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.
Module type plant system dynamics analysis code (MSG-COPD). Code manual
International Nuclear Information System (INIS)
Sakai, Takaaki
2002-11-01
MSG-COPD is a module type plant system dynamics analysis code which involves a multi-dimensional thermal-hydraulics calculation module to analyze pool type of fast breeder reactors. Explanations of each module and the methods for the input data are described in this code manual. (author)
Developments of HTGR thermofluid dynamic analysis codes and HTGR plant dynamic simulation code
International Nuclear Information System (INIS)
Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi
1983-01-01
In nuclear power plants as well as high temperature gas-cooled reactor plants, the design is mostly performed on the basis of the results after their characteristics have been grasped by carrying out the numerical simulation using the analysis code. Also in Kawasaki Heavy Industries Ltd., on the basis of the system engineering accumulated with gas-cooled reactors since several years ago, the preparation and systematization of analysis codes have been advanced, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In this report, a part of the results is described. The example of the analysis applying the two-dimensional compressible flow analysis codes SOLA-VOF and SALE-2D, which were developed by Los Alamos National Laboratory in USA and modified for use in Kawasaki, to HTGR system is reported. Besides, Kawasaki has developed the control characteristics analyzing code DYSCO by which the change of system composition is easy and high versatility is available. The outline, fundamental equations, fundamental algorithms and examples of application of the SOLA-VOF and SALE-2D, the present status of system characteristic simulation codes and the outline of the DYSCO are described. (Kako, I.)
Computer codes for beam dynamics analysis of cyclotronlike accelerators
Smirnov, V.
2017-12-01
Computer codes suitable for the study of beam dynamics in cyclotronlike (classical and isochronous cyclotrons, synchrocyclotrons, and fixed field alternating gradient) accelerators are reviewed. Computer modeling of cyclotron segments, such as the central zone, acceleration region, and extraction system is considered. The author does not claim to give a full and detailed description of the methods and algorithms used in the codes. Special attention is paid to the codes already proven and confirmed at the existing accelerating facilities. The description of the programs prepared in the worldwide known accelerator centers is provided. The basic features of the programs available to users and limitations of their applicability are described.
BWR plant dynamic analysis code BWRDYN user's manual
International Nuclear Information System (INIS)
Yokobayashi, Masao; Yoshida, Kazuo; Fujiki, Kazuo
1989-06-01
Computer code BWRDYN has been developed for thermal-hydraulic analysis of a BWR plant. It can analyze the various types of transient caused by not only small but also large disturbances such as operating mode changes and/or system malfunctions. The verification of main analytical models of the BWRDYN code has been performed with measured data of actual BWR plant. Furthermore, the installation of BOP (Balance of Plant) model has made it possible to analyze the effect of BOP on reactor system. This report describes on analytical models and instructions for user of the BWRDYN code. (author)
Convolutional Sparse Coding for Static and Dynamic Images Analysis
Directory of Open Access Journals (Sweden)
B. A. Knyazev
2014-01-01
Full Text Available The objective of this work is to improve performance of static and dynamic objects recognition. For this purpose a new image representation model and a transformation algorithm are proposed. It is examined and illustrated that limitations of previous methods make it difficult to achieve this objective. Static images, specifically handwritten digits of the widely used MNIST dataset, are the primary focus of this work. Nevertheless, preliminary qualitative results of image sequences analysis based on the suggested model are presented.A general analytical form of the Gabor function, often employed to generate filters, is described and discussed. In this research, this description is required for computing parameters of responses returned by our algorithm. The recursive convolution operator is introduced, which allows extracting free shape features of visual objects. The developed parametric representation model is compared with sparse coding based on energy function minimization.In the experimental part of this work, errors of estimating the parameters of responses are determined. Also, parameters statistics and their correlation coefficients for more than 106 responses extracted from the MNIST dataset are calculated. It is demonstrated that these data correspond well with previous research studies on Gabor filters as well as with works on visual cortex primary cells of mammals, in which similar responses were observed. A comparative test of the developed model with three other approaches is conducted; speed and accuracy scores of handwritten digits classification are presented. A support vector machine with a linear or radial basic function is used for classification of images and their representations while principal component analysis is used in some cases to prepare data beforehand. High accuracy is not attained due to the specific difficulties of combining our model with a support vector machine (a 3.99% error rate. However, another method is
Energy Technology Data Exchange (ETDEWEB)
Picard, Richard Roy [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bhat, Kabekode Ghanasham [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-07-18
We examine sensitivity analysis and uncertainty quantification for molecular dynamics simulation. Extreme (large or small) output values for the LAMMPS code often occur at the boundaries of input regions, and uncertainties in those boundary values are overlooked by common SA methods. Similarly, input values for which code outputs are consistent with calibration data can also occur near boundaries. Upon applying approaches in the literature for imprecise probabilities (IPs), much more realistic results are obtained than for the complacent application of standard SA and code calibration.
