Ontario Institute of Technology, Oshawa, Canada. The workshop was designed to examine both normal and abnormal reactor operating conditions, and to observe the conditions in the control room that may have impact on the subsequent offsite emergency response. The workshop was attended by participants from a number of countries encompassing diverse job functions related to nuclear emergency response. Objectives of the workshop were to provide opportunities for participants to act in the roles of control room personnel under different reactor operating scenarios, providing a unique experience for participants to interact with the simulator in real-time, and providing increased awareness of control room operations during accident conditions. The ability to "pause" the simulator during exercises allowed the instructors to evaluate and critique the performance of participants, and to provide context with respect to potential offsite emergency actions. Feedback from the participants highlighted (i) advantages of observing and participating "hands-on" with operational exercises, (ii) their general unfamiliarity with control room operational procedures and arrangements prior to the workshop, (iii) awareness of the vast quantity of detailed control room procedures for both normal and transient conditions, and (iv) appreciation of the increased workload for the operators in the control room during a transient from normal operations. Based upon participant feedback, it was determined that the objectives of the training had been met, and that future workshops should be conducted.

  1. 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

  2. LOCA simulation in the NRU reactor: materials test-1

    International Nuclear Information System (INIS)

    Russcher, G.E.; Marshall, R.K.; Hesson, G.M.; Wildung, N.J.; Rausch, W.N.

    1981-10-01

    A simulated loss-of-coolant accident was performed with a full-length test bundle of pressurized water reactor fuel rods. This second experiment of the program produced peak fuel cladding temperatures of 1148K (1607 0 F) and resulted in six ruptured fuel rods. Test data and initial results from the experiment are presented here in the form of photographs and graphical summaries. These results are also compared with the preceding prototypic thermal-hydraulic test results and with computer model test predictions

  3. 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

  4. The denitration of simulated fast reactor highly active liquor waste

    International Nuclear Information System (INIS)

    Saum, C.J.; Ford, L.H.; Blatts, N.

    1981-01-01

    A short series of tests have been made with simulated HAL containing representative concentrations of palladium and phosphate ion. The information obtained has been confirmed in a small scale continuous denitration plant. These cases of four stirred pot reactors arranged in cascade. One possible advantage of this plant would be the low mean acidity in the first stage compared to the feed material which would limit to some extent the violence of the reaction. This would lead to a lower rate of gas evolution and may permit operation even with liquors where foaming is a problem. (DG)

  5. Simulations for the transmutation of nuclear wastes with hybrid reactors

    International Nuclear Information System (INIS)

    Vuillier, St.

    1998-06-01

    A Monte Carlo simulation, devoted to the spallation, has been built in the framework of the hybrid systems proposed for the nuclear wastes incineration. This system GSPARTE, described the reactions evolution. It takes into account and improves the nuclear codes and the low and high energy particles transport in the GEANT code environment, adapted to the geometry of the hybrid reactors. Many applications and abacus useful for the wastes transmutation, have been realized with this system: production of thick target neutrons, source definition, material damages. (A.L.B.)

  6. 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)

  7. Parallel linear solvers for simulations of reactor thermal hydraulics

    International Nuclear Information System (INIS)

    Yan, Y.; Antal, S.P.; Edge, B.; Keyes, D.E.; Shaver, D.; Bolotnov, I.A.; Podowski, M.Z.

    2011-01-01

    The state-of-the-art multiphase fluid dynamics code, NPHASE-CMFD, performs multiphase flow simulations in complex domains using implicit nonlinear treatment of the governing equations and in parallel, which is a very challenging environment for the linear solver. The present work illustrates how the Portable, Extensible Toolkit for Scientific Computation (PETSc) and scalable Algebraic Multigrid (AMG) preconditioner from Hypre can be utilized to construct robust and scalable linear solvers for the Newton correction equation obtained from the discretized system of governing conservation equations in NPHASE-CMFD. The overall long-tem objective of this work is to extend the NPHASE-CMFD code into a fully-scalable solver of multiphase flow and heat transfer problems, applicable to both steady-state and stiff time-dependent phenomena in complete fuel assemblies of nuclear reactors and, eventually, the entire reactor core (such as the Virtual Reactor concept envisioned by CASL). This campaign appropriately begins with the linear algebraic equation solver, which is traditionally a bottleneck to scalability in PDE-based codes. The computational complexity of the solver is usually superlinear in problem size, whereas the rest of the code, the “physics” portion, usually has its complexity linear in the problem size. (author)

  8. A hybrid computer simulation of reactor spatial dynamics

    International Nuclear Information System (INIS)

    Hinds, H.W.

    1977-08-01

    The partial differential equations describing the one-speed spatial dynamics of thermal neutron reactors were converted to a set of ordinary differential equations, using finite-difference approximations for the spatial derivatives. The variables were then normalized to a steady-state reference condition in a novel manner, to yield an equation set particularly suitable for implementation on a hybrid computer. One Applied Dynamics AD/FIVE analog-computer console is capable of solving, all in parallel, up to 30 simultaneous differential equations. This corresponds roughly to eight reactor nodes, each with two active delayed-neutron groups. To improve accuracy, an increase in the number of nodes is usually required. Using the Hsu-Howe multiplexing technique, an 8-node, one-dimensional module was switched back and forth between the left and right halves of the reactor, to simulate a 16-node model, also in one dimension. These two versions (8 or 16 nodes) of the model were tested on benchmark problems of the loss-of-coolant type, which were also solved using the digital code FORSIM, with two energy groups and 26 nodes. Good agreement was obtained between the two solution techniques. (author)

  9. 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)

  10. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    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 numericaly. 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 resistence added to the film coeficient. 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) [pt

  11. 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

  12. 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

  13. 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

  14. Analysis of dynamic stability and safety of reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    In order to enable qualitative analysis of dynamic properties of reactors RA and RB, mathematical models of these reactors were formulated and adapted for solution on analog computer. This report contains basic assessments for creating the model and complete equations for each reactor. Model was used to analyse three possible accidents at the RA reactor and possible hypothetical accidents at the RB reactor

  15. Reactor controller design using genetic algorithms with simulated annealing

    International Nuclear Information System (INIS)

    Erkan, K.; Buetuen, E.

    2000-01-01

    This chapter presents a digital control system for ITU TRIGA Mark-II reactor using genetic algorithms with simulated annealing. The basic principles of genetic algorithms for problem solving are inspired by the mechanism of natural selection. Natural selection is a biological process in which stronger individuals are likely to be winners in a competing environment. Genetic algorithms use a direct analogy of natural evolution. Genetic algorithms are global search techniques for optimisation but they are poor at hill-climbing. Simulated annealing has the ability of probabilistic hill-climbing. Thus, the two techniques are combined here to get a fine-tuned algorithm that yields a faster convergence and a more accurate search by introducing a new mutation operator like simulated annealing or an adaptive cooling schedule. In control system design, there are currently no systematic approaches to choose the controller parameters to obtain the desired performance. The controller parameters are usually determined by test and error with simulation and experimental analysis. Genetic algorithm is used automatically and efficiently searching for a set of controller parameters for better performance. (orig.)

  16. 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

  17. Simulation of mixing effects in a VVER-1000 reactor

    International Nuclear Information System (INIS)

    Ulrich Bieder; Gauthier Fauchet; Sylvie Betin; Nikola Kolev; Dimitar Popov

    2005-01-01

    Full text of publication follows: The work presented has been performed in the framework of the OECD/NEA thermalhydraulic benchmark V1000CT-2. This benchmark is related to fluid mixing in the reactor vessel during a MSLB accident scenario in a VVER-1000 reactor. The purpose of the first exercise is to test the capability of CFD codes to represent the coolant mixing in the reactor vessel, in particular in the downcomer and the lower plenum. Coolant mixing in a VVER-1000 V320 reactor was investigated in plant experiments during the commissioning of Kozloduy Unit 5 and 6. Starting from nearly symmetric states, asymmetric loop operation in different combinations was caused by disturbing the steam flow from one or more steam generators. Non-uniform and asymmetric loop flow mixing in the reactor vessel has been observed in the event of asymmetric loop operation. For certain flow patterns there is a shift (swirl) of the main loop flows with respect to the cold leg axes. This azimuthal shift as well as mixing coefficients from cold legs to the fuel assembly inlets have been measured. The presented reference problem is a pure TH problem with given boundary conditions and power distributions. During a stabilization phase, the thermal power of the reactor was 281 MW i.e. 9.36% of the nominal power according to primary balance. Then, a transient was initiated by closing the steam isolation valve of the steam generator one (SG-1) and isolating SG-1 from feed water. The coolant temperature in the cold and hot legs of Loop no 1 rose by 13-13.5 C. After about 20 minutes a stabilized state was reached which is considered as 'final state'. This final state has been analysed with the Trio-U code. Trio-U is a CFD code developed by the CEA Grenoble, aimed to supply an efficient computational tool to simulate transient thermalhydraulic mono-phase turbulent flows encountered in nuclear systems as well as in industrial processes. For the presented study, a LES approach was used. Therefore

  18. 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.

  19. Development of dynamic simulation code for fuel cycle fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1999-02-01

    A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  20. Simulation of pressurized water reactor in accidental state

    International Nuclear Information System (INIS)

    Chakir, E.

    1994-01-01

    The aim of this work is to develop the 1300 MWe 4 loops 'PWR' simulator called 'SATRAPE', witch the adopted physics modelisation allows a simplified neutronic calculation, and focus essentially on the reactor thermal hydraulic behavior in the case of the following accidents: - Loss of Coolant Accident (LOCA). - Steam Generator Tube Failure (SGTF). - Steam Line Break (SLB). In case of the 'LOCA' or 'SLB' accident, this modelisation enables the calculation of the pressure and the temperature in the containment building, and also the debit of the released dose in this latter in case of the 'LOCA' accident. The adopted models are relatively simple so as to allow an explicit resolve. In SATRAPE, two graphical interfaces enables to launch orders, whereas the other permits to visualize, the principal state variables of installations. The results obtained show a very good consistency with the envisaged commonly scenario at the time of the considered accidents. 33 refs., 52 figs., 1 tab. (author)

  1. Simulation of styrene polymerization reactors: kinetic and thermodynamic modeling

    Directory of Open Access Journals (Sweden)

    A. S. Almeida

    2008-06-01

    Full Text Available A mathematical model for the free radical polymerization of styrene is developed to predict the steady-state and dynamic behavior of a continuous process. Special emphasis is given for the kinetic and thermodynamic models, where the most sensitive parameters were estimated using data from an industrial plant. The thermodynamic model is based on a cubic equation of state and a mixing rule applied to the low-pressure vapor-liquid equilibrium of polymeric solutions, suitable for modeling the auto-refrigerated polymerization reactors, which use the vaporization rate to remove the reaction heat from the exothermic reactions. The simulation results show the high predictive capability of the proposed model when compared with plant data for conversion, average molecular weights, polydispersity, melt flow index, and thermal properties for different polymer grades.

  2. 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.)

  3. 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)

  4. 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

  5. Test Results from a Direct Drive Gas Reactor Simulator Coupled to a Brayton Power Conversion Unit

    Science.gov (United States)

    Hervol, David S.; Briggs, Maxwell H.; Owen, Albert K.; Bragg-Sitton, Shannon M.; Godfroy, Thomas J.

    2010-01-01

    Component level testing of power conversion units proposed for use in fission surface power systems has typically been done using relatively simple electric heaters for thermal input. These heaters do not adequately represent the geometry or response of proposed reactors. As testing of fission surface power systems transitions from the component level to the system level it becomes necessary to more accurately replicate these reactors using reactor simulators. The Direct Drive Gas-Brayton Power Conversion Unit test activity at the NASA Glenn Research Center integrates a reactor simulator with an existing Brayton test rig. The response of the reactor simulator to a change in Brayton shaft speed is shown as well as the response of the Brayton to an insertion of reactivity, corresponding to a drum reconfiguration. The lessons learned from these tests can be used to improve the design of future reactor simulators which can be used in system level fission surface power tests.

  6. 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

  7. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J; Park, W S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  8. Numerical simulations of subcritical reactor kinetics in thermal hydraulic transient phases

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, J.; Park, W. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute (KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons, external neutrons form spallation reactions are essentially required for operating the reactor in its steady state. Furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance to the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases. 5 refs., 8 figs. (Author)

  9. ALE3D: An Arbitrary Lagrangian-Eulerian Multi-Physics Code

    Energy Technology Data Exchange (ETDEWEB)

    Noble, Charles R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anderson, Andrew T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Barton, Nathan R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bramwell, Jamie A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Capps, Arlie [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chang, Michael H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chou, Jin J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dawson, David M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Diana, Emily R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dunn, Timothy A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Faux, Douglas R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Fisher, Aaron C. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Greene, Patrick T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Heinz, Ines [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kanarska, Yuliya [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Khairallah, Saad A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Liu, Benjamin T. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Margraf, Jon D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nichols, Albert L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Nourgaliev, Robert N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Puso, Michael A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Reus, James F. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Robinson, Peter B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Shestakov, Alek I. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Solberg, Jerome M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Taller, Daniel [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Tsuji, Paul H. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Christopher A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); White, Jeremy L. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-05-23

    ALE3D is a multi-physics numerical simulation software tool utilizing arbitrary-Lagrangian- Eulerian (ALE) techniques. The code is written to address both two-dimensional (2D plane and axisymmetric) and three-dimensional (3D) physics and engineering problems using a hybrid finite element and finite volume formulation to model fluid and elastic-plastic response of materials on an unstructured grid. As shown in Figure 1, ALE3D is a single code that integrates many physical phenomena.

  10. Acceleration methods for multi-physics compressible flow

    Science.gov (United States)

    Peles, Oren; Turkel, Eli

    2018-04-01

    In this work we investigate the Runge-Kutta (RK)/Implicit smoother scheme as a convergence accelerator for complex multi-physics flow problems including turbulent, reactive and also two-phase flows. The flows considered are subsonic, transonic and supersonic flows in complex geometries, and also can be either steady or unsteady flows. All of these problems are considered to be a very stiff. We then introduce an acceleration method for the compressible Navier-Stokes equations. We start with the multigrid method for pure subsonic flow, including reactive flows. We then add the Rossow-Swanson-Turkel RK/Implicit smoother that enables performing all these complex flow simulations with a reasonable CFL number. We next discuss the RK/Implicit smoother for time dependent problem and also for low Mach numbers. The preconditioner includes an intrinsic low Mach number treatment inside the smoother operator. We also develop a modified Roe scheme with a corresponding flux Jacobian matrix. We then give the extension of the method for real gas and reactive flow. Reactive flows are governed by a system of inhomogeneous Navier-Stokes equations with very stiff source terms. The extension of the RK/Implicit smoother requires an approximation of the source term Jacobian. The properties of the Jacobian are very important for the stability of the method. We discuss what the chemical physics theory of chemical kinetics tells about the mathematical properties of the Jacobian matrix. We focus on the implication of the Le-Chatelier's principle on the sign of the diagonal entries of the Jacobian. We present the implementation of the method for turbulent flow. We use a two RANS turbulent model - one equation model - Spalart-Allmaras and a two-equation model - k-ω SST model. The last extension is for two-phase flows with a gas as a main phase and Eulerian representation of a dispersed particles phase (EDP). We present some examples for such flow computations inside a ballistic evaluation

  11. Thermo-hydraulic simulations of the experimental fast reactor core

    International Nuclear Information System (INIS)

    Silveira Luz, M. da; Braz Filho, F.A.; Borges, E.M.

    1985-01-01

    A study of the core and performance of metallic fuel of the experimental fast reactor, from the thermal-hydraulic point of view, was carried out employing the COBRA IV-I code. The good safety characteristics of this reactor and the feasibility of using metallic fuel in experimental fast reactor were demonstrated. (Author) [pt

  12. An evaluation of control rod motion simulator of research reactor

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

  13. Evaluation of control rod motion simulator research reactors

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator has been carried out testing of the reactor control rod using a servomotor. Reactor control rod motion at any point should be in the right position, one of the motors that can move in a precise and correct the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servomotor function test for motor work to ensure the performance of the appliance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage Vout nets at 24 V, 6.5 A with 12 Ω load deviation obtained V0 = V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125%, next to the breakdown voltage Vout nets at 12V, 4.2 A with a 6 Ω load deviation obtained V0 = V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on Vout 24 V, 4.5 A with 12 Ω load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%. (author)

  14. Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

    International Nuclear Information System (INIS)

    Guyot, Maxime

    2014-01-01

    This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

  15. 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)

  16. 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)

  17. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

    2013-01-01

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  18. 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

  19. 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.

  20. Analysis of dynamic stability and safety of the reactor system by reactor simulator

    International Nuclear Information System (INIS)

    Raisic, N.

    1963-11-01

    This document defines the approximations done for establishing a mathematical model of a reactor. Since the model should be used for safety analysis, it was important to choose a mathematical model less stable than the reactor itself. The analysis was performed on the analog computer RAS. Results obtained and conclusions concerned with three possible reactor accidents are presented [sr

  1. The double chooz experiment: simulation of reactor antineutrino spectra

    International Nuclear Information System (INIS)

    Mueller, T.

    2010-01-01

    The Double Chooz experiment aims to study the oscillations of electron antineutrinos produced by the Chooz nuclear power station, located in France, in the Ardennes region. It will lead to an unprecedented accuracy on the value of the mixing angle θ 13 . Improving the current knowledge on this parameter, given by the Chooz experiment, requires a reduction of both statistical and systematic errors, that is to say not only observing a large data sample, but also controlling the experimental uncertainties involved in the production and detection of electron antineutrinos. The use of two identical detectors will cancel most of the experimental systematic uncertainties limiting the sensitivity to the value of the mixing angle. We present in this thesis, simulations of reactor antineutrino spectra that were carried out in order to control the sources of systematic uncertainty related to the production of these particles by the plant. We also discuss our work on controlling the normalization error of the experiment through the precise determination of the number of target protons by a weighing measurement and through the study of the fiducial volume of the detectors which requires an accurate modeling of neutron physics. After three years of data taking with two detectors, Double Chooz will be able to disentangle an oscillation signal for sin 2 2θ 13 ≥ 0.05 (at 3σ) or, if no oscillations are observed, to put a limit of sin 2 2θ 13 ≤ 0.03 at 90% C.L. (author) [fr

  2. Hydride blister formation simulation in Candu type reactors

    International Nuclear Information System (INIS)

    Otero, D.; Bollini, C.; Sangregorio, D.

    1992-01-01

    We have developed a computer code for the probability study of hydride blister formation in pressure tubes named BLIFO. The basic hypothesis of the model are: the pressure tube is divided into five areas according to the existence of four garter springs. For each area the probability of blister formation is the probability of the hydrogen content exceeding a critical threshold when contact tube is present; the probability of a blister in a tube is the OR combination of the probabilities of a blister in each area; the tube contact is a function of the garter springs location, and the time; the critical hydrogen threshold is sorted over the areas within the pressure tube; hydrogen pick-up rate was sorted with a Gaussian distribution; the initial hydrogen content values for each tube were measured before the ensamble and they are used in the code. For Embalse evaluation, we build up a subroutine that simulate Gaussian distribution using the parameters of a typical nuclear power Candu reactor garter spring distribution. (author)

  3. 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)

  4. 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)

  5. Core followup studies of the Tarapur Reactors with the three dimensional BWR simulator COMTEG

    Energy Technology Data Exchange (ETDEWEB)

    Dwivedi, S. R.; Jagannathan, V.; Mohanakrishnan, P.; Srinivasan, K. R.; Rastogi, B. P.

    1976-07-01

    Both the units of the Tarapur Atomic Power Station started operation in the year 1969. Since then, these units have completed three cycles. For efficient operation and fuel management of these reactors, a three dimensional BWR simulator COMETG has been developed. The reactors are closely being followed using the simulator. The detailed analyses for cycle 3/4 operation of both the units are described in the paper. The results show very good agreement between calculated and measured values. It is concluded that reactor core behaviour could be predicted in a satisfactory manner with the core simulator COMETG.

  6. Integrated approach for fusion multi-physics coupled analyses based on hybrid CAD and mesh geometries

    Energy Technology Data Exchange (ETDEWEB)

    Qiu, Yuefeng, E-mail: yuefeng.qiu@kit.edu; Lu, Lei; Fischer, Ulrich

    2015-10-15

    Highlights: • Integrated approach for neutronics, thermal and structural analyses was developed. • MCNP5/6, TRIPOLI-4 were coupled with CFX, Fluent and ANSYS Workbench. • A novel meshing approach has been proposed for describing MC geometry. - Abstract: Coupled multi-physics analyses on fusion reactor devices require high-fidelity neutronic models, and flexible, accurate data exchanging between various calculation codes. An integrated coupling approach has been developed to enable the conversion of CAD, mesh, or hybrid geometries for Monte Carlo (MC) codes MCNP5/6, TRIPOLI-4, and translation of nuclear heating data for CFD codes Fluent, CFX and structural mechanical software ANSYS Workbench. The coupling approach has been implemented based on SALOME platform with CAD modeling, mesh generation and data visualization capabilities. A novel meshing approach has been developed for generating suitable meshes for MC geometry descriptions. The coupling approach has been concluded to be reliable and efficient after verification calculations of several application cases.

  7. 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

  8. Fuel-management simulations for once-through thorium fuel cycle in CANDU reactors

    International Nuclear Information System (INIS)

    Chan, P.S.W.; Boczar, P.G.; Ellis, R.J.; Ardeshiri, F.

    1999-01-01

    High neutron economy, on-power refuelling and a simple fuel bundle design result in unsurpassed fuel cycle flexibility for CANDU reactors. These features facilitate the introduction and exploitation of thorium fuel cycles in existing CANDU reactors in an evolutionary fashion. Detailed full-core fuel-management simulations concluded that a once-through thorium fuel cycle can be successfully implemented in an existing CANDU reactor without requiring major modifications. (author)

  9. Training and teaching with SILOETTE reactor and associated simulators at the Nuclear Research Centre of Grenoble

    International Nuclear Information System (INIS)

    Destot, M.

