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Sample records for dalton-thermix coupled code

  1. Quality Improvement of MARS Code and Establishment of Code Coupling

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Kim, Kyung Doo

    2010-04-01

    The improvement of MARS code quality and coupling with regulatory auditing code have been accomplished for the establishment of self-reliable technology based regulatory auditing system. The unified auditing system code was realized also by implementing the CANDU specific models and correlations. As a part of the quality assurance activities, the various QA reports were published through the code assessments. The code manuals were updated and published a new manual which describe the new models and correlations. The code coupling methods were verified though the exercise of plant application. The education-training seminar and technology transfer were performed for the code users. The developed MARS-KS is utilized as reliable auditing tool for the resolving the safety issue and other regulatory calculations. The code can be utilized as a base technology for GEN IV reactor applications

  2. Coupled geochemical and solute transport code development

    International Nuclear Information System (INIS)

    Morrey, J.R.; Hostetler, C.J.

    1985-01-01

    A number of coupled geochemical hydrologic codes have been reported in the literature. Some of these codes have directly coupled the source-sink term to the solute transport equation. The current consensus seems to be that directly coupling hydrologic transport and chemical models through a series of interdependent differential equations is not feasible for multicomponent problems with complex geochemical processes (e.g., precipitation/dissolution reactions). A two-step process appears to be the required method of coupling codes for problems where a large suite of chemical reactions must be monitored. Two-step structure requires that the source-sink term in the transport equation is supplied by a geochemical code rather than by an analytical expression. We have developed a one-dimensional two-step coupled model designed to calculate relatively complex geochemical equilibria (CTM1D). Our geochemical module implements a Newton-Raphson algorithm to solve heterogeneous geochemical equilibria, involving up to 40 chemical components and 400 aqueous species. The geochemical module was designed to be efficient and compact. A revised version of the MINTEQ Code is used as a parent geochemical code

  3. SIMULATE-3 K coupled code applications

    Energy Technology Data Exchange (ETDEWEB)

    Joensson, Christian [Studsvik Scandpower AB, Vaesteraas (Sweden); Grandi, Gerardo; Judd, Jerry [Studsvik Scandpower Inc., Idaho Falls, ID (United States)

    2017-07-15

    This paper describes the coupled code system TRACE/SIMULATE-3 K/VIPRE and the application of this code system to the OECD PWR Main Steam Line Break. A short description is given for the application of the coupled system to analyze DNBR and the flexibility the system creates for the user. This includes the possibility to compare and evaluate the result with the TRACE/SIMULATE-3K (S3K) coupled code, the S3K standalone code (core calculation) as well as performing single-channel calculations with S3K and VIPRE. This is the typical separate-effect-analyses required for advanced calculations in order to develop methodologies to be used for safety analyses in general. The models and methods of the code systems are presented. The outline represents the analysis approach starting with the coupled code system, reactor and core model calculation (TRACE/S3K). This is followed by a more detailed core evaluation (S3K standalone) and finally a very detailed thermal-hydraulic investigation of the hot pin condition (VIPRE).

  4. A method for scientific code coupling in a distributed environment

    International Nuclear Information System (INIS)

    Caremoli, C.; Beaucourt, D.; Chen, O.; Nicolas, G.; Peniguel, C.; Rascle, P.; Richard, N.; Thai Van, D.; Yessayan, A.

    1994-12-01

    This guide book deals with coupling of big scientific codes. First, the context is introduced: big scientific codes devoted to a specific discipline coming to maturity, and more and more needs in terms of multi discipline studies. Then we describe different kinds of code coupling and an example of code coupling: 3D thermal-hydraulic code THYC and 3D neutronics code COCCINELLE. With this example we identify problems to be solved to realize a coupling. We present the different numerical methods usable for the resolution of coupling terms. This leads to define two kinds of coupling: with the leak coupling, we can use explicit methods, and with the strong coupling we need to use implicit methods. On both cases, we analyze the link with the way of parallelizing code. For translation of data from one code to another, we define the notion of Standard Coupling Interface based on a general structure for data. This general structure constitutes an intermediary between the codes, thus allowing a relative independence of the codes from a specific coupling. The proposed method for the implementation of a coupling leads to a simultaneous run of the different codes, while they exchange data. Two kinds of data communication with message exchange are proposed: direct communication between codes with the use of PVM product (Parallel Virtual Machine) and indirect communication with a coupling tool. This second way, with a general code coupling tool, is based on a coupling method, and we strongly recommended to use it. This method is based on the two following principles: re-usability, that means few modifications on existing codes, and definition of a code usable for coupling, that leads to separate the design of a code usable for coupling from the realization of a specific coupling. This coupling tool available from beginning of 1994 is described in general terms. (authors). figs., tabs

  5. Mesh-based parallel code coupling interface

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, K.; Steckel, B. (eds.) [GMD - Forschungszentrum Informationstechnik GmbH, St. Augustin (DE). Inst. fuer Algorithmen und Wissenschaftliches Rechnen (SCAI)

    2001-04-01

    MpCCI (mesh-based parallel code coupling interface) is an interface for multidisciplinary simulations. It provides industrial end-users as well as commercial code-owners with the facility to combine different simulation tools in one environment. Thereby new solutions for multidisciplinary problems will be created. This opens new application dimensions for existent simulation tools. This Book of Abstracts gives a short overview about ongoing activities in industry and research - all presented at the 2{sup nd} MpCCI User Forum in February 2001 at GMD Sankt Augustin. (orig.) [German] MpCCI (mesh-based parallel code coupling interface) definiert eine Schnittstelle fuer multidisziplinaere Simulationsanwendungen. Sowohl industriellen Anwender als auch kommerziellen Softwarehersteller wird mit MpCCI die Moeglichkeit gegeben, Simulationswerkzeuge unterschiedlicher Disziplinen miteinander zu koppeln. Dadurch entstehen neue Loesungen fuer multidisziplinaere Problemstellungen und fuer etablierte Simulationswerkzeuge ergeben sich neue Anwendungsfelder. Dieses Book of Abstracts bietet einen Ueberblick ueber zur Zeit laufende Arbeiten in der Industrie und in der Forschung, praesentiert auf dem 2{sup nd} MpCCI User Forum im Februar 2001 an der GMD Sankt Augustin. (orig.)

  6. The SWAN coupling code: user's guide

    International Nuclear Information System (INIS)

    Litaudon, X.; Moreau, D.

    1988-11-01

    Coupling of slow waves in a plasma near the lower hybrid frequency is well known and linear theory with density step followed by a constant gradient can be used with some confidence. With the aid of the computer code SWAN, which stands for 'Slow Wave Antenna', the following parameters can be numerically calculated: n parallel power spectrum, directivity (weighted by the current drive efficiency), reflection coefficients (amplitude and phase) both before and after the E-plane junctions, scattering matrix at the plasma interface, scattering matrix at the E-plane junctions, maximum electric fields in secondary waveguides and location where it occurs, effect of passive waveguides on each side of the antenna, and the effect of a finite magnetic field in front of the antenna (for homogeneous plasma). This manual gives the basic information on the main assumptions of the coupling theory and on the use and general structure of the code itself. It answers the questions what are the main assumptions of the physical model? how to execute a job? what are the input parameters of the code? and what are the output results and where are they written? (author)

  7. Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Austregesilo, H.; Velkov, K. [GRS, Garching (Germany)] [and others

    1997-07-01

    The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes.

  8. Interface requirements to couple thermal-hydraulic codes to 3D neutronic codes

    International Nuclear Information System (INIS)

    Langenbuch, S.; Austregesilo, H.; Velkov, K.

    1997-01-01

    The present situation of thermalhydraulics codes and 3D neutronics codes is briefly described and general considerations for coupling of these codes are discussed. Two different basic approaches of coupling are identified and their relative advantages and disadvantages are discussed. The implementation of the coupling for 3D neutronics codes in the system ATHLET is presented. Meanwhile, this interface is used for coupling three different 3D neutronics codes

  9. RELAP5/MOD3 code coupling model

    International Nuclear Information System (INIS)

    Martin, R.P.; Johnsen, G.W.

    1994-01-01

    A new capability has been incorporated into RELAP5/MOD3 that enables the coupling of RELAP5/MOD3 to other computer codes. The new capability has been designed to support analysis of the new advanced reactor concepts. Its user features rely solely on new RELAP5 open-quotes styledclose quotes input and the Parallel Virtual Machine (PVM) software, which facilitates process management and distributed communication of multiprocess problems. RELAP5/MOD3 manages the input processing, communication instruction, process synchronization, and its own send and receive data processing. The flexible capability requires that an explicit coupling be established, which updates boundary conditions at discrete time intervals. Two test cases are presented that demonstrate the functionality, applicability, and issues involving use of this capability

  10. Application of coupled codes for safety analysis and licensing issues

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    2006-01-01

    An overview is given on the development and the advantages of coupled codes which integrate 3D neutron kinetics into thermal-hydraulic system codes. The work performed within GRS by coupling the thermal-hydraulic system code ATHLET and the 3D neutronics code QUABOX/CUBBOX is described as an example. The application of the coupled codes as best-estimate simulation tools for safety analysis is discussed. Some examples from German licensing practices are given which demonstrate how the improved analytical methods of coupled codes have contributed to solve licensing issues related to optimized and more economical use of fuel. (authors)

  11. Coupling a Basin Modeling and a Seismic Code using MOAB

    KAUST Repository

    Yan, Mi; Jordan, Kirk; Kaushik, Dinesh; Perrone, Michael; Sachdeva, Vipin; Tautges, Timothy J.; Magerlein, John

    2012-01-01

    We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.

  12. Coupling a Basin Modeling and a Seismic Code using MOAB

    KAUST Repository

    Yan, Mi

    2012-06-02

    We report on a demonstration of loose multiphysics coupling between a basin modeling code and a seismic code running on a large parallel machine. Multiphysics coupling, which is one critical capability for a high performance computing (HPC) framework, was implemented using the MOAB open-source mesh and field database. MOAB provides for code coupling by storing mesh data and input and output field data for the coupled analysis codes and interpolating the field values between different meshes used by the coupled codes. We found it straightforward to use MOAB to couple the PBSM basin modeling code and the FWI3D seismic code on an IBM Blue Gene/P system. We describe how the coupling was implemented and present benchmarking results for up to 8 racks of Blue Gene/P with 8192 nodes and MPI processes. The coupling code is fast compared to the analysis codes and it scales well up to at least 8192 nodes, indicating that a mesh and field database is an efficient way to implement loose multiphysics coupling for large parallel machines.

  13. Development of a coupled code system based on system transient code, RETRAN, and 3-D neutronics code, MASTER

    International Nuclear Information System (INIS)

    Kim, K. D.; Jung, J. J.; Lee, S. W.; Cho, B. O.; Ji, S. K.; Kim, Y. H.; Seong, C. K.

    2002-01-01

    A coupled code system of RETRAN/MASTER has been developed for best-estimate simulations of interactions between reactor core neutron kinetics and plant thermal-hydraulics by incorporation of a 3-D reactor core kinetics analysis code, MASTER into system transient code, RETRAN. The soundness of the consolidated code system is confirmed by simulating the MSLB benchmark problem developed to verify the performance of a coupled kinetics and system transient codes by OECD/NEA

  14. Interface requirements to couple thermal hydraulics codes to severe accident codes: ICARE/CATHARE

    Energy Technology Data Exchange (ETDEWEB)

    Camous, F.; Jacq, F.; Chatelard, P. [IPSN/DRS/SEMAR CE-Cadarache, St Paul Lez Durance (France)] [and others

    1997-07-01

    In order to describe with the same code the whole sequence of severe LWR accidents, up to the vessel failure, the Institute of Protection and Nuclear Safety has performed a coupling of the severe accident code ICARE2 to the thermalhydraulics code CATHARE2. The resulting code, ICARE/CATHARE, is designed to be as pertinent as possible in all the phases of the accident. This paper is mainly devoted to the description of the ICARE2-CATHARE2 coupling.

  15. Preliminary Coupling of MATRA Code for Multi-physics Analysis

    International Nuclear Information System (INIS)

    Kim, Seongjin; Choi, Jinyoung; Yang, Yongsik; Kwon, Hyouk; Hwang, Daehyun

    2014-01-01

    The boundary conditions such as the inlet temperature, mass flux, averaged heat flux, power distributions of the rods, and core geometry is given by constant values or functions of time. These conditions are separately calculated and provided by other codes, such as a neutronics or a system codes, into the MATRA code. In addition, the coupling of several codes in the different physics field is focused and embodied. In this study, multiphysics coupling methods were developed for a subchannel code (MATRA) with neutronics codes (MASTER, DeCART) and a fuel performance code (FRAPCON-3). Preliminary evaluation results for representative sample cases are presented. The MASTER and DeCART codes provide the power distribution of the rods in the core to the MATRA code. In case of the FRAPCON-3 code, the variation of the rod diameter induced by the thermal expansion is yielded and provided. The MATRA code transfers the thermal-hydraulic conditions that each code needs. Moreover, the coupling method with each code is described

  16. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  17. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  18. Spatially coupled LDPC coding in cooperative wireless networks

    NARCIS (Netherlands)

    Jayakody, D.N.K.; Skachek, V.; Chen, B.

    2016-01-01

    This paper proposes a novel technique of spatially coupled low-density parity-check (SC-LDPC) code-based soft forwarding relaying scheme for a two-way relay system. We introduce an array-based optimized SC-LDPC codes in relay channels. A more precise model is proposed to characterize the residual

  19. An analytical demonstration of coupling schemes between magnetohydrodynamic codes and eddy current codes

    International Nuclear Information System (INIS)

    Liu Yueqiang; Albanese, R.; Rubinacci, G.; Portone, A.; Villone, F.

    2008-01-01

    In order to model a magnetohydrodynamic (MHD) instability that strongly couples to external conducting structures (walls and/or coils) in a fusion device, it is often necessary to combine a MHD code solving for the plasma response, with an eddy current code computing the fields and currents of conductors. We present a rigorous proof of the coupling schemes between these two types of codes. One of the coupling schemes has been introduced and implemented in the CARMA code [R. Albanese, Y. Q. Liu, A. Portone, G. Rubinacci, and F. Villone, IEEE Trans. Magn. 44, 1654 (2008); A. Portone, F. Villone, Y. Q. Liu, R. Albanese, and G. Rubinacci, Plasma Phys. Controlled Fusion 50, 085004 (2008)] that couples the MHD code MARS-F[Y. Q. Liu, A. Bondeson, C. M. Fransson, B. Lennartson, and C. Breitholtz, Phys. Plasmas 7, 3681 (2000)] and the eddy current code CARIDDI[R. Albanese and G. Rubinacci, Adv. Imaging Electron Phys. 102, 1 (1998)]. While the coupling schemes are described for a general toroidal geometry, we give the analytical proof for a cylindrical plasma.

  20. Transient and fuel performance analysis with VTT's coupled code system

    International Nuclear Information System (INIS)

    Daavittila, A.; Hamalainen, A.; Raty, H.

    2005-01-01

    VTT (technical research center of Finland) maintains and further develops a comprehensive safety analysis code system ranging from the basic neutronic libraries to 3-dimensional transient analysis and fuel behaviour analysis codes. The code system is based on various types of couplings between the relevant physical phenomena. The main tools for analyses of reactor transients are presently the 3-dimensional reactor dynamics code HEXTRAN for cores with a hexagonal fuel assembly geometry and TRAB-3D for cores with a quadratic fuel assembly geometry. HEXTRAN has been applied to safety analyses of VVER type reactors since early 1990's. TRAB-3D is the latest addition to the code system, and has been applied to BWR and PWR analyses in recent years. In this paper it is shown that TRAB-3D has calculated accurately the power distribution during the Olkiluoto-1 load rejection test. The results from the 3-dimensional analysis can be used as boundary conditions for more detailed fuel rod analysis. For this purpose a general flow model GENFLO, developed at VTT, has been coupled with USNRC's FRAPTRAN fuel accident behaviour model. The example case for FRAPTRAN-GENFLO is for an ATWS at a BWR plant. The basis for the analysis is an oscillation incident in the Olkiluoto-1 BWR during reactor startup on February 22, 1987. It is shown that the new coupled code FRAPTRAN/GENFLO is quite a promising tool that can handle flow situations and give a detailed analysis of reactor transients

  1. Code Coupling for Multi-Dimensional Core Transient Analysis

    International Nuclear Information System (INIS)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il

    2015-01-01

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident

  2. Code Coupling for Multi-Dimensional Core Transient Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  3. A finite range coupled channel Born approximation code

    International Nuclear Information System (INIS)

    Nagel, P.; Koshel, R.D.

    1978-01-01

    The computer code OUKID calculates differential cross sections for direct transfer nuclear reactions in which multistep processes, arising from strongly coupled inelastic states in both the target and residual nuclei, are possible. The code is designed for heavy ion reactions where full finite range and recoil effects are important. Distorted wave functions for the elastic and inelastic scattering are calculated by solving sets of coupled differential equations using a Matrix Numerov integration procedure. These wave functions are then expanded into bases of spherical Bessel functions by the plane-wave expansion method. This approach allows the six-dimensional integrals for the transition amplitude to be reduced to products of two one-dimensional integrals. Thus, the inelastic scattering is treated in a coupled channel formalism while the transfer process is treated in a finite range born approximation formalism. (Auth.)

  4. 3D neutronic codes coupled with thermal-hydraulic system codes for PWR, and BWR and VVER reactors

    Energy Technology Data Exchange (ETDEWEB)

    Langenbuch, S.; Velkov, K. [GRS, Garching (Germany); Lizorkin, M. [Kurchatov-Institute, Moscow (Russian Federation)] [and others

    1997-07-01

    This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed.

  5. Coding "We-ness" in couple's relationship stories: A method for assessing mutuality in couple therapy.

    Science.gov (United States)

    Gildersleeve, Sara; Singer, Jefferson A; Skerrett, Karen; Wein, Shelter

    2017-05-01

    "We-ness," a couple's mutual investment in their relationship and in each other, has been found to be a potent dimension of couple resilience. This study examined the development of a method to capture We-ness in psychotherapy through the coding of relationship narratives co-constructed by couples ("We-Stories"). It used a coding system to identify the core thematic elements that make up these narratives. Couples that self-identified as "happy" (N = 53) generated We-Stories and completed measures of relationship satisfaction and mutuality. These stories were then coded using the We-Stories coding manual. Findings indicated that security, an element that involves aspects of safety, support, and commitment, was most common, appearing in 58.5% of all narratives. This element was followed by the elements of pleasure (49.1%) and shared meaning/vision (37.7%). The number of "We-ness" elements was also correlated with and predictive of discrepancy scores on measures of relationship mutuality, indicating the validity of the We-Stories coding manual. Limitations and future directions are discussed.

  6. Status of the coupled fluid-structure dynamics code SEURBNUK

    International Nuclear Information System (INIS)

    Smith, B.L.; Yerkess, A.; Adamson, J.

    1983-07-01

    The computer code SEURBNUK-2 is used collaboratively for the study of fast reactor containment integrity. Continuous extension and improvement of the numerical modelling has been required to match the performance of the code against the COVA series of scale model experiments and the requirements of reactor safety analysis. The present capabilities of SEURBNUK-2 are outlined and the most recent development topics are summarised. For internal structures amenable to thin shell treatment, a recent addition to the code permits these to be perforated, which is useful in modelling dip-plates and above-core structures in the reactor. In safety analysis much attention is paid to the response of the roof structure to impact loading from a rising coolant slug. The typical relationship between duration of the loading and the natural period of the roof shows that a coupled fluid/structure analysis is required. This must include the roof hold-down device which can introduce a low frequency component that considerably modifies the response of the closure system. A recent major extension to the SEURBNUK modelling is the installation of a moving roof option which, together with development of the logic to link structures external to the containment vessel, provides such coupling. (Auth.)

  7. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  8. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    International Nuclear Information System (INIS)

    Valtonen, Keijo; Hamalainen, Anitta; Cunningham, Mitchel E.

    2002-01-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  9. FRAPTRAN Fuel Rod Code and its Coupled Transient Analysis with the GENFLO Thermal-Hydraulic Code

    Energy Technology Data Exchange (ETDEWEB)

    Valtonen, Keijo (Radiation and Nuclear Safety Authority, Finland); Hamalainen, Anitta (VTT Energy, Finland); Cunningham, Mitchel E.(BATTELLE (PACIFIC NW LAB))

    2002-05-01

    The FRAPTRAN computer code has been developed for the U.S. Nuclear Regulatory Commission (NRC) to calculate fuel behavior during power and/or cooling transients at burnup levels up to 65 MWd/kgM. FRAPTRAN has now been assessed and peer reviewed. STUK/VTT have coupled GENFLO to FRAPTRAN for calculations with improved coolant boundary conditions and prepared example calculations to show the effect of improving the coolant boundary conditions.

  10. Study of the noise propagation in PWR with coupled codes

    International Nuclear Information System (INIS)

    Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

    2011-01-01

    The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

  11. Investigation of coupling scheme for neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Wang Guoli; Yu Jianfeng; Pen Muzhang; Zhang Yuman.

    1988-01-01

    Recently, a number of coupled neutronics/thermal-hydraulics codes have been used in reaction design and safty analysis, which have been obtained by coupling previous neutronic and thermal-hydraulic codes. The different coupling schemes affect computer time and accuracy of calculation results. Numberical experiments of several different coupling schemes and some heuristic results are described

  12. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1997-01-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

  13. Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

    Energy Technology Data Exchange (ETDEWEB)

    Baratta, A.J.

    1997-07-01

    To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

  14. Coupling of 3D neutronics models with the system code ATHLET

    International Nuclear Information System (INIS)

    Langenbuch, S.; Velkov, K.

    1999-01-01

    The system code ATHLET for plant transient and accident analysis has been coupled with 3D neutronics models, like QUABOX/CUBBOX, for the realistic evaluation of some specific safety problems under discussion. The considerations for the coupling approach and its realization are discussed. The specific features of the coupled code system established are explained and experience from first applications is presented. (author)

  15. Validation and applicability of the 3D core kinetics and thermal hydraulics coupled code SPARKLE

    International Nuclear Information System (INIS)

    Miyata, Manabu; Maruyama, Manabu; Ogawa, Junto; Otake, Yukihiko; Miyake, Shuhei; Tabuse, Shigehiko; Tanaka, Hirohisa

    2009-01-01

    The SPARKLE code is a coupled code system based on three individual codes whose physical models have already been verified and validated. Mitsubishi Heavy Industries (MHI) confirmed the coupling calculation, including data transfer and the total reactor coolant system (RCS) behavior of the SPARKLE code. The confirmation uses the OECD/NEA MSLB benchmark problem, which is based on Three Mile Island Unit 1 (TMI-1) nuclear power plant data. This benchmark problem has been used to verify coupled codes developed and used by many organizations. Objectives of the benchmark program are as follows. Phase 1 is to compare the results of the system transient code using point kinetics. Phase 2 is to compare the results of the coupled three-dimensional (3D) core kinetics code and 3D core thermal-hydraulics (T/H) code, and Phase 3 is to compare the results of the combined coupled system transient code, 3D core kinetics code, and 3D core T/H code as a total validation of the coupled calculation. The calculation results of the SPARKLE code indicate good agreement with other benchmark participants' results. Therefore, the SPARKLE code is validated through these benchmark problems. In anticipation of applying the SPARKLE code to licensing analyses, MHI and Japanese PWR utilities have established a safety analysis method regarding the calculation conditions such as power distributions, reactivity coefficients, and event-specific features. (author)

  16. 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.

  17. 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)

  18. Development and verification of a coupled code system RETRAN-MASTER-TORC

    International Nuclear Information System (INIS)

    Cho, J.Y.; Song, J.S.; Joo, H.G.; Zee, S.Q.

    2004-01-01

    Recently, coupled thermal-hydraulics (T-H) and three-dimensional kinetics codes have been widely used for the best-estimate simulations such as the main steam line break (MSLB) and locked rotor problems. This work is to develop and verify one of such codes by coupling the system T-H code RETRAN, the 3-D kinetics code MASTER and sub-channel analysis code TORC. The MASTER code has already been applied to such simulations after coupling with the MARS or RETRAN-3D multi-dimensional system T-H codes. The MASTER code contains a sub-channel analysis code COBRA-III C/P, and the coupled systems MARSMASTER-COBRA and RETRAN-MASTER-COBRA had been already developed and verified. With these previous studies, a new coupled system of RETRAN-MASTER-TORC is to be developed and verified for the standard best-estimate simulation code package in Korea. The TORC code has already been applied to the thermal hydraulics design of the several ABB/CE type plants and Korean Standard Nuclear Power Plants (KSNP). This justifies the choice of TORC rather than COBRA. Because the coupling between RETRAN and MASTER codes are already established and verified, this work is simplified to couple the TORC sub-channel T-H code with the MASTER neutronics code. The TORC code is a standalone code that solves the T-H equations for a given core problem from reading the input file and finally printing the converged solutions. However, in the coupled system, because TORC receives the pin power distributions from the neutronics code MASTER and transfers the T-H results to MASTER iteratively, TORC needs to be controlled by the MASTER code and does not need to solve the given problem completely at each iteration step. By this reason, the coupling of the TORC code with the MASTER code requires several modifications in the I/O treatment, flow iteration and calculation logics. The next section of this paper describes the modifications in the TORC code. The TORC control logic of the MASTER code is then followed. The

  19. Light-water-reactor coupled neutronic and thermal-hydraulic codes

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1982-01-01

    An overview is presented of computer codes that model light water reactor cores with coupled neutronics and thermal-hydraulics. This includes codes for transient analysis and codes for steady state analysis which include fuel depletion and fission product buildup. Applications in nuclear design, reactor operations and safety analysis are given and the major codes in use in the USA are identified. The neutronic and thermal-hydraulic methodologies and other code features are outlined for three steady state codes (PDQ7, NODE-P/B and SIMULATE) and four dynamic codes (BNL-TWIGL, MEKIN, RAMONA-3B, RETRAN-02). Speculation as to future trends with such codes is also presented

  20. A method for scientific code coupling in a distributed environment; Une methodologie pour le couplage de codes scientifiques en environnement distribue

    Energy Technology Data Exchange (ETDEWEB)

    Caremoli, C; Beaucourt, D; Chen, O; Nicolas, G; Peniguel, C; Rascle, P; Richard, N; Thai Van, D; Yessayan, A

    1994-12-01

    This guide book deals with coupling of big scientific codes. First, the context is introduced: big scientific codes devoted to a specific discipline coming to maturity, and more and more needs in terms of multi discipline studies. Then we describe different kinds of code coupling and an example of code coupling: 3D thermal-hydraulic code THYC and 3D neutronics code COCCINELLE. With this example we identify problems to be solved to realize a coupling. We present the different numerical methods usable for the resolution of coupling terms. This leads to define two kinds of coupling: with the leak coupling, we can use explicit methods, and with the strong coupling we need to use implicit methods. On both cases, we analyze the link with the way of parallelizing code. For translation of data from one code to another, we define the notion of Standard Coupling Interface based on a general structure for data. This general structure constitutes an intermediary between the codes, thus allowing a relative independence of the codes from a specific coupling. The proposed method for the implementation of a coupling leads to a simultaneous run of the different codes, while they exchange data. Two kinds of data communication with message exchange are proposed: direct communication between codes with the use of PVM product (Parallel Virtual Machine) and indirect communication with a coupling tool. This second way, with a general code coupling tool, is based on a coupling method, and we strongly recommended to use it. This method is based on the two following principles: re-usability, that means few modifications on existing codes, and definition of a code usable for coupling, that leads to separate the design of a code usable for coupling from the realization of a specific coupling. This coupling tool available from beginning of 1994 is described in general terms. (authors). figs., tabs.

  1. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J. [Purdue Univ., West Lafayette, IN (United States). Dept. of Nuclear Engineering; Wang, W. [SCIENTECH, Inc., Rockville, MD (United States); Mousseau, V.A.; Ebert, D.D. [Nuclear Regulatory Commission, Washington, DC (United States). Office of Nuclear Regulatory Research

    1999-03-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model.

  2. A generalized interface module for the coupling of spatial kinetics and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Barber, D.A.; Miller, R.M.; Joo, H.G.; Downar, T.J.; Mousseau, V.A.; Ebert, D.D.

    1999-01-01

    A generalized interface module has been developed for the coupling of any thermal-hydraulics code to any spatial kinetics code. The coupling scheme was designed and implemented with emphasis placed on maximizing flexibility while minimizing modifications to the respective codes. In this design, the thermal-hydraulics, general interface, and spatial kinetics codes function independently and utilize the Parallel Virtual Machine software to manage cross-process communication. Using this interface, the USNRC version of the 3D neutron kinetics code, PARCX, has been coupled to the USNRC system analysis codes RELAP5 and TRAC-M. RELAP5/PARCS assessment results are presented for two NEACRP rod ejection benchmark problems and an NEA/OECD main steam line break benchmark problem. The assessment of TRAC-M/PARCS has only recently been initiated, nonetheless, the capabilities of the coupled code are presented for a typical PWR system/core model

  3. A coupled systems code-CFD MHD solver for fusion blanket design

    Energy Technology Data Exchange (ETDEWEB)

    Wolfendale, Michael J., E-mail: m.wolfendale11@imperial.ac.uk; Bluck, Michael J.

    2015-10-15

    Highlights: • A coupled systems code-CFD MHD solver for fusion blanket applications is proposed. • Development of a thermal hydraulic systems code with MHD capabilities is detailed. • A code coupling methodology based on the use of TCP socket communications is detailed. • Validation cases are briefly discussed for the systems code and coupled solver. - Abstract: The network of flow channels in a fusion blanket can be modelled using a 1D thermal hydraulic systems code. For more complex components such as junctions and manifolds, the simplifications employed in such codes can become invalid, requiring more detailed analyses. For magnetic confinement reactor blanket designs using a conducting fluid as coolant/breeder, the difficulties in flow modelling are particularly severe due to MHD effects. Blanket analysis is an ideal candidate for the application of a code coupling methodology, with a thermal hydraulic systems code modelling portions of the blanket amenable to 1D analysis, and CFD providing detail where necessary. A systems code, MHD-SYS, has been developed and validated against existing analyses. The code shows good agreement in the prediction of MHD pressure loss and the temperature profile in the fluid and wall regions of the blanket breeding zone. MHD-SYS has been coupled to an MHD solver developed in OpenFOAM and the coupled solver validated for test geometries in preparation for modelling blanket systems.

  4. Single-phase mixing studies by means of a directly coupled CFD/system-code tool

    International Nuclear Information System (INIS)

    Bertolotto, Davide; Chawla, Rakesh; Manera, Annalisa; Smith, Brian; Prasser, Horst-Michael

    2008-01-01

    The present paper describes the coupling of the 3D computational fluid dynamics (CFD) code CFX with the best estimate thermal-hydraulic code TRACE. Two different coupling schemes, i.e. an explicit and a semi-implicit one, have been tested. Verification of the coupled CFX/TRACE code has first been carried out on the basis of a simple test case consisting of a straight pipe filled with liquid subject to a sudden acceleration. As a second validation step, measurements using advanced instrumentation (wire-mesh sensors) have been performed in a simple, specially constructed test facility consisting of two loops connected by a double T-junction. Comparisons of the measurements are made with calculation results obtained using the coupled codes, as well as the individual codes in stand-alone mode, thereby clearly bringing out the effectiveness of the achieved coupling for simulating situations in which three-dimensional mixing phenomena are important. (authors)

  5. Coupling methods for parallel running RELAPSim codes in nuclear power plant simulation

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yankai; Lin, Meng, E-mail: linmeng@sjtu.edu.cn; Yang, Yanhua

    2016-02-15

    When the plant is modeled detailedly for high precision, it is hard to achieve real-time calculation for one single RELAP5 in a large-scale simulation. To improve the speed and ensure the precision of simulation at the same time, coupling methods for parallel running RELAPSim codes were proposed in this study. Explicit coupling method via coupling boundaries was realized based on a data-exchange and procedure-control environment. Compromise of synchronization frequency was well considered to improve the precision of simulation and guarantee the real-time simulation at the same time. The coupling methods were assessed using both single-phase flow models and two-phase flow models and good agreements were obtained between the splitting–coupling models and the integrated model. The mitigation of SGTR was performed as an integral application of the coupling models. A large-scope NPP simulator was developed adopting six splitting–coupling models of RELAPSim and other simulation codes. The coupling models could improve the speed of simulation significantly and make it possible for real-time calculation. In this paper, the coupling of the models in the engineering simulator is taken as an example to expound the coupling methods, i.e., coupling between parallel running RELAPSim codes, and coupling between RELAPSim code and other types of simulation codes. However, the coupling methods are also referable in other simulator, for example, a simulator employing ATHLETE instead of RELAP5, other logic code instead of SIMULINK. It is believed the coupling method is commonly used for NPP simulator regardless of the specific codes chosen in this paper.

  6. Investigation of spatial coupling aspects for coupled code application in PWR safety analysis

    International Nuclear Information System (INIS)

    Todorova, N.K.; Ivanov, K.N.

    2003-01-01

    The simulation of nuclear power plant accident conditions requires three-dimensional (3-D) modeling of the reactor core to ensure a realistic description of physical phenomena. This paper describes a part of the research activities carried out on the sensitivity of coupled neutronics/thermal-hydraulic system code's results to the spatial mesh overlays used for modeling pressurized water reactor (PWR) cores for analysis of different transients. The coupled TRAC-PF1/NEM was used to model PWR rod ejection accident (REA). Modeling schemes for pressurized water reactor are described in detail, followed by a comparative analysis of both steady state and transient calculations. By using different TRAC-PF1/NEM vessel modeling options it was demonstrated that the geometric refinement plays a great role in determining the local parameters and control rod worth in the case of spatially asymmetric transients. The capability of TRAC-PF1/NEM to introduce local refinement of heat structure models was explored while preserving the original coarse-mesh structure of the hydraulic model. The obtained results indicated that the thermal-hydraulic feedback phenomenon is non-linear and cannot be separated even in rod ejection accident analysis, where the Doppler feedback plays a dominant role. While the impact of neutronics mesh refinement is well known, this research found that the local predictions, as well as the global predictions are also very sensitive to the thermal-hydraulic refinement

  7. Assessment of systems codes and their coupling with CFD codes in thermal–hydraulic applications to innovative reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bandini, G., E-mail: giacomino.bandini@enea.it [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Polidori, M. [Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) (Italy); Gerschenfeld, A.; Pialla, D.; Li, S. [Commissariat à l’Energie Atomique (CEA) (France); Ma, W.M.; Kudinov, P.; Jeltsov, M.; Kööp, K. [Royal Institute of Technology (KTH) (Sweden); Huber, K.; Cheng, X.; Bruzzese, C.; Class, A.G.; Prill, D.P. [Karlsruhe Institute of Technology (KIT) (Germany); Papukchiev, A. [Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) (Germany); Geffray, C.; Macian-Juan, R. [Technische Universität München (TUM) (Germany); Maas, L. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN) (France)

    2015-01-15

    Highlights: • The assessment of RELAP5, TRACE and CATHARE system codes on integral experiments is presented. • Code benchmark of CATHARE, DYN2B, and ATHLET on PHENIX natural circulation experiment. • Grid-free pool modelling based on proper orthogonal decomposition for system codes is explained. • The code coupling methodologies are explained. • The coupling of several CFD/system codes is tested against integral experiments. - Abstract: The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal–hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal–hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal–hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.

  8. SIERRA Code Coupling Module: Arpeggio User Manual Version 4.44

    Energy Technology Data Exchange (ETDEWEB)

    Sierra Thermal/Fluid Team

    2017-04-01

    The SNL Sierra Mechanics code suite is designed to enable simulation of complex multiphysics scenarios. The code suite is composed of several specialized applications which can operate either in standalone mode or coupled with each other. Arpeggio is a supported utility that enables loose coupling of the various Sierra Mechanics applications by providing access to Framework services that facilitate the coupling. More importantly Arpeggio orchestrates the execution of applications that participate in the coupling. This document describes the various components of Arpeggio and their operability. The intent of the document is to provide a fast path for analysts interested in coupled applications via simple examples of its usage.

  9. Modelling of the Rod Control System in the coupled code RELAP5-QUABOX/CUBBOX

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    1999-01-01

    There is a general agreement that for many light water reactor transient calculations, it is necessary to use a multidimensional neutron kinetics model coupled to sophisticated thermal-hydraulic models in order to obtain satisfactory results. These calculations are needed for a variety of applications for licensing safety analyses, probabilistic risk assessment, operational support, and training. At FER, Zagreb, a coupling of 3D neutronics code QUABOX/CUBBOX and system code RELAP5 was performed. In the paper the Rod Control System model in the RELAP5 part of the coupled code is presented. A first testing of the model was performed by calculation of reactor trip from full power for NPP Krsko. Results of 3D neutronics calculation obtained by coupled code QUABOX/CUBBOX were compared with point kinetics calculation performed with RELAP5 stand alone code.(author)

  10. Application of the SALOME platform to the loose coupling of the CATHENA and ELOCA codes

    International Nuclear Information System (INIS)

    Zhuchkova, A.

    2012-01-01

    Use of coupled codes for the safety analysis of nuclear power plants is highly desirable, as it permits multi-disciplinary studies of complex reactor behaviors and, in particular, accident simulations. The present work demonstrates the potential of the SALOME platform as an interface for creating integrated, multi-disciplinary simulations of reactor scenarios. For this purpose two codes currently in use within the Canadian nuclear industry, CATHENA and ELOCA, were coupled by means of SALOME. The coupled codes were used to model the Power Burst Facility (PBF)-CANDU Test, which was to test the thermal-mechanical behavior of PHWR (pressurized heavy water reactor) fuel during a simulated Large Loss-Of-Coolant Accident (LLOCA). The results of the SALOME-coupled simulations are compared with a previous analysis in which the two codes were coupled using a package of scripts. (author)

  11. Development of the CRIPTE Code for Electromagnetic Coupling

    National Research Council Canada - National Science Library

    Parmantier, Jean-Philippe

    2005-01-01

    .... This code was originally developed as part of an experiment performed under the joint US-France international data exchange program on the atmospheric electricity/aircraft interactions, DEA-AF-79-7336...

  12. An approach for coupled-code multiphysics core simulations from a common input

    International Nuclear Information System (INIS)

    Schmidt, Rodney; Belcourt, Kenneth; Hooper, Russell; Pawlowski, Roger; Clarno, Kevin; Simunovic, Srdjan; Slattery, Stuart; Turner, John; Palmtag, Scott

    2015-01-01

    Highlights: • We describe an approach for coupled-code multiphysics reactor core simulations. • The approach can enable tight coupling of distinct physics codes with a common input. • Multi-code multiphysics coupling and parallel data transfer issues are explained. • The common input approach and how the information is processed is described. • Capabilities are demonstrated on an eigenvalue and power distribution calculation. - Abstract: This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak

  13. COUPLED SIMULATION OF GAS COOLED FAST REACTOR FUEL ASSEMBLY WITH NESTLE CODE SYSTEM

    Directory of Open Access Journals (Sweden)

    Filip Osusky

    2018-05-01

    Full Text Available The paper is focused on coupled calculation of the Gas Cooled Fast Reactor. The proper modelling of coupled neutronics and thermal-hydraulics is the corner stone for future safety assessment of the control and emergency systems. Nowadays, the system and channel thermal-hydraulic codes are accepted by the national regulatory authorities in European Union for license purposes, therefore the code NESTLE was used for the simulation. The NESTLE code is a coupled multigroup neutron diffusion code with thermal-hydraulic sub-channel code. In the paper, the validation of NESTLE code 5.2.1 installation is presented. The processing of fuel assembly homogeneous parametric cross-section library for NESTLE code simulation is made by the sequence TRITON of SCALE code package system. The simulated case in the NESTLE code is one fuel assembly of GFR2400 concept with reflective boundary condition in radial direction and zero flux boundary condition in axial direction. The results of coupled calculation are presented and are consistent with the GFR2400 study of the GoFastR project.

  14. Development of a general coupling interface for the fuel performance code TRANSURANUS – Tested with the reactor dynamics code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.; Macián-Juan, R.

    2015-01-01

    Highlights: • A general coupling interface was developed for couplings of the TRANSURANUS code. • With this new tool simplified fuel behavior models in codes can be replaced. • Applicable e.g. for several reactor types and from normal operation up to DBA. • The general coupling interface was applied to the reactor dynamics code DYN3D. • The new coupled code system DYN3D–TRANSURANUS was successfully tested for RIA. - Abstract: A general interface is presented for coupling the TRANSURANUS fuel performance code with thermal hydraulics system, sub-channel thermal hydraulics, computational fluid dynamics (CFD) or reactor dynamics codes. As first application the reactor dynamics code DYN3D was coupled at assembly level in order to describe the fuel behavior in more detail. In the coupling, DYN3D provides process time, time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in case of the two-way coupling approach transfers parameters like fuel temperature and cladding temperature back to DYN3D. Results of the coupled code system are presented for the reactivity transient scenario, initiated by control rod ejection. More precisely, the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy. These differences can be explained thanks to the greater detail in fuel behavior modeling. The numerical performance for DYN3D–TRANSURANUS was proved to be fast and stable. The coupled code system can therefore improve the assessment of safety criteria, at a reasonable computational cost

  15. Coupling of RELAP5-3D and GAMMA codes for Nuclear Hydrogen System Analysis

    International Nuclear Information System (INIS)

    Jin, Hyung Gon

    2007-02-01

    RELAP5-3D is one of the most important system analysis codes in nuclear field, which has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The GAMMA code is a multi-dimensional multi-component mixture analysis code with the complete set of chemical reaction models which is developed for safety analysis of HTGR (High Temperature Gas Cooled Reactor) air-ingress. The two codes, RELAP5-3D and GAMMA, are coupled to be used for nuclear-hydrogen system analysis, which requires the capability of the analysis of multi-component gas mixture and two-phase flow. In order to couple the two codes, 4 steps are needed. Before coupling, the GAMMA code was transformed into DLL (dynamic link liberally) from executive type and RELAP5-3D was recompiled into Compaq Visual Fortran environments for our debugging purpose. As the second step, two programs - RELAP5-3D and GAMMA codes - must be synchronized in terms of time and time step. Based on that time coupling, the coupled code can calculate simultaneously. Time-step coupling had been accomplished successfully and it is tested by using a simple test input. As a next step, source-term coupling was done and it was also tested in two different test inputs. The fist case is a simple test condition, which has no chemical reaction. And the other test set is the chemical reaction model, including four non-condensable gas species, which are He, O2, CO, CO2. Finally, in order to analyze combined cycle system, heat-flux coupling has been made and a simple heat exchanger model was demonstrated

  16. Qualification of the coupled RELAP5/PANTHER/COBRA code package for licensing applications

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Zhang Jinzhao

    2004-01-01

    A coupled thermal hydraulics-neutronics code package has been developed at Tractebel Engineering (TE), in which the best-estimate thermal-hydraulic system code, RELAP5/mod2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via the dynamic data exchange interface, TALINK. The Departure from Nucleate Boiling Ratio (DNBR) is calculated by the sub-channel thermal-hydraulic analysis code COBRA-3C. The package provides the capability to accurately simulate the key physical phenomena in nuclear power plant accidents with strong asymmetric behaviours and system-core interactions. This paper presents the TE coupled code package and focuses on the methodology followed for qualifying it for licensing applications. The qualification of the coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric PWR accidents with strong core-system interactions

  17. Simulation of LOCA power transients of CANDU6 by SCAN/RELAP-CANDU coupled code system

    International Nuclear Information System (INIS)

    Hong, In Seob; Kim, Chang Hyo; Hwang, Su Hyun; Kim, Man Woong; Chung, Bub Dong

    2004-01-01

    As can be seen in the standalone application of RELAP-CANDU for LOCA analysis of CANDU-PHWR, the system thermal-hydraulic code alone cannot predict the transient behavior accurately. Therefore, best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. The purpose of this research is to develop and test a coupled neutronics and thermal-hydraulics analysis code, SCAN (SNU CANDU-PHWR Neutronics) and RELAP-CANDU, for transient analysis of CANDU-PHWR's. For this purpose, a spatial kinetics calculation module of SCAN, a 3-D CANDU-PHWR neutronics design and analysis code, is dynamically coupled with RELAP-CANDU, the system thermal-hydraulic code for CANDU-PHWR. The performance of the coupled code system is examined by simulation of reactor power transients caused by a hypothetical Loss Of Coolant Accident (LOCA) in Wolsong units, which involves the insertion of positive void reactivity into the core in the course of transients. Specifically, a 40% Reactor Inlet Header (RIH) break LOCA was assumed for the test of the SCAN/RELAP-CANDU coupled code system analysis

  18. Development of a coupling code for PWR reactor cavity radiation streaming calculation

    International Nuclear Information System (INIS)

    Zheng, Z.; Wu, H.; Cao, L.; Zheng, Y.; Zhang, H.; Wang, M.

    2012-01-01

    PWR reactor cavity radiation streaming is important for the safe of the personnel and equipment, thus calculation has to be performed to evaluate the neutron flux distribution around the reactor. For this calculation, the deterministic codes have difficulties in fine geometrical modeling and need huge computer resource; and the Monte Carlo codes require very long sampling time to obtain results with acceptable precision. Therefore, a coupling method has been developed to eliminate the two problems mentioned above in each code. In this study, we develop a coupling code named DORT2MCNP to link the Sn code DORT and Monte Carlo code MCNP. DORT2MCNP is used to produce a combined surface source containing top, bottom and side surface simultaneously. Because SDEF card is unsuitable for the combined surface source, we modify the SOURCE subroutine of MCNP and compile MCNP for this application. Numerical results demonstrate the correctness of the coupling code DORT2MCNP and show reasonable agreement between the coupling method and the other two codes (DORT and MCNP). (authors)

  19. The gradual development steps of the external coupled RELAP5 - DYN3D code

    International Nuclear Information System (INIS)

    Strmensky, C.

    2001-01-01

    This paper describes the on-going and finished parts of project: 'The external coupled RELAP5-DYN3D code'. The development progress was divided into four steps. In present time, second and third steps are performed and four step is started. The two parameters of coolant was selected and are exchanged between codes RELAP5 and DYN3D. (authors)

  20. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Cheng, X.

    2015-01-01

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  1. Sub-channel/system coupled code development and its application to SCWR-FQT loop

    Energy Technology Data Exchange (ETDEWEB)

    Liu, X.J., E-mail: xiaojingliu@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dong Chuan Road, Shanghai 200240 (China); Cheng, X. [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Prießnitz-Str. 3, 76131 Karlsruhe (Germany)

    2015-04-15

    Highlights: • A coupled code is developed for SCWR accident simulation. • The feasibility of the code is shown by application to SCWR-FQT loop. • Some measures are selected by sensitivity analysis. • The peak cladding temperature can be reduced effectively by the proposed measures. - Abstract: In the frame of Super-Critical Reactor In Pipe Test Preparation (SCRIPT) project in China, one of the challenge tasks is to predict the transient performance of SuperCritical Water Reactor-Fuel Qualification Test (SCWR-FQT) loop under some accident conditions. Several thermal–hydraulic codes (system code, sub-channel code) are selected to perform the safety analysis. However, the system code cannot simulate the local behavior of the test bundle, and the sub-channel code is incapable of calculating the whole system behavior of the test loop. Therefore, to combine the merits of both codes, and minimizes their shortcomings, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code COBRA-SC and system code ATHLET-SC are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the new developed coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal–hydraulic parameters are predicted by the sub-channel code COBRA-SC. The codes are utilized to get the local thermal–hydraulic parameters in the SCWR-FQT fuel bundle under some accident case (e.g. a flow blockage during LOCA). Some measures to mitigate the accident consequence are proposed by the sensitivity study and trialed to demonstrate their effectiveness in the coupled simulation. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel bundle can be reduced effectively by the safety measures

  2. Verification of CTF/PARCSv3.2 coupled code in a Turbine Trip scenario

    International Nuclear Information System (INIS)

    Abarca, A.; Hidalga, P.; Miro, R.; Verdu, G.; Sekhri, A.

    2017-01-01

    Multiphysics codes had revealed as a best-estimate approach to simulate core behavior in LWR. Coupled neutronics and thermal-hydraulics codes are being used and improved to achieve reliable results for reactor safety transient analysis. The implementation of the feedback procedure between the coupled codes at each time step allows a more accurate simulation and a better prediction of the safety limits of analyzed scenarios. With the objective of testing the recently developed CTF/PARCSv3.2 coupled code, a code-to-code verification against TRACE has been developed in a BWR Turbine Trip scenario. CTF is a thermal-hydraulic subchannel code that features two-fluid, three-field representation of the two-phase flow, while PARCS code solves the neutronic diffusion equation in a 3D nodal distribution. PARCS features allow as well the use of extended sets of cross section libraries for a more precise neutronic performance in different formats like PMAX or NEMTAB. Using this option the neutronic core composition of KKL will be made taking advantage of the core follow database. The results of the simulation will be verified against TRACE results. TRACE will be used as a reference code for the validation process since it has been a recommended code by the USNRC. The model used for TRACE includes a full core plus relevant components such as the steam lines and the valves affecting and controlling the turbine trip evolution. The coupled code performance has been evaluated using the Turbine Trip event that took place in Kern Kraftwerk Leibstadt (KKL), at the fuel cycle 18. KKL is a Nuclear Power Plant (NPP) located in Leibstadt, Switzerland. This NPP operates with a BWR developing 3600 MWt in fuel cycles of one year. The Turbine Trip is a fast transient developing a pressure peak in the reactor followed by a power decreasing due to the selected control rod insertion. This kind of transient is very useful to check the feedback performance between both coupled codes due to the fast

  3. Uncertainty and sensitivity analysis applied to coupled code calculations for a VVER plant transient

    International Nuclear Information System (INIS)

    Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K. D.

    2004-01-01

    The development of coupled codes, combining thermal-hydraulic system codes and 3D neutron kinetics, is an important step to perform best-estimate plant transient calculations. It is generally agreed that the application of best-estimate methods should be supplemented by an uncertainty and sensitivity analysis to quantify the uncertainty of the results. The paper presents results from the application of the GRS uncertainty and sensitivity method for a VVER-440 plant transient, which was already studied earlier for the validation of coupled codes. For this application, the main steps of the uncertainty method are described. Typical results of the method applied to the analysis of the plant transient by several working groups using different coupled codes are presented and discussed The results demonstrate the capability of an uncertainty and sensitivity analysis. (authors)

  4. Comparative study of the Peach Bottom turbine trip experiment using two different coupled codes approaches

    International Nuclear Information System (INIS)

    Bambara, M.; Bousbia-Salah, A.; D'Auria, F.

    2005-01-01

    Full text of publication follows: In the last years a great concern about the neutron-3D/thermal-hydraulic codes coupling took place. Owing to the improved computational technology, 'best estimate' analyses are today a common tool to assess safety features, and they are necessary if an asymmetric behaviour in the core region exists, or if strong interactions between the core neutronics and reactor thermal-hydraulic occur. In order to validate the coupled codes performances, several international programmes were issued. Among these activities, the OECD/NEA BWR Turbine Trip (TT) was chosen for further sensitivity analyses. It consists of a turbine trip (TT) experiment carried out at the Peach Bottom 2 BWR. In this paper, the results of two different coupled codes systems are summarized and compared. The BWR TT simulations were carried out coupling the thermal-hydraulic system code RELAP5/mode 3.2 to the 3D neutron kinetics code Parcs/2.3, and also the system code ATHLET to the neutronics code QUABOX-CUBBOX. An exhaustive overview of the main features is given, and those aspects, which need further developments and experiences, are pointed out. (authors)

  5. Extensions to the coupled chemical equilibria and migration code CHEQMATE

    International Nuclear Information System (INIS)

    Haworth, A.; Sharland, S.M.; Tasker, P.W.; Tweed, C.J.

    1988-08-01

    The CHEQMATE program was developed to model the evolution of spatially inhomogeneous aqueous chemical systems. The original CHEQMATE models one-dimensional diffusion and electromigration of ionic species with chemical equilibration provided by the geochemical program PHREEQE. CHEQMATE has principally been used to study the evolution of the chemical environment in and around a nuclear waste repository. In this paper, we describe extensions to CHEQMATE to increase the range of situations that can be modelled. These extensions are the addition of advection of species in a constant groundwater flow, the facility to model migration of species through a series of media with different transport properties and migration in a spherical geometry which allows investigation of dilution effects. For each extension, we describe the alterations in the transport part of the code and consider how the model is set up. An example of a problem using the different versions is given. (author)

  6. Interface code between WIMS-AECL and RFSP-IST for coupling computing

    International Nuclear Information System (INIS)

    Xu Liangwang; Liu Yu; Jia Baoshan

    2007-01-01

    A code based on the protocols of Telnet and FTP is developed with C++ for coupling computing between WIMS-AECL and RFSP-IST. the input document of WIMS-AECL and RFSP-ISP cna be generated automatically and be submitted to server, the output document will be downloaded by the end of computing. the function of analyzing standard output document is also included in this code. After simple updating, this code can meet the requirement of other code using input document, e.g. CATHENA. A pilot study of the relation between void fraction and reactivity in TACR, some valuable conclusions has been achieved. (authors)

  7. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    Energy Technology Data Exchange (ETDEWEB)

    Hoogenboom, J. Eduard, E-mail: J.E.Hoogenboom@tudelft.nl [Delft University of Technology (Netherlands); Ivanov, Aleksandar; Sanchez, Victor, E-mail: Aleksandar.Ivanov@kit.edu, E-mail: Victor.Sanchez@kit.edu [Karlsruhe Institute of Technology, Institute of Neutron Physics and Reactor Technology, Eggenstein-Leopoldshafen (Germany); Diop, Cheikh, E-mail: Cheikh.Diop@cea.fr [CEA/DEN/DANS/DM2S/SERMA, Commissariat a l' Energie Atomique, Gif-sur-Yvette (France)

    2011-07-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  8. Nodal kinetics model upgrade in the Penn State coupled TRAC/NEM codes

    International Nuclear Information System (INIS)

    Beam, Tara M.; Ivanov, Kostadin N.; Baratta, Anthony J.; Finnemann, Herbert

    1999-01-01

    The Pennsylvania State University currently maintains and does development and verification work for its own versions of the coupled three-dimensional kinetics/thermal-hydraulics codes TRAC-PF1/NEM and TRAC-BF1/NEM. The subject of this paper is nodal model enhancements in the above mentioned codes. Because of the numerous validation studies that have been performed on almost every aspect of these codes, this upgrade is done without a major code rewrite. The upgrade consists of four steps. The first two steps are designed to improve the accuracy of the kinetics model, based on the nodal expansion method. The polynomial expansion solution of 1D transverse integrated diffusion equation is replaced with a solution, which uses a semi-analytic expansion. Further the standard parabolic polynomial representation of the transverse leakage in the above 1D equations is replaced with an improved approximation. The last two steps of the upgrade address the code efficiency by improving the solution of the time-dependent NEM equations and implementing a multi-grid solver. These four improvements are implemented into the standalone NEM kinetics code. Verification of this code was accomplished based on the original verification studies. The results show that the new methods improve the accuracy and efficiency of the code. The verification of the upgraded NEM model in the TRAC-PF1/NEM and TRAC-BF1/NEM coupled codes is underway

  9. A flexible coupling scheme for Monte Carlo and thermal-hydraulics codes

    International Nuclear Information System (INIS)

    Hoogenboom, J. Eduard; Ivanov, Aleksandar; Sanchez, Victor; Diop, Cheikh

    2011-01-01

    A coupling scheme between a Monte Carlo code and a thermal-hydraulics code is being developed within the European NURISP project for comprehensive and validated reactor analysis. The scheme is flexible as it allows different Monte Carlo codes and different thermal-hydraulics codes to be used. At present the MCNP and TRIPOLI4 Monte Carlo codes can be used and the FLICA4 and SubChanFlow thermal-hydraulics codes. For all these codes only an original executable is necessary. A Python script drives the iterations between Monte Carlo and thermal-hydraulics calculations. It also calls a conversion program to merge a master input file for the Monte Carlo code with the appropriate temperature and coolant density data from the thermal-hydraulics calculation. Likewise it calls another conversion program to merge a master input file for the thermal-hydraulics code with the power distribution data from the Monte Carlo calculation. Special attention is given to the neutron cross section data for the various required temperatures in the Monte Carlo calculation. Results are shown for an infinite lattice of PWR fuel pin cells and a 3 x 3 fuel BWR pin cell cluster. Various possibilities for further improvement and optimization of the coupling system are discussed. (author)

  10. Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

    International Nuclear Information System (INIS)

    Keco, M.; Debrecin, N.; Grgic, D.