Interface between computational fluid dynamics (CFD) and plant analysis computer codes
International Nuclear Information System (INIS)
Coffield, R.D.; Dunckhorst, F.F.; Tomlinson, E.T.; Welch, J.W.
1993-01-01
Computational fluid dynamics (CFD) can provide valuable input to the development of advanced plant analysis computer codes. The types of interfacing discussed in this paper will directly contribute to modeling and accuracy improvements throughout the plant system and should result in significant reduction of design conservatisms that have been applied to such analyses in the past
Energy Technology Data Exchange (ETDEWEB)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard, E-mail: B.L.Sjenitzer@TUDelft.nl, E-mail: J.E.Hoogenboom@TUDelft.nl [Delft University of Technology (Netherlands)
2011-07-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
International Nuclear Information System (INIS)
Sjenitzer, Bart L.; Hoogenboom, J. Eduard
2011-01-01
A new Dynamic Monte Carlo method is implemented in the general purpose Monte Carlo code Tripoli 4.6.1. With this new method incorporated, a general purpose code can be used for safety transient analysis, such as the movement of a control rod or in an accident scenario. To make the Tripoli code ready for calculating on dynamic systems, the Tripoli scheme had to be altered to incorporate time steps, to include the simulation of delayed neutron precursors and to simulate prompt neutron chains. The modified Tripoli code is tested on two sample cases, a steady-state system and a subcritical system and the resulting neutron fluxes behave just as expected. The steady-state calculation has a constant neutron flux over time and this result shows the stability of the calculation. The neutron flux stays constant with acceptable variance. This also shows that the starting conditions are determined correctly. The sub-critical case shows that the code can also handle dynamic systems with a varying neutron flux. (author)
Development of a CFD Code for Analysis of Fluid Dynamic Forces in Seals
Athavale, Mahesh M.; Przekwas, Andrzej J.; Singhal, Ashok K.
1991-01-01
The aim is to develop a 3-D computational fluid dynamics (CFD) code for the analysis of fluid flow in cylindrical seals and evaluation of the dynamic forces on the seals. This code is expected to serve as a scientific tool for detailed flow analysis as well as a check for the accuracy of the 2D industrial codes. The features necessary in the CFD code are outlined. The initial focus was to develop or modify and implement new techniques and physical models. These include collocated grid formulation, rotating coordinate frames and moving grid formulation. Other advanced numerical techniques include higher order spatial and temporal differencing and an efficient linear equation solver. These techniques were implemented in a 2D flow solver for initial testing. Several benchmark test cases were computed using the 2D code, and the results of these were compared to analytical solutions or experimental data to check the accuracy. Tests presented here include planar wedge flow, flow due to an enclosed rotor, and flow in a 2D seal with a whirling rotor. Comparisons between numerical and experimental results for an annular seal and a 7-cavity labyrinth seal are also included.