    1983-10-01

    Thanks to its three reactors SILOE (35 MW), MELUSINE (8 MW) and SILOETTE (100 KW), the Reactor Department of the Nuclear Research Centre of Grenoble has gained a considerable experience in the operation and utilization of research and material testing reactors. Inside of this general framework, the Reactor Department of Grenoble has built up a training and teaching centre that has been permanently active since 1975, with the aim of satisfying the considerable needs arising from the development of electro-nuclear power stations. The course is mainly intended for engineers and technicians who will be responsible for running power stations. A thorough series of practical exercices, carried out in the SILOETTE training reactor and in a PWR or in a Gas Cooled Reactor Simulator, desmonstrates the application of the theorical courses and familiarises the trainees with the behaviour of reactors and power stations

  10. New exploration on TMSR: modelling and simulation

    Energy Technology Data Exchange (ETDEWEB)

    Si, S.; Chen, Q.; Bei, H.; Zhao, J., E-mail: ssy@snerdi.com.cn [Shanghai Nuclear Engineering Research & Design Inst., Shanghai (China)

    2015-07-01

    A tightly coupled multi-physics model for MSR (Molten Salt Reactor) system involving the reactor core and the rest of the primary loop has been developed and employed in an in-house developed computer code TANG-MSR. In this paper, the computer code is used to simulate the behavior of steady state operation and transient for our redesigned TMSR. The presented simulation results demonstrate that the models employed in TANG-MSR can capture major physics phenomena in MSR and the redesigned TMSR has excellent performance of safety and sustainability. (author)

  11. Design of Simulink module for dynamic reactivity simulation of marine reactor automatic control rod

    International Nuclear Information System (INIS)

    Chen Zhiyun; Luo Lei; Chen Wenzhen; Gui Xuewen

    2010-01-01

    The power of marine reactor varies frequently and acutely, which induces the frequent and acute adjustment of the automatic control rod. According to the characteristics of marine reactor and the problem of improper control rod reactivity insertion in previous literatures, the Simulink module for dynamic reactivity simulation of automatic control rod was designed and adopted as a sub-module of Simulink program for the fast calculation of the physical and thermal parameters of marine reactor. A typical dynamic process of the marine reactor was used as the benchmark, which indicates that the designed Simulink module is capable of the dynamic simulation of automatic control rod position and reactivity, and is adequate to the fast calculation of physic and thermal parameters. The Simulink module is of significant meaning to the simulation of the dynamic process of marine reactor and the fast calculation of the operating parameters. (authors)

  12. 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)

  13. Simulating Neutronic Core Parameters in a Research and Test Reactor

    International Nuclear Information System (INIS)

    Selim, H.K.; Amin, E.A.; Koutb, M.E.

    2011-01-01

    The present study proposes an Artificial Neural Network (ANN) modeling technique that predicts the control rods positions in a nuclear research reactor. The neutron, flux in the core of the reactor is used as the training data for the neural network model. The data used to train and validate the network are obtained by modeling the reactor core with the neutronic calculation code: CITVAP. The type of the network used in this study is the feed forward multilayer neural network with the backpropagation algorithm. The results show that the proposed ANN has good generalization capability to estimate the control rods positions knowing neutron flux for a research and test reactor. This method can be used to predict critical control rods positions to be used for reactor operation after reload

  14. Computer simulation of radiation processes in reactor facilities

    International Nuclear Information System (INIS)

    Gann, V.V.; Abdulaev, A.M.; Zhukov, A.I.; Marekhin, S.V.; Soldatov, S.A.

    2009-01-01

    The paper describes experience of the code system ALPHA-H/PHOENIX-H/ANC-H (APA) and the code MCNP usage for fuel assembly neutronic calculations and modeling of VVER-1000 reactor core. Using Monte Carlo code MCNP, calculations of neutron field and pin-by-pin energy deposition distributions are provided for different type of assemblies in reactor core. An MCNP model for unit 3 Zaporozhye NPP reactor core was designed. Calculations for pin-by-pin energy deposition in the reactor core were performed using the code system APA and the code MCNP. Comparison of these calculations shows rather high precision of APA calculation for energy deposition in the fuel rods and assemblies operated in the reactor core

  15. 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

  16. 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.

  17. 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

  18. 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

  19. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

    1988-01-01

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  20. Reactor control and protection of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhu Jinping; Sun Jiliang

    1996-01-01

    The control and protection simulation of Qinshan 300 MW Nuclear Power Unit, including the nuclear control, the pressurizer pressure control, the pressurizer level control, the rod control, the reactor shutdown protection and engineered safety feature etc are briefly introduced

  1. Multi-Physics Analysis of the Fermilab Booster RF Cavity

    International Nuclear Information System (INIS)

    Awida, M.; Reid, J.; Yakovlev, V.; Lebedev, V.; Khabiboulline, T.; Champion, M.

    2012-01-01

    After about 40 years of operation the RF accelerating cavities in Fermilab Booster need an upgrade to improve their reliability and to increase the repetition rate in order to support a future experimental program. An increase in the repetition rate from 7 to 15 Hz entails increasing the power dissipation in the RF cavities, their ferrite loaded tuners, and HOM dampers. The increased duty factor requires careful modelling for the RF heating effects in the cavity. A multi-physic analysis investigating both the RF and thermal properties of Booster cavity under various operating conditions is presented in this paper.

  2. Development of 3D CFD simulation method in nuclear reactor safety analysis

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mariah Adam

    2012-01-01

    One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

  3. Modification of UASB reactor by using CFD simulations for enhanced treatment of municipal sewage.

    Science.gov (United States)

    Das, Suprotim; Sarkar, Supriya; Chaudhari, Sanjeev

    2018-02-01

    Up-flow anaerobic sludge blanket (UASB) has been in use since last few decades for the treatment of organic wastewaters. However, the performance of UASB reactor is quite low for treatment of low strength wastewaters (LSWs) due to less biogas production leading to poor mixing. In the present research work, a modification was done in the design of UASB to improve mixing of reactor liquid which is important to enhance the reactor performance. The modified UASB (MUASB) reactor was designed by providing a slanted baffle along the height of the reactor having an angle of 5.7° with the vertical wall. A two-dimensional computational fluid dynamics (CFD) simulation of three phase gas-liquid-solid flow in MUASB reactor was performed and compared with conventional UASB reactor. The CFD study indicated better mixing in terms of vorticity magnitude in MUASB reactor as compared to conventional UASB, which was reflected in the reactor performance. The performance of MUASB was compared with conventional UASB reactor for the onsite treatment of domestic sewage as LSW. Around 16% higher total chemical oxygen demand removal efficiency was observed in MUASB reactor as compared to conventional UASB during this study. Therefore, this MUASB model demonstrates a qualitative relationship between mixing and performance during the treatment of LSW. From the study, it seems that MUASB holds promise for field applications.

  4. Simulation for Remote Operation for REX10 Nuclear Reactor

    International Nuclear Information System (INIS)

    Lee, Sim Won; Kim, Dong Su; Na, Man Gyun; Lee, Yoon Joon; Lee, Yeon Gun; Park, Goon Cherl

    2010-01-01

    The newly designed REX10 (Regional Energy Reactor, 10MWth) is an environmentally-friendly and stable small nuclear reactor for a small-scale reactor based Multi-purpose regional energy system. The REX10 has been developed to maintain system safety in order to be placed in densely populated region, island, etc. In addition, it is significantly hard to recruit many operation and maintenance personnel for small power reactors differently from usual commercial reactors because of its remote location and of economic reasons. In order to overcome these constraints, to decrease the operation cost by reducing operation and maintenance personnel, and to increase plant reliability through autonomous plant control, it is needed to design the control system of the small power reactors and to establish its unmanned remote operation system. In this study, the REX10 reactor core thermal power controller is designed by using a REX10 code analyzer. The remote control facility through man-machine interface (MMI) design and interface between programming languages was established and it was used to verify remote operation of REX10

  5. Propagation of neutron-reaction uncertainties through multi-physics models of novel LWR's

    Directory of Open Access Journals (Sweden)

    Hernandez-Solis Augusto

    2017-01-01

    Full Text Available The novel design of the renewable boiling water reactor (RBWR allows a breeding ratio greater than unity and thus, it aims at providing for a self-sustained fuel cycle. The neutron reactions that compose the different microscopic cross-sections and angular distributions are uncertain, so when they are employed in the determination of the spatial distribution of the neutron flux in a nuclear reactor, a methodology should be employed to account for these associated uncertainties. In this work, the Total Monte Carlo (TMC method is used to propagate the different neutron-reactions (as well as angular distributions covariances that are part of the TENDL-2014 nuclear data (ND library. The main objective is to propagate them through coupled neutronic and thermal-hydraulic models in order to assess the uncertainty of important safety parameters related to multi-physics, such as peak cladding temperature along the axial direction of an RBWR fuel assembly. The objective of this study is to quantify the impact that ND covariances of important nuclides such as U-235, U-238, Pu-239 and the thermal scattering of hydrogen in H2O have in the deterministic safety analysis of novel nuclear reactors designs.

  6. 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

  7. Development of a computer code for transients simulation in PWR type reactors

    International Nuclear Information System (INIS)

    Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

    1981-01-01

    A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

  8. A three-dimensional operational transient simulation of the CANDU core with typical reactor regulating system

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Choong Sub; Kim, Hyun Dae; Park, Kyung Seok; Park, Jong Woon [Institute for Advanced Engineering, Taejon (Korea, Republic of)

    1995-07-01

    This paper describes the results of simulation of a CANDU operational transient problem (re-startup after short shutdown) using the Coupled Reactor Kinetics(CRKIN) code developed previously with CANDU Reactor Regulating System (RRS) logic. The performance in the simulation is focused on investigating the behaviours of neutron power and regulating devices in accordance with the changes of xenon concentration following the operation of the RRS.

  9. 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.

  10. 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

  11. Dynamical and quasi-static multi-physical models of a diesel internal combustion engine using Energetic Macroscopic Representation

    International Nuclear Information System (INIS)

    Horrein, L.; Bouscayrol, A.; Cheng, Y.; El Fassi, M.

    2015-01-01

    Highlights: • Internal Combustion Engine (ICE) dynamical and static models. • Organization of ICE model using Energetic Macroscopic Representation. • Description of the distribution of the chemical, thermal and mechanical power. • Implementation of the ICE model in a global vehicle model. - Abstract: In the simulation of new vehicles, the Internal Combustion Engine (ICE) is generally modeled by a static map. This model yields the mechanical power and the fuel consumption. But some studies require the heat energy from the ICE to be considered (i.e. waste heat recovery, thermal regulation of the cabin). A dynamical multi-physical model of a diesel engine is developed to consider its heat energy. This model is organized using Energetic Macroscopic Representation (EMR) in order to be interconnected to other various models of vehicle subsystems. An experimental validation is provided. Moreover a multi-physical quasi-static model is also derived. According to different modeling aims, a comparison of the dynamical and the quasi-static model is discussed in the case of the simulation of a thermal vehicle. These multi-physical models with different simulation time consumption provide good basis for studying the effects of the thermal energy on the vehicle behaviors, including the possibilities of waste heat recovery

  12. Thermal hydraulic simulation of the CANDU nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Carvalho, Athos M.S.S. de; Ramos, Mario C.; Costa, Antonella L.; Fernandes, Gustavo H.N., E-mail: athos1495@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciência e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Rio de janeiro, RJ (Brazil)

    2017-07-01

    The CANDU (Canada Deuterium Uranium) is a Canadian-designed power reactor of PHWR type (Pressurized Heavy Water Reactor) that uses heavy water (deuterium oxide) for moderator and coolant, and natural uranium for fuel. There are about 47 reactors of this type in operation around the world generating more than 23 GWe, highlighting the importance of this kind of device. In this way, the main purpose of this study is to develop a thermal hydraulic model for a CANDU reactor to aggregate knowledge in this line of research. In this way, a core modeling was performed using RELAP5-3D code. Results were compared with reference data to verify the model behavior in steady state operation. Thermal hydraulic parameters as temperature, pressure and mass flow rate were verified and the results are in good agreement with reference data, as it is being presented in this work. (author)

  13. Simulation of MILD combustion using Perfectly Stirred Reactor model

    KAUST Repository

    Chen, Z.; Vanteru, Mahendra Reddy; Ruan, S.; Doan, N. A K; Roberts, William L.; Swaminathan, N.

    2016-01-01

    A simple model based on a Perfectly Stirred Reactor (PSR) is proposed for moderate or intense low-oxygen dilution (MILD) combustion. The PSR calculation is performed covering the entire flammability range and the tabulated chemistry approach is used

  14. 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)

  15. 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)

  16. Computational Fluid Dynamics simulation of hydrothermal liquefaction of microalgae in a continuous plug-flow reactor.

    Science.gov (United States)

    Ranganathan, Panneerselvam; Savithri, Sivaraman

    2018-06-01

    Computational Fluid Dynamics (CFD) technique is used in this work to simulate the hydrothermal liquefaction of Nannochloropsis sp. microalgae in a lab-scale continuous plug-flow reactor to understand the fluid dynamics, heat transfer, and reaction kinetics in a HTL reactor under hydrothermal condition. The temperature profile in the reactor and the yield of HTL products from the present simulation are obtained and they are validated with the experimental data available in the literature. Furthermore, the parametric study is carried out to study the effect of slurry flow rate, reactor temperature, and external heat transfer coefficient on the yield of products. Though the model predictions are satisfactory in comparison with the experimental results, it still needs to be improved for better prediction of the product yields. This improved model will be considered as a baseline for design and scale-up of large-scale HTL reactor. Copyright © 2018 Elsevier Ltd. All rights reserved.

  17. 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

  18. Simulation of coolant mixing in pressure vessel reactors

    International Nuclear Information System (INIS)

    Hoehne, T.

    2003-06-01

    The work was aimed at the experimental investigation and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing conductivity measurements by wire mesh sensors and velocity measurements by the LDA technique. The CFD calculations were carried out with the CFD-code CFX-4. For the design of the facility, calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. Flow measurements and the corresponding CFD calculations in the ROCOM downcomer under steady state conditions showed a Re number independency at nominal flow rates. The flow field is dominated by recirculation areas below the inlet nozzles. Transient flow measurements with high performance LDA-technique showed in agreement with CFX-4 results, that in the case of the start up of a pump after a laminar stage large vortices dominate the flow. In the case of stationary mixing, the maximum value of the averaged mixing scalar at the core inlet was found in the sector below the inlet nozzle, where the tracer was injected. At the start-up case of one pump due to a strong impulse driven flow at the inlet nozzle the horizontal part of the flow dominates in the downcomer. The injection is distributed into two main jets, the maximum of the tracer concentration at the core inlet appears at the opposite part of the loop where the tracer was injected. Additionally, the stationary three-dimensional flow distribution in the downcomer and the lower plenum of a VVER-440/V-230 reactor was calculated with CFX-4. The comparison with experimental data and an analytical mixing model showed a

  19. 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.).

  20. 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)

  1. 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

  2. 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

  3. Development of dynamic simulation code for fuel cycle of fusion reactor. 1. Single pulse operation simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan

    1997-11-01

    A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)

  4. 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)

  5. 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)

  6. A simulator-independent optimization tool based on genetic algorithm applied to nuclear reactor design

    International Nuclear Information System (INIS)

    Abreu Pereira, Claudio Marcio Nascimento do; Schirru, Roberto; Martinez, Aquilino Senra

    1999-01-01

    Here is presented an engineering optimization tool based on a genetic algorithm, implemented according to the method proposed in recent work that has demonstrated the feasibility of the use of this technique in nuclear reactor core designs. The tool is simulator-independent in the sense that it can be customized to use most of the simulators which have the input parameters read from formatted text files and the outputs also written from a text file. As the nuclear reactor simulators generally use such kind of interface, the proposed tool plays an important role in nuclear reactor designs. Research reactors may often use non-conventional design approaches, causing different situations that may lead the nuclear engineer to face new optimization problems. In this case, a good optimization technique, together with its customizing facility and a friendly man-machine interface could be very interesting. Here, the tool is described and some advantages are outlined. (author)

  7. 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

  8. 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

  9. Simulation of a nuclear accident by an academic simulator of a VVER-1000 reactor; Simulacion de un accidente nuclear, mediante un simulador academico de un reactor VVER-1000

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez G, L. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Mexico D. F. (Mexico); Salazar S, E., E-mail: laurahg42@gmail.com [UNAM, Facultad de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, 62250 Jiutepec, Morelos (Mexico)

    2014-10-15

    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)

  10. Foundational development of an advanced nuclear reactor integrated safety code

    International Nuclear Information System (INIS)

    Clarno, Kevin; Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth; Hooper, Russell Warren; Humphries, Larry LaRon

    2010-01-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  11. Foundational development of an advanced nuclear reactor integrated safety code.

    Energy Technology Data Exchange (ETDEWEB)

    Clarno, Kevin (Oak Ridge National Laboratory, Oak Ridge, TN); Lorber, Alfred Abraham; Pryor, Richard J.; Spotz, William F.; Schmidt, Rodney Cannon; Belcourt, Kenneth (Ktech Corporation, Albuquerque, NM); Hooper, Russell Warren; Humphries, Larry LaRon

    2010-02-01

    This report describes the activities and results of a Sandia LDRD project whose objective was to develop and demonstrate foundational aspects of a next-generation nuclear reactor safety code that leverages advanced computational technology. The project scope was directed towards the systems-level modeling and simulation of an advanced, sodium cooled fast reactor, but the approach developed has a more general applicability. The major accomplishments of the LDRD are centered around the following two activities. (1) The development and testing of LIME, a Lightweight Integrating Multi-physics Environment for coupling codes that is designed to enable both 'legacy' and 'new' physics codes to be combined and strongly coupled using advanced nonlinear solution methods. (2) The development and initial demonstration of BRISC, a prototype next-generation nuclear reactor integrated safety code. BRISC leverages LIME to tightly couple the physics models in several different codes (written in a variety of languages) into one integrated package for simulating accident scenarios in a liquid sodium cooled 'burner' nuclear reactor. Other activities and accomplishments of the LDRD include (a) further development, application and demonstration of the 'non-linear elimination' strategy to enable physics codes that do not provide residuals to be incorporated into LIME, (b) significant extensions of the RIO CFD code capabilities, (c) complex 3D solid modeling and meshing of major fast reactor components and regions, and (d) an approach for multi-physics coupling across non-conformal mesh interfaces.

  12. Modeling and simulation of loss of the ultimate heat sink in a typical material testing reactor

    International Nuclear Information System (INIS)

    El-Khatib, Hisham; El-Morshedy, Salah El-Din; Higazy, Maher G.; El-Shazly, Karam

    2013-01-01

    Highlights: ► A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in MTR. ► The model involves three coupled sub-models for core, heat exchanger and cooling tower. ► The model is validated against PARET for steady-state and verified by operation data for transients. ► The model is used to simulate the behavior of the reactor under a loss of the ultimate heat sink. ► The model results are analyzed and discussed. -- Abstract: A thermal–hydraulic model has been developed to simulate loss of the ultimate heat sink in a typical material testing reactor (MTR). The model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The model is validated against PARET code for steady-state operation and verified by the reactor operation records for transients. Then, the model is used to simulate the thermal–hydraulic behavior of the reactor under a loss of the ultimate heat sink event. The simulation is performed for two operation regimes: regime I representing 11 MW power and three cooling tower cells operated, and regime II representing 22 MW power and six cooling tower cells operated. In regime I, the simulation is performed for 1, 2 and 3 cooling tower cells failed while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower cells failed. The simulation is performed under protected conditions where the safety action called power reduction is triggered by reactor protection system to decrease the reactor power by 20% when the coolant inlet temperature to the core reaches 43 °C and scram is triggered if the core inlet temperature reaches 44 °C. The model results are analyzed and discussed.

  13. Simulation of power excursions - Osiris reactor; Simulation des excursions de puissance - pile Osiris

    Energy Technology Data Exchange (ETDEWEB)

    Pascouet, A [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    Following the experimental work accomplished in the U.S.A. on Borax 1 and SPERT 1 and the accident of SL 1, the 'Commissariat a l'Energie Atomique' started a research program about the safety of its own swimming Pool reactors, with regard to power excursions. The first research work led to the design of programmed explosive charges, adapted to the simulation of a power excursion. This report describes the application of these methods to the investigation of Osiris safety. (author) [French] A la suite des essais effectues aux U.S.A. sur BORAX 1 et SPERT 1 et de l'accident survenu a SL 1, le Commissariat a l'Energie Atomique a lance un programme d'etudes sur la surete de ses reacteurs piscines vis-a-vis des excursions de puissance. Les premieres etudes ont abouti A la mise au point de charges programmees capables de simuler une excursion de puissance. On trouvera dans le present rapport l'application de ces methodes a l'etude de la surete d'OSIRIS. (auteur)

  14. SAFSIM: A computer program for engineering simulations of space reactor system performance

    International Nuclear Information System (INIS)

    Dobranich, D.

    1992-01-01

    SAFSIM (System Analysis Flow SIMulator) is a FORTRAN computer program that provides engineering simulations of user-specified flow networks at the system level. It includes fluid mechanics, heat transfer, and reactor dynamics capabilities. SAFSIM provides sufficient versatility to allow the simulation of almost any flow system, from a backyard sprinkler system to a clustered nuclear reactor propulsion system. In addition to versatility, speed and robustness are primary goals of SAFSIM. The current capabilities of SAFSIM are summarized, and some illustrative example results are presented

  15. 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

  16. 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.

  17. Reactor controller design using genetic algorithm with simulated annealing

    International Nuclear Information System (INIS)

    Willjuice Iruthyarajan, M.

    2012-01-01

    Many reactor control design work, specifically the problem of synthesis and optimization of reactor networks involving the classical reaction schemes was studied, considering a superstructure formed by a CSTR and a PFR and their possible arrangements. A genetic algorithm was proposed, together with a systematic procedure. Two case studies were solved with the proposed systematic. Both of them present similar results than the published in the literature. The first case studied was the Trambouze reaction scheme. Although selectivity values are smaller then the values published in the referred papers, the reactors system combined volume is always minor them the other ones. The second case studied was the Van de Vusse reaction scheme. In this case, the obtained value for the total volume is always minor then the considered papers. One can conclude that when compared with the other works presented in the literature results are compatible and very interesting. The developed algorithms can be used as a good alternative for reactor networks design and optimization problem

  18. Coupling between a multi-physics workflow engine and an optimization framework

    Science.gov (United States)

    Di Gallo, L.; Reux, C.; Imbeaux, F.; Artaud, J.-F.; Owsiak, M.; Saoutic, B.; Aiello, G.; Bernardi, P.; Ciraolo, G.; Bucalossi, J.; Duchateau, J.-L.; Fausser, C.; Galassi, D.; Hertout, P.; Jaboulay, J.-C.; Li-Puma, A.; Zani, L.