    2005-01-01

    Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

  11. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    International Nuclear Information System (INIS)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

    2001-01-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  12. Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

    2001-07-01

    A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

  13. Coupling of system thermal–hydraulics and Monte-Carlo code: Convergence criteria and quantification of correlation between statistical uncertainty and coupled error

    International Nuclear Information System (INIS)

    Wu, Xu; Kozlowski, Tomasz

    2015-01-01

    Highlights: • Coupling of Monte Carlo code Serpent and thermal–hydraulics code RELAP5. • A convergence criterion is developed based on the statistical uncertainty of power. • Correlation between MC statistical uncertainty and coupled error is quantified. • Both UO 2 and MOX single assembly models are used in the coupled simulation. • Validation of coupling results with a multi-group transport code DeCART. - Abstract: Coupled multi-physics approach plays an important role in improving computational accuracy. Compared with deterministic neutronics codes, Monte Carlo codes have the advantage of a higher resolution level. In the present paper, a three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code, Serpent, is coupled with a thermal–hydraulics safety analysis code, RELAP5. The coupled Serpent/RELAP5 code capability is demonstrated by the improved axial power distribution of UO 2 and MOX single assembly models, based on the OECD-NEA/NRC PWR MOX/UO 2 Core Transient Benchmark. Comparisons of calculation results using the coupled code with those from the deterministic methods, specifically heterogeneous multi-group transport code DeCART, show that the coupling produces more precise results. A new convergence criterion for the coupled simulation is developed based on the statistical uncertainty in power distribution in the Monte Carlo code, rather than ad-hoc criteria used in previous research. The new convergence criterion is shown to be more rigorous, equally convenient to use but requiring a few more coupling steps to converge. Finally, the influence of Monte Carlo statistical uncertainty on the coupled error of power and thermal–hydraulics parameters is quantified. The results are presented such that they can be used to find the statistical uncertainty to use in Monte Carlo in order to achieve a desired precision in coupled simulation

  14. BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

    International Nuclear Information System (INIS)

    Hartmann, C.; Sanchez, V.; Tietsch, W.; Stieglitz, R.

    2012-01-01

    The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

  15. The development of depletion program coupled with Monte Carlo computer code

    International Nuclear Information System (INIS)

    Nguyen Kien Cuong; Huynh Ton Nghiem; Vuong Huu Tan

    2015-01-01

    The paper presents the development of depletion code for light water reactor coupled with MCNP5 code called the MCDL code (Monte Carlo Depletion for Light Water Reactor). The first order differential depletion system equations of 21 actinide isotopes and 50 fission product isotopes are solved by the Radau IIA Implicit Runge Kutta (IRK) method after receiving neutron flux, reaction rates in one group energy and multiplication factors for fuel pin, fuel assembly or whole reactor core from the calculation results of the MCNP5 code. The calculation for beryllium poisoning and cooling time is also integrated in the code. To verify and validate the MCDL code, high enriched uranium (HEU) and low enriched uranium (LEU) fuel assemblies VVR-M2 types and 89 fresh HEU fuel assemblies, 92 LEU fresh fuel assemblies cores of the Dalat Nuclear Research Reactor (DNRR) have been investigated and compared with the results calculated by the SRAC code and the MCNP R EBUS linkage system code. The results show good agreement between calculated data of the MCDL code and reference codes. (author)

  16. XGC developments for a more efficient XGC-GENE code coupling

    Science.gov (United States)

    Dominski, Julien; Hager, Robert; Ku, Seung-Hoe; Chang, Cs

    2017-10-01

    In the Exascale Computing Program, the High-Fidelity Whole Device Modeling project initially aims at delivering a tightly-coupled simulation of plasma neoclassical and turbulence dynamics from the core to the edge of the tokamak. To permit such simulations, the gyrokinetic codes GENE and XGC will be coupled together. Numerical efforts are made to improve the numerical schemes agreement in the coupling region. One of the difficulties of coupling those codes together is the incompatibility of their grids. GENE is a continuum grid-based code and XGC is a Particle-In-Cell code using unstructured triangular mesh. A field-aligned filter is thus implemented in XGC. Even if XGC originally had an approximately field-following mesh, this field-aligned filter permits to have a perturbation discretization closer to the one solved in the field-aligned code GENE. Additionally, new XGC gyro-averaging matrices are implemented on a velocity grid adapted to the plasma properties, thus ensuring same accuracy from the core to the edge regions.

  17. Evaluation of hydrogen production system coupling with HTTR using dynamic analysis code

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Ohashi, Hirofumi; Inaba, Yoshitomo; Nishihara, Tetsuo; Hayashi, Koji; Inagaki, Yoshiyuki

    2006-01-01

    The Japan Atomic Energy Agency (JAEA) was entrusted 'Development of Nuclear Heat Utilization Technology' by Ministry of Education, Culture, Sports, Science and Technology. In this development, the JAEA investigated the system integration technology to couple the hydrogen production system by steam reforming with the High Temperature Engineering Test Reactor (HTTR). Prior to the construction of the hydrogen production system coupling with the HTTR, a dynamic analysis code had to be developed to evaluate the system transient behaviour of the hydrogen production system because there are no examples of chemical facilities coupled with nuclear reactor in the world. This report describes the evaluation of the hydrogen production system coupling with HTTR using analysis code, N-HYPAC, which can estimate transient behaviour of the hydrogen production system by steam reforming. The results of this investigation provide that the influence of the thermal disturbance caused by the hydrogen production system on the HTTR can be estimated well. (author)

  18. Influence of reactor vessel nodalization in the coupled code analysis of Asymmetric Main Feedwater Isolation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    2001-01-01

    Asymmetric Main Feedwater Isolation (AMFWI) transient in one Steam Generator (SG) for NPP Krsko using RELAP5 standalone code and coupled code RELAP5- QUABOX/CUBBOX (R5QC) was analyzed. In the RELAP5 standalone calculation, a point kinetics model was used, while in the coupled code a three-dimensional (3D) neutronics model of QUABOX with different RELAP5 nodalization schemes of reactor vessel was used. Both code versions use best-estimate thermal-hydraulic system code for all components in the plant and include realistic description of plant protection and control systems. Two different types of calculations were performed: with and without automatic control rod system available. The AMFWI transient causes the great asymmetry of the transferred heat in the SGs and subsequently the asymmetry of the power produced across the core due to different reactivity feedback resulting from the thermal-hydraulic channels assigned to different loops. The work presented in the paper is a part of validation of the 3D coupled code R5QC in the analysis of asymmetric transients.(author)

  19. Nonlinear to Linear Elastic Code Coupling in 2-D Axisymmetric Media.

    Energy Technology Data Exchange (ETDEWEB)

    Preston, Leiph [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-08-01

    Explosions within the earth nonlinearly deform the local media, but at typical seismological observation distances, the seismic waves can be considered linear. Although nonlinear algorithms can simulate explosions in the very near field well, these codes are computationally expensive and inaccurate at propagating these signals to great distances. A linearized wave propagation code, coupled to a nonlinear code, provides an efficient mechanism to both accurately simulate the explosion itself and to propagate these signals to distant receivers. To this end we have coupled Sandia's nonlinear simulation algorithm CTH to a linearized elastic wave propagation code for 2-D axisymmetric media (axiElasti) by passing information from the nonlinear to the linear code via time-varying boundary conditions. In this report, we first develop the 2-D axisymmetric elastic wave equations in cylindrical coordinates. Next we show how we design the time-varying boundary conditions passing information from CTH to axiElasti, and finally we demonstrate the coupling code via a simple study of the elastic radius.

  20. Incorporation of coupled nonequilibrium chemistry into a two-dimensional nozzle code (SEAGULL)

    Science.gov (United States)

    Ratliff, A. W.

    1979-01-01

    A two-dimensional multiple shock nozzle code (SEAGULL) was extended to include the effects of finite rate chemistry. The basic code that treats multiple shocks and contact surfaces was fully coupled with a generalized finite rate chemistry and vibrational energy exchange package. The modified code retains all of the original SEAGULL features plus the capability to treat chemical and vibrational nonequilibrium reactions. Any chemical and/or vibrational energy exchange mechanism can be handled as long as thermodynamic data and rate constants are available for all participating species.

  1. Coupled neutronic and thermal-hydraulic code benchmark activities at the International Nuclear Safety Center

    International Nuclear Information System (INIS)

    Podlazov, L. N.

    1998-01-01

    Two realistic benchmark problems are defined and used to assess the performance of coupled thermal-hydraulic and neutronic codes used in simulating dynamic processes in VVER-1000 and RBMK reactor systems. One of the problems simulates a design basis accident involving the ejection of three control and protection system rods from a VVER-1000 reactor. The other is based on a postulated rod withdrawal from an operating RBMK reactor. Preliminary results calculated by various codes are compared. While these results show significant differences, the intercomparisons performed so far provide a basis for further evaluation of code limitations and modeling assumptions

  2. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.

  3. Development of a dynamic coupled hydro-geomechanical code and its application to induced seismicity

    Science.gov (United States)

    Miah, Md Mamun

    This research describes the importance of a hydro-geomechanical coupling in the geologic sub-surface environment from fluid injection at geothermal plants, large-scale geological CO2 sequestration for climate mitigation, enhanced oil recovery, and hydraulic fracturing during wells construction in the oil and gas industries. A sequential computational code is developed to capture the multiphysics interaction behavior by linking a flow simulation code TOUGH2 and a geomechanics modeling code PyLith. Numerical formulation of each code is discussed to demonstrate their modeling capabilities. The computational framework involves sequential coupling, and solution of two sub-problems- fluid flow through fractured and porous media and reservoir geomechanics. For each time step of flow calculation, pressure field is passed to the geomechanics code to compute effective stress field and fault slips. A simplified permeability model is implemented in the code that accounts for the permeability of porous and saturated rocks subject to confining stresses. The accuracy of the TOUGH-PyLith coupled simulator is tested by simulating Terzaghi's 1D consolidation problem. The modeling capability of coupled poroelasticity is validated by benchmarking it against Mandel's problem. The code is used to simulate both quasi-static and dynamic earthquake nucleation and slip distribution on a fault from the combined effect of far field tectonic loading and fluid injection by using an appropriate fault constitutive friction model. Results from the quasi-static induced earthquake simulations show a delayed response in earthquake nucleation. This is attributed to the increased total stress in the domain and not accounting for pressure on the fault. However, this issue is resolved in the final chapter in simulating a single event earthquake dynamic rupture. Simulation results show that fluid pressure has a positive effect on slip nucleation and subsequent crack propagation. This is confirmed by

  4. Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes

    Science.gov (United States)

    2015-11-01

    Memorandum Simulation of Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes...Weld Mechanical Behavior to Include Welding-Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes by Charles R. Fisher...Welding- Induced Residual Stress and Distortion: Coupling of SYSWELD and Abaqus Codes 5a. CONTRACT NUMBER N/A 5b. GRANT NUMBER N/A 5c

  5. ITS - The integrated TIGER series of coupled electron/photon Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Halbleib, J.A.; Mehlhorn, T.A.

    1985-01-01

    The TIGER series of time-independent coupled electron/photon Monte Carlo transport codes is a group of multimaterial, multidimensional codes designed to provide a state-of-the-art description of the production and transport of the electron/photon cascade. The codes follow both electrons and photons from 1.0 GeV down to 1.0 keV, and the user has the option of combining the collisional transport with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence. Source particles can be either electrons or photons. The most important output data are (a) charge and energy deposition profiles, (b) integral and differential escape coefficients for both electrons and photons, (c) differential electron and photon flux, and (d) pulse-height distributions for selected regions of the problem geometry. The base codes of the series differ from one another primarily in their dimensionality and geometric modeling. They include (a) a one-dimensional multilayer code, (b) a code that describes the transport in two-dimensional axisymmetric cylindrical material geometries with a fully three-dimensional description of particle trajectories, and (c) a general three-dimensional transport code which employs a combinatorial geometry scheme. These base codes were designed primarily for describing radiation transport for those situations in which the detailed atomic structure of the transport medium is not important. For some applications, it is desirable to have a more detailed model of the low energy transport. The system includes three additional codes that contain a more elaborate ionization/relaxation model than the base codes. Finally, the system includes two codes that combine the collisional transport of the multidimensional base codes with transport in macroscopic electric and magnetic fields of arbitrary spatial dependence

  6. Enhancement of safety analysis reliability for a CANDU-6 reactor using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Kim, Man Woong; Choi, Yong Seog; Sin, Chul; Kim, Hyun Koon; Kim, Hho Jung; Hwang, Su Hyun; Hong, In Seob; Kim, Chang Hyo

    2005-01-01

    In LOCA analysis of the CANDU reactor, the system thermal-hydraulic code, RELAP-CANDU, alone cannot predict the transient behavior accurately. Therefore, the best estimate neutronics and system thermal-hydraulic coupled code system is necessary to describe the transient behavior with higher accuracy and reliability. To perform on-line calculation of safety analysis for CANDU reactor, a coupled thermal hydraulics-neutronics code system was developed in such a way that the best-estimate thermal-hydraulic system code for CANDU reactor, RELAP-CANDU, is coupled with the full three-dimensional reactor core kinetic code

  7. BALTORO a general purpose code for coupling discrete ordinates and Monte-Carlo radiation transport calculations

    International Nuclear Information System (INIS)

    Zazula, J.M.

    1983-01-01

    The general purpose code BALTORO was written for coupling the three-dimensional Monte-Carlo /MC/ with the one-dimensional Discrete Ordinates /DO/ radiation transport calculations. The quantity of a radiation-induced /neutrons or gamma-rays/ nuclear effect or the score from a radiation-yielding nuclear effect can be analysed in this way. (author)

  8. Simulation and verification studies of reactivity initiated accident by comparative approach of NK/TH coupling codes and RELAP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Ud-Din Khan, Salah [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center; Peng, Minjun [Harbin Engineering Univ. (China). College of Nuclear Science and Technology; Yuntao, Song; Ud-Din Khan, Shahab [Chinese Academy of Sciences, Hefei (China). Inst. of Plasma Physics; Haider, Sajjad [King Saud Univ., Riyadh (Saudi Arabia). Sustainable Energy Technologies Center

    2017-02-15

    The objective is to analyze the safety of small modular nuclear reactors of 220 MWe power. Reactivity initiated accidents (RIA) were investigated by neutron kinetic/thermal hydraulic (NK/TH) coupling approach and thermal hydraulic code i.e., RELAP5. The results obtained by these approaches were compared for validation and accuracy of simulation. In the NK/TH coupling technique, three codes (HELIOS, REMARK, THEATRe) were used. These codes calculate different parameters of the reactor core (fission power, reactivity, fuel temperature and inlet/outlet temperatures). The data exchanges between the codes were assessed by running the codes simultaneously. The results obtained from both (NK/TH coupling) and RELAP5 code analyses complement each other, hence confirming the accuracy of simulation.

  9. An introduction to LIME 1.0 and its use in coupling codes for multiphysics simulations.

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-11-01

    LIME is a small software package for creating multiphysics simulation codes. The name was formed as an acronym denoting 'Lightweight Integrating Multiphysics Environment for coupling codes.' LIME is intended to be especially useful when separate computer codes (which may be written in any standard computer language) already exist to solve different parts of a multiphysics problem. LIME provides the key high-level software (written in C++), a well defined approach (with example templates), and interface requirements to enable the assembly of multiple physics codes into a single coupled-multiphysics simulation code. In this report we introduce important software design characteristics of LIME, describe key components of a typical multiphysics application that might be created using LIME, and provide basic examples of its use - including the customized software that must be written by a user. We also describe the types of modifications that may be needed to individual physics codes in order for them to be incorporated into a LIME-based multiphysics application.

  10. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    International Nuclear Information System (INIS)

    Li, Jia; Jiang, Kecheng; Zhang, Xiaokang; Nie, Xingchen; Zhu, Qinjun; Liu, Songlin

    2016-01-01

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  11. Nuclear-thermal-coupled optimization code for the fusion breeding blanket conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    Li, Jia, E-mail: lijia@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Jiang, Kecheng; Zhang, Xiaokang [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China); Nie, Xingchen [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230027, Anhui (China); Zhu, Qinjun; Liu, Songlin [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, Anhui (China)

    2016-12-15

    Highlights: • A nuclear-thermal-coupled predesign code has been developed for optimizing the radial build arrangement of fusion breeding blanket. • Coupling module aims at speeding up the efficiency of design progress by coupling the neutronics calculation code with the thermal-hydraulic analysis code. • Radial build optimization algorithm aims at optimal arrangement of breeding blanket considering one or multiple specified objectives subject to the design criteria such as material temperature limit and available TBR. - Abstract: Fusion breeding blanket as one of the key in-vessel components performs the functions of breeding the tritium, removing the nuclear heat and heat flux from plasma chamber as well as acting as part of shielding system. The radial build design which determines the arrangement of function zones and material properties on the radial direction is the basis of the detailed design of fusion breeding blanket. For facilitating the radial build design, this study aims for developing a pre-design code to optimize the radial build of blanket with considering the performance of nuclear and thermal-hydraulic simultaneously. Two main features of this code are: (1) Coupling of the neutronics analysis with the thermal-hydraulic analysis to speed up the analysis progress; (2) preliminary optimization algorithm using one or multiple specified objectives subject to the design criteria in the form of constrains imposed on design variables and performance parameters within the possible engineering ranges. This pre-design code has been applied to the conceptual design of water-cooled ceramic breeding blanket in project of China fusion engineering testing reactor (CFETR).

  12. CEPXS/ONELD: A one-dimensional coupled electron-photon discrete ordinates code package

    International Nuclear Information System (INIS)

    Lorence, L.J. Jr.; Morel, J.E.

    1992-01-01

    CEPXS/ONELD is a discrete ordinates transport code package that can model the electron-photon cascade from 100 MeV to 1 keV. The CEPXS code generates fully-coupled multigroup-Legendre cross section data. This data is used by the general-purpose discrete ordinates code, ONELD, which is derived from the Los Alamos ONEDANT and ONBTRAN codes. Version 1.0 of CEPXS/ONELD was released in 1989 and has been primarily used to analyze the effect of radiation environments on electronics. Version 2.0 is under development and will include user-friendly features such as the automatic selection of group structure, spatial mesh structure, and S N order

  13. Monitoring and preventing numerical oscillations in 3D simulations with coupled Monte Carlo codes

    International Nuclear Information System (INIS)

    Kotlyar, D.; Shwageraus, E.

    2014-01-01

    Highlights: • Conventional coupling methods used in all MC codes can be numerically unstable. • Application of new stochastic implicit (SIMP) methods may be required. • The implicit methods require additional computational effort. • Monitoring diagnostic of the numerical stability was developed here. • The procedure allows to create an hybrid explicit–implicit coupling scheme. - Abstract: Previous studies have reported that different schemes for coupling Monte Carlo (MC) neutron transport with burnup and thermal hydraulic feedbacks may potentially be numerically unstable. This issue can be resolved by application of implicit methods, such as the stochastic implicit mid-point (SIMP) methods. In order to assure numerical stability, the new methods do require additional computational effort. The instability issue however, is problem-dependent and does not necessarily occur in all cases. Therefore, blind application of the unconditionally stable coupling schemes, and thus incurring extra computational costs, may not always be necessary. In this paper, we attempt to develop an intelligent diagnostic mechanism, which will monitor numerical stability of the calculations and, if necessary, switch from simple and fast coupling scheme to more computationally expensive but unconditionally stable one. To illustrate this diagnostic mechanism, we performed a coupled burnup and TH analysis of a single BWR fuel assembly. The results indicate that the developed algorithm can be easily implemented in any MC based code for monitoring of numerical instabilities. The proposed monitoring method has negligible impact on the calculation time even for realistic 3D multi-region full core calculations

  14. Sub-step methodology for coupled Monte Carlo depletion and thermal hydraulic codes

    International Nuclear Information System (INIS)

    Kotlyar, D.; Shwageraus, E.

    2016-01-01

    Highlights: • Discretization of time in coupled MC codes determines the results’ accuracy. • The error is due to lack of information regarding the time-dependent reaction rates. • The proposed sub-step method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. • The reaction rates are varied as functions of nuclide densities and TH conditions. - Abstract: The governing procedure in coupled Monte Carlo (MC) codes relies on discretization of the simulation time into time steps. Typically, the MC transport solution at discrete points will generate reaction rates, which in most codes are assumed to be constant within the time step. This assumption can trigger numerical instabilities or result in a loss of accuracy, which, in turn, would require reducing the time steps size. This paper focuses on reducing the time discretization error without requiring additional MC transport solutions and hence with no major computational overhead. The sub-step method presented here accounts for the reaction rate variation due to the variation in nuclide densities and thermal hydraulic (TH) conditions. This is achieved by performing additional depletion and TH calculations within the analyzed time step. The method was implemented in BGCore code and subsequently used to analyze a series of test cases. The results indicate that computational speedup of up to a factor of 10 may be achieved over the existing coupling schemes.

  15. Spatiotemporal coding of inputs for a system of globally coupled phase oscillators

    Science.gov (United States)

    Wordsworth, John; Ashwin, Peter

    2008-12-01

    We investigate the spatiotemporal coding of low amplitude inputs to a simple system of globally coupled phase oscillators with coupling function g(ϕ)=-sin(ϕ+α)+rsin(2ϕ+β) that has robust heteroclinic cycles (slow switching between cluster states). The inputs correspond to detuning of the oscillators. It was recently noted that globally coupled phase oscillators can encode their frequencies in the form of spatiotemporal codes of a sequence of cluster states [P. Ashwin, G. Orosz, J. Wordsworth, and S. Townley, SIAM J. Appl. Dyn. Syst. 6, 728 (2007)]. Concentrating on the case of N=5 oscillators we show in detail how the spatiotemporal coding can be used to resolve all of the information that relates the individual inputs to each other, providing that a long enough time series is considered. We investigate robustness to the addition of noise and find a remarkable stability, especially of the temporal coding, to the addition of noise even for noise of a comparable magnitude to the inputs.

  16. Heterogeneous redox reactions in groundwater flow systems - Investigation and application of two different coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Pfingsten, W.; Carnahan, C.L. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1995-05-01

    Two simulators of reactive chemical transport are applied to a set of problems involving heterogeneous reactions of uranium species. The simulators use similar algorithms to compute the heterogeneous chemical equilibria, but they use different approaches to the computation of solute transport and to the coupling of transport with chemical reactions. One simulator (MCOTAC) sequentially couples calculations of static chemical equilibria to a random-walk simulation of solute advection and dispersion. The other simulator (THCC) directly couples mass action relations for chemical equilibria to finite-difference representations of the solute transport equations. The aim of the comparison was to demonstrate the applicability of the newly developed code MCOTAC to redox problems, and to identify and investigate general differences between the two types of codes within these applications. The chosen heterogeneous redox systems are hypothetically generate systems which provide numerical difficulties within the coupled code calculation. Uranium, an important component of heterogeneous redox systems consisting of uraniferous solids and natural groundwaters, was chosen as a main component in the example redox systems because of practical interest for performance assessment of geological repositories for nuclear wastes. The calculations show reasonable agreement, in general, between the two computational approaches. Specific areas of disagreement arise from numerical difficulties to each approach. Such `benchmarking` can enhance confidence in the overall performance of individual simulators while identifying aspects that may require further investigations and possible modifications. (author) figs., tabs., 7 refs.

  17. CRA Control Logic Realization for MARS 1-D/MASTER coupled Code System

    International Nuclear Information System (INIS)

    Han, Soonkyoo; Jeong, Sungsu; Lee, Suyong

    2013-01-01

    Both Multi-dimensional Analysis Reactor Safety (MARS) code and Multi-purpose Analyzer for Static and Transient Effects of Reactors (MASTER) code, developed by Korea Atomic Energy Research Institute (KAERI), can be coupled for various simulations of nuclear reactor system. In the MARS 1-D/MASTER coupled code system, MARS is used for the thermal hydraulic calculations and MASTER is used for reactor core calculations. In case of using this coupled code system, the movements of control rod assembly (CRA) are controlled by MASTER. MASTER, however, has a CRA control function which is inputted by user as a form of time dependent table. When simulations related to sequential CRA insertion or withdrawal which are not ejection or drop are performed, this CRA control function is not sufficient to demonstrate the process of CRA movements. Therefore an alternative way is proposed for realization of CRA control logic in MASTER. In this study, the manually realized CRA control logic was applied by inputting the time dependent CRA positions into MASTER. And the points of CRA movements were decided by iterations. At the end of CRA movement, the reactor power difference and the average coolant temperature difference were not out of the range of their dead bands. Therefore it means that this manually realized CRA control logic works appropriately in the dead bands of the logic. Therefore the proper CRA movement points could be decided by using this manually realized CRA control logic. Based on these results, it is verified that the proper CRA movement points can be chosen by using the proposed CRA control logic in this article. In conclusion, it is expected that this proposed CRA control logic in MASTER can be used to properly demonstrate the process related to CRA sequential movements in the MARS 1-D/MASTER coupled code system

  18. Preliminary Development of the MARS/FREK Spatial Kinetics Coupled System Code for Square Fueled Fast Reactor Applications

    International Nuclear Information System (INIS)

    Bae, Moo Hoon; Joo, Han Gyu

    2009-01-01

    Incorporation of a three-dimensional (3-D) reactor kinetics model into a system thermal-hydraulic (T/H) code enhances the capability to perform realistic analyses of the core neutronic behavior and the plant system dynamics which are coupled each other. For this advantage, several coupled system T/H and spatial kinetics codes, such as RELAP/PARCS, RELAP5/ PANBOX, and MARS/MASTER have been developed. These codes, however, so far limited to LWR applications. The objective of this work is to develop such a coupled code for fast reactor applications. Particularly, applications to lead-bismuth eutectic (LBE) cooled fast reactor are of interest which employ open square lattices. A fast reactor kinetics code applicable to square fueled cores called FREK is coupled the LBE version of the MARS code. The MARS/MASTER coupled code is used as the reference for the integration. The coupled code MARS/FREK is examined for a conceptual reactor called P-DEMO which is being developed by NUTRECK. In order to check the validity of the coupled code, however, the OECD MSLB benchmark exercise III calculation is solved first

  19. User Guide for the R5EXEC Coupling Interface in the RELAP5-3D Code

    Energy Technology Data Exchange (ETDEWEB)

    Forsmann, J. Hope [Idaho National Lab. (INL), Idaho Falls, ID (United States); Weaver, Walter L. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-04-01

    This report describes the R5EXEC coupling interface in the RELAP5-3D computer code from the users perspective. The information in the report is intended for users who want to couple RELAP5-3D to other thermal-hydraulic, neutron kinetics, or control system simulation codes.

  20. Coupling a system code with computational fluid dynamics for the simulation of complex coolant reactivity effects

    International Nuclear Information System (INIS)

    Bertolotto, D.

    2011-11-01

    The current doctoral research is focused on the development and validation of a coupled computational tool, to combine the advantages of computational fluid dynamics (CFD) in analyzing complex flow fields and of state-of-the-art system codes employed for nuclear power plant (NPP) simulations. Such a tool can considerably enhance the analysis of NPP transient behavior, e.g. in the case of pressurized water reactor (PWR) accident scenarios such as Main Steam Line Break (MSLB) and boron dilution, in which strong coolant flow asymmetries and multi-dimensional mixing effects strongly influence the reactivity of the reactor core, as described in Chap. 1. To start with, a literature review on code coupling is presented in Chap. 2, together with the corresponding ongoing projects in the international community. Special reference is made to the framework in which this research has been carried out, i.e. the Paul Scherrer Institute's (PSI) project STARS (Steady-state and Transient Analysis Research for the Swiss reactors). In particular, the codes chosen for the coupling, i.e. the CFD code ANSYS CFX V11.0 and the system code US-NRC TRACE V5.0, are part of the STARS codes system. Their main features are also described in Chap. 2. The development of the coupled tool, named CFX/TRACE from the names of the two constitutive codes, has proven to be a complex and broad-based task, and therefore constraints had to be put on the target requirements, while keeping in mind a certain modularity to allow future extensions to be made with minimal efforts. After careful consideration, the coupling was defined to be on-line, parallel and with non-overlapping domains connected by an interface, which was developed through the Parallel Virtual Machines (PVM) software, as described in Chap. 3. Moreover, two numerical coupling schemes were implemented and tested: a sequential explicit scheme and a sequential semi-implicit scheme. Finally, it was decided that the coupling would be single

  1. NSLINK, Coupling of NJOY Cross-Sections Generator Code to SCALE-3 System

    International Nuclear Information System (INIS)

    De Leege, P.F.A

    1991-01-01

    1 - Description of program or function: NSLINK (NJOY - SCALE - LINK) is a set of computer codes to couple the NJOY cross-section generation code to the SCALE-3 code system (using AMPX-2 master library format) retaining the Nordheim resolved resonance treatment option. 2 - Method of solution: The following module and codes are included in NSLINK: XLACSR: This module is a stripped-down version of the XLACS-2 code. The module passes all l=0 resonance parameters as well as the contribution from all other resonances to the group cross-sections, the contribution from the wings of the l=0 resonances, the background cross-section and possible interference for multilevel Breit-Wigner resonance parameters. The group cross-sections are stored in the appropriate 1-D cross-section arrays. The output file has AMPX-2 master format. The original NJOY code is used to calculate all other data. The XLACSR module is included in the NJOY code. MILER: This code converts NJOY output (GENDF format) to AMPX-2 master format. The code is an extensively revised version of the original MILER code. In addition, the treatment of thermal scattering matrices at different temperatures is included. UNITABR: This code is a revised version of the UNITAB code. It merges the output of XLACSR and MILER in such a way that contributions from the bodies of the l=0 resonances in the resolved energy range, calculated by XLACSR, are subtracted from the 1-D group cross-section arrays for fission (MT=18) and neutron capture (MT=102). The l=0 resonance parameters and the contributions from the bodies of these resonances are added separately (MT=1023, 1022 and 1021). The total cross-section (MT=1), the absorption cross- section (MT=27) and the neutron removal cross-section (MT=101) values are adjusted. In the case of Bondarenko data, infinite dilution values of the cross-sections (MT=1, 18 and 102) are changed in the same way as the 1-D cross-section. The output file of UNITABR is in AMPX-2 master format and

  2. A guide to the coupled chemical equilibria and migration code CHEQMATE

    International Nuclear Information System (INIS)

    Haworth, A.; Sharland, S.M.; Tasker, P.W.; Tweed, C.J.

    1988-02-01

    The CHEQMATE (CHemical EQuilibrium with Migration and Transport Equations) program has been developed to model the evolution of spatially inhomogeneous aqueous chemical systems. CHEQMATE models one-dimensional diffusion and electromigration of ionic species with chemical equilibration provided by the geochemical code PHREEQE. The transport and chemical parts of the CHEQMATE code are iteratively coupled, so that local chemical equilibrium is maintained as the transport processes evolve. CHEQMATE is very flexible and can easily be applied to many different evolving chemical systems. It has principally been used to study the evolution of the chemical environment in and around a nuclear waste repository. (author)

  3. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    International Nuclear Information System (INIS)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae

    1998-01-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others

  4. MARS 1.3 system analysis code coupling with CONTEMPT4/MOD5/PCCS containment analysis code using dynamic link library

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Bub Dong; Jeong, Jae Jun; Lee, Won Jae [KAERI, Taejon (Korea, Republic of)

    1998-10-01

    The two independent codes, MARS 1.3 and CONTEMPT4/MOD5/PCCS, have been coupled using the method of dynamic-link-library (DLL) technique. Overall configuration of the code system is designed so that MARS will be a main driver program which use CONTEMPT as associated routines. Using Digital Visual Fortran compiler, DLL was generated from the CONTEMPT source code with the interfacing routine names and arguments. Coupling of MARS with CONTEMPT was realized by calling the DLL routines at the appropriate step in the MARS code. Verification of coupling was carried out for LBLOCA transient of a typical plant design. It was found that the DLL technique is much more convenient than the UNIX process control techniques and effective for Window operating system. Since DLL can be used by more than one application and an application program can use many DLLs simultaneously, this technique would enable the existing codes to use more broadly with linking others.

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

    International Nuclear Information System (INIS)

    Hussein, M.S; Lewis, B.J.; Bonin, H.W.

    2013-01-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k eff calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k eff calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k eff calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Hussein, M.S, E-mail: mohamed.hussein@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada); Lewis, B.J., E-mail: Brent.Lewis@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada); Bonin, H.W., E-mail: bonin-h@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    The theory of multipoint coupled reactors developed by multi-group transport is verified by using the probabilistic transport code MCNP5 and the continuous-energy Monte Carlo reactor physics burnup calculation Serpent code. The verification was performed by calculating the multiplication factors (or criticality factors) and coupling coefficients for a two-region test reactor known as the Deuterium Critical Assembly, DCA. The multiplication factors k{sub eff} calculated numerically and independently from simulations of the DCA by MCNP5 and Serpent codes are compared with the multiplication factors k{sub eff} calculated based on the coupled reactor theory. Excellent agreement was obtained between the multiplication factors k{sub eff} calculated with the Serpent code, with MCNP5, and from the coupled reactor theory. This analysis demonstrates that the Serpent code is valid for the multipoint coupled reactor calculations. (author)

  7. Development of Coupled Interface System between the FADAS Code and a Source-term Evaluation Code XSOR for CANDU Reactors

    International Nuclear Information System (INIS)

    Son, Han Seong; Song, Deok Yong; Kim, Ma Woong; Shin, Hyeong Ki; Lee, Sang Kyu; Kim, Hyun Koon

    2006-01-01

    An accident prevention system is essential to the industrial security of nuclear industry. Thus, the more effective accident prevention system will be helpful to promote safety culture as well as to acquire public acceptance for nuclear power industry. The FADAS(Following Accident Dose Assessment System) which is a part of the Computerized Advisory System for a Radiological Emergency (CARE) system in KINS is used for the prevention against nuclear accident. In order to enhance the FADAS system more effective for CANDU reactors, it is necessary to develop the various accident scenarios and reliable database of source terms. This study introduces the construction of the coupled interface system between the FADAS and the source-term evaluation code aimed to improve the applicability of the CANDU Integrated Safety Analysis System (CISAS) for CANDU reactors

  8. DYNREL - the reference calculation (coupled code utilization on analysis of RIA-transient)

    International Nuclear Information System (INIS)

    Strmensky, C.; Darilek, P.

    2003-01-01

    DYNREL is coupled code, comprising DYN3D and RELAP5 programs. The coupled code has been developed during four years. Now DYNREL is tested on selected RIA and thermo-hydraulic transient calculations. This material describes some results from selected RIA transient calculation (initiated by control rod movement). DYNREL modelled the whole nuclear reactors. The core is modeled as 313 or 349 independent thermo-hydraulic channels with 10 or 20 axial layers. Thermo-hydraulic part contains about 700 components that covered the six loops' model of nuclear power plant in detail. The calculated results are compared with DYN3D/M3, DYN3D/H1.1 results (Authors)

  9. Uncertainty propagation applied to multi-scale thermal-hydraulics coupled codes. A step towards validation

    Energy Technology Data Exchange (ETDEWEB)

    Geffray, Clotaire Clement

    2017-03-20

    The work presented here constitutes an important step towards the validation of the use of coupled system thermal-hydraulics and computational fluid dynamics codes for the simulation of complex flows in liquid metal cooled pool-type facilities. First, a set of methods suited for uncertainty and sensitivity analysis and validation activities with regards to the specific constraints of the work with coupled and expensive-to-run codes is proposed. Then, these methods are applied to the ATHLET - ANSYS CFX model of the TALL-3D facility. Several transients performed at this latter facility are investigated. The results are presented, discussed and compared to the experimental data. Finally, assessments of the validity of the selected methods and of the quality of the model are offered.

  10. ITS Version 6 : the integrated TIGER series of coupled electron/photon Monte Carlo transport codes.

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2008-04-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of lineartime-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 6, the latest version of ITS, contains (1) improvements to the ITS 5.0 codes, and (2) conversion to Fortran 90. The general user friendliness of the software has been enhanced through memory allocation to reduce the need for users to modify and recompile the code.

  11. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    International Nuclear Information System (INIS)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min

    2016-01-01

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI

  12. Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI.

  13. Multidimensional method of spatially coupled approximation to the transverse escape in nodal codes

    International Nuclear Information System (INIS)

    Jatuff, F.E.

    1990-01-01

    A natural extension of the polynomic development programmed in RHENO code is presented, which adds to the variable order one-dimensional functions sum, a number of terms that represent functions of production. These new terms, which provide a direct determination of transverse escapes, are calculated from the new variables coupling among nodes: the 4 fluxes in rectangle vortices (bidimensional Cartesian geometry) or the 12 fluxes half-way through the parallelepiped edges (tridimensional Cartesian geometry). (Author) [es

  14. Learning of spatio-temporal codes in a coupled oscillator system.

    Science.gov (United States)

    Orosz, Gábor; Ashwin, Peter; Townley, Stuart

    2009-07-01

    In this paper, we consider a learning strategy that allows one to transmit information between two coupled phase oscillator systems (called teaching and learning systems) via frequency adaptation. The dynamics of these systems can be modeled with reference to a number of partially synchronized cluster states and transitions between them. Forcing the teaching system by steady but spatially nonhomogeneous inputs produces cyclic sequences of transitions between the cluster states, that is, information about inputs is encoded via a "winnerless competition" process into spatio-temporal codes. The large variety of codes can be learned by the learning system that adapts its frequencies to those of the teaching system. We visualize the dynamics using "weighted order parameters (WOPs)" that are analogous to "local field potentials" in neural systems. Since spatio-temporal coding is a mechanism that appears in olfactory systems, the developed learning rules may help to extract information from these neural ensembles.

  15. A perturbation-based susbtep method for coupled depletion Monte-Carlo codes

    International Nuclear Information System (INIS)

    Kotlyar, Dan; Aufiero, Manuele; Shwageraus, Eugene; Fratoni, Massimiliano

    2017-01-01

    Highlights: • The GPT method allows to calculate the sensitivity coefficients to any perturbation. • Full Jacobian of sensitivities, cross sections (XS) to concentrations, may be obtained. • The time dependent XS is obtained by combining the GPT and substep methods. • The proposed GPT substep method considerably reduces the time discretization error. • No additional MC transport solutions are required within the time step. - Abstract: Coupled Monte Carlo (MC) methods are becoming widely used in reactor physics analysis and design. Many research groups therefore, developed their own coupled MC depletion codes. Typically, in such coupled code systems, neutron fluxes and cross sections are provided to the depletion module by solving a static neutron transport problem. These fluxes and cross sections are representative only of a specific time-point. In reality however, both quantities would change through the depletion time interval. Recently, Generalized Perturbation Theory (GPT) equivalent method that relies on collision history approach was implemented in Serpent MC code. This method was used here to calculate the sensitivity of each nuclide and reaction cross section due to the change in concentration of every isotope in the system. The coupling method proposed in this study also uses the substep approach, which incorporates these sensitivity coefficients to account for temporal changes in cross sections. As a result, a notable improvement in time dependent cross section behavior was obtained. The method was implemented in a wrapper script that couples Serpent with an external depletion solver. The performance of this method was compared with other existing methods. The results indicate that the proposed method requires substantially less MC transport solutions to achieve the same accuracy.

  16. Computation of a BWR Turbine Trip with CATHARE-CRONOS2-FLICA4 Coupled Codes

    International Nuclear Information System (INIS)

    Mignot, G.; Royer, E.; Rameau, B.; Todorova, N.

    2004-01-01

    The CEA/DEN modeling and computation results with the CATHARE, CRONOS2, and FLICA4 codes of the Organisation for Economic Co-operation and Development boiling water reactor turbine trip benchmark are presented. The first exercise of the benchmark to model the whole reactor thermal hydraulics with specified power has been performed with the CATHARE system code. Exercise 2, devoted to core thermal-hydraulic neutronic analysis with provided boundary conditions and neutronic cross sections, has been carried out with the CRONOS2 and FLICA4 codes. Finally, exercise 3, combining system thermal hydraulics and core three-dimensional thermal-hydraulics-neutronics, was computed with the three coupled codes: CATHARE, CRONOS2, and FLICA4.Our one-dimensional thermal-hydraulic reactor computation agrees well with the benchmark reference data and demonstrates the capacities of CATHARE to model a turbine trip transient. Coupled three-dimensional thermal-hydraulic and neutronic analysis displays a high sensitivity of the power peak to the core thermal-hydraulic model. The use of at least 100 channels is recommended to achieve reasonable results for integral and local parameters. Deviations between experimental data and exercise 3 results are discussed: timing of events, core pressure drop, and neutronic model. Finally, analysis of extreme scenarios as sensitivity studies on the transient to assess the effect of the scram, the bypass relief valve, and the steam relief valves is presented

  17. Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

    International Nuclear Information System (INIS)

    Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y.

    2014-01-01

    To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications

  18. Applicability of Coupled Thermalhydraulic Codes for Safety Analysis of Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gairola, A.; Bhowmik, P. K.; Shamim, J. A.; Suh, K. Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    To this end computational codes like RELAP and TRACE are used to model thermal-hydraulic response of nuclear power plant during an accident. By careful modeling and significant user experience these system codes are able to simulate the behavior of primary system and the containment to a reasonable extent. Comparatively decoupled simulation is simple but might not produce reality and the physics involved in an accurate manner. Thus simulation using two different system codes is interesting as the whole system is coupled through the pressure in the containment and flow through the break. Using this methodology it might be possible to get new insight about the primary and containment behavior by the precise simulation of the accident both in the current reactors and future Gen-III/III+ reactors. Couple thermalhydraulic code methodology is still new and require further investigations. Applicability of such methodology to the GEN-II plants have met with limited success, however a number of situations in which this methodology could be applied are still unexplored and thus provides a room for improvement and modifications.

  19. Analysis of some antecipated transients without scram for PWR type reactors by coupling of the CORAN code to the ALMOD code system

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    This study investigates some antecipated transients without scram for a pressurized water cooled reactor, using coupling of the containment CORAN code to the ALMOD code system, under severe random conditions. This coupling has the objective of including containment model as part of an unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle, a failure in the closure of the pressurizer relief valve was also investigated. (Author) [pt

  20. Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Pasichnyk, Ihor; Zwermann, Winfried; Velkov, Kiril [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Nikonov, Sergey [VNIIAES, Moscow (Russian Federation)

    2016-09-15

    An uncertainty and sensitivity analysis is performed for the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). A switch off of one main coolant pump (MCP) at nominal reactor power is calculated using a coupled thermohydraulic and neutron-kinetic ATHLET-PARCS code. The objectives are to study uncertainty of total reactor power and to identify the main sources of reactor power uncertainty. The GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. A set of most important thermal-hydraulic parameters of the primary circuit is identified and a total of 23 thermohydraulic parameters are statistically varied using GRS code SUSA. The ATHLET model contains also a balance-of-plant (BOP) model which is simulated using ATHLET GCSM module. In particular the operation of the main steam generator regulators is modelled in detail. A set of 200 varied coupled ATHLET-PARCS calculations is analyzed. The results obtained show a clustering effect in the behavior of global reactor parameters. It is found that the GCSM system together with varied input parameters strongly influence the overall nuclear power plant behavior and can even lead to a new scenario. Possible reasons of the clustering effect are discussed in the paper. This work is a step forward in establishing a ''best-estimate calculations in combination with performing uncertainty analysis'' methodology for coupled full core calculations.

  1. Validation of coupled neutronic / thermal-hydraulic codes for VVER reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Mittag, S.; Grundmann, U.; Kliem, S.; Kozmenkov, Y.; Rindelhardt, U.; Rohde, U.; Weiss, F.-P.; Langenbuch, S.; Krzykacz-Hausmann, B.; Schmidt, K.-D.; Vanttola, T.; Haemaelaeinen, A.; Kaloinen, E.; Kereszturi, A.; Hegyi, G.; Panka, I.; Hadek, J.; Strmensky, C.; Darilek, P.; Petkov, P.; Stefanova, S.; Kuchin, A.; Khalimonchuk, V.; Hlbocky, P.; Sico, D.; Danilin, S.; Ionov, V.; Nikonov, S.; Powney, D.

    2004-08-01

    In recent years, the simulation methods for the safety analysis of nuclear power plants have been continuously improved to perform realistic calculations. Therefore in VALCO work package 2 (WP 2), the usual application of coupled neutron-kinetic / thermal-hydraulic codes to VVER has been supplemented by systematic uncertainty and sensitivity analyses. A comprehensive uncertainty analysis has been carried out. The GRS uncertainty and sensitivity method based on the statistical code package SUSA was applied to the two transients studied earlier in SRR-1/95: A load drop of one turbo-generator in Loviisa-1 (VVER-440), and a switch-off of one feed water pump in Balakovo-4 (VVER-1000). The main steps of these analyses and the results obtained by applying different coupled code systems (SMABRE - HEXTRAN, ATHLET - DYN3D, ATHLET - KIKO3D, ATHLET - BIPR-8) are described in this report. The application of this method is only based on variations of input parameter values. No internal code adjustments are needed. An essential result of the analysis using the GRS SUSA methodology is the identification of the input parameters, such as the secondary-circuit pressure, the control-assembly position (as a function of time), and the control-assembly efficiency, that most sensitively affect safety-relevant output parameters, like reactor power, coolant heat-up, and primary pressure. Uncertainty bands for these output parameters have been derived. The variation of potentially uncertain input parameter values as a consequence of uncertain knowledge can activate system actions causing quite different transient evolutions. This gives indications about possible plant conditions that might be reached from the initiating event assuming only small disturbances. In this way, the uncertainty and sensitivity analysis reveals the spectrum of possible transient evolutions. Deviations of SRR-1/95 coupled code calculations from measurements also led to the objective to separate neutron kinetics from

  2. The coupled code system DORT-TD/THERMIX and its application to the OECD/NEA/NSC PBMR400 MW coupled neutronics thermal hydraulics transient benchmark

    International Nuclear Information System (INIS)

    Pautz, A.; Tyobeka, B.; Ivanov, K.

    2009-01-01

    In new reactor designs that are still under review such as the Pebble Bed Modular Reactor (PBMR), not much experimental data exists to benchmark newly developed computer codes against. Such a situation requires that nuclear engineers and designers of this novel reactor design must resort to the validation of a newly developed code through a code-to-code benchmarking exercise because there are validated codes that are currently in use to analyze this reactor design, albeit very few of them. There are numerous HTR core physics benchmarks that are currently being pursued by different organizations, for different purposes. One such benchmark exercise is the PBMR-400MW OECD/NEA coupled neutronics/thermal hydraulics transient benchmark. In this paper, a newly developed coupled neutronics thermal hydraulics code system, DORT-TD/THERMIX with both transport and diffusion theory options, is used to simulate both the steady-state as well as several transient scenarios in this benchmark problem. (orig.)

  3. Advanced methodology to simulate boiling water reactor transient using coupled thermal-hydraulic/neutron-kinetic codes

    Energy Technology Data Exchange (ETDEWEB)

    Hartmann, Christoph Oliver

    2016-06-13

    Coupled Thermal-hydraulic/Neutron-kinetic (TH/NK) simulations of Boiling Water Reactor transients require well validated and accurate simulation tools. The generation of cross-section (XS) libraries, depending on the individual thermal-hydraulic state parameters, is of paramount importance for coupled simulations. Problem-dependent XS-sets for 3D core simulations are being generated mainly by well validated, fast running commercial and user-friendly lattice codes such as CASMO and HELIOS. In this dissertation a computational route, based on the lattice code SCALE6/TRITON, the cross-section interface GenPMAXS, the best-estimate thermal-hydraulic system code TRACE and the core simulator PARCS, for best-estimate simulations of Boiling Water (BWR) transients has been developed and validated. The computational route has been supplemented by a subsequent uncertainty and sensitivity study based on Monte Carlo sampling and propagation of the uncertainties of input parameters to the output (SUSA code). The analysis of a single BWR fuel assembly depletion problem with PARCS using SCALE/TRITON cross-sections has been shown a good agreement with the results obtained with CASMO cross-section sets. However, to compensate the deficiencies of the interface program GenPMAXS, PYTHON scripts had to be developed to incorporate missing data, as the yields of Iodine, Xenon and Promethium, into the cross-section-data sets (PMAXS-format) generated by GenPMAXS from the SCALE/TRITON output. The results of the depletion analysis of a full BWR core with PARCS have indicated the importance of considering history effects, adequate modeling of the reflector region and the control rods, as the PARCS simulations for depleted fuel and all control rods inserted (ARI) differs significantly at the fuel assembly top and bottom. Systematic investigations with the coupled codes TRACE/PARCS have been performed to analyse the core behaviour at different thermal conditions using nuclear data (XS

  4. Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs

    International Nuclear Information System (INIS)

    Mejia S, D. M.; Del Valle G, E.

    2015-09-01

    The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)

  5. Development of a general coupling interface for the fuel performance code transuranus tested with the reactor dynamic code DYN3D

    International Nuclear Information System (INIS)

    Holt, L.; Rohde, U.; Seidl, M.; Schubert, A.; Van Uffelen, P.

    2013-01-01

    Several institutions plan to couple the fuel performance code TRANSURANUS developed by the European Institute for Transuranium Elements with their own codes. One of these codes is the reactor dynamic code DYN3D maintained by the Helmholtz-Zentrum Dresden - Rossendorf. DYN3D was developed originally for VVER type reactors and was extended later to western type reactors. Usually, the fuel rod behavior is modeled in thermal hydraulics and neutronic codes in a simplified manner. The main idea of this coupling is to describe the fuel rod behavior in the frame of core safety analysis in a more detailed way, e.g. including the influence of the high burn-up structure, geometry changes and fission gas release. It allows to take benefit from the improved computational power and software achieved over the last two decades. The coupling interface was developed in a general way from the beginning. Thence it can be easily used also by other codes for a coupling with TRANSURANUS. The user can choose between a one-way as well as a two-way online coupling option. For a one-way online coupling, DYN3D provides only the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, but the fuel performance code doesn’t transfer any variable back to DYN3D. In a two-way online coupling, TRANSURANUS in addition transfers parameters like fuel temperature and cladding temperature back to DYN3D. This list of variables can be extended easily by geometric and further variables of interest. First results of the code system DYN3D-TRANSURANUS will be presented for a control rod ejection transient in a modern western type reactor. Pre-analyses show already that a detailed fuel rod behavior modeling will influence the thermal hydraulics and thence also the neutronics due to the Doppler reactivity effect of the fuel temperature. The coupled code system has therefore a potential to improve the assessment of safety criteria. The developed code system DYN3D-TRANSURANUS can be used also

  6. A computational study on altered theta-gamma coupling during learning and phase coding.

    Directory of Open Access Journals (Sweden)

    Xuejuan Zhang

    Full Text Available There is considerable interest in the role of coupling between theta and gamma oscillations in the brain in the context of learning and memory. Here we have used a neural network model which is capable of producing coupling of theta phase to gamma amplitude firstly to explore its ability to reproduce reported learning changes and secondly to memory-span and phase coding effects. The spiking neural network incorporates two kinetically different GABA(A receptor-mediated currents to generate both theta and gamma rhythms and we have found that by selective alteration of both NMDA receptors and GABA(A,slow receptors it can reproduce learning-related changes in the strength of coupling between theta and gamma either with or without coincident changes in theta amplitude. When the model was used to explore the relationship between theta and gamma oscillations, working memory capacity and phase coding it showed that the potential storage capacity of short term memories, in terms of nested gamma-subcycles, coincides with the maximal theta power. Increasing theta power is also related to the precision of theta phase which functions as a potential timing clock for neuronal firing in the cortex or hippocampus.

  7. Radiation Coupling with the FUN3D Unstructured-Grid CFD Code

    Science.gov (United States)

    Wood, William A.