Development of an advanced fluid-dynamic analysis code: α-flow
International Nuclear Information System (INIS)
Akiyama, Mamoru
1990-01-01
A Project for development of large scale three-dimensional fluid-dynamic analysis code, α-FLOW, coping with the recent advancement of supercomputers and workstations, has been in progress. This project is called the α-Project, which has been promoted by the Association for Large Scale Fluid Dynamics Analysis Code comprising private companies and research institutions such as universities. The developmental period for the α-FLOW is four years, March 1989 to March 1992. To date, the major portions of basic design and program preparation have been completed and the project is in the stage of testing each module. In this paper, the present status of the α-Project, design policy and outline of α-FLOW are described. (author)
Code Pulse: Software Assurance (SWA) Visual Analytics for Dynamic Analysis of Code
2014-09-01
31 4.5.1 Market Analysis...competitive market analysis to assess the tool potential. The final transition targets were selected and expressed along with our research on the topic...public release milestones. Details of our testing methodology is in our Software Test Plan deliv- erable, CP- STP -0001. A summary of this approach is
SALT [System Analysis Language Translater]: A steady state and dynamic systems code
International Nuclear Information System (INIS)
Berry, G.; Geyer, H.
1983-01-01
SALT (System Analysis Language Translater) is a lumped parameter approach to system analysis which is totally modular. The modules are all precompiled and only the main program, which is generated by SALT, needs to be compiled for each unique system configuration. This is a departure from other lumped parameter codes where all models are written by MACROS and then compiled for each unique configuration, usually after all of the models are lumped together and sorted to eliminate undetermined variables. The SALT code contains a robust and sophisticated steady-sate finder (non-linear equation solver), optimization capability and enhanced GEAR integration scheme which makes use of sparsity and algebraic constraints. The SALT systems code has been used for various technologies. The code was originally developed for open-cycle magnetohydrodynamic (MHD) systems. It was easily extended to liquid metal MHD systems by simply adding the appropriate models and property libraries. Similarly, the model and property libraries were expanded to handle fuel cell systems, flue gas desulfurization systems, combined cycle gasification systems, fluidized bed combustion systems, ocean thermal energy conversion systems, geothermal systems, nuclear systems, and conventional coal-fired power plants. Obviously, the SALT systems code is extremely flexible to be able to handle all of these diverse systems. At present, the dynamic option has only been used for LMFBR nuclear power plants and geothermal power plants. However, it can easily be extended to other systems and can be used for analyzing control problems. 12 refs
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)
International Nuclear Information System (INIS)
Nakamura, Yasunori; Tam, Nguyen Ngoc; Ohata, Tomiso; Morita, Kiminori; Nakamachi, Eiji
2004-01-01
The crystallographic texture evolution induced by plastic deformation in the sheet metal forming process has a great influence on its formability. In the present study, a dynamic explicit finite element (FE) analysis code is newly developed by introducing a crystallographic homogenization method to estimate the polycrystalline sheet metal formability, such as the extreme thinning and 'earing'. This code can predict the plastic deformation induced texture evolution at the micro scale and the plastic anisotropy at the macro scale, simultaneously. This multi-scale analysis can couple the microscopic crystal plasticity inhomogeneous deformation with the macroscopic continuum deformation. In this homogenization process, the stress at the macro scale is defined by the volume average of those of the corresponding microscopic crystal aggregations in satisfying the equation of motion and compatibility condition in the micro scale 'unit cell', where the periodicity of deformation is satisfied. This homogenization algorithm is implemented in the conventional dynamic explicit finite element code by employing the updated Lagrangian formulation and the rate type elastic/viscoplastic constitutive equation.At first, it has been confirmed through a texture evolution analyses in cases of typical deformation modes that Taylor's 'constant strain homogenization algorithm' yields extreme concentration toward the preferred crystal orientations compared with our homogenization one. Second, we study the plastic anisotropy effects on 'earing' in the hemispherical cup deep drawing process of pure ferrite phase sheet metal. By the comparison of analytical results with those of Taylor's assumption, conclusions are drawn that the present newly developed dynamic explicit crystallographic homogenization FEM shows more reasonable prediction of plastic deformation induced texture evolution and plastic anisotropy at the macro scale
International Nuclear Information System (INIS)
Lundsager, P.; Krenk, S.