    2016-03-01

    A generic coupling method between a multi-physics workflow engine and an optimization framework is presented in this paper. The coupling architecture has been developed in order to preserve the integrity of the two frameworks. The objective is to provide the possibility to replace a framework, a workflow or an optimizer by another one without changing the whole coupling procedure or modifying the main content in each framework. The coupling is achieved by using a socket-based communication library for exchanging data between the two frameworks. Among a number of algorithms provided by optimization frameworks, Genetic Algorithms (GAs) have demonstrated their efficiency on single and multiple criteria optimization. Additionally to their robustness, GAs can handle non-valid data which may appear during the optimization. Consequently GAs work on most general cases. A parallelized framework has been developed to reduce the time spent for optimizations and evaluation of large samples. A test has shown a good scaling efficiency of this parallelized framework. This coupling method has been applied to the case of SYCOMORE (SYstem COde for MOdeling tokamak REactor) which is a system code developed in form of a modular workflow for designing magnetic fusion reactors. The coupling of SYCOMORE with the optimization platform URANIE enables design optimization along various figures of merit and constraints.

  19. Benchmark for evaluation and validation of reactor simulations (BEAVRS)

    Energy Technology Data Exchange (ETDEWEB)

    Horelik, N.; Herman, B.; Forget, B.; Smith, K. [Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 77 Massachusetts Avenue, Cambridge, MA 02139 (United States)

    2013-07-01

    Advances in parallel computing have made possible the development of high-fidelity tools for the design and analysis of nuclear reactor cores, and such tools require extensive verification and validation. This paper introduces BEAVRS, a new multi-cycle full-core Pressurized Water Reactor (PWR) depletion benchmark based on two operational cycles of a commercial nuclear power plant that provides a detailed description of fuel assemblies, burnable absorbers, in-core fission detectors, core loading patterns, and numerous in-vessel components. This benchmark enables analysts to develop extremely detailed reactor core models that can be used for testing and validation of coupled neutron transport, thermal-hydraulics, and fuel isotopic depletion. The benchmark also provides measured reactor data for Hot Zero Power (HZP) physics tests, boron letdown curves, and three-dimensional in-core flux maps from fifty-eight instrumented assemblies. Initial comparisons between calculations performed with MIT's OpenMC Monte Carlo neutron transport code and measured cycle 1 HZP test data are presented, and these results display an average deviation of approximately 100 pcm for the various critical configurations and control rod worth measurements. Computed HZP radial fission detector flux maps also agree reasonably well with the available measured data. All results indicate that this benchmark will be extremely useful in validation of coupled-physics codes and uncertainty quantification of in-core physics computational predictions. The detailed BEAVRS specification and its associated data package is hosted online at the MIT Computational Reactor Physics Group web site (http://crpg.mit.edu/), where future revisions and refinements to the benchmark specification will be made publicly available. (authors)

  20. 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.

  1. 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.

  2. MATLAB/SIMULINK platform for simulation of CANDU reactor control system

    International Nuclear Information System (INIS)

    Javidnia, H.; Jiang, J.

    2007-01-01

    In this paper a simulation platform for CANDU reactors' control system is presented. The platform is built on MATLAB/SIMULINK interactive graphical interface. Since MATLAB/SIMULINK are powerful tools to describe systems mathematically, all the subsystems in a CANDU reactor are represented in MATLAB's language and are implemented in SIMULINK graphical representation. The focus of the paper is on the flux control loop of CANDU reactors. However, the ideas can be extended to include other parts in CANDU power plants and the same technique can be applied to other types of nuclear reactors and their control systems. The CANDU reactor model and xenon feedback model are also discussed in this paper. (author)

  3. Module-based Hybrid Uncertainty Quantification for Multi-physics Applications: Theory and Software

    Energy Technology Data Exchange (ETDEWEB)

    Tong, Charles [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chen, Xiao [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Iaccarino, Gianluca [Stanford Univ., CA (United States); Mittal, Akshay [Stanford Univ., CA (United States)

    2013-10-08

    In this project we proposed to develop an innovative uncertainty quantification methodology that captures the best of the two competing approaches in UQ, namely, intrusive and non-intrusive approaches. The idea is to develop the mathematics and the associated computational framework and algorithms to facilitate the use of intrusive or non-intrusive UQ methods in different modules of a multi-physics multi-module simulation model in a way that physics code developers for different modules are shielded (as much as possible) from the chores of accounting for the uncertain ties introduced by the other modules. As the result of our research and development, we have produced a number of publications, conference presentations, and a software product.

  4. Contribution to the study of multi-physical phenomena in cementitious materials

    International Nuclear Information System (INIS)

    Bary, B.

    2010-09-01

    This document is a synthesis of the applied research studies undertaken by the author during ten years, first at the University of Marne-La-Vallee during the period 1999-2002, then at the CEA. These studies concern the modeling and the numerical simulations of the cementitious materials behavior subjected on the one hand to moderate thermomechanical and hydric loadings, and on the other hand to chemical attacks due to the migration of calcium, carbonate and sulfate ions. The developed approaches may be viewed as multi-physical in the sense that the models used for describing the behavior couple various fields and phenomena such as mechanics, thermal, hydric and ionic transfers, and chemistry. In addition, analytical up-scaling techniques are applied to estimate the physical properties associated with these phenomena (mechanical, hydraulic and diffusive parameters) as a function of the microstructure and the hydric state of the material. (author)

  5. CFD Simulation of an Anaerobic Membrane BioReactor (AnMBR to Treat Industrial Wastewater

    Directory of Open Access Journals (Sweden)

    Laura C. Zuluaga

    2015-06-01

    Full Text Available A Computational Fluid Dynamics (CFD simulation has been developed for an Anaerobic Membrane BioReactor (AnMBR to treat industrial wastewater. As the process consists of a side-stream MBR, two separate simulations were created: (i reactor and (ii membrane. Different cases were conducted for each one, so the surrounding temperature and the total suspended solids (TSS concentration were checked. For the reactor, the most important aspects to consider were the dead zones and the mixing, whereas for the ceramic membrane, it was the shear stress over the membrane surface. Results show that the reactor's mixing process was adequate and that the membrane presented higher shear stress in the 'triangular' channel.

  6. 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)

  7. 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)

  8. A two-dimensional simulator of the neutronic behaviour of low power fast reactors

    International Nuclear Information System (INIS)

    Penha, M.A.V.R. da.

    1984-01-01

    A model to simulate the temporal neutronic behaviour of fast breeder reactors was developed. The effective cross-sections are corrected, whenever the reactor state change; by using linear correlations and interpolation schemes with data contained in a library previously compiled. This methodology was coupled with a simplified spatial neutronic calculation to investigate the temporal behaviour of neutronic parameters such as breeding gain, flux and power. (Author) [pt

  9. Simulation of inlet and outlet riser break sequences in the N Reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.

    1988-02-01

    This report documents work performed at the Idaho National Engineering Laboratory (INEL) in support of the Westinghouse Hanford Company safety analyses of the N Reactor. The RELAP5/MOD2 computer code was used in analyzing two hypothetical transients. The computer code was modified specifically to simulate the refill behavior in the N Reactor process tubes. The transients analyzed were a double-ended rupture of an inlet riser column and a double-ended rupture of an outlet riser column

  10. Further experience in simulation of rod drop experiments in the Loviisa and Mochovce reactors

    International Nuclear Information System (INIS)

    Siltanen, P.; Kaloinen, E.; Tanskanen, A.; Mattila, R.

    2001-01-01

    Simulations of reactor scram experiments using the 3-dimensional kinetics code HEXTRAN have been updated for the initial cores of Loviisa-1 and 2 Mochovce-1 and have been extended to burned cores of Loviisa-1. In these simulations, the entire experiment is simulated dynamically, including the behaviour of the core, the signal of the ionization chamber, and the inverse point kinetics of the reactivity meter. The predicted output of the reactivity meter is compared with the output observed during the experiment (Authors)

  11. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  12. High performance multi-scale and multi-physics computation of nuclear power plant subjected to strong earthquake. An Overview

    International Nuclear Information System (INIS)

    Yoshimura, Shinobu; Kawai, Hiroshi; Sugimoto, Shin'ichiro; Hori, Muneo; Nakajima, Norihiro; Kobayashi, Kei

    2010-01-01

    Recently importance of nuclear energy has been recognized again due to serious concerns of global warming and energy security. In parallel, it is one of critical issues to verify safety capability of ageing nuclear power plants (NPPs) subjected to strong earthquake. Since 2007, we have been developing the multi-scale and multi-physics based numerical simulator for quantitatively predicting actual quake-proof capability of ageing NPPs under operation or just after plant trip subjected to strong earthquake. In this paper, we describe an overview of the simulator with some preliminary results. (author)

  13. The problem of a digital simulation of Xe oscillations in power reactors

    International Nuclear Information System (INIS)

    Elzmann, H.J.

    1974-04-01

    Xe-induced power oscillations are simulated in a pressurized water reactor. The coupled balance equation for the neutrons and the decay products iodine/xenon are decoupled via a quasi-stationary approach. The stationary multigroup diffusion equation is solved with a difference method. The whole model is realized with the aid of already existing modules of the reactor program system RSYST. Its basic usefulness is established. A further expansion of the method is discussed with the aim to develop rod drive programs for real reactors. (orig./LN) [de

  14. 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)

  15. A simulation Model of the Reactor Hall Ventilation and air Conditioning Systems of ETRR-2

    International Nuclear Information System (INIS)

    Abd El-Rahman, M.F.

    2004-01-01

    Although the conceptual design for any system differs from one designer to another. each of them aims to achieve the function of the system required. the ventilation and air conditioning system of reactors hall is one of those systems that really differs but always dose its function for which it is designed. thus, ventilation and air conditioning in some reactor hall constitute only one system whereas in some other ones, they are separate systems. the Egypt Research Reactor-2 (ETRR-2)represents the second type. most studies conducted on ventilation and air conditioning simulation models either in traditional building or for research rectors show that those models were not designed similarly to the model of the hall of ETRR-2 in which ventilation and air conditioning constitute two separate systems.besides, those studies experimented on ventilation and air conditioning simulation models of reactor building predict the temperature and humidity inside these buildings at certain outside condition and it is difficult to predict when the outside conditions are changed . also those studies do not discuss the influences of reactor power changes. therefore, the present work deals with a computational study backed by infield experimental measurements of the performance of the ventilation and air conditioning systems of reactor hall during normal operation at different outside conditions as well as at different levels of reactor power

  16. Simulation model of dynamical behaviour of reactor fuel assemblies

    International Nuclear Information System (INIS)

    Planchard, J.

    1994-01-01

    This report briefly describes the homogenized dynamical equations of a tube bundle placed in a perfect irrotational fluid, on case of small displacements. This approach can be used to study the mechanical behaviour of fuel assemblies of PWR reactor submitted to earthquake or depressurization blow-down. The numerical calculations require to define the added mass matrix of the fuel assemblies, for which the principle of computation is presented. (author). 14 refs., 4 figs

  17. Assessing reactor physics codes capabilities to simulate fast reactors on the example of the BN-600 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Ivanov, Vladimir [Scientific and Engineering Centre for Nuclear and Radiation Safety (SES NRS), Moscow (Russian Federation); Bousquet, Jeremy [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work aims to assess the capabilities of reactor physics codes (initially validated for thermal reactors) to simulate fast sodium cooled reactors. The BFS-62-3A critical experiment from the BN-600 Hybrid Core Benchmark Analyses was chosen for the investigation. Monte-Carlo codes (KENO from SCALE and SERPENT 2.1.23) and the deterministic diffusion code DYN3D-MG are applied to calculate the neutronic parameters. It was found that the multiplication factor and reactivity effects calculated by KENO and SERPENT using the ENDF/B-VII.0 continuous energy library are in a good agreement with each other and with the measured benchmark values. Few-groups macroscopic cross sections, required for DYN3D-MG, were prepared in applying different methods implemented in SCALE and SERPENT. The DYN3D-MG results of a simplified benchmark show reasonable agreement with results from Monte-Carlo calculations and measured values. The former results are used to justify DYN3D-MG implementation for sodium cooled fast reactors coupled deterministic analysis.

  18. 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)

  19. Pressurized water reactor simulation in the training environment

    International Nuclear Information System (INIS)

    Wills, A.G.

    1990-01-01

    The paper gives a brief history of PWR Simulation within the DNST and an outline of the training courses leading to the requirement for the Display Array Simulation System. Focus is then placed upon the flexible use of real time simulation in the teaching of plant dynamics by the use of model generated data. The use of interactive consoles and a large scale colour graphic display has led to the success of the Display Array Simulation System within the DNST. Realisation of the potential of the system has led to many other proposed uses for the installed system and the paper concludes by discussing some of these. (orig./DG)

  20. Internal transport barrier formation and pellet injection simulation in helical and tokamak reactors

    International Nuclear Information System (INIS)

    Higashiyama, You; Yamazaki, Kozo; Arimoto, Hideki; Garcia, Jeronimo

    2008-01-01

    In the future fusion reactor, plasma density peaking is important for increase in the fusion power gain and for achievement of confinement improvement mode. Density control and internal transport barrier (ITB) formation due to pellet injection have been simulated in tokamak and helical reactors using the toroidal transport linkage code TOTAL. First, pellet injection simulation is carried out, including the neutral gas shielding model and the mass relocation model in the TOTAL code, and the effectiveness of high-field side (HFS) pellet injection is clarified. Second, ITB simulation with pellet injection is carried out with the confinement improvement model based on the E x B shear effects, and it is found that deep pellet penetration is helpful for ITB formation as well as plasma core fuelling in the reversed-shear tokamak and helical reactors. (author)

  1. Comparison of Instrumentation and Control Parameters Based on Simulation and Experimental Data for Reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Anith Khairunnisa Ghazali; Mohd Sabri Minhat

    2015-01-01

    Reactor TRIGA PUSPATI (RTP) undergoes safe operation for more than 30 years and the only research reactor in Malaysia. The main safety feature of Instrumentation and Control (I and C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. There are no best models for RTP simulation was designed for study and research. Therefore, the comparison for I&C parameters are very essential, to design the best RTP model using MATLAB/ Simulink as close as the RTP. The simulation of TRIGA reactor type already develop using desktop reactor simulator such as Personal Computer Transient Analyzer (PCTRAN). The experimental data from RTP and simulation of PCTRAN shows some similarities and differences due to certain limitation. Currently, the structured RTP simulation was designed using MATLAB and Simulink tool that consist of ideal fission chamber, controller, control rod position, height to worth and RTP model. The study on this paper focus on comparison between real data from RTP and simulation result from PCTRAN on I&C parameters such as water level, fuel temperature, bulk temperature, power rated and rod position. The error analysis due to some similarities and differences of I&C parameters shall be obtained and analysed. The result will be used as reference for proposed new structured of RTP model. (author)

  2. A dynamic model of the reactor coolant system flow for KMRR plant simulation

    International Nuclear Information System (INIS)

    Rhee, B.W.; Noh, T.W.; Park, C.; Sim, B.S.; Oh, S.K.

    1990-01-01

    To support computer simulation studies for reactor control system design and performance evaluation, a dynamic model of the reactor coolant system (RCS) and reflector cooling system has been developed. This model is composed of the reactor coolant loop momentum equation, RCS pump dynamic equation, RCS pump characteristic equation, and the energy equation for the coolant inside the various components and piping. The model is versatile enough to simulate the normal steady-state conditions as well as most of the anticipated flow transients without pipe rupture. This model has been successfully implemented as the plant simulation code KMRRSIM for the Korea Multi-purpose Research Reactor and is now under extensive validation testing. The initial stage of validation has been comparison of its result with that of already validated, more detailed reactor system transient codes such as RELAP5. The results, as compared to the predictions by RELAP5 simulation, have been generally found to be very encouraging and the model is judged to be accurate enough to fulfill its intended purpose. However, this model will continue to be validated against other plant's data and eventually will be assessed by test data from KMRR

  3. Package Flow Model and its fuzzy implementation for simulating nuclear reactor system dynamics

    International Nuclear Information System (INIS)

    Matsuoka, Hiroshi; Ishiguro, Misako.

    1996-01-01

    A simple intuitive simulation model, which we call 'Package Flow Model', has been developed to evaluate physical processes in nuclear reactor system from a macroscopic point of view. In the previous paper, we showed the physical process of each energy generation and transfer stage in a PWR could be modeled by PFM, and its dynamics could be approximately simulated by fuzzy implementation. In this paper, a PFMs network approach for a total PWR system simulation is proposed and some transients of nuclear ship 'MUTSU' reactor system are evaluated. The simulated results are consistent with those from Nuclear Ship Engineering Simulation System developed by JAERI. Furthermore, a visual representation method is proposed to intuitively capture the profile of fuel safety transient. Using the PFMs network, we can handily calculate the transient phenomena of the system even by a notebook-type personal computer. In addition, we can easily interpret the results of calculation surveying a small number of parameters. (author)

  4. Training simulator for nuclear power plant reactor control model and method

    International Nuclear Information System (INIS)

    Czerbuejewski, F.R.

    1975-01-01

    A description is given of a method and system for the real-time dynamic simulation of a nuclear power plant for training purposes, wherein a control console has a plurality of manual and automatic remote control devices for operating simulated control rods and has indicating devices for monitoring the physical operation of a simulated reactor. Digital computer means are connected to the control console to calculate data values for operating the monitoring devices in accordance with the control devices. The simulation of the reactor control rod mechanism is disclosed whereby the digital computer means operates the rod position monitoring devices in a real-time that is a fraction of the computer time steps and simulates the quick response of a control rod remote control lever together with the delayed response upon a change of direction

  5. 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)

  6. 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)

  7. Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabet, F.

    2004-02-01

    Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed

  8. General meeting. Technical reunion: the numerical and experimental simulation applied to the Reactor Physics

    International Nuclear Information System (INIS)

    2001-10-01

    The SFEN (French Society on Nuclear Energy), organized the 18 october 2001 at Paris, a technical day on the numerical and experimental simulation, applied to the reactor Physics. Nine aspects were discussed, giving a state of the art in the domain:the french nuclear park; the future technology; the controlled thermonuclear fusion; the new organizations and their implications on the research and development programs; Framatome-ANP markets and industrial code packages; reactor core simulation at high temperature; software architecture; SALOME; DESCARTES. (A.L.B.)

  9. Training methods and facilities on reactor and simulators at the Grenoble Nuclear Research Centre

    International Nuclear Information System (INIS)

    Destot, M.; Siebert, S.

    1987-01-01

    Siloette is a CEA unit with a threshold vocation: operation of the Siloette 100 KW pool-type research reactor; basic training in reactor physics for nuclear power plant operators; and production of nuclear power plant simulators: PWR, GCR and more generally of all types of industrial unit simulators, thermal power plant, network, chemical plant, etc. From this experience, they would emphasize in particular the synergy arising from these complementary activities, the essential role of training in basic principles as a complement to operation training, and the ever-increasing importance of design ergonomics of the training means

  10. The chemical energy unit partial oxidation reactor operation simulation modeling

    Science.gov (United States)

    Mrakin, A. N.; Selivanov, A. A.; Batrakov, P. A.; Sotnikov, D. G.

    2018-01-01

    The chemical energy unit scheme for synthesis gas, electric and heat energy production which is possible to be used both for the chemical industry on-site facilities and under field conditions is represented in the paper. The partial oxidation reactor gasification process mathematical model is described and reaction products composition and temperature determining algorithm flow diagram is shown. The developed software product verification showed good convergence of the experimental values and calculations according to the other programmes: the temperature determining relative discrepancy amounted from 4 to 5 %, while the absolute composition discrepancy ranged from 1 to 3%. The synthesis gas composition was found out practically not to depend on the supplied into the partial oxidation reactor (POR) water vapour enthalpy and compressor air pressure increase ratio. Moreover, air consumption coefficient α increase from 0.7 to 0.9 was found out to decrease synthesis gas target components (carbon and hydrogen oxides) specific yield by nearly 2 times and synthesis gas target components required ratio was revealed to be seen in the water vapour specific consumption area (from 5 to 6 kg/kg of fuel).

  11. Simplified dynamic simulation of a traveling wave nuclear reactor

    International Nuclear Information System (INIS)

    Sanchez M, H.; Espinosa P, G.; Francois, J. L.; Lopez S, R.

    2016-09-01

    In this work the nuclear fuel burn wave in a fast traveling wave reactor (TWR) is presented, using the reduced model of the neutron diffusion equation, considering only the axial component, and the equations of the transuranic dynamics of U-Pu and a radionuclide of Pu. Two critical zones of the reactor are considered, one enriched with U-Pu called ignition zone and the other impoverished zone or of U-238, named breeding zone. Occupying Na as refrigerant within TWR, and Fe as structural material; both are present in the ignition and breeding zones. Considering as a fissile material the Pu, since by neutron capture the U is transformed into Pu, thus increasing the quantity of Pu more than that of U; in this way the fuel burn stability with the wave dynamics is understood. The calculation of the results was approached numerically to determine the temporal space evolution of the neutron flux in this system and of the main isotopes involved in the burning process. (Author)

  12. 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)

  13. RPV-1: a first virtual reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, St.

    2005-01-01

    The presented work was aimed at building a first VTR (virtual test reactor) to simulate irradiation effects in pressure vessel steels of nuclear reactor. It mainly consisted in: - modeling the formation of the irradiation induced damage in such steels, as well as their plasticity behavior - selecting codes and models to carry out the simulations of the involved mechanisms. Since the main focus was to build a first tool (rather than a perfect tool), it was decided to use, as much as possible, existing codes and models in spite of their imperfections. - developing and parameterizing two missing codes: INCAS and DUPAIR. - proposing an architecture to link the selected codes and models. - constructing and validating the tool. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or transmit data. A user friendly Python interface facilitates the running of the simulations and the visualization of the results. RPV-1 relies on many simplifications and approximations and has to be considered as a prototype aimed at clearing the way. According to the functionalities targeted for RPV-1, the main weakness is a bad Ni and Mn sensitivity. However, the tool can already be used for many applications (understanding of experimental results, assessment of effects of material and irradiation conditions,....). (O.M.)

  14. LWR [Light Water Reactor] power plant simulations using the AD10 and AD100 systems

    International Nuclear Information System (INIS)

    Wulff, W.; Cheng, H.S.; Chien, C.J.; Jang, J.Y.; Lin, H.C.; Mallen, A.N.; Wang, S.J.