    2012-01-01

    The HARA radiation code is fully-coupled to the FUN3D unstructured-grid CFD code for the purpose of simulating high-energy hypersonic flows. The radiation energy source terms and surface heat transfer, under the tangent slab approximation, are included within the fluid dynamic ow solver. The Fire II flight test, at the Mach-31 1643-second trajectory point, is used as a demonstration case. Comparisons are made with an existing structured-grid capability, the LAURA/HARA coupling. The radiative surface heat transfer rates from the present approach match the benchmark values within 6%. Although radiation coupling is the focus of the present work, convective surface heat transfer rates are also reported, and are seen to vary depending upon the choice of mesh connectivity and FUN3D ux reconstruction algorithm. On a tetrahedral-element mesh the convective heating matches the benchmark at the stagnation point, but under-predicts by 15% on the Fire II shoulder. Conversely, on a mixed-element mesh the convective heating over-predicts at the stagnation point by 20%, but matches the benchmark away from the stagnation region.

  8. The European source term code ESTER - basic ideas and tools for coupling of ATHLET and ESTER

    International Nuclear Information System (INIS)

    Schmidt, F.; Schuch, A.; Hinkelmann, M.

    1993-04-01

    The French software house CISI and IKE of the University of Stuttgart have developed during 1990 and 1991 in the frame of the Shared Cost Action Reactor Safety the informatic structure of the European Source TERm Evaluation System (ESTER). Due to this work tools became available which allow to unify on an European basis both code development and code application in the area of severe core accident research. The behaviour of reactor cores is determined by thermal hydraulic conditions. Therefore for the development of ESTER it was important to investigate how to integrate thermal hydraulic code systems with ESTER applications. This report describes the basic ideas of ESTER and improvements of ESTER tools in view of a possible coupling of the thermal hydraulic code system ATHLET and ESTER. Due to the work performed during this project the ESTER tools became the most modern informatic tools presently available in the area of severe accident research. A sample application is given which demonstrates the use of the new tools. (orig.) [de

  9. Development and validation of the fast doppler broadening module coupled within RMC code

    International Nuclear Information System (INIS)

    Yu Jiankai; Liang Jin'gang; Yu Ganglin; Wang Kan

    2015-01-01

    It is one of the efficient approach to reduce the memory consumption in Monte Carlo based reactor physical simulations by using the On-the-fly Doppler broadening for temperature dependent nuclear cross sections. RXSP is a nuclear cross sections processing code being developed by REAL team in Department of Engineering Physics in Tsinghua University, which has an excellent performance in Doppler broadening the temperature dependent continuous energy neutron cross sections. To meet the dual requirements of both accuracy and efficiency during the Monte Carlo simulations with many materials and many temperatures in it, this work enables the capability of on-the-fly pre-Doppler broadening cross sections during the neutron transport by coupling the Fast Doppler Broaden module in RXSP code embedded in the RMC code also being developed by REAL team in Tsinghua University. Additionally, the original OpenMP-based parallelism has been successfully converted into the MPI-based framework, being fully compatible with neutron transport in RMC code, which has achieved a vast parallel efficiency improvement. This work also provides a flexible approach to solve Monte Carlo based full core depletion calculation with many temperatures feedback in many isotopes. (author)

  10. Coupling Hydrodynamic and Wave Propagation Codes for Modeling of Seismic Waves recorded at the SPE Test.

    Science.gov (United States)

    Larmat, C. S.; Rougier, E.; Delorey, A.; Steedman, D. W.; Bradley, C. R.

    2016-12-01

    The goal of the Source Physics Experiment (SPE) is to bring empirical and theoretical advances to the problem of detection and identification of underground nuclear explosions. For this, the SPE program includes a strong modeling effort based on first principles calculations with the challenge to capture both the source and near-source processes and those taking place later in time as seismic waves propagate within complex 3D geologic environments. In this paper, we report on results of modeling that uses hydrodynamic simulation codes (Abaqus and CASH) coupled with a 3D full waveform propagation code, SPECFEM3D. For modeling the near source region, we employ a fully-coupled Euler-Lagrange (CEL) modeling capability with a new continuum-based visco-plastic fracture model for simulation of damage processes, called AZ_Frac. These capabilities produce high-fidelity models of various factors believed to be key in the generation of seismic waves: the explosion dynamics, a weak grout-filled borehole, the surrounding jointed rock, and damage creation and deformations happening around the source and the free surface. SPECFEM3D, based on the Spectral Element Method (SEM) is a direct numerical method for full wave modeling with mathematical accuracy. The coupling interface consists of a series of grid points of the SEM mesh situated inside of the hydrodynamic code's domain. Displacement time series at these points are computed using output data from CASH or Abaqus (by interpolation if needed) and fed into the time marching scheme of SPECFEM3D. We will present validation tests with the Sharpe's model and comparisons of waveforms modeled with Rg waves (2-8Hz) that were recorded up to 2 km for SPE. We especially show effects of the local topography, velocity structure and spallation. Our models predict smaller amplitudes of Rg waves for the first five SPE shots compared to pure elastic models such as Denny &Johnson (1991).

  11. Applications guide to the RSIC-distributed version of the MCNP code (coupled Monte Carlo neutron-photon Code)

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1985-09-01

    An overview of the RSIC-distributed version of the MCNP code (a soupled Monte Carlo neutron-photon code) is presented. All general features of the code, from machine hardware requirements to theoretical details, are discussed. The current nuclide cross-section and other libraries available in the standard code package are specified, and a realistic example of the flexible geometry input is given. Standard and nonstandard source, estimator, and variance-reduction procedures are outlined. Examples of correct usage and possible misuse of certain code features are presented graphically and in standard output listings. Finally, itemized summaries of sample problems, various MCNP code documentation, and future work are given

  12. Energy Conservation Tests of a Coupled Kinetic-kinetic Plasma-neutral Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    Stotler, D. P.; Chang, C. S.; Ku, S. H.; Lang, J.; Park, G.

    2012-08-29

    A Monte Carlo neutral transport routine, based on DEGAS2, has been coupled to the guiding center ion-electron-neutral neoclassical PIC code XGC0 to provide a realistic treatment of neutral atoms and molecules in the tokamak edge plasma. The DEGAS2 routine allows detailed atomic physics and plasma-material interaction processes to be incorporated into these simulations. The spatial pro le of the neutral particle source used in the DEGAS2 routine is determined from the uxes of XGC0 ions to the material surfaces. The kinetic-kinetic plasma-neutral transport capability is demonstrated with example pedestal fueling simulations.

  13. MAC/GMC Code Enhanced for Coupled Electromagnetothermoelastic Analysis of Smart Composites

    Science.gov (United States)

    Bednarcyk, Brett A.; Arnold, Steven M.; Aboudi, Jacob

    2002-01-01

    Intelligent materials are those that exhibit coupling between their electromagnetic response and their thermomechanical response. This coupling allows smart materials to react mechanically (e.g., an induced displacement) to applied electrical or magnetic fields (for instance). These materials find many important applications in sensors, actuators, and transducers. Recently interest has arisen in the development of smart composites that are formed via the combination of two or more phases, one or more of which is a smart material. To design with and utilize smart composites, designers need theories that predict the coupled smart behavior of these materials from the electromagnetothermoelastic properties of the individual phases. The micromechanics model known as the generalized method of cells (GMC) has recently been extended to provide this important capability. This coupled electromagnetothermoelastic theory has recently been incorporated within NASA Glenn Research Center's Micromechanics Analysis Code with Generalized Method of Cells (MAC/GMC). This software package is user friendly and has many additional features that render it useful as a design and analysis tool for composite materials in general, and with its new capabilities, for smart composites as well.

  14. TRACE/VALKIN: a neutronics-thermohydraulics coupled code to analyze strong 3D transients

    Energy Technology Data Exchange (ETDEWEB)

    Rafael Miro; Gumersindo Verdu; Ana Maria Sanchez [Chemical and Nuclear Engineering Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain); Damian Ginestar [Applied Mathematics Department. Polytechnic University of Valencia. Cami de Vera s/n. 46022 Valencia (Spain)

    2005-07-01

    Full text of publication follows: A nuclear reactor simulator consists mainly of two different blocks, which solve the models used for the basic physical phenomena taking place in the reactor. In this way, there is a neutronic module which simulates the neutron balance in the reactor core, and a thermal-hydraulics module, which simulates the heat transfer in the fuel, the heat transfer from the fuel to the water, and the different condensation and evaporation processes taking place in the reactor core and in the condenser systems. TRACE is a two-phase, two-fluid thermal-hydraulic reactor systems analysis code. The TRACE acronym stands for TRAC/RELAP Advanced Computational Engine, reflecting its ability to run both RELAP5 and TRAC legacy input models. It includes a three-dimensional kinetics module called PARCS for performing advanced analysis of coupled core thermal-hydraulic/kinetics problems. TRACE-VALKIN code is a new time domain analysis code to study transients in LWR reactors. This code uses the best estimate code TRACE to give account of the heat transfer and thermal-hydraulic processes, and a 3D neutronics module. This module has two options, the MODKIN option that makes use of a modal method based on the assumption that the neutronic flux can be approximately expanded in terms of the dominant lambda modes associated with a static configuration of the reactor core, and the NOKIN option that uses a one-step backward discretization of the neutron diffusion equation. The lambda modes are obtained using the Implicit Restarted Arnoldi approach or the Jacob-Davidson algorithm. To check the performance of the coupled code TRACE-VALKIN against complex 3D neutronic transients, using the cross-sections tables generated with the translator SIMTAB from SIMULATE to TRACE/VALKIN, the Cofrentes NPP SCRAM-61 transient is simulated. Cofrentes NPP is a General Electric BWR-6 design located in Valencia-land (Spain). It is in operation since 1985 and currently in its fifteenth

  15. A comparison of two fully coupled codes for integrated dynamic analysis of floating vertical axis wind turbines

    NARCIS (Netherlands)

    Koppenol, Boy; Cheng, Zhengshun; Gao, Zhen; Simao Ferreira, C.; Moan, T; Tande, John Olav Giæver; Kvamsdal, Trond; Muskulus, Michael

    2017-01-01

    This paper presents a comparison of two state-of-the-art codes that are capable of modelling floating vertical axis wind turbines (VAWTs) in fully coupled time-domain simulations, being the HAWC2 by DTU and the SIMO-RIFLEX-AC code by NTNU/MARINTEK. The comparative study focusses on the way

  16. The Use of Coupled Code Technique for Best Estimate Safety Analysis of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Bousbia Salah, A.; D'Auria, F.

    2006-01-01

    Issues connected with the thermal-hydraulics and neutronics of nuclear plants still challenge the design, safety and the operation of Light Water nuclear Reactors (LWR). The lack of full understanding of complex mechanisms related to the interaction between these issues imposed the adoption of conservative safety limits. Those safety margins put restrictions on the optimal exploitation of the plants and consequently reduced economic profit of the plant. In the light of the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. In fact, during the last decades Best Estimate (BE) neutronic and thermal-hydraulic calculations were so far carried out following rather parallel paths with only few interactions between them. Nowadays, it becomes possible to switch to new generation of computational tools, namely, Coupled Code technique. The application of such method is mandatory for the analysis of accident conditions where strong coupling between the core neutronics and the primary circuit thermal-hydraulics, and more especially when asymmetrical processes take place in the core leading to local space-dependent power generation. Through the current study, a demonstration of the maturity level achieved in the calculation of 3-D core performance during complex accident scenarios in NPPs is emphasized. Typical applications are outlined and discussed showing the main features and limitations of this technique. (author)

  17. Higher-order harmonics coupling in different free-electron laser codes

    Science.gov (United States)

    Giannessi, L.; Freund, H. P.; Musumeci, P.; Reiche, S.

    2008-08-01

    The capability for simulation of the dynamics of a free-electron laser including the higher-order harmonics in linear undulators exists in several existing codes as MEDUSA [H.P. Freund, S.G. Biedron, and S.V. Milton, IEEE J. Quantum Electron. 27 (2000) 243; H.P. Freund, Phys. Rev. ST-AB 8 (2005) 110701] and PERSEO [L. Giannessi, Overview of Perseo, a system for simulating FEL dynamics in Mathcad, , in: Proceedings of FEL 2006 Conference, BESSY, Berlin, Germany, 2006, p. 91], and has been recently implemented in GENESIS 1.3 [See ]. MEDUSA and GENESIS also include the dynamics of even harmonics induced by the coupling through the betatron motion. In addition MEDUSA, which is based on a non-wiggler averaged model, is capable of simulating the generation of even harmonics in the transversally cold beam regime, i.e. when the even harmonic coupling arises from non-linear effects associated with longitudinal particle dynamics and not to a finite beam emittance. In this paper a comparison between the predictions of the codes in different conditions is given.

  18. Geometrical modification transfer between specific meshes of each coupled physical codes. Application to the Jules Horowitz research reactor experimental devices

    International Nuclear Information System (INIS)

    Duplex, B.

    2011-01-01

    The CEA develops and uses scientific software, called physical codes, in various physical disciplines to optimize installation and experimentation costs. During a study, several physical phenomena interact, so a code coupling and some data exchanges between different physical codes are required. Each physical code computes on a particular geometry, usually represented by a mesh composed of thousands to millions of elements. This PhD Thesis focuses on the geometrical modification transfer between specific meshes of each coupled physical code. First, it presents a physical code coupling method where deformations are computed by one of these codes. Next, it discusses the establishment of a model, common to different physical codes, grouping all the shared data. Finally, it covers the deformation transfers between meshes of the same geometry or adjacent geometries. Geometrical modifications are discrete data because they are based on a mesh. In order to permit every code to access deformations and to transfer them, a continuous representation is computed. Two functions are developed, one with a global support, and the other with a local support. Both functions combine a simplification method and a radial basis function network. A whole use case is dedicated to the Jules Horowitz reactor. The effect of differential dilatations on experimental device cooling is studied. (author) [fr

  19. Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Yang, T.; Cheng, X.

    2014-01-01

    To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

  20. Simulating Coupling Complexity in Space Plasmas: First Results from a new code

    Science.gov (United States)

    Kryukov, I.; Zank, G. P.; Pogorelov, N. V.; Raeder, J.; Ciardo, G.; Florinski, V. A.; Heerikhuisen, J.; Li, G.; Petrini, F.; Shematovich, V. I.; Winske, D.; Shaikh, D.; Webb, G. M.; Yee, H. M.

    2005-12-01

    The development of codes that embrace 'coupling complexity' via the self-consistent incorporation of multiple physical scales and multiple physical processes in models has been identified by the NRC Decadal Survey in Solar and Space Physics as a crucial necessary development in simulation/modeling technology for the coming decade. The National Science Foundation, through its Information Technology Research (ITR) Program, is supporting our efforts to develop a new class of computational code for plasmas and neutral gases that integrates multiple scales and multiple physical processes and descriptions. We are developing a highly modular, parallelized, scalable code that incorporates multiple scales by synthesizing 3 simulation technologies: 1) Computational fluid dynamics (hydrodynamics or magneto-hydrodynamics-MHD) for the large-scale plasma; 2) direct Monte Carlo simulation of atoms/neutral gas, and 3) transport code solvers to model highly energetic particle distributions. We are constructing the code so that a fourth simulation technology, hybrid simulations for microscale structures and particle distributions, can be incorporated in future work, but for the present, this aspect will be addressed at a test-particle level. This synthesis we will provide a computational tool that will advance our understanding of the physics of neutral and charged gases enormously. Besides making major advances in basic plasma physics and neutral gas problems, this project will address 3 Grand Challenge space physics problems that reflect our research interests: 1) To develop a temporal global heliospheric model which includes the interaction of solar and interstellar plasma with neutral populations (hydrogen, helium, etc., and dust), test-particle kinetic pickup ion acceleration at the termination shock, anomalous cosmic ray production, interaction with galactic cosmic rays, while incorporating the time variability of the solar wind and the solar cycle. 2) To develop a coronal

  1. Syrthes thermal code and Estet or N3S fluid mechanics codes coupling; Couplage du code de thermique Syrthes et des codes de mecanique des fluides N3S et ou Estet

    Energy Technology Data Exchange (ETDEWEB)

    Peniguel, C [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I [SIMULOG, 78 - Guyancourt (France)

    1997-06-01

    EDF has developed numerical codes for modeling the conductive, radiative and convective thermal transfers and their couplings in complex industrial configurations: the convection in a fluid is solved by Estet in finite volumes or N3S in finite elements, the conduction is solved by Syrthes and the wall-to-wall thermal radiation is modelled by Syrthes with the help of a radiosity method. Syrthes controls the different heat exchanges which may occur between fluid and solid domains, using an explicit iterative method. An extension of Syrthes has been developed in order to allow the consideration of configurations where several fluid codes operate simultaneously, using ``message passing`` tools such as PVM (Parallel Virtual Machine) and the Calcium code coupler developed at EDF. Application examples are given

  2. Phase-amplitude coupling supports phase coding in human ECoG

    Science.gov (United States)

    Watrous, Andrew J; Deuker, Lorena; Fell, Juergen; Axmacher, Nikolai

    2015-01-01

    Prior studies have shown that high-frequency activity (HFA) is modulated by the phase of low-frequency activity. This phenomenon of phase-amplitude coupling (PAC) is often interpreted as reflecting phase coding of neural representations, although evidence for this link is still lacking in humans. Here, we show that PAC indeed supports phase-dependent stimulus representations for categories. Six patients with medication-resistant epilepsy viewed images of faces, tools, houses, and scenes during simultaneous acquisition of intracranial recordings. Analyzing 167 electrodes, we observed PAC at 43% of electrodes. Further inspection of PAC revealed that category specific HFA modulations occurred at different phases and frequencies of the underlying low-frequency rhythm, permitting decoding of categorical information using the phase at which HFA events occurred. These results provide evidence for categorical phase-coded neural representations and are the first to show that PAC coincides with phase-dependent coding in the human brain. DOI: http://dx.doi.org/10.7554/eLife.07886.001 PMID:26308582

  3. Cross-verification of the GENE and XGC codes in preparation for their coupling

    Science.gov (United States)

    Jenko, Frank; Merlo, Gabriele; Bhattacharjee, Amitava; Chang, Cs; Dominski, Julien; Ku, Seunghoe; Parker, Scott; Lanti, Emmanuel

    2017-10-01

    A high-fidelity Whole Device Model (WDM) of a magnetically confined plasma is a crucial tool for planning and optimizing the design of future fusion reactors, including ITER. Aiming at building such a tool, in the framework of the Exascale Computing Project (ECP) the two existing gyrokinetic codes GENE (Eulerian delta-f) and XGC (PIC full-f) will be coupled, thus enabling to carry out first principle kinetic WDM simulations. In preparation for this ultimate goal, a benchmark between the two codes is carried out looking at ITG modes in the adiabatic electron limit. This verification exercise is also joined by the global Lagrangian PIC code ORB5. Linear and nonlinear comparisons have been carried out, neglecting for simplicity collisions and sources. A very good agreement is recovered on frequency, growth rate and mode structure of linear modes. A similarly excellent agreement is also observed comparing the evolution of the heat flux and of the background temperature profile during nonlinear simulations. Work supported by the US DOE under the Exascale Computing Project (17-SC-20-SC).

  4. ARCADIAR - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

    International Nuclear Information System (INIS)

    Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas; Thareau, Sebastien

    2007-01-01

    Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code system ARCADIA R and concludes on customer benefits. ARCADIA R is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA R system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)

  5. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    Goethem, G. van; Lassmann, K.

    1989-01-01

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  6. Filtering authentic sepsis arising in the ICU using administrative codes coupled to a SIRS screening protocol.

    Science.gov (United States)

    Sudduth, Christopher L; Overton, Elizabeth C; Lyu, Peter F; Rimawi, Ramzy H; Buchman, Timothy G

    2017-06-01

    Using administrative codes and minimal physiologic and laboratory data, we sought a high-specificity identification strategy for patients whose sepsis initially appeared during their ICU stay. We studied all patients discharged from an academic hospital between September 1, 2013 and October 31, 2014. Administrative codes and minimal physiologic and laboratory criteria were used to identify patients at high risk of developing the onset of sepsis in the ICU. Two clinicians then independently reviewed the patient record to verify that the screened-in patients appeared to become septic during their ICU admission. Clinical chart review verified sepsis in 437/466 ICU stays (93.8%). Of these 437 encounters, only 151 (34.6%) were admitted to the ICU with neither SIRS nor evidence of infection and therefore appeared to become septic during their ICU stay. Selected administrative codes coupled to SIRS criteria and applied to patients admitted to ICU can yield up to 94% authentic sepsis patients. However, only 1/3 of patients thus identified appeared to become septic during their ICU stay. Studies that depend on high-intensity monitoring for description of the time course of sepsis require clinician review and verification that sepsis initially appeared during the monitoring period. Copyright © 2017 Elsevier Inc. All rights reserved.

  7. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

    2006-01-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  8. The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

    2006-07-01

    At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

  9. Simulation of the OECD Main-Steam-Line-Break Benchmark Exercise 3 Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, Christophe R.; Zhang Jinzhao

    2004-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package has been qualified and will be used at Tractebel Engineering (TE) for analyzing asymmetric pressurized water reactor (PWR) accidents with strong core-system interactions. The Organization for Economic Cooperation and Development/U.S. Nuclear Regulatory Commission PWR main-steam-line-break benchmark problem was analyzed as part of the qualification efforts to demonstrate the capability of the coupled code package of simulating such transients. This paper reports the main results of TE's contribution to the benchmark Exercise 3

  10. The CNCSN: one, two- and three-dimensional coupled neutral and charged particle discrete ordinates code package

    International Nuclear Information System (INIS)

    Voloschenko, A.M.; Gukov, S.V.; Kryuchkov, V.P.; Dubinin, A.A.; Sumaneev, O.V.

    2005-01-01

    The CNCSN package is composed of the following codes: -) KATRIN-2.0: a three-dimensional neutral and charged particle transport code; -) KASKAD-S-2.5: a two-dimensional neutral and charged particle transport code; -) ROZ-6.6: a one-dimensional neutral and charged particle transport code; -) ARVES-2.5: a preprocessor for the working macroscopic cross-section format FMAC-M for transport calculations; -) MIXERM: a utility code for preparing mixtures on the base of multigroup cross-section libraries in ANISN format; -) CEPXS-BFP: a version of the Sandia National Lab. multigroup coupled electron-photon cross-section generating code CEPXS, adapted for solving the charged particles transport in the Boltzmann-Fokker-Planck formulation with the use of discrete ordinate method; -) SADCO-2.4: Institute for High-Energy Physics modular system for generating coupled nuclear data libraries to provide high-energy particles transport calculations by multigroup method; -) KATRIF: the post-processor for the KATRIN code; -) KASF: the post-processor for the KASKAD-S code; and ROZ6F: the post-processor for the ROZ-6 code. The coding language is Fortran-90

  11. TART 2000: A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    International Nuclear Information System (INIS)

    Cullen, D.E

    2000-01-01

    TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files

  12. TART 2000 A Coupled Neutron-Photon, 3-D, Combinatorial Geometry, Time Dependent, Monte Carlo Transport Code

    CERN Document Server

    Cullen, D

    2000-01-01

    TART2000 is a coupled neutron-photon, 3 Dimensional, combinatorial geometry, time dependent Monte Carlo radiation transport code. This code can run on any modern computer. It is a complete system to assist you with input Preparation, running Monte Carlo calculations, and analysis of output results. TART2000 is also incredibly FAST; if you have used similar codes, you will be amazed at how fast this code is compared to other similar codes. Use of the entire system can save you a great deal of time and energy. TART2000 is distributed on CD. This CD contains on-line documentation for all codes included in the system, the codes configured to run on a variety of computers, and many example problems that you can use to familiarize yourself with the system. TART2000 completely supersedes all older versions of TART, and it is strongly recommended that users only use the most recent version of TART2000 and its data files.

  13. Transient cases analyses of the TRIGA IPR-R1 using thermal hydraulic and neutron kinetic coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Scari, Maria E., E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: melizabethscari@yahoo.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores (INCT/CNPq), Belo Horizonte (Brazil); Miro, Rafael; Verdu, Gumersindo, E-mail: rmiro@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Universidad Politecnica de Valencia (Spain). Departamento de Ingenieria Quimica y Nuclear

    2015-07-01

    Simulations and analyses of nuclear reactors have been improved by utilization of coupled thermal-hydraulic (TH) and neutron kinetics (NK) system codes especially to simulate transients that involve strong feedback effects between NK and TH. The TH-NK coupling technique was initially developed and used to simulate the behavior of power reactors; however, several coupling methodologies are now being applied for research reactors. This work presents the coupling methodology application between RELAP5 and PARCS codes using as a model the TRIGA IPR-R1 research reactor. Analyses of steady state and transient conditions and comparisons with results from simulations using only the RELAP5 code are being presented in this paper. (author)

  14. Trading speed and accuracy by coding time: a coupled-circuit cortical model.

    Directory of Open Access Journals (Sweden)

    Dominic Standage

    2013-04-01

    Full Text Available Our actions take place in space and time, but despite the role of time in decision theory and the growing acknowledgement that the encoding of time is crucial to behaviour, few studies have considered the interactions between neural codes for objects in space and for elapsed time during perceptual decisions. The speed-accuracy trade-off (SAT provides a window into spatiotemporal interactions. Our hypothesis is that temporal coding determines the rate at which spatial evidence is integrated, controlling the SAT by gain modulation. Here, we propose that local cortical circuits are inherently suited to the relevant spatial and temporal coding. In simulations of an interval estimation task, we use a generic local-circuit model to encode time by 'climbing' activity, seen in cortex during tasks with a timing requirement. The model is a network of simulated pyramidal cells and inhibitory interneurons, connected by conductance synapses. A simple learning rule enables the network to quickly produce new interval estimates, which show signature characteristics of estimates by experimental subjects. Analysis of network dynamics formally characterizes this generic, local-circuit timing mechanism. In simulations of a perceptual decision task, we couple two such networks. Network function is determined only by spatial selectivity and NMDA receptor conductance strength; all other parameters are identical. To trade speed and accuracy, the timing network simply learns longer or shorter intervals, driving the rate of downstream decision processing by spatially non-selective input, an established form of gain modulation. Like the timing network's interval estimates, decision times show signature characteristics of those by experimental subjects. Overall, we propose, demonstrate and analyse a generic mechanism for timing, a generic mechanism for modulation of decision processing by temporal codes, and we make predictions for experimental verification.

  15. Development a computer codes to couple PWR-GALE output and PC-CREAM input

    Science.gov (United States)

    Kuntjoro, S.; Budi Setiawan, M.; Nursinta Adi, W.; Deswandri; Sunaryo, G. R.

    2018-02-01

    Radionuclide dispersion analysis is part of an important reactor safety analysis. From the analysis it can be obtained the amount of doses received by radiation workers and communities around nuclear reactor. The radionuclide dispersion analysis under normal operating conditions is carried out using the PC-CREAM code, and it requires input data such as source term and population distribution. Input data is derived from the output of another program that is PWR-GALE and written Population Distribution data in certain format. Compiling inputs for PC-CREAM programs manually requires high accuracy, as it involves large amounts of data in certain formats and often errors in compiling inputs manually. To minimize errors in input generation, than it is make coupling program for PWR-GALE and PC-CREAM programs and a program for writing population distribution according to the PC-CREAM input format. This work was conducted to create the coupling programming between PWR-GALE output and PC-CREAM input and programming to written population data in the required formats. Programming is done by using Python programming language which has advantages of multiplatform, object-oriented and interactive. The result of this work is software for coupling data of source term and written population distribution data. So that input to PC-CREAM program can be done easily and avoid formatting errors. Programming sourceterm coupling program PWR-GALE and PC-CREAM is completed, so that the creation of PC-CREAM inputs in souceterm and distribution data can be done easily and according to the desired format.

  16. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    International Nuclear Information System (INIS)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-01

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too

  17. Coupling calculation of CFD-ACE computational fluid dynamics code and DeCART whole-core neutron transport code for development of numerical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Hwan; Seo, Kyong Won; Chun, Tae Hyun; Kim, Kang Seog

    2005-03-15

    Code coupling activities have so far focused on coupling the neutronics modules with the CFD module. An interface module for the CFD-ACE/DeCART coupling was established as an alternative to the original STAR-CD/DeCART interface. The interface module for DeCART/CFD-ACE was validated by single-pin model. The optimized CFD mesh was decided through the calculation of multi-pin model. It was important to consider turbulent mixing of subchannels for calculation of fuel temperature. For the parallel calculation, the optimized decompose process was necessary to reduce the calculation costs and setting of the iteration and convergence criterion for each code was important, too.

  18. Neutronics/Thermo-fluid Coupled Analysis of PMR-200 Equilibrium Cycle by CAPP/GAMMA+ Code System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Tak, Nam-il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The equilibrium core was obtained by performing CAPP stand-alone multi-cycle depletion calculation with critical rod position search. In this work, a code system for coupled neutronics and thermo-fluids simulation was developed using CAPP and GAMMA+ codes. A server program, INTCA, controls the two codes for coupled calculations and performs the mapping between the variables of the two codes based on the nodalization of the two codes. In order to extend the knowledge about the coupled behavior of a prismatic VHTR, the CAPP/GAMMA+ code system was applied to steady state performance analysis of PMR-200. The coupled calculation was carried out for the equilibrium core of PMR-200 from BOC to EOC. The peak fuel temperature was predicted to be 1372 .deg. C near MOC. However, the cycle-average fuel temperature was calculated as 1230 .deg. C, which is slightly below the design target of 1250 .deg. C. In addition, significant impact of the bypass flow on the central reflector temperature was found. Without bypass flow, the temperature of the active core region was slightly decreased while the temperature of the central and side reflector region was increased much. The both changes in the temperature increase the multiplication factor and the total change of the multiplication factor was more than 300 pcm. On the other hand, the effect of the bypass flow on the power density profile was not significant.

  19. The TALL-3D facility design and commissioning tests for validation of coupled STH and CFD codes

    Energy Technology Data Exchange (ETDEWEB)

    Grishchenko, Dmitry, E-mail: dmitry@safety.sci.kth.se; Jeltsov, Marti, E-mail: marti@safety.sci.kth.se; Kööp, Kaspar, E-mail: kaspar@safety.sci.kth.se; Karbojian, Aram, E-mail: karbojan@kth.se; Villanueva, Walter, E-mail: walter@safety.sci.kth.se; Kudinov, Pavel, E-mail: pavel@safety.sci.kth.se

    2015-08-15

    Highlights: • Design of a heavy liquid thermal-hydraulic loop for CFD/STH code validation. • Description of the loop instrumentation and assessment of measurement error. • Experimental data from forced to natural circulation transient. - Abstract: Application of coupled CFD (Computational Fluid Dynamics) and STH (System Thermal Hydraulics) codes is a prerequisite for computationally affordable and sufficiently accurate prediction of thermal-hydraulics of complex systems. Coupled STH and CFD codes require validation for understanding and quantification of the sources of uncertainties in the code prediction. TALL-3D is a liquid Lead Bismuth Eutectic (LBE) loop developed according to the requirements for the experimental data for validation of coupled STH and CFD codes. The goals of the facility design are to provide (i) mutual feedback between natural circulation in the loop and complex 3D mixing and stratification phenomena in the pool-type test section, (ii) a possibility to validate standalone STH and CFD codes for each subsection of the facility, and (iii) sufficient number of experimental data to separate the process of input model calibration and code validation. Description of the facility design and its main components, approach to estimation of experimental uncertainty and calibration of model input parameters that are not directly measured in the experiment are discussed in the paper. First experimental data from the forced to natural circulation transient is also provided in the paper.

  20. Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

    International Nuclear Information System (INIS)

    Ramos, Enrique; Roman, Jose E.; Abarca, Agustín; Miró, Rafael; Bermejo, Juan A.

    2016-01-01

    Highlights: • An MPI parallel version of the thermal–hydraulic subchannel code COBRA-TF has been developed. • The parallel code has been coupled to the 3D neutron diffusion code PARCSv2.7. • The new codes are validated with a control rod drop transient. - Abstract: In order to reduce the response time when simulating large reactors in detail, a parallel version of the thermal–hydraulic subchannel code COBRA-TF (CTF) has been developed using the standard Message Passing Interface (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian matrix is parallelized, in such a way that each processor is in charge of generating the data associated with a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit. With the goal of creating a powerful tool to simulate the reactor core behavior during asymmetrical transients, the 3D neutron diffusion code PARCSv2.7 (PARCS) has been coupled with the parallel version of CTF (pCTF) using the Parallel Virtual Machine (PVM) technology. In order to validate the correctness of the parallel coupled code, a control rod drop transient has been simulated comparing the results with the real experimental measures acquired during an NPP real test.

  1. Calculations of the giant-dipole-resonance photoneutrons using a coupled EGS4-morse code

    International Nuclear Information System (INIS)

    Liu, J.C.; Nelson, W.R.; Kase, K.R.; Mao, X.S.

    1995-10-01

    The production and transport of the photoneutrons from the giant-dipoleresonance reaction have been implemented in a coupled EGS4-MORSE code. The total neutron yield (including both the direct neutron and evaporation neutron components) is calculated by folding the photoneutron yield cross sections with the photon track length distribution in the target. Empirical algorithms based on the measurements have been developed to estimate the fraction and energy of the direct neutron component for each photon. The statistical theory in the EVAP4 code, incorporated as a MORSE subroutine, is used to determine the energies of the evaporation neutrons. These represent major improvements over other calculations that assumed no direct neutrons, a constant fraction of direct neutrons, monoenergetic direct neutron, or a constant nuclear temperature for the evaporation neutrons. It was also assumed that the slow neutrons ( 2 θ, which have a peak emission at 900. Comparisons between the calculated and the measured photoneutron results (spectra of the direct, evaporation and total neutrons; nuclear temperatures; direct neutron fractions) for materials of lead, tungsten, tantalum and copper have been made. The results show that the empirical algorithms, albeit simple, can produce reasonable results over the interested photon energy range

  2. Code Coupling via Jacobian-Free Newton-Krylov Algorithms with Application to Magnetized Fluid Plasma and Kinetic Neutral Models

    Energy Technology Data Exchange (ETDEWEB)

    Joseph, Ilon [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2014-05-27

    Jacobian-free Newton-Krylov (JFNK) algorithms are a potentially powerful class of methods for solving the problem of coupling codes that address dfferent physics models. As communication capability between individual submodules varies, different choices of coupling algorithms are required. The more communication that is available, the more possible it becomes to exploit the simple sparsity pattern of the Jacobian, albeit of a large system. The less communication that is available, the more dense the Jacobian matrices become and new types of preconditioners must be sought to efficiently take large time steps. In general, methods that use constrained or reduced subsystems can offer a compromise in complexity. The specific problem of coupling a fluid plasma code to a kinetic neutrals code is discussed as an example.

  3. Coupling of the computational fluid dynamics code ANSYS CFX with the 3D neutron kinetic core model DYN3D

    International Nuclear Information System (INIS)

    Kliem, S.; Grahn, A.; Rohde, U.; Schuetze, J.; Frank, Th.

    2010-01-01

    The computational fluid dynamics code ANSYS CFX has been coupled with the neutron-kinetic core model DYN3D. ANSYS CFX calculates the fluid dynamics and related transport phenomena in the reactors coolant and provides the corresponding data to DYN3D. In the fluid flow simulation of the coolant, the core itself is modeled within the porous body approach. 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 prototype that is currently available, the coupling is restricted to single-phase flow problems. In the time domain an explicit coupling of the codes has been implemented so far. Steady-state and transient verification calculations for two small-size test problems confirm the correctness of the implementation of the prototype coupling. The first test problem was a mini-core consisting of nine real-size fuel assemblies with quadratic cross section. Comparison was performed with the DYN3D stand-alone code. In the steady state, the effective multiplication factor obtained by the DYN3D/ANSYS CFX codes hows a deviation of 9.8 pcm from the DYN3D stand-alone solution. This difference can be attributed to the use of different water property packages in the two codes. The transient test case simulated the withdrawal of the control rod from the central fuel assembly at hot zero power in the same mini-core. Power increase during the introduction of positive reactivity and power reduction due to fuel temperature increase are calculated in the same manner by the coupled and the stand-alone codes. The maximum values reached during the power rise differ by about 1 MW at a power level of 50 MW. Beside the different water property packages, these differences are caused by the use of different flow solvers. The same calculations were carried for a mini-core with seven real-size fuel assemblies with hexagonal cross section in

  4. Feasibility Study of Coupling the CASMO-4/TABLES-3/SIMULATE-3 Code System to TRACE/PARCS

    International Nuclear Information System (INIS)

    Demaziere, Christophe; Staalek, Mathias

    2004-12-01

    This report investigates the feasibility of coupling the Studsvik Scandpower CASMO-4/TABLES-3/SIMULATE-3 codes to the US NRC TRACE/PARCS codes. The data required by TRACE/PARCS are actually the ones necessary to run its neutronic module PARCS. Such data are the macroscopic nuclear cross-sections, some microscopic nuclear cross-sections important for the Xenon and Samarium poisoning effects, the Assembly Discontinuity Factors, and the kinetic parameters. All these data can be retrieved from the Studsvik Scandpower codes. The data functionalization is explained in detail for both systems of codes and the possibility of coupling each of these codes to TRACE/PARCS is discussed. Due to confidentiality restrictions in the use of the CASMO-4 files and to an improper format of the TABLES-3 output files, it is demonstrated that TRACE/PARCS can only be coupled to SIMULATE-3. Specifically-dedicated SIMULATE-3 input decks allow easily editing the neutronic data at specific operating statepoints. Although the data functionalization is different between both systems of codes, such a procedure permits reconstructing a set of data directly compatible with PARCS

  5. HYDROCOIN [HYDROlogic COde INtercomparison] Level 1: Benchmarking and verification test results with CFEST [Coupled Fluid, Energy, and Solute Transport] code: Draft report

    International Nuclear Information System (INIS)

    Yabusaki, S.; Cole, C.; Monti, A.M.; Gupta, S.K.

    1987-04-01

    Part of the safety analysis is evaluating groundwater flow through the repository and the host rock to the accessible environment by developing mathematical or analytical models and numerical computer codes describing the flow mechanisms. This need led to the establishment of an international project called HYDROCOIN (HYDROlogic COde INtercomparison) organized by the Swedish Nuclear Power Inspectorate, a forum for discussing techniques and strategies in subsurface hydrologic modeling. The major objective of the present effort, HYDROCOIN Level 1, is determining the numerical accuracy of the computer codes. The definition of each case includes the input parameters, the governing equations, the output specifications, and the format. The Coupled Fluid, Energy, and Solute Transport (CFEST) code was applied to solve cases 1, 2, 4, 5, and 7; the Finite Element Three-Dimensional Groundwater (FE3DGW) Flow Model was used to solve case 6. Case 3 has been ignored because unsaturated flow is not pertinent to SRP. This report presents the Level 1 results furnished by the project teams. The numerical accuracy of the codes is determined by (1) comparing the computational results with analytical solutions for cases that have analytical solutions (namely cases 1 and 4), and (2) intercomparing results from codes for cases which do not have analytical solutions (cases 2, 5, 6, and 7). Cases 1, 2, 6, and 7 relate to flow analyses, whereas cases 4 and 5 require nonlinear solutions. 7 refs., 71 figs., 9 tabs

  6. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    Energy Technology Data Exchange (ETDEWEB)

    Marguet, S.D. [Electricite de France (EDF), 92 - Clamart (France)

    1997-12-31

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF`s neutronic code COCCINELLE uses the Rowland`s formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission`s products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on `low` configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.) 7 refs.

  7. A new coupling of the 3D thermal-hydraulic code THYC and the thermo-mechanical code CYRANO3 for PWR calculations

    International Nuclear Information System (INIS)

    Marguet, S.D.

    1997-01-01

    Among all parameters, the fuel temperature has a significant influence on the reactivity of the core, because of the Doppler effect on cross-sections. Most neutronic codes use a straightforward method to calculate an average fuel temperature used in their specific feed-back models. For instance, EDF's neutronic code COCCINELLE uses the Rowland's formula using the temperatures of the center and the surface of the pellet. COCCINELLE is coupled to the 3D thermal-hydraulic code THYC with calculates TDoppler with is standard thermal model. In order to improve the accuracy of such calculations, we have developed the coupling of our two latest codes in thermal-hydraulics (THYC) and thermo-mechanics (CYRANO3). THYC calculates two-phase flows in pipes or rod bundles and is used for transient calculations such as steam-line break, boron dilution accidents, DNB predictions, steam generator and condenser studies. CYRANO3 calculates most of the phenomena that take place in the fuel such as: 1) heat transfer induced by nuclear power; 2) thermal expansion of the fuel and the cladding; 3) release of gaseous fission's products; 4) mechanical interaction between the pellet and the cladding. These two codes are now qualified in their own field and the coupling, using Parallel Virtual Machine (PVM) libraries customized in an home-made-easy-to-use package called CALCIUM, has been validated on 'low' configurations (no thermal expansion, constant thermal characteristics) and used on accidental transients such as rod ejection and loss of coolant accident. (K.A.)

  8. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    International Nuclear Information System (INIS)

    Peng Muzhang; Zhang Quan; Wang Guoli; Zhang Yuman

    1988-01-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory

  9. TISKTH-3: a couple neutronics/thermal-hydraulics code for the transient analysis of light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Muzhang, Peng; Quan, Zhang; Guoli, Wang; Yuman, Zhang

    1988-03-01

    TISKTH-3 is a coupled neutronics/thermal-hydraulics code for the transient analysis. A 3-dimensional neutron kinetics equation solved by the Nodal Green's Function Method is used for the neutronics model of the code. A homogeneous equilibrium model with a complete boiling curve and two numerical solutions of the implicit and explicit scheme is used for the thermal-hydraulics model of the code. A 2-dimensional heat conduction equation with variable conductivity solved by the method of weighted residuals is used for the fuel rod heat transfer model of the code. TISKTH-3 is able to analyze the fast transient process and complicate accident situations in the core. The initative applications have shown that the stability and convergency in the calculations with the code are satisfactory.

  10. Coupling of the SYRTHES thermal code with the ESTET or N3S fluid mechanics codes; Couplage du code de thermique SYRTHES et des codes de mecanique des fluides ESTET ou N3S

    Energy Technology Data Exchange (ETDEWEB)

    Peniguel, C [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I [Simulog, 78 (France)

    1998-12-31

    Thermal aspects take place in several industrial applications in which Electricite de France (EdF) is concerned. In most cases, several physical phenomena like conduction, radiation and convection are involved in thermal transfers. The aim of this paper is to present a numerical tool adapted to industrial configurations and which uses the coupling between fluid convection (resolved with ESTET in finite-volumes or with N3S in finite-elements) and radiant heat transfers between walls (resolved with SYRTHES using a radiosity method). SYRTHES manages the different thermal exchanges that can occur between fluid and solid domains thanks to an explicit iterative method. An extension of SYRTHES has been developed which allows to take into account simultaneously several fluid codes using `message passing` computer tools like Parallel Virtual Machine (PVM) and the code coupling software CALCIUM developed by the Direction of Studies and Researches (DER) of EdF. Various examples illustrate the interest of such a numerical tool. (J.S.) 12 refs.

  11. Coupling of the SYRTHES thermal code with the ESTET or N3S fluid mechanics codes; Couplage du code de thermique SYRTHES et des codes de mecanique des fluides ESTET ou N3S

    Energy Technology Data Exchange (ETDEWEB)

    Peniguel, C. [Electricite de France (EDF), 78 - Chatou (France). Direction des Etudes et Recherches; Rupp, I. [Simulog, 78 (France)

    1997-12-31

    Thermal aspects take place in several industrial applications in which Electricite de France (EdF) is concerned. In most cases, several physical phenomena like conduction, radiation and convection are involved in thermal transfers. The aim of this paper is to present a numerical tool adapted to industrial configurations and which uses the coupling between fluid convection (resolved with ESTET in finite-volumes or with N3S in finite-elements) and radiant heat transfers between walls (resolved with SYRTHES using a radiosity method). SYRTHES manages the different thermal exchanges that can occur between fluid and solid domains thanks to an explicit iterative method. An extension of SYRTHES has been developed which allows to take into account simultaneously several fluid codes using `message passing` computer tools like Parallel Virtual Machine (PVM) and the code coupling software CALCIUM developed by the Direction of Studies and Researches (DER) of EdF. Various examples illustrate the interest of such a numerical tool. (J.S.) 12 refs.

  12. Lessons learned in the verification, validation and application of a coupled heat and fluid flow code

    International Nuclear Information System (INIS)

    Tsang, C.F.

    1986-01-01

    A summary is given of the authors recent studies in the verification, validation and application of a coupled heat and fluid flow code. Verification has been done against eight analytic and semi-analytic solutions. These solutions include those involving thermal buoyancy flow and fracture flow. Comprehensive field validation studies over a period of four years are discussed. The studies are divided into three stages: (1) history matching, (2) double-blind prediction and confirmation, (3) design optimization. At each stage, parameter sensitivity studies are performed. To study the applications of mathematical models, a problem proposed by the International Energy Agency (IEA) is solved using this verified and validated numerical model as well as two simpler models. One of the simpler models is a semi-analytic method assuming the uncoupling of the heat and fluid flow processes. The other is a graphical method based on a large number of approximations. Variations are added to the basic IEA problem to point out the limits of ranges of applications of each model. A number of lessons are learned from the above investigations. These are listed and discussed

  13. Coupling External Radiation Transport Code Results to the GADRAS Detector Response Function

    International Nuclear Information System (INIS)

    Mitchell, Dean J.; Thoreson, Gregory G.; Horne, Steven M.

    2014-01-01

    Simulating gamma spectra is useful for analyzing special nuclear materials. Gamma spectra are influenced not only by the source and the detector, but also by the external, and potentially complex, scattering environment. The scattering environment can make accurate representations of gamma spectra difficult to obtain. By coupling the Monte Carlo Nuclear Particle (MCNP) code with the Gamma Detector Response and Analysis Software (GADRAS) detector response function, gamma spectrum simulations can be computed with a high degree of fidelity even in the presence of a complex scattering environment. Traditionally, GADRAS represents the external scattering environment with empirically derived scattering parameters. By modeling the external scattering environment in MCNP and using the results as input for the GADRAS detector response function, gamma spectra can be obtained with a high degree of fidelity. This method was verified with experimental data obtained in an environment with a significant amount of scattering material. The experiment used both gamma-emitting sources and moderated and bare neutron-emitting sources. The sources were modeled using GADRAS and MCNP in the presence of the external scattering environment, producing accurate representations of the experimental data.

  14. Coupling the MCNP Monte Carlo code and the FISPACT activation code with automatic visualization of the results of simulations

    International Nuclear Information System (INIS)

    Bourauel, Peter; Nabbi, Rahim; Biel, Wolfgang; Forrest, Robin

    2009-01-01

    The MCNP 3D Monte Carlo computer code is used not only for criticality calculations of nuclear systems but also to simulate transports of radiation and particles. The findings so obtained about neutron flux distribution and the associated spectra allow information about materials activation, nuclear heating, and radiation damage to be obtained by means of activation codes such as FISPACT. The stochastic character of particle and radiation transport processes normally links findings to the materials cells making up the geometry model of MCNP. Where high spatial resolution is required for the activation calculations with FISPACT, fine segmentation of the MCNP geometry becomes compulsory, which implies considerable expense for the modeling process. For this reason, an alternative simulation technique has been developed in an effort to automate and optimize data transfer between MCNP and FISPACT. (orig.)

  15. Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Kliem, S

    1998-10-01

    Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

  16. Calculation of an accident with delayed scram at NPP Greifswald using the coupled code DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Kliem, S.

    1998-01-01

    Complex computer codes modeling the whole reactor system including 3D neutron kinetics in combination with advanced thermohydraulic plant models become more and more important for the safety assessment of nuclear reactors. Transients or experiments with both neutron kinetic and thermalhydraulic data are needed for the validation of such coupled codes like DYN3D/ATHLET. First of all measured results from nuclear power plant (NPP) transients should be used, because the experimental thermalhydraulic facilities do not offer the possibility to model space-dependent neutron kinetic effects and research reactors with reliably measured 3D neutron kinetic data do not allow to study thermalhydraulic feedback effects. In this paper, an accident with delayed scram which occurred in 1989 at the NPP Greifswald is analyzed. Calculations of this accident were carried out with the goal to validate the coupled code DYN3D/ATHLET. (orig.)

  17. Development of a coupled dynamics code with transport theory capability and application to accelerator driven systems transients

    International Nuclear Information System (INIS)

    Cahalan, J.E.; Ama, T.; Palmiotti, G.; Taiwo, T.A.; Yang, W.S.

    2000-01-01

    The VARIANT-K and DIF3D-K nodal spatial kinetics computer codes have been coupled to the SAS4A and SASSYS-1 liquid metal reactor accident and systems analysis codes. SAS4A and SASSYS-1 have been extended with the addition of heavy liquid metal (Pb and Pb-Bi) thermophysical properties, heat transfer correlations, and fluid dynamics correlations. The coupling methodology and heavy liquid metal modeling additions are described. The new computer code suite has been applied to analysis of neutron source and thermal-hydraulics transients in a model of an accelerator-driven minor actinide burner design proposed in an OECD/NEA/NSC benchmark specification. Modeling assumptions and input data generation procedures are described. Results of transient analyses are reported, with emphasis on comparison of P1 and P3 variational nodal transport theory results with nodal diffusion theory results, and on significance of spatial kinetics effects

  18. Development of finite element code for the analysis of coupled thermo-hydro-mechanical behaviors of saturated-unsaturated medium

    International Nuclear Information System (INIS)

    Ohnishi, Y.; Shibata, H.; Kobayashi, A.

    1985-01-01

    A model is presented which describes fully coupled thermo-hydro-mechanical behavior of porous geologic medium. The mathematical formulation for the model utilizes the Biot theory for the consolidation and the energy balance equation. The medium is in the condition of saturated-unsaturated flow, then the free surfaces are taken into consideration in the model. The model, incorporated in a finite element numerical procedure, was implemented in a two-dimensional computer code. The code was developed under the assumptions that the medium is poro-elastic and in plane strain condition; water in the ground does not change its phase; heat is transferred by conductive and convective flow. Analytical solutions pertaining to consolidation theory for soils and rocks, thermoelasticity for solids and hydrothermal convection theory provided verification of stress and fluid flow couplings, respectively in the coupled model. Several types of problems are analyzed. The one is a study of some of the effects of completely coupled thermo-hydro-mechanical behavior on the response of a saturated-unsaturated porous rock containing a buried heat source. Excavation of an underground opening which has radioactive wastes at elevated temperatures is modeled and analyzed. The results shows that the coupling phenomena can be estimated at some degree by the numerical procedure. The computer code has a powerful ability to analyze of the repository the complex nature of the repository

  19. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

  20. ARTEMIS: The core simulator of AREVA NP's next generation coupled neutronics/thermal-hydraulics code system ARCADIAR

    International Nuclear Information System (INIS)

    Hobson, Greg; Merk, Stephan; Bolloni, Hans-Wilhelm; Breith, Karl-Albert; Curca-Tivig, Florin; Van Geemert, Rene; Heinecke, Jochen; Hartmann, Bettina; Porsch, Dieter; Tiles, Viatcheslav; Dall'Osso, Aldo; Pothet, Baptiste

    2008-01-01

    AREVA NP has developed a next-generation coupled neutronics/thermal-hydraulics code system, ARCADIA R , to fulfil customer's current demands and even anticipate their future demands in terms of accuracy and performance. The new code system will be implemented world-wide and will replace several code systems currently used in various global regions. An extensive phase of verification and validation of the new code system is currently in progress. One of the principal components of this new system is the core simulator, ARTEMIS. Besides the stand-alone tests on the individual computational modules, integrated tests on the overall code are being performed in order to check for non-regression as well as for verification of the code. Several benchmark problems have been successfully calculated. Full-core depletion cycles of different plant types from AREVA's French, American and German regions (e.g. N4 and KONVOI types) have been performed with ARTEMIS (using APOLLO2-A cross sections) and compared directly with current production codes, e.g. with SCIENCE and CASCADE-3D, and additionally with measurements. (authors)

  1. Development of a finite element code to solve thermo-hydro-mechanical coupling and simulate induced seismicity.

    Science.gov (United States)

    María Gómez Castro, Berta; De Simone, Silvia; Rossi, Riccardo; Larese De Tetto, Antonia; Carrera Ramírez, Jesús

    2015-04-01

    Coupled thermo-hydro-mechanical modeling is essential for CO2 storage because of (1) large amounts of CO2 will be injected, which will cause large pressure buildups and might compromise the mechanical stability of the caprock seal, (2) the most efficient technique to inject CO2 is the cold injection, which induces thermal stress changes in the reservoir and seal. These stress variations can cause mechanical failure in the caprock and can also trigger induced earthquakes. To properly assess these effects, numerical models that take into account the short and long-term thermo-hydro-mechanical coupling are an important tool. For this purpose, there is a growing need of codes that couple these processes efficiently and accurately. This work involves the development of an open-source, finite element code written in C ++ for correctly modeling the effects of thermo-hydro-mechanical coupling in the field of CO2 storage and in others fields related to these processes (geothermal energy systems, fracking, nuclear waste disposal, etc.), and capable to simulate induced seismicity. In order to be able to simulate earthquakes, a new lower dimensional interface element will be implemented in the code to represent preexisting fractures, where pressure continuity will be imposed across the fractures.

  2. Development and verification of coupled fluid-structural dynamic codes for stress analysis of reactor vessel internals under blowdown loading

    International Nuclear Information System (INIS)

    Krieg, R.; Schlechtendahl, E.G.