1975-08-01
The static and dynamic response of a cylindrical/ spherical containment to a Boeing 720 impact is computed using 3 different linear elastic computer codes: FINEL, SAP and STARDYNE. Stress and displacement fields are shown together with time histories for a point in the impact zone. The main conclusions from this study are: - In this case the maximum dynamic load factors for stress and displacements were close to 1, but a static analysis alone is not fully sufficient. - More realistic load time histories should be considered. - The main effects seem to be local. The present study does not indicate general collapse from elastic stresses alone. - Further study of material properties at high rates is needed. (author)
Structural dynamics in LMFBR containment analysis. A brief survey of computational methods and codes
International Nuclear Information System (INIS)
Chang, Y.W.
1977-01-01
This paper gives a brief survey of the computational methods and codes available for LMFBR containment analysis. The various numerical methods commonly used in the computer codes are compared. It provides the reactor engineers to up-to-date information on the development of structural dynamics in LMFBR containment analysis. It can also be used as a basis for the selection of the numerical method in the future code development. First, the commonly used finite-difference expressions in the Lagrangian codes will be compared. Sample calculations will be used as a basis for discussing and comparing the accuracy of the various finite-difference representations. The distortion of the meshes will also be compared; the techniques used for eliminating the numerical instabilities will be discussed and compared using examples. Next, the numerical methods used in the Eulerian formulation will be compared, first among themselves and then with the Lagrangian formulations. Special emphasis is placed on the effect of mass diffusion of the Eulerian calculation on the propagation of discontinuities. Implicit and explicit numerical integrations will be discussed and results obtained from these two techniques will be compared. Then, the finite-element methods are compared with the finite-difference methods. The advantages and disadvantages of the two methods will be discussed in detail, together with the versatility and ease of application of the method to containment analysis having complex geometries. It will also be shown that the finite-element equations for a constant-pressure fluid element is identical to the finite-difference equations using contour integrations. Finally, conclusions based on this study will be given
The NEST Dry-Run Mode: Efficient Dynamic Analysis of Neuronal Network Simulation Code
Directory of Open Access Journals (Sweden)
Susanne Kunkel
2017-06-01
Full Text Available NEST is a simulator for spiking neuronal networks that commits to a general purpose approach: It allows for high flexibility in the design of network models, and its applications range from small-scale simulations on laptops to brain-scale simulations on supercomputers. Hence, developers need to test their code for various use cases and ensure that changes to code do not impair scalability. However, running a full set of benchmarks on a supercomputer takes up precious compute-time resources and can entail long queuing times. Here, we present the NEST dry-run mode, which enables comprehensive dynamic code analysis without requiring access to high-performance computing facilities. A dry-run simulation is carried out by a single process, which performs all simulation steps except communication as if it was part of a parallel environment with many processes. We show that measurements of memory usage and runtime of neuronal network simulations closely match the corresponding dry-run data. Furthermore, we demonstrate the successful application of the dry-run mode in the areas of profiling and performance modeling.
An evaluation and analysis of three dynamic watershed acidification codes (MAGIC, ETD, and ILWAS)
Energy Technology Data Exchange (ETDEWEB)
Jenne, E.A.; Eary, L.E.; Vail, L.W.; Girvin, D.C.; Liebetrau, A.M.; Hibler, L.F.; Miley, T.B.; Monsour, M.J.