    1989-01-01

    Boiling (BWR) and Pressurized (PWR) Water Reactor Power Plants are being simulated at BNL with the AD10 and AD100 Peripheral Processor Systems. The AD10 system has been used for BWR simulations since 1984 for safety analyses, emergency training and optimization studies. BWR simulation capabilities have been implemented recently on the AD100 system and PWR simulation capabilities are currently being developed under the auspices of international cooperation. Modeling and simulation methods are presented with emphasis on the simulation of the Nuclear Steam Supply System. Results are presented for BWR simulation and performance characteristics are compared of the AD10 and AD100 systems. It will be shown that the AD100 simulates two times faster than two AD10 processors operating in parallel and that the computing capacity of one AD100 (with FMU processor) is twice as large as that of two AD10 processors. 9 refs., 5 figs., 1 tab

  15. Simulation of the automatic depressurization system (Ads) for a boiling water reactor (BWR) based on RELAP

    International Nuclear Information System (INIS)

    Ramirez G, C.; Chavez M, C.

    2012-10-01

    The automatic depressurization system (Ads) of the boiling water reactor (BWR) like part of the emergency cooling systems is designed to liberate the vapor pressure of the reactor vessel, as well as the main vapor lines. At the present time in the Engineering Faculty, UNAM personnel works in the simulation of the Laguna Verde reactor based on the nuclear code RELAP/SCADAP and in the incorporation to the same of the emergency cooling systems. The simulation of the emergency cooling systems began with the inclusion of two hydrodynamic volumes, one source and another drain, and the incorporation of the initiation logic for each emergency system. In this work is defined and designed a simplified model of Ads of the reactor, considering a detail level based on the main elements that compose it. As tool to implement the proposed model, the RELAP code was used. The simulated main functions of Ads are centered in the quick depressurization of the reactor by means of the vapor discharge through the relief/safety valves to the suppression pool, and, in the event of break of the main vapor line, the reduction of the vessel pressure operates for that the cooling systems of the core to low pressure (Lpcs and Lpci) they can begin their operation. (Author)

  16. V.S.O.P.-computer code system for reactor physics and fuel cycle simulation

    International Nuclear Information System (INIS)

    Teuchert, E.; Hansen, U.; Haas, K.A.

    1980-03-01

    V.S.O.P. (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on neutron flux synthesis with depletion and shutdown features, incore and out-of-pile fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted to Pebble Bed HTRs). Various techniques have been employed to accelerate the iterative processes and to optimize the internal data transfer. A limitation of the storage requirement to 360 K-bites is achieved by an overlay structure. The code system has been used extensively for comparison studies of reactors, their fuel cycles, and related detailed features. Beside its use in research and development work for the high temperature reactor the system has been applied successfully to LWR and Heavy Water Reactors. (orig.) [de

  17. Numerical simulation of flow field in the China advanced research reactor flow-guide tank

    International Nuclear Information System (INIS)

    Xu Changjiang

    2002-01-01

    The flow-guide tank in China advanced research reactor (CARR) acts as a reactor inlet coolant distributor and play an important role in reducing the flow-induced vibration of the internal components of the reactor core. Numerical simulations of the flow field in the flow-guide tank under different conceptual designing configurations are carried out using the PHOENICS3.2. It is seen that the inlet coolant is well distributed circumferentially into the flow-guide tank with the inlet buffer plate and the flow distributor barrel. The maximum cross-flow velocity within the flow-guide tank is reduced significantly, and the reduction of flow-induced vibration of reactor internals is expected

  18. Modeling, simulation, and optimization of a front-end system for acetylene hydrogenation reactors

    Directory of Open Access Journals (Sweden)

    R. Gobbo

    2004-12-01

    Full Text Available The modeling, simulation, and dynamic optimization of an industrial reaction system for acetylene hydrogenation are discussed in the present work. The process consists of three adiabatic fixed-bed reactors, in series, with interstage cooling. These reactors are located after the compression and the caustic scrubbing sections of an ethylene plant, characterizing a front-end system; in contrast to the tail-end system where the reactors are placed after the de-ethanizer unit. The acetylene conversion and selectivity profiles for the reactors are optimized, taking into account catalyst deactivation and process constraints. A dynamic optimal temperature profile that maximizes ethylene production and meets product specifications is obtained by controlling the feed and intercoolers temperatures. An industrial acetylene hydrogenation system is used to provide the necessary data to adjust kinetics and transport parameters and to validate the approach.

  19. Simulators and their use in the training of CEGB reactor operations engineers

    International Nuclear Information System (INIS)

    Madden, V.J.; Tompsett, P.A.

    1988-01-01

    The development of simulators in the Central Electricity Generating Board's nuclear power training are traced, and, in describing the overall training programme of an advanced gas-cooled reactor operations engineer, the contribution made by a range of simulation devices from concept through to full-scope replica simulators is indicated. The capabilities of today's simulators are such that they are also making other contributions to the commissioning and safe operation of nuclear power plants. They are being successfully used for ergonomic and procedure validation work and the testing and commissioning of software for automatic control systems, and data and alarm processing systems. (author)

  20. Transport simulation of ITER [International Thermonuclear Engineering Reactor] startup

    International Nuclear Information System (INIS)

    Attenberger, S.E.; Houlberg, W.A.

    1989-01-01

    The present International Thermonuclear Engineering Reactor (ITER) reference configurations are the ''Technology Phase,'' in which the plasma current is maintained noninductively at a subignition density, and the ''Physics Phase,'' which is ignited but requires inductive maintenance of the current. The WHIST 1.5-D transport code is used to evaluate the volt-second requirements of both configurations. A slow current ramp (60-80's) is required for fixed-radius startup in ITER to avoid hollow current density profiles. To reach the operating point requires about 203 V·s for the Technology Phase (18 MA) and about 270 V·s for the Physics Phase (22 MA). The resistive losses can be reduced with expanding-radius startup. 5 refs., 4 figs

  1. Start-up simulations of the PULSAR pulsed tokamak reactor

    International Nuclear Information System (INIS)

    Werley, K.A.; Bathke, C.G.

    1993-01-01

    Start-up conditions are examined for a pulsed tokamak reactor that uses only inductively driven plasma current (and bootstrap current). A zero-dimensional (profile-averaged) model containing plasma power and particle balance equations is used to study several aspects of plasma start-up, including: (1) optimization of the start-up pathway; (2) tradeoffs of auxiliary start-up heating power versus start-up time; (3) volt-second consumption; (4) thermal stability of the operating point; (5) estimates of the diverter heat flux and temperature during the start-up transient; (6) the sensitivity of the available operating space to allowed values of the H confinement factor; and (7) partial-power operations

  2. Modelling of reactor control and protection systems in the core simulator program GARLIC

    International Nuclear Information System (INIS)

    Beraha, D.; Lupas, O.; Ploegert, K.

    1984-01-01

    For analysis of the interaction between control and limitation systems and the power distribution in the reactor core, a valuable tool is provided by the joint simulation of the core and the interacting systems. To this purpose, the core simulator GARLIC has been enhanced by models of the systems for controlling and limiting the reactor power and the power distribution in the core as well as by modules for calculating safety related core parameters. The computer-based core protection system, first installed in the Grafenrheinfeld NPP, has been included in the simulation. In order to evaluate the accuracy of GARLIC-simulations, the code has been compared with a design code in the train of a verification phase. The report describes the program extensions and the results of the verification. (orig.) [de

  3. Simulation an Accelerator driven Subcritical Reactor core with thorium fuel

    International Nuclear Information System (INIS)

    Shirmohammadi, L.; Pazirandeh, A.

    2011-01-01

    The main purpose of this work is simulation An Accelerator driven Subcritical core with Thorium as a new generation nuclear fuel. In this design core , A subcritical core coupled to an accelerator with proton beam (E p =1 GeV) is simulated by MCNPX code .Although the main purpose of ADS systems are transmutation and use MA (Minor Actinides) as a nuclear fuel but another use of these systems are use thorium fuel. This simulated core has two fuel assembly type : (Th-U) and (U-Pu) . Consequence , Neutronic parameters related to ADS core are calculated. It has shown that Thorium fuel is use able in this core and less nuclear waste ,Although Iran has not Thorium reserves but study on Thorium fuel cycle can open a new horizontal in use nuclear energy as a clean energy and without nuclear waste

  4. 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

  5. Peculiarities of approximation for reactor neutron energy spectra during computerized simulation of radiation defects

    International Nuclear Information System (INIS)

    Kupchishin, A.A.; Kupchishin, A.I.; Stusik, G.; Omarbekova, Zh.

    2001-01-01

    Peculiarities of approximation for reactor neutron energy spectra during radiation defects computerized simulation were discussed. Approximation of neutron spectra N(E) was carried out by N(E)=α·exp(-β·E)·sh(γ·E) formula (1), where α, β, γ - approximation coefficients. In the capacity of operating reactor data experimental data on 235 U and 239 Pu were applied. The algorithm was designed, and acting soft ware for spectra parameters calculation was developed. The following values of approximation parameters were obtained: α=80.8; β=0.935;γ=2.04 (for uranium and plutonium these coefficients are less distinguishing). Then with use of formula 1 and α, β, γ coefficients the approximation curves were constructed. These curves satisfactorily describe existing experimental data and allowing to use its for radiation defects simulation in the reactor materials

  6. Simulation of the preliminary General Electric SP-100 space reactor concept using the ATHENA computer code

    International Nuclear Information System (INIS)

    Fletcher, C.D.

    1986-01-01

    The capability to perform thermal-hydraulic analyses of a space reactor using the ATHENA computer code is demonstrated. The fast reactor, liquid-lithium coolant loops, and lithium-filled heat pipes of the preliminary General electric SP-100 design were modeled with ATHENA. Two demonstration transient calculations were performed simulating accident conditions. Calculated results are available for display using the Nuclear Plant Analyzer color graphics analysis tool in addition to traditional plots. ATHENA-calculated results appear reasonable, both for steady state full power conditions, and for the two transients. This analysis represents the first known transient thermal-hydraulic simulation using an integral space reactor system model incorporating heat pipes. 6 refs., 17 figs., 1 tab

  7. Transient thermal-hydraulic simulations of direct cycle gas cooled reactors

    International Nuclear Information System (INIS)

    Tauveron, Nicolas; Saez, Manuel; Marchand, Muriel; Chataing, Thierry; Geffraye, Genevieve; Bassi, Christophe

    2005-01-01

    This work concerns the design and safety analysis of gas cooled reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbo-machine trip, 10 in. cold duct break, 10 in. break in cold duct combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation. The turbo-machine contribution is discussed and can offer a help or an alternative to 'active' heat extraction systems

  8. BRENDA: a dynamic simulator for a sodium-cooled fast reactor power plant

    International Nuclear Information System (INIS)

    Hetrick, D.L.; Sowers, G.W.

    1978-06-01

    This report is a users' manual for one version of BRENDA (Breeder Reactor Nuclear Dynamic Analysis), which is a digital program for simulating the dynamic behavior of a sodium-cooled fast reactor power plant. This version, which contains 57 differential equations, represents a simplified model of the Clinch River Breeder Reactor Project (CRBRP). BRENDA is an input deck for DARE P (Differential Analyzer Replacement, Portable), which is a continuous-system simulation language developed at the University of Arizona. This report contains brief descriptions of DARE P and BRENDA, instructions for using BRENDA in conjunction with DARE P, and some sample output. A list of variable names and a listing for BRENDA are included as appendices

  9. Sequential UASB and dual media packed-bed reactors for domestic wastewater treatment - experiment and simulation.

    Science.gov (United States)

    Rodríguez-Gómez, Raúl; Renman, Gunno

    2016-01-01

    A wastewater treatment system composed of an upflow anaerobic sludge blanket (UASB) reactor followed by a packed-bed reactor (PBR) filled with Sorbulite(®) and Polonite(®) filter material was tested in a laboratory bench-scale experiment. The system was operated for 50 weeks and achieved very efficient total phosphorus (P) removal (99%), 7-day biochemical oxygen demand removal (99%) and pathogenic bacteria reduction (99%). However, total nitrogen was only moderately reduced in the system (40%). A model focusing on simulation of organic material, solids and size of granules was then implemented and validated for the UASB reactor. Good agreement between the simulated and measured results demonstrated the capacity of the model to predict the behaviour of solids and chemical oxygen demand, which is critical for successful P removal and recovery in the PBR.

  10. Steady-state and transient simulations of gas cooled reactor with the computer code CATHARE

    International Nuclear Information System (INIS)

    Tauveron, N.; Saez, M.; Marchand, M.; Chataing, T.; Geffraye, G.; Cherel, J. M.

    2003-01-01

    This work concerns the design and safety analysis of Gas Cooled Reactors. The CATHARE code is used to test the design and safety of two different concepts, a High Temperature Gas Reactor concept (HTGR) and a Gas Fast Reactor concept (GFR). Relative to the HTGR concept, three transient simulations are performed and described in this paper: loss of electrical load without turbomachine trip, 10 inch cold duct break, 10 inch cold duct break combined with a tube rupture of a cooling exchanger. A second step consists in modelling a GFR concept. A nominal steady state situation at a power of 600 MW is obtained and first transient simulations are carried out to study decay heat removal situations after primary loop depressurisation

  11. Simulation of hot-channel transients for PHWR reactors

    International Nuclear Information System (INIS)

    Masriera, N.A.

    1988-01-01

    For the simulation of transients a whole-plant code is needed. These codes model the core in a very simplified way. When local variables have to be calculated a different kind of code is needed: a subchannel-code. This report studies the use of the cobra code as a subchannel-code, for the simulation of a PHWR fuel channel, considering that this code was developed for PWR cores calculation. A special effort is made to obtain optimized models for different calculations: steady state, soft transients and severe transients. These models differ in number of subchannels, axial nodes, and the choice of the most important variables. (Author) [es

  12. High Conduction Neutron Absorber to Simulate Fast Reactor Environment in an Existing Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Guillen, Donna; Greenwood, Lawrence R.; Parry, James

    2014-06-22

    A need was determined for a thermal neutron absorbing material that could be cooled in a gas reactor environment without using large amounts of a coolant that would thermalize the neutron flux. A new neutron absorbing material was developed that provided high conduction so a small amount of water would be sufficient for cooling thereby thermalizing the flux as little as possible. An irradiation experiment was performed to assess the effects of radiation and the performance of a new neutron absorbing material. Neutron fluence monitors were placed inside specially fabricated holders within a set of drop-in capsules and irradiated for up to four cycles in the Advanced Test Reactor. Following irradiation, the neutron fluence monitor wires were analyzed by gamma and x-ray spectrometry to determine the activities of the activation products. The adjusted neutron fluences were calculated and grouped into three bins – thermal, epithermal and fast to evaluate the spectral shift created by the new material. Fluence monitors were evaluated after four different irradiation periods to evaluate the effects of burn-up in the absorbing material. Additionally, activities of the three highest activity isotopes present in the specimens are given.

  13. High conduction neutron absorber to simulate fast reactor environment in an existing test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen; Larry R. Greenwood; James R. Parry

    2014-06-22

    A new metal matrix composite material has been developed to serve as a thermal neutron absorber for testing fast reactor fuels and materials in an existing pressurized water reactor. The performance of this material was evaluated by placing neutron fluence monitors within shrouded and unshrouded holders and irradiating for up to four cycles. The monitor wires were analyzed by gamma and X-ray spectrometry to determine the activities of the activation products. Adjusted neutron fluences were calculated and grouped into three bins—thermal, epithermal, and fast—to evaluate the spectral shift created by the new material. A comparison of shrouded and unshrouded fluence monitors shows a thermal fluence decrease of ~11 % for the shielded monitors. Radioisotope activity and mass for each of the major activation products is given to provide insight into the evolution of thermal absorption cross-section during irradiation. The thermal neutron absorption capability of the composite material appears to diminish at total neutron fluence levels of ~8 × 1025 n/m2. Calculated values for dpa in excess of 2.0 were obtained for two common structural materials (iron and nickel) of interest for future fast flux experiments.

  14. Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

    International Nuclear Information System (INIS)

    Liebig, E.

    1977-01-01

    A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

  15. 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.

  16. The simulation study on the Nuclear Heating Reactor's power auto-control system

    International Nuclear Information System (INIS)

    Yang Zhijun; Liu Longzhi; Hu Guifen

    2000-01-01

    The power automatic control system on nuclear heating reactor (NHR) is a multi-input and multi-output non-linear system. The power automatic control system on NHR is studied by modern control theory. Through the simulation experiments, it is clear that adopting μ outer-loop and LQR inner-loop feedback, the best control results are obtained

  17. A Comparative Study on the Refueling Simulation Method for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Do, Quang Binh; Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The Canada deuterium uranium (CANDU) reactor calculation is typically performed by the RFSP code to obtain the power distribution upon a refueling. In order to assess the equilibrium behavior of the CANDU reactor, a few methods were suggested for a selection of the refueling channel. For example, an automatic refueling channel selection method (AUTOREFUEL) and a deterministic method (GENOVA) were developed, which were based on a reactor's operation experience and the generalized perturbation theory, respectively. Both programs were designed to keep the zone controller unit (ZCU) water level within a reasonable range during a continuous refueling simulation. However, a global optimization of the refueling simulation, that includes constraints on the discharge burn-up, maximum channel power (MCP), maximum bundle power (MBP), channel power peaking factor (CPPF) and the ZCU water level, was not achieved. In this study, an evolutionary algorithm, which is indeed a hybrid method based on the genetic algorithm, the elitism strategy and the heuristic rules for a multi-cycle and multi-objective optimization of the refueling simulation has been developed for the CANDU reactor. This paper presents the optimization model of the genetic algorithm and compares the results with those obtained by other simulation methods.

  18. 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...

  19. 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)

  20. Advances in thermal hydraulic and neutronic simulation for reactor analysis and safety

    International Nuclear Information System (INIS)

    Tentner, A.M.; Blomquist, R.N.; Canfield, T.R.; Ewing, T.F.; Garner, P.L.; Gelbard, E.M.; Gross, K.C.; Minkoff, M.; Valentin, R.A.

    1993-01-01

    This paper describes several large-scale computational models developed at Argonne National Laboratory for the simulation and analysis of thermal-hydraulic and neutronic events in nuclear reactors and nuclear power plants. The impact of advanced parallel computing technologies on these computational models is emphasized

  1. A simulated test of physical starting and reactor physics on zero power facility of PWR

    International Nuclear Information System (INIS)

    Yao Zewu; Ji Huaxiang; Chen Zhicheng; Yao Zhiquan; Chen Chen; Li Yuwen

    1995-01-01

    The core neutron economics has been verified through experiments conducted at a zero power reactor with baffles of various thickness. A simulated test of physical starting of Qinshan PWR has been introduced. The feasibility and safety of the programme are verified. The research provides a valuable foundation for developing physical starting programme

  2. Design of a Modular Monolithic Implicit Solver for Multi-Physics Applications

    Science.gov (United States)

    Carton De Wiart, Corentin; Diosady, Laslo T.; Garai, Anirban; Burgess, Nicholas; Blonigan, Patrick; Ekelschot, Dirk; Murman, Scott M.

    2018-01-01

    The design of a modular multi-physics high-order space-time finite-element framework is presented together with its extension to allow monolithic coupling of different physics. One of the main objectives of the framework is to perform efficient high- fidelity simulations of capsule/parachute systems. This problem requires simulating multiple physics including, but not limited to, the compressible Navier-Stokes equations, the dynamics of a moving body with mesh deformations and adaptation, the linear shell equations, non-re effective boundary conditions and wall modeling. The solver is based on high-order space-time - finite element methods. Continuous, discontinuous and C1-discontinuous Galerkin methods are implemented, allowing one to discretize various physical models. Tangent and adjoint sensitivity analysis are also targeted in order to conduct gradient-based optimization, error estimation, mesh adaptation, and flow control, adding another layer of complexity to the framework. The decisions made to tackle these challenges are presented. The discussion focuses first on the "single-physics" solver and later on its extension to the monolithic coupling of different physics. The implementation of different physics modules, relevant to the capsule/parachute system, are also presented. Finally, examples of coupled computations are presented, paving the way to the simulation of the full capsule/parachute system.

  3. Optimization by simulation of the coupling between a sub-critical reactor and its spallation source. Towards a pilot reactor

    International Nuclear Information System (INIS)

    Kerdraon, D.

    2001-10-01

    Accelerator Driven Systems (ADS), based on a proton accelerator and a sub-critical core coupled with a spallation target, offer advantages in order to reduce the nuclear waste radiotoxicity before repository closure. Many studies carried out on the ADS should lead to the definition of an experimental plan which would federate the different works in progress. This thesis deals with the neutronic Monte Carlo simulations with the MCNPX code to optimize such a system in view of a pilot reactor building. First, we have recalled the main neutronic properties of an hybrid reactor. The concept of gas-cooled eXperimental Accelerator Driven System (XADS) chosen for our investigations comes from the preliminary studies done by the Framatome company. In order to transmute minor actinides, we have considered the time evolution of the main fuels which could be reasonably used for the demonstration phases. The neutronic parameters of the reactor, concerning minor actinide transmutation, are reported. Also, we have calculated the characteristic times and the transmutation rates in the case of 99 Tc and 129 I isotopes. We have identified some neutronic differences between an experimental and a power ADS according to the infinite multiplication coefficient, the shape factor and the level of flux to extend the demonstrator concept. We have proposed geometric solutions to keep the radial shape factor of a power ADS acceptable. In the last part, beyond the experimental XADS scope, we have examined the possible transition towards an uranium/thorium cycle based on Molten Salt Reactors using a power ADS in order to generate the required 233 U proportion. (author)

  4. Real-time dynamic simulator for the Topaz II reactor power system

    International Nuclear Information System (INIS)

    Kwok, K.S.