    1977-01-01

    YAQUIR has been applied to large PWR blowdown problems and compared with LECK results. The structural model of CYLDY2 and the fluid model of YAQUIR have been coupled in the code STRUYA. First tests with the fluid dynamic systems code FLUST have been successful. The incompressible fluid version of the 3D coupled code FLUX for HDR-geometry was checked against some analytical test cases and was used for evaluation of the eigenfrequencies of the coupled system. Several test cases were run with the two phase flow code SOLA-DF with satisfactory results. Remarkable agreement was found between YAQUIR results and experimental data obtained from shallow water analogy experiments. A test for investigation of nonequilibrium twophase flow dynamics has been specified in some detail. The test is to be performed early 1978 in the water loop of the IRB. Good agreement was found between the natural frequency predictions for the core barrel obtained from CYLDY2 and STRUDL/DYNAL. Work started on improvement of the beam mode treatment in CYLDY2. The name of this modified version will be CYLDY3. The fluiddynamic code SING1, based on an advanced singularity method and applicable to a broad class of highly transient, incompressible 3D-problems with negligible viscosity has been developed and tested. It will be used in connection with the planned laboratory experiments in order to investigate the effect of the core structure on the blowdown process. Coupling of SING1 with structural dynamics is on the way. (orig./RW) [de

  3. Coupling of electromagnetic and thermal codes. Induction heating; Couplage des codes electromagnetique et thermique. Le chauffage par induction

    Energy Technology Data Exchange (ETDEWEB)

    Colombani, M. [CEDRAT, (France)

    1997-12-31

    The development and adjustment of induction heating systems is quite delicate because two different subjects of physics are involved: magnetism (Foucault currents) and thermal engineering. Moreover, the magnetic and electrical properties depends on the temperature and the dissipated power depends on the magnetic and electrical properties and on the electrical excitation sources (geometry, intensity, frequency). The CEDRAT company has been involved since several years in the development of modeling softwares which allow to analyze these kind of problems. The most used is the FLUX2D software, developed by CEDRAT RECHERCHE in collaboration with the LEG (CNRS-INPG) and EdF, and which is used in several domains of applications (electric motors, actuators, high-voltage devices, magnetic recording, induction heating etc..). This software is based on a finite-element calculation method and, in the case of induction heating, it can perform different types of modeling: magnetic, thermal, temperature-dependant properties, weak and strong coupling, coupling with the electric circuit equations etc.. (J.S.)

  4. A Slater parameter optimisation interface for the CIV3 atomic structure code and its possible use with the R-matrix close coupling collision code

    International Nuclear Information System (INIS)

    Fawcett, B.C.; Hibbert, A.

    1989-11-01

    Details are here provided of amendments to the atomic structure code CIV3 which allow the optional adjustment of Slater parameters and average energies of configurations so that they result in improved energy levels and eigenvectors. It is also indicated how, in principle, the resultant improved eigenvectors can be utilised by the R-matrix collision code, thus providing an optimised target for close coupling collision strength calculations. An analogous computational method was recently reported for distorted wave collision strength calculations and applied to Fe XIII. The general method is suitable for the computation of collision strengths for complex ions and in some cases can then provide a basis for collision strength calculations in ions where ab initio computations break down or result in unnecessarily large errors. (author)

  5. ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes

    International Nuclear Information System (INIS)

    Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A.; Seltzer, S.M.; Berger, M.J.

    1993-01-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures

  6. Performance of the coupled thermalhydraulics/neutron kinetics code R/P/C on workstation clusters and multiprocessor systems

    International Nuclear Information System (INIS)

    Hammer, C.; Paffrath, M.; Boeer, R.; Finnemann, H.; Jackson, C.J.

    1996-01-01

    The light water reactor core simulation code PANBOX has been coupled with the transient analysis code RELAP5 for the purpose of performing plant safety analyses with a three-dimensional (3-D) neutron kinetics model. The system has been parallelized to improve the computational efficiency. The paper describes the features of this system with emphasis on performance aspects. Performance results are given for different types of parallelization, i. e. for using an automatic parallelizing compiler, using the portable PVM platform on a workstation cluster, using PVM on a shared memory multiprocessor, and for using machine dependent interfaces. (author)

  7. ITS Version 3.0: The Integrated TIGER Series of coupled electron/photon Monte Carlo transport codes

    Energy Technology Data Exchange (ETDEWEB)

    Halbleib, J.A.; Kensek, R.P.; Valdez, G.D.; Mehlhorn, T.A. [Sandia National Labs., Albuquerque, NM (United States); Seltzer, S.M.; Berger, M.J. [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Ionizing Radiation Div.

    1993-06-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields. It combines operational simplicity and physical accuracy in order to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Flexibility of construction permits tailoring of the codes to specific applications and extension of code capabilities to more complex applications through simple update procedures.

  8. Coupling of 3-D core computational codes and a reactor simulation software for the computation of PWR reactivity accidents induced by thermal-hydraulic transients

    International Nuclear Information System (INIS)

    Raymond, P.; Caruge, D.; Paik, H.J.

    1994-01-01

    The French CEA has recently developed a set of new computer codes for reactor physics computations called the Saphir system which includes CRONOS-2, a three-dimensional neutronic code, FLICA-4, a three-dimensional core thermal hydraulic code, and FLICA-S, a primary loops thermal-hydraulic transient computation code, which are coupled and applied to analyze a severe reactivity accident induced by a thermal hydraulic transient: the Steamline Break accident for a pressurized water reactor until soluble boron begins to accumulate in the core. The coupling of these codes has proved to be numerically stable. 15 figs., 7 refs

  9. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics

    International Nuclear Information System (INIS)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-01-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  10. Detailed modeling of KALININ-3 NPP VVER-1000 reactor pressure vessel by the coupled system code ATHLET/BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.P.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper gives an overview of the recent developments of a new reactor pressure vessel (RPV) model of VVER-1000 for the coupled system code ATHLET/BIPR-VVER. Based on the previous experience a methodology is worked out for modeling the RPV in a pseudo-3D way with the help of a multiple parallel thermal-hydraulic channel scheme that follows the hexagonal fuel assembly structure from the bottom to the top of the reactor. The results of the first application of the new modeling are discussed on the base of the OECD/NEA coupled code benchmark for Kalinin-3 NPP transient. Coolant mass flow distributions in reactor volume of VVER 1000 reactor are presented and discussed. It is shown that along the core height a mass flow re-distribution of the coolant takes place starting approximately at an axial layer located 1 meter below the core outlet. (author)

  11. The Verification of Coupled Neutronics Thermal-Hydraulics Code NODAL3 in the PWR Rod Ejection Benchmark

    Directory of Open Access Journals (Sweden)

    Surian Pinem

    2014-01-01

    Full Text Available A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the few-group neutron diffusion equation in 3-dimensional geometry for typical PWR static and transient analyses. The spatial variables are treated by using a polynomial nodal method while for the neutron dynamic solver the adiabatic and improved quasistatic methods are adopted. In this paper we report the benchmark calculation results of the code against the OECD/NEA CRP PWR rod ejection cases. The objective of this work is to determine the accuracy of NODAL3 code in analysing the reactivity initiated accident due to the control rod ejection. The NEACRP PWR rod ejection cases are chosen since many organizations participated in the NEA project using various methods as well as approximations, so that, in addition to the reference solutions, the calculation results of NODAL3 code can also be compared to other codes’ results. The transient parameters to be verified are time of power peak, power peak, final power, final average Doppler temperature, maximum fuel temperature, and final coolant temperature. The results of NODAL3 code agree well with the PHANTHER reference solutions in 1993 and 1997 (revised. Comparison with other validated codes, DYN3D/R and ANCK, shows also a satisfactory agreement.

  12. PAD: a one-dimensional, coupled neutronic-thermodynamic-hydrodynamic computer code

    International Nuclear Information System (INIS)

    Peterson, D.M.; Stratton, W.R.; McLaughlin, T.P.

    1976-12-01

    Theoretical and numerical foundations, utilization guide, sample problems, and program listing and glossary are given for the PAD computer code which describes dynamic systems with interactive neutronics, thermodynamics, and hydrodynamics in one-dimensional spherical, cylindrical, and planar geometries. The code has been applied to prompt critical excursions in various fissioning systems (solution, metal, LMFBR, etc.) as well as to nonfissioning systems

  13. The coupled code system TORT-TD/ATTICA3D for 3-D transient analysis of pebble-bed HTGR

    International Nuclear Information System (INIS)

    Seubert, A.; Sureda, A.; Lapins, J.; Buck, M.; Laurien, E.; Bader, J.; EnBW Kernkraft GmbH, Philippsburg

    2012-01-01

    This paper describes the time-dependent 3-D discrete-ordinates based coupled code system TORT-TD/ATTICA3D and its application to HTGR of pebble bed type. TORT-TD/ATTICA3D is represented by a single executable and adapts the so-called internal coupling approach. Three-dimensional distributions of temperatures from ATTICA3D and power density from TORT-TD are efficiently exchanged by direct memory access of array elements via interface routines. Applications of TORT-TD/ATTICA3D to three transients based on the PBMR-400 benchmark (total and partial control rod withdrawal and cold helium ingress) and the full power steady state of the HTR-10 are presented. For the partial control rod withdrawal, 3-D effects of local neutron flux redistributions are clearly identified. The results are very promising and demonstrate that the coupled code system TORT-TD/ATTICA3D may represent a key component in a future comprehensive 3-D code system for HTGR of pebble bed type. (orig.)

  14. VVER-1000 main steam line break analysis using the coupled code system DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Kliem, Soeren; Hoehne, Thomas; Rohde, Ulrich; Weiss, Frank-Peter; Kozmenkov, Yaroslav

    2008-01-01

    Calculations using the coupled code system DYN3D/ATHLET were performed in the frame of the OECD/NEA MSLB benchmark for a VVER-1000 reactor. The coolant mixing inside the reactor pressure vessel was treated using a validated empirical mixing model implemented into the DYN3D/ATHLET code. Using very conservative boundary conditions (reduced scram worth, two stuck rods, running MCP throughout the whole transient) a return-to-power was predicted. For the assessment of the empirical mixing model a time dependent calculation using the computational fluid dynamics code CFX-10 was performed. For that analysis, a detailed model of the reactor pressure vessel consisting of the inlets nozzles, downcomer, lower plenum and a part of the core and having 4.67 million unstructured tetra cell elements was used. For the considered case with running main coolant pumps, this calculation shows a sector formation at the core inlet with a certain amount of mixing at the edges of the sector. A core calculation using these CFX results as boundary conditions predicted also a return-to-power with a maximum value being about 200 MW lower than in the coupled code calculation. This variation calculation confirms the applicability of the empirical mixing model. The comparison shows also, that in this way results with a reasonable degree of conservatism can be obtained. (authors)

  15. Assessment of the Effects on PCT Evaluation of Enhanced Fuel Model Facilitated by Coupling the MARS Code with the FRAPTRAN Code

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyong Chol; Lee, Young Jin; Han, Sam Hee [NSE Technology Inc., Daejeon (Korea, Republic of)

    2016-10-15

    The principal objectives of the two safety criteria, peak cladding temperature (PCT) and total oxidation limits, are to ensure that the fuel rod claddings remain sufficiently ductile so that they do not crack and fragment during a LOCA. Another important purpose of the PCT limit is to ensure that the fuel cladding does not enter the regime of runaway oxidation and uncontrollable heat-up. However, even when the PCT limit is satisfied, it is known that cladding failures may still occur in a certain percentage of the fuel rods during a LOCA. This is largely because a 100% fuel failure is assumed for the radiological consequence analysis in the US regulatory practices. In this study, we analyze the effects of cladding failure and other fuel model features on PCT during a LOCA using the MARS-FRAPTRAN coupled code. MARS code has been coupled with FRAPTRAN code to extend fuel modeling capability. The coupling allows feedback of FRAPTRAN results in real time. Because of the significant impact of fuel models on key safety parameters such as PCT, detailed and accurate fuel models should be employed when evaluating PCT in LOCA analysis. It is noteworthy that the ECCS evaluation models laid out in the Appendix K to 10CFR50 require a provision for predicting cladding swelling and rupture and require to assume that the inside of the cladding react with steam after the rupture. The metal-water reaction energy can have significantly large effect on the reflood PCT, especially when fuel failure occurs. Effects of applying an advanced fuel model on the PCT evaluation can be clearly seen when comparing the MARS and the FRAPTRAN results in both the one-way calculation and the feedback calculation. As long as MARS and FRAPTRAN are used respectively in the ranges where they have been validated, the coupled calculation results are expected to be valid and to reveal various aspects of phenomena which have not been discovered in previous uncoupled calculations by MARS or FRAPTRAN.

  16. PC-based support programs coupled with the sets code for large fault tree analysis

    International Nuclear Information System (INIS)

    Hioki, K.; Nakai, R.

    1989-01-01

    Power Reactor and Nuclear Fuel Development Corporation (PNC) has developed four PC programs: IEIQ (Initiating Event Identification and Quantification), MODESTY (Modular Even Description for a Variety of Systems), FAUST (Fault Summary Tables Generation Program) and ETAAS (Event Tree Analysis Assistant System). These programs prepare the input data for the SETS (Set Equation Transformation System) code and construct and quantify event trees (E/Ts) using the output of the SETS code. The capability of these programs is described and some examples of the results are presented in this paper. With these PC programs and the SETS code, PSA can now be performed with more consistency and less manpower

  17. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    International Nuclear Information System (INIS)

    Sanchez-Espinoza, Victor Hugo

    2008-07-01

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  18. Investigations of the VVER-1000 coolant transient benchmark phase 1 with the coupled code system RELAP5/PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez-Espinoza, Victor Hugo

    2008-07-15

    As part of the reactor dynamics activities of FZK/IRS, the qualification of best-estimate coupled code systems for reactor safety evaluations is a key step toward improving their prediction capability and acceptability. The VVER-1000 Coolant Transient Benchmark Phase 1 represents an excellent opportunity to validate the simulation capability of the coupled code system RELAP5/PACRS regarding both the thermal hydraulic plant response (RELAP5) using measured data obtained during commissioning tests at the Kozloduy nuclear power plant unit 6 and the neutron kinetics models of PARCS for hexagonal geometries. The Phase 1 is devoted to the analysis of the switching on of one main coolant pump while the other three pumps are in operation. It includes the following exercises: (a) investigation of the integral plant response using a best-estimate thermal hydraulic system code with a point kinetics model (b) analysis of the core response for given initial and transient thermal hydraulic boundary conditions using a coupled code system with 3D-neutron kinetics model and (c) investigation of the integral plant response using a best-estimate coupled code system with 3D-neutron kinetics. Already before the test, complex flow conditions exist within the RPV e.g. coolant mixing in the upper plenum caused by the reverse flow through the loop-3 with the stopped pump. The test is initiated by switching on the main coolant pump of loop-3 that leads to a reversal of the flow through the respective piping. After about 13 s the mass flow rate through this loop reaches values comparable with the one of the other loops. During this time period, the increased primary coolant flow causes a reduction of the core averaged coolant temperature and thus an increase of the core power. Later on, the power stabilizes at a level higher than the initial power. In this analysis, special attention is paid on the prediction of the spatial asymmetrical core cooling during the test and its effects on the

  19. Three Mile Island Unit 1 Main Steam Line Break Three-Dimensional Neutronics/Thermal-Hydraulics Analysis: Application of Different Coupled Codes

    International Nuclear Information System (INIS)

    D'Auria, Francesco; Moreno, Jose Luis Gago; Galassi, Giorgio Maria; Grgic, Davor; Spadoni, Antonino

    2003-01-01

    A comprehensive analysis of the double ended main steam line break (MSLB) accident assumed to occur in the Babcock and Wilcox Three Mile Island Unit 1 (TMI-1) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy, in cooperation with the University of Zagreb, Croatia. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development-Committee on the Safety of Nuclear Installations-Nuclear Science Committee pressurized water reactor MSLB benchmark.Thermal-hydraulic system codes (various versions of Relap5), three-dimensional (3-D) neutronics codes (Parcs, Quabbox, and Nestle), and one subchannel code (Cobra) have been adopted for the analysis. Results from the following codes (or code versions) are assumed as reference:1. Relap5/mod3.2.2, beta version, coupled with the 3-D neutron kinetics Parcs code parallel virtual machine (PVM) coupling2. Relap5/mod3.2.2, gamma version, coupled with the 3-D neutron kinetics Quabbox code (direct coupling)3. Relap5/3D code coupled with the 3-D neutron kinetics Nestle code.The influence of PVM and of direct coupling is also discussed.Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by Pennsylvania State University in cooperation with GPU Nuclear Corporation (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission. The comparison among the results obtained by adopting the same thermal-hydraulic nodalization and the coupled code version is discussed in this paper.The capability of the control rods to recover the accident has been demonstrated in all the cases as well as the capability of all the codes to predict the time evolution of the assigned transient. However, one stuck control rod caused some 'recriticality' or 'return to power' whose magnitude is largely affected by boundary and initial conditions

  20. Analysis of OECD/CSNI ISP-42 phase A PANDA experiment using coupled code R5G (RELAP5-GOTHIC)

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Grgic, D.; Bajs, T.

    2010-01-01

    In the paper, the results of the analysis of OECD/CSNI ISP-42 Phase A experiment at PANDA facility using stand-alone codes RELAP5/mod3.3 and GOTHIC 7.2b as well as coupled code R5G (RELAP5/mod3.3-GOTHIC 7.2b) are presented. PANDA is a large-scale thermal-hydraulic test facility installed at PSI (Paul Scherrer Institute) in Switzerland. The OECD/CSNI ISP-42 test consists of six sequential phases (Phase A through F). The present work deals with the post-test calculation of the Phase A, including the break of the main steam line and the Passive Containment Cooling (PCC) System Start-Up. The objective of the test is to investigate the start-up phenomenology of passive cooling system when steam is injected into cold vessel filled with air. The calculation was performed using stand-alone RELAP5/mod3.3 and GOTHIC 7.2b models, and then the same calculation was performed using coupled code with RELAP5 being responsible for reactor part of the model and GOTHIC being responsible for containment part of the model. The prediction capability, running time and modeling aspects were discussed for all three cases. (authors)

  1. Parallelization of the MAAP-A code neutronics/thermal hydraulics coupling

    International Nuclear Information System (INIS)

    Froehle, P.H.; Wei, T.Y.C.; Weber, D.P.; Henry, R.E.

    1998-01-01

    A major new feature, one-dimensional space-time kinetics, has been added to a developmental version of the MAAP code through the introduction of the DIF3D-K module. This code is referred to as MAAP-A. To reduce the overall job time required, a capability has been provided to run the MAAP-A code in parallel. The parallel version of MAAP-A utilizes two machines running in parallel, with the DIF3D-K module executing on one machine and the rest of the MAAP-A code executing on the other machine. Timing results obtained during the development of the capability indicate that reductions in time of 30--40% are possible. The parallel version can be run on two SPARC 20 (SUN OS 5.5) workstations connected through the ethernet. MPI (Message Passing Interface standard) needs to be implemented on the machines. If necessary the parallel version can also be run on only one machine. The results obtained running in this one-machine mode identically match the results obtained from the serial version of the code

  2. Neutronic and thermal-hydraulic coupling using Milonga and OpenFOAM codes: an approach using free software

    International Nuclear Information System (INIS)

    Silva, Vitor Vasconcelos Araújo

    2016-01-01

    The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)

  3. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-01-01

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  4. KENOREST - A new coupled code system based on KENO and OREST for criticality and burnup inventory calculations

    International Nuclear Information System (INIS)

    Hesse, U.; Gmal, B.; Voggenberger, Th.; Baleanu, M.; Langenbuch, S.

    2001-01-01

    The program system KENOREST version 1998 will be presented, which is a useful tool for burnup and reactivity calculations for LWR fuel. The three-dimensional Monte Carlo code KENO-V.a is coupled with the one-dimensional GRS burnup program system OREST-98. The objective is to achieve a better modelling of plutonium and actinide build-up or burnout for advanced heterogeneous fuel assembly designs. Further objectives are directed to reliable calculations of the pin power distributions and of reactor safety parameters including axial and radial rod temperatures for fuel assemblies of modern design. The stand-alone-code KENO-V.a version is used without any changes in the program source. The OREST-98 system was developed to handle multirod problems and additional burnup dependent moderator conditions which can be applied to stretch-out simulations in the reactor. A new interface module RESPEFF between KENO and OREST transforms the 2-d or 3-d KENO flux results to the one-dimensional lattice code OREST in a fully automated manner to maintain reaction rate balance between the codes. First results for assembly multiplication factors, isotope inventories are compared with OECD results. (author)

  5. Safety related investigations of the VVER-1000 reactor type by the coupled code system TRACE/PARCS

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Lischke, Wolfgang; Sanchez Espinoza, Victor Hugo

    2007-01-01

    This study was performed at the Institute of Reactor Safety at the Research Center Karlsruhe. It is embedded in the ongoing investigations of the international code application and maintenance program (CAMP) for qualification and validation of system codes like TRACE [1] and PARCS [2]. The predestinated reactor type for the validation of these two codes was the Russian designed VVER-1000 because the OECD/NEA VVER-1000 Coolant Transient Benchmark Phase 2 [3] includes detailed information of the Bulgarian nuclear power plant (NPP) Kozloduy unit 6. The posttest-investigations of a coolant mixing experiment have shown that the predicted parameters (coolant temperature, pressure drop, etc.) are in good agreement to the measured data. The coolant mixing pattern especially in the downcomer has been also reproduced quiet well by TRACE. The coupled code system TRACE/PARCS which was applied on a postulated main steam line break (MSLB) provides good results compared to reference values and the ones of other participants of the benchmark. It can be pointed out that the developed three-dimensional nodalisation of the reactor pressure vessel (RPV) is appropriate for the description of transients where the thermal-hydraulics and the neutronics are strongly linked. (author)

  6. Improvement of MSLB transient analysis for VVER by the coupled code system KIKO3D/ATHLET

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2001-01-01

    An overview is given on the investigations of the Main Steam Line Break transient in a VVER- 440 NPP by using the KIKO3D/ATHLET 1.2.A coupled code system. Special attention was paid for the influence of modeling the outcore detector signals and the malfunctioning of the emergency control system (scram with stuck rod). The conservatism of the calculations was assured even in the case of application of the 3D best estimate KIKO3D code. The consequence of MSLB accident is investigated at the end of cycle (EOC), at full power (FP) and shut down initial conditions. Even if very strong conservative assumptions were applied, dangerous hot spots were not found in the supposed scenarios.(author)

  7. TITAN: an advanced three-dimensional coupled neutronic/thermal-hydraulics code for light water nuclear reactor core analysis

    International Nuclear Information System (INIS)

    Griggs, D.P.; Kazimi, M.S.; Henry, A.F.

    1984-06-01

    The three-dimensional nodal neutronics code QUANDRY and the three-dimensional two-fluid thermal-hydraulics code THERMIT are combined into TITAN. Steady-state and transient coupling methodologies based upon a tandem structure were devised and implemented. Additional models for nuclear feedback, equilibrium xenon and direct moderator heating were added. TITAN was tested using a boiling water two channel problem and the coupling methodologies were shown to be effective. Simulated turbine trip transients and several control rod withdrawal transients were analyzed with good results. Sensitivity studies indicated that the time-step size can affect transient results significantly. TITAN was also applied to a quarter core PWR problem based on a real reactor geometry. The steady-state results were compared to a solution produced by MEKIN-B and poor agreement between the horizontal power shapes was found. Calculations with various mesh spacings showed that the mesh spacings in the MEKIN-B analysis were too large to produce accurate results with a finite difference method. The TITAN results were shown to be reasonable. A pair of control rod ejection accidents were also analyzed with TITAN. A comparison of the TITAN PWR control rod ejection results with results from coupled point kinetics/thermal-hydraulics analyses showed that the point kinetics method used (adiabatic method for control rod reactivities, steady-state flux shape for core-averaged reactivity feedback) underpredicted the power excursion in one case and overpredicted it in the other. It was therefore concluded that point kinetics methods should be used with caution and that three-dimensional codes like TITAN are superior for analyzing PWR control rod ejection transients

  8. An Analysis of Language Code Used by the Cross- Married Couples, Banjarese- Javanese Ethnics: A Case Study in South Kalimantan Province, Indonesia

    Directory of Open Access Journals (Sweden)

    - Supiani

    2016-08-01

    Full Text Available This research aims to describe the use of language code applied by the participants and to find out the factors influencing the choice of language codes. This research is qualitative research that describe the use of language code in the cross married couples. The data are taken from the discourses about language code phenomena dealing with the cross- married couples, Banjarese- Javanese ethnics in Tanah Laut regency South Kalimantan, Indonesia. The conversations occur in the family and social life such as between a husband and a wife, a father and his son/daughter, a mother and her son/daughter, a husband and his friends, a wife and her neighbor, and so on. There are 23 data observed and recoded by the researcher based on a certain criteria. Tanah Laut regency is chosen as a purposive sample where this regency has many different ethnics so that they do cross cultural marriage for example between Banjarese- Javanese ethnics. Findings reveal that mostly the cross married couple used code mixing and code switching in their conversation of daily activities. Code mixing is uttered by Javanese father or mother to their children. Mixed codes are used namely Banjarese+Javanese+Indonesian. Meanwhile, code switching occurs when there is another factor or a new participant who join in the discourse. The codes change from Banjarese to Indonesian codes or Javanese to Indonesian codes due to new participant who involve himself/herself in the dialogue. The influential factors are situational factors, the environment (neighborhood, relative status, and ethnicity. Keywords: Language codes, Cross- married couples, Banjarese and Javanese ethics, Dialects

  9. Enforcing dust mass conservation in 3D simulations of tightly coupled grains with the PHANTOM SPH code

    Science.gov (United States)

    Ballabio, G.; Dipierro, G.; Veronesi, B.; Lodato, G.; Hutchison, M.; Laibe, G.; Price, D. J.

    2018-06-01

    We describe a new implementation of the one-fluid method in the SPH code PHANTOM to simulate the dynamics of dust grains in gas protoplanetary discs. We revise and extend previously developed algorithms by computing the evolution of a new fluid quantity that produces a more accurate and numerically controlled evolution of the dust dynamics. Moreover, by limiting the stopping time of uncoupled grains that violate the assumptions of the terminal velocity approximation, we avoid fatal numerical errors in mass conservation. We test and validate our new algorithm by running 3D SPH simulations of a large range of disc models with tightly and marginally coupled grains.

  10. Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

    International Nuclear Information System (INIS)

    Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane

    2015-01-01

    Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

  11. Simulation of the first step of the coupling of the PARCS/RELAP5 codes to ANGRA 2 facility

    Energy Technology Data Exchange (ETDEWEB)

    Del Pozzo, Andrea Sanchez; Andrade, Delvonei A. de; Sabundjian, Gaiane, E-mail: delvonei@ipen.br, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Since the Three Mile Island (1979) and Chernobyl (1986) accidents, the International Agency of Energy Atomic (IAEA) has worked with the authorities of other countries that use nuclear power plants in order to guarantee the safe of those facilities. The utilities have simulated design basic accidents to verify the integrity of the nuclear power plant to these events. However, after Fukushima accident in Japan (2011), the people have felt insecure and been afraid in relation to nuclear power plants. Today, the international and national organizations, such as the International Agency of Energy Atomic (IAEA) and Comissao Nacional de Energia Nuclear (CNEN), respectively, have worked very hard to prevent some accidents and transients in nuclear power plants in order to ensure the security of the general population. In case of accidents, as the Rod Ejection Accident (REA), it is very important to do the coupling between neutronic and thermal hydraulic areas of nuclear reactors. To solve this type of problem there is the coupling between PARCS/RELAP5 codes. However, to perform this analysis it is necessary to simulate three steps. The first step is simulating the steady state of one nuclear power plant by using RELAP5 code. The second step is to run the steady state of this reactor using the coupling PARCS/RELAP5, and the final step is simulating the REA of this facility with PARCS/RELAP5 coupling. The aim of this work is to show the results of the first step of this analysis, i.e., by means of simulation the steady state of Angra 2 nuclear power plant using RELAP5 version 3.3. In this case, the modeling from the core was more detailed than in the original version developed some years ago for Angra 2. The results obtained in this work were satisfactory. (author)

  12. STRUYA a code for two-dimensional fluid flow analysis with and without structure coupling

    International Nuclear Information System (INIS)

    Katz, F.W.; Schlechtendahl, E.G.; Stoelting, K.

    1979-11-01

    STRUYA is a code for two-dimensional subsonic and supersonic flow analysis. Both Eulerian and Lagrangian grids are allowed. In the third dimension the flow domain may be bounded by a moving wall. The wall movement may be prescribed in a time-and space varying way or computed by a structural model. STRUYA offers a general scheme for adapting various structural models. As a standard feature it includes a cylindrical shell model (CYLDY2). (orig.) [de

  13. ICECO-CEL: a coupled Eulerian-Lagrangian code for analyzing primary system response in fast reactors

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1981-02-01

    This report describes a coupled Eulerian-Lagrangian code, ICECO-CEL, for analyzing the response of the primary system during hypothetical core disruptive accidents. The implicit Eulerian method is used to calculate the fluid motion so that large fluid distortion, two-dimensional sliding interface, flow around corners, flow through coolant passageways, and out-flow boundary conditions can be treated. The explicit Lagrangian formulation is employed to compute the response of the containment vessel and other elastic-plastic solids inside the reactor containment. Large displacements, as well as geometrical and material nonlinearities are considered in the analysis. Marker particles are utilized to define the free surface or the material interface and to visualize the fluid motion. The basic equations and numerical techniques used in the Eulerian hydrodynamics and Lagrangian structural dynamics are described. Treatment of the above-core hydrodynamics, sodium spillage, fluid cavitation, free-surface boundary conditions and heat transfer are also presented. Examples are given to illustrate the capabilities of the computer code. Comparisons of the code predictions with available experimental data are also made

  14. Analysis of PBMR transients using a coupled neutron transport/thermal-hydraulics code DORT-TD/thermix

    International Nuclear Information System (INIS)

    Tyobeka, B.; Ivanov, K.; Pautz, A.

    2007-01-01

    In the advent of increased demand for safety and economics of nuclear power plants, nuclear engineers and designers are called upon to develop advanced computation tools. In these developments, space-time effects in the dynamics of nuclear reactors must be considered within the framework of a full 3-dimensional treatment of both neutron kinetics and thermal hydraulics. In a recent effort at the Pennsylvania State University, a time-dependent version of the discrete ordinates transport code DORT, DORT-TD was coupled to a 2-dimensional core thermal hydraulics code THERMIX-DIREKT. In the coupling process, a feedback model was developed to account for the feedback effects and was implemented into DORT-TD. During the calculation process for each spatial node of the DORT-TD core model, feedback parameters representative of this node are passed to the feedback module. Using these values, cross section tables are then interpolated for the appropriate macroscopic cross section values. The updated macroscopic cross sections are passed back to DORT-TD to perform transport core calculations, and the power distribution is transferred to THERMIX-DIREKT to obtain the relevant thermal-hydraulics data in turn, and this calculation loop continues. In this paper, DORT-TD/THERMIX is used to simulate transients of interest in the PBMR (Pebble Bed Modular Reactor) safety using established benchmark problems: load change from 100% to 40% power and fast control rod ejection (PBMR-268 benchmark problem). The results obtained are compared with those obtained using the diffusion-based module of the code. The results are only preliminary and so far show that diffusion theory is not such a bad approximation for PBMR for the prediction of integral parameters

  15. Partial Encryption of Entropy-Coded Video Compression Using Coupled Chaotic Maps

    Directory of Open Access Journals (Sweden)

    Fadi Almasalha

    2014-10-01

    Full Text Available Due to pervasive communication infrastructures, a plethora of enabling technologies is being developed over mobile and wired networks. Among these, video streaming services over IP are the most challenging in terms of quality, real-time requirements and security. In this paper, we propose a novel scheme to efficiently secure variable length coded (VLC multimedia bit streams, such as H.264. It is based on code word error diffusion and variable size segment shuffling. The codeword diffusion and the shuffling mechanisms are based on random operations from a secure and computationally efficient chaos-based pseudo-random number generator. The proposed scheme is ubiquitous to the end users and can be deployed at any node in the network. It provides different levels of security, with encrypted data volume fluctuating between 5.5–17%. It works on the compressed bit stream without requiring any decoding. It provides excellent encryption speeds on different platforms, including mobile devices. It is 200% faster and 150% more power efficient when compared with AES software-based full encryption schemes. Regarding security, the scheme is robust to well-known attacks in the literature, such as brute force and known/chosen plain text attacks.

  16. Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

    Energy Technology Data Exchange (ETDEWEB)

    McCaskey, Alexander J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Slattery, Stuart [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Billings, Jay Jay [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

    The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy's Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report details a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To achieve this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. This report describes this work in full. We begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from across the NEAMS toolkit to enable BISON-PROTEUS integration. Furthermore, we show how

  17. Efficient full wave code for the coupling of large multirow multijunction LH grills.

    Czech Academy of Sciences Publication Activity Database

    Preinhaelter, Josef; Hillairet, J.; Milanesio, D.; Maggiora, R.; Urban, Jakub; Vahala, L.; Vahala, G.

    2017-01-01

    Roč. 57, č. 11 (2017), č. článku 116060. ISSN 0029-5515 R&D Projects: GA MŠk(CZ) 8D15001; GA MŠk(CZ) LM2015045 Institutional support: RVO:61389021 Keywords : lower hybrid waves * coupling * large multirow multijunction grills * tokamak * full-wave Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 3.307, year: 2016 http://iopscience.iop.org/article/10.1088/1741-4326/aa7f4f/meta

  18. Offshore code comparison collaboration continuation (OC4), phase I - Results of coupled simulations of an offshore wind turbine with jacket support structure

    DEFF Research Database (Denmark)

    Popko, Wojciech; Vorpahl, Fabian; Zuga, Adam

    2012-01-01

    In this paper, the exemplary results of the IEA Wind Task 30 "Offshore Code Comparison Collaboration Continuation" (OC4) Project - Phase I, focused on the coupled simulation of an offshore wind turbine (OWT) with a jacket support structure, are presented. The focus of this task has been the verif......In this paper, the exemplary results of the IEA Wind Task 30 "Offshore Code Comparison Collaboration Continuation" (OC4) Project - Phase I, focused on the coupled simulation of an offshore wind turbine (OWT) with a jacket support structure, are presented. The focus of this task has been...... the verification of OWT modeling codes through code-to-code comparisons. The discrepancies between the results are shown and the sources of the differences are discussed. The importance of the local dynamics of the structure is depicted in the simulation results. Furthermore, attention is given to aspects...

  19. Comparison of a Coupled Near and Far Wake Model With a Free Wake Vortex Code

    DEFF Research Database (Denmark)

    Pirrung, Georg; Riziotis, Vasilis; Aagaard Madsen, Helge

    2016-01-01

    to be updated during the computation. Further, the effect of simplifying the exponential function approximation of the near wake model to increase the computation speed is investigated in this work. A modification of the dynamic inflow weighting factors of the far wake model is presented that ensures good...... computations performed using a free wake panel code. The focus of the description of the aerodynamics model is on the numerical stability, the computation speed and the accuracy of 5 unsteady simulations. To stabilize the near wake model, it has to be iterated to convergence, using a relaxation factor that has...... and a BEM model is centered around the NREL 5 MW reference turbine. The response to pitch steps at different pitching speeds is compared. By means of prescribed vibration cases, the effect of the aerodynamic model on the predictions of the aerodynamic work is investigated. The validation shows that a BEM...

  20. Spectral history model in DYN3D: Verification against coupled Monte-Carlo thermal-hydraulic code BGCore

    International Nuclear Information System (INIS)

    Bilodid, Y.; Kotlyar, D.; Margulis, M.; Fridman, E.; Shwageraus, E.

    2015-01-01

    Highlights: • Pu-239 based spectral history method was tested on 3D BWR single assembly case. • Burnup of a BWR fuel assembly was performed with the nodal code DYN3D. • Reference solution was obtained by coupled Monte-Carlo thermal-hydraulic code BGCore. • The proposed method accurately reproduces moderator density history effect for BWR test case. - Abstract: This research focuses on the verification of a recently developed methodology accounting for spectral history effects in 3D full core nodal simulations. The traditional deterministic core simulation procedure includes two stages: (1) generation of homogenized macroscopic cross section sets and (2) application of these sets to obtain a full 3D core solution with nodal codes. The standard approach adopts the branch methodology in which the branches represent all expected combinations of operational conditions as a function of burnup (main branch). The main branch is produced for constant, usually averaged, operating conditions (e.g. coolant density). As a result, the spectral history effects that associated with coolant density variation are not taken into account properly. Number of methods to solve this problem (such as micro-depletion and spectral indexes) were developed and implemented in modern nodal codes. Recently, we proposed a new and robust method to account for history effects. The methodology was implemented in DYN3D and involves modification of the few-group cross section sets. The method utilizes the local Pu-239 concentration as an indicator of spectral history. The method was verified for PWR and VVER applications. However, the spectrum variation in BWR core is more pronounced due to the stronger coolant density change. The purpose of the current work is investigating the applicability of the method to BWR analysis. The proposed methodology was verified against recently developed BGCore system, which couples Monte Carlo neutron transport with depletion and thermal-hydraulic solvers and

  1. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    International Nuclear Information System (INIS)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

    2005-01-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  2. Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

    Energy Technology Data Exchange (ETDEWEB)

    Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

    2005-07-01

    Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

  3. Misconception regarding conventional coupling of fields and particles in XFEL codes

    Energy Technology Data Exchange (ETDEWEB)

    Geloni, Gianluca [Europeam XFEL GmbH, Hamburg (Germany); Kocharyan, Vitali; Saldin, Evgeni [DESY Hamburg (Germany)

    2016-01-15

    Maxwell theory is usually treated in the laboratory frame under the standard time order, that is the usual light-signal clock synchronization. In contrast, particle tracking in the laboratory frame usually treats time as an independent variable. As a result, here we argue that the evolution of electron beams is usually treated according to the absolute time convention i.e. using a different time order defined by a non-standard clock synchronization procedure. This essential point has never received attention in the accelerator community. There are two possible ways of coupling fields and particles in this situation. The first, Lorentz's prerelativistic way, consists in a 'translation' of Maxwell's electrodynamics to the absolute time world-picture. The second, Einstein's way, consists in a 'translation' of particle tracking results to the electromagnetic world-picture, obeying the standard time order. Conventional particle tracking shows that the electron beam direction changes after a transverse kick, while the orientation of the microbunching phase front stays unvaried. Here we show that in the ultrarelativistic asymptotic v → c, the orientation of the planes of simultaneity, i.e. the orientation of the microbunching fronts, is always perpendicular to the electron beam velocity when the evolution of the modulated electron beam is treated under Einstein's time order. This effect allows for the production of coherent undulator radiation from a modulated electron beam in the kicked direction without suppression. We hold a recent FEL study at the LCLS as a direct experimental evidence that the microbunching wavefront indeed readjusts its direction after the electron beam is kicked by a large angle, limited only by the beamline aperture. In a previous paper we quantitatively described this result invoking the aberration of light effect, which corresponds to Lorentz's way of coupling fields and particles. The purpose of

  4. Misconception regarding conventional coupling of fields and particles in XFEL codes

    International Nuclear Information System (INIS)

    Geloni, Gianluca; Kocharyan, Vitali; Saldin, Evgeni

    2016-01-01

    Maxwell theory is usually treated in the laboratory frame under the standard time order, that is the usual light-signal clock synchronization. In contrast, particle tracking in the laboratory frame usually treats time as an independent variable. As a result, here we argue that the evolution of electron beams is usually treated according to the absolute time convention i.e. using a different time order defined by a non-standard clock synchronization procedure. This essential point has never received attention in the accelerator community. There are two possible ways of coupling fields and particles in this situation. The first, Lorentz's prerelativistic way, consists in a 'translation' of Maxwell's electrodynamics to the absolute time world-picture. The second, Einstein's way, consists in a 'translation' of particle tracking results to the electromagnetic world-picture, obeying the standard time order. Conventional particle tracking shows that the electron beam direction changes after a transverse kick, while the orientation of the microbunching phase front stays unvaried. Here we show that in the ultrarelativistic asymptotic v → c, the orientation of the planes of simultaneity, i.e. the orientation of the microbunching fronts, is always perpendicular to the electron beam velocity when the evolution of the modulated electron beam is treated under Einstein's time order. This effect allows for the production of coherent undulator radiation from a modulated electron beam in the kicked direction without suppression. We hold a recent FEL study at the LCLS as a direct experimental evidence that the microbunching wavefront indeed readjusts its direction after the electron beam is kicked by a large angle, limited only by the beamline aperture. In a previous paper we quantitatively described this result invoking the aberration of light effect, which corresponds to Lorentz's way of coupling fields and particles. The purpose of

  5. Code Cactus; Code Cactus

    Energy Technology Data Exchange (ETDEWEB)

    Fajeau, M; Nguyen, L T; Saunier, J [Commissariat a l' Energie Atomique, Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France)

    1966-09-01

    This code handles the following problems: -1) Analysis of thermal experiments on a water loop at high or low pressure; steady state or transient behavior; -2) Analysis of thermal and hydrodynamic behavior of water-cooled and moderated reactors, at either high or low pressure, with boiling permitted; fuel elements are assumed to be flat plates: - Flowrate in parallel channels coupled or not by conduction across plates, with conditions of pressure drops or flowrate, variable or not with respect to time is given; the power can be coupled to reactor kinetics calculation or supplied by the code user. The code, containing a schematic representation of safety rod behavior, is a one dimensional, multi-channel code, and has as its complement (FLID), a one-channel, two-dimensional code. (authors) [French] Ce code permet de traiter les problemes ci-dessous: 1. Depouillement d'essais thermiques sur boucle a eau, haute ou basse pression, en regime permanent ou transitoire; 2. Etudes thermiques et hydrauliques de reacteurs a eau, a plaques, a haute ou basse pression, ebullition permise: - repartition entre canaux paralleles, couples on non par conduction a travers plaques, pour des conditions de debit ou de pertes de charge imposees, variables ou non dans le temps; - la puissance peut etre couplee a la neutronique et une representation schematique des actions de securite est prevue. Ce code (Cactus) a une dimension d'espace et plusieurs canaux, a pour complement Flid qui traite l'etude d'un seul canal a deux dimensions. (auteurs)

  6. A fully-implicit Particle-In-Cell Monte Carlo Collision code for the simulation of inductively coupled plasmas

    Science.gov (United States)

    Mattei, S.; Nishida, K.; Onai, M.; Lettry, J.; Tran, M. Q.; Hatayama, A.

    2017-12-01

    We present a fully-implicit electromagnetic Particle-In-Cell Monte Carlo collision code, called NINJA, written for the simulation of inductively coupled plasmas. NINJA employs a kinetic enslaved Jacobian-Free Newton Krylov method to solve self-consistently the interaction between the electromagnetic field generated by the radio-frequency coil and the plasma response. The simulated plasma includes a kinetic description of charged and neutral species as well as the collision processes between them. The algorithm allows simulations with cell sizes much larger than the Debye length and time steps in excess of the Courant-Friedrichs-Lewy condition whilst preserving the conservation of the total energy. The code is applied to the simulation of the plasma discharge of the Linac4 H- ion source at CERN. Simulation results of plasma density, temperature and EEDF are discussed and compared with optical emission spectroscopy measurements. A systematic study of the energy conservation as a function of the numerical parameters is presented.

  7. Benchmark of coupling codes (ALOHA, TOPLHA and GRILL3D) with ITER-relevant Lower Hybrid antenna

    International Nuclear Information System (INIS)

    Milanesio, D.; Hillairet, J.; Panaccione, L.; Maggiora, R.; Artaud, J.F.; Bae, Y.S.; Barbera, A.M.A.; Belo, J.; Berger-By, G.; Bernard, J.M.; Cara, Ph.; Cardinali, A.; Castaldo, C.; Ceccuzzi, S.; Cesario, R.; Decker, J.; Delpech, L.; Ekedahl, A.; Garcia, J.; Garibaldi, P.

    2011-01-01

    In order to assist the design of the future ITER Lower Hybrid launcher, coupling codes ALOHA, from CEA/IRFM, TOPLHA, from Politecnico di Torino, and GRILL3D, developed by Dr. Mikhail Irzak (A.F. Ioffe Physico-Technical Institute, St. Petersburg, Russia) and operated by ENEA Frascati, have been compared with the updated (six modules with four active waveguides per module) Passive-Active Multi-junction (PAM) Lower Hybrid antennas. Both ALOHA and GRILL3D formulate the problem in terms of rectangular waveguides modes, while TOPLHA is based on boundary-value problem with the adoption of a triangular cell-mesh to represent the relevant waveguides surfaces. Several plasma profiles, with varying edge density and density increase, have been adopted to provide a complete description of the simulated launcher in terms of reflection coefficient, computed at the beginning of each LH module, and of power spectra. Good agreement can be observed among codes for all the simulated profiles.

  8. The FLUKA Monte Carlo code coupled with the local effect model for biological calculations in carbon ion therapy

    CERN Document Server

    Mairani, A; Kraemer, M; Sommerer, F; Parodi, K; Scholz, M; Cerutti, F; Ferrari, A; Fasso, A

    2010-01-01

    Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fur Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed C-12 ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-d...

  9. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    2007-01-01

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  10. AMPX: a modular code system for generating coupled multigroup neutron-gamma libraries from ENDF/B

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Lucius, J.L.; Petrie, L.M.; Ford, W.E. III; White, J.E.; Wright, R.Q.

    1976-03-01

    AMPX is a modular system for producing coupled multigroup neutron-gamma cross section sets. Basic neutron and gamma cross-section data for AMPX are obtained from ENDF/B libraries. Most commonly used operations required to generate and collapse multigroup cross-section sets are provided in the system. AMPX is flexibly dimensioned; neutron group structures, and gamma group structures, and expansion orders to represent anisotropic processes are all arbitrary and limited only by available computer core and budget. The basic processes provided will (1) generate multigroup neutron cross sections; (2) generate multigroup gamma cross sections; (3) generate gamma yields for gamma-producing neutron interactions; (4) combine neutron cross sections, gamma cross sections, and gamma yields into final ''coupled sets''; (5) perform one-dimensional discrete ordinates transport or diffusion theory calculations for neutrons and gammas and, on option, collapse the cross sections to a broad-group structure, using the one-dimensional results as weighting functions; (6) plot cross sections, on option, to facilitate the ''evaluation'' of a particular multigroup set of data; (7) update and maintain multigroup cross section libraries in such a manner as to make it not only easy to combine new data with previously processed data but also to do it in a single pass on the computer; and (8) output multigroup cross sections in convenient formats for other codes. (auth)

  11. The coupling of the Star-Cd software to a whole-core neutron transport code Decart for PWR applications

    International Nuclear Information System (INIS)

    Thomas, J.W.; Lee, H.C.; Downar, T.J.; Sofu, T.; Weber, D.P.; Joo, H.G.; Cho, J.Y.

    2003-01-01

    As part of a U.S.- Korea collaborative U.S. Department of Energy INERI project, a comprehensive high-fidelity reactor-core modeling capability is being developed for detailed analysis of existing and advanced PWR reactor designs. An essential element of the project has been the development of an interface between the computational fluid dynamics (CFD) module, STAR-CD, and the neutronics module, DeCART. Since the computational mesh for CFD and neutronics calculations are generally different, the capability to average and decompose data on these different meshes has been an important part of code coupling activities. An averaging process has been developed to extract neutronics zone temperatures in the fuel and coolant and to generate appropriate multi group cross sections and densities. Similar procedures have also been established to map the power distribution from the neutronics zones to the mesh structure used in the CFD module. Since MPI is used as the parallel model in STAR-CD and conflicts arise during initiation of a second level of MPI, the interface developed here is based on using TCP/IP protocol sockets to establish communication between the CFD and neutronics modules. Preliminary coupled calculations have been performed for PWR fuel assembly size problems and converged solutions have been achieved for a series of steady-state problems ranging from a single pin to a 1/8 model of a 17 x 17 PWR fuel assembly. (authors)

  12. Validation of the coupling of mesh models to GEANT4 Monte Carlo code for simulation of internal sources of photons

    International Nuclear Information System (INIS)

    Caribe, Paulo Rauli Rafeson Vasconcelos; Cassola, Vagner Ferreira; Kramer, Richard; Khoury, Helen Jamil

    2013-01-01

    The use of three-dimensional models described by polygonal meshes in numerical dosimetry enables more accurate modeling of complex objects than the use of simple solid. The objectives of this work were validate the coupling of mesh models to the Monte Carlo code GEANT4 and evaluate the influence of the number of vertices in the simulations to obtain absorbed fractions of energy (AFEs). Validation of the coupling was performed to internal sources of photons with energies between 10 keV and 1 MeV for spherical geometries described by the GEANT4 and three-dimensional models with different number of vertices and triangular or quadrilateral faces modeled using Blender program. As a result it was found that there were no significant differences between AFEs for objects described by mesh models and objects described using solid volumes of GEANT4. Since that maintained the shape and the volume the decrease in the number of vertices to describe an object does not influence so meant dosimetric data, but significantly decreases the time required to achieve the dosimetric calculations, especially for energies less than 100 keV

  13. Coupling of an aeroacoustic model and a parabolic equation code for long range wind turbine noise propagation

    Science.gov (United States)

    Cotté, B.

    2018-05-01

    This study proposes to couple a source model based on Amiet's theory and a parabolic equation code in order to model wind turbine noise emission and propagation in an inhomogeneous atmosphere. Two broadband noise generation mechanisms are considered, namely trailing edge noise and turbulent inflow noise. The effects of wind shear and atmospheric turbulence are taken into account using the Monin-Obukhov similarity theory. The coupling approach, based on the backpropagation method to preserve the directivity of the aeroacoustic sources, is validated by comparison with an analytical solution for the propagation over a finite impedance ground in a homogeneous atmosphere. The influence of refraction effects is then analyzed for different directions of propagation. The spectrum modification related to the ground effect and the presence of a shadow zone for upwind receivers are emphasized. The validity of the point source approximation that is often used in wind turbine noise propagation models is finally assessed. This approximation exaggerates the interference dips in the spectra, and is not able to correctly predict the amplitude modulation.

  14. Validation of the U.S. NRC coupled code system TRITON/TRACE/PARCS with the special power excursion reactor test III (SPERT III)

    Energy Technology Data Exchange (ETDEWEB)

    Wang, R. C.; Xu, Y.; Downar, T. [Dept. of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Ann Arbor, MI 48104 (United States); Hudson, N. [RES Div., U.S. NRC, Rockville, MD (United States)

    2012-07-01

    The Special Power Excursion Reactor Test III (SPERT III) was a series of reactivity insertion experiments conducted in the 1950's. This paper describes the validation of the U.S. NRC Coupled Code system TRITON/PARCS/TRACE to simulate reactivity insertion accidents (RIA) by using several of the SPERT III tests. The work here used the SPERT III E-core configuration tests in which the RIA was initiated by ejecting a control rod. The resulting super-prompt reactivity excursion and negative reactivity feedback produced the familiar bell shaped power increase and decrease. The energy deposition during such a power peak has important safety consequences and provides validation basis for core coupled multi-physics codes. The transients of five separate tests are used to benchmark the PARCS/TRACE coupled code. The models were thoroughly validated using the original experiment documentation. (authors)

  15. Development of neutronics and thermal hydraulics coupled code – SAC-RIT for plate type fuel and its application to reactivity initiated transient analysis

    International Nuclear Information System (INIS)

    Singh, Tej; Kumar, Jainendra; Mazumdar, Tanay; Raina, V.K.

    2013-01-01

    Highlights: • A point reactor kinetics code coupled with thermal hydraulics of plate type fuel is developed. • This code is applicable for two phase flow of coolant. • Safety analysis of IAEA benchmark reactor core is carried out. • Results agree well with the results available in literature. - Abstract: A point reactor kinetics code SAC-RIT, acronym of Safety Analysis Code for Reactivity Initiated Transient, coupled with thermal hydraulics of two phase coolant flow for plate type fuel, is developed to calculate reactivity initiated transient analysis of nuclear research and test reactors. Point kinetics equations are solved by fourth order Runge Kutta method. Reactivity feedback effect is included into the code. Solution of kinetics equations gives neutronic power and it is then fed into a thermal hydraulic code where mass, momentum and thermal energy conservation equations are solved by explicit finite difference method to find out fuel, clad and coolant temperatures during transients. In this code, all possible flow regimes including laminar flow, transient flow and turbulent flow have been covered. Various heat transfer coefficients suitable for single liquid, sub-cooled boiling, saturation boiling, film boiling and single vapor phases are incorporated in the thermal hydraulic code

  16. International training program: 3D S.UN.COP - Scaling, uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminar

    International Nuclear Information System (INIS)

    Petruzzi, A.; D'Auria, F.; Bajs, T.; Reventos, F.

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP 2005 (Scaling, Uncertainty and 3D COuPled code calculations) seminar has been organized by University of Pisa and University of Zagreb as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). It was recognized that such a course represented both a source of continuing education for current code users and a means for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The seminar-training was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and holding the training and the final examination. A certificate (LA Code User grade) was released

  17. The FLUKA Monte Carlo code coupled with the local effect model for biological calculations in carbon ion therapy

    Energy Technology Data Exchange (ETDEWEB)

    Mairani, A [University of Pavia, Department of Nuclear and Theoretical Physics, and INFN, via Bassi 6, 27100 Pavia (Italy); Brons, S; Parodi, K [Heidelberg Ion Beam Therapy Center and Department of Radiation Oncology, Im Neuenheimer Feld 450, 69120 Heidelberg (Germany); Cerutti, F; Ferrari, A; Sommerer, F [CERN, 1211 Geneva 23 (Switzerland); Fasso, A [SLAC National Accelerator Laboratory, 2575 Sand Hill Road, Menlo Park, CA 94025 (United States); Kraemer, M; Scholz, M, E-mail: Andrea.Mairani@mi.infn.i [GSI Biophysik, Planck-Str. 1, D-64291 Darmstadt (Germany)

    2010-08-07

    Clinical Monte Carlo (MC) calculations for carbon ion therapy have to provide absorbed and RBE-weighted dose. The latter is defined as the product of the dose and the relative biological effectiveness (RBE). At the GSI Helmholtzzentrum fuer Schwerionenforschung as well as at the Heidelberg Ion Therapy Center (HIT), the RBE values are calculated according to the local effect model (LEM). In this paper, we describe the approach followed for coupling the FLUKA MC code with the LEM and its application to dose and RBE-weighted dose calculations for a superimposition of two opposed {sup 12}C ion fields as applied in therapeutic irradiations. The obtained results are compared with the available experimental data of CHO (Chinese hamster ovary) cell survival and the outcomes of the GSI analytical treatment planning code TRiP98. Some discrepancies have been observed between the analytical and MC calculations of absorbed physical dose profiles, which can be explained by the differences between the laterally integrated depth-dose distributions in water used as input basic data in TRiP98 and the FLUKA recalculated ones. On the other hand, taking into account the differences in the physical beam modeling, the FLUKA-based biological calculations of the CHO cell survival profiles are found in good agreement with the experimental data as well with the TRiP98 predictions. The developed approach that combines the MC transport/interaction capability with the same biological model as in the treatment planning system (TPS) will be used at HIT to support validation/improvement of both dose and RBE-weighted dose calculations performed by the analytical TPS.