1989-01-01
The US Environmental Protection Agency is currently using the dynamic watershed acidification codes MAGIC, ILWAS, and ETD to assess the potential future impact of the acidic deposition on surface water quality by simulating watershed acid neutralization processes. The reliability of forecasts made with these codes is of considerable concern. The present study evaluates the process formulations (i.e., conceptual and numerical representation of atmospheric, hydrologic geochemical and biogeochemical processes), compares their approaches to calculating acid neutralizing capacity (ANC), and estimates the relative effects (sensitivity) of perturbations in the input data on selected output variables for each code. Input data were drawn from three Adirondack (upstate New York) watersheds: Panther Lake, Clear Pond, and Woods Lake. Code calibration was performed by the developers of the codes. Conclusions focus on summarizing the adequacy of process formulations, differences in ANC simulation among codes and recommendations for further research to increase forecast reliability. 87 refs., 11 figs., 77 tabs.
DYNAPCON: a computer code for dynamic analysis of prestressed concrete structures
International Nuclear Information System (INIS)
Marchertas, A.H.
1982-09-01
A finite element computer code for the transient analysis of prestressed concrete reactor vessels (PCRVs) for LMFBR containment is described. The method assumes rotational symmetry of the structure. Time integration is by an explicit method. The quasistatic prestressing operation of the PCRV model is performed by a dynamic relaxation technique. The material model accounts for the crushing and tensile cracking in arbitrary direction in concrete and the elastic-plastic behavior of reinforcing steel. The variation of the concrete tensile cracking and compressive crushing limits with strain rate is taken into account. Relative slip is permitted between the concrete and tendons. Several example solutions are presented and compared with experimental results. These sample problems range from simply supported beams to small scale models of PCRV's. It is shown that the analytical methods correlate quite well with experimental results, although in the vicinity of the failure load the response of the models tend to be quite sensitive to input parameters
Development of a computer code for dynamic analysis of the primary circuit of advanced reactors
Energy Technology Data Exchange (ETDEWEB)
Rocha, Jussie Soares da; Lira, Carlos A.B.O.; Magalhaes, Mardson A. de Sa, E-mail: cabol@ufpe.b [Universidade Federal de Pernambuco (DEN/UFPE), Recife, PE (Brazil). Dept. de Energia Nuclear
2011-07-01
Currently, advanced reactors are being developed, seeking for enhanced safety, better performance and low environmental impacts. Reactor designs must follow several steps and numerous tests before a conceptual project could be certified. In this sense, computational tools become indispensable in the preparation of such projects. Thus, this study aimed at the development of a computational tool for thermal-hydraulic analysis by coupling two computer codes to evaluate the influence of transients caused by pressure variations and flow surges in the region of the primary circuit of IRIS reactor between the core and the pressurizer. For the simulation, it was used a situation of 'insurge', characterized by the entry of water in the pressurizer, due to the expansion of the refrigerant in the primary circuit. This expansion was represented by a pressure disturbance in step form, through the block 'step' of SIMULINK, thus enabling the transient startup. The results showed that the dynamic tool, obtained through the coupling of the codes, generated very satisfactory responses within model limitations, preserving the most important phenomena in the process. (author)
Development of a computer code for dynamic analysis of the primary circuit of advanced reactors
International Nuclear Information System (INIS)
Rocha, Jussie Soares da; Lira, Carlos A.B.O.; Magalhaes, Mardson A. de Sa
2011-01-01
Currently, advanced reactors are being developed, seeking for enhanced safety, better performance and low environmental impacts. Reactor designs must follow several steps and numerous tests before a conceptual project could be certified. In this sense, computational tools become indispensable in the preparation of such projects. Thus, this study aimed at the development of a computational tool for thermal-hydraulic analysis by coupling two computer codes to evaluate the influence of transients caused by pressure variations and flow surges in the region of the primary circuit of IRIS reactor between the core and the pressurizer. For the simulation, it was used a situation of 'insurge', characterized by the entry of water in the pressurizer, due to the expansion of the refrigerant in the primary circuit. This expansion was represented by a pressure disturbance in step form, through the block 'step' of SIMULINK, thus enabling the transient startup. The results showed that the dynamic tool, obtained through the coupling of the codes, generated very satisfactory responses within model limitations, preserving the most important phenomena in the process. (author)
Energy Technology Data Exchange (ETDEWEB)
Jankovsky, Zachary Kyle [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2017-05-01
It is difficult to assess the consequences of a transient in a sodium-cooled fast reactor (SFR) using traditional probabilistic risk assessment (PRA) methods, as numerous safety-related sys- tems have passive characteristics. Often there is significant dependence on the value of con- tinuous stochastic parameters rather than binary success/failure determinations. One form of dynamic PRA uses a system simulator to represent the progression of a transient, tracking events through time in a discrete dynamic event tree (DDET). In order to function in a DDET environment, a simulator must have characteristics that make it amenable to changing physical parameters midway through the analysis. The SAS4A SFR system analysis code did not have these characteristics as received. This report describes the code modifications made to allow dynamic operation as well as the linking to a Sandia DDET driver code. A test case is briefly described to demonstrate the utility of the changes.