    1994-01-01

    A dynamic simulator of the TOPAZ II reactor system has been developed for the Nuclear Electric Propulsion Space Test Program. The simulator is a self-contained IBM-PC compatible based system that executes at a speed faster than real-time. The simulator combines first-principle modeling and empirical correlations in its algorithm to attain the modeling accuracy and computational through-put that are required for real-time execution. The overall execution time of the simulator for each time step is 15 ms when no data is written to the disk, and 18 ms when nine double precision data points are written to the disk once in every time step. The simulation program has been tested and it is able to handle a step decrease of $8 worth of reactivity. It also provides simulation of fuel, emitter, collector, stainless steel, and ZrH moderator failures. Presented in this paper are the models used in the calculations, a sample simulation session, and a discussion of the performance and limitations of the simulator. The simulator has been found to provide realistic real-time dynamic response of the TOPAZ II reactor system under both normal and causality conditions

  5. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  6. Development and verification of the neutron diffusion solver for the GeN-Foam multi-physics platform

    International Nuclear Information System (INIS)

    Fiorina, Carlo; Kerkar, Nordine; Mikityuk, Konstantin; Rubiolo, Pablo; Pautz, Andreas

    2016-01-01

    Highlights: • Development and verification of a neutron diffusion solver based on OpenFOAM. • Integration in the GeN-Foam multi-physics platform. • Implementation and verification of acceleration techniques. • Implementation of isotropic discontinuity factors. • Automatic adjustment of discontinuity factors. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and the EPFL has been developing in recent years a new code system for reactor analysis based on OpenFOAM®. The objective is to supplement available legacy codes with a modern tool featuring state-of-the-art characteristics in terms of scalability, programming approach and flexibility. As part of this project, a new solver has been developed for the eigenvalue and transient solution of multi-group diffusion equations. Several features distinguish the developed solver from other available codes, in particular: object oriented programming to ease code modification and maintenance; modern parallel computing capabilities; use of general unstructured meshes; possibility of mesh deformation; cell-wise parametrization of cross-sections; and arbitrary energy group structure. In addition, the solver is integrated into the GeN-Foam multi-physics solver. The general features of the solver and its integration with GeN-Foam have already been presented in previous publications. The present paper describes the diffusion solver in more details and provides an overview of new features recently implemented, including the use of acceleration techniques and discontinuity factors. In addition, a code verification is performed through a comparison with Monte Carlo results for both a thermal and a fast reactor system.

  7. Reactor physics simulations with coupled Monte Carlo calculation and computational fluid dynamics

    International Nuclear Information System (INIS)

    Seker, V.; Thomas, J. W.; Downar, T. J.

    2007-01-01

    The interest in high fidelity modeling of nuclear reactor cores has increased over the last few years and has become computationally more feasible because of the dramatic improvements in processor speed and the availability of low cost parallel platforms. In the research here high fidelity, multi-physics analyses was performed by solving the neutron transport equation using Monte Carlo methods and by solving the thermal-hydraulics equations using computational fluid dynamics. A computation tool based on coupling the Monte Carlo code MCNP5 and the Computational Fluid Dynamics (CFD) code STAR-CD was developed as an audit tool for lower order nuclear reactor calculations. This paper presents the methodology of the developed computer program 'McSTAR' along with the verification and validation efforts. McSTAR is written in PERL programming language and couples MCNP5 and the commercial CFD code STAR-CD. MCNP uses a continuous energy cross section library produced by the NJOY code system from the raw ENDF/B data. A major part of the work was to develop and implement methods to update the cross section library with the temperature distribution calculated by STAR-CD for every region. Three different methods were investigated and two of them are implemented in McSTAR. The user subroutines in STAR-CD are modified to read the power density data and assign them to the appropriate variables in the program and to write an output data file containing the temperature, density and indexing information to perform the mapping between MCNP and STAR-CD cells. The necessary input file manipulation, data file generation, normalization and multi-processor calculation settings are all done through the program flow in McSTAR. Initial testing of the code was performed using a single pin cell and a 3X3 PWR pin-cell problem. The preliminary results of the single pin-cell problem are compared with those obtained from a STAR-CD coupled calculation with the deterministic transport code De

  8. Benchmark exercise for fluid flow simulations in a liquid metal fast reactor fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Merzari, E., E-mail: emerzari@anl.gov [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Fischer, P. [Mathematics and Computer Science Division, Argonne National Laboratory, 9700 S. Cass Avenue, Lemont, IL 60439 (United States); Yuan, H. [Nuclear Engineering Division, Argonne National Laboratory, Lemont, IL (United States); Van Tichelen, K.; Keijers, S. [SCK-CEN, Boeretang 200, Mol (Belgium); De Ridder, J.; Degroote, J.; Vierendeels, J. [Ghent University, Ghent (Belgium); Doolaard, H.; Gopala, V.R.; Roelofs, F. [NRG, Petten (Netherlands)

    2016-03-15

    Highlights: • A EUROTAM-US INERI consortium has performed a benchmark exercise related to fast reactor assembly simulations. • LES calculations for a wire-wrapped rod bundle are compared with RANS calculations. • Results show good agreement for velocity and cross flows. - Abstract: As part of a U.S. Department of Energy International Nuclear Energy Research Initiative (I-NERI), Argonne National Laboratory (Argonne) is collaborating with the Dutch Nuclear Research and consultancy Group (NRG), the Belgian Nuclear Research Centre (SCK·CEN), and Ghent University (UGent) in Belgium to perform and compare a series of fuel-pin-bundle calculations representative of a fast reactor core. A wire-wrapped fuel bundle is a complex configuration for which little data is available for verification and validation of new simulation tools. UGent and NRG performed their simulations with commercially available computational fluid dynamics (CFD) codes. The high-fidelity Argonne large-eddy simulations were performed with Nek5000, used for CFD in the Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) suite. SHARP is a versatile tool that is being developed to model the core of a wide variety of reactor types under various scenarios. It is intended both to serve as a surrogate for physical experiments and to provide insight into experimental results. Comparison of the results obtained by the different participants with the reference Nek5000 results shows good agreement, especially for the cross-flow data. The comparison also helps highlight issues with current modeling approaches. The results of the study will be valuable in the design and licensing process of MYRRHA, a flexible fast research reactor under design at SCK·CEN that features wire-wrapped fuel bundles cooled by lead-bismuth eutectic.

  9. Direct numerical simulation of reactor two-phase flows enabled by high-performance computing

    Energy Technology Data Exchange (ETDEWEB)

    Fang, Jun; Cambareri, Joseph J.; Brown, Cameron S.; Feng, Jinyong; Gouws, Andre; Li, Mengnan; Bolotnov, Igor A.

    2018-04-01

    Nuclear reactor two-phase flows remain a great engineering challenge, where the high-resolution two-phase flow database which can inform practical model development is still sparse due to the extreme reactor operation conditions and measurement difficulties. Owing to the rapid growth of computing power, the direct numerical simulation (DNS) is enjoying a renewed interest in investigating the related flow problems. A combination between DNS and an interface tracking method can provide a unique opportunity to study two-phase flows based on first principles calculations. More importantly, state-of-the-art high-performance computing (HPC) facilities are helping unlock this great potential. This paper reviews the recent research progress of two-phase flow DNS related to reactor applications. The progress in large-scale bubbly flow DNS has been focused not only on the sheer size of those simulations in terms of resolved Reynolds number, but also on the associated advanced modeling and analysis techniques. Specifically, the current areas of active research include modeling of sub-cooled boiling, bubble coalescence, as well as the advanced post-processing toolkit for bubbly flow simulations in reactor geometries. A novel bubble tracking method has been developed to track the evolution of bubbles in two-phase bubbly flow. Also, spectral analysis of DNS database in different geometries has been performed to investigate the modulation of the energy spectrum slope due to bubble-induced turbulence. In addition, the single-and two-phase analysis results are presented for turbulent flows within the pressurized water reactor (PWR) core geometries. The related simulations are possible to carry out only with the world leading HPC platforms. These simulations are allowing more complex turbulence model development and validation for use in 3D multiphase computational fluid dynamics (M-CFD) codes.

  10. Modelling transport phenomena in a multi-physics context

    Science.gov (United States)

    Marra, Francesco

    2015-01-01

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating.

  11. Modelling transport phenomena in a multi-physics context

    Energy Technology Data Exchange (ETDEWEB)

    Marra, Francesco [Dipartimento di Ingegneria Chimica e Alimentare - Università degli studi di Salerno Via Ponte Don Melillo - 84084 Fisciano SA (Italy)

    2015-01-22

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating.

  12. Modelling transport phenomena in a multi-physics context

    International Nuclear Information System (INIS)

    Marra, Francesco

    2015-01-01

    Innovative heating research on cooking, pasteurization/sterilization, defrosting, thawing and drying, often focuses on areas which include the assessment of processing time, evaluation of heating uniformity, studying the impact on quality attributes of the final product as well as considering the energy efficiency of these heating processes. During the last twenty years, so-called electro-heating-processes (radio-frequency - RF, microwaves - MW and ohmic - OH) gained a wide interest in industrial food processing and many applications using the above mentioned technologies have been developed with the aim of reducing processing time, improving process efficiency and, in many cases, the heating uniformity. In the area of innovative heating, electro-heating accounts for a considerable portion of both the scientific literature and commercial applications, which can be subdivided into either direct electro-heating (as in the case of OH heating) where electrical current is applied directly to the food or indirect electro-heating (e.g. MW and RF heating) where the electrical energy is firstly converted to electromagnetic radiation which subsequently generates heat within a product. New software packages, which make easier solution of PDEs based mathematical models, and new computers, capable of larger RAM and more efficient CPU performances, allowed an increasing interest about modelling transport phenomena in systems and processes - as the ones encountered in food processing - that can be complex in terms of geometry, composition, boundary conditions but also - as in the case of electro-heating assisted applications - in terms of interaction with other physical phenomena such as displacement of electric or magnetic field. This paper deals with the description of approaches used in modelling transport phenomena in a multi-physics context such as RF, MW and OH assisted heating

  13. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  14. Development of Multipurpose PLC trainer for the simulator of reactor safety system

    International Nuclear Information System (INIS)

    Syaiful Bakhri; Deswandri; Ahmad Abtokhi

    2014-01-01

    PLC becomes one of the essential components for the current type of reactor which based on digital instrumentation and control. Several studies have demonstrated the promising results including the implementation of PLC's for RSG-GAS research reactor. However, research for the safety and reliability analysis can not be carried out freely in the existing systems.Therefore, this research aims to develop a PLC trainer employing micro PLC OMRON CP1MA which can be useful for simulator of various topics in reactor safety. Two experimental tests were carried out to show the PLC’s performances. The first experimental testing implementing reactor protection system of research reactor RSG-GAS shows the capacity of PLC system to identify the initiator of the SCRAM logic as well as giving a promptly response. Secondly, the application of PLC to controls the water level in dual reservoir system simulation, demonstrates the simplicity of the operation and design while maintaining the best performances. (author)

  15. 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.)

  16. Simulation of In-Vessel Corium Retention through External Reactor Vessel Cooling for SMART using SIMPLE

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin-Sung; Son, Donggun; Park, Rae-Joon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Thermal load analysis from the corium pool to the outer reactor vessel in the lower plenum of the reactor vessel is necessary to evaluate the effect of the IVR-ERVC during a severe accident for SMART. A computational code called SIMPLE (Sever Invessel Melt Progression in Lower plenum Environment) has been developed for analyze transient behavior of molten corium in the lower plenum, interaction between corium and coolant, and heat-up and ablation of reactor vessel wall. In this study, heat load analysis of the reactor vessel for SMART has been conducted using the SIMPLE. Transient behavior of the molten corium in the lower plenum and IVR-ERVC for SMART has been simulated using SIMPLE. Heat flux from the corium pool to the outer reactor vessel is concentrated in metallic layer by the focusing effect. As a result, metallic layer shows higher temperature than the oxidic layer. Also, vessel wall of metallic layer has been ablated by the high in-vessel temperature. Ex-vessel temperature of the metallic layer was maintained 390 K and vessel thickness was maintained 14 cm. It means that the reactor vessel integrity is maintained by the IVR-ERVC.

  17. 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)

  18. Modelling of slaughterhouse solid waste anaerobic digestion: determination of parameters and continuous reactor simulation.

    Science.gov (United States)

    López, Iván; Borzacconi, Liliana

    2010-10-01

    A model based on the work of Angelidaki et al. (1993) was applied to simulate the anaerobic biodegradation of ruminal contents. In this study, two fractions of solids with different biodegradation rates were considered. A first-order kinetic was used for the easily biodegradable fraction and a kinetic expression that is function of the extracellular enzyme concentration was used for the slowly biodegradable fraction. Batch experiments were performed to obtain an accumulated methane curve that was then used to obtain the model parameters. For this determination, a methodology derived from the "multiple-shooting" method was successfully used. Monte Carlo simulations allowed a confidence range to be obtained for each parameter. Simulations of a continuous reactor were performed using the optimal set of model parameters. The final steady-states were determined as functions of the operational conditions (solids load and residence time). The simulations showed that methane flow peaked at a flow rate of 0.5-0.8 Nm(3)/d/m(reactor)(3) at a residence time of 10-20 days. Simulations allow the adequate selection of operating conditions of a continuous reactor. (c) 2010 Elsevier Ltd. All rights reserved.

  19. Hardware-in-the-Loop Simulation for the Automatic Power Control System of Research Reactors

    International Nuclear Information System (INIS)

    Fikry, R.M.; Shehata, S.A.; Elaraby, S.M.; Mahmoud, M.I.; Elbardini, M.M.

    2009-01-01

    Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this paper, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In- The-Loop (HIL). Some of the control loop components are real hardware components and thc others are simulated. First, the whole system is modeled and tested by Real- Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-Ioop simulation is tested using Real-Time Windows Target in MATLAB and Visual C++. The control parts are included as hardware components which are the reactor control rod and its drivers. Two kinds of controllers are studied, Proportional derivative (PD) and Fuzzy controller, An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain

  20. Feasibility study of teleoperational maintenance using real-time simulator for experimental fusion reactor

    International Nuclear Information System (INIS)

    Hamada, Tomoyuki; Tanaka, Keiji; Oka, Kiyoshi; Shibanuma, Kiyoshi

    2004-01-01

    The maintenance manipulator for the experimental fusion reactor has long vertical and horizontal telescopic booms to access the neutral beam injector of the fusion reactor. Due to this boom structure, the vibration and deflection of the manipulator are the critical issues for the accurate operation. A real-time simulation system was constructed to evaluate the maneuverability of the manipulator under these vibration and deflection. In this simulation system, the dynamic behavior of the flexible manipulator is calculated synchronized with the real-time control input of the human operator. A vibration and position compensation method was adapted to improve the maneuverability. Through the evaluation using the real-time simulation system, it was verified that the manipulator is maneuverable by using vibration and position compensation. (author)

  1. Numerical Simulation of Two-branch Hot Gas Mixing at Reactor Outlet of HTR-PM

    International Nuclear Information System (INIS)

    Hao Pengefei; Zhou Yangping; Li Fu; Shi Lei; He Heng

    2014-01-01

    A series of two-branch model experiment has been finished to investigate the thermal mixing efficiency of the HTR-PM reactor outlet. This paper introduces the numerical simulation on the design of thermal mixing structure of HTR-PM and the test facility with Fluent software. The profiles of temperature, pressure and velocity in the mixing structure design and the test facility are discussed by comparing with the model experiment results. The numerical simulation results of the test facility have good agreement to the experiment results. In addition, the thermal-fluid characters obtained by numerical simulation show the thermal mixing structure of HTR-PM has similarity with the test facility. Finally, it is concluded that the thermal mixing design at HTR-PM reactor outlet can fulfilled the requirements for high thermal mixing efficiency and appropriate pressure drop. (author)

  2. TRIGASIM: A computer program to simulate a TRIGA Mark I Reactor

    International Nuclear Information System (INIS)

    Ruby, Lawrence

    1992-01-01

    A Fortran-77 computer program has been written which simulates the operation of a TRIGA Mark I Reactor. The 'operator' has options at 1-second intervals, of raising rods, lowering rods, maintaining rods steady, dropping a rod, or scramming the reactor. Results are printed to the screen, and to 2 output files - a tabular record and a logarithmic plot of the power. The Point Kinetic Equations are programmed with 6 delayed groups, quasi-static power feedback, and forward differencing. A pulsing option is available, with simulation which employs the Fuchs Model. A pulse-tail model has been devised to simulate behavior for a few minutes following a pulse. Both graphic and tabular output are also available for the pulses. (author)

  3. Modular high-temperature gas-cooled reactor simulation using parallel processors

    International Nuclear Information System (INIS)

    Ball, S.J.; Conklin, J.C.

    1989-01-01

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulations routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (operator) interaction. In this paper the benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses

  4. Technical Basis for Physical Fidelity of NRC Control Room Training Simulators for Advanced Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Minsk, Brian S.; Branch, Kristi M.; Bates, Edward K.; Mitchell, Mark R.; Gore, Bryan F.; Faris, Drury K.

    2009-10-09

    The objective of this study is to determine how simulator physical fidelity influences the effectiveness of training the regulatory personnel responsible for examination and oversight of operating personnel and inspection of technical systems at nuclear power reactors. It seeks to contribute to the U.S. Nuclear Regulatory Commission’s (NRC’s) understanding of the physical fidelity requirements of training simulators. The goal of the study is to provide an analytic framework, data, and analyses that inform NRC decisions about the physical fidelity requirements of the simulators it will need to train its staff for assignment at advanced reactors. These staff are expected to come from increasingly diverse educational and experiential backgrounds.

  5. 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.

  6. Finding Solutions to Different Problems Simultaneously in a Multi-molecule Simulated Reactor

    Directory of Open Access Journals (Sweden)

    Jaderick P. Pabico

    2014-12-01

    Full Text Available – In recent years, the chemical metaphor has emerged as a computational paradigm based on the observation of different researchers that the chemical systems of living organisms possess inherent computational properties. In this metaphor, artificial molecules are considered as data or solutions, while the interactions among molecules are defined by an algorithm. In recent studies, the chemical metaphor was used as a distributed stochastic algorithm that simulates an abstract reactor to solve the traveling salesperson problem (TSP. Here, the artificial molecules represent Hamiltonian cycles, while the reactor is governed by reactions that can re-order Hamiltonian cycles. In this paper, a multi-molecule reactor (MMR-n that simulates chemical catalysis is introduced. The MMR-n solves in parallel three NP-hard computational problems namely, the optimization of the genetic parameters of a plant growth simulation model, the solution to large instances of symmetric and asymmetric TSP, and the static aircraft landing scheduling problems (ALSP. The MMR-n was shown as a computational metaphor capable of optimizing the cultivar coefficients of CERES-Rice model, and at the same time, able to find solutions to TSP and ALSP. The MMR-n as a computational paradigm has a better computational wall clock time compared to when these three problems are solved individually by a single-molecule reactor (MMR-1.

  7. PHEDRE model for the simulation of PWR reactors

    International Nuclear Information System (INIS)

    Bernard, Patrice; Dupraz, Remy; Vasile, Alfredo.

    1979-11-01

    This note presents the model of PHEDRE, simulator of a PWR, set on the hybrid computers of CISI, at the Nuclear Research Center of Cadarache. The model mainly concerns the primary part and the steam production of the PWR constructed in France. It includes an axial modelization of the core, the pressurizer, two loops of steam production and the inlet of the turbine, and the regulations concerning these components. The note presents the equations of the model, the structures of the codes concerning the initialization and the dynamic resolution, and describes the control panel of PHEDRE [fr

  8. Simulation of fusion first-wall environment in a fission reactor

    International Nuclear Information System (INIS)

    Hassanein, A.M.; Kulcinski, G.L.; Longhurst, G.R.

    1982-01-01

    A novel concept to produce a realistic simulation of a fusion first-wall test environment has been proposed recently. This concept takes advantage of the (/eta/, α) reaction in 59 Ni to produce a high internal helium content in the metal while using the 3 He (/eta/, /rho/)T reaction in the gas surrounding the specimen to produce an external heat and particle flux. Models to calculate heat flux, erosion rate, implantation, and damage rate to the walls of the test module are presented. Preliminary results show that a number of important fusion technology issues could be tested experimentally in a fission reactor such as the Engineering Test Reactor

  9. Size limitations for microwave cavity to simulate heating of blanket material in fusion reactor

    International Nuclear Information System (INIS)

    Wolf, D.

    1987-01-01

    The power profile in the blanket material of a nuclear fusion reactor can be simulated by using microwaves at 200 MHz. Using these microwaves, ceramic breeder materials can be thermally tested to determine their acceptability as blanket materials without entering a nuclear fusion environment. A resonating cavity design is employed which can achieve uniform cross sectional heating in the plane transverse to the neutron flux. As the sample size increases in height and width, higher order modes, above the dominant mode, are propagated and destroy the approximation to the heating produced in a fusion reactor. The limits at which these modes develop are determined in the paper

  10. Conceptual design and neutronics analyses of a fusion reactor blanket simulation facility

    International Nuclear Information System (INIS)

    Beller, D.E.; Ott, K.O.; Terry, W.K.

    1987-01-01

    A new conceptual design of a fusion reactor blanket simulation facility has been developed. This design follows the principles that have been successfully employed in the Purdue Fast Breeder Blanket Facility (FBBF), where experiments have resulted in the discovery of substantial deficiencies in neutronics predictions. With this design, discrepancies between calculation and experimental data can be nearly fully attributed to calculation methods because design deficiencies that could affect results are insignificant. The conceptual design of this FBBF analog, the Fusion Reactor Blanket Facility, is presented

  11. Simulation of the fuzzy-smith control system for the high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Li Deheng; Xu Xiaolin; Zheng Jie; Guo Renjun; Zhang Guifen

    1997-01-01

    The Fuzzy-Smith pre-estimate controller to solve the control of the big delay system is developed, accompanied with the development of the mathematical model of the 10 MW high temperature gas cooled test reactor (HTR-10) and the design of its control system. The simulation results show the Fuzzy-Smith pre-estimate controller has the advantages of both fuzzy control and Smith pre-estimate controller; it has better compensation to the delay and better adaptability to the parameter change of the control object. So it is applicable to the design of the control system for the high temperature gas cooled reactor

  12. Simulation model for the dynamic behavior of the hydraUlic circuito of PWR reactors

    International Nuclear Information System (INIS)

    Hirdes, V.R.T.R.

    1987-01-01

    The present work consist of the development of a computer code for the simulations of hydraulic transients caused by stoppages of the primary coolant pumps of nuclear reactors and it applied to the hydraulic circuits typical of PWR reactor. The code calculates the time-histories of the mass flux, rotation speed, electric and hydraulic torque and dynamic head of the pumps. It can be used for any combination of active and inactive pumps. Several transients were analysed and the results were compared with comparared with data from the Angra-I nuclear power plant. The results were considered satisfactory. (author) [pt

  13. 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

  14. Final report on LDRD project : coupling strategies for multi-physics applications.

    Energy Technology Data Exchange (ETDEWEB)

    Hopkins, Matthew Morgan; Moffat, Harry K.; Carnes, Brian; Hooper, Russell Warren; Pawlowski, Roger P.