  18. A novel domain overlapping strategy for the multiscale coupling of CFD with 1D system codes with applications to transient flows

    International Nuclear Information System (INIS)

    Grunloh, T.P.; Manera, A.

    2016-01-01

    Highlights: • A novel domain overlapping coupling method is presented. • Method calculates closure coefficients for system codes based on CFD results. • Convergence and stability are compared with a domain decomposition implementation. • Proposed method is tested in several 1D cases. • Proposed method found to exhibit more favorable convergence and stability behavior. - Abstract: A novel multiscale coupling methodology based on a domain overlapping approach has been developed to couple a computational fluid dynamics code with a best-estimate thermal hydraulic code. The methodology has been implemented in the coupling infrastructure code Janus, developed at the University of Michigan, providing methods for the online data transfer between the commercial computational fluid dynamics code STAR-CCM+ and the US NRC best-estimate thermal hydraulic system code TRACE. Coupling between these two software packages is motivated by the desire to extend the range of applicability of TRACE to scenarios in which local momentum and energy transfer are important, such as three-dimensional mixing. These types of flows are relevant, for example, in the simulation of passive safety systems including large containment pools, or for flow mixing in the reactor pressure vessel downcomer of current light water reactors and integral small modular reactors. The intrafluid shear forces neglected by TRACE equations of motion are readily calculated from computational fluid dynamics solutions. Consequently, the coupling methods used in this study are built around correcting TRACE solutions with data from a corresponding STAR-CCM+ solution. Two coupling strategies are discussed in the paper: one based on a novel domain overlapping approach specifically designed for transient operation, and a second based on the well-known domain decomposition approach. In the present paper, we discuss the application of the two coupling methods to the simulation of open and closed loops in both steady

  19. Accounting for the inertia of the thermocouples' measurements by modelling of a NPP Kalinin-3 transient with the coupled system code ATHLET-BIPR-VVER

    International Nuclear Information System (INIS)

    Nikonov, S.; Velkov, K.

    2008-01-01

    The ATHLET-BIPR-VVER coupled system code is applied for performing of safety analysis for different WWER reactors. During the last years its validation matrix is continuously being enlarged. The measurements performed during the commissioning phase of NPP Kalinin Unit 3 for the transient 'Switching-off of one Main Circulation Pump at nominal power' are very well documented and have a variety of recorded integral and local thermo-hydraulic and neutron-physic parameters including the measurements' errors. This data is being used for further validation of the coupled code system ATHLET-BIPR-VVER. In the paper are discussed the problems and our solutions by the correct interpretation of the measured thermocouples' records at NPP Kalinin-3 and the comparison with the predicted results by the coupled thermal-hydraulic/neutron-kinetic code ATHLET-BIPR-VVER. Of primary importance by such comparisons is the correct accounting of the fluid mixing process that take place in the surrounding of the measuring sensors and also the consideration of the time delay (inertia term) of the measuring devices. On the bases of previous experience and many simulations of the defined transient a method is discussed and proposed to consider correctly the inertia term of the thermocouples' measurements. The new modelling is implemented in the coupled system code ATHLET-BIPR-VVER for further validation. (Author)

  20. Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

    1995-04-01

    The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

  1. Analyses of deformation and thermal-hydraulics within a wire-wrapped fuel subassembly in a liquid metal fast reactor by the coupled code system

    Energy Technology Data Exchange (ETDEWEB)

    Uwaba, Tomoyuki, E-mail: uwaba.tomoyuki@jaea.go.jp; Ohshima, Hiroyuki; Ito, Masahiro

    2017-06-15

    Highlights: • The coupled computational code system allowed for mechanical and thermal-hydraulic analyses in a fast reactor fuel subassembly. • In this system interactive calculations between flow area deformations and coolant temperature changes are repeated to their convergence state. • Effects on bundle-duct interaction on coolant temperature distributions were investigated by using the code system. - Abstract: The coupled numerical analysis of mechanical and thermal-hydraulic behaviors was performed for a wire-wrapped fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal-hydraulic analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that the radial distribution of coolant temperature in the subassembly tended to flatten as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such flattening of temperature distribution was slightly observed as a result of fuel pin bowings due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal-hydraulics was also investigated in this study.

  2. Providing thermal-hydraulic boundary conditions to the reactor code TINTE through a Flownex-TINTE coupling - HTR2008-58110

    International Nuclear Information System (INIS)

    Marais, D.; Greyvenstein, G. P.

    2008-01-01

    TINTE is a well established reactor analysis code which models the transient behaviour of pebble bed reactor cores but it does not include the capabilities to model a power conversion unit (PCU). This raises the issue that TINTE cannot model full system transients. One way to overcome this problem is to supply TINTE with time-dependant thermal-hydraulic boundary conditions which are obtained from PCU simulations. This study investigates a method to provide boundary conditions for the nuclear code TINTE during full system transients. This was accomplished by creating a high level interface between the systems CFD code Flownex and TINTE. An indirect coupling method is explored whereby characteristics of the PCU are matched to characteristics of the nuclear core. This method eliminates the need to iterate between the two codes. A number of transients are simulated using the coupled code and then compared against stand-alone Flownex simulations. The coupling method introduces relatively small errors when reproducing mass flow, temperature and pressure in steady state analysis, but become more pronounced when dealing with fast thermal-hydraulic transients. Decreasing the maximum time step length of TINTE reduces this problem, but increases the computational time. Copyright ASME 2008. (authors)

  3. Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

    Energy Technology Data Exchange (ETDEWEB)

    Samokhin, Aleksei [Scientific and Engineering Centre for Nuclear and Radiation Safety (SEC NRS), Moscow (Russian Federation); Zilly, Matias [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany)

    2016-11-15

    This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK- 1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

  4. Couplings

    Science.gov (United States)

    Stošić, Dušan; Auroux, Aline

    Basic principles of calorimetry coupled with other techniques are introduced. These methods are used in heterogeneous catalysis for characterization of acidic, basic and red-ox properties of solid catalysts. Estimation of these features is achieved by monitoring the interaction of various probe molecules with the surface of such materials. Overview of gas phase, as well as liquid phase techniques is given. Special attention is devoted to coupled calorimetry-volumetry method. Furthermore, the influence of different experimental parameters on the results of these techniques is discussed, since it is known that they can significantly influence the evaluation of catalytic properties of investigated materials.

  5. Analysis of the OECD/NRC BWR Turbine Trip Transient Benchmark with the Coupled Thermal-Hydraulics and Neutronics Code TRAC-M/PARCS

    International Nuclear Information System (INIS)

    Lee, Deokjung; Downar, Thomas J.; Ulses, Anthony; Akdeniz, Bedirhan; Ivanov, Kostadin N.

    2004-01-01

    An analysis of the Peach Bottom Unit 2 Turbine Trip 2 (TT2) experiment has been performed using the U.S. Nuclear Regulatory Commission coupled thermal-hydraulics and neutronics code TRAC-M/PARCS. The objective of the analysis was to assess the performance of TRAC-M/PARCS on a BWR transient with significance in two-phase flow and spatial variations of the neutron flux. TRAC-M/PARCS results are found to be in good agreement with measured plant data for both steady-state and transient phases of the benchmark. Additional analyses of four fictitious extreme scenarios are performed to provide a basis for code-to-code comparisons and comprehensive testing of the thermal-hydraulics/neutronics coupling. The obtained results of sensitivity studies on the effect of direct moderator heating on transient simulation indicate the importance of this modeling aspect

  6. Domain decomposition with local refinement for flow simulation around a nuclear waste disposal site: direct computation versus simulation using code coupling with OCamlP3L

    Energy Technology Data Exchange (ETDEWEB)

    Clement, F.; Vodicka, A.; Weis, P. [Institut National de Recherches Agronomiques (INRA), 78 - Le Chesnay (France); Martin, V. [Institut National de Recherches Agronomiques (INRA), 92 - Chetenay Malabry (France); Di Cosmo, R. [Institut National de Recherches Agronomiques (INRA), 78 - Le Chesnay (France); Paris-7 Univ., 75 (France)

    2003-07-01

    We consider the application of a non-overlapping domain decomposition method with non-matching grids based on Robin interface conditions to the problem of flow surrounding an underground nuclear waste disposal. We show with a simple example how one can refine the mesh locally around the storage with this technique. A second aspect is studied in this paper. The coupling between the sub-domains can be achieved by computing in two ways: either directly (i.e. the domain decomposition algorithm is included in the code that solves the problems on the sub-domains) or using code coupling. In the latter case, each sub-domain problem is solved separately and the coupling is performed by another program. We wrote a coupling program in the functional language Ocaml, using the OcamIP31 environment devoted to ease the parallelism. This at the same time we test the code coupling and we use the natural parallel property of domain decomposition methods. Some simple 2D numerical tests show promising results, and further studies are under way. (authors)

  7. Domain decomposition with local refinement for flow simulation around a nuclear waste disposal site: direct computation versus simulation using code coupling with OCamlP3L

    International Nuclear Information System (INIS)

    Clement, F.; Vodicka, A.; Weis, P.; Martin, V.; Di Cosmo, R.

    2003-01-01

    We consider the application of a non-overlapping domain decomposition method with non-matching grids based on Robin interface conditions to the problem of flow surrounding an underground nuclear waste disposal. We show with a simple example how one can refine the mesh locally around the storage with this technique. A second aspect is studied in this paper. The coupling between the sub-domains can be achieved by computing in two ways: either directly (i.e. the domain decomposition algorithm is included in the code that solves the problems on the sub-domains) or using code coupling. In the latter case, each sub-domain problem is solved separately and the coupling is performed by another program. We wrote a coupling program in the functional language Ocaml, using the OcamIP31 environment devoted to ease the parallelism. This at the same time we test the code coupling and we use the natural parallel property of domain decomposition methods. Some simple 2D numerical tests show promising results, and further studies are under way. (authors)

  8. Development of finite element code for the analysis of coupled thermo-hydro-mechanical behaviors of a saturated-unsaturated medium

    International Nuclear Information System (INIS)

    Ohnishi, Y.; Shibata, H.; Kobsayashi, A.

    1987-01-01

    A model is presented which describes fully coupled thermo-hydro-mechanical behavior of a porous geologic medium. The mathematical formulation for the model utilizes the Biot theory for the consolidation and the energy balance equation. If the medium is in the condition of saturated-unsaturated flow, then the free surfaces are taken into consideration in the model. The model, incorporated in a finite element numerical procedure, was implemented in a two-dimensional computer code. The code was developed under the assumptions that the medium is poro-elastic and in the plane strain condition; that water in the ground does not change its phase; and that heat is transferred by conductive and convective flow. Analytical solutions pertaining to consolidation theory for soils and rocks, thermoelasticity for solids and hydrothermal convection theory provided verification of stress and fluid flow couplings, respectively, in the coupled model. Several types of problems are analyzed

  9. Comparison of 'system thermal-hydraulics-3 dimensional reactor kinetics' coupled calculations using the MARS 1D and 3D modules and the MASTER code

    International Nuclear Information System (INIS)

    Jung, J. J.; Joo, H. K.; Lee, W. J.; Ji, S. K.; Jung, B. D.

    2002-01-01

    KAERI has developed the coupled 'system thermal-hydraulics - 3 dimensional reactor kinetics' code, MARS/MASTER since 1998. However, there is a limitation in the existing MARS/MASTER code; that is, to perform the coupled calculations using MARS/MASTER, we have to utilize the hydrodynamic model and the heat structure model of the MARS '3D module'. In some transients, reactor kinetics behavior is strongly multi-dimensional, but core thermal-hydraulic behavior remains in one-dimensional manner. For efficient analysis of such transients, we coupled the MARS 1D module with MASTER. The new feature has been assessed by the 'OECD NEA Main Steam Line Break (MSLB) benchmark exercise III' simulations

  10. Biological dose estimation for charged-particle therapy using an improved PHITS code coupled with a microdosimetric kinetic model

    International Nuclear Information System (INIS)

    Sato, Tatsuhiko; Watanabe, Ritsuko; Kase, Yuki; Niita, Koji; Sihver, Lembit

    2009-01-01

    -particle therapy was established using the improved PHITS coupled with a microdosimetric kinetic (MK) model. The detailed procedures for improving the PHITS code and establishing the biological-dose-estimation method were presented elsewhere. Hence, this report focuses on describing their capabilities to optimize the treatment planning of charged particle therapy, thereby maximizing the therapeutic effect on tumor while minimizing unintended harmful effects on surrounding normal tissues. (author)

  11. International training program in support of safety analysis. 3D S.UN.COP-scaling uncertainty and 3D thermal-hydraulics/neutron-kinetics coupled codes seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc; Hassan, Yassin

    2007-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysis to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users. Six seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005), at the School of Industrial Engineering of Barcelona (January-February 2006) and in Buenos Aires, Argentina (October 2006), being this last one requested by ARN (Autoridad Regulatoria Nuclear), NA-SA (Nucleoelectrica Argentina S.A) and CNEA (Comision Nacional de Energia Atomica). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 in Barcelona was successfully held with the attendance of 33

  12. International Training Program in Support of Safety Analysis: 3D S.UN.COP-Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminars

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Bajs, Tomislav; Reventos, Francesc

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users [1]. Five seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004), at University of Zagreb (2005) and at the School of Industrial Engineering of Barcelona (2006). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2006 was successfully held with the attendance of 33 participants coming from 18 countries and 28 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 30 scientists (coming from 13 countries and 23 different institutions) were

  13. Coupling an analytical description of anti-scatter grids with simulation software of radiographic systems using Monte Carlo code

    International Nuclear Information System (INIS)

    Rinkel, J.; Dinten, J.M.; Tabary, J.

    2004-01-01

    The use of focused anti-scatter grids on digital radiographic systems with two-dimensional detectors produces acquisitions with a decreased scatter to primary ratio and thus improved contrast and resolution. Simulation software is of great interest in optimizing grid configuration according to a specific application. Classical simulators are based on complete detailed geometric descriptions of the grid. They are accurate but very time consuming since they use Monte Carlo code to simulate scatter within the high-frequency grids. We propose a new practical method which couples an analytical simulation of the grid interaction with a radiographic system simulation program. First, a two dimensional matrix of probability depending on the grid is created offline, in which the first dimension represents the angle of impact with respect to the normal to the grid lines and the other the energy of the photon. This matrix of probability is then used by the Monte Carlo simulation software in order to provide the final scattered flux image. To evaluate the gain of CPU time, we define the increasing factor as the increase of CPU time of the simulation with as opposed to without the grid. Increasing factors were calculated with the new model and with classical methods representing the grid with its CAD model as part of the object. With the new method, increasing factors are shorter by one to two orders of magnitude compared with the second one. These results were obtained with a difference in calculated scatter of less than five percent between the new and the classical method. (authors)

  14. Coupling analysis of deformation and thermal-hydraulics in a FBR fuel pin bundle using BAMBOO and ASFRE-IV Codes

    International Nuclear Information System (INIS)

    Ito, Masahiro; Imai, Yasutomo; Uwaba, Tomoyuki; Ohshima, Hiroyuki

    2004-03-01

    The bundle-duct interaction may occur in sodium cooled wire-wrapped FBR fuel subassemblies in high burn-up conditions. JNC has been developing a bundle deformation analysis code BAMBOO (Behavior Analysis code for Mechanical interaction of fuel Bundle under On-power Operation), a thermal hydraulics analysis code ASFRE-IV (Analysis of Sodium Flow in Reactor Elements - ver. IV) and their coupling method as a simulation system for the evaluation on the integrity of deformed FBR fuel pin bundles. In this study, the simulation system was applied to a coupling analysis of deformation and thermal-hydraulics in the fuel pin-bundle under a steady-state condition just after startup for the purpose of the verification of the simulation system. The iterative calculations of deformation and thermal-hydraulics employed in the coupling analysis provided numerically unstable solutions. From the result, it was found that improvement of the coupling algorithm of BAMBOO and ASFRE-IV is necessary to reduce numerical fluctuations and to obtain better convergence by introducing such computational technique as the optimized under-relaxation method. (author)

  15. GEYSER/TONUS: A coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Petit, M.; Durin, M.; Gauvain, J. [Commissariat a l`Energie Atomique, Gif sur Yvette (France)

    1995-09-01

    In many countries, the safety requirements for future light water reactors include accounting for severe accidents in the design process. As far as the containment is concerned, the design must now include mitigation features to limit the pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. In this context, new needs appear for the modeling of the thermal hydraulics inside the containment. It requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. Moreover, the effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled, as for example hydrogen stratification and condensation. To model such a complex situation, the use of multi-dimensional computer codes seems to be necessary in case of large volumes. The aim of the GEYSER/TONUS computer code is to fulfill this need. This code is currently under development at CEA in Saclay. It will allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, as the objective is to be able to treat complete scenarios. Physical models of classical lumped parameters codes will adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows, thanks to its modular conception, to construct sophisticated applications based upon it.

  16. Statistical safety evaluation of BWR turbine trip scenario using coupled neutron kinetics and thermal hydraulics analysis code SKETCH-INS/TRACE5.0

    International Nuclear Information System (INIS)

    Ichikawa, Ryoko; Masuhara, Yasuhiro; Kasahara, Fumio

    2012-01-01

    The Best Estimate Plus Uncertainty (BEPU) method has been prepared for the regulatory cross-check analysis at Japan Nuclear Energy Safety Organization (JNES) on base of the three-dimensional neutron-kinetics/thermal-hydraulics coupled code SKETCH-INS/TRACE5.0. In the preparation, TRACE5.0 is verified against the large-scale thermal-hydraulic tests carried out with NUPEC facility. These tests were focused on the pressure drop of steam-liquid two phase flow and void fraction distribution. From the comparison of the experimental data with other codes (RELAP5/MOD3.3 and TRAC-BF1), TRACE5.0 was judged better than other codes. It was confirmed that TRACE5.0 has high reliability for thermal hydraulics behavior and are used as a best-estimate code for the statistical safety evaluation. Next, the coupled code SKETCH-INS/TRACE5.0 was applied to turbine trip tests performed at the Peach Bottom-2 BWR4 Plant. The turbine trip event shows the rapid power peak due to the voids collapse with the pressure increase. The analyzed peak value of core power is better simulated than the previous version SKETCH-INS/TRAC-BF1. And the statistical safety evaluation using SKETCH-INS/TRACE5.0 was applied to the loss of load transient for examining the influence of the choice of sampling method. (author)

  17. International Training Program: 3D S. Un. Cop - Scaling, Uncertainty and 3D Thermal-Hydraulics/Neutron-Kinetics Coupled Codes Seminar

    International Nuclear Information System (INIS)

    Pertuzzi, A.; D'Auria, F.; Bajs, T.; Reventos, F.

    2006-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers and vendors, nuclear fuel companies, research organizations, consulting companies, and technical support organizations. The computer code user represents a source of uncertainty that can influence the results of system code calculations. This influence is commonly known as the 'user effect' and stems from the limitations embedded in the codes as well as from the limited capability of the analysts to use the codes. Code user training and qualification is an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In other words, the program aims at contributing towards solving the problem of user effect. The 3D S.UN.COP (Scaling, Uncertainty and 3D COuPled code calculations) seminars have been organized as follow-up of the proposal to IAEA for the Permanent Training Course for System Code Users (D'Auria, 1998). Four seminars have been held at University of Pisa (2003, 2004), at The Pennsylvania State University (2004) and at University of Zagreb (2005). It was recognized that such courses represented both a source of continuing education for current code users and a mean for current code users to enter the formal training structure of a proposed 'permanent' stepwise approach to user training. The 3D S.UN.COP 2005 was successfully held with the participation of 19 persons coming from 9 countries and 14 different institutions (universities, vendors, national laboratories and regulatory bodies). More than 15 scientists were involved in the organization of the seminar, presenting theoretical aspects of the proposed methodologies and

  18. Application of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the OECD/NRC BWR turbine trip benchmark and its performance on multi-processor computers

    International Nuclear Information System (INIS)

    Langenbuch, S.; Schmidt, K.D.; Velkov, K.

    2003-01-01

    The OECD/NRC BWR Turbine Trip (TT) Benchmark is investigated to perform code-to-code comparison of coupled codes including a comparison to measured data which are available from turbine trip experiments at Peach Bottom 2. This Benchmark problem for a BWR over-pressure transient represents a challenging application of coupled codes which integrate 3-dimensional neutron kinetics into thermal-hydraulic system codes for best-estimate simulation of plant transients. This transient represents a typical application of coupled codes which are usually performed on powerful workstations using a single CPU. Nowadays, the availability of multi-CPUs is much easier. Indeed, powerful workstations already provide 4 to 8 CPU, computer centers give access to multi-processor systems with numbers of CPUs in the order of 16 up to several 100. Therefore, the performance of the coupled code Athlet-Quabox/Cubbox on multi-processor systems is studied. Different cases of application lead to changing requirements of the code efficiency, because the amount of computer time spent in different parts of the code is varying. This paper presents main results of the coupled code Athlet-Quabox/Cubbox for the extreme scenarios of the BWR TT Benchmark together with evaluations of the code performance on multi-processor computers. (authors)

  19. Development of the coupled 'system thermal-hydraulics, 3D reactor kinetics, and hot channel' analysis capability of the MARS code

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, J. J.; Chung, B. D.; Lee, W.J

    2005-02-01

    The subchannel analysis capability of the MARS 3D module has been improved. Especially, the turbulent mixing and void drift models for flow mixing phenomena in rod bundles have been assessed using some well-known rod bundle test data. Then, the subchannel analysis feature was combined to the existing coupled 'system Thermal-Hydraulics (T/H) and 3D reactor kinetics' calculation capability of MARS. These features allow the coupled 'system T/H, 3D reactor kinetics, and hot channel' analysis capability and, thus, realistic simulations of hot channel behavior as well as global system T/H behavior. In this report, the MARS code features for the coupled analysis capability are described first. The code modifications relevant to the features are also given. Then, a coupled analysis of the Main Steam Line Break (MSLB) is carried out for demonstration. The results of the coupled calculations are very reasonable and realistic, and show these methods can be used to reduce the over-conservatism in the conventional safety analysis.

  20. Development of sub-channel code SACoS and its application in coupled neutronics/thermal hydraulics system for SCWR

    International Nuclear Information System (INIS)

    Chaudri, Khurrum Saleem; Su Yali; Chen Ronghua; Tian Wenxi; Su Guanghui; Qiu Suizheng

    2012-01-01

    Highlights: ► A tool is developed for coupled neutronics/thermal-hydraulic analysis for SCWR. ► For thermal hydraulic analysis, a sub-channel code SACoS is developed and verified. ► Coupled analysis agree quite well with the reference calculations. ► Different choice of important parameters makes huge difference in design calculations. - Abstract: Supercritical Water Reactor (SCWR) is one of the promising reactors from the list of fourth generation of nuclear reactors. High thermal efficiency and low cost of electricity make it an attractive option in the era of growing energy demand. An almost seven fold density variation for coolant/moderator along the active height does not allow the use of constant density assumption for design calculations, as used for previous generations of reactors. The advancement in computer technology gives us the superior option of performing coupled analysis. Thermal hydraulics calculations of supercritical water systems present extra challenges as not many computational tools are available to perform that job. This paper introduces a new sub-channel code called Sub-channel Analysis Code of SCWR (SACoS) and its application in coupled analyses of High Performance Light Water Reactor (HPLWR). SACoS can compute the basic thermal hydraulic parameters needed for design studies of a supercritical water reactor. Multiple heat transfer and pressure drop correlations are incorporated in the code according to the flow regime. It has the additional capability of calculating the thermal hydraulic parameters of moderator flowing in water box and between fuel assemblies under co-current or counter current flow conditions. Using MCNP4c and SACoS, a coupled system has been developed for SCWR design analyses. The developed coupled system is verified by performing and comparing HPLWR calculations. The results were found to be in very good agreement. Significant difference between the results was seen when Doppler feedback effect was included in

  1. FLUST-2D - A computer code for the calculation of the two-dimensional flow of a compressible medium in coupled retangular areas

    International Nuclear Information System (INIS)

    Enderle, G.

    1979-01-01

    The computer-code FLUST-2D is able to calculate the two-dimensional flow of a compressible fluid in arbitrary coupled rectangular areas. In a finite-difference scheme the program computes pressure, density, internal energy and velocity. Starting with a basic set of equations, the difference equations in a rectangular grid are developed. The computational cycle for coupled fluid areas is described. Results of test calculations are compared to analytical solutions and the influence of time step and mesh size are investigated. The program was used to precalculate the blowdown experiments of the HDR experimental program. Downcomer, plena, internal vessel region, blowdown pipe and a containment area have been modelled two-dimensionally. The major results of the precalculations are presented. This report also contains a description of the code structure and user information. (orig.) [de

  2. Effects of Secondary Circuit Modeling on Results of Pressurized Water Reactor Main Steam Line Break Benchmark Calculations with New Coupled Code TRAB-3D/SMABRE

    International Nuclear Information System (INIS)

    Daavittila, Antti; Haemaelaeinen, Anitta; Kyrki-Rajamaeki, Riitta

    2003-01-01

    All of the three exercises of the Organization for Economic Cooperation and Development/Nuclear Regulatory Commission pressurized water reactor main steam line break (PWR MSLB) benchmark were calculated at VTT, the Technical Research Centre of Finland. For the first exercise, the plant simulation with point-kinetic neutronics, the thermal-hydraulics code SMABRE was used. The second exercise was calculated with the three-dimensional reactor dynamics code TRAB-3D, and the third exercise with the combination TRAB-3D/SMABRE. VTT has over ten years' experience of coupling neutronic and thermal-hydraulic codes, but this benchmark was the first time these two codes, both developed at VTT, were coupled together. The coupled code system is fast and efficient; the total computation time of the 100-s transient in the third exercise was 16 min on a modern UNIX workstation. The results of all the exercises are similar to those of the other participants. In order to demonstrate the effect of secondary circuit modeling on the results, three different cases were calculated. In case 1 there is no phase separation in the steam lines and no flow reversal in the aspirator. In case 2 the flow reversal in the aspirator is allowed, but there is no phase separation in the steam lines. Finally, in case 3 the drift-flux model is used for the phase separation in the steam lines, but the aspirator flow reversal is not allowed. With these two modeling variations, it is possible to cover a remarkably broad range of results. The maximum power level reached after the reactor trip varies from 534 to 904 MW, the range of the time of the power maximum being close to 30 s. Compared to the total calculated transient time of 100 s, the effect of the secondary side modeling is extremely important

  3. Experimental studies and modelling of cation interactions with solid materials: application to the MIMICC project. (Multidimensional Instrumented Module for Investigations on chemistry-transport Coupled Codes)

    International Nuclear Information System (INIS)

    Hardin, Emmanuelle

    1999-01-01

    The study of cation interactions with solid materials is useful in order to define the chemistry interaction component of the MIMICC project (Multidimensional Instrumented Module for Investigations on chemistry-transport Coupled Codes). This project will validate the chemistry-transport coupled codes. Database have to be supplied on the cesium or ytterbium interactions with solid materials in suspension. The solid materials are: a strong cation exchange resin, a natural sand which presents small impurities, and a zirconium phosphate. The cation exchange resin is useful to check that the surface complexation theory can be applied on a pure cation exchanger. The sand is a natural material, and its isotherms will be interpreted using pure oxide-cation system data, such as pure silica-cation data. Then the study on the zirconium phosphate salt is interesting because of the increasing complexity in the processes (dissolution, sorption and co-precipitation). These data will enable to approach natural systems, constituted by several complex solids which can interfere on each other. These data can also be used for chemistry-transport coupled codes. Potentiometric titration, sorption isotherms, sorption kinetics, cation surface saturation curves are made, in order to obtain the different parameters relevant to the cation sorption at the solid surface, for each solid-electrolyte-cation system. The influence of different parameters such as ionic strength, pH, and electrolyte is estimated. All the experimental curves are fitted with FITEQL code based on the surface complexation theory using the constant capacitance model, in order to give a mechanistic interpretation of the ion retention phenomenon at the solid surface. The speciation curves of all systems are plotted, using the FITEQL code too. Systems with an increasing complexity are studied: dissolution, sorption and coprecipitation coexist in the cation-salt systems. Then the data obtained on each single solid, considered

  4. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark - a consistent approach for assessing coupled codes for RIA analysis

    International Nuclear Information System (INIS)

    Boyan D Ivanov; Kostadin N Ivanov; Eric Royer; Sylvie Aniel; Nikola Kolev; Pavlin Groudev

    2005-01-01

    Full text of publication follows: The Rod Ejection Accident (REA) and Main Steam Line Break (MSLB) are two of the most important Design Basis Accidents (DBA) for VVER-1000 exhibiting significant localized space-time effects. A consistent approach for assessing coupled three-dimensional (3-D) neutron kinetics/thermal hydraulics codes for these Reactivity Insertion Accidents (RIA) is to first validate the codes using the available plant test (measured) data and after that perform cross code comparative analysis for REA and MSLB scenarios. In the framework of joint effort between the Nuclear Energy Agency (NEA) of OECD, the United States Department of Energy (US DOE), and the Commissariat a l'Energie Atomique (CEA), France a coupled 3-D neutron kinetics/thermal hydraulics benchmark was defined. The benchmark is based on data from the Unit 6 of the Bulgarian Kozloduy Nuclear Power Plant (NPP). In performing this work the PSU, USA and CEA-Saclay, France have collaborated with Bulgarian organizations, in particular with the KNPP and the INRNE. The benchmark consists of two phases: Phase 1: Main Coolant Pump Switching On; Phase 2: Coolant Mixing Tests and MSLB. In addition to the measured (experiment) scenario, an extreme calculation scenario was defined for better testing 3-D neutronics/thermal-hydraulics techniques: rod ejection simulation with control rod being ejected in the core sector cooled by the switched on MCP. Since the previous coupled code benchmarks indicated that further development of the mixing computation models in the integrated codes is necessary, a coolant mixing experiment and MSLB transients are selected for simulation in Phase 2 of the benchmark. The MSLB event is characterized by a large asymmetric cooling of the core, stuck rods and a large primary coolant flow variation. Two scenarios are defined in Phase 2: the first scenario is taken from the current licensing practice and the second one is derived from the original one using aggravating

  5. Application of the coupled code system KIKO3D/ATHLET to the boron dilution transients in VVER-440 Type NPP

    International Nuclear Information System (INIS)

    Hegyi, G.; Kereszturi, A.; Trosztel, I.

    2003-01-01

    The transient caused by a perturbation of boron concentration and coolant temperature at the inlet of a Russian developed reactor (VVER-440) is reanalysed as a part of the modernisation (introduction of a new type profiled fuel) and power upgrading (up to 108 %) project. This task is one of the basis cases to be investigated in the safety analysis of the pressurized water reactor (PWR) and the criteria of Anticipated Operational Occurrences (AOO) have to be fulfilled for it. First detailed planning calculations were performed with the thermal hydraulic system code ATHLET and neutron physical code system KARATE-440 to find out the appropriate initial parameter set taking into account the active safety system of the NPP. Finally the most reactive case is analysed by the KIKO3D/ATHLET coupled system code. Whereas the investigation is done for safety analysis, conservative assumptions are imposed on reactivity characteristics. Moreover at the core inlet no-mixing is supposed from the unaffected loops. The presented calculations show, how the coupled code system with a detailed description of plant functions and core behaviour can help to understand better the local phenomena in the study of a potential risk of dilution accident, as it offers the possibility to evaluate the plant safety in a more realistic and versatile manner. (author)

  6. ITS version 5.0 :the integrated TIGER series of coupled electron/Photon monte carlo transport codes with CAD geometry.

    Energy Technology Data Exchange (ETDEWEB)

    Franke, Brian Claude; Kensek, Ronald Patrick; Laub, Thomas William

    2005-09-01

    ITS is a powerful and user-friendly software package permitting state-of-the-art Monte Carlo solution of linear time-independent coupled electron/photon radiation transport problems, with or without the presence of macroscopic electric and magnetic fields of arbitrary spatial dependence. Our goal has been to simultaneously maximize operational simplicity and physical accuracy. Through a set of preprocessor directives, the user selects one of the many ITS codes. The ease with which the makefile system is applied combines with an input scheme based on order-independent descriptive keywords that makes maximum use of defaults and internal error checking to provide experimentalists and theorists alike with a method for the routine but rigorous solution of sophisticated radiation transport problems. Physical rigor is provided by employing accurate cross sections, sampling distributions, and physical models for describing the production and transport of the electron/photon cascade from 1.0 GeV down to 1.0 keV. The availability of source code permits the more sophisticated user to tailor the codes to specific applications and to extend the capabilities of the codes to more complex applications. Version 5.0, the latest version of ITS, contains (1) improvements to the ITS 3.0 continuous-energy codes, (2) multigroup codes with adjoint transport capabilities, (3) parallel implementations of all ITS codes, (4) a general purpose geometry engine for linking with CAD or other geometry formats, and (5) the Cholla facet geometry library. Moreover, the general user friendliness of the software has been enhanced through increased internal error checking and improved code portability.

  7. OECD/NRC BWR Turbine Trip Transient Benchmark as a Basis for Comprehensive Qualification and Studying Best-Estimate Coupled Codes

    International Nuclear Information System (INIS)

    Ivanov, Kostadin; Olson, Andy; Sartori, Enrico

    2004-01-01

    An Organisation for Economic Co-operation and Development (OECD)/U.S. Nuclear Regulatory Commission (NRC)-sponsored coupled-code benchmark has been initiated for a boiling water reactor (BWR) turbine trip (TT) transient. Turbine trip transients in a BWR are pressurization events in which the coupling between core space-dependent neutronic phenomena and system dynamics plays an important role. In addition, the available real plant experimental data make this benchmark problem very valuable. Over the course of defining and coordinating the BWR TT benchmark, a systematic approach has been established to validate best-estimate coupled codes. This approach employs a multilevel methodology that not only allows for a consistent and comprehensive validation process but also contributes to the study of different numerical and computational aspects of coupled best-estimate simulations. This paper provides an overview of the OECD/NRC BWR TT benchmark activities with emphasis on the discussion of the numerical and computational aspects of the benchmark

  8. PRELIMINARY COUPLING OF THE MONTE CARLO CODE OPENMC AND THE MULTIPHYSICS OBJECT-ORIENTED SIMULATION ENVIRONMENT (MOOSE) FOR ANALYZING DOPPLER FEEDBACK IN MONTE CARLO SIMULATIONS

    Energy Technology Data Exchange (ETDEWEB)

    Matthew Ellis; Derek Gaston; Benoit Forget; Kord Smith

    2011-07-01

    In recent years the use of Monte Carlo methods for modeling reactors has become feasible due to the increasing availability of massively parallel computer systems. One of the primary challenges yet to be fully resolved, however, is the efficient and accurate inclusion of multiphysics feedback in Monte Carlo simulations. The research in this paper presents a preliminary coupling of the open source Monte Carlo code OpenMC with the open source Multiphysics Object-Oriented Simulation Environment (MOOSE). The coupling of OpenMC and MOOSE will be used to investigate efficient and accurate numerical methods needed to include multiphysics feedback in Monte Carlo codes. An investigation into the sensitivity of Doppler feedback to fuel temperature approximations using a two dimensional 17x17 PWR fuel assembly is presented in this paper. The results show a functioning multiphysics coupling between OpenMC and MOOSE. The coupling utilizes Functional Expansion Tallies to accurately and efficiently transfer pin power distributions tallied in OpenMC to unstructured finite element meshes used in MOOSE. The two dimensional PWR fuel assembly case also demonstrates that for a simplified model the pin-by-pin doppler feedback can be adequately replicated by scaling a representative pin based on pin relative powers.

  9. The CNCSN-2: One, two-and three-dimensional coupled neutral and charged particle discrete ordinates code system

    International Nuclear Information System (INIS)

    Voloschenko, A. M.; Gukov, S. V.; Russkov, A. A.; Gurevich, M. I.; Shkarovsky, D. A.; Kryuchkov, V. P.; Sumaneev, O. V.; Dubinin, A. A.

    2009-01-01

    KATRIN, KASKAD-Sand ROZ-6 codes solve the multigroup transport equation for neutrons, photons and charged particles in 3D. BOT3P-5., ConDat can be used as preprocessor. ARVES-2.5, a cross-section preprocessor (the package of utilities for operating with the cross section file in FMAC-M format) is included. Auxiliary codes MIXERM, CEPXS-BFP, CEPXS-BFP, SADCO-2.4 and CNCSN-2 are used

  10. RADHEAT-V3, a code system for generating coupled neutron and gamma-ray group constants and analyzing radiation transport

    International Nuclear Information System (INIS)

    Koyama, Kinji; Taji, Yukichi; Miyasaka, Shun-ichi; Minami, Kazuyoshi.

    1977-07-01

    The modular code system RADHEAT is for producing coupled multigroup neutron and gamma-ray cross section sets, analyzing the neutron and gamma-ray transport, and calculating the energy deposition and atomic displacements due to these radiations in a nuclear reactor or shield. The basic neutron cross sections and secondary gamma-ray production data are taken from ENDF/B and POPOP4 libraries respectively. The system (1) generates multigroup neutron cross sections, energy deposition coefficients and atomic displacement factors due to neutron reactions, (2) generates multigroup gamma-ray cross sections and energy transfer coefficients, (3) generates secondary gamma-ray production cross sections, (4) combines these cross sections into the coupled set, (5) outputs and updates the multigroup cross section libraries in convenient formats for other transport codes, (6) analyzes the neutron and gamma-ray transport and calculates the energy deposition and the number density of atomic displacements in a medium, (7) collapses the cross sections to a broad-group structure, by option, using the weighting functions obtained by one-dimensional transport calculation, and (8) plots, by option, multigroup cross sections, and neutron and gamma-ray distributions. Definitions of the input data required in various options of the code system are also given. (auth.)

  11. GEYSER/TONUS: a coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Petit, M.; Durin, M.; Gauvain, J.

    1995-12-31

    The safety requirements for future light water reactors include accounting for severe accidents in the design process. The design must now include mitigation features to limit pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. Modelling the thermal hydraulics inside the containment requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. The effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled. The GEYSER/TONUS multi-dimensional computer code is under development at CEA Saclay to model this complex situation. It allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, and physical models of classical lumped parameters codes will be adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows to construct sophisticated applications based upon it. (J.S.). 22 refs., 1 fig.

  12. GEYSER/TONUS: a coupled multi-D lumped parameter code for reactor thermal hydraulics analysis in case of severe accidents

    International Nuclear Information System (INIS)

    Petit, M.; Durin, M.; Gauvain, J.

    1995-01-01

    The safety requirements for future light water reactors include accounting for severe accidents in the design process. The design must now include mitigation features to limit pressure and temperature inside the building. Hydrogen concentration is also a major issue for severe accidents. Modelling the thermal hydraulics inside the containment requires the description of complex phenomena such as condensation, stratification, transport of gases and aerosols, heat transfers. The effect of mitigation systems will increase the heterogeneities in the building, and most of those phenomena can be coupled. The GEYSER/TONUS multi-dimensional computer code is under development at CEA Saclay to model this complex situation. It allow the coupling of parts of the containment described in a lumped parameter manner, together with meshed parts. Emphasis is put on the numerical methods used to solve the transient problem, and physical models of classical lumped parameters codes will be adapted for the spatially described zones. The code is developed in the environment of the CASTEM 2000/TRIO EF system which allows to construct sophisticated applications based upon it. (J.S.). 22 refs., 1 fig

  13. TORT-TD/ATTICA3D: a coupled neutron transport and thermal hydraulics code system for 3-D transient analysis of gas cooled high temperature reactors

    International Nuclear Information System (INIS)

    Lapins, J.; Seubert, A.; Buck, M.; Bader, J.; Laurien, E.

    2011-01-01

    Comprehensive safety studies of high temperature gas cooled reactors (HTR) require full three dimensional coupled treatments of both neutron kinetics and thermal-hydraulics. In a common effort, GRS and IKE developed the coupled code system TORT-TD/ATTICA3D for pebble bed type HTR that connects the 3-D transient discrete-ordinates transport code TORT-TD with the 3-D porous medium thermal-hydraulics code ATTICA3D. In this paper, the physical models and calculation capabilities of TORT-TD and ATTICA3D are presented, focusing on model improvements in ATTICA3D and extensions made in TORT-TD related to HTR application. For first applications, the OECD/NEA/NSC PBMR-400 benchmark has been chosen. Results obtained with TORT-TD/ATTICA3D will be shown for transient exercises, e.g. control rod withdrawal and a control rod ejection. Results are compared to other benchmark participants' solutions with special focus on fuel temperature modelling features of ATTICA3D. The provided “grey-curtain” nuclear cross section libraries have been used. First results on 3-D effects during a control rod withdrawal transient will be presented. (author)

  14. Analysis of PWR control rod ejection accident with the coupled code system SKETCH-INS/TRACE by incorporating pin power reconstruction model

    International Nuclear Information System (INIS)

    Nakajima, T.; Sakai, T.

    2010-01-01

    The pin power reconstruction model was incorporated in the 3-D nodal kinetics code SKETCH-INS in order to produce accurate calculation of three-dimensional pin power distributions throughout the reactor core. In order to verify the employed pin power reconstruction model, the PWR MOX/UO_2 core transient benchmark problem was analyzed with the coupled code system SKETCH-INS/TRACE by incorporating the model and the influence of pin power reconstruction model was studied. SKETCH-INS pin power distributions for 3 benchmark problems were compared with the PARCS solutions which were provided by the host organisation of the benchmark. SKETCH-INS results were in good agreement with the PARCS results. The capability of employed pin power reconstruction model was confirmed through the analysis of benchmark problems. A PWR control rod ejection benchmark problem was analyzed with the coupled code system SKETCH-INS/ TRACE by incorporating the pin power reconstruction model. The influence of pin power reconstruction model was studied by comparing with the result of conventional node averaged flux model. The results indicate that the pin power reconstruction model has significant effect on the pin powers during transient and hence on the fuel enthalpy

  15. Neutronic / thermal-hydraulic coupling with the code system Trace / Parcs; Acoplamiento neutronico / termohidraulico con el sistema de codigos TRACE / PARCS

    Energy Technology Data Exchange (ETDEWEB)

    Mejia S, D. M. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col. Narvarte, 03020 Ciudad de Mexico (Mexico); Del Valle G, E., E-mail: dulcemaria.mejia@cnsns.gob.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. Lindavista, 07738 Ciudad de Mexico (Mexico)

    2015-09-15

    The developed models for Parcs and Trace codes corresponding for the cycle 15 of the Unit 1 of the Laguna Verde nuclear power plant are described. The first focused to the neutronic simulation and the second to thermal hydraulics. The model developed for Parcs consists of a core of 444 fuel assemblies wrapped in a radial reflective layer and two layers, a superior and another inferior, of axial reflector. The core consists of 27 total axial planes. The model for Trace includes the vessel and its internal components as well as various safety systems. The coupling between the two codes is through two maps that allow its intercommunication. Both codes are used in coupled form performing a dynamic simulation that allows obtaining acceptably a stable state from which is carried out the closure of all the main steam isolation valves (MSIVs) followed by the performance of safety relief valves (SRVs) and ECCS. The results for the power and reactivities introduced by the moderator density, the fuel temperature and total temperature are shown. Data are also provided like: the behavior of the pressure in the steam dome, the water level in the downcomer, the flow through the MSIVs and SRVs. The results are explained for the power, the pressure in the steam dome and the water level in the downcomer which show agreement with the actions of the MSIVs, SRVs and ECCS. (Author)

  16. Comparisons with measured data of the simulated local core parameters by the coupled code ATHLET-BIPR-VVER applying a new enhanced model of the reactor pressure vessel

    International Nuclear Information System (INIS)

    Nikonov, S.; Pasichnyk, I.; Velkov, K.; Pautz, A.

    2011-01-01

    The paper describes the performed comparisons of measured and simulated local core data based on the OECD/NEA Benchmark on Kalinin-3 NPP: 'Switching off of one of the four operating main circulation pumps at nominal reactor power'. The local measurements of in core self-powered neutron detectors (SPND) in 64 fuel assemblies on 7 axial levels are used for the comparisons of the assemblies axial power distributions and the thermocouples readings at 93 fuel assembly heads are applied for the fuel assembly coolant temperature comparisons. The analyses are done on the base of benchmark transient calculations performed with the coupled system code ATHLET/BIPR-VVER. In order to describe more realistically the fluid mixing phenomena in a reactor pressure vessel a new enhanced nodalization scheme is being developed. It could take into account asymmetric flow behaviour in the reactor pressure vessel structures like downcomer, reactor core inlet and outlet, control rods' guided tubes, support grids etc. For this purpose details of the core geometry are modelled. About 58000 control volumes and junctions are applied. Cross connection are used to describe the interaction between the fluid objects. The performed comparisons are of great interest because they show some advantages by performing coupled code production pseudo-3D analysis of NPPs applying the parallel thermo-hydraulic channel methodology (or 1D thermo-hydraulic system code modeling). (Authors)

  17. Coupling Legacy and Contemporary Deterministic Codes to Goldsim for Probabilistic Assessments of Potential Low-Level Waste Repository Sites

    Science.gov (United States)

    Mattie, P. D.; Knowlton, R. G.; Arnold, B. W.; Tien, N.; Kuo, M.

    2006-12-01

    Sandia National Laboratories (Sandia), a U.S. Department of Energy National Laboratory, has over 30 years experience in radioactive waste disposal and is providing assistance internationally in a number of areas relevant to the safety assessment of radioactive waste disposal systems. International technology transfer efforts are often hampered by small budgets, time schedule constraints, and a lack of experienced personnel in countries with small radioactive waste disposal programs. In an effort to surmount these difficulties, Sandia has developed a system that utilizes a combination of commercially available codes and existing legacy codes for probabilistic safety assessment modeling that facilitates the technology transfer and maximizes limited available funding. Numerous codes developed and endorsed by the United States Nuclear Regulatory Commission and codes developed and maintained by United States Department of Energy are generally available to foreign countries after addressing import/export control and copyright requirements. From a programmatic view, it is easier to utilize existing codes than to develop new codes. From an economic perspective, it is not possible for most countries with small radioactive waste disposal programs to maintain complex software, which meets the rigors of both domestic regulatory requirements and international peer review. Therefore, re-vitalization of deterministic legacy codes, as well as an adaptation of contemporary deterministic codes, provides a creditable and solid computational platform for constructing probabilistic safety assessment models. External model linkage capabilities in Goldsim and the techniques applied to facilitate this process will be presented using example applications, including Breach, Leach, and Transport-Multiple Species (BLT-MS), a U.S. NRC sponsored code simulating release and transport of contaminants from a subsurface low-level waste disposal facility used in a cooperative technology transfer

  18. Relating system-to-CFD coupled code analyses to theoretical framework of a multi-scale method

    International Nuclear Information System (INIS)

    Cadinu, F.; Kozlowski, T.; Dinh, T.N.

    2007-01-01

    Over past decades, analyses of transient processes and accidents in a nuclear power plant have been performed, to a significant extent and with a great success, by means of so called system codes, e.g. RELAP5, CATHARE, ATHLET codes. These computer codes, based on a multi-fluid model of two-phase flow, provide an effective, one-dimensional description of the coolant thermal-hydraulics in the reactor system. For some components in the system, wherever needed, the effect of multi-dimensional flow is accounted for through approximate models. The later are derived from scaled experiments conducted for selected accident scenarios. Increasingly, however, we have to deal with newer and ever more complex accident scenarios. In some such cases the system codes fail to serve as simulation vehicle, largely due to its deficient treatment of multi-dimensional flow (in e.g. downcomer, lower plenum). A possible way of improvement is to use the techniques of Computational Fluid Dynamics (CFD). Based on solving Navier-Stokes equations, CFD codes have been developed and used, broadly, to perform analysis of multi-dimensional flow, dominantly in non-nuclear industry and for single-phase flow applications. It is clear that CFD simulations can not substitute system codes but just complement them. Given the intrinsic multi-scale nature of this problem, we propose to relate it to the more general field of research on multi-scale simulations. Even though multi-scale methods are developed on case-by-case basis, the need for a unified framework brought to the development of the heterogeneous multi-scale method (HMM)

  19. Extension of BEPU methods to Sub-channel Thermal-Hydraulics and to Coupled Three-Dimensional Neutronics/Thermal-Hydraulics Codes

    International Nuclear Information System (INIS)

    Avramova, M.; Ivanov, K.; Arenas, C.

    2013-01-01

    The principles that support the risk-informed regulation are to be considered in an integrated decision-making process. Thus, any evaluation of licensing issues supported by a safety analysis would take into account both deterministic and probabilistic aspects of the problem. The deterministic aspects will be addressed using Best Estimate code calculations and considering the associated uncertainties i.e. Plus Uncertainty (BEPU) calculations. In recent years there has been an increasing demand from nuclear research, industry, safety and regulation for best estimate predictions to be provided with their confidence bounds. This applies also to the sub-channel thermal-hydraulic codes, which are used to evaluate local safety parameters. The paper discusses the extension of BEPU methods to the sub-channel thermal-hydraulic codes on the example of the Pennsylvania State University (PSU) version of COBRA-TF (CTF). The use of coupled codes supplemented with uncertainty analysis allows to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1000, EPR-1600 and etc.). The paper presents the application of Generalized Perturbation Theory (GPT) to generate uncertainties associated with the few-group assembly homogenized neutron cross-section data used as input in coupled reactor core calculations. This is followed by a discussion of uncertainty propagation methodologies, being implemented by PSU in cooperation of Technical University of Catalonia (UPC) for reactor core calculations and for comprehensive multi-physics simulations. (authors)

  20. Using a multiphase flow code to model the coupled effects of repository consolidation and multiphase brine and gas flow at the Waste Isolation Pilot Plant

    International Nuclear Information System (INIS)

    Freeze, G.A.; Larson, K.W.; Davies, P.B.; Webb, S.W.

    1995-01-01

    Long-term repository assessment must consider the processes of (1) gas generation, (2) room closure and expansions due to salt creep, and (3) multiphase (brine and gas) fluid flow, as well as the complex coupling between these three processes. The mechanical creep closure code SANCHO was used to simulate the closure of a single, perfectly sealed disposal room filled with water and backfill. SANCHO uses constitutive models to describe salt creep, waste consolidation, and backfill consolidation, Five different gas-generation rate histories were simulated, differentiated by a rate multiplier, f, which ranged from 0.0 (no gas generation) to 1.0 (expected gas generation under brine-dominated conditions). The results of the SANCHO f-series simulations provide a relationship between gas generation, room closure, and room pressure for a perfectly sealed room. Several methods for coupling this relationship with multiphase fluid flow into and out of a room were examined. Two of the methods are described

  1. Whole core pin-by-pin coupled neutronic-thermal-hydraulic steady state and transient calculations using COBAYA3 code

    International Nuclear Information System (INIS)

    Jimenez, J.; Herrero, J. J.; Cuervo, D.; Aragones, J. M.

    2010-10-01

    Nowadays coupled 3-dimensional neutron kinetics and thermal-hydraulic core calculations are performed by applying a radial average channel approach using a meshing of one quarter of assembly in the best case. This approach does not take into account the subchannels effects due to the averaging of the physical fields and the loose of heterogeneity in the thermal-hydraulic model. Therefore the models do not have enough resolution to predict those subchannels effects which are important for the fuel design safety margins, because it is in the local scale, where we can search the hottest pellet or the maximum heat flux. The Polytechnic University of Madrid advanced multi-scale neutron-kinetics and thermal-hydraulics methodologies being implemented in COBAYA3 include domain decomposition by alternate core dissections for the local 3-dimensional fine-mesh scale problems (pin cells/subchannels) and an analytical nodal diffusion solver for the coarse mesh scale coupled with the thermal-hydraulic using a model of one channel per assembly or per quarter of assembly. In this work, we address the domain decomposition by the alternate core dissections methodology applied to solve coupled 3-dimensional neutronic-thermal-hydraulic problems at the fine-mesh scale. The neutronic-thermal-hydraulic coupling at the cell-subchannel scale allows the treatment of the effects of the detailed thermal-hydraulic feedbacks on cross-sections, thus resulting in better estimates of the local safety margins at the pin level. (Author)

  2. Simulation of decay heat removal by natural convection in a pool type fast reactor model-ramona-with coupled 1D/2D thermal hydraulic code system

    Energy Technology Data Exchange (ETDEWEB)

    Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.

  3. Validation of the coupled neutron kinetic thermohydraulic code ATHLET/DYN3D with help of measured data of the OECD Turbine Trip Benchmarks. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2003-12-01

    The project consisted in the validation of the coupled neutron kinetic/thermal hydraulic code system ATHLET/DYN3D for boiling water reactors by the participation at the OECD/NRC turbine trip benchmark. The benchmark defined by the OECD and the American NRC is based on an experiment with closure of the turbine stop valve which was carried out in 1977 in the nuclear power plant Peach Bottom 2 within the framework of a series of 3 experiments. In the experiment, the closure of the valve caused a pressure wave which propagated with attenuation into the reactor core. The condensation of steam in the reactor core caused by the increase of pressure lead to a positive reactivity insertion. The following rise of power was limited by the feedback and the insertion of the control rods. In the frame of the benchmark, the codes could be validated by comparisons with the measured results and the result of the other participants. The benchmark was divided into 3 phases or exercises. Phase I was used for checking the thermo-hydraulic model of the system using a given power release in the core. In phase II, three-dimensional core calculations were performed for given thermal-hydraulic boundary conditions. Coupled calculations were carried out for the selected experiment and four extreme scenarios in the phase III. In the frame of the project, FZR took part in phases II and III of the benchmark. The calculations for phase II were performed with DYN3D by using the assembly discontinuity factors (ADF) and 764 thermal-hydraulic channels (1 channel/assembly). The ATHLET input data set for the coolant system was obtained form the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). It was slightly modified for the phase III calculations carried out with the parallel coupling of ATHLET and DYN3D. For spatially averaged parameters, a good agreement with the results of measurement and the results of other codes was achieved. The influence of the different models was investigated with the

  4. DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part II: Comparison of different temporal schemes

    International Nuclear Information System (INIS)

    Gomez-Torres, Armando Miguel; Sanchez-Espinoza, Victor Hugo; Ivanov, Kostadin; Macian-Juan, Rafael

    2012-01-01

    Highlights: ► A fixed point iteration (FPI) is implemented in DYNSUB. ► Comparisons between the explicit scheme and the FPI are done. ► The FPI scheme allows moving from one time step to the other with converged solution. ► FPI allows the use of larger time steps without compromising the accuracy of results. ► FPI results are promising and represent an option in order to optimize calculations. -- Abstract: DYNSUB is a novel two-way pin-based coupling of the simplified transport (SP 3 ) version of DYN3D with the subchannel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transients conditions, and has been widely described in Part I of this paper. Additionally to the explicit coupling developed and described in Part I, a nested loop iteration or fixed point iteration (FPI) is implemented in DYNSUB. A FPI is not an implicit scheme but approximates it by adding an iteration loop to the current explicit scheme. The advantage of the method is that it allows the use of larger time steps; however the nested loop iteration could take much more time in getting a converged solution that could be less efficient than the explicit scheme with small time steps. A comparison of the two temporal schemes is performed. The results using FPI are very promising and represent a very good option in order to optimize computational times without losing accuracy. However it is also shown that a FPI scheme can produce inaccurate results if the time step is not chosen in agreement with the analyzed transient.