Development of HTGR plant dynamics simulation code
International Nuclear Information System (INIS)
Ohashi, Kazutaka; Tazawa, Yujiro; Mitake, Susumu; Suzuki, Katsuo.
1987-01-01
Plant dynamics simulation analysis plays an important role in the design work of nuclear power plant especially in the plant safety analysis, control system analysis, and transient condition analysis. The authors have developed the plant dynamics simulation code named VESPER, which is applicable to the design work of High Temperature Engineering Test Reactor, and have been improving the code corresponding to the design changes made in the subsequent design works. This paper describes the outline of VESPER code and shows its sample calculation results selected from the recent design work. (author)
PROMETHEUS - a code system for dynamic 3-D analysis of nuclear reactors
Energy Technology Data Exchange (ETDEWEB)
Khotylev, V.A.; Hoogenboom, J.E.; Leege, P.F.A. de [Technische Univ. Delft (Netherlands). Interfacultair Reactor Inst.
1996-09-01
The paper presents a multidimensional, general-purpose neutronics code system. It solves a number of steady-state and/or transient problems with coupled thermal hydraulics in one-, two-, or three-dimensional geometry. Due to a number of specialized features such as cavity treatment, automated convergence control, burnup treatment using the full isotopic transition matrix, the code system can be applied for the analysis of fast and slow transients in small, large, and innovative reactor cores. (author)
Structural dynamics in LMFBR containment analysis: a brief survey of computational methods and codes
International Nuclear Information System (INIS)
Chang, Y.W.; Gvildys, J.
1977-01-01
In recent years, the use of computer codes to study the response of primary containment of large, liquid-metal fast breeder reactors (LMFBR) under postulated accident conditions has been adopted by most fast reactor projects. Since the first introduction of REXCO-H containment code in 1969, a number of containment codes have evolved and been reported in the literature. The paper briefly summarizes the various numerical methods commonly used in containment analysis in computer programs. They are compared on the basis of truncation errors resulting in the numerical approximation, the method of integration, the resolution of the computed results, and the ease of programming in computer codes. The aim of the paper is to provide enough information to an analyst so that he can suitably define his choice of method, and hence his choice of programs
International Nuclear Information System (INIS)
Buonomano, Annamaria; Palombo, Adolfo
2014-01-01
Highlights: • A new dynamic simulation code for building energy performance analysis is presented. • The thermal behavior of each building element is modeled by a thermal RC network. • The physical models implemented in the code are illustrated. • The code was validated by the BESTEST standard procedure. • We investigate residential buildings, offices and stores in different climates. - Abstract: A novel dynamic simulation model for the building envelope energy performance analysis is presented in this paper. This tool helps the investigation of many new building technologies to increase the system energy efficiency and it can be carried out for scientific research purposes. In addition to the yearly heating and cooling load and energy demand, the obtained output is the dynamic temperature profile of indoor air and surfaces and the dynamic profile of the thermal fluxes through the building elements. The presented simulation model is also validated through the BESTEST standard procedure. Several new case studies are developed for