    2007-11-01

    Many current and future modeling applications at Sandia including ASC milestones will critically depend on the simultaneous solution of vastly different physical phenomena. Issues due to code coupling are often not addressed, understood, or even recognized. The objectives of the LDRD has been both in theory and in code development. We will show that we have provided a fundamental analysis of coupling, i.e., when strong coupling vs. a successive substitution strategy is needed. We have enabled the implementation of tighter coupling strategies through additions to the NOX and Sierra code suites to make coupling strategies available now. We have leveraged existing functionality to do this. Specifically, we have built into NOX the capability to handle fully coupled simulations from multiple codes, and we have also built into NOX the capability to handle Jacobi Free Newton Krylov simulations that link multiple applications. We show how this capability may be accessed from within the Sierra Framework as well as from outside of Sierra. The critical impact from this LDRD is that we have shown how and have delivered strategies for enabling strong Newton-based coupling while respecting the modularity of existing codes. This will facilitate the use of these codes in a coupled manner to solve multi-physic applications.

  15. A full-spectrum analysis of high-speed train interior noise under multi-physical-field coupling excitations

    Science.gov (United States)

    Zheng, Xu; Hao, Zhiyong; Wang, Xu; Mao, Jie

    2016-06-01

    High-speed-railway-train interior noise at low, medium, and high frequencies could be simulated by finite element analysis (FEA) or boundary element analysis (BEA), hybrid finite element analysis-statistical energy analysis (FEA-SEA) and statistical energy analysis (SEA), respectively. First, a new method named statistical acoustic energy flow (SAEF) is proposed, which can be applied to the full-spectrum HST interior noise simulation (including low, medium, and high frequencies) with only one model. In an SAEF model, the corresponding multi-physical-field coupling excitations are firstly fully considered and coupled to excite the interior noise. The interior noise attenuated by sound insulation panels of carriage is simulated through modeling the inflow acoustic energy from the exterior excitations into the interior acoustic cavities. Rigid multi-body dynamics, fast multi-pole BEA, and large-eddy simulation with indirect boundary element analysis are first employed to extract the multi-physical-field excitations, which include the wheel-rail interaction forces/secondary suspension forces, the wheel-rail rolling noise, and aerodynamic noise, respectively. All the peak values and their frequency bands of the simulated acoustic excitations are validated with those from the noise source identification test. Besides, the measured equipment noise inside equipment compartment is used as one of the excitation sources which contribute to the interior noise. Second, a full-trimmed FE carriage model is firstly constructed, and the simulated modal shapes and frequencies agree well with the measured ones, which has validated the global FE carriage model as well as the local FE models of the aluminum alloy-trim composite panel. Thus, the sound transmission loss model of any composite panel has indirectly been validated. Finally, the SAEF model of the carriage is constructed based on the accurate FE model and stimulated by the multi-physical-field excitations. The results show

  16. Simulation of petcoke gasification in slagging moving bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nagpal, Soumitro; Sarkar, T.K.; Sen, P.K. [Research and Development Center, Engineers India Limited, Gurgaon 122001 (India)

    2005-03-25

    A mathematical model for simulation of moving bed petcoke gasifiers was developed. The model introduces a new feed characterization method, gas-phase resistance and volatilization models. The model is validated using reported data for a slagging gasifier. Effect of feed oxygen-to-coke and steam-to-coke ratios and feed coke rates on gasification performance was examined. Slagging zone moving bed gasifier operation with very high petcoke fluxes of over 4000 kg/m{sup 2}/h was possible with high petcoke conversion. Peak gas temperatures exceeded 1500 {sup o}C. Fluxes higher than 5000 kg/m{sup 2}/h are limited by an approach to fluidization of small particles in the combustion zone. The moving bed gasifier performance was found superior to performance of an entrained flow gasifier (EFG) with respect to energy efficiency and oxygen consumption.

  17. Leaching characteristics of actinides from simulated reactor waste glass

    International Nuclear Information System (INIS)

    Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S.

    1979-01-01

    Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0 0 C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0 0 C, whereas they show a correlation with flow rate for all three days at 75.0 0 C. 237 Np rates at 75.0 0 C are greater than those at 25.0 0 C, but 239 Pu rates at 75.0 0 C are less than or equal to those at 25.0 0 C

  18. Simulation software of 3-D two-neutron energy groups for ship reactor with hexagonal fuel subassembly

    International Nuclear Information System (INIS)

    Zhang Fan; Cai Zhangsheng; Yu Lei; Gui Xuewen

    2005-01-01

    Core simulation software for 3-D two-neutron energy groups is developed. This software is used to simulate the ship reactor with hexagonal fuel subassembly after 10, 150 and 200 burnup days, considering the hydraulic and thermal feedback. It accurately simulates the characteristics of the fast and thermal neutrons and the detailed power distribution in a reactor under normal and abnormal operation condition. (authors)

  19. Numerical simulation of vortex pyrolysis reactors for condensable tar production from biomass

    Energy Technology Data Exchange (ETDEWEB)

    Miller, R.S.; Bellan, J. [California Inst. of Tech., Pasadena, CA (United States). Jet Propulsion Lab.

    1998-08-01

    A numerical study is performed in order to evaluate the performance and optimal operating conditions of vortex pyrolysis reactors used for condensable tar production from biomass. A detailed mathematical model of porous biomass particle pyrolysis is coupled with a compressible Reynolds stress transport model for the turbulent reactor swirling flow. An initial evaluation of particle dimensionality effects is made through comparisons of single- (1D) and multi-dimensional particle simulations and reveals that the 1D particle model results in conservative estimates for total pyrolysis conversion times and tar collection. The observed deviations are due predominantly to geometry effects while directional effects from thermal conductivity and permeability variations are relatively small. Rapid ablative particle heating rates are attributed to a mechanical fragmentation of the biomass particles that is modeled using a critical porosity for matrix breakup. Optimal thermal conditions for tar production are observed for 900 K. Effects of biomass identity, particle size distribution, and reactor geometry and scale are discussed.

  20. Simulating the behaviour of zirconium-alloy components in nuclear reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.

    2001-12-01

    To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

  1. Trickle bed reactor model to simulate the performance of commercial diesel hydrotreating unit

    Energy Technology Data Exchange (ETDEWEB)

    C. Murali; R.K. Voolapalli; N. Ravichander; D.T. Gokak; N.V. Choudary [Bharat Petroleum Corporation Ltd., Udyog Kendra (India). Corporate R& amp; D Centre

    2007-05-15

    A two phase mathematical model was developed to simulate the performance of bench scale and commercial hydrotreating reactors. Major hydrotreating reactions, namely, hydrodesulphurization, hydrodearomatization and olefins saturation were modeled. Experiments were carried out in a fixed bed reactor to study the effect of different process variables and these results were used for estimating kinetic parameters. Significant amount of feed vaporization (20-50%) was estimated under normal operating conditions of DHDS suggesting the importance of considering feed vaporization in DHDS modeling. The model was validated with plant operating data, under close to ultra low sulphur levels by correctly accounting for feed vaporization in heat balance relations and appropriate use of hydrodynamic correlations. The model could predict the product quality, reactor bed temperature profiles and chemical hydrogen consumption in commercial plant adequately. 14 refs., 7 figs., 6 tabs.

  2. Natural circulation in a VVER reactor geometry: Experiments with the PACTEL facility and Cathare simulations

    Energy Technology Data Exchange (ETDEWEB)

    Raussi, P.; Kainulainen, S. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    There are some 40 reactors based on the VVER design in use. Database available for computer code assessment for VVER reactors is rather limited. Experiments were conducted to study natural circulation behaviour in the PACTEL facility, a medium-scale integral test loop patterned after VVER pressurized water reactors. Flow behaviour over a range of coolant inventories was studied with a small-break experiment. In the small-break experiments, flow stagnation and system repressurization were observed when the water level in the upper plenum fell below the entrances to the hot legs. The cause was attributed to the hot leg loop seals, which are a unique feature of the VVER geometry. At low primary inventories, core cooling was achieved through the boiler-condenser mode. The experiment was simulated using French thermalhydraulic system code CATHARE.

  3. The DSNP simulation language and its application to liquid-metal fast breeder reactor transient analyses

    International Nuclear Information System (INIS)

    Saphier, D.; Madell, J.T.

    1982-01-01

    A new, special purpose block-oriented simulation language, the Dynamic Simulator for Nuclear Power Plants (DSNP), was used to perform a dynamic analysis of several conceptual design studies of liquid metal fast breeder reactors. The DSNP being a high level language enables the user to transform a power plant flow chart directly into a simulation program using a small number of DSNP statements. In addition to the language statements, the DSNP system has its own precompiler and an extensive library containing models of power plant components, algorithms of physical processes, material property functions, and various auxiliary functions. The comparative analysis covered oxide-fueled versus metal-fueled core designs and loop- versus pool-type reactors. The question of interest was the rate of change of the temperatures in the components in the upper plenum and the primary loop, in particular the reactor outlet nozzle and the intermediate heat exchanger inlet nozzle during different types of transients. From the simulations performed it can be concluded that metal-fueled cores will have much faster temperature transients than oxide-fueled cores due mainly to the much higher thermal diffusivity of the metal fuel. The transients in the pool-type design (either with oxide fuel or metal fuel) will be much slower than in the loop-type design due to the large heat capacity of the sodium pool. The DSNP language was demonstrated to be well suited to perform many types of transient analysis in nuclear power plants

  4. Conceptual design of a nucleo electric simulator with PBMR reactor based in Reduced order models

    International Nuclear Information System (INIS)

    Valle H, J.; Morales S, J.B.

    2005-01-01

    This project has as purpose to know to depth the operation of a PBMR nucleo electric type (Pebble Bed Modular Reactor), which has a reactor of moderate graphite spheres and fuel of uranium dioxide cooled with Helium and Brayton thermodynamic cycle. The simulator seeks to describe the dynamics of the one process of energy generation in the nuclear fuel, the process of transport toward the coolant one and the conversion to mechanical energy in the turbo-generators as well as in the heat exchangers indispensable for the process. The dynamics of reload of the fuel elements it is not modeled in detail but their effects are represented in the parameters of the pattern. They are modeled also the turbo-compressors of the primary circuit of the work fluid. The control of the power of the nuclear reactor is modeled by means of reactivity functions specified in the simulation platform. The proposed mathematical models will be settled in the platform of simulation of Simulink-Mat Lab. The proposed control panels for this simulator can be designed and to implement using the box of tools of Simulink that facilitates this process. The work presents the mathematical models more important used for their future implementation in Simulink. (Author)

  5. Development of a methodology for simulation of gas cooled reactors with purpose of transmutation

    International Nuclear Information System (INIS)

    Silva, Clarysson Alberto da

    2009-01-01

    This work proposes a methodology of MHR (Modular Helium Reactor) simulation using the WIMSD-5B (Winfrith Improved Multi/group Scheme) nuclear code which is validated by MCNPX 2.6.0 (Monte Carlo N-Particle transport eXtend) nuclear code. The goal is verify the capability of WIMSD-5B to simulate a reactor type GT-MHR (Gas Turbine Modular Helium Reactor), considering all the fuel recharges possibilities. Also is evaluated the possibility of WIMSD-5B to represent adequately the fuel evolution during the fuel recharge. Initially was verified the WIMSD-5B capability to simulate the recharge specificities of this model by analysis of neutronic parameters and isotopic composition during the burnup. After the model was simulated using both WIMSD-5B and MCNPX 2.6.0 codes and the results of k eff , neutronic flux and isotopic composition were compared. The results show that the deterministic WIMSD-5B code can be applied to a qualitative evaluation, representing adequately the core behavior during the fuel recharges being possible in a short period of time to inquire about the burned core that, once optimized, can be quantitatively evaluated by a code type MCNPX 2.6.0. (author)

  6. Test Plan for the Boiling Water Reactor Dry Cask Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Durbin, Samuel [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lindgren, Eric R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-11-01

    The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the

  7. Leaching characteristics of actinides from simulated reactor waste glass

    Energy Technology Data Exchange (ETDEWEB)

    Weed, H.C.; Coles, D.G.; Bradley, D.J.; Mensing, R.W.; Schweiger, J.S.

    1979-01-18

    Two methods for measuring the leach rates of simulated high level waste glass are compared. One is a modification of the standard IAEA method and the other is a one-pass method in which fresh leachant solution is pumped over the sample at a controlled flow rate and temperature. For times up to 3 days, there is close agreement between results from the two methods at 25.0/sup 0/C. Leach rates from the one-pass method show a correlation with flow rate only on day 1 at 25.0/sup 0/C, whereas they show a correlation with flow rate for all three days at 75.0/sup 0/C. /sup 237/Np rates at 75.0/sup 0/C are greater than those at 25.0/sup 0/C, but /sup 239/Pu rates at 75.0/sup 0/C are less than or equal to those at 25.0/sup 0/C.

  8. DynMo: Dynamic Simulation Model for Space Reactor Power Systems

    International Nuclear Information System (INIS)

    El-Genk, Mohamed; Tournier, Jean-Michel

    2005-01-01

    A Dynamic simulation Model (DynMo) for space reactor power systems is developed using the SIMULINK registered platform. DynMo is modular and could be applied to power systems with different types of reactors, energy conversion, and heat pipe radiators. This paper presents a general description of DynMo-TE for a space power system powered by a Sectored Compact Reactor (SCoRe) and that employs off-the-shelf SiGe thermoelectric converters. SCoRe is liquid metal cooled and designed for avoidance of a single point failure. The reactor core is divided into six equal sectors that are neutronically, but not thermal-hydraulically, coupled. To avoid a single point failure in the power system, each reactor sector has its own primary and secondary loops, and each loop is equipped with an electromagnetic (EM) pump. A Power Conversion assembly (PCA) and a Thermoelectric Conversion Assembly (TCA) of the primary and secondary EM pumps thermally couple each pair of a primary and a secondary loop. The secondary loop transports the heat rejected by the PCA and the pumps TCA to a rubidium heat pipes radiator panel. The primary loops transport the thermal power from the reactor sector to the PCAs for supplying a total of 145-152 kWe to the load at 441-452 VDC, depending on the selections of the primary and secondary liquid metal coolants. The primary and secondary coolant combinations investigated are lithium (Li)/Li, Li/sodium (Na), Na-Na, Li/NaK-78 and Na/NaK-78, for which the reactor exit temperature is kept below 1250 K. The results of a startup transient of the system from an initial temperature of 500 K are compared and discussed

  9. Calculations of steady-state and reactivity insertion transients in a research reactor simulating the PWR

    International Nuclear Information System (INIS)

    Mladin, Mirea; Mladin, Daniela; Prodea, Ilie

    2010-01-01

    In 2008, IAEA started a Coordinated Research Project for benchmarking the thermalhydraulic and neutronic computer codes for research reactor analysis against the experimental data. In this framework, for the first year of research contract, the Institute for Nuclear Research engaged in steady-state analysis of SPERT-III reactor and also in the simulation of the reactivity insertion tests performed in this reactor during mid sixties. In the first part, the paper describes a Monte Carlo input model of the oxide core selected for investigation and the results of the steady-state neutronic calculations with respect to hot and cold core reactivity excess and control rods worth. Also, prompt neutron life and reactivity feed-back coefficients were examined. These results were compared with the data provided in the reactor specification document concerning neutronic design calculated data. The second part of the paper is dedicated to calculation of the reactivity insertion transients with RELAP5 and CATHARE2 thermalhydraulic codes, both including point reactor kinetics models, and to comparison with experimental data. (authors)

  10. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  11. Development of the supporting system of the Monju advanced reactor simulator (MARS)

    International Nuclear Information System (INIS)

    Koyagoshi, Naoki; Sasaki, Kazuichi; Sawada, Makoto

    2002-10-01

    The MARS has been operating for operator training and operation procedure's verification of the prototype fast breeder reactor 'Monju' since April 1991. In order to carry out the above results more effectively, the MARS supporting system which consists of several computer system has being developed. This report covers the following three supporting systems developed from 1994 to 2001 and study on evaluation method of Monju operator training data. Expanded Monju visual animation system. The Monju visual animation system was developed to visualize the inner structure of equipments and the parameters without measuring points. This system is used for training form 1993. And then, the training limits of the system has been extended. Development of the Monju min simulator for reactor core analysis. Development of the Monju min simulator which analyzes thermo-hydraulic behavior in the Monju reactor in detail is proceeding with the aims; of upgrading Monju operator training effect. The obtained results will be reflected to remodeling of MARS's reactor core analysis mode. Development of the severe accident CAI (Computer Assisted Instruction) system. The prototype system which supports study on accident management was developed. This system will be converted when the severe accident procedure of Monju is fixed, and it will be used for training. Study on evaluation method of Monju operate training data. In order to reconstruct the operator training system, the evaluation method of training data was considered. The availability has been checked as a result of evaluating crew communication using this method. (author)

  12. Modernization of the NESTLE-CANDU reactor simulator and coupling to scale-processed cross sections

    International Nuclear Information System (INIS)

    Hart, S.; Maldonado, G.I.

    2012-01-01

    The original version of the NESTLE computer code for CANDU applications, herein referred as the NESTLE-CANDU or NESTLE-C program, was developed under sponsorship by the CNSC as a “stand-alone” program. In fact, NESTLE-C emerged from the original version of NESTLE, applicable to light water reactors, which was written in FORTRAN 77 to solve the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). Accordingly, NESTLE-C can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source or eigenvalue initiated transient problems for CANDU reactor fuel arrangements and geometries. This article reports a recent conversion of the NESTLE-C code to the Fortran 90 standard, in addition, we highlight other code updates carried out to modularize and modernize NESTLE-C in a manner consistent with the latest updates performed with the parent NESTLE code for light water reactor (LWR) applications. Also reported herein, is a simulation of a CANDU reactor employing 37-element fuel bundles, which was carried out to highlight the SCALE to NESTLE-C coupling developed for two-group collapsed and bundle homogenized cross-section generation. The results presented are consistent with corresponding simulations that employed HELIOS generated cross-sections. (author)

  13. Modernization of the NESTLE-CANDU reactor simulator and coupling to scale-processed cross sections

    Energy Technology Data Exchange (ETDEWEB)

    Hart, S.; Maldonado, G.I. [Univ. of Tennessee, Knoxville, Tennessee (United States)

    2012-07-01

    The original version of the NESTLE computer code for CANDU applications, herein referred as the NESTLE-CANDU or NESTLE-C program, was developed under sponsorship by the CNSC as a “stand-alone” program. In fact, NESTLE-C emerged from the original version of NESTLE, applicable to light water reactors, which was written in FORTRAN 77 to solve the few-group neutron diffusion equation utilizing the Nodal Expansion Method (NEM). Accordingly, NESTLE-C can solve the eigenvalue (criticality); eigenvalue adjoint; external fixed-source or eigenvalue initiated transient problems for CANDU reactor fuel arrangements and geometries. This article reports a recent conversion of the NESTLE-C code to the Fortran 90 standard, in addition, we highlight other code updates carried out to modularize and modernize NESTLE-C in a manner consistent with the latest updates performed with the parent NESTLE code for light water reactor (LWR) applications. Also reported herein, is a simulation of a CANDU reactor employing 37-element fuel bundles, which was carried out to highlight the SCALE to NESTLE-C coupling developed for two-group collapsed and bundle homogenized cross-section generation. The results presented are consistent with corresponding simulations that employed HELIOS generated cross-sections. (author)

  14. Experimental study on air ingress during a primary pipe rupture accident with a graphite reactor core simulator

    International Nuclear Information System (INIS)

    Takeda, Tetsuaki; Hishida, Makoto; Baba, Shinichi

    1991-11-01

    When a primary coolant pipe of a High Temperature Gas Cooled Reactor (HTGR) ruptures, helium gas in the reactor core blows out into the container, and the primary cooling system reduces the pressure. After the pressures are balanced between the reactor and the container, air is expected to enter into the reactor core from the breach. It seems to be probable that the graphite structures is oxidized by air. Hence, it is necessary to investigate the air ingress process and the behavior of the generating gases by the oxidation reactions. The previous experimental study is performed on the molecular diffusion and natural convection of the two component gas mixtures using a test model simulating simply the reactor. Objective of the study was to investigate the air ingress process during the early stage of the primary pipe rupture accident. However, since the model did not have any kind of graphite components, the reaction between graphite and oxygen was not simulated. The present model includes the reactor core and the high temperature plenum simulators made of graphite. The major results obtained in the present study are summarized in the followings: (1) The air ingress process with graphite oxidation reaction is similar to that without the reaction qualitatively. (2) When the reactor core simulator is maintained at low temperatures (lower than 450degC), the initiation time of the natural circulation of air is almost equal to that of the natural circulation of nitrogen. On the other hand, when the temperature of the reactor core simulator is high (more than 500degC), the initiation time of the natural circulation of air is earlier than that of nitrogen. (3) When the temperature of the reactor core simulator is higher than 600degC, oxygen is almost dissipated by the graphite structures. When the temperature of the reactor core simulator is below 700degC, carbon dioxide mainly is generated by the oxidation reactions. (author)

  15. Monte Carlo radiative transfer simulation of a cavity solar reactor for the reduction of cerium oxide

    Energy Technology Data Exchange (ETDEWEB)

    Villafan-Vidales, H.I.; Arancibia-Bulnes, C.A.; Dehesa-Carrasco, U. [Centro de Investigacion en Energia, Universidad Nacional Autonoma de Mexico, Privada Xochicalco s/n, Col. Centro, A.P. 34, Temixco, Morelos 62580 (Mexico); Romero-Paredes, H. [Departamento de Ingenieria de Procesos e Hidraulica, Universidad Autonoma Metropolitana-Iztapalapa, Av. San Rafael Atlixco No.186, Col. Vicentina, A.P. 55-534, Mexico D.F 09340 (Mexico)

    2009-01-15

    Radiative heat transfer in a solar thermochemical reactor for the thermal reduction of cerium oxide is simulated with the Monte Carlo method. The directional characteristics and the power distribution of the concentrated solar radiation that enters the cavity is obtained by carrying out a Monte Carlo ray tracing of a paraboloidal concentrator. It is considered that the reactor contains a gas/particle suspension directly exposed to concentrated solar radiation. The suspension is treated as a non-isothermal, non-gray, absorbing, emitting, and anisotropically scattering medium. The transport coefficients of the particles are obtained from Mie-scattering theory by using the optical properties of cerium oxide. From the simulations, the aperture radius and the particle concentration were optimized to match the characteristics of the considered concentrator. (author)

  16. 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

  17. Multi-Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNP

    International Nuclear Information System (INIS)

    Downar, Thomas; Seker, Volkan

    2013-01-01

    Radioactive gaseous fission products are released out of the fuel element at a significantly higher rate when the fuel temperature exceeds 1600°C in high-temperature gas-cooled reactors (HTGRs). Therefore, it is of paramount importance to accurately predict the peak fuel temperature during all operational and design-basis accident conditions. The current methods used to predict the peak fuel temperature in HTGRs, such as the Next-Generation Nuclear Plant (NGNP), estimate the average fuel temperature in a computational mesh modeling hundreds of fuel pebbles or a fuel assembly in a pebble-bed reactor (PBR) or prismatic block type reactor (PMR), respectively. Experiments conducted in operating HTGRs indicate considerable uncertainty in the current methods and correlations used to predict actual temperatures. The objective of this project is to improve the accuracy in the prediction of local 'hot' spots by developing multi-scale, multi-physics methods and implementing them within the framework of established codes used for NGNP analysis.The multi-scale approach which this project will implement begins with defining suitable scales for a physical and mathematical model and then deriving and applying the appropriate boundary conditions between scales. The macro scale is the greatest length that describes the entire reactor, whereas the meso scale models only a fuel block in a prismatic reactor and ten to hundreds of pebbles in a pebble bed reactor. The smallest scale is the micro scale--the level of a fuel kernel of the pebble in a PBR and fuel compact in a PMR--which needs to be resolved in order to calculate the peak temperature in a fuel kernel.