  5. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor

    International Nuclear Information System (INIS)

    Salazar C, J.H.; Nunez C, A.; Chavez M, C.

    2004-01-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  6. Solution of the fifth dynamic Atomic Energy Research benchmark problem using the coupled code DIN3/ATHLET

    International Nuclear Information System (INIS)

    Kliem, S.

    1998-01-01

    The fifth dynamic benchmark is the first benchmark for coupled thermohydraulic system/three dimensional hexagonal neutron kinetic core models. In this benchmark the interaction between the components of a WWER-440 NPP with the reactor core has been investigated. The initiating event is a symmetrical break of the main steam header at the end of the first fuel cycle and the shutdown conditions with one control rod group s tucking. This break causes an overcooling of the primary circuit. During this overcooling the scram reactivity is compensated and the scrammed reactor becomes re critical. The calculation was continued until the highly-borated water from the high pressure injection system terminated the power excursion. Several aspects of the very complex and complicated benchmark problem are analyzed in detail. Sensitivity studies with different hydraulic parameters are made. The influence on the course of the transient and on the solution is discussed.(Author)

  7. Isotope coded protein labeling coupled immunoprecipitation (ICPL-IP): a novel approach for quantitative protein complex analysis from native tissue.

    Science.gov (United States)

    Vogt, Andreas; Fuerholzner, Bettina; Kinkl, Norbert; Boldt, Karsten; Ueffing, Marius

    2013-05-01

    High confidence definition of protein interactions is an important objective toward the understanding of biological systems. Isotope labeling in combination with affinity-based isolation of protein complexes has increased in accuracy and reproducibility, yet, larger organisms--including humans--are hardly accessible to metabolic labeling and thus, a major limitation has been its restriction to small animals, cell lines, and yeast. As composition as well as the stoichiometry of protein complexes can significantly differ in primary tissues, there is a great demand for methods capable to combine the selectivity of affinity-based isolation as well as the accuracy and reproducibility of isotope-based labeling with its application toward analysis of protein interactions from intact tissue. Toward this goal, we combined isotope coded protein labeling (ICPL)(1) with immunoprecipitation (IP) and quantitative mass spectrometry (MS). ICPL-IP allows sensitive and accurate analysis of protein interactions from primary tissue. We applied ICPL-IP to immuno-isolate protein complexes from bovine retinal tissue. Protein complexes of immunoprecipitated β-tubulin, a highly abundant protein with known interactors as well as the lowly expressed small GTPase RhoA were analyzed. The results of both analyses demonstrate sensitive and selective identification of known as well as new protein interactions by our method.

  8. Isotope Coded Protein Labeling Coupled Immunoprecipitation (ICPL-IP): A Novel Approach for Quantitative Protein Complex Analysis From Native Tissue*

    Science.gov (United States)

    Vogt, Andreas; Fuerholzner, Bettina; Kinkl, Norbert; Boldt, Karsten; Ueffing, Marius

    2013-01-01

    High confidence definition of protein interactions is an important objective toward the understanding of biological systems. Isotope labeling in combination with affinity-based isolation of protein complexes has increased in accuracy and reproducibility, yet, larger organisms—including humans—are hardly accessible to metabolic labeling and thus, a major limitation has been its restriction to small animals, cell lines, and yeast. As composition as well as the stoichiometry of protein complexes can significantly differ in primary tissues, there is a great demand for methods capable to combine the selectivity of affinity-based isolation as well as the accuracy and reproducibility of isotope-based labeling with its application toward analysis of protein interactions from intact tissue. Toward this goal, we combined isotope coded protein labeling (ICPL)1 with immunoprecipitation (IP) and quantitative mass spectrometry (MS). ICPL-IP allows sensitive and accurate analysis of protein interactions from primary tissue. We applied ICPL-IP to immuno-isolate protein complexes from bovine retinal tissue. Protein complexes of immunoprecipitated β-tubulin, a highly abundant protein with known interactors as well as the lowly expressed small GTPase RhoA were analyzed. The results of both analyses demonstrate sensitive and selective identification of known as well as new protein interactions by our method. PMID:23268931

  9. Evaluation of the fluence to dose conversion coefficients for high energy neutrons using a voxel phantom coupled with the GEANT4 code

    CERN Document Server

    Paganini, S

    2005-01-01

    Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from Galactic cosmic radiation. Crews of future high-speed commercial flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the male adult voxels phantom MAX, developed in the Nuclear Energy Department of Pernambuco Federal University in Brazil, has been coupled with the Monte Carlo simulation code GEANT4. This toolkit, distributed and upgraded from the international scientific community of CERN/Switzerland, simulates thermal to ultrahigh energy neutrons transport and interactions in the matter. The high energy neutrons are pointed as the component that contribute about 70% of the neutron effective dose that represent the 35% to 60% total dose at aircraft altitude. In this research calculations of conversion coefficients from fluence to effective dose are performed for neutrons of energies from 100 MeV ...

  10. Selected examples for safety analysis in VVER-440 type reactors simulated by the coupled ATHLET/KIKO3D code system

    International Nuclear Information System (INIS)

    Hegyi, Gy.; Kereszturi, A.; Trosztel, I.

    2005-01-01

    Recently several projects have been initiated in Hungary aiming at the introduction of new fuel type, increased maximum allowed power and economic fuel cycle. The planned upgraded power and parallel application of new fuel type require the renewal of the relevant chapter of the Final Safety Analysis Report (FSAR). One of the main tools used for analyzing transient scenarios initiating by reactivity and power distribution anomalies was the ATHLET/KIKO3D coupled neutron kinetic / thermal-hydraulic code. This paper gives an overview of two analyses, which was prepared in the frame of the revision of Paks FSAR, namely the ''withdrawal of one control rod'' and ''initial phase of main steam line break'' events. (author)

  11. Development of a new code to solve hydro-mechanical coupling, shear failure and tensile failure due to hydraulic fracturing operations.

    Science.gov (United States)

    María Gómez Castro, Berta; De Simone, Silvia; Carrera, Jesús

    2016-04-01

    Nowadays, there are still some unsolved relevant questions which must be faced if we want to proceed to the hydraulic fracturing in a safe way. How much will the fracture propagate? This is one of the most important questions that have to be solved in order to avoid the formation of pathways leading to aquifer targets and atmospheric release. Will the fracture failure provoke a microseismic event? Probably this is the biggest fear that people have in fracking. The aim of this work (developed as a part of the EU - FracRisk project) is to understand the hydro-mechanical coupling that controls the shear of existing fractures and their propagation during a hydraulic fracturing operation, in order to identify the key parameters that dominate these processes and answer the mentioned questions. This investigation focuses on the development of a new C++ code which simulates hydro-mechanical coupling, shear movement and propagation of a fracture. The framework employed, called Kratos, uses the Finite Element Method and the fractures are represented with an interface element which is zero thickness. This means that both sides of the element lie together in the initial configuration (it seems a 1D element in a 2D domain, and a 2D element in a 3D domain) and separate as the adjacent matrix elements deform. Since we are working in hard, fragile rocks, we can assume an elastic matrix and impose irreversible displacements in fractures when rock failure occurs. The formulation used to simulate shear and tensile failures is based on the analytical solution proposed by Okada, 1992 and it is part of an iterative process. In conclusion, the objective of this work is to employ the new code developed to analyze the main uncertainties related with the hydro-mechanical behavior of fractures derived from the hydraulic fracturing operations.

  12. Validation of coupled Relap5-3D code in the analysis of RBMK-1500 specific transients

    International Nuclear Information System (INIS)

    Evaldas, Bubelis; Algirdas, Kaliatka; Eugenijus, Uspuras

    2003-01-01

    This paper deals with the modelling of RBMK-1500 specific transients taking place at Ignalina NPP. These transients include: measurements of void and fast power reactivity coefficients, change of graphite cooling conditions and reactor power reduction transients. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is very unique and important from the gap between fuel channel and the graphite bricks model validation point of view. The measurement results, obtained during this transient, allowed to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. Reactor power reduction is a regular operation procedure during the entire lifetime of the reactor. In all cases it starts by either a scram or a power reduction signal activation by the reactor control and protection system or by an operator. The obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviours of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modelling of the neutronic processes taking place in RBMK- 1500 reactor core. And finally, the performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500

  13. A posteriori error analysis for hydro-mechanical couplings and implementation in Code-Aster; Analyse d'erreur a posteriori pour les couplages hydro-mecaniques et mise en oeuvre dans Code-Aster

    Energy Technology Data Exchange (ETDEWEB)

    Meunier, S

    2007-11-15

    We analyse approximations by finite elements in space and finite differences in time of coupled Hydro-Mechanical (HM) problems related to the quasi-static linear poro-elasticity theory. The physical bases of this theory are briefly restated and an abstract setting is proposed to perform the mathematical study of the stationary and un-stationary versions of the HM problem. For the stationary version, the well-posedness of the continuous and discrete problems are established and the a priori error analysis is performed. Then, we propose the a posteriori error analysis by using two different techniques suited to estimate the displacement error and the pressure error, respectively, both in the H{sub x}{sup 1}-norm. The classical properties of reliability and optimality are proved for the associated error estimators. Some numerical experiments using Code-Aster illustrate the theoretical results. For the un-stationary version, we first establish a stability result for the continuous problem. Then, we present an optimal a priori error analysis using elliptic projection techniques. Finally, the a posteriori error analysis is performed by using two different approaches: a direct approach and an elliptic reconstruction approach. The first is suited to estimate the pressure error in the L{sub t}{sup 2}(H{sub x}{sup 1})-norm and the second is suited to estimate the displacement error in the L{sub t}{sup {infinity}}(H{sub x}{sup 1})-norm and the pressure error in the L{sub t}{sup {infinity}}(H{sub x}{sup 1})-norm. Numerical experiments using Code-Aster complete the theoretical results. (author)

  14. The Aster code; Code Aster

    Energy Technology Data Exchange (ETDEWEB)

    Delbecq, J.M

    1999-07-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  15. BLINDAGE: A neutron and gamma-ray transport code for shieldings with the removal-diffusion technique coupled with the point-kernel technique

    International Nuclear Information System (INIS)

    Fanaro, L.C.C.B.

    1984-01-01

    It was developed the BLINDAGE computer code for the radiation transport (neutrons and gammas) calculation. The code uses the removal - diffusion method for neutron transport and point-kernel technique with buil-up factors for gamma-rays. The results obtained through BLINDAGE code are compared with those obtained with the ANISN and SABINE computer codes. (Author) [pt

  16. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    International Nuclear Information System (INIS)

    White, Morgan C.

    2000-01-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V and V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second, the ability to

  17. Development and Implementation of Photonuclear Cross-Section Data for Mutually Coupled Neutron-Photon Transport Calculations in the Monte Carlo N-Particle (MCNP) Radiation Transport Code

    Energy Technology Data Exchange (ETDEWEB)

    White, Morgan C. [Univ. of Florida, Gainesville, FL (United States)

    2000-07-01

    The fundamental motivation for the research presented in this dissertation was the need to development a more accurate prediction method for characterization of mixed radiation fields around medical electron accelerators (MEAs). Specifically, a model is developed for simulation of neutron and other particle production from photonuclear reactions and incorporated in the Monte Carlo N-Particle (MCNP) radiation transport code. This extension of the capability within the MCNP code provides for the more accurate assessment of the mixed radiation fields. The Nuclear Theory and Applications group of the Los Alamos National Laboratory has recently provided first-of-a-kind evaluated photonuclear data for a select group of isotopes. These data provide the reaction probabilities as functions of incident photon energy with angular and energy distribution information for all reaction products. The availability of these data is the cornerstone of the new methodology for state-of-the-art mutually coupled photon-neutron transport simulations. The dissertation includes details of the model development and implementation necessary to use the new photonuclear data within MCNP simulations. A new data format has been developed to include tabular photonuclear data. Data are processed from the Evaluated Nuclear Data Format (ENDF) to the new class ''u'' A Compact ENDF (ACE) format using a standalone processing code. MCNP modifications have been completed to enable Monte Carlo sampling of photonuclear reactions. Note that both neutron and gamma production are included in the present model. The new capability has been subjected to extensive verification and validation (V&V) testing. Verification testing has established the expected basic functionality. Two validation projects were undertaken. First, comparisons were made to benchmark data from literature. These calculations demonstrate the accuracy of the new data and transport routines to better than 25 percent. Second

  18. Extension of the supercritical carbon dioxide brayton cycle to low reactor power operation: investigations using the coupled anl plant dynamics code-SAS4A/SASSYS-1 liquid metal reactor code system

    International Nuclear Information System (INIS)

    Moisseytsev, A.; Sienicki, J.J.

    2012-01-01

    Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO 2 ) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO 2 cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO 2 . It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO 2 heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO 2 -to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO 2 turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the calculations reveal that the

  19. WWER reactor physics code applications

    International Nuclear Information System (INIS)

    Gado, J.; Kereszturi, A.; Gacs, A.; Telbisz, M.

    1994-01-01

    The coupled steady-state reactor physics and thermohydraulic code system KARATE has been developed and applied for WWER-1000 and WWER-440 operational calculations. The 3 D coupled kinetic code KIKO3D has been developed and validated for WWER-440 accident analysis applications. The coupled kinetic code SMARTA developed by VTT Helsinki has been applied for WWER-440 accident analysis. The paper gives a summary of the experience in code development and application. (authors). 10 refs., 2 tabs., 5 figs

  20. Improved Transient Performance of a Fuzzy Modified Model Reference Adaptive Controller for an Interacting Coupled Tank System Using Real-Coded Genetic Algorithm

    Directory of Open Access Journals (Sweden)

    Asan Mohideen Khansadurai

    2014-01-01

    Full Text Available The main objective of the paper is to design a model reference adaptive controller (MRAC with improved transient performance. A modification to the standard direct MRAC called fuzzy modified MRAC (FMRAC is used in the paper. The FMRAC uses a proportional control based Mamdani-type fuzzy logic controller (MFLC to improve the transient performance of a direct MRAC. The paper proposes the application of real-coded genetic algorithm (RGA to tune the membership function parameters of the proposed FMRAC offline so that the transient performance of the FMRAC is improved further. In this study, a GA based modified MRAC (GAMMRAC, an FMRAC, and a GA based FMRAC (GAFMRAC are designed for a coupled tank setup in a hybrid tank process and their transient performances are compared. The results show that the proposed GAFMRAC gives a better transient performance than the GAMMRAC or the FMRAC. It is concluded that the proposed controller can be used to obtain very good transient performance for the control of nonlinear processes.

  1. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, Alessandro; D' Auria, Francesco [University of Pisa, San Piero a Grado (Italy). Nuclear Research Group San Piero a Grado (GRNSPG); Galetti, Regina, E-mail: regina@cnen.gov.b [National Commission for Nuclear Energy (CNEN), Rio de Janeiro, RJ (Brazil); Bajs, Tomislav [University of Zagreb (Croatia). Fac. of Electrical Engineering and Computing. Dept. of Power Systems; Reventos, Francesc [Technical University of Catalonia, Barcelona (Spain). Dept. of Physics and Nuclear Engineering

    2011-07-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  2. Acquired experience on organizing 3D S.UN.COP: international course to support nuclear license by user training in the areas of scaling, uncertainty, and 3D thermal-hydraulics/neutron-kinetics coupled codes

    International Nuclear Information System (INIS)

    Petruzzi, Alessandro; D'Auria, Francesco; Galetti, Regina; Bajs, Tomislav; Reventos, Francesc

    2011-01-01

    Thermal-hydraulic system computer codes are extensively used worldwide for analysis of nuclear facilities by utilities, regulatory bodies, nuclear power plant designers, vendors, and research organizations. Computer code user represents a source of uncertainty that may significantly affect the results of system code calculations. Code user training and qualification represent an effective means for reducing the variation of results caused by the application of the codes by different users. This paper describes the experience in applying a systematic approach to training code users who, upon completion of the training, should be able to perform calculations making the best possible use of the capabilities of best estimate codes. In addition, this paper presents the organization and the main features of the 3D S.UN.COP (scaling, uncertainty, and 3D coupled code calculations) seminars during which particular emphasis is given to practical applications in connection with the licensing process of best estimate plus uncertainty methodologies, showing the designer, utility and regulatory approaches. (author)

  3. Optimized Core Design and Fuel Management of a Pebble-Bed Type Nuclear Reactor

    International Nuclear Information System (INIS)

    Boer, Brian

    2007-01-01

    The Very High Temperature Reactor (VHTR) has been selected by the international Generation IV research initiative as one of the six most promising nuclear reactor concepts that are expected to enter service in the second half of the 21st century. The VHTR is characterized by a high plant efficiency and a high fuel discharge burnup level. More specifically, the (pebble-bed type) High Temperature Reactor (HTR) is known for its inherently safe characteristics, coming from a negative temperature reactivity feedback, a low power density and a large thermal inertia of the core. The core of a pebble-bed reactor consists of graphite spheres (pebbles) that form a randomly packed porous bed, which is cooled by high pressure helium. The pebbles contain thousands of fuel particles, which are coated with several pyrocarbon and silicon carbon layers that are designed to contain the fission products that are formed during operation of the reactor. The inherent safety concept has been demonstrated in small pebble-bed reactors in practice, but an increase in the reactor size and power is required for cost-effective power production. An increase of the power density in order to increase the helium coolant outlet temperature is attractive with regard to the efficiency and possible process heat applications. However, this increase leads in general to higher fuel temperatures, which could lead to a consequent increase of the fuel coating failure probability. This thesis deals with the pebble-bed type VHTR that aims at an increased coolant outlet temperature of 1000 degrees C and beyond. For the simulation of the neutronic and thermal-hydraulic behavior of the reactor the DALTON-THERMIX coupled code system has been developed and has been validated against experiments performed in the AVR and HTR-10 reactors. An analysis of the 400 MWth Pebble Bed Modular Reactor (PBMR) design shows that the inherent safety concept that has been demonstrated in practice in the smaller AVR and HTR-10

  4. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [de

  5. MCNP code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MCNP code is the major Monte Carlo coupled neutron-photon transport research tool at the Los Alamos National Laboratory, and it represents the most extensive Monte Carlo development program in the United States which is available in the public domain. The present code is the direct descendent of the original Monte Carlo work of Fermi, von Neumaum, and Ulam at Los Alamos in the 1940s. Development has continued uninterrupted since that time, and the current version of MCNP (or its predecessors) has always included state-of-the-art methods in the Monte Carlo simulation of radiation transport, basic cross section data, geometry capability, variance reduction, and estimation procedures. The authors of the present code have oriented its development toward general user application. The documentation, though extensive, is presented in a clear and simple manner with many examples, illustrations, and sample problems. In addition to providing the desired results, the output listings give a a wealth of detailed information (some optional) concerning each state of the calculation. The code system is continually updated to take advantage of advances in computer hardware and software, including interactive modes of operation, diagnostic interrupts and restarts, and a variety of graphical and video aids

  6. Approach to the calculation of energy deposition in a container of fuel irradiated by the neutronic codes coupling fluid-dynamics; Aprpoximacion al calculo de la deposicion energetica en un contenedor de combustible irradiado mediante el acoplamiento de codigos neutronico fluido-dinamicos

    Energy Technology Data Exchange (ETDEWEB)

    Hueso, C.; Aleman, A.; Colomer, C.; Fabbri, M.; Martin, M.; Saellas, J.

    2013-07-01

    In this work identifies a possible area of improvement through the creation of a code of coupling between deposition energy codes which calculate neutron (MCNP), and data from heading into fluid dynamics (ANSYS-Fluent) or codes thermomechanical, called MAFACS (Monte Carlo ANSYS Fluent Automatic Coupling Software), being possible to so summarize the process by shortening the needs of computing time, increasing the precision of the results and therefore improving the design of the components.

  7. Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark. Re-connection of an isolated loop

    Energy Technology Data Exchange (ETDEWEB)

    Kotsarev, Alexander; Lizorkin, Mikhail [National Research Centre ' Kurchatov Institute' , Moscow (Russian Federation); Bencik, Marek; Hadek, Jan [UJV Rez, a.s., Rez (Czech Republic); Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

    2016-09-15

    The 7th AER dynamic benchmark is a continuation of the efforts to validate the codes systematically for the estimation of the transient behavior of VVER type nuclear power plants. The main part of the benchmark is the simulation of the re-connection of an isolated circulation loop with low temperature in a VVER-440 plant. This benchmark was calculated by the National Research Centre ''Kurchatov Institute'' (with the code ATHLET/BIPR-VVER), UJV Rez (with the code RELAP5-3D {sup copyright}) and HZDR (with the code DYN3D/ATHLET). The paper gives an overview of the behavior of the main thermal hydraulic and neutron kinetic parameters in the provided solutions.

  8. OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark for assessing coupled neutronics/thermal-hydraulics system codes for VVER-1000 RIA analysis

    International Nuclear Information System (INIS)

    Ivanov, B.; Ivanov, K.; Aniel, S.; Royer, E.; Kolev, N.; Groudev, P.

    2004-01-01

    The present paper describes the two phases of the OECD/DOE/CEA VVER-1000 coolant transient benchmark labeled as V1000CT. This benchmark is based on a data from the Bulgarian Kozloduy NPP Unit 6. The first phase of the benchmark was designed for the purpose of assessing neutron kinetics and thermal-hydraulic modeling for a VVER-1000 reactor, and specifically for their use in analyzing reactivity transients in a VVER-1000 reactor. Most of the results of Phase 1 will be compared against experimental data and the rest of the results will be used for code-to-code comparison. The second phase of the benchmark is planned for evaluation and improvement of the mixing computational models. Code-to-code and code-to-data comparisons will be done based on data of a mixing experiment conducted at Kozloduy-6. Main steam line break will be also analyzed in the second phase of the V1000CT benchmark. The results from it will be used for code-to-code comparison. The benchmark team has been involved in analyzing different aspects and performing sensitivity studies of the different benchmark exercises. The paper presents a comparison of selected results, obtained with two different system thermal-hydraulics codes, with the plant data for the Exercise 1 of Phase 1 of the benchmark as well as some results for Exercises 2 and 3. Overall, this benchmark has been well accepted internationally, with many organizations representing 11 countries participating in the first phase of the benchmark. (authors)

  9. The Drosophila genes CG14593 and CG30106 code for G-protein-coupled receptors specifically activated by the neuropeptides CCHamide-1 and CCHamide-2

    DEFF Research Database (Denmark)

    Hansen, Karina K; Hauser, Frank; Williamson, Michael

    2011-01-01

    Recently, a novel neuropeptide, CCHamide, was discovered in the silkworm Bombyx mori (L. Roller et al., Insect Biochem. Mol. Biol. 38 (2008) 1147-1157). We have now found that all insects with a sequenced genome have two genes, each coding for a different CCHamide, CCHamide-1 and -2. We have also...

  10. The CORSYS neutronics code system

    International Nuclear Information System (INIS)

    Caner, M.; Krumbein, A.D.; Saphier, D.; Shapira, M.

    1994-01-01

    The purpose of this work is to assemble a code package for LWR core physics including coupled neutronics, burnup and thermal hydraulics. The CORSYS system is built around the cell code WIMS (for group microscopic cross section calculations) and 3-dimension diffusion code CITATION (for burnup and fuel management). We are implementing such a system on an IBM RS-6000 workstation. The code was rested with a simplified model of the Zion Unit 2 PWR. (authors). 6 refs., 8 figs., 1 tabs

  11. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 3. Analysis of the OECD TMI-1 Main-Steam- Line-Break Benchmark Accident Using the Coupled RELAP5/PANTHER Codes

    International Nuclear Information System (INIS)

    Schneidesch, C.R.; Guisset, J.P.; Zhang, J.; Bryce, P.; Parkes, M.

    2001-01-01

    The RELAP5 best-estimate thermal-hydraulic system code has been coupled with the PANTHER three-dimensional (3-D) neutron kinetics code via the TALINK dynamic data exchange control and processing tool. The coupled RELAP5/PANTHER code package is being qualified and will be used at British Energy (BE) and Tractebel Energy Engineering (TEE), independently, to analyze pressurized water reactor (PWR) transients where strong core-system interactions occur. The Organization for Economic Cooperation and Development/Nuclear Energy Agency PWR Main-Steam-Line-Break (MSLB) Benchmark problem was performed to demonstrate the capability of the coupled code package to simulate such transients, and this paper reports the BE and TEE contributions. In the first exercise, a point-kinetics (PK) calculation is performed using the RELAP5 code. Two solutions have been derived for the PK case. The first corresponds to scenario, 1 where calculations are carried out using the original (BE) rod worth and where no significant return to power (RTP) occurs. The second corresponds to scenario 2 with arbitrarily reduced rod worth in order to obtain RTP (and was not part of the 'official' results). The results, as illustrated in Fig. 1, show that the thermalhydraulic system response and rod worth are essential in determining the core response. The second exercise consists of a 3-D neutron kinetics transient calculation driven by best-estimate time-dependent core inlet conditions on a 18 T and H zones basis derived from TRAC-PF1/MOD2 (PSU), again analyzing two scenarios of different rod worths. Two sets of PANTHER solutions were submitted for exercise 2. The first solution uses a spatial discretization of one node per assembly and 24 core axial layers for both flux and T and H mesh. The second is characterized by spatial refinement (2 x 2 nodes per assembly, 48 core layers for flux, and T and H calculation), time refinement (half-size time steps), and an increased radial discretization for solution

  12. Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

    Energy Technology Data Exchange (ETDEWEB)

    Van Hove, W.; Van Laeken, K.; Bartsoen, L. [Belgatom, Brussels (Belgium)] [and others

    1995-09-01

    To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with time dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.

  13. Maxwell's equations in axisymmetrical geometry: coupling H(curl) finite element in volume and H(div) finite element in surface. The numerical code FuMel

    International Nuclear Information System (INIS)

    Cambon, S.; Lacoste, P.

    2011-01-01

    We propose a finite element method to solve the axisymmetric scattering problem posed on a regular bounded domain. Here we shall show how to reduce the initial 3D problem into a truncated sum of 2D independent problems posed into a meridian plane of the object. Each of these problem results in the coupling of a partial differential equation into the interior domain and an integral equation on the surface simulating the free space. Then variational volume and boundary integral formulations of Maxwell's equation on regular surfaces are derived. We introduce some general finite element adapted to cylindrical coordinates and constructed from nodal and mixed finite element both for the interior (volume) and for the integral equation (surface). (authors)

  14. Numerical and computational aspects of the coupled three-dimensional core/ plant simulations: organization for economic cooperation and development/ U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-II. 5. TMI-1 Benchmark Performed by Different Coupled Three-Dimensional Neutronics Thermal- Hydraulic Codes

    International Nuclear Information System (INIS)

    D'Auria, F.; Galassi, G.M.; Spadoni, A.; Gago, J.L.; Grgic, D.

    2001-01-01

    A comprehensive analysis of a double-ended main-steam-line-break (MSLB) accident assumed to have occurred in the Babcock and Wilcox Three Mile Island (TMI) Unit 1 nuclear power plant (NPP) has been carried out at the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the University of Pisa, Italy. The research has been carried out in cooperation with the University of Zagreb, Croatia, and with partial financial support from the European Union through a grant to one of the authors. The overall activity has been completed within the framework of the participation in the Organization for Economic Cooperation and Development Committee on the Safety of Nuclear Installations-Nuclear Science Committee PWR MSLB Benchmark. Different code versions have been adopted in the analysis. Results from the following codes (or code versions) are described in this paper: 1. RELAP5/mod 3.2.2, gamma version, coupled with the three-dimensional (3-D) neutron kinetics PARCS code; 2. RELAP5/mod 3.2.2, gamma version, coupled with the 3-D neutron kinetics QUABBOX code; 3. RELAP5/3D code coupled with the 3-D neutron kinetics NESTLE code. Boundary and initial conditions of the system, including those relevant to the fuel status, have been supplied by The Pennsylvania State University in cooperation with GPU Nuclear (the utility, owner of TMI) and the U.S. Nuclear Regulatory Commission (NRC). The main challenge for the calculation was the prediction of the return to power (RTP) following the inlet of cold water into the core and one 'stuck-withdrawn' control rod. Non-realistic assumptions were proposed to augment the core power peak following scram. Zero-dimensional neutronics codes were capable of detecting the RTP after scram. However, the application of 3-D neutronics codes to the same scenario allowed the calculation of a similar value for overall core power peak but showed power increase occurrence in about one-tenth of the core volume. The results achieved in phase 1 of

  15. Coding Partitions

    Directory of Open Access Journals (Sweden)

    Fabio Burderi

    2007-05-01

    Full Text Available Motivated by the study of decipherability conditions for codes weaker than Unique Decipherability (UD, we introduce the notion of coding partition. Such a notion generalizes that of UD code and, for codes that are not UD, allows to recover the ``unique decipherability" at the level of the classes of the partition. By tacking into account the natural order between the partitions, we define the characteristic partition of a code X as the finest coding partition of X. This leads to introduce the canonical decomposition of a code in at most one unambiguouscomponent and other (if any totally ambiguouscomponents. In the case the code is finite, we give an algorithm for computing its canonical partition. This, in particular, allows to decide whether a given partition of a finite code X is a coding partition. This last problem is then approached in the case the code is a rational set. We prove its decidability under the hypothesis that the partition contains a finite number of classes and each class is a rational set. Moreover we conjecture that the canonical partition satisfies such a hypothesis. Finally we consider also some relationships between coding partitions and varieties of codes.

  16. Geochemical computer codes. A review

    International Nuclear Information System (INIS)

    Andersson, K.

    1987-01-01

    In this report a review of available codes is performed and some code intercomparisons are also discussed. The number of codes treating natural waters (groundwater, lake water, sea water) is large. Most geochemical computer codes treat equilibrium conditions, although some codes with kinetic capability are available. A geochemical equilibrium model consists of a computer code, solving a set of equations by some numerical method and a data base, consisting of thermodynamic data required for the calculations. There are some codes which treat coupled geochemical and transport modeling. Some of these codes solve the equilibrium and transport equations simultaneously while other solve the equations separately from each other. The coupled codes require a large computer capacity and have thus as yet limited use. Three code intercomparisons have been found in literature. It may be concluded that there are many codes available for geochemical calculations but most of them require a user that us quite familiar with the code. The user also has to know the geochemical system in order to judge the reliability of the results. A high quality data base is necessary to obtain a reliable result. The best results may be expected for the major species of natural waters. For more complicated problems, including trace elements, precipitation/dissolution, adsorption, etc., the results seem to be less reliable. (With 44 refs.) (author)

  17. Numerical Modeling and Investigation of Fluid-Driven Fracture Propagation in Reservoirs Based on a Modified Fluid-Mechanically Coupled Model in Two-Dimensional Particle Flow Code

    Directory of Open Access Journals (Sweden)

    Jian Zhou

    2016-09-01

    Full Text Available Hydraulic fracturing is a useful tool for enhancing rock mass permeability for shale gas development, enhanced geothermal systems, and geological carbon sequestration by the high-pressure injection of a fracturing fluid into tight reservoir rocks. Although significant advances have been made in hydraulic fracturing theory, experiments, and numerical modeling, when it comes to the complexity of geological conditions knowledge is still limited. Mechanisms of fluid injection-induced fracture initiation and propagation should be better understood to take full advantage of hydraulic fracturing. This paper presents the development and application of discrete particle modeling based on two-dimensional particle flow code (PFC2D. Firstly, it is shown that the modeled value of the breakdown pressure for the hydraulic fracturing process is approximately equal to analytically calculated values under varied in situ stress conditions. Furthermore, a series of simulations for hydraulic fracturing in competent rock was performed to examine the influence of the in situ stress ratio, fluid injection rate, and fluid viscosity on the borehole pressure history, the geometry of hydraulic fractures, and the pore-pressure field, respectively. It was found that the hydraulic fractures in an isotropic medium always propagate parallel to the orientation of the maximum principal stress. When a high fluid injection rate is used, higher breakdown pressure is needed for fracture propagation and complex geometries of fractures can develop. When a low viscosity fluid is used, fluid can more easily penetrate from the borehole into the surrounding rock, which causes a reduction of the effective stress and leads to a lower breakdown pressure. Moreover, the geometry of the fractures is not particularly sensitive to the fluid viscosity in the approximate isotropic model.

  18. The RealGas and RealGasH2O Options of the TOUGH+ Code for the Simulation of Coupled Fluid and Heat Flow in Tight/Shale Gas Systems

    Energy Technology Data Exchange (ETDEWEB)

    Moridis, George; Freeman, Craig

    2013-09-30

    We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas . The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include: coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas

  19. Conceptual design of a passively safe thorium breeder Pebble Bed Reactor

    International Nuclear Information System (INIS)

    Wols, F.J.; Kloosterman, J.L.; Lathouwers, D.; Hagen, T.H.J.J. van der

    2015-01-01

    Highlights: • This work proposes three possible designs for a thorium Pebble Bed Reactor. • A high-conversion PBR (CR > 0.96), passively safe and within practical constraints. • A thorium breeder PBR (220 cm core) in practical regime, but not passively safe. • A passively safe breeder, requiring higher fuel reprocessing and recycling rates. - Abstract: More sustainable nuclear power generation might be achieved by combining the passive safety and high temperature applications of the Pebble Bed Reactor (PBR) design with the resource availability and favourable waste characteristics of the thorium fuel cycle. It has already been known that breeding can be achieved with the thorium fuel cycle inside a Pebble Bed Reactor if reprocessing is performed. This is also demonstrated in this work for a cylindrical core with a central driver zone, with 3 g heavy metal pebbles for enhanced fission, surrounded by a breeder zone containing 30 g thorium pebbles, for enhanced conversion. The main question of the present work is whether it is also possible to combine passive safety and breeding, within a practical operating regime, inside a thorium Pebble Bed Reactor. Therefore, the influence of several fuel design, core design and operational parameters upon the conversion ratio and passive safety is evaluated. A Depressurized Loss of Forced Cooling (DLOFC) is considered the worst safety scenario that can occur within a PBR. So, the response to a DLOFC with and without scram is evaluated for several breeder PBR designs using a coupled DALTON/THERMIX code scheme. With scram it is purely a heat transfer problem (THERMIX) demonstrating the decay heat removal capability of the design. In case control rods cannot be inserted, the temperature feedback of the core should also be able to counterbalance the reactivity insertion by the decaying xenon without fuel temperatures exceeding 1600 °C. Results show that high conversion ratios (CR > 0.96) and passive safety can be combined in

  20. The Aster code

    International Nuclear Information System (INIS)

    Delbecq, J.M.

    1999-01-01

    The Aster code is a 2D or 3D finite-element calculation code for structures developed by the R and D direction of Electricite de France (EdF). This dossier presents a complete overview of the characteristics and uses of the Aster code: introduction of version 4; the context of Aster (organisation of the code development, versions, systems and interfaces, development tools, quality assurance, independent validation); static mechanics (linear thermo-elasticity, Euler buckling, cables, Zarka-Casier method); non-linear mechanics (materials behaviour, big deformations, specific loads, unloading and loss of load proportionality indicators, global algorithm, contact and friction); rupture mechanics (G energy restitution level, restitution level in thermo-elasto-plasticity, 3D local energy restitution level, KI and KII stress intensity factors, calculation of limit loads for structures), specific treatments (fatigue, rupture, wear, error estimation); meshes and models (mesh generation, modeling, loads and boundary conditions, links between different modeling processes, resolution of linear systems, display of results etc..); vibration mechanics (modal and harmonic analysis, dynamics with shocks, direct transient dynamics, seismic analysis and aleatory dynamics, non-linear dynamics, dynamical sub-structuring); fluid-structure interactions (internal acoustics, mass, rigidity and damping); linear and non-linear thermal analysis; steels and metal industry (structure transformations); coupled problems (internal chaining, internal thermo-hydro-mechanical coupling, chaining with other codes); products and services. (J.S.)

  1. Analysis of the three dimensional core kinetics NESTLE code coupling with the advanced thermo-hydraulic code systems, RELAP5/SCDAPSIM and its application to the Laguna Verde Central reactor; Analisis para el acoplamiento del codigo NESTLE para la cinetica tridimensional del nucleo al codigo avanzado de sistemas termo-hidraulicos, RELAP5/SCDAPSIM y su aplicacion al reactor de la CNLV

    Energy Technology Data Exchange (ETDEWEB)

    Salazar C, J H; Nunez C, A [CNSNS, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D.F. (Mexico); Chavez M, C [UNAM, Facultad de Ingenieria, DEPFI Campus Morelos (Mexico)

    2004-07-01

    The objective of the written present is to propose a methodology for the joining of the codes RELAP5/SCDAPSIM and NESTLE. The development of this joining will be carried out inside a doctoral program of Engineering in Energy with nuclear profile of the Ability of Engineering of the UNAM together with the National Commission of Nuclear Security and Safeguards (CNSNS). The general purpose of this type of developments, is to have tools that are implemented by multiple programs or codes such a that systems or models of the three-dimensional kinetics of the core can be simulated and those of the dynamics of the reactor (water heater-hydraulics). In the past, by limitations for the calculation of the complete answer of both systems, the developed models they were carried out for separate, putting a lot of emphasis in one but neglecting the other one. These methodologies, calls of better estimate, will be good to the nuclear industry to evaluate, with more high grades of detail, the designs of the nuclear power plant (for modifications to those already existent or for new concepts in the designs of advanced reactors), besides analysing events (transitory and have an accident), among other applications. The coupled system was applied to design studies and investigation of the Laguna Verde Nuclear power plant (CNLV). (Author)

  2. Multiple component codes based generalized LDPC codes for high-speed optical transport.

    Science.gov (United States)

    Djordjevic, Ivan B; Wang, Ting

    2014-07-14

    A class of generalized low-density parity-check (GLDPC) codes suitable for optical communications is proposed, which consists of multiple local codes. It is shown that Hamming, BCH, and Reed-Muller codes can be used as local codes, and that the maximum a posteriori probability (MAP) decoding of these local codes by Ashikhmin-Lytsin algorithm is feasible in terms of complexity and performance. We demonstrate that record coding gains can be obtained from properly designed GLDPC codes, derived from multiple component codes. We then show that several recently proposed classes of LDPC codes such as convolutional and spatially-coupled codes can be described using the concept of GLDPC coding, which indicates that the GLDPC coding can be used as a unified platform for advanced FEC enabling ultra-high speed optical transport. The proposed class of GLDPC codes is also suitable for code-rate adaption, to adjust the error correction strength depending on the optical channel conditions.

  3. Introduction of SCIENCE code package

    International Nuclear Information System (INIS)

    Lu Haoliang; Li Jinggang; Zhu Ya'nan; Bai Ning

    2012-01-01

    The SCIENCE code package is a set of neutronics tools based on 2D assembly calculations and 3D core calculations. It is made up of APOLLO2F, SMART and SQUALE and used to perform the nuclear design and loading pattern analysis for the reactors on operation or under construction of China Guangdong Nuclear Power Group. The purpose of paper is to briefly present the physical and numerical models used in each computation codes of the SCIENCE code pack age, including the description of the general structure of the code package, the coupling relationship of APOLLO2-F transport lattice code and SMART core nodal code, and the SQUALE code used for processing the core maps. (authors)

  4. Speaking Code

    DEFF Research Database (Denmark)

    Cox, Geoff

    Speaking Code begins by invoking the “Hello World” convention used by programmers when learning a new language, helping to establish the interplay of text and code that runs through the book. Interweaving the voice of critical writing from the humanities with the tradition of computing and software...

  5. Improving the refueling cycle of a WWER-1000 using cuckoo search method and thermal-neutronic coupling of PARCS v2.7, COBRA-EN and WIMSD-5B codes

    Energy Technology Data Exchange (ETDEWEB)

    Yarizadeh-Beneh, M.; Mazaheri-Beni, H.; Poursalehi, N., E-mail: n_poursalehi@sbu.ac.ir

    2016-12-15

    Highlights: • The cuckoo search algorithm is applied to the loading pattern optimization of a nuclear reactor core. • Calculations during the cycle show a good agreement between results and reference for the original LP. • Results indicate the efficient performance of cuckoo search approach coupled with thermal-neutronic solvers. • Neutronic parameters of proposed core pattern are improved relative to original core pattern. - Abstract: The fuel loading pattern optimization is an important process in the refueling design of a nuclear reactor core. Also the analysis of reactor core performance during the operation cycle can be a significant step in the core loading pattern optimization (LPO). In this work, for the first time, a new method i.e. cuckoo search algorithm (CS) has been applied to the fuel loading pattern design of Bushehr WWER-1000 core. In this regard, two objectives have been chosen for finding the best configuration including the improvement of operation cycle length associated with flattening the radial power distribution of fuel assemblies. The core pattern optimization has been performed by coupling the CS algorithm to thermal-neutronic codes including PARCS v2.7, COBRA-EN and WIMSD-5B for earning desired parameters along the operation cycle. The calculations have been done for the beginning of cycle (BOC) to the end of cycle (EOC) states. According to numerical results, the longer operation cycle for the semi-optimized loading pattern has been achieved along with less power peaking factor (PPF) in comparison to the original core pattern of Bushehr WWER-1000. Gained results confirm the efficient and suitable performance of the developed program and also the introduced CS method in the LPO of a nuclear WWER type.

  6. Qualification of the core model DYN3D coupled with the code ATHLET as an advanced tool for the accident analysis of VVER type reactors. Pt. 2. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.; Rohde, U.

    2002-10-01

    Benchmark calculations for the validation of the coupled neutron kinetics/thermohydraulic code complex DYN3D-ATHLET are described. Two benchmark problems concerning hypothetical accident scenarios with leaks in the steam system for a VVER-440 type reactor and the TMI-1 PWR have been solved. The first benchmark task has been defined by FZR in the frame of the international association 'Atomic Energy Research' (AER), the second exercise has been organized under the auspices of the OECD. While in the first benchmark the break of the main steam collector in the sub-critical hot zero power state of the reactor was considered, the break of one of the two main steam lines at full reactor power was assumed in the OECD benchmark. Therefore, in this exercise the mixing of the coolant from the intact and the defect loops had to be considered, while in the AER benchmark the steam collector break causes a homogeneous overcooling of the primary circuit. In the AER benchmark, each participant had to use its own macroscopic cross section libraries. In the OECD benchmark, the cross sections were given in the benchmark definition. The main task of both benchmark problems was to analyse the re-criticality of the scrammed reactor due to the overcooling. For both benchmark problems, a good agreement of the DYN3D-ATHLET solution with the results of other codes was achieved. Differences in the time of re-criticality and the height of the power peak between various solutions of the AER benchmark can be explained by the use of different cross section data. Significant differences in the thermohydraulic parameters (coolant temperature, pressure) occurred only at the late stage of the transient during the emergency injection of highly borated water. In the OECD benchmark, a broader scattering of the thermohydraulic results can be observed, while a good agreement between the various 3D reactor core calculations with given thermohydraulic boundary conditions was achieved. Reasons for the

  7. International benchmark for coupled codes and uncertainty analysis in modelling: switching-Off of one of the four operating main circulation pumps at nominal reactor power at NPP Kalinin unit 3

    International Nuclear Information System (INIS)

    Tereshonok, V. A.; Nikonov, S. P.; Lizorkin, M. P.; Velkov, K; Pautz, A.; Ivanov, V.

    2008-01-01

    The paper briefly describes the Specification of an international NEA/OECD benchmark based on measured plant data. During the commissioning tests for nominal power at NPP Kalinin Unit 3 a lot of measurements of neutron and thermo-hydraulic parameters have been carried out in the reactor pressure vessel, primary and the secondary circuits. One of the measured data sets for the transient 'Switching-off of one Main Circulation Pump (MCP) at nominal power' has been chosen to be applied for validation of coupled thermal-hydraulic and neutron-kinetic system codes and additionally for performing of uncertainty analyses as a part of the NEA/OECD Uncertainty Analysis in Modeling Benchmark. The benchmark is opened for all countries and institutions. The experimental data and the final specification with the cross section libraries will be provided to the participants from NEA/OECD only after official declaration of real participation in the benchmark and delivery of the simulated results of the transient for comparison. (Author)

  8. Aespoe Hard Rock Laboratory. Aespoe Task Force on Engineered Barrier System. Modelling of THM-coupled processes for benchmark 2.2 with the code GeoSys/RockFlow

    Energy Technology Data Exchange (ETDEWEB)

    Nowak, Thomas; Kunz, Herbert (Federal Inst. for Geosciences and Natural Resources, Hannover (Germany))

    2010-02-15

    In 2004 the Swedish Nuclear Fuel and Waste Management Co. (SKB) initiated the project 'Task Force on Engineered Barrier Systems'. This project has the objective to verify the feasibility of modelling THM-coupled processes (task 1) and gas migration processes (task 2) in clay-rich buffer materials. The tasks are performed on the basis of appropriate benchmarks. This report documents the modelling results of the THM-benchmark 2.2 - the Canister Retrieval Test - using the code GeoSys/RockFlow. The Temperature Buffer Test which was performed in the immediate vicinity of the Canister Retrieval Test is included in the model. Especially the heat transport requires the handling of the problem in 3-D. Due to limitations imposed by post-processing different spatial discretisations of the model had to be used during the processing of the benchmark. The calculated temperatures agree well with measured data. Concerning hydraulic parameters the values of permeability and tortuosity were varied in the calculations. The time necessary to saturate the buffer is very sensitive to both of these values. In comparison to thermal and hydraulic processes the model only has limited capacity to predict the measured evolution of total pressure

  9. Aespoe Hard Rock Laboratory. Aespoe Task Force on Engineered Barrier System. Modelling of THM-coupled processes for benchmark 2.2 with the code GeoSys/RockFlow

    International Nuclear Information System (INIS)

    Nowak, Thomas; Kunz, Herbert

    2010-02-01

    In 2004 the Swedish Nuclear Fuel and Waste Management Co. (SKB) initiated the project 'Task Force on Engineered Barrier Systems'. This project has the objective to verify the feasibility of modelling THM-coupled processes (task 1) and gas migration processes (task 2) in clay-rich buffer materials. The tasks are performed on the basis of appropriate benchmarks. This report documents the modelling results of the THM-benchmark 2.2 - the Canister Retrieval Test - using the code GeoSys/RockFlow. The Temperature Buffer Test which was performed in the immediate vicinity of the Canister Retrieval Test is included in the model. Especially the heat transport requires the handling of the problem in 3-D. Due to limitations imposed by post-processing different spatial discretisations of the model had to be used during the processing of the benchmark. The calculated temperatures agree well with measured data. Concerning hydraulic parameters the values of permeability and tortuosity were varied in the calculations. The time necessary to saturate the buffer is very sensitive to both of these values. In comparison to thermal and hydraulic processes the model only has limited capacity to predict the measured evolution of total pressure

  10. Power feedback effects in the LEM code

    International Nuclear Information System (INIS)

    Kromar, M.

    1999-01-01

    The nodal diffusion code LEM has been extended with the power feedback option. Thermohydraulic and neutronic coupling is covered with the Reactivity Coefficient Method. Presented are results of the code testing. Verification is done on the typical non-uprated NPP Krsko reload cycles. Results show that the code fulfill objectives arising in the process of reactor core analysis.(author)

  11. Coding Labour

    Directory of Open Access Journals (Sweden)

    Anthony McCosker

    2014-03-01

    Full Text Available As well as introducing the Coding Labour section, the authors explore the diffusion of code across the material contexts of everyday life, through the objects and tools of mediation, the systems and practices of cultural production and organisational management, and in the material conditions of labour. Taking code beyond computation and software, their specific focus is on the increasingly familiar connections between code and labour with a focus on the codification and modulation of affect through technologies and practices of management within the contemporary work organisation. In the grey literature of spreadsheets, minutes, workload models, email and the like they identify a violence of forms through which workplace affect, in its constant flux of crisis and ‘prodromal’ modes, is regulated and governed.

  12. Simulation of spray phenomena using the containment code system COCOSYS. 1{sup st} Technical report.Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS (VAMKoS); Simulation von Spruehstrahlphaenomenen mit dem Containment Code System COCOSYS. 1. Technischer Fachbericht. Validierung und Analyse ausgewaehlter Modelle sowie der Kopplung der Systemcodes ATHLET-CD und COCOSYS (VAMKoS)

    Energy Technology Data Exchange (ETDEWEB)

    Risken, Tobias; Koch, Marco K.

    2014-12-15

    The present report is the first Technical Report within the research project ''Validation and interpretation of selected models and of the coupling of the system codes ATHLET-CD and COCOSYS'', funded by the German Federal Ministry for Economic Affairs and Energy (BMWi 1501465) and projected at the Reactor Simulation and Safety Group, Chair of Energy Systems and Energy Economics (LEE) at the Ruhr-Universitaet Bochum (RUB). This report deals with the simulation of spray phenomena with the containment code system COCOSYS, which is developed by the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH. First, post-test calculations of the OECD THAI-2 tests HD-30 and HD-31 are presented. The simulation results are compared to experimental values and thereby assessed. The analysis focuses an the assessment of the simultaneous use of the COCOSYS models IVO (spray model) and FRONT (combustion model) as well as a spray entrainment model developed at RUB regarding the simulation of the phenomena related to the interaction of spray and combustion processes. The simulation results show the necessity to consider the induced turbulences. The simulation of these turbulences is performed by modifying the FRONT input parameters leading to an improvement of the simulation results. The consideration of the entrainment positively influences the simulated flow pattern. Subsequently, the simulation of the entrainment of interacting sprays, as occurring in containment spray systems, is considered. The entrainment of interacting sprays is influenced by droplet collisions and changes of the drag between the droplets and the atmosphere. For the simulation an entrainment factor, which has to be determined externally, is implemented into COCOSYS. Exemplary simulations of the OECD SETH-2 ST3 tests show, that in general the use of entrainment factors enables the calculation of alternated gas distributions.

  13. Speech coding

    Energy Technology Data Exchange (ETDEWEB)

    Ravishankar, C., Hughes Network Systems, Germantown, MD

    1998-05-08

    Speech is the predominant means of communication between human beings and since the invention of the telephone by Alexander Graham Bell in 1876, speech services have remained to be the core service in almost all telecommunication systems. Original analog methods of telephony had the disadvantage of speech signal getting corrupted by noise, cross-talk and distortion Long haul transmissions which use repeaters to compensate for the loss in signal strength on transmission links also increase the associated noise and distortion. On the other hand digital transmission is relatively immune to noise, cross-talk and distortion primarily because of the capability to faithfully regenerate digital signal at each repeater purely based on a binary decision. Hence end-to-end performance of the digital link essentially becomes independent of the length and operating frequency bands of the link Hence from a transmission point of view digital transmission has been the preferred approach due to its higher immunity to noise. The need to carry digital speech became extremely important from a service provision point of view as well. Modem requirements have introduced the need for robust, flexible and secure services that can carry a multitude of signal types (such as voice, data and video) without a fundamental change in infrastructure. Such a requirement could not have been easily met without the advent of digital transmission systems, thereby requiring speech to be coded digitally. The term Speech Coding is often referred to techniques that represent or code speech signals either directly as a waveform or as a set of parameters by analyzing the speech signal. In either case, the codes are transmitted to the distant end where speech is reconstructed or synthesized using the received set of codes. A more generic term that is applicable to these techniques that is often interchangeably used with speech coding is the term voice coding. This term is more generic in the sense that the

  14. Optimal codes as Tanner codes with cyclic component codes

    DEFF Research Database (Denmark)

    Høholdt, Tom; Pinero, Fernando; Zeng, Peng

    2014-01-01

    In this article we study a class of graph codes with cyclic code component codes as affine variety codes. Within this class of Tanner codes we find some optimal binary codes. We use a particular subgraph of the point-line incidence plane of A(2,q) as the Tanner graph, and we are able to describe ...

  15. Warthog: Coupling Status Update

    Energy Technology Data Exchange (ETDEWEB)

    Hart, Shane W. D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Reardon, Bradley T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-06-30

    The Warthog code was developed to couple codes that are developed in both the Multi-Physics Object-Oriented Simulation Environment (MOOSE) from Idaho National Laboratory (INL) and SHARP from Argonne National Laboratory (ANL). The initial phase of this work, focused on coupling the neutronics code PROTEUS with the fuel performance code BISON. The main technical challenge involves mapping the power density solution determined by PROTEUS to the fuel in BISON. This presents a challenge since PROTEUS uses the MOAB mesh format, but BISON, like all other MOOSE codes, uses the libMesh format. When coupling the different codes, one must consider that Warthog is a light-weight MOOSE-based program that uses the Data Transfer Kit (DTK) to transfer data between the various mesh types. Users set up inputs for the codes they want to run, and then Warthog transfers the data between them. Currently Warthog supports XSProc from SCALE or the Sub-Group Application Programming Interface (SGAPI) in PROTEUS for generating cross sections. It supports arbitrary geometries using PROTEUS and BISON. DTK will transfer power densities and temperatures between the codes where the domains overlap. In the past fiscal year (FY), much work has gone into demonstrating two-way coupling for simple pin cells of various materials. XSProc was used to calculate the cross sections, which were then passed to PROTEUS in an external file. PROTEUS calculates the fission/power density, and Warthog uses DTK to pass this information to BISON, where it is used as the heat source. BISON then calculates the temperature profile of the pin cell and sends it back to XSProc to obtain the temperature corrected cross sections. This process is repeated until the convergence criteria (tolerance on BISON solve, or number of time steps) is reached. Models have been constructed and run for both uranium oxide and uranium silicide fuels. These models demonstrate a clear difference in power shape that is not accounted for in a

  16. MARS Code in Linux Environment

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Bae, Sung Won; Jung, Jae Joon; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    The two-phase system analysis code MARS has been incorporated into Linux system. The MARS code was originally developed based on the RELAP5/MOD3.2 and COBRA-TF. The 1-D module which evolved from RELAP5 alone could be applied for the whole NSSS system analysis. The 3-D module developed based on the COBRA-TF, however, could be applied for the analysis of the reactor core region where 3-D phenomena would be better treated. The MARS code also has several other code units that could be incorporated for more detailed analysis. The separate code units include containment analysis modules and 3-D kinetics module. These code modules could be optionally invoked to be coupled with the main MARS code. The containment code modules (CONTAIN and CONTEMPT), for example, could be utilized for the analysis of the plant containment phenomena in a coupled manner with the nuclear reactor system. The mass and energy interaction during the hypothetical coolant leakage accident could, thereby, be analyzed in a more realistic manner. In a similar way, 3-D kinetics could be incorporated for simulating the three dimensional reactor kinetic behavior, instead of using the built-in point kinetics model. The MARS code system, developed initially for the MS Windows environment, however, would not be adequate enough for the PC cluster system where multiple CPUs are available. When parallelism is to be eventually incorporated into the MARS code, MS Windows environment is not considered as an optimum platform. Linux environment, on the other hand, is generally being adopted as a preferred platform for the multiple codes executions as well as for the parallel application. In this study, MARS code has been modified for the adaptation of Linux platform. For the initial code modification, the Windows system specific features have been removed from the code. Since the coupling code module CONTAIN is originally in a form of dynamic load library (DLL) in the Windows system, a similar adaptation method

  17. MARS Code in Linux Environment

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Bae, Sung Won; Jung, Jae Joon; Chung, Bub Dong

    2005-01-01

    The two-phase system analysis code MARS has been incorporated into Linux system. The MARS code was originally developed based on the RELAP5/MOD3.2 and COBRA-TF. The 1-D module which evolved from RELAP5 alone could be applied for the whole NSSS system analysis. The 3-D module developed based on the COBRA-TF, however, could be applied for the analysis of the reactor core region where 3-D phenomena would be better treated. The MARS code also has several other code units that could be incorporated for more detailed analysis. The separate code units include containment analysis modules and 3-D kinetics module. These code modules could be optionally invoked to be coupled with the main MARS code. The containment code modules (CONTAIN and CONTEMPT), for example, could be utilized for the analysis of the plant containment phenomena in a coupled manner with the nuclear reactor system. The mass and energy interaction during the hypothetical coolant leakage accident could, thereby, be analyzed in a more realistic manner. In a similar way, 3-D kinetics could be incorporated for simulating the three dimensional reactor kinetic behavior, instead of using the built-in point kinetics model. The MARS code system, developed initially for the MS Windows environment, however, would not be adequate enough for the PC cluster system where multiple CPUs are available. When parallelism is to be eventually incorporated into the MARS code, MS Windows environment is not considered as an optimum platform. Linux environment, on the other hand, is generally being adopted as a preferred platform for the multiple codes executions as well as for the parallel application. In this study, MARS code has been modified for the adaptation of Linux platform. For the initial code modification, the Windows system specific features have been removed from the code. Since the coupling code module CONTAIN is originally in a form of dynamic load library (DLL) in the Windows system, a similar adaptation method

  18. Aztheca Code

    International Nuclear Information System (INIS)

    Quezada G, S.; Espinosa P, G.; Centeno P, J.; Sanchez M, H.

    2017-09-01

    This paper presents the Aztheca code, which is formed by the mathematical models of neutron kinetics, power generation, heat transfer, core thermo-hydraulics, recirculation systems, dynamic pressure and level models and control system. The Aztheca code is validated with plant data, as well as with predictions from the manufacturer when the reactor operates in a stationary state. On the other hand, to demonstrate that the model is applicable during a transient, an event occurred in a nuclear power plant with a BWR reactor is selected. The plant data are compared with the results obtained with RELAP-5 and the Aztheca model. The results show that both RELAP-5 and the Aztheca code have the ability to adequately predict the behavior of the reactor. (Author)

  19. Vocable Code

    DEFF Research Database (Denmark)

    Soon, Winnie; Cox, Geoff

    2018-01-01

    a computational and poetic composition for two screens: on one of these, texts and voices are repeated and disrupted by mathematical chaos, together exploring the performativity of code and language; on the other, is a mix of a computer programming syntax and human language. In this sense queer code can...... be understood as both an object and subject of study that intervenes in the world’s ‘becoming' and how material bodies are produced via human and nonhuman practices. Through mixing the natural and computer language, this article presents a script in six parts from a performative lecture for two persons...

  20. NSURE code

    International Nuclear Information System (INIS)

    Rattan, D.S.

    1993-11-01

    NSURE stands for Near-Surface Repository code. NSURE is a performance assessment code. developed for the safety assessment of near-surface disposal facilities for low-level radioactive waste (LLRW). Part one of this report documents the NSURE model, governing equations and formulation of the mathematical models, and their implementation under the SYVAC3 executive. The NSURE model simulates the release of nuclides from an engineered vault, their subsequent transport via the groundwater and surface water pathways tot he biosphere, and predicts the resulting dose rate to a critical individual. Part two of this report consists of a User's manual, describing simulation procedures, input data preparation, output and example test cases

  1. Tandem Mirror Reactor Systems Code (Version I)

    International Nuclear Information System (INIS)

    Reid, R.L.; Finn, P.A.; Gohar, M.Y.

    1985-09-01

    A computer code was developed to model a Tandem Mirror Reactor. Ths is the first Tandem Mirror Reactor model to couple, in detail, the highly linked physics, magnetics, and neutronic analysis into a single code. This report describes the code architecture, provides a summary description of the modules comprising the code, and includes an example execution of the Tandem Mirror Reactor Systems Code. Results from this code for two sensitivity studies are also included. These studies are: (1) to determine the impact of center cell plasma radius, length, and ion temperature on reactor cost and performance at constant fusion power; and (2) to determine the impact of reactor power level on cost

  2. Coding Class

    DEFF Research Database (Denmark)

    Ejsing-Duun, Stine; Hansbøl, Mikala

    Denne rapport rummer evaluering og dokumentation af Coding Class projektet1. Coding Class projektet blev igangsat i skoleåret 2016/2017 af IT-Branchen i samarbejde med en række medlemsvirksomheder, Københavns kommune, Vejle Kommune, Styrelsen for IT- og Læring (STIL) og den frivillige forening...... Coding Pirates2. Rapporten er forfattet af Docent i digitale læringsressourcer og forskningskoordinator for forsknings- og udviklingsmiljøet Digitalisering i Skolen (DiS), Mikala Hansbøl, fra Institut for Skole og Læring ved Professionshøjskolen Metropol; og Lektor i læringsteknologi, interaktionsdesign......, design tænkning og design-pædagogik, Stine Ejsing-Duun fra Forskningslab: It og Læringsdesign (ILD-LAB) ved Institut for kommunikation og psykologi, Aalborg Universitet i København. Vi har fulgt og gennemført evaluering og dokumentation af Coding Class projektet i perioden november 2016 til maj 2017...

  3. Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the objectives, meeting goals and overall NASA goals for the NASA Data Standards Working Group. The presentation includes information on the technical progress surrounding the objective, short LDPC codes, and the general results on the Pu-Pw tradeoff.

  4. ANIMAL code

    International Nuclear Information System (INIS)

    Lindemuth, I.R.

    1979-01-01

    This report describes ANIMAL, a two-dimensional Eulerian magnetohydrodynamic computer code. ANIMAL's physical model also appears. Formulated are temporal and spatial finite-difference equations in a manner that facilitates implementation of the algorithm. Outlined are the functions of the algorithm's FORTRAN subroutines and variables

  5. Network Coding

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 15; Issue 7. Network Coding. K V Rashmi Nihar B Shah P Vijay Kumar. General Article Volume 15 Issue 7 July 2010 pp 604-621. Fulltext. Click here to view fulltext PDF. Permanent link: https://www.ias.ac.in/article/fulltext/reso/015/07/0604-0621 ...

  6. Expander Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 10; Issue 1. Expander Codes - The Sipser–Spielman Construction. Priti Shankar. General Article Volume 10 ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science Bangalore 560 012, India.

  7. GAMERA - The New Magnetospheric Code

    Science.gov (United States)

    Lyon, J.; Sorathia, K.; Zhang, B.; Merkin, V. G.; Wiltberger, M. J.; Daldorff, L. K. S.

    2017-12-01

    The Lyon-Fedder-Mobarry (LFM) code has been a main-line magnetospheric simulation code for 30 years. The code base, designed in the age of memory to memory vector ma- chines,is still in wide use for science production but needs upgrading to ensure the long term sustainability. In this presentation, we will discuss our recent efforts to update and improve that code base and also highlight some recent results. The new project GAM- ERA, Grid Agnostic MHD for Extended Research Applications, has kept the original design characteristics of the LFM and made significant improvements. The original de- sign included high order numerical differencing with very aggressive limiting, the ability to use arbitrary, but logically rectangular, grids, and maintenance of div B = 0 through the use of the Yee grid. Significant improvements include high-order upwinding and a non-clipping limiter. One other improvement with wider applicability is an im- proved averaging technique for the singularities in polar and spherical grids. The new code adopts a hybrid structure - multi-threaded OpenMP with an overarching MPI layer for large scale and coupled applications. The MPI layer uses a combination of standard MPI and the Global Array Toolkit from PNL to provide a lightweight mechanism for coupling codes together concurrently. The single processor code is highly efficient and can run magnetospheric simulations at the default CCMC resolution faster than real time on a MacBook pro. We have run the new code through the Athena suite of tests, and the results compare favorably with the codes available to the astrophysics community. LFM/GAMERA has been applied to many different situations ranging from the inner and outer heliosphere and magnetospheres of Venus, the Earth, Jupiter and Saturn. We present example results the Earth's magnetosphere including a coupled ring current (RCM), the magnetospheres of Jupiter and Saturn, and the inner heliosphere.

  8. Improved hydrogen distribution calculation in the containment using the coupled codes MELCOR and GASFLOW for the analysis of severe accidents in nuclear power plants; Verbesserte Berechnung der Wasserstoffverteilung im Sicherheitsbehaelter bei der Analyse schwerer Stoerfaelle in Kernkraftwerken durch Kopplung von MELCOR und GASFLOW

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, Tobias

    2014-09-01

    The risk of a hydrogen combustion within a containment of a pressurized water reactor during a severe loss of coolant accident (LOCA) is evaluated using numerical simulations. The code MELCOR provides integral analysis capabilities for severe accidents. Yet, its Lumped Parameter (LP) model provides less accurate information about on thermal hydraulics within the containment during a LOCA. GASFLOW is a CFD code that simulates both the local hydrogen distribution and the pressure inside the containment more realistically. Currently, to perform these GASFLOW simulations, the common procedure is to use a source term from a previous MELCOR calculation. However, with this approach, the influence of the more realistic GASFLOW pressure on the mass flow through the leak cannot be taken into account. This inconsistency is overcome by coupling both codes in this thesis. Here, the MELCOR instance is responsible for the primary and secondary systems. At the same time, the GASFLOW instance predicts the thermal hydraulics of the containment. The more accurate containment pressure from the GASFLOW instance is used in the MELCOR instance to calculate consistent outflow rates through the leak. In order to couple both codes, the existing interface in MELCOR is modified and a new interface for GASFLOW is developed and implemented. To begin with, the hydrogen distribution inside a generic containment is calculated by MELCOR using a typical coarse LP nodalization and a refined one. The results obtained are compared to a GASFLOW simulation. It is shown that the refinement only leads to a better agreement with the GASFLOW result if the correct flow directions are predefined by the nodalization. The safety relevant, local peak concentrations of hydrogen cannot be resolved by MELCOR. Consequently, the use of the CFD code is indispensable. The correct functioning of the coupling is proven within four steps. At first, the modified MELCOR interface is checked by computing a test case using two

  9. ESCADRE and ICARE code systems

    International Nuclear Information System (INIS)

    Reocreux, M.; Gauvain, J.

    1992-01-01

    The French sever accident code development program is following two parallel approaches: the first one is dealing with ''integral codes'' which are designed for giving immediate engineer answers, the second one is following a more mechanistic way in order to have the capability of detailed analysis of experiments, in order to get a better understanding of the scaling problem and reach a better confidence in plant calculations. In the first approach a complete system has been developed and is being used for practical cases: this is the ESCADRE system. In the second approach, a set of codes dealing first with primary circuit is being developed: a mechanistic core degradation code, ICARE, has been issued and is being coupled with the advanced thermalhydraulic code CATHARE. Fission product codes have been also coupled to CATHARE. The ''integral'' ESCADRE system and the mechanistic ICARE and associated codes are described. Their main characteristics are reviewed and the status of their development and assessment given. Future studies are finally discussed. 36 refs, 4 figs, 1 tab

  10. Panda code

    International Nuclear Information System (INIS)

    Altomare, S.; Minton, G.

    1975-02-01

    PANDA is a new two-group one-dimensional (slab/cylinder) neutron diffusion code designed to replace and extend the FAB series. PANDA allows for the nonlinear effects of xenon, enthalpy and Doppler. Fuel depletion is allowed. PANDA has a completely general search facility which will seek criticality, maximize reactivity, or minimize peaking. Any single parameter may be varied in a search. PANDA is written in FORTRAN IV, and as such is nearly machine independent. However, PANDA has been written with the present limitations of the Westinghouse CDC-6600 system in mind. Most computation loops are very short, and the code is less than half the useful 6600 memory size so that two jobs can reside in the core at once. (auth)

  11. CANAL code

    International Nuclear Information System (INIS)

    Gara, P.; Martin, E.

    1983-01-01

    The CANAL code presented here optimizes a realistic iron free extraction channel which has to provide a given transversal magnetic field law in the median plane: the current bars may be curved, have finite lengths and cooling ducts and move in a restricted transversal area; terminal connectors may be added, images of the bars in pole pieces may be included. A special option optimizes a real set of circular coils [fr

  12. Coupling an analytical description of anti-scatter grids with simulation software of radiographic systems using Monte Carlo code; Couplage d'une methode de description analytique de grilles anti diffusantes avec un logiciel de simulation de systemes radiographiques base sur un code Monte Carlo

    Energy Technology Data Exchange (ETDEWEB)

    Rinkel, J.; Dinten, J.M.; Tabary, J

    2004-07-01

    The use of focused anti-scatter grids on digital radiographic systems with two-dimensional detectors produces acquisitions with a decreased scatter to primary ratio and thus improved contrast and resolution. Simulation software is of great interest in optimizing grid configuration according to a specific application. Classical simulators are based on complete detailed geometric descriptions of the grid. They are accurate but very time consuming since they use Monte Carlo code to simulate scatter within the high-frequency grids. We propose a new practical method which couples an analytical simulation of the grid interaction with a radiographic system simulation program. First, a two dimensional matrix of probability depending on the grid is created offline, in which the first dimension represents the angle of impact with respect to the normal to the grid lines and the other the energy of the photon. This matrix of probability is then used by the Monte Carlo simulation software in order to provide the final scattered flux image. To evaluate the gain of CPU time, we define the increasing factor as the increase of CPU time of the simulation with as opposed to without the grid. Increasing factors were calculated with the new model and with classical methods representing the grid with its CAD model as part of the object. With the new method, increasing factors are shorter by one to two orders of magnitude compared with the second one. These results were obtained with a difference in calculated scatter of less than five percent between the new and the classical method. (authors)

  13. Generalized N-coupled maps with invariant measure in Bose ...

    Indian Academy of Sciences (India)

    chronization problem of an array of the linearly coupled map lattices of ... groups and graphs and also in the design of experiments, coding theory, partition ... coupled map lattice which covers internal and external couplings in a form of asso-.

  14. Trellis and turbo coding iterative and graph-based error control coding

    CERN Document Server

    Schlegel, Christian B

    2015-01-01

    This new edition has been extensively revised to reflect the progress in error control coding over the past few years. Over 60% of the material has been completely reworked, and 30% of the material is original. Convolutional, turbo, and low density parity-check (LDPC) coding and polar codes in a unified framework. Advanced research-related developments such as spatial coupling. A focus on algorithmic and implementation aspects of error control coding.

  15. From concatenated codes to graph codes

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom

    2004-01-01

    We consider codes based on simple bipartite expander graphs. These codes may be seen as the first step leading from product type concatenated codes to more complex graph codes. We emphasize constructions of specific codes of realistic lengths, and study the details of decoding by message passing...

  16. Modified PARMILA code for new accelerating structures

    International Nuclear Information System (INIS)

    Takeda, H.; Stovall, J.E.

    1995-01-01

    The PARMILA code was originally developed as a numerical tool to design and simulate the beam performance of the drift-tube linac (DTL). The authors have extended PARMILA to the design of both the coupled-cavity linac (CCL) and the coupled-cavity drift-tube linac (CCDTL). They describe the new design and simulation features associated with these linac structures and improvements to the code that facilitate a seamless linac design process

  17. Fluid structure coupling algorithm

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.; Landram, C.S.; Quinones, D.F.

    1980-01-01

    A fluid-structure-interaction algorithm has been developed and incorporated into the two-dimensional code PELE-IC. This code combines an Eulerian incompressible fluid algorithm with a Lagrangian finite element shell algorithm and incorporates the treatment of complex free surfaces. The fluid structure and coupling algorithms have been verified by the calculation of solved problems from the literature and from air and steam blowdown experiments. The code has been used to calculate loads and structural response from air blowdown and the oscillatory condensation of steam bubbles in water suppression pools typical of boiling water reactors. The techniques developed have been extended to three dimensions and implemented in the computer code PELE-3D

  18. Distinct timescales of population coding across cortex.

    Science.gov (United States)

    Runyan, Caroline A; Piasini, Eugenio; Panzeri, Stefano; Harvey, Christopher D

    2017-08-03

    The cortex represents information across widely varying timescales. For instance, sensory cortex encodes stimuli that fluctuate over few tens of milliseconds, whereas in association cortex behavioural choices can require the maintenance of information over seconds. However, it remains poorly understood whether diverse timescales result mostly from features intrinsic to individual neurons or from neuronal population activity. This question remains unanswered, because the timescales of coding in populations of neurons have not been studied extensively, and population codes have not been compared systematically across cortical regions. Here we show that population codes can be essential to achieve long coding timescales. Furthermore, we find that the properties of population codes differ between sensory and association cortices. We compared coding for sensory stimuli and behavioural choices in auditory cortex and posterior parietal cortex as mice performed a sound localization task. Auditory stimulus information was stronger in auditory cortex than in posterior parietal cortex, and both regions contained choice information. Although auditory cortex and posterior parietal cortex coded information by tiling in time neurons that were transiently informative for approximately 200 milliseconds, the areas had major differences in functional coupling between neurons, measured as activity correlations that could not be explained by task events. Coupling among posterior parietal cortex neurons was strong and extended over long time lags, whereas coupling among auditory cortex neurons was weak and short-lived. Stronger coupling in posterior parietal cortex led to a population code with long timescales and a representation of choice that remained consistent for approximately 1 second. In contrast, auditory cortex had a code with rapid fluctuations in stimulus and choice information over hundreds of milliseconds. Our results reveal that population codes differ across cortex

  19. Improvements to SOIL: An Eulerian hydrodynamics code

    International Nuclear Information System (INIS)

    Davis, C.G.

    1988-04-01

    Possible improvements to SOIL, an Eulerian hydrodynamics code that can do coupled radiation diffusion and strength of materials, are presented in this report. Our research is based on the inspection of other Eulerian codes and theoretical reports on hydrodynamics. Several conclusions from the present study suggest that some improvements are in order, such as second-order advection, adaptive meshes, and speedup of the code by vectorization and/or multitasking. 29 refs., 2 figs

  20. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  1. Neutronic and thermal-hydraulic coupling using Milonga and OpenFOAM codes: an approach using free software; Acoplamento neutrônico e termo-hidráulico usando os códigos Milonga e OpenFOAM: uma abordagem com software livre

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Vitor Vasconcelos Araújo

    2016-07-01

    The development of a fine mesh coupled neutronics/thermal-hydraulics framework mainly using open source software is presented. The contributions proposed go in two different directions: one, is the focus on the open software development, a concept widely spread in many fields of knowledge but rarely explored in the nuclear engineering field; the second, is the use of operating system shared memory as a fast and reliable storage area to couple the computational fluid dynamics (CFD) software OpenFOAM to the free and flexible reactor core analysis code Milonga. This concept was applied to simulate the behavior of the TRIGA Mark 1 IPR-R1 reactor fuel pin in steady-state mode. The macroscopic cross-sections for the model, a set of two-group cross-sections data, were generated using WIMSD-5B code. The results show that this innovative coupled system gives consistent results, encouraging system further development and its use for complex nuclear systems. (author)

  2. Automatic coding method of the ACR Code

    International Nuclear Information System (INIS)

    Park, Kwi Ae; Ihm, Jong Sool; Ahn, Woo Hyun; Baik, Seung Kook; Choi, Han Yong; Kim, Bong Gi

    1993-01-01

    The authors developed a computer program for automatic coding of ACR(American College of Radiology) code. The automatic coding of the ACR code is essential for computerization of the data in the department of radiology. This program was written in foxbase language and has been used for automatic coding of diagnosis in the Department of Radiology, Wallace Memorial Baptist since May 1992. The ACR dictionary files consisted of 11 files, one for the organ code and the others for the pathology code. The organ code was obtained by typing organ name or code number itself among the upper and lower level codes of the selected one that were simultaneous displayed on the screen. According to the first number of the selected organ code, the corresponding pathology code file was chosen automatically. By the similar fashion of organ code selection, the proper pathologic dode was obtained. An example of obtained ACR code is '131.3661'. This procedure was reproducible regardless of the number of fields of data. Because this program was written in 'User's Defined Function' from, decoding of the stored ACR code was achieved by this same program and incorporation of this program into program in to another data processing was possible. This program had merits of simple operation, accurate and detail coding, and easy adjustment for another program. Therefore, this program can be used for automation of routine work in the department of radiology

  3. Error-correction coding

    Science.gov (United States)

    Hinds, Erold W. (Principal Investigator)

    1996-01-01

    This report describes the progress made towards the completion of a specific task on error-correcting coding. The proposed research consisted of investigating the use of modulation block codes as the inner code of a concatenated coding system in order to improve the overall space link communications performance. The study proposed to identify and analyze candidate codes that will complement the performance of the overall coding system which uses the interleaved RS (255,223) code as the outer code.

  4. Dynamic Shannon Coding

    OpenAIRE

    Gagie, Travis

    2005-01-01

    We present a new algorithm for dynamic prefix-free coding, based on Shannon coding. We give a simple analysis and prove a better upper bound on the length of the encoding produced than the corresponding bound for dynamic Huffman coding. We show how our algorithm can be modified for efficient length-restricted coding, alphabetic coding and coding with unequal letter costs.

  5. Fundamentals of convolutional coding

    CERN Document Server

    Johannesson, Rolf

    2015-01-01

    Fundamentals of Convolutional Coding, Second Edition, regarded as a bible of convolutional coding brings you a clear and comprehensive discussion of the basic principles of this field * Two new chapters on low-density parity-check (LDPC) convolutional codes and iterative coding * Viterbi, BCJR, BEAST, list, and sequential decoding of convolutional codes * Distance properties of convolutional codes * Includes a downloadable solutions manual

  6. Codes Over Hyperfields

    Directory of Open Access Journals (Sweden)

    Atamewoue Surdive

    2017-12-01

    Full Text Available In this paper, we define linear codes and cyclic codes over a finite Krasner hyperfield and we characterize these codes by their generator matrices and parity check matrices. We also demonstrate that codes over finite Krasner hyperfields are more interesting for code theory than codes over classical finite fields.

  7. Fuel performance analysis code 'FAIR'

    International Nuclear Information System (INIS)

    Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1994-01-01

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  8. H0 precessor computer code

    International Nuclear Information System (INIS)

    van Dyck, O.B.; Floyd, R.A.

    1981-05-01

    A spin precessor using H - to H 0 stripping, followed by small precession magnets, has been developed for the LAMPF 800-MeV polarized H - beam. The performance of the system was studied with the computer code documented in this report. The report starts from the fundamental physics of a system of spins with hyperfine coupling in a magnetic field and contains many examples of beam behavior as calculated by the program

  9. Vector Network Coding Algorithms

    OpenAIRE

    Ebrahimi, Javad; Fragouli, Christina

    2010-01-01

    We develop new algebraic algorithms for scalar and vector network coding. In vector network coding, the source multicasts information by transmitting vectors of length L, while intermediate nodes process and combine their incoming packets by multiplying them with L x L coding matrices that play a similar role as coding c in scalar coding. Our algorithms for scalar network jointly optimize the employed field size while selecting the coding coefficients. Similarly, for vector coding, our algori...

  10. Improved fluid-structure coupling

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.; Landram, C.S.

    1981-01-01

    In the computer code PELE-IC, an incompressible Eulerian hydrodynamic algorithm was coupled to a Lagrangian finite element shell algorithm for the analysis of pressure suppression in boiling water reactors. This effort also required the development of a free surface algorithm capable of handling expanding gas bubbles. These algorithms have been improved to strengthen the coupling and to add the capability for following the more complex free surfaces resulting from steam condensation. These improvements have also permitted more economical 2D calculations and have made it feasible to develop a 3D version. A compressible option using the acoustic approximation has also been added, furthering the usefulness of the code. The coupling improvements were made in three areas which are identified as (1) preferential coupling, (2) merged cell coupling, and (3) free surface-structure coupling, and are described. These algorithms have been additionally implemented in a three dimensional version of the code called PELE3D. This version has a free surface capability to follow expanding and contracting bubbles and is coupled to a curved rigid surface

  11. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    Energy Technology Data Exchange (ETDEWEB)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available.

  12. AMPX-77: A modular code system for generating coupled multigroup neutron-gamma cross-section libraries from ENDF/B-IV and/or ENDF/B-V

    International Nuclear Information System (INIS)

    Greene, N.M.; Ford, W.E. III; Petrie, L.M.; Arwood, J.W.

    1992-10-01

    AMPX-77 is a modular system of computer programs that pertain to nuclear analyses, with a primary emphasis on tasks associated with the production and use of multigroup cross sections. AH basic cross-section data are to be input in the formats used by the Evaluated Nuclear Data Files (ENDF/B), and output can be obtained in a variety of formats, including its own internal and very general formats, along with a variety of other useful formats used by major transport, diffusion theory, and Monte Carlo codes. Processing is provided for both neutron and gamma-my data. The present release contains codes all written in the FORTRAN-77 dialect of FORTRAN and wig process ENDF/B-V and earlier evaluations, though major modules are being upgraded in order to process ENDF/B-VI and will be released when a complete collection of usable routines is available

  13. Homological stabilizer codes

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Jonas T., E-mail: jonastyleranderson@gmail.com

    2013-03-15

    In this paper we define homological stabilizer codes on qubits which encompass codes such as Kitaev's toric code and the topological color codes. These codes are defined solely by the graphs they reside on. This feature allows us to use properties of topological graph theory to determine the graphs which are suitable as homological stabilizer codes. We then show that all toric codes are equivalent to homological stabilizer codes on 4-valent graphs. We show that the topological color codes and toric codes correspond to two distinct classes of graphs. We define the notion of label set equivalencies and show that under a small set of constraints the only homological stabilizer codes without local logical operators are equivalent to Kitaev's toric code or to the topological color codes. - Highlights: Black-Right-Pointing-Pointer We show that Kitaev's toric codes are equivalent to homological stabilizer codes on 4-valent graphs. Black-Right-Pointing-Pointer We show that toric codes and color codes correspond to homological stabilizer codes on distinct graphs. Black-Right-Pointing-Pointer We find and classify all 2D homological stabilizer codes. Black-Right-Pointing-Pointer We find optimal codes among the homological stabilizer codes.

  14. The RealGas and RealGasH2O options of the TOUGH+ code for the simulation of coupled fluid and heat flow in tight/shale gas systems

    Science.gov (United States)

    We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas. The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and sh...

  15. Diagnostic Coding for Epilepsy.

    Science.gov (United States)

    Williams, Korwyn; Nuwer, Marc R; Buchhalter, Jeffrey R

    2016-02-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  16. Coding of Neuroinfectious Diseases.

    Science.gov (United States)

    Barkley, Gregory L

    2015-12-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  17. Vector Network Coding

    OpenAIRE

    Ebrahimi, Javad; Fragouli, Christina

    2010-01-01

    We develop new algebraic algorithms for scalar and vector network coding. In vector network coding, the source multicasts information by transmitting vectors of length L, while intermediate nodes process and combine their incoming packets by multiplying them with L X L coding matrices that play a similar role as coding coefficients in scalar coding. Our algorithms for scalar network jointly optimize the employed field size while selecting the coding coefficients. Similarly, for vector co...

  18. Entropy Coding in HEVC

    OpenAIRE

    Sze, Vivienne; Marpe, Detlev

    2014-01-01

    Context-Based Adaptive Binary Arithmetic Coding (CABAC) is a method of entropy coding first introduced in H.264/AVC and now used in the latest High Efficiency Video Coding (HEVC) standard. While it provides high coding efficiency, the data dependencies in H.264/AVC CABAC make it challenging to parallelize and thus limit its throughput. Accordingly, during the standardization of entropy coding for HEVC, both aspects of coding efficiency and throughput were considered. This chapter describes th...

  19. Generalized concatenated quantum codes

    International Nuclear Information System (INIS)

    Grassl, Markus; Shor, Peter; Smith, Graeme; Smolin, John; Zeng Bei

    2009-01-01

    We discuss the concept of generalized concatenated quantum codes. This generalized concatenation method provides a systematical way for constructing good quantum codes, both stabilizer codes and nonadditive codes. Using this method, we construct families of single-error-correcting nonadditive quantum codes, in both binary and nonbinary cases, which not only outperform any stabilizer codes for finite block length but also asymptotically meet the quantum Hamming bound for large block length.

  20. Rateless feedback codes

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Koike-Akino, Toshiaki; Orlik, Philip

    2012-01-01

    This paper proposes a concept called rateless feedback coding. We redesign the existing LT and Raptor codes, by introducing new degree distributions for the case when a few feedback opportunities are available. We show that incorporating feedback to LT codes can significantly decrease both...... the coding overhead and the encoding/decoding complexity. Moreover, we show that, at the price of a slight increase in the coding overhead, linear complexity is achieved with Raptor feedback coding....

  1. Advanced video coding systems

    CERN Document Server

    Gao, Wen

    2015-01-01

    This comprehensive and accessible text/reference presents an overview of the state of the art in video coding technology. Specifically, the book introduces the tools of the AVS2 standard, describing how AVS2 can help to achieve a significant improvement in coding efficiency for future video networks and applications by incorporating smarter coding tools such as scene video coding. Topics and features: introduces the basic concepts in video coding, and presents a short history of video coding technology and standards; reviews the coding framework, main coding tools, and syntax structure of AV

  2. Coding for dummies

    CERN Document Server

    Abraham, Nikhil

    2015-01-01

    Hands-on exercises help you learn to code like a pro No coding experience is required for Coding For Dummies,your one-stop guide to building a foundation of knowledge inwriting computer code for web, application, and softwaredevelopment. It doesn't matter if you've dabbled in coding or neverwritten a line of code, this book guides you through the basics.Using foundational web development languages like HTML, CSS, andJavaScript, it explains in plain English how coding works and whyit's needed. Online exercises developed by Codecademy, a leading online codetraining site, help hone coding skill

  3. Adaptive Evolution Coupled with Retrotransposon Exaptation Allowed for the Generation of a Human-Protein-Specific Coding Gene That Promotes Cancer Cell Proliferation and Metastasis in Both Haematological Malignancies and Solid Tumours: The Extraordinary Case of MYEOV Gene

    Directory of Open Access Journals (Sweden)

    Spyros I. Papamichos

    2015-01-01

    Full Text Available The incidence of cancer in human is high as compared to chimpanzee. However previous analysis has documented that numerous human cancer-related genes are highly conserved in chimpanzee. Till date whether human genome includes species-specific cancer-related genes that could potentially contribute to a higher cancer susceptibility remains obscure. This study focuses on MYEOV, an oncogene encoding for two protein isoforms, reported as causally involved in promoting cancer cell proliferation and metastasis in both haematological malignancies and solid tumours. First we document, via stringent in silico analysis, that MYEOV arose de novo in Catarrhini. We show that MYEOV short-isoform start codon was evolutionarily acquired after Catarrhini/Platyrrhini divergence. Throughout the course of Catarrhini evolution MYEOV acquired a gradually elongated translatable open reading frame (ORF, a gradually shortened translation-regulatory upstream ORF, and alternatively spliced mRNA variants. A point mutation introduced in human allowed for the acquisition of MYEOV long-isoform start codon. Second, we demonstrate the precious impact of exonized transposable elements on the creation of MYEOV gene structure. Third, we highlight that the initial part of MYEOV long-isoform coding DNA sequence was under positive selection pressure during Catarrhini evolution. MYEOV represents a Primate Orphan Gene that acquired, via ORF expansion, a human-protein-specific coding potential.

  4. Entrepreneurial Couples

    DEFF Research Database (Denmark)

    Dahl, Michael S.; Van Praag, Mirjam; Thompson, Peter

    2015-01-01

    We study possible motivations for co-entreprenurial couples to start up a joint firm, using a sample of 1,069 Danish couples that established a joint enterprise between 2001 and 2010. We compare their pre-entry characteristics, firm performance and post-dissolution private and financial outcomes...

  5. APR1400 Containment Simulation with CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Moon Kyu; Chung, Bub Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  6. APR1400 Containment Simulation with CONTAIN code

    International Nuclear Information System (INIS)

    Hwang, Moon Kyu; Chung, Bub Dong

    2010-01-01

    The more realistic containment pressure variation predicted by the CONTAIN code through the coupled analysis during a large break loss of coolant accident in the nuclear power plant is expected to provide more accurate prediction for the plant behavior than a standalone MARS-KS calculation. The input deck has been generated based on the already available ARP- 1400 input for CONTEMPT code. Similarly to the CONTEMPT input deck, a simple two-cell model was adopted to model the containment behavior, one cell for the containment inner volume and another cell for the environment condition. The developed input for the CONTAIN code is to be eventually applied for the coupled code calculation of MARS-KS/CONTAIN

  7. Discussion on LDPC Codes and Uplink Coding

    Science.gov (United States)

    Andrews, Ken; Divsalar, Dariush; Dolinar, Sam; Moision, Bruce; Hamkins, Jon; Pollara, Fabrizio

    2007-01-01

    This slide presentation reviews the progress that the workgroup on Low-Density Parity-Check (LDPC) for space link coding. The workgroup is tasked with developing and recommending new error correcting codes for near-Earth, Lunar, and deep space applications. Included in the presentation is a summary of the technical progress of the workgroup. Charts that show the LDPC decoder sensitivity to symbol scaling errors are reviewed, as well as a chart showing the performance of several frame synchronizer algorithms compared to that of some good codes and LDPC decoder tests at ESTL. Also reviewed is a study on Coding, Modulation, and Link Protocol (CMLP), and the recommended codes. A design for the Pseudo-Randomizer with LDPC Decoder and CRC is also reviewed. A chart that summarizes the three proposed coding systems is also presented.

  8. Locally orderless registration code

    DEFF Research Database (Denmark)

    2012-01-01

    This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows.......This is code for the TPAMI paper "Locally Orderless Registration". The code requires intel threadding building blocks installed and is provided for 64 bit on mac, linux and windows....

  9. Decoding Codes on Graphs

    Indian Academy of Sciences (India)

    Shannon limit of the channel. Among the earliest discovered codes that approach the. Shannon limit were the low density parity check (LDPC) codes. The term low density arises from the property of the parity check matrix defining the code. We will now define this matrix and the role that it plays in decoding. 2. Linear Codes.

  10. Manually operated coded switch

    International Nuclear Information System (INIS)

    Barnette, J.H.

    1978-01-01

    The disclosure related to a manually operated recodable coded switch in which a code may be inserted, tried and used to actuate a lever controlling an external device. After attempting a code, the switch's code wheels must be returned to their zero positions before another try is made

  11. Coding in Muscle Disease.

    Science.gov (United States)

    Jones, Lyell K; Ney, John P

    2016-12-01

    Accurate coding is critically important for clinical practice and research. Ongoing changes to diagnostic and billing codes require the clinician to stay abreast of coding updates. Payment for health care services, data sets for health services research, and reporting for medical quality improvement all require accurate administrative coding. This article provides an overview of administrative coding for patients with muscle disease and includes a case-based review of diagnostic and Evaluation and Management (E/M) coding principles in patients with myopathy. Procedural coding for electrodiagnostic studies and neuromuscular ultrasound is also reviewed.

  12. QR Codes 101

    Science.gov (United States)

    Crompton, Helen; LaFrance, Jason; van 't Hooft, Mark

    2012-01-01

    A QR (quick-response) code is a two-dimensional scannable code, similar in function to a traditional bar code that one might find on a product at the supermarket. The main difference between the two is that, while a traditional bar code can hold a maximum of only 20 digits, a QR code can hold up to 7,089 characters, so it can contain much more…

  13. Entrepreneurial Couples

    DEFF Research Database (Denmark)

    Dahl, Michael S.; Van Praag, Mirjam; Thompson, Peter

    with a selected set of comparable firms and couples. We find evidence that couples often establish a business together because one spouse – most commonly the female – has limited outside opportunities in the labor market. However, the financial benefits for each of the spouses, and especially the female......We study possible motivations for co-entrepenurial couples to start up a joint firm, using a sample of 1,069 Danish couples that established a joint enterprise between 2001 and 2010. We compare their pre-entry characteristics, firm performance and postdissolution private and financial outcomes......, are larger in co-entrepreneurial firms, both during the life of the business and post-dissolution. The start-up of co-entrepreneurial firms seems therefore a sound investment in the human capital of both spouses as well as in the reduction of income inequality in the household. We find no evidence of non...

  14. Entrepreneurial Couples

    DEFF Research Database (Denmark)

    Dahl, Michael S.; Van Praag, Mirjam; Thompson, Peter

    with a selected set of comparable firms and couples. We find evidence that couples often establish a business together because one spouse - most commonly the female - has limited outside opportunities in the labor market. However, the financial benefits for each of the spouses, and especially the female......We study possible motivations for co-entrepenurial couples to start up a joint firm, us-ing a sample of 1,069 Danish couples that established a joint enterprise between 2001 and 2010. We compare their pre-entry characteristics, firm performance and post-dissolution private and financial outcomes......, are larger in co-entrepreneurial firms, both during the life of the business and post-dissolution. The start-up of co-entrepreneurial firms seems therefore a sound in-vestment in the human capital of both spouses as well as in the reduction of income inequality in the household. We find no evidence of non...

  15. CTCN: Colloid transport code -- nuclear

    International Nuclear Information System (INIS)

    Jain, R.

    1993-01-01

    This report describes the CTCN computer code, designed to solve the equations of transient colloidal transport of radionuclides in porous and fractured media. This Fortran 77 package solves systems of coupled nonlinear differential-algebraic equations with a wide range of boundary conditions. The package uses the Method of Lines technique with a special section which forms finite-difference discretizations in up to four spatial dimensions to automatically convert the system into a set of ordinary differential equations. The CTCN code then solves these equations using a robust, efficient ODE solver. Thus CTCN can be used to solve population balance equations along with the usual transport equations to model colloid transport processes or as a general problem solver to treat up to four-dimensional differential-algebraic systems

  16. Codes and curves

    CERN Document Server

    Walker, Judy L

    2000-01-01

    When information is transmitted, errors are likely to occur. Coding theory examines efficient ways of packaging data so that these errors can be detected, or even corrected. The traditional tools of coding theory have come from combinatorics and group theory. Lately, however, coding theorists have added techniques from algebraic geometry to their toolboxes. In particular, by re-interpreting the Reed-Solomon codes, one can see how to define new codes based on divisors on algebraic curves. For instance, using modular curves over finite fields, Tsfasman, Vladut, and Zink showed that one can define a sequence of codes with asymptotically better parameters than any previously known codes. This monograph is based on a series of lectures the author gave as part of the IAS/PCMI program on arithmetic algebraic geometry. Here, the reader is introduced to the exciting field of algebraic geometric coding theory. Presenting the material in the same conversational tone of the lectures, the author covers linear codes, inclu...

  17. Future trends in image coding

    Science.gov (United States)

    Habibi, Ali

    1993-01-01

    The objective of this article is to present a discussion on the future of image data compression in the next two decades. It is virtually impossible to predict with any degree of certainty the breakthroughs in theory and developments, the milestones in advancement of technology and the success of the upcoming commercial products in the market place which will be the main factors in establishing the future stage to image coding. What we propose to do, instead, is look back at the progress in image coding during the last two decades and assess the state of the art in image coding today. Then, by observing the trends in developments of theory, software, and hardware coupled with the future needs for use and dissemination of imagery data and the constraints on the bandwidth and capacity of various networks, predict the future state of image coding. What seems to be certain today is the growing need for bandwidth compression. The television is using a technology which is half a century old and is ready to be replaced by high definition television with an extremely high digital bandwidth. Smart telephones coupled with personal computers and TV monitors accommodating both printed and video data will be common in homes and businesses within the next decade. Efficient and compact digital processing modules using developing technologies will make bandwidth compressed imagery the cheap and preferred alternative in satellite and on-board applications. In view of the above needs, we expect increased activities in development of theory, software, special purpose chips and hardware for image bandwidth compression in the next two decades. The following sections summarize the future trends in these areas.

  18. Development of authentication code for multi-access optical code division multiplexing based quantum key distribution

    Science.gov (United States)

    Taiwo, Ambali; Alnassar, Ghusoon; Bakar, M. H. Abu; Khir, M. F. Abdul; Mahdi, Mohd Adzir; Mokhtar, M.

    2018-05-01

    One-weight authentication code for multi-user quantum key distribution (QKD) is proposed. The code is developed for Optical Code Division Multiplexing (OCDMA) based QKD network. A unique address assigned to individual user, coupled with degrading probability of predicting the source of the qubit transmitted in the channel offer excellent secure mechanism against any form of channel attack on OCDMA based QKD network. Flexibility in design as well as ease of modifying the number of users are equally exceptional quality presented by the code in contrast to Optical Orthogonal Code (OOC) earlier implemented for the same purpose. The code was successfully applied to eight simultaneous users at effective key rate of 32 bps over 27 km transmission distance.

  19. Sensitive determination of thiols in wine samples by a stable isotope-coded derivatization reagent d0/d4-acridone-10-ethyl-N-maleimide coupled with high-performance liquid chromatography-electrospray ionization-tandem mass spectrometry analysis.

    Science.gov (United States)

    Lv, Zhengxian; You, Jinmao; Lu, Shuaimin; Sun, Weidi; Ji, Zhongyin; Sun, Zhiwei; Song, Cuihua; Chen, Guang; Li, Guoliang; Hu, Na; Zhou, Wu; Suo, Yourui

    2017-03-31

    As the key aroma compounds, varietal thiols are the crucial odorants responsible for the flavor of wines. Quantitative analysis of thiols can provide crucial information for the aroma profiles of different wine styles. In this study, a rapid and sensitive method for the simultaneous determination of six thiols in wine using d 0 /d 4 -acridone-10-ethyl-N-maleimide (d 0 /d 4 -AENM) as stable isotope-coded derivatization reagent (SICD) by high performance liquid chromatography-electrospray ionization-tandem mass spectrometry (HPLC-ESI-MS/MS) has been developed. Quantification of thiols was performed by using d 4 -AENM labeled thiols as the internal standards (IS), followed by stable isotope dilution HPLC-ESI-MS/MS analysis. The AENM derivatization combined with multiple reactions monitoring (MRM) not only allowed trace analysis of thiols due to the extremely high sensitivity, but also efficiently corrected the matrix effects during HPLC-MS/MS and the fluctuation in MS/MS signal intensity due to instrument. The obtained internal standard calibration curves for six thiols were linear over the range of 25-10,000pmol/L (R 2 ≥0.9961). Detection limits (LODs) for most of analytes were below 6.3pmol/L. The proposed method was successfully applied for the simultaneous determination of six kinds of thiols in wine samples with precisions ≤3.5% and recoveries ≥78.1%. In conclusion, the developed method is expected to be a promising tool for detection of trace thiols in wine and also in other complex matrix. Copyright © 2017 Elsevier B.V. All rights reserved.

  20. Fluid-structure-coupling algorithm

    International Nuclear Information System (INIS)

    McMaster, W.H.; Gong, E.Y.; Landram, C.S.; Quinones, D.F.

    1980-01-01

    A fluid-structure-interaction algorithm has been developed and incorporated into the two dimensional code PELE-IC. This code combines an Eulerian incompressible fluid algorithm with a Lagrangian finite element shell algorithm and incorporates the treatment of complex free surfaces. The fluid structure, and coupling algorithms have been verified by the calculation of solved problems from the literature and from air and steam blowdown experiments. The code has been used to calculate loads and structural response from air blowdown and the oscillatory condensation of steam bubbles in water suppression pools typical of boiling water reactors. The techniques developed here have been extended to three dimensions and implemented in the computer code PELE-3D

  1. Coupling of THALES and FROST using MPI Method

    International Nuclear Information System (INIS)

    Park, Jin Woo; Ryu, Seok Hee; Jung, Chan Do; Jung, Jee Hoon; Um, Kil Sup; Lee, Jae Il

    2013-01-01

    This paper presents the coupling method between THALES and FROST and the simulation results with the coupled code system. In this study, subchannel analysis code THALES and transient fuel performance code FROST were coupled using MPI method as the first stage of the development of the multi-dimensional safety analysis methodology. As a part of the validation, the CEA ejection accident was simulated using the coupled THALES-FROST code and the results were compared with the ShinKori 3 and 4 FSAR. Comparison results revealed that CHASER using MPI method predicts fuel temperatures and heat flux quantitatively well. Thus it was confirmed that the THALES and FROST are properly coupled. In near future, ASTRA, multi-dimensional core neutron kinetics code, will be linked to THALESFROST code for the detailed three-dimensional CEA ejection analysis. The current safety analysis methodology for a CEA ejection accident based on numerous conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KNF is developing the multi-dimensional safety analysis methodology to enhance the consequences of the CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and transient fuel performance analysis code FROST are being coupled using message passing interface(MPI). For the first step, THALES and FROST are coupled and tested

  2. The materiality of Code

    DEFF Research Database (Denmark)

    Soon, Winnie

    2014-01-01

    This essay studies the source code of an artwork from a software studies perspective. By examining code that come close to the approach of critical code studies (Marino, 2006), I trace the network artwork, Pupufu (Lin, 2009) to understand various real-time approaches to social media platforms (MSN......, Twitter and Facebook). The focus is not to investigate the functionalities and efficiencies of the code, but to study and interpret the program level of code in order to trace the use of various technological methods such as third-party libraries and platforms’ interfaces. These are important...... to understand the socio-technical side of a changing network environment. Through the study of code, including but not limited to source code, technical specifications and other materials in relation to the artwork production, I would like to explore the materiality of code that goes beyond technical...

  3. Coding for optical channels

    CERN Document Server

    Djordjevic, Ivan; Vasic, Bane

    2010-01-01

    This unique book provides a coherent and comprehensive introduction to the fundamentals of optical communications, signal processing and coding for optical channels. It is the first to integrate the fundamentals of coding theory and optical communication.

  4. SEVERO code - user's manual

    International Nuclear Information System (INIS)

    Sacramento, A.M. do.

    1989-01-01

    This user's manual contains all the necessary information concerning the use of SEVERO code. This computer code is related to the statistics of extremes = extreme winds, extreme precipitation and flooding hazard risk analysis. (A.C.A.S.)

  5. The θ-γ neural code.

    Science.gov (United States)

    Lisman, John E; Jensen, Ole

    2013-03-20

    Theta and gamma frequency oscillations occur in the same brain regions and interact with each other, a process called cross-frequency coupling. Here, we review evidence for the following hypothesis: that the dual oscillations form a code for representing multiple items in an ordered way. This form of coding has been most clearly demonstrated in the hippocampus, where different spatial information is represented in different gamma subcycles of a theta cycle. Other experiments have tested the functional importance of oscillations and their coupling. These involve correlation of oscillatory properties with memory states, correlation with memory performance, and effects of disrupting oscillations on memory. Recent work suggests that this coding scheme coordinates communication between brain regions and is involved in sensory as well as memory processes. Copyright © 2013 Elsevier Inc. All rights reserved.

  6. Synthesizing Certified Code

    OpenAIRE

    Whalen, Michael; Schumann, Johann; Fischer, Bernd

    2002-01-01

    Code certification is a lightweight approach for formally demonstrating software quality. Its basic idea is to require code producers to provide formal proofs that their code satisfies certain quality properties. These proofs serve as certificates that can be checked independently. Since code certification uses the same underlying technology as program verification, it requires detailed annotations (e.g., loop invariants) to make the proofs possible. However, manually adding annotations to th...

  7. FERRET data analysis code

    International Nuclear Information System (INIS)

    Schmittroth, F.

    1979-09-01

    A documentation of the FERRET data analysis code is given. The code provides a way to combine related measurements and calculations in a consistent evaluation. Basically a very general least-squares code, it is oriented towards problems frequently encountered in nuclear data and reactor physics. A strong emphasis is on the proper treatment of uncertainties and correlations and in providing quantitative uncertainty estimates. Documentation includes a review of the method, structure of the code, input formats, and examples

  8. Stylize Aesthetic QR Code

    OpenAIRE

    Xu, Mingliang; Su, Hao; Li, Yafei; Li, Xi; Liao, Jing; Niu, Jianwei; Lv, Pei; Zhou, Bing

    2018-01-01

    With the continued proliferation of smart mobile devices, Quick Response (QR) code has become one of the most-used types of two-dimensional code in the world. Aiming at beautifying the appearance of QR codes, existing works have developed a series of techniques to make the QR code more visual-pleasant. However, these works still leave much to be desired, such as visual diversity, aesthetic quality, flexibility, universal property, and robustness. To address these issues, in this paper, we pro...

  9. Enhancing QR Code Security

    OpenAIRE

    Zhang, Linfan; Zheng, Shuang

    2015-01-01

    Quick Response code opens possibility to convey data in a unique way yet insufficient prevention and protection might lead into QR code being exploited on behalf of attackers. This thesis starts by presenting a general introduction of background and stating two problems regarding QR code security, which followed by a comprehensive research on both QR code itself and related issues. From the research a solution taking advantages of cloud and cryptography together with an implementation come af...

  10. Opening up codings?

    DEFF Research Database (Denmark)

    Steensig, Jakob; Heinemann, Trine

    2015-01-01

    doing formal coding and when doing more “traditional” conversation analysis research based on collections. We are more wary, however, of the implication that coding-based research is the end result of a process that starts with qualitative investigations and ends with categories that can be coded...

  11. Gauge color codes

    DEFF Research Database (Denmark)

    Bombin Palomo, Hector

    2015-01-01

    Color codes are topological stabilizer codes with unusual transversality properties. Here I show that their group of transversal gates is optimal and only depends on the spatial dimension, not the local geometry. I also introduce a generalized, subsystem version of color codes. In 3D they allow...

  12. Refactoring test code

    NARCIS (Netherlands)

    A. van Deursen (Arie); L.M.F. Moonen (Leon); A. van den Bergh; G. Kok

    2001-01-01

    textabstractTwo key aspects of extreme programming (XP) are unit testing and merciless refactoring. Given the fact that the ideal test code / production code ratio approaches 1:1, it is not surprising that unit tests are being refactored. We found that refactoring test code is different from

  13. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report; Qualifizierung des Kernmodells DYN3D im Komplex mit dem Stoerfallcode ATHLET als fortgeschrittenes Werkzeug fuer die Stoerfallanalyse von WWER-Reaktoren. T. 1. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Grundmann, U.; Kliem, S.; Krepper, E.; Mittag, S; Rohde, U.; Schaefer, F.; Seidel, A.

    1998-03-01

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.) [Deutsch] Das Reaktorkernmodell DYN3D mit 3D Neutronenkinetik wurde an den Thermohydraulik-Systemcode ATHLET angekoppelt. Im vorliegenden Bericht werden Arbeiten zur Qualifizierung des gekoppelten Codekomplexes zu einem validierten Hilfsmittel fuer Stoerfallablaufanalysen zu Reaktoren des russischen Typs WWER dargestellt. Diese umfassten im einzelnen: - Beitraege zur Validierung der Einzelcodes ATHLET und DYN3D anhand der Nachrechnung von Experimenten zum

  14. Software Certification - Coding, Code, and Coders

    Science.gov (United States)

    Havelund, Klaus; Holzmann, Gerard J.

    2011-01-01

    We describe a certification approach for software development that has been adopted at our organization. JPL develops robotic spacecraft for the exploration of the solar system. The flight software that controls these spacecraft is considered to be mission critical. We argue that the goal of a software certification process cannot be the development of "perfect" software, i.e., software that can be formally proven to be correct under all imaginable and unimaginable circumstances. More realistically, the goal is to guarantee a software development process that is conducted by knowledgeable engineers, who follow generally accepted procedures to control known risks, while meeting agreed upon standards of workmanship. We target three specific issues that must be addressed in such a certification procedure: the coding process, the code that is developed, and the skills of the coders. The coding process is driven by standards (e.g., a coding standard) and tools. The code is mechanically checked against the standard with the help of state-of-the-art static source code analyzers. The coders, finally, are certified in on-site training courses that include formal exams.

  15. 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.)

  16. 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.)

  17. APROS couplings from core to containment

    International Nuclear Information System (INIS)

    Puska, E.K.; Ylijoki, J.