  18. Multi-Scale Multi-physics Methods Development for the Calculation of Hot-Spots in the NGNP

    Energy Technology Data Exchange (ETDEWEB)

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Seker, Volkan [Univ. of Michigan, Ann Arbor, MI (United States)

    2013-04-30

    Radioactive gaseous fission products are released out of the fuel element at a significantly higher rate when the fuel temperature exceeds 1600°C in high-temperature gas-cooled reactors (HTGRs). Therefore, it is of paramount importance to accurately predict the peak fuel temperature during all operational and design-basis accident conditions. The current methods used to predict the peak fuel temperature in HTGRs, such as the Next-Generation Nuclear Plant (NGNP), estimate the average fuel temperature in a computational mesh modeling hundreds of fuel pebbles or a fuel assembly in a pebble-bed reactor (PBR) or prismatic block type reactor (PMR), respectively. Experiments conducted in operating HTGRs indicate considerable uncertainty in the current methods and correlations used to predict actual temperatures. The objective of this project is to improve the accuracy in the prediction of local "hot" spots by developing multi-scale, multi-physics methods and implementing them within the framework of established codes used for NGNP analysis.The multi-scale approach which this project will implement begins with defining suitable scales for a physical and mathematical model and then deriving and applying the appropriate boundary conditions between scales. The macro scale is the greatest length that describes the entire reactor, whereas the meso scale models only a fuel block in a prismatic reactor and ten to hundreds of pebbles in a pebble bed reactor. The smallest scale is the micro scale--the level of a fuel kernel of the pebble in a PBR and fuel compact in a PMR--which needs to be resolved in order to calculate the peak temperature in a fuel kernel.

  19. Modeling, simulation, and analysis of a reactor system for the generation of white liquor of a pulp and paper industry

    Directory of Open Access Journals (Sweden)

    Ricardo Andreola

    2011-02-01

    Full Text Available An industrial system for the production of white liquor of a pulp and paper industry, Klabin Paraná Papéis, formed by ten reactors was modeled, simulated, and analyzed. The developed model considered possible water losses by the evaporation and reaction, in addition to variations in the volumetric flow of lime mud across the reactors due to the composition variations. The model predictions agreed well with the process measurements at the plant and the results showed that the slaking reaction was nearly complete at the third causticizing reactor, while causticizing ends by the seventh reactor. Water loss due to slaking reaction and evaporation occurred more pronouncedly in the slaker reactor than in the final causticizing reactors; nevertheless, the lime mud flow remained nearly constant across the reactors.

  20. Final safety and hazards analysis for the Battelle LOCA simulation tests in the NRU reactor

    International Nuclear Information System (INIS)

    Axford, D.J.; Martin, I.C.; McAuley, S.J.

    1981-04-01

    This is the final safety and hazards report for the proposed Battelle LOCA simulation tests in NRU. A brief description of equipment test design and operating procedure precedes a safety analysis and hazards review of the project. The hazards review addresses potential equipment failures as well as potential for a metal/water reaction and evaluates the consequences. The operation of the tests as proposed does not present an unacceptable risk to the NRU Reactor, CRNL personnel or members of the public. (author)

  1. On geometric simulating in nuclear reactor calculations by the Monte-Carlo method

    International Nuclear Information System (INIS)

    Ostashenko, S.V.

    1988-01-01

    Analysis of existing geometric modules makes it possible to reveal their disadvantages and to formulate requirements list, which should be satisfied by any usefull geometry system. Short description of GDL language used for complex reactor systems simulating is given. GDL language applies hierarchical representation scheme to assemblies, which aids to reduce significantly amount of input data. The language is part of GDL geometry system designed for MCU package and implemented on ES computers

  2. Hybrid computer simulation of the dynamics of the Hoger Onderwijs Reactor

    International Nuclear Information System (INIS)

    Moers, J.C.; Vries, J.W. de.

    1976-01-01

    A distributed parameter model for the dynamics of the Hoger Onderwijs Reactor (HOR) at Delft is presented. The neutronic and the thermodynamic part of this model have been separately implemented on the AD4-IBM1800 Hybrid Computer of the Delft University of Technology Computation Centre. A continuous Space/Discrete Time solution method has been employed. Some test results of the simulation are included

  3. Numerical simulation of moderator flow and temperature distributions in a CANDU reactor vessel

    International Nuclear Information System (INIS)

    Carlucci, L.N.

    1982-10-01

    This paper describes numerical predictions of the two-dimensional flow and temperature fields of an internally-heated liquid in a typical CANDU reactor vessel. Turbulence momentum and energy transport are simulated using the k-epsilon model. Both steady-state and transient results are discussed. The finite control volume analogues of the conservation equations are solved using a modified version of the TEACH code

  4. Method for accounting for macroscopic heterogeneities in reactor material balance generation in fuel cycle simulations

    Energy Technology Data Exchange (ETDEWEB)

    Bagdatlioglu, Cem, E-mail: cemb@utexas.edu; Schneider, Erich

    2016-06-15

    Highlights: • Describes addition of spatially dependent power sharing to a previous methodology. • The methodology is used for calculating the input and output isotopics and burnup. • Generalizes to simulate reactors with strong spatial and flux heterogeneities. • Presents cases where the old approach would not have been sufficient. - Abstract: This paper describes the addition of spatially dependent power sharing to a methodology used for calculating the input and output isotopics and burnup of nuclear reactors within a nuclear fuel cycle simulator. Neutron balance and depletion calculations are carried out using pre-calculated fluence-based libraries. These libraries track the transmutation and neutron economy evolution of unit masses of nuclides available in input fuel. The work presented in the paper generalizes the method to simulate reactors that contain more than one type of fuel as well as strong spatial and flux heterogeneities, for instance breeders with a driver–blanket configuration. To achieve this, spatial flux calculations are used to determine the fluence-dependent relative average fluxes inside macroscopic spatial regions. These fluxes are then used to determine the average power of macroscopic spatial regions as well as to more accurately calculate region-specific transmutation rates. The paper presents several cases where the fluence based approach alone would not have been sufficient to determine results.

  5. Analysis and application of a simulator of a nuclear reactor AP-600

    International Nuclear Information System (INIS)

    Medina S, V. S.; Salazar S, E.

    2011-11-01

    In front of the resurgence of interest in the nuclear power production, several national organizations have considered convenient to have highly specialized human resources in the technologies of nuclear reactors of III + and IV generation. For this task, the intensive and extensive applications of the computation should been considered, as the virtual instrumentation. The present work analyzes the possible applications of a nuclear simulator provided by the IAEA with base in the design of the reactor AP-600, using a focusing of modular model developed in FORTRAN. One part of the work that was made with the simulator includes the evaluation of 21 transitory events of operation, including the recreation of the accident happened in the nuclear power plant of Three Mile Island in 1979, comparing the actions flow and the answer of the systems under the intrinsic security of a III + generation reactor. The impact that had the mentioned accident was analyzed in the growing of the nuclear energy sector and in the public image with regard to the nuclear power plants. An application for this simulator was proposed, its use as tool for the instruction in the nuclear engineering courses using it to observe the operation of the different security systems and its interrelation inside the power plant as well as a theoretical/practical approach for the student. (Author)

  6. Estimate of fuel burnup spatial a multipurpose reactor in computer simulation

    International Nuclear Information System (INIS)

    Santos, Nadia Rodrigues dos; Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes

    2015-01-01

    In previous research, which aimed, through computer simulation, estimate the spatial fuel burnup for the research reactor benchmark, material test research - International Atomic Energy Agency (MTR/IAEA), it was found that the use of the code in FORTRAN language, based on the diffusion theory of neutrons and WIMSD-5B, which makes cell calculation, bespoke be valid to estimate the spatial burnup other nuclear research reactors. That said, this paper aims to present the results of computer simulation to estimate the space fuel burnup of a typical multipurpose reactor, plate type and dispersion. the results were considered satisfactory, being in line with those presented in the literature. for future work is suggested simulations with other core configurations. are also suggested comparisons of WIMSD-5B results with programs often employed in burnup calculations and also test different methods of interpolation values obtained by FORTRAN. Another proposal is to estimate the burning fuel, taking into account the thermohydraulics parameters and the appearance of xenon. (author)

  7. Extraction of Water from Martian Regolith Simulant via Open Reactor Concept

    Science.gov (United States)

    Trunek, Andrew J.; Linne, Diane L.; Kleinhenz, Julie E.; Bauman, Steven W.

    2018-01-01

    To demonstrate proof of concept water extraction from simulated Martian regolith, an open reactor design is presented along with experimental results. The open reactor concept avoids sealing surfaces and complex moving parts. In an abrasive environment like the Martian surface, those reactor elements would be difficult to maintain and present a high probability of failure. A general lunar geotechnical simulant was modified by adding borax decahydrate (Na2B4O7·10H2O) (BDH) to mimic the 3 percent water content of hydrated salts in near surface soils on Mars. A rotating bucket wheel excavated the regolith from a source bin and deposited the material onto an inclined copper tray, which was fitted with heaters and a simple vibration system. The combination of vibration, tilt angle and heat was used to separate and expose as much regolith surface area as possible to liberate the water contained in the hydrated minerals, thereby increasing the efficiency of the system. The experiment was conducted in a vacuum system capable of maintaining a Martian like atmosphere. Evolved water vapor was directed to a condensing system using the ambient atmosphere as a sweep gas. The water vapor was condensed and measured. Processed simulant was captured in a collection bin and weighed in real time. The efficiency of the system was determined by comparing pre- and post-processing soil mass along with the volume of water captured.

  8. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  9. Implementation of geomechanical models for engineered clay barriers in multi-physic partial differential equation solvers

    International Nuclear Information System (INIS)

    Navarro, V.; Alonso, J.; Asensio, L.; Yustres, A.; Pintado, X.

    2012-01-01

    when using the mixed method. The constitutive formulation is implemented as a balance equation. This way, the user can freely implement models involving implicit relationships between variables. To illustrate the application of this method, we have analysed the implementation of a modified formulation of the Barcelona Expansive Model, a critical state model (CSM) of reference for expansive clays, as it is the case for bentonite clays for engineered barriers for radioactive nuclear spent fuel confinement. The tool developed was used to satisfactorily model the coupled hydro-mechanical problem of the free-swelling of a MX-80 bentonite disc. The bentonite sample had an initial dry density of 1600 kg/m 3 and a water content of 10% in mass. The initial dimensions of the disk were a height of 15.85 mm and a diameter of 100 mm. The hydration with deionised water applied on the top face of the sample was modelled with a saturation boundary condition. Figure 1 illustrates the case modelled and results obtained on pore water pressure and swelling of the sample at an intermediate time in the simulation. In addition, the modelling results have been compared with the experimental results obtained in the laboratory test carried out by B+TECH, as shown in Figure 2. In conclusion, the present proposal means a useful approach for the introduction of advanced Soil Mechanics models to the modelling of bentonite clays in multi-physics partial differential solvers. The use of it enables to overcome the limitations of some MPDES to integrate state functions defined implicitly. This makes it possible to combine the versatility of MPDES with powerful constitutive grounds. (authors)

  10. Importance of thermal nonequilibrium considerations for the simulation of nuclear reactor LOCA transients

    International Nuclear Information System (INIS)

    Fischer, S.R.; Nelson, R.A.; Sullivan, L.H.

    1980-01-01

    The purpose of this paper is to show the importance of considering thermal nonequilibrium effects in computer simulations of the refill and reflood portions of pressurized water reactor (PWR) loss-of-coolnat accident (LOCA) transients. Although RELAP4 assumes thermodynamic equilibrium between phases, models that account for the nonequilibrium phenomena associated with the mixing of subcooled emergency cooling water with steam and the superheating of vapor in the presence of liquid droplets have recently been incorporated into the code. Code calculated results, both with and without these new models, have been compared with experimental test data to assess the importance of including thermal nonequilibrium phenomena in computer code simulations

  11. Simulation for Supporting Scale-Up of a Fluidized Bed Reactor for Advanced Water Oxidation

    Directory of Open Access Journals (Sweden)

    Farhana Tisa

    2014-01-01

    Full Text Available Simulation of fluidized bed reactor (FBR was accomplished for treating wastewater using Fenton reaction, which is an advanced oxidation process (AOP. The simulation was performed to determine characteristics of FBR performance, concentration profile of the contaminants, and various prominent hydrodynamic properties (e.g., Reynolds number, velocity, and pressure in the reactor. Simulation was implemented for 2.8 L working volume using hydrodynamic correlations, continuous equation, and simplified kinetic information for phenols degradation as a model. The simulation shows that, by using Fe3+ and Fe2+ mixtures as catalyst, TOC degradation up to 45% was achieved for contaminant range of 40–90 mg/L within 60 min. The concentration profiles and hydrodynamic characteristics were also generated. A subsequent scale-up study was also conducted using similitude method. The analysis shows that up to 10 L working volume, the models developed are applicable. The study proves that, using appropriate modeling and simulation, data can be predicted for designing and operating FBR for wastewater treatment.

  12. Standardization of accelerator irradiation procedures for simulation of neutron induced damage in reactor structural materials

    Science.gov (United States)

    Shao, Lin; Gigax, Jonathan; Chen, Di; Kim, Hyosim; Garner, Frank A.; Wang, Jing; Toloczko, Mychailo B.

    2017-10-01

    Self-ion irradiation is widely used as a method to simulate neutron damage in reactor structural materials. Accelerator-based simulation of void swelling, however, introduces a number of neutron-atypical features which require careful data extraction and, in some cases, introduction of innovative irradiation techniques to alleviate these issues. We briefly summarize three such atypical features: defect imbalance effects, pulsed beam effects, and carbon contamination. The latter issue has just been recently recognized as being relevant to simulation of void swelling and is discussed here in greater detail. It is shown that carbon ions are entrained in the ion beam by Coulomb force drag and accelerated toward the target surface. Beam-contaminant interactions are modeled using molecular dynamics simulation. By applying a multiple beam deflection technique, carbon and other contaminants can be effectively filtered out, as demonstrated in an irradiation of HT-9 alloy by 3.5 MeV Fe ions.

  13. Computer simulation of nuclear reactor control by means of heuristic learning controller

    International Nuclear Information System (INIS)

    Bubak, M.; Moscinski, J.

    1976-01-01

    A trial of application of two techniques of Artificial Intelligence: heuristic Programming and Learning Machines Theory for nuclear reactor control is presented. Considering complexity of the mathematical models describing satisfactorily the nuclear reactors, value changes of these models parameters in course of operation, knowledge of some parameters value with too small exactness, there appear diffucluties in the classical approach application for these objects control systems design. The classical approach consists in definition of the permissible control actions set on the base of the set performance index and the object mathematical model. The Artificial Intelligence methods enable construction of the control system, which gets during work an information being a priori inaccessible and uses it for its action change for the control to be the optimum one. Applying these methods we have elaborated the reactor power control system. As the performance index there has been taken the integral of the error square. For the control system there are only accessible: the set power trajectory, the reactor power and the control rod position. The set power trajectory has been divided into time intervals called heuristic intervals. At the beginning of every heuristic interval, on the base of the obtained experience, the control system chooses from the control (heuristic) set the optimum control. The heuristic set it is the set of relations between the control rod rate and the state variables, the set and the obtained power, similar to simplifications applied by nuclear reactors operators. The results obtained for the different control rod rates and different reactor (simulated on the digital computer) show the proper work of the system. (author)

  14. Neutron spectrometry around the VENUS reactor using Monte Carlo simulations and Bonner spheres measurements

    International Nuclear Information System (INIS)

    Coeck, M.; Lacoste, V.; Muller, H.

    2005-01-01

    Full text: Reliable determination of neutron doses in workplaces is still an issue in the field of radiation protection. The EVIDOS project ('evaluation of individual dosimetry in mixed neutron and photon radiation fields', 5FP supported by the EC) aims to evaluate different methods for individual dosimetry in mixed neutron-photon workplaces in nuclear industry, and focuses on the neutron component. This objective cannot be reached on the basis of investigations in calibration fields only, but requires studies in representative workplaces of the nuclear industry. The VENUS reactor, a zero-power research reactor established by the SCK·CEN, was chosen as one of these workplaces. This paper presents the assessment of the neutron field near the VENUS reactor, particularly in areas near the reactor shielding and in the control room where operators are frequently present during a reactor run. From the neutron spectrum, an evaluation of H*(10) can be made. MCNPX simulations were performed to obtain a reference spectrum at the two areas of interest. Using a k eff calculation the source term was acquired which was subsequently used in a fixed source MCNPX model of the complete shielding geometry of the reactor hall. Reference spectrometry was also performed using a Bonner spheres system. The unfolding spectra were obtained using the NUBAY and GRAVEL codes. The NUBAY program, based on Bayesian parameter estimation methods, assumes a parameterized spectrum and provides posterior probability distributions for both the set of parameters and a set of integral quantities. The code GRAVEL, an iterative algorithm based on SAND-II, was used with various default spectra, among them the NUBAY solution. Bonner spheres data GRAVEL unfolding was also performed using the MCNPX spectra as an initial guess. In this paper the outcome of both calculations and measurements is compared. (author)

  15. A Multi-Physics PWR Model for the Load Following

    OpenAIRE

    Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

    2016-01-01

    International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

  16. Burnup Estimation of Rhodium Self-Powered Neutron Detector Emitter in VVER Reactor Core Using Monte Carlo Simulations

    OpenAIRE

    Khrutchinsky, А. А.; Kuten, S. A.; Babichev, L. F.

    2011-01-01

    Estimation of burn-up in a rhodium-103 emitter of self-powered neutron detector in VVER-1000 reactor core has been performed using Monte Carlo simulations within approximation of a constant neutron flux.

  17. Improvement of nuclear ship engineering simulation system. Hardware renewal and interface improvement of the integral type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, Hiroki; Kyoya, Masahiko; Shimazaki, Junya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Kano, Tadashi [KCS, Co., Mito, Ibaraki (Japan); Takahashi, Teruo [Energis, Co., Kobe, Hyogo (Japan)

    2001-10-01

    JAERI had carried out the design study about a lightweight and compact integral type reactor (an advanced marine reactor) with passive safety equipment as a power source for the future nuclear ships, and completed an engineering design. We have developed the simulator for the integral type reactor to confirm the design and operation performance and to utilize the study of automation of the reactor operation. The simulator can be used also for future research and development of a compact reactor. However, the improvement in a performance of hardware and a human machine interface of software of the simulator were needed for future research and development. Therefore, renewal of hardware and improvement of software have been conducted. The operability of the integral-reactor simulator has been improved. Furthermore, this improvement with the hardware and software on the market brought about better versatility, maintainability, extendibility and transfer of the system. This report mainly focuses on contents of the enhancement in a human machine interface, and describes hardware renewal and the interface improvement of the integral type reactor simulator. (author)

  18. SARIE upgrade: Nuclear reactor and water systems 'engineering and training' simulator

    International Nuclear Information System (INIS)

    Roth, P.

    2006-01-01

    Confronted as of its origins with the on-board layout constraints of the French Navy ships, TECHNICATOME integrates, as of the design, the ergonomics and the risks control related to the human factors. During more than 30 years, TECHNICATOME demonstrated a one of a kind know-how from the design to the execution of powerful, flexible and highly available nuclear compact reactors. A total control which includes up to the supervision and monitoring systems, the acoustic discreetly of the systems and its components, implemented on on-board reactors, testing reactors as well as experimental reactors. The functionalities of simulation were right from the start used by TECHNICATOME during the design phase of these installations to carry out operation engineering analyses on the thermal hydraulic and neutron aspects, to validate the principles of operation of the supervision systems like by the use of digital models in 3D CAD to validate the kinematics of operation or the interactions between systems. More recently, and starting from the end of the Nineties, a thought needs was launched to determine the interests related to the development of a training simulator associated with these installations with objectives, among others, to ensure the phase of initial training of the new operators, to widen the field of the training to the accidental situations, the management of crisis and crews behaviour supervision, the possibilities of replay which support the consolidation of the acquired knowledge(debriefing) with situation resume, and to increase the overall training capacity. An upgrade and modernisation project of these various simulation means was thus launched since 2001 with the objective to optimize the whole of the tasks supported by these means. (author)

  19. Analytical and numerical study of validation test-cases for multi-physic problems: application to magneto-hydro-dynamic

    Directory of Open Access Journals (Sweden)

    D Cébron

    2016-04-01

    Full Text Available The present paper is concerned with the numerical simulation of Magneto-Hydro-Dynamic (MHD problems with industrial tools. MHD has receivedattention some twenty to thirty years ago as a possible alternative inpropulsion applications; MHD propelled ships have even been designed forthat purpose. However, such propulsion systems have been proved of lowefficiency and fundamental researches in the area have progressivelyreceived much less attention over the past decades. Numerical simulationof MHD problem could however provide interesting solutions in the field ofturbulent flow control. The development of recent efficient numericaltechniques for multi-physic applications provide promising tool for theengineer for that purpose. In the present paper, some elementary testcases in laminar flow with magnetic forcing terms are analysed; equationsof the coupled problem are exposed, analytical solutions are derived ineach case and are compared to numerical solutions obtained with anumerical tool for multi-physic applications. The present work can be seenas a validation of numerical tools (based on the finite element method foracademic as well as industrial application purposes.

  20. Transient multi-physics analysis of a magnetorheological shock absorber with the inverse Jiles-Atherton hysteresis model

    Science.gov (United States)

    Zheng, Jiajia; Li, Yancheng; Li, Zhaochun; Wang, Jiong

    2015-10-01

    This paper presents multi-physics modeling of an MR absorber considering the magnetic hysteresis to capture the nonlinear relationship between the applied current and the generated force under impact loading. The magnetic field, temperature field, and fluid dynamics are represented by the Maxwell equations, conjugate heat transfer equations, and Navier-Stokes equations. These fields are coupled through the apparent viscosity and the magnetic force, both of which in turn depend on the magnetic flux density and the temperature. Based on a parametric study, an inverse Jiles-Atherton hysteresis model is used and implemented for the magnetic field simulation. The temperature rise of the MR fluid in the annular gap caused by core loss (i.e. eddy current loss and hysteresis loss) and fluid motion is computed to investigate the current-force behavior. A group of impulsive tests was performed for the manufactured MR absorber with step exciting currents. The numerical and experimental results showed good agreement, which validates the effectiveness of the proposed multi-physics FEA model.

  1. Simulation research about China Experimental Fast Reactor steam turboset based on Flowmaster platform

    International Nuclear Information System (INIS)

    Yan Hao; Tian Zhaofei

    2014-01-01

    In the third loop of China Experimental Fast Reactor (CEFR), steam turboset take an important role in converting heat energy into electric energy. However, turbo sets have not been operated on the condition of more than 40%P_0 (P_0 is full power) since they were installed. Thus it is necessary to make an analogue simulation. Based on the real models of turbo sets in CEFR, simulation models were created with the help of Flowmaster platform. By using such simulation models, a steady state result in full power circumstance was got, which is in accordance with design parameters. Meanwhile, a transient state simulation with operating condition ranging from full power to 40%P_0 was accomplished and a result which verifies part of performance and running conditions of turbo sets was got. The result of analogue simulation shows that based on Flowmaster platform, the running condition of simulation models can comply with design requirement, and offer reference values to the actual running. Such simulation models can also offer reference values to other simulation models in the third loop of CEFR. (authors)

  2. 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)

  3. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  4. 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)

  5. SIMIFR: A code to simulate material movement in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    White, A.M.; Orechwa, Yuri.

    1991-01-01

    The SIMIFR code has been written to simulate the movement of material through a process. This code can be used to investigate inventory differences in material balances, assist in process design, and to produce operational scheduling. The particular process that is of concern to the authors is that centered around Argonne National Laboratory's Integral Fast Reactor. This is a process which involves the irradiation of fissile material for power production, and the recycling of the irradiated reactor fuel pins into fresh fuel elements. To adequately simulate this process it is necessary to allow for locations which can contain either discrete items or homogeneous mixtures. It is also necessary to allow for a very flexible process control algorithm. Further, the code must have the capability of transmuting isotopic compositions and computing internally the fraction of material from a process ending up in a given location. The SIMIFR code has been developed to perform all of these tasks. In addition to simulating the process, the code is capable of generating random measurement values and sampling errors for all locations, and of producing a restart deck so that terminated problems may be continued. In this paper the authors first familiarize the reader with the IFR fuel cycle. The different capabilities of the SIMIFR code are described. Finally, the simulation of the IFR fuel cycle using the SIMIFR code is discussed. 4 figs

  6. A full scope nuclear power plant simulator for multiple reactor types with virtual control panels

    International Nuclear Information System (INIS)

    Yonezawa, Hisanori; Ueda, Hiroki; Kato, Takahisa

    2017-01-01

    This paper summarizes a full scope nuclear power plant simulator for multiple reactor types with virtual control panels which Toshiba developed and delivered. After the Fukushima DAIICHI nuclear power plants accident, it is required that all the people who are engaged in the design, manufacturing, operation, maintenance, management and regulation for the nuclear power plant should learn the wide and deep knowledge about the nuclear power plant design including the severe accident. For this purpose, the training with a full scope simulator is one of the most suitable ways. However the existing full scope simulators which are consist of the control panels replica of the referenced plants are costly and they are hard to remodel to fit to the real plant of the latest condition. That's why Toshiba developed and delivered the new concept simulator system which covers multiple referenced plants even though they have different design like BWR and PWR. The control panels of the simulator are made by combining 69 large Liquid Crystal Display (LCD) panels with touch screen instead of a control panel replica of referenced plant. The screen size of the each panel is 42 inches and 3 displays are arranged in tandem for one unit and 23 units are connected together. Each panel displays switches, indicators, recorders and lamps with the Computer Graphics (CG) and trainees operate them with touch operations. The simulator includes a BWR and a PWR simulator model, which enable trainees to learn the wide and deep knowledge about the nuclear power plant of BWR and PWR reactor types. (author)

  7. CFD simulation of fluid dynamic and biokinetic processes within activated sludge reactors under intermittent aeration regime.

    Science.gov (United States)

    Sánchez, F; Rey, H; Viedma, A; Nicolás-Pérez, F; Kaiser, A S; Martínez, M

    2018-08-01

    Due to the aeration system, biological reactors are the most energy-consuming facilities of convectional WWTPs. Many biological reactors work under intermittent aeration regime; the optimization of the aeration process (air diffuser layout, air flow rate per diffuser, aeration length …) is necessary to ensure an efficient performance; satisfying the effluent requirements with the minimum energy consumption. This work develops a CFD modelling of an activated sludge reactor (ASR) which works under intermittent aeration regime. The model considers the fluid dynamic and biological processes within the ASR. The biological simulation, which is transient, takes into account the intermittent aeration regime. The CFD modelling is employed for the selection of the aeration system of an ASR. Two different aeration configurations are simulated. The model evaluates the aeration power consumption necessary to satisfy the effluent requirements. An improvement of 2.8% in terms of energy consumption is achieved by modifying the air diffuser layout. An analysis of the influence of the air flow rate per diffuser on the ASR performance is carried out. The results show a reduction of 14.5% in the energy consumption of the aeration system when the air flow rate per diffuser is reduced. The model provides an insight into the aeration inefficiencies produced within ASRs. Copyright © 2018 Elsevier Ltd. All rights reserved.

  8. Use of ion beams to simulate reaction of reactor fuels with their cladding

    International Nuclear Information System (INIS)

    Birtcher, R.C.; Baldo, P.

    2006-01-01

    Processes occurring within reactor cores are not amenable to direct experimental observation. Among major concerns are damage, fission gas accumulation and reaction between the fuel and its cladding all of which lead to swelling. These questions can be investigated through simulation with ion beams. As an example, we discuss the irradiation driven interaction of uranium-molybdenum alloys, intended for use as low-enrichment reactor fuels, with aluminum, which is used as fuel cladding. Uranium-molybdenum coated with a 100 nm thin film of aluminum was irradiated with 3 MeV Kr ions to simulate fission fragment damage. Mixing and diffusion of aluminum was followed as a function of irradiation with RBS and nuclear reaction analysis using the 27 Al(p,γ) 28 Si reaction which occurs at a proton energy of 991.9 keV. During irradiation at 150 deg. C, aluminum diffused into the uranium alloy at a irradiation driven diffusion rate of 30 nm 2 /dpa. At a dose of 90 dpa, uranium diffusion into the aluminum layer resulted in formation of an aluminide phase at the initial interface. The thickness of this phase grew until it consumed the aluminum layer. The rapid diffusion of Al into these reactor fuels may offer explanation of the observation that porosity is not observed in the fuel particles but on their periphery

  9. Thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Dominguez, Dany S.; Mazaira, Leorlen Y. Rojas; Hernandez, Carlos R.G.; Lira, Carlos Alberto Brayner de Oliveira

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal–hydraulic characteristics. In this article, it was performed the thermal–hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a column of FCC (Face Centered Cubic) cells, with 41 layers and 82 pebbles. The input data used were taken from the thermohydraulic IAEA Benchmark (TECDOC-1694). The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  10. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  11. CFD Numerical Simulation of Biodiesel Synthesis in a Spinning Disc Reactor

    Directory of Open Access Journals (Sweden)

    Wen Zhuqing

    2015-03-01

    Full Text Available In this paper a two-disc spinning disc reactor for intensified biodiesel synthesis is described and numerically simulated. The reactor consists of two flat discs, located coaxially and parallel to each other with a gap of 0.2 mm between the discs. The upper disc is located on a rotating shaft while the lower disc is stationary. The feed liquids, triglycerides (TG and methanol are introduced coaxially along the centre line of rotating disc and stationary disc. Fluid hydrodynamics in the reactor for synthesis of biodiesel from TG and methanol in the presence of a sodium hydroxide catalyst are simulated, using convection-diffusion-reaction species transport model by the CFD software ANSYS©Fluent v. 13.0. The effect of the upper disc’s spinning speed is evaluated. The results show that the rotational speed increase causes an increase of TG conversion despite the fact that the residence time decreases. Compared to data obtained from adequate experiments, the model shows a satisfactory agreement.

  12. Review of turbulence modelling for numerical simulation of nuclear reactor thermal-hydraulics

    International Nuclear Information System (INIS)

    Bernard, J.P.; Haapalehto, T.

    1996-01-01

    The report deals with the modelling of turbulent flows in nuclear reactor thermal-hydraulic applications. The goal is to give tools and knowledge about turbulent flows and their modelling in practical applications for engineers, and especially nuclear engineers. The emphasize is on the theory of turbulence, the existing different turbulence models, the state-of-art of turbulence in research centres, the available models in the commercial code CFD-FLOW3D, and the latest applications of turbulence modelling in nuclear reactor thermal-hydraulics. It turns out that it is difficult to elaborate an universal turbulence model and each model has its advantages and drawbacks in each application. However, the increasing power of computers can permit the emergence of new methods of turbulence modelling such as Direct Numerical Simulation (DNS) and Large Eddy Simulation (LES) which open new horizons in this field. These latter methods are beginning to be available in commercial codes and are used in different nuclear applications such as 3-D modelling of the nuclear reactor cores and the steam generators. (orig.) (22 refs.)

  13. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    International Nuclear Information System (INIS)

    Silva, Alexandro S.; Mazaira, Leorlen Y.R.; Dominguez, Dany S.; Hernandez, Carlos R.G.

    2015-01-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  14. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

    2017-01-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  15. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-11-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  16. Simulation of the flow obstruction of a jet pump in a BWR reactor with the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Cardenas V, J.; Filio L, C.

    2016-09-01

    This work simulates the flow obstruction of a jet pump in one of the recirculation loops of a nuclear power plant with a reactor of type BWR at 100% of operating power, in order to analyze the behavior of the total flow of the refrigerant passing through the reactor core, the total flow in each recirculation loop of the reactor, together with the 10 jet pumps of each loop. The behavior of the power and the reactivity insertion due to the change of the refrigerant flow pattern is also analyzed. The simulation was carried out using the RELAP/SCDAPSIM version 3.5 code, using a reactor model with 10 jet pumps in each recirculation loop and a core consisting of 6 radial zones and 25 axial zones. The scenario postulates the flow obstruction in a jet pump in a recirculation loop A when the reactor operates at 100% rated power, causing a change in the total flow of refrigerant in the reactor core, leading to a decrease in power. Once the reactor conditions are established to its new power, the operator tries to recover the nominal power using the flow control valve of the recirculation loop A, opening stepwise as a strategy to safely recover the reactor power. In this analysis is assumed that the intention of the nuclear plant operator is to maintain the operation of the reactor during the established cycle. (Author)

  17. 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

  18. 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.

  19. Multi-physics modelling of a compliant humanoid robot

    Energy Technology Data Exchange (ETDEWEB)

    Zobova, Alexandra A., E-mail: azobova@mech.math.msu.su [Lomonosov Moscow State University, Faculty of Mechanics and Mathematics (Russian Federation); Habra, Timothée, E-mail: timothee.habra@uclouvain.be [Université catholique de Louvain (UCL), Center for Research in Mechatronics, Institute of Mechanics, Materials, and Civil Engineering (Belgium); Van der Noot, Nicolas, E-mail: nicolas.vandernoot@uclouvain.be, E-mail: nicolas.vandernoot@epfl.ch [EPFL STI IBI BIOROB, Biorobotics Laboratory, Institute of Bioengineering, École polytechnique fédérale de Lausanne (EPFL) (Switzerland); Dallali, Houman, E-mail: houman.dallali@iit.it; Tsagarakis, Nikolaos G., E-mail: nikos.tsagarakis@iit.it [Istituto Italiano di Tecnologia, Department of Advanced Robotics (Italy); Fisette, Paul, E-mail: paul.fisette@uclouvain.be; Ronsse, Renaud, E-mail: renaud.ronsse@uclouvain.be [Université catholique de Louvain (UCL), Center for Research in Mechatronics, Institute of Mechanics, Materials, and Civil Engineering (Belgium)

    2017-01-15

    We present a multibody simulator being used for compliant humanoid robot modelling and report our reasoning for choosing the settings of the simulator’s key features. First, we provide a study on how the numerical integration speed and accuracy depend on the coordinate representation of the multibody system. This choice is particularly critical for mechanisms with long serial chains (e.g. legs and arms). Our second contribution is a full electromechanical model of the inner dynamics of the compliant actuators embedded in the COMAN robot, since joints’ compliance is needed for the robot safety and energy efficiency. Third, we discuss the different approaches for modelling contacts and selecting an appropriate contact library. The recommended solution is to couple our simulator with an open-source contact library offering both accurate and fast contact modelling. The simulator performances are assessed by two different tasks involving contacts: a bimanual manipulation task and a squatting tasks. The former shows reliability of the simulator. For the latter, we report a comparison between the robot behaviour as predicted by our simulation environment, and the real one.

  20. Multi-physics modelling of a compliant humanoid robot

    International Nuclear Information System (INIS)

    Zobova, Alexandra A.; Habra, Timothée; Van der Noot, Nicolas; Dallali, Houman; Tsagarakis, Nikolaos G.; Fisette, Paul; Ronsse, Renaud

    2017-01-01

    We present a multibody simulator being used for compliant humanoid robot modelling and report our reasoning for choosing the settings of the simulator’s key features. First, we provide a study on how the numerical integration speed and accuracy depend on the coordinate representation of the multibody system. This choice is particularly critical for mechanisms with long serial chains (e.g. legs and arms). Our second contribution is a full electromechanical model of the inner dynamics of the compliant actuators embedded in the COMAN robot, since joints’ compliance is needed for the robot safety and energy efficiency. Third, we discuss the different approaches for modelling contacts and selecting an appropriate contact library. The recommended solution is to couple our simulator with an open-source contact library offering both accurate and fast contact modelling. The simulator performances are assessed by two different tasks involving contacts: a bimanual manipulation task and a squatting tasks. The former shows reliability of the simulator. For the latter, we report a comparison between the robot behaviour as predicted by our simulation environment, and the real one.

  1. Real-time simulation of response to load variation for a ship reactor based on point-reactor double regions and lumped parameter model

    Energy Technology Data Exchange (ETDEWEB)

    Wang Qiao; Zhang De [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Wenzhen, E-mail: Cwz2@21cn.com [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China); Chen Zhiyun [Department of Nuclear Energy Science and Engineering, Naval University of Engineering, Wuhan 430033 (China)

    2011-05-15

    Research highlights: > We calculate the variation of main parameters of the reactor core by the Simulink. > The Simulink calculation software (SCS) can deal well with the stiff problem. > The high calculation precision is reached with less time, and the results can be easily displayed. > The quick calculation of ship reactor transient can be achieved by this method. - Abstract: Based on the point-reactor double regions and lumped parameter model, while the nuclear power plant second loop load is increased or decreased quickly, the Simulink calculation software (SCS) is adopted to calculate the variation of main physical and thermal-hydraulic parameters of the reactor core. The calculation results are compared with those of three-dimensional simulation program. It is indicated that the SCS can deal well with the stiff problem of the point-reactor kinetics equations and the coupled problem of neutronics and thermal-hydraulics. The high calculation precision can be reached with less time, and the quick calculation of parameters of response to load disturbance for the ship reactor can be achieved. The clear image of the calculation results can also be displayed quickly by the SCS, which is very significant and important to guarantee the reactor safety operation.

  2. Development of Reactor Core Model based on Optimal Analysis for Shinhanul no. 1, 2 Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung-min [Korea Hydro Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    As one of the outputs of 'Development of the Shin Hanul Nuclear Plant(SHN) 1,2 Simulator' project which is being done by KHNP Central Research Institute, the SHN1,2 Simulator is being developed including the KNICS methodology and advanced Alarm Systems first applied to the Nuclear Power Plant in Korea, and the SHN 1,2 simulator adopts the virtually stimulated HMI(Human-Machine Interface) for the non-safety MMIS system, whose key-programs are identical to those applied to the real SHN 1,2 plants. The purpose of this paper is to develop localization core model by integrating the Simulator system with the Simulator core model though technology agreement of KAERI. To develop ShinHanul 1 and 2 reactor core simulator model, KHNP and KAERI create MASTER-SIM model and tried validation. And calculations of MASSIM{sub S}S program for MASTER{sub S}IM validation, are within tolerance range. Test has not yet been completed. And many verification will be conducted MASTER-SIM software is expected to be the highest economic software and satisfy international simulator standards.

  3. 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)

  4. Electrically heated ex-reactor pellet-cladding interaction (PCI) simulations utilizing irradiated Zircaloy cladding

    International Nuclear Information System (INIS)

    Barner, J.O.; Fitzsimmons, D.E.

    1985-02-01

    In a program sponsored by the Fuel Systems Research Branch of the US Nuclear Regulatory Commission, a series of six electrically heated fuel rod simulation tests were conducted at Pacific Northwest Laboratory. The primary objective of these tests was to determine the susceptibility of irradiated pressurized-water reactor (PWR) Zircaloy-4 cladding to failures caused by pellet-cladding mechanical interaction (PCMI). A secondary objective was to acquire kinetic data (e.g., ridge growth or relaxation rates) that might be helpful in the interpretation of in-reactor performance results and/or the modeling of PCMI. No cladding failures attributable to PCMI occurred during the six tests. This report describes the testing methods, testing apparatus, fuel rod diametral strain-measuring device, and test matrix. Test results are presented and discussed

  5. Measurements and simulations of electromagnetic perturbations on neutronic and thermodynamic control systems of nuclear reactors

    International Nuclear Information System (INIS)

    Lecompte, J.-C.; Buisson, Jacques.

    1981-01-01

    Neutronics and thermodynamics measuring devices realize the control and nuclear safety of the reactor and the core. They cause the emergency stop of the reactor if the informations given by the sensors are out of the normal operation tolerances. Defining and assuring a protection level sufficiently high against electromagnetic disturbances is essential for avoiding those measuring devices of giving erroneous indications in the environment where they are located. To obtain this result, one procceds according to the following three steps: a) In situ measurements of the ambient level of disturbances (oscilloscope analysis, statistical measurements). b) Measurement in laboratory of the immunity against the disturbances caused by every constituent of the device (electronic cables, connectors) and then the complete device. c) In situ simulation on the devices of the perturbations caused by the measured ambient level [fr

  6. Thermal Hydraulic Characteristics of Fuel Defects in Plate Type Nuclear Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bodey, Isaac T [ORNL

    2014-05-01

    Turbulent flow coupled with heat transfer is investigated for a High Flux Isotope Reactor (HFIR) fuel plate. The Reynolds Averaged Navier-Stokes Models are used for fluid dynamics and the transfer of heat from a thermal nuclear fuel plate using the Multi-physics code COMSOL. Simulation outcomes are compared with experimental data from the Advanced Neutron Source Reactor Thermal Hydraulic Test Loop. The computational results for the High Flux Isotope Reactor core system provide a more physically accurate simulation of this system by modeling the turbulent flow field in conjunction with the diffusion of thermal energy within the solid and fluid phases of the model domain. Recommendations are made regarding Nusselt number correlations and material properties for future thermal hydraulic modeling efforts

  7. 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.

  8. Dynamic model for the control system simulation and design of a 200 MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Yuai; Liu Longzhi; Ma Changwen

    1999-01-01

    The author develops a nonlinear dynamic model used in a wide range control system simulation for a 200 MW Nuclear Heating Reactor (NHR-200). Besides a one-point neutron kinetics equation and temperature feedback based on the lumped fuel and coolant temperature, which are the usual methods used in modeling of PWR, two other factors are also considered in order to suit the wide range operation. The first consideration is the natural circulation in the primary loop because it affects the heat transfer coefficients in the core and in the primary heat exchanger (PHE). The second consideration is the flow rate variation in the secondary loop which leads to some nonlinear properties. The simulation results show that the model is accurate enough for control system simulation. Some model reduction basis can be obtained through the dynamic analysis

  9. Optimization of pressurized water reactor shuffling by simulated annealing with heuristics

    International Nuclear Information System (INIS)

    Stevens, J.G.; Smith, K.S.; Rempe, K.R.; Downar, T.J.

    1995-01-01

    Simulated-annealing optimization of reactor core loading patterns is implemented with support for design heuristics during candidate pattern generation. The SIMAN optimization module uses the advanced nodal method of SIMULATE-3 and the full cross-section detail of CASMO-3 to evaluate accurately the neutronic performance of each candidate, resulting in high-quality patterns. The use of heuristics within simulated annealing is explored. Heuristics improve the consistency of optimization results for both fast- and slow-annealing runs with no penalty from the exclusion of unusual candidates. Thus, the heuristic application of designer judgment during automated pattern generation is shown to be effective. The capability of the SIMAN module to find and evaluate families of loading patterns that satisfy design constraints and have good objective performance within practical run times is demonstrated. The use of automated evaluations of successive cycles to explore multicycle effects of design decisions is discussed

  10. 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