    2005-01-01

    APROS simulation environment is able to describe the 1-D and 3-D neutronics of the reactor core. It is also able to describe the thermal hydraulics of the core and circuits either with 5- equation or 6-equation thermal hydraulics. It can also describe the plant automation and electrical systems, as well as the behaviour of the containment. The peculiar feature of APROS in comparison to other coupled systems is that all parts in the coupled system are described with the same code instead of coupling two or three separate codes together with information exchange between the separate codes. The most recent possibility is the coupled calculation of the process and the containment. The more traditional coupling, the coupling of the process containing both the process description and the automation description with more or less detailed description of the 3-D core either for safety analysis or real-time simulation purposes has been discussed in previous work. The paper presents and discusses the capabilities of the code in coupling the plant process and automation description with the plant containment description with two example transient cases. An improved boron concentration solution with second order upwind discretization has been recently included in APROS. An example on the increased accuracy acquired in the 3-D core model has been included. (authors)

  18. The network code

    International Nuclear Information System (INIS)

    1997-01-01

    The Network Code defines the rights and responsibilities of all users of the natural gas transportation system in the liberalised gas industry in the United Kingdom. This report describes the operation of the Code, what it means, how it works and its implications for the various participants in the industry. The topics covered are: development of the competitive gas market in the UK; key points in the Code; gas transportation charging; impact of the Code on producers upstream; impact on shippers; gas storage; supply point administration; impact of the Code on end users; the future. (20 tables; 33 figures) (UK)

  19. Coding for Electronic Mail

    Science.gov (United States)

    Rice, R. F.; Lee, J. J.

    1986-01-01

    Scheme for coding facsimile messages promises to reduce data transmission requirements to one-tenth current level. Coding scheme paves way for true electronic mail in which handwritten, typed, or printed messages or diagrams sent virtually instantaneously - between buildings or between continents. Scheme, called Universal System for Efficient Electronic Mail (USEEM), uses unsupervised character recognition and adaptive noiseless coding of text. Image quality of resulting delivered messages improved over messages transmitted by conventional coding. Coding scheme compatible with direct-entry electronic mail as well as facsimile reproduction. Text transmitted in this scheme automatically translated to word-processor form.

  20. Spread-spectrum communication using binary spatiotemporal chaotic codes

    International Nuclear Information System (INIS)

    Wang Xingang; Zhan Meng; Gong Xiaofeng; Lai, C.H.; Lai, Y.-C.

    2005-01-01

    We propose a scheme to generate binary code for baseband spread-spectrum communication by using a chain of coupled chaotic maps. We compare the performances of this type of spatiotemporal chaotic code with those of a conventional code used frequently in digital communication, the Gold code, and demonstrate that our code is comparable or even superior to the Gold code in several key aspects: security, bit error rate, code generation speed, and the number of possible code sequences. As the field of communicating with chaos faces doubts in terms of performance comparison with conventional digital communication schemes, our work gives a clear message that communicating with chaos can be advantageous and it deserves further attention from the nonlinear science community

  1. NAGRADATA. Code key. Geology

    International Nuclear Information System (INIS)

    Mueller, W.H.; Schneider, B.; Staeuble, J.

    1984-01-01

    This reference manual provides users of the NAGRADATA system with comprehensive keys to the coding/decoding of geological and technical information to be stored in or retreaved from the databank. Emphasis has been placed on input data coding. When data is retreaved the translation into plain language of stored coded information is done automatically by computer. Three keys each, list the complete set of currently defined codes for the NAGRADATA system, namely codes with appropriate definitions, arranged: 1. according to subject matter (thematically) 2. the codes listed alphabetically and 3. the definitions listed alphabetically. Additional explanation is provided for the proper application of the codes and the logic behind the creation of new codes to be used within the NAGRADATA system. NAGRADATA makes use of codes instead of plain language for data storage; this offers the following advantages: speed of data processing, mainly data retrieval, economies of storage memory requirements, the standardisation of terminology. The nature of this thesaurian type 'key to codes' makes it impossible to either establish a final form or to cover the entire spectrum of requirements. Therefore, this first issue of codes to NAGRADATA must be considered to represent the current state of progress of a living system and future editions will be issued in a loose leave ringbook system which can be updated by an organised (updating) service. (author)

  2. XSOR codes users manual

    International Nuclear Information System (INIS)

    Jow, Hong-Nian; Murfin, W.B.; Johnson, J.D.

    1993-11-01

    This report describes the source term estimation codes, XSORs. The codes are written for three pressurized water reactors (Surry, Sequoyah, and Zion) and two boiling water reactors (Peach Bottom and Grand Gulf). The ensemble of codes has been named ''XSOR''. The purpose of XSOR codes is to estimate the source terms which would be released to the atmosphere in severe accidents. A source term includes the release fractions of several radionuclide groups, the timing and duration of releases, the rates of energy release, and the elevation of releases. The codes have been developed by Sandia National Laboratories for the US Nuclear Regulatory Commission (NRC) in support of the NUREG-1150 program. The XSOR codes are fast running parametric codes and are used as surrogates for detailed mechanistic codes. The XSOR codes also provide the capability to explore the phenomena and their uncertainty which are not currently modeled by the mechanistic codes. The uncertainty distributions of input parameters may be used by an. XSOR code to estimate the uncertainty of source terms

  3. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    1999-01-01

    The present lecture has a main goal to show how the transport lattice calculations are realised in a standard computer code. This is illustrated on the example of the WIMSD code, belonging to the most popular tools for reactor calculations. Most of the approaches discussed here can be easily modified to any other lattice code. The description of the code assumes the basic knowledge of reactor lattice, on the level given in the lecture on 'Reactor lattice transport calculations'. For more advanced explanation of the WIMSD code the reader is directed to the detailed descriptions of the code cited in References. The discussion of the methods and models included in the code is followed by the generally used homogenisation procedure and several numerical examples of discrepancies in calculated multiplication factors based on different sources of library data. (author)

  4. DLLExternalCode

    Energy Technology Data Exchange (ETDEWEB)

    2014-05-14

    DLLExternalCode is the a general dynamic-link library (DLL) interface for linking GoldSim (www.goldsim.com) with external codes. The overall concept is to use GoldSim as top level modeling software with interfaces to external codes for specific calculations. The DLLExternalCode DLL that performs the linking function is designed to take a list of code inputs from GoldSim, create an input file for the external application, run the external code, and return a list of outputs, read from files created by the external application, back to GoldSim. Instructions for creating the input file, running the external code, and reading the output are contained in an instructions file that is read and interpreted by the DLL.

  5. Coding as a Trojan Horse for Mathematics Education Reform

    Science.gov (United States)

    Gadanidis, George

    2015-01-01

    The history of mathematics educational reform is replete with innovations taken up enthusiastically by early adopters without significant transfer to other classrooms. This paper explores the coupling of coding and mathematics education to create the possibility that coding may serve as a Trojan Horse for mathematics education reform. That is,…

  6. Computer Code for Nanostructure Simulation

    Science.gov (United States)

    Filikhin, Igor; Vlahovic, Branislav

    2009-01-01

    Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.

  7. NORTICA - a new code for cyclotron analysis

    International Nuclear Information System (INIS)

    Gorelov, D.; Johnson, D.; Marti, F.

    2001-01-01

    The new package NORTICA (Numerical ORbit Tracking In Cyclotrons with Analysis) of computer codes for beam dynamics simulations is under development at NSCL. The package was started as a replacement for the code MONSTER developed in the laboratory in the past. The new codes are capable of beam dynamics simulations in both CCF (Coupled Cyclotron Facility) accelerators, the K500 and K1200 superconducting cyclotrons. The general purpose of this package is assisting in setting and tuning the cyclotrons taking into account the main field and extraction channel imperfections. The computer platform for the package is Alpha Station with UNIX operating system and X-Windows graphic interface. A multiple programming language approach was used in order to combine the reliability of the numerical algorithms developed over the long period of time in the laboratory and the friendliness of modern style user interface. This paper describes the capability and features of the codes in the present state

  8. RCS modeling with the TSAR FDTD code

    Energy Technology Data Exchange (ETDEWEB)

    Pennock, S.T.; Ray, S.L.

    1992-03-01

    The TSAR electromagnetic modeling system consists of a family of related codes that have been designed to work together to provide users with a practical way to set up, run, and interpret the results from complex 3-D finite-difference time-domain (FDTD) electromagnetic simulations. The software has been in development at the Lawrence Livermore National Laboratory (LLNL) and at other sites since 1987. Active internal use of the codes began in 1988 with limited external distribution and use beginning in 1991. TSAR was originally developed to analyze high-power microwave and EMP coupling problems. However, the general-purpose nature of the tools has enabled us to use the codes to solve a broader class of electromagnetic applications and has motivated the addition of new features. In particular a family of near-to-far field transformation routines have been added to the codes, enabling TSAR to be used for radar-cross section and antenna analysis problems.

  9. Flashing coupled density wave oscillation

    International Nuclear Information System (INIS)

    Jiang Shengyao; Wu Xinxin; Zhang Youjie

    1997-07-01

    The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The phenomenon and mechanism of different kinds of two-phase flow instabilities, namely geyser instability, flashing instability and flashing coupled density wave instability are described. The especially interpreted flashing coupled density wave instability has never been studied well, it is analyzed by using a one-dimensional non-thermo equilibrium two-phase flow drift model computer code. Calculations are in good agreement with the experiment results. (5 refs.,5 figs., 1 tab.)

  10. LINK codes TRAC-BF1/PARCSv2.7 in LINUX without external communication interface

    International Nuclear Information System (INIS)

    Barrachina, T.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Verdu, G.; Concejal, A.; Solar, A.

    2014-01-01

    The TRAC-BF1 code is still widely used by the nuclear industry for safety analysis. The plant models developed using this code are highly validated, so it is advisable to continue improving this code before migrating to another completely different code. The coupling with the NRC neutronic code PARCSv2.7 increases the simulation capabilities in transients in which the power distribution plays an important role. In this paper, the procedure for the coupling of TRAC-BF1 and PARCSv2.7 codes without PVM and in Linux is presented. (Author)

  11. Toric Varieties and Codes, Error-correcting Codes, Quantum Codes, Secret Sharing and Decoding

    DEFF Research Database (Denmark)

    Hansen, Johan Peder

    We present toric varieties and associated toric codes and their decoding. Toric codes are applied to construct Linear Secret Sharing Schemes (LSSS) with strong multiplication by the Massey construction. Asymmetric Quantum Codes are obtained from toric codes by the A.R. Calderbank P.W. Shor and A.......M. Steane construction of stabilizer codes (CSS) from linear codes containing their dual codes....

  12. Unsteady interfacial coupling of two-phase flow models

    International Nuclear Information System (INIS)

    Hurisse, O.

    2006-01-01

    The primary coolant circuit in a nuclear power plant contains several distinct components (vessel, core, pipes,...). For all components, specific codes based on the discretization of partial differential equations have already been developed. In order to obtain simulations for the whole circuit, the interfacial coupling of these codes is required. The approach examined within this work consists in coupling codes by providing unsteady information through the coupling interface. The numerical technique relies on the use of an interface model, which is combined with the basic strategy that was introduced by Greenberg and Leroux in order to compute approximations of steady solutions of non-homogeneous hyperbolic systems. Three different coupling cases have been examined: (i) the coupling of a one-dimensional Euler system with a two-dimensional Euler system; (ii) the coupling of two distinct homogeneous two-phase flow models; (iii) the coupling of a four-equation homogeneous model with the standard two-fluid model. (author)

  13. Parallelization of Subchannel Analysis Code MATRA

    International Nuclear Information System (INIS)

    Kim, Seongjin; Hwang, Daehyun; Kwon, Hyouk

    2014-01-01

    A stand-alone calculation of MATRA code used up pertinent computing time for the thermal margin calculations while a relatively considerable time is needed to solve the whole core pin-by-pin problems. In addition, it is strongly required to improve the computation speed of the MATRA code to satisfy the overall performance of the multi-physics coupling calculations. Therefore, a parallel approach to improve and optimize the computability of the MATRA code is proposed and verified in this study. The parallel algorithm is embodied in the MATRA code using the MPI communication method and the modification of the previous code structure was minimized. An improvement is confirmed by comparing the results between the single and multiple processor algorithms. The speedup and efficiency are also evaluated when increasing the number of processors. The parallel algorithm was implemented to the subchannel code MATRA using the MPI. The performance of the parallel algorithm was verified by comparing the results with those from the MATRA with the single processor. It is also noticed that the performance of the MATRA code was greatly improved by implementing the parallel algorithm for the 1/8 core and whole core problems

  14. An Optimal Linear Coding for Index Coding Problem

    OpenAIRE

    Pezeshkpour, Pouya

    2015-01-01

    An optimal linear coding solution for index coding problem is established. Instead of network coding approach by focus on graph theoric and algebraic methods a linear coding program for solving both unicast and groupcast index coding problem is presented. The coding is proved to be the optimal solution from the linear perspective and can be easily utilize for any number of messages. The importance of this work is lying mostly on the usage of the presented coding in the groupcast index coding ...

  15. Track 3: growth of nuclear technology and research numerical and computational aspects of the coupled three-dimensional core/plant simulations: organization for economic cooperation and development/U.S. nuclear regulatory commission pressurized water reactor main-steam-line-break benchmark-I. 5. Analyses of the OECD MSLB Benchmark with the Codes DYN3D and DYN3D/ATHLET

    International Nuclear Information System (INIS)

    Grundmann, U.; Kliem, S.

    2001-01-01

    The code DYN3D coupled with ATHLET was used for the analysis of the OECD Main-Steam-Line-Break (MSLB) Benchmark, which is based on real plant design and operational data of the TMI-1 pressurized water reactor (PWR). Like the codes RELAP or TRAC,ATHLET is a thermal-hydraulic system code with point or one-dimensional neutron kinetic models. ATHLET, developed by the Gesellschaft for Anlagen- und Reaktorsicherheit, is widely used in Germany for safety analyses of nuclear power plants. DYN3D consists of three-dimensional nodal kinetic models and a thermal-hydraulic part with parallel coolant channels of the reactor core. DYN3D was coupled with ATHLET for analyzing more complex transients with interactions between coolant flow conditions and core behavior. It can be applied to the whole spectrum of operational transients and accidents, from small and intermediate leaks to large breaks of coolant loops or steam lines at PWRs and boiling water reactors. The so-called external coupling is used for the benchmark, where the thermal hydraulics is split into two parts: DYN3D describes the thermal hydraulics of the core, while ATHLET models the coolant system. Three exercises of the benchmark were simulated: Exercise 1: point kinetics plant simulation (ATHLET) Exercise 2: coupled three-dimensional neutronics/core thermal-hydraulics evaluation of the core response for given core thermal-hydraulic boundary conditions (DYN3D) Exercise 3: best-estimate coupled core-plant transient analysis (DYN3D/ATHLET). Considering the best-estimate cases (scenarios 1 of exercises 2 and 3), the reactor does not reach criticality after the reactor trip. Defining more serious tests for the codes, the efficiency of the control rods was decreased (scenarios 2 of exercises 2 and 3) to obtain recriticality during the transient. Besides the standard simulation given by the specification, modifications are introduced for sensitivity studies. The results presented here show (a) the influence of a reduced

  16. The Aesthetics of Coding

    DEFF Research Database (Denmark)

    Andersen, Christian Ulrik

    2007-01-01

    Computer art is often associated with computer-generated expressions (digitally manipulated audio/images in music, video, stage design, media facades, etc.). In recent computer art, however, the code-text itself – not the generated output – has become the artwork (Perl Poetry, ASCII Art, obfuscated...... code, etc.). The presentation relates this artistic fascination of code to a media critique expressed by Florian Cramer, claiming that the graphical interface represents a media separation (of text/code and image) causing alienation to the computer’s materiality. Cramer is thus the voice of a new ‘code...... avant-garde’. In line with Cramer, the artists Alex McLean and Adrian Ward (aka Slub) declare: “art-oriented programming needs to acknowledge the conditions of its own making – its poesis.” By analysing the Live Coding performances of Slub (where they program computer music live), the presentation...

  17. Majorana fermion codes

    International Nuclear Information System (INIS)

    Bravyi, Sergey; Terhal, Barbara M; Leemhuis, Bernhard

    2010-01-01

    We initiate the study of Majorana fermion codes (MFCs). These codes can be viewed as extensions of Kitaev's one-dimensional (1D) model of unpaired Majorana fermions in quantum wires to higher spatial dimensions and interacting fermions. The purpose of MFCs is to protect quantum information against low-weight fermionic errors, that is, operators acting on sufficiently small subsets of fermionic modes. We examine to what extent MFCs can surpass qubit stabilizer codes in terms of their stability properties. A general construction of 2D MFCs is proposed that combines topological protection based on a macroscopic code distance with protection based on fermionic parity conservation. Finally, we use MFCs to show how to transform any qubit stabilizer code to a weakly self-dual CSS code.

  18. Theory of epigenetic coding.

    Science.gov (United States)

    Elder, D

    1984-06-07

    The logic of genetic control of development may be based on a binary epigenetic code. This paper revises the author's previous scheme dealing with the numerology of annelid metamerism in these terms. Certain features of the code had been deduced to be combinatorial, others not. This paradoxical contrast is resolved here by the interpretation that these features relate to different operations of the code; the combinatiorial to coding identity of units, the non-combinatorial to coding production of units. Consideration of a second paradox in the theory of epigenetic coding leads to a new solution which further provides a basis for epimorphic regeneration, and may in particular throw light on the "regeneration-duplication" phenomenon. A possible test of the model is also put forward.

  19. DISP1 code

    International Nuclear Information System (INIS)

    Vokac, P.

    1999-12-01

    DISP1 code is a simple tool for assessment of the dispersion of the fission product cloud escaping from a nuclear power plant after an accident. The code makes it possible to tentatively check the feasibility of calculations by more complex PSA3 codes and/or codes for real-time dispersion calculations. The number of input parameters is reasonably low and the user interface is simple enough to allow a rapid processing of sensitivity analyses. All input data entered through the user interface are stored in the text format. Implementation of dispersion model corrections taken from the ARCON96 code enables the DISP1 code to be employed for assessment of the radiation hazard within the NPP area, in the control room for instance. (P.A.)

  20. Phonological coding during reading.

    Science.gov (United States)

    Leinenger, Mallorie

    2014-11-01

    The exact role that phonological coding (the recoding of written, orthographic information into a sound based code) plays during silent reading has been extensively studied for more than a century. Despite the large body of research surrounding the topic, varying theories as to the time course and function of this recoding still exist. The present review synthesizes this body of research, addressing the topics of time course and function in tandem. The varying theories surrounding the function of phonological coding (e.g., that phonological codes aid lexical access, that phonological codes aid comprehension and bolster short-term memory, or that phonological codes are largely epiphenomenal in skilled readers) are first outlined, and the time courses that each maps onto (e.g., that phonological codes come online early [prelexical] or that phonological codes come online late [postlexical]) are discussed. Next the research relevant to each of these proposed functions is reviewed, discussing the varying methodologies that have been used to investigate phonological coding (e.g., response time methods, reading while eye-tracking or recording EEG and MEG, concurrent articulation) and highlighting the advantages and limitations of each with respect to the study of phonological coding. In response to the view that phonological coding is largely epiphenomenal in skilled readers, research on the use of phonological codes in prelingually, profoundly deaf readers is reviewed. Finally, implications for current models of word identification (activation-verification model, Van Orden, 1987; dual-route model, e.g., M. Coltheart, Rastle, Perry, Langdon, & Ziegler, 2001; parallel distributed processing model, Seidenberg & McClelland, 1989) are discussed. (PsycINFO Database Record (c) 2014 APA, all rights reserved).

  1. The aeroelastic code FLEXLAST

    Energy Technology Data Exchange (ETDEWEB)

    Visser, B. [Stork Product Eng., Amsterdam (Netherlands)

    1996-09-01

    To support the discussion on aeroelastic codes, a description of the code FLEXLAST was given and experiences within benchmarks and measurement programmes were summarized. The code FLEXLAST has been developed since 1982 at Stork Product Engineering (SPE). Since 1992 FLEXLAST has been used by Dutch industries for wind turbine and rotor design. Based on the comparison with measurements, it can be concluded that the main shortcomings of wind turbine modelling lie in the field of aerodynamics, wind field and wake modelling. (au)

  2. MORSE Monte Carlo code

    International Nuclear Information System (INIS)

    Cramer, S.N.

    1984-01-01

    The MORSE code is a large general-use multigroup Monte Carlo code system. Although no claims can be made regarding its superiority in either theoretical details or Monte Carlo techniques, MORSE has been, since its inception at ORNL in the late 1960s, the most widely used Monte Carlo radiation transport code. The principal reason for this popularity is that MORSE is relatively easy to use, independent of any installation or distribution center, and it can be easily customized to fit almost any specific need. Features of the MORSE code are described

  3. QR codes for dummies

    CERN Document Server

    Waters, Joe

    2012-01-01

    Find out how to effectively create, use, and track QR codes QR (Quick Response) codes are popping up everywhere, and businesses are reaping the rewards. Get in on the action with the no-nonsense advice in this streamlined, portable guide. You'll find out how to get started, plan your strategy, and actually create the codes. Then you'll learn to link codes to mobile-friendly content, track your results, and develop ways to give your customers value that will keep them coming back. It's all presented in the straightforward style you've come to know and love, with a dash of humor thrown

  4. Tokamak Systems Code

    International Nuclear Information System (INIS)

    Reid, R.L.; Barrett, R.J.; Brown, T.G.

    1985-03-01

    The FEDC Tokamak Systems Code calculates tokamak performance, cost, and configuration as a function of plasma engineering parameters. This version of the code models experimental tokamaks. It does not currently consider tokamak configurations that generate electrical power or incorporate breeding blankets. The code has a modular (or subroutine) structure to allow independent modeling for each major tokamak component or system. A primary benefit of modularization is that a component module may be updated without disturbing the remainder of the systems code as long as the imput to or output from the module remains unchanged

  5. Efficient Coding of Information: Huffman Coding -RE ...

    Indian Academy of Sciences (India)

    to a stream of equally-likely symbols so as to recover the original stream in the event of errors. The for- ... The source-coding problem is one of finding a mapping from U to a ... probability that the random variable X takes the value x written as ...

  6. NR-code: Nonlinear reconstruction code

    Science.gov (United States)

    Yu, Yu; Pen, Ue-Li; Zhu, Hong-Ming

    2018-04-01

    NR-code applies nonlinear reconstruction to the dark matter density field in redshift space and solves for the nonlinear mapping from the initial Lagrangian positions to the final redshift space positions; this reverses the large-scale bulk flows and improves the precision measurement of the baryon acoustic oscillations (BAO) scale.

  7. 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.

  8. 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.

  9. Co-simulation coupling spectral/finite elements for 3D soil/structure interaction problems

    Science.gov (United States)

    Zuchowski, Loïc; Brun, Michael; De Martin, Florent

    2018-05-01

    The coupling between an implicit finite elements (FE) code and an explicit spectral elements (SE) code has been explored for solving the elastic wave propagation in the case of soil/structure interaction problem. The coupling approach is based on domain decomposition methods in transient dynamics. The spatial coupling at the interface is managed by a standard coupling mortar approach, whereas the time integration is dealt with an hybrid asynchronous time integrator. An external coupling software, handling the interface problem, has been set up in order to couple the FE software Code_Aster with the SE software EFISPEC3D.

  10. Research and Design in Unified Coding Architecture for Smart Grids

    Directory of Open Access Journals (Sweden)

    Gang Han

    2013-09-01

    Full Text Available Standardized and sharing information platform is the foundation of the Smart Grids. In order to improve the dispatching center information integration of the power grids and achieve efficient data exchange, sharing and interoperability, a unified coding architecture is proposed. The architecture includes coding management layer, coding generation layer, information models layer and application system layer. Hierarchical design makes the whole coding architecture to adapt to different application environments, different interfaces, loosely coupled requirements, which can realize the integration model management function of the power grids. The life cycle and evaluation method of survival of unified coding architecture is proposed. It can ensure the stability and availability of the coding architecture. Finally, the development direction of coding technology of the Smart Grids in future is prospected.

  11. Synthesizing Certified Code

    Science.gov (United States)

    Whalen, Michael; Schumann, Johann; Fischer, Bernd

    2002-01-01

    Code certification is a lightweight approach to demonstrate software quality on a formal level. Its basic idea is to require producers to provide formal proofs that their code satisfies certain quality properties. These proofs serve as certificates which can be checked independently. Since code certification uses the same underlying technology as program verification, it also requires many detailed annotations (e.g., loop invariants) to make the proofs possible. However, manually adding theses annotations to the code is time-consuming and error-prone. We address this problem by combining code certification with automatic program synthesis. We propose an approach to generate simultaneously, from a high-level specification, code and all annotations required to certify generated code. Here, we describe a certification extension of AUTOBAYES, a synthesis tool which automatically generates complex data analysis programs from compact specifications. AUTOBAYES contains sufficient high-level domain knowledge to generate detailed annotations. This allows us to use a general-purpose verification condition generator to produce a set of proof obligations in first-order logic. The obligations are then discharged using the automated theorem E-SETHEO. We demonstrate our approach by certifying operator safety for a generated iterative data classification program without manual annotation of the code.

  12. Code of Ethics

    Science.gov (United States)

    Division for Early Childhood, Council for Exceptional Children, 2009

    2009-01-01

    The Code of Ethics of the Division for Early Childhood (DEC) of the Council for Exceptional Children is a public statement of principles and practice guidelines supported by the mission of DEC. The foundation of this Code is based on sound ethical reasoning related to professional practice with young children with disabilities and their families…

  13. Interleaved Product LDPC Codes

    OpenAIRE

    Baldi, Marco; Cancellieri, Giovanni; Chiaraluce, Franco

    2011-01-01

    Product LDPC codes take advantage of LDPC decoding algorithms and the high minimum distance of product codes. We propose to add suitable interleavers to improve the waterfall performance of LDPC decoding. Interleaving also reduces the number of low weight codewords, that gives a further advantage in the error floor region.

  14. Insurance billing and coding.

    Science.gov (United States)

    Napier, Rebecca H; Bruelheide, Lori S; Demann, Eric T K; Haug, Richard H

    2008-07-01

    The purpose of this article is to highlight the importance of understanding various numeric and alpha-numeric codes for accurately billing dental and medically related services to private pay or third-party insurance carriers. In the United States, common dental terminology (CDT) codes are most commonly used by dentists to submit claims, whereas current procedural terminology (CPT) and International Classification of Diseases, Ninth Revision, Clinical Modification (ICD.9.CM) codes are more commonly used by physicians to bill for their services. The CPT and ICD.9.CM coding systems complement each other in that CPT codes provide the procedure and service information and ICD.9.CM codes provide the reason or rationale for a particular procedure or service. These codes are more commonly used for "medical necessity" determinations, and general dentists and specialists who routinely perform care, including trauma-related care, biopsies, and dental treatment as a result of or in anticipation of a cancer-related treatment, are likely to use these codes. Claim submissions for care provided can be completed electronically or by means of paper forms.

  15. Error Correcting Codes

    Indian Academy of Sciences (India)

    Science and Automation at ... the Reed-Solomon code contained 223 bytes of data, (a byte ... then you have a data storage system with error correction, that ..... practical codes, storing such a table is infeasible, as it is generally too large.

  16. Scrum Code Camps

    DEFF Research Database (Denmark)

    Pries-Heje, Lene; Pries-Heje, Jan; Dalgaard, Bente

    2013-01-01

    is required. In this paper we present the design of such a new approach, the Scrum Code Camp, which can be used to assess agile team capability in a transparent and consistent way. A design science research approach is used to analyze properties of two instances of the Scrum Code Camp where seven agile teams...

  17. RFQ simulation code

    International Nuclear Information System (INIS)

    Lysenko, W.P.

    1984-04-01

    We have developed the RFQLIB simulation system to provide a means to systematically generate the new versions of radio-frequency quadrupole (RFQ) linac simulation codes that are required by the constantly changing needs of a research environment. This integrated system simplifies keeping track of the various versions of the simulation code and makes it practical to maintain complete and up-to-date documentation. In this scheme, there is a certain standard version of the simulation code that forms a library upon which new versions are built. To generate a new version of the simulation code, the routines to be modified or added are appended to a standard command file, which contains the commands to compile the new routines and link them to the routines in the library. The library itself is rarely changed. Whenever the library is modified, however, this modification is seen by all versions of the simulation code, which actually exist as different versions of the command file. All code is written according to the rules of structured programming. Modularity is enforced by not using COMMON statements, simplifying the relation of the data flow to a hierarchy diagram. Simulation results are similar to those of the PARMTEQ code, as expected, because of the similar physical model. Different capabilities, such as those for generating beams matched in detail to the structure, are available in the new code for help in testing new ideas in designing RFQ linacs

  18. Error Correcting Codes

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 2; Issue 3. Error Correcting Codes - Reed Solomon Codes. Priti Shankar. Series Article Volume 2 Issue 3 March ... Author Affiliations. Priti Shankar1. Department of Computer Science and Automation, Indian Institute of Science, Bangalore 560 012, India ...

  19. 78 FR 18321 - International Code Council: The Update Process for the International Codes and Standards

    Science.gov (United States)

    2013-03-26

    ... Energy Conservation Code. International Existing Building Code. International Fire Code. International... Code. International Property Maintenance Code. International Residential Code. International Swimming Pool and Spa Code International Wildland-Urban Interface Code. International Zoning Code. ICC Standards...

  20. GAROS, an aeroelastic code for coupled fixed-rotating structures

    Energy Technology Data Exchange (ETDEWEB)

    Rees, M. [Aerodyn Energiestyseme GmbH, Rendsburg (Germany); Vollan, A. [Pilatus Flugzeugwerke, Stans (Switzerland)

    1996-09-01

    The GAROS (General Analysis of Rotating Structures) program system has been specially designed to calculate aeroelastic stability and dynamic response of horizontal axis wind energy converters. Nevertheless it is also suitable for the dynamic analysis of helicopter rotors and has been used in the analysis of car bodies taking account of rotating wheels. GAROS was developed over the last 17 years. In the following the mechanical and the aerodynamic model will be discussed in detail. A short overview of the solution methods for the equation of motion in time and frequency domain will ge given. After this one example for the FEM model of the rotor and tower will be discussed. (EG)

  1. Coupling of partitioned physics codes with quasi-Newton methods

    CSIR Research Space (South Africa)

    Haelterman, R

    2017-03-01

    Full Text Available , A class of methods for solving nonlinear simultaneous equations. Math. Comp. 19, pp. 577–593 (1965) [3] C.G. Broyden, Quasi-Newton methods and their applications to function minimization. Math. Comp. 21, pp. 368–381 (1967) [4] J.E. Dennis, J.J. More...´, Quasi-Newton methods: motivation and theory. SIAM Rev. 19, pp. 46–89 (1977) [5] J.E. Dennis, R.B. Schnabel, Least Change Secant Updates for quasi- Newton methods. SIAM Rev. 21, pp. 443–459 (1979) [6] G. Dhondt, CalculiX CrunchiX USER’S MANUAL Version 2...

  2. Validation of thermalhydraulic codes

    International Nuclear Information System (INIS)

    Wilkie, D.

    1992-01-01

    Thermalhydraulic codes require to be validated against experimental data collected over a wide range of situations if they are to be relied upon. A good example is provided by the nuclear industry where codes are used for safety studies and for determining operating conditions. Errors in the codes could lead to financial penalties, to the incorrect estimation of the consequences of accidents and even to the accidents themselves. Comparison between prediction and experiment is often described qualitatively or in approximate terms, e.g. ''agreement is within 10%''. A quantitative method is preferable, especially when several competing codes are available. The codes can then be ranked in order of merit. Such a method is described. (Author)

  3. Fracture flow code

    International Nuclear Information System (INIS)

    Dershowitz, W; Herbert, A.; Long, J.

    1989-03-01

    The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)

  4. Dark coupling

    International Nuclear Information System (INIS)

    Gavela, M.B.; Hernández, D.; Honorez, L. Lopez; Mena, O.; Rigolin, S.

    2009-01-01

    The two dark sectors of the universe—dark matter and dark energy—may interact with each other. Background and linear density perturbation evolution equations are developed for a generic coupling. We then establish the general conditions necessary to obtain models free from non-adiabatic instabilities. As an application, we consider a viable universe in which the interaction strength is proportional to the dark energy density. The scenario does not exhibit ''phantom crossing'' and is free from instabilities, including early ones. A sizeable interaction strength is compatible with combined WMAP, HST, SN, LSS and H(z) data. Neutrino mass and/or cosmic curvature are allowed to be larger than in non-interacting models. Our analysis sheds light as well on unstable scenarios previously proposed

  5. Review of SKB's Code Documentation and Testing

    International Nuclear Information System (INIS)

    Hicks, T.W.

    2005-01-01

    SKB is in the process of developing the SR-Can safety assessment for a KBS 3 repository. The assessment will be based on quantitative analyses using a range of computational codes aimed at developing an understanding of how the repository system will evolve. Clear and comprehensive code documentation and testing will engender confidence in the results of the safety assessment calculations. This report presents the results of a review undertaken on behalf of SKI aimed at providing an understanding of how codes used in the SR 97 safety assessment and those planned for use in the SR-Can safety assessment have been documented and tested. Having identified the codes us ed by SKB, several codes were selected for review. Consideration was given to codes used directly in SKB's safety assessment calculations as well as to some of the less visible codes that are important in quantifying the different repository barrier safety functions. SKB's documentation and testing of the following codes were reviewed: COMP23 - a near-field radionuclide transport model developed by SKB for use in safety assessment calculations. FARF31 - a far-field radionuclide transport model developed by SKB for use in safety assessment calculations. PROPER - SKB's harness for executing probabilistic radionuclide transport calculations using COMP23 and FARF31. The integrated analytical radionuclide transport model that SKB has developed to run in parallel with COMP23 and FARF31. CONNECTFLOW - a discrete fracture network model/continuum model developed by Serco Assurance (based on the coupling of NAMMU and NAPSAC), which SKB is using to combine hydrogeological modelling on the site and regional scales in place of the HYDRASTAR code. DarcyTools - a discrete fracture network model coupled to a continuum model, recently developed by SKB for hydrogeological modelling, also in place of HYDRASTAR. ABAQUS - a finite element material model developed by ABAQUS, Inc, which is used by SKB to model repository buffer

  6. Huffman coding in advanced audio coding standard

    Science.gov (United States)

    Brzuchalski, Grzegorz

    2012-05-01

    This article presents several hardware architectures of Advanced Audio Coding (AAC) Huffman noiseless encoder, its optimisations and working implementation. Much attention has been paid to optimise the demand of hardware resources especially memory size. The aim of design was to get as short binary stream as possible in this standard. The Huffman encoder with whole audio-video system has been implemented in FPGA devices.

  7. Hamor-2: a computer code for LWR inventory calculation

    International Nuclear Information System (INIS)

    Guimaraes, L.N.F.; Marzo, M.A.S.

    1985-01-01

    A method for calculating the accuracy inventory of LWR reactors is presented. This method uses the Hamor-2 computer code. Hamor-2 is obtained from the coupling of two other computer codes Hammer-Techion and Origen-2 for testing Hamor-2, its results were compared to concentration values measured from activides of two PWR reactors; Kernkraftwerk Obrighein (KWO) and H.B. Robinson (HBR). These actinides are U 235 , U 236 , U 238 , Pu 239 , Pu 241 and PU 242 . The computer code Hammor-2 shows better results than the computer code Origem-2, when both are compared with experimental results. (E.G.) [pt

  8. CBP Phase I Code Integration

    International Nuclear Information System (INIS)

    Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

    2011-01-01

    was developed to link GoldSim with external codes (Smith III et al. 2010). The DLL uses a list of code inputs provided by GoldSim to create an input file for the external application, runs the external code, and returns a list of outputs (read from files created by the external application) back to GoldSim. In this way GoldSim provides: (1) a unified user interface to the applications, (2) the capability of coupling selected codes in a synergistic manner, and (3) the capability of performing probabilistic uncertainty analysis with the codes. GoldSim is made available by the GoldSim Technology Group as a free 'Player' version that allows running but not editing GoldSim models. The player version makes the software readily available to a wider community of users that would wish to use the CBP application but do not have a license for GoldSim.

  9. CBP PHASE I CODE INTEGRATION

    Energy Technology Data Exchange (ETDEWEB)

    Smith, F.; Brown, K.; Flach, G.; Sarkar, S.

    2011-09-30

    developed to link GoldSim with external codes (Smith III et al. 2010). The DLL uses a list of code inputs provided by GoldSim to create an input file for the external application, runs the external code, and returns a list of outputs (read from files created by the external application) back to GoldSim. In this way GoldSim provides: (1) a unified user interface to the applications, (2) the capability of coupling selected codes in a synergistic manner, and (3) the capability of performing probabilistic uncertainty analysis with the codes. GoldSim is made available by the GoldSim Technology Group as a free 'Player' version that allows running but not editing GoldSim models. The player version makes the software readily available to a wider community of users that would wish to use the CBP application but do not have a license for GoldSim.

  10. Scalable Coupling of Multi-Scale AEH and PARADYN Impact Analyses

    National Research Council Canada - National Science Library

    Valisetty, Rama; Cheung, Peter; Namburu, Raju

    2004-01-01

    .... An asymptotic expansion homogenization (AEH) based micro-structural code available for modeling micro-structural aspects of modern armor materials is coupled with PARADYN, a parallel explicit Lagrangian finite element code...

  11. Report number codes

    Energy Technology Data Exchange (ETDEWEB)

    Nelson, R.N. (ed.)

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name.

  12. Report number codes

    International Nuclear Information System (INIS)

    Nelson, R.N.

    1985-05-01

    This publication lists all report number codes processed by the Office of Scientific and Technical Information. The report codes are substantially based on the American National Standards Institute, Standard Technical Report Number (STRN)-Format and Creation Z39.23-1983. The Standard Technical Report Number (STRN) provides one of the primary methods of identifying a specific technical report. The STRN consists of two parts: The report code and the sequential number. The report code identifies the issuing organization, a specific program, or a type of document. The sequential number, which is assigned in sequence by each report issuing entity, is not included in this publication. Part I of this compilation is alphabetized by report codes followed by issuing installations. Part II lists the issuing organization followed by the assigned report code(s). In both Parts I and II, the names of issuing organizations appear for the most part in the form used at the time the reports were issued. However, for some of the more prolific installations which have had name changes, all entries have been merged under the current name

  13. Status report on SHARP coupling framework.

    Energy Technology Data Exchange (ETDEWEB)

    Caceres, A.; Tautges, T. J.; Lottes, J.; Fischer, P.; Rabiti, C.; Smith, M. A.; Siegel, A.; Yang, W. S.; Palmiotti, G.

    2008-05-30

    This report presents the software engineering effort under way at ANL towards a comprehensive integrated computational framework (SHARP) for high fidelity simulations of sodium cooled fast reactors. The primary objective of this framework is to provide accurate and flexible analysis tools to nuclear reactor designers by simulating multiphysics phenomena happening in complex reactor geometries. Ideally, the coupling among different physics modules (such as neutronics, thermal-hydraulics, and structural mechanics) needs to be tight to preserve the accuracy achieved in each module. However, fast reactor cores in steady state mode represent a special case where weak coupling between neutronics and thermal-hydraulics is usually adequate. Our framework design allows for both options. Another requirement for SHARP framework has been to implement various coupling algorithms that are parallel and scalable to large scale since nuclear reactor core simulations are among the most memory and computationally intensive, requiring the use of leadership-class petascale platforms. This report details our progress toward achieving these goals. Specifically, we demonstrate coupling independently developed parallel codes in a manner that does not compromise performance or portability, while minimizing the impact on individual developers. This year, our focus has been on developing a lightweight and loosely coupled framework targeted at UNIC (our neutronics code) and Nek (our thermal hydraulics code). However, the framework design is not limited to just using these two codes.

  14. Finite volume thermal-hydraulics and neutronics coupled calculations - 15300

    International Nuclear Information System (INIS)

    Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.

    2015-01-01

    The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly

  15. Cryptography cracking codes

    CERN Document Server

    2014-01-01

    While cracking a code might seem like something few of us would encounter in our daily lives, it is actually far more prevalent than we may realize. Anyone who has had personal information taken because of a hacked email account can understand the need for cryptography and the importance of encryption-essentially the need to code information to keep it safe. This detailed volume examines the logic and science behind various ciphers, their real world uses, how codes can be broken, and the use of technology in this oft-overlooked field.

  16. Coded Splitting Tree Protocols

    DEFF Research Database (Denmark)

    Sørensen, Jesper Hemming; Stefanovic, Cedomir; Popovski, Petar

    2013-01-01

    This paper presents a novel approach to multiple access control called coded splitting tree protocol. The approach builds on the known tree splitting protocols, code structure and successive interference cancellation (SIC). Several instances of the tree splitting protocol are initiated, each...... instance is terminated prematurely and subsequently iterated. The combined set of leaves from all the tree instances can then be viewed as a graph code, which is decodable using belief propagation. The main design problem is determining the order of splitting, which enables successful decoding as early...

  17. Transport theory and codes

    International Nuclear Information System (INIS)

    Clancy, B.E.

    1986-01-01

    This chapter begins with a neutron transport equation which includes the one dimensional plane geometry problems, the one dimensional spherical geometry problems, and numerical solutions. The section on the ANISN code and its look-alikes covers problems which can be solved; eigenvalue problems; outer iteration loop; inner iteration loop; and finite difference solution procedures. The input and output data for ANISN is also discussed. Two dimensional problems such as the DOT code are given. Finally, an overview of the Monte-Carlo methods and codes are elaborated on

  18. Gravity inversion code

    International Nuclear Information System (INIS)

    Burkhard, N.R.

    1979-01-01

    The gravity inversion code applies stabilized linear inverse theory to determine the topography of a subsurface density anomaly from Bouguer gravity data. The gravity inversion program consists of four source codes: SEARCH, TREND, INVERT, and AVERAGE. TREND and INVERT are used iteratively to converge on a solution. SEARCH forms the input gravity data files for Nevada Test Site data. AVERAGE performs a covariance analysis on the solution. This document describes the necessary input files and the proper operation of the code. 2 figures, 2 tables

  19. Models and applications of the UEDGE code

    International Nuclear Information System (INIS)

    Rensink, M.E.; Knoll, D.A.; Porter, G.D.; Rognlien, T.D.; Smith, G.R.; Wising, F.

    1996-09-01

    The transport of particles and energy from the core of a tokamak to nearby material surfaces is an important problem for understanding present experiments and for designing reactor-grade devices. A number of fluid transport codes have been developed to model the plasma in the edge and scrape-off layer (SOL) regions. This report will focus on recent model improvements and illustrative results from the UEDGE code. Some geometric and mesh considerations are introduced, followed by a general description of the plasma and neutral fluid models. A few comments on computational issues are given and then two important applications are illustrated concerning benchmarking and the ITER radiative divertor. Finally, we report on some recent work to improve the models in UEDGE by coupling to a Monte Carlo neutrals code and by utilizing an adaptive grid

  20. Validation of the reactor dynamics code TRAB

    International Nuclear Information System (INIS)

    Raety, H.; Kyrki-Rajamaeki, R.; Rajamaeki, M.

    1991-05-01

    The one-dimensional reactor dynamics code TRAB (Transient Analysis code for BWRs) developed at VTT was originally designed for BWR analyses, but it can in its present version be used for various modelling purposes. The core model of TRAB can be used separately for LWR calculations. For PWR modelling the core model of TRAB has been coupled to circuit model SMABRE to form the SMATRA code. The versatile modelling capabilities of TRAB have been utilized also in analyses of e.g. the heating reactor SECURE and the RBMK-type reactor (Chernobyl). The report summarizes the extensive validation of TRAB. TRAB has been validated with benchmark problems, comparative calculations against independent analyses, analyses of start-up experiments of nuclear power plants and real plant transients. Comparative RBMES type reactor calculations have been made against Soviet simulations and the initial power excursion of the Chernobyl reactor accident has also been calculated with TRAB

  1. A computer code to design liquid containers for vehicles

    International Nuclear Information System (INIS)

    Parizi, H.B.; Fard, M.P.; Dolatabadi, A.

    2003-01-01

    We are presenting the development of a modular code for the simulation of the fluid sloshing that occurs in the liquid containers in vehicles. Sloshing occurs when a partially filled container of liquid goes through transient or steady external forces. Under such conditions, the free surface of the liquid may move and the liquid may impact on the walls of the container, exchanging forces. These forces may cause numerous harmful and undesirable consequences in the operation of the vehicle, such as vehicle turn over. The fluid mechanic equations that describe the fluid sloshing in the container and the dynamic equations that describe the movement of the container are solved separately in two different codes. The codes are coupled weekly, such that the output of one code will be used as the input to the other code in the same time step. The outputs of the fluid code are the forces and torques that are applied to the body of the container due to sloshing, whereas the output of the dynamic code are the translational and rotational velocities and accelerations of the container. The proposed software can be used to test the performance of the designed container under various operating condition and allow effective improvements to the container design. The proposed code is different than the presently available codes, in that it will provide a true simulation of the coupled fluid and structure interaction. (author)

  2. Recent developments in KTF. Code optimization and improved numerics

    International Nuclear Information System (INIS)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin

    2012-01-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  3. Recent developments in KTF. Code optimization and improved numerics

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez, Javier; Avramova, Maria; Sanchez, Victor Hugo; Ivanov, Kostadin [Karlsruhe Institute of Technology (KIT) (Germany). Inst. for Neutron Physics and Reactor Technology (INR)

    2012-11-01

    The rapid increase of computer power in the last decade facilitated the development of high fidelity simulations in nuclear engineering allowing a more realistic and accurate optimization as well as safety assessment of reactor cores and power plants compared to the legacy codes. Thermal hydraulic subchannel codes together with time dependent neutron transport codes are the options of choice for an accurate prediction of local safety parameters. Moreover, fast running codes with the best physical models are needed for high fidelity coupled thermal hydraulic / neutron kinetic solutions. Hence at KIT, different subchannel codes such as SUBCHANFLOW and KTF are being improved, validated and coupled with different neutron kinetics solutions. KTF is a subchannel code developed for best-estimate analysis of both Pressurized Water Reactor (PWR) and BWR. It is based on the Pennsylvania State University (PSU) version of COBRA-TF (Coolant Boling in Rod Arrays Two Fluids) named CTF. In this paper, the investigations devoted to the enhancement of the code numeric and informatics structure are presented and discussed. By some examples the gain on code speed-up will be demonstrated and finally an outlook of further activities concentrated on the code improvements will be given. (orig.)

  4. Calculation of fluid-structure interaction for reactor safety with the Cassiopee code

    International Nuclear Information System (INIS)

    Graveleau, J.L.; Louvet, P.D.

    1979-01-01

    The cassiopee code is an eulerian-lagrangian coupled code for computations where the hydrodynamic is coupled with structural domains. It is completely explicit. The fluid zones may be computed either in lagrangian or in eulerian coordinates; thin shells can be computed wih their flexural behaviour; elastic plastic zones must be calculated in a lagrangian way. This code is under development in Cadarache. Its purpose is to compute the hypothetical core disruptive accident of a LMFBR when lagrangian codes are not sufficient. This paper contains a description of the code and two examples of computations, one of which has been compared with experimental results

  5. Fulcrum Network Codes

    DEFF Research Database (Denmark)

    2015-01-01

    Fulcrum network codes, which are a network coding framework, achieve three objectives: (i) to reduce the overhead per coded packet to almost 1 bit per source packet; (ii) to operate the network using only low field size operations at intermediate nodes, dramatically reducing complexity...... in the network; and (iii) to deliver an end-to-end performance that is close to that of a high field size network coding system for high-end receivers while simultaneously catering to low-end ones that can only decode in a lower field size. Sources may encode using a high field size expansion to increase...... the number of dimensions seen by the network using a linear mapping. Receivers can tradeoff computational effort with network delay, decoding in the high field size, the low field size, or a combination thereof....

  6. Supervised Convolutional Sparse Coding

    KAUST Repository

    Affara, Lama Ahmed; Ghanem, Bernard; Wonka, Peter

    2018-01-01

    coding, which aims at learning discriminative dictionaries instead of purely reconstructive ones. We incorporate a supervised regularization term into the traditional unsupervised CSC objective to encourage the final dictionary elements

  7. SASSYS LMFBR systems code

    International Nuclear Information System (INIS)

    Dunn, F.E.; Prohammer, F.G.; Weber, D.P.

    1983-01-01

    The SASSYS LMFBR systems analysis code is being developed mainly to analyze the behavior of the shut-down heat-removal system and the consequences of failures in the system, although it is also capable of analyzing a wide range of transients, from mild operational transients through more severe transients leading to sodium boiling in the core and possible melting of clad and fuel. The code includes a detailed SAS4A multi-channel core treatment plus a general thermal-hydraulic treatment of the primary and intermediate heat-transport loops and the steam generators. The code can handle any LMFBR design, loop or pool, with an arbitrary arrangement of components. The code is fast running: usually faster than real time

  8. OCA Code Enforcement

    Data.gov (United States)

    Montgomery County of Maryland — The Office of the County Attorney (OCA) processes Code Violation Citations issued by County agencies. The citations can be viewed by issued department, issued date...

  9. The fast code

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, L.N.; Wilson, R.E. [Oregon State Univ., Dept. of Mechanical Engineering, Corvallis, OR (United States)

    1996-09-01

    The FAST Code which is capable of determining structural loads on a flexible, teetering, horizontal axis wind turbine is described and comparisons of calculated loads with test data are given at two wind speeds for the ESI-80. The FAST Code models a two-bladed HAWT with degrees of freedom for blade bending, teeter, drive train flexibility, yaw, and windwise and crosswind tower motion. The code allows blade dimensions, stiffnesses, and weights to differ and models tower shadow, wind shear, and turbulence. Additionally, dynamic stall is included as are delta-3 and an underslung rotor. Load comparisons are made with ESI-80 test data in the form of power spectral density, rainflow counting, occurrence histograms, and azimuth averaged bin plots. It is concluded that agreement between the FAST Code and test results is good. (au)

  10. Code Disentanglement: Initial Plan

    Energy Technology Data Exchange (ETDEWEB)

    Wohlbier, John Greaton [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelley, Timothy M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rockefeller, Gabriel M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Calef, Matthew Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-01-27

    The first step to making more ambitious changes in the EAP code base is to disentangle the code into a set of independent, levelized packages. We define a package as a collection of code, most often across a set of files, that provides a defined set of functionality; a package a) can be built and tested as an entity and b) fits within an overall levelization design. Each package contributes one or more libraries, or an application that uses the other libraries. A package set is levelized if the relationships between packages form a directed, acyclic graph and each package uses only packages at lower levels of the diagram (in Fortran this relationship is often describable by the use relationship between modules). Independent packages permit independent- and therefore parallel|development. The packages form separable units for the purposes of development and testing. This is a proven path for enabling finer-grained changes to a complex code.

  11. Induction technology optimization code

    International Nuclear Information System (INIS)

    Caporaso, G.J.; Brooks, A.L.; Kirbie, H.C.

    1992-01-01

    A code has been developed to evaluate relative costs of induction accelerator driver systems for relativistic klystrons. The code incorporates beam generation, transport and pulsed power system constraints to provide an integrated design tool. The code generates an injector/accelerator combination which satisfies the top level requirements and all system constraints once a small number of design choices have been specified (rise time of the injector voltage and aspect ratio of the ferrite induction cores, for example). The code calculates dimensions of accelerator mechanical assemblies and values of all electrical components. Cost factors for machined parts, raw materials and components are applied to yield a total system cost. These costs are then plotted as a function of the two design choices to enable selection of an optimum design based on various criteria. (Author) 11 refs., 3 figs

  12. VT ZIP Code Areas

    Data.gov (United States)

    Vermont Center for Geographic Information — (Link to Metadata) A ZIP Code Tabulation Area (ZCTA) is a statistical geographic entity that approximates the delivery area for a U.S. Postal Service five-digit...

  13. Bandwidth efficient coding

    CERN Document Server

    Anderson, John B

    2017-01-01

    Bandwidth Efficient Coding addresses the major challenge in communication engineering today: how to communicate more bits of information in the same radio spectrum. Energy and bandwidth are needed to transmit bits, and bandwidth affects capacity the most. Methods have been developed that are ten times as energy efficient at a given bandwidth consumption as simple methods. These employ signals with very complex patterns and are called "coding" solutions. The book begins with classical theory before introducing new techniques that combine older methods of error correction coding and radio transmission in order to create narrowband methods that are as efficient in both spectrum and energy as nature allows. Other topics covered include modulation techniques such as CPM, coded QAM and pulse design.

  14. Reactor lattice codes

    International Nuclear Information System (INIS)

    Kulikowska, T.

    2001-01-01

    The description of reactor lattice codes is carried out on the example of the WIMSD-5B code. The WIMS code in its various version is the most recognised lattice code. It is used in all parts of the world for calculations of research and power reactors. The version WIMSD-5B is distributed free of charge by NEA Data Bank. The description of its main features given in the present lecture follows the aspects defined previously for lattice calculations in the lecture on Reactor Lattice Transport Calculations. The spatial models are described, and the approach to the energy treatment is given. Finally the specific algorithm applied in fuel depletion calculations is outlined. (author)

  15. Critical Care Coding for Neurologists.

    Science.gov (United States)

    Nuwer, Marc R; Vespa, Paul M

    2015-10-01

    Accurate coding is an important function of neurologic practice. This contribution to Continuum is part of an ongoing series that presents helpful coding information along with examples related to the issue topic. Tips for diagnosis coding, Evaluation and Management coding, procedure coding, or a combination are presented, depending on which is most applicable to the subject area of the issue.

  16. Lattice Index Coding

    OpenAIRE

    Natarajan, Lakshmi; Hong, Yi; Viterbo, Emanuele

    2014-01-01

    The index coding problem involves a sender with K messages to be transmitted across a broadcast channel, and a set of receivers each of which demands a subset of the K messages while having prior knowledge of a different subset as side information. We consider the specific case of noisy index coding where the broadcast channel is Gaussian and every receiver demands all the messages from the source. Instances of this communication problem arise in wireless relay networks, sensor networks, and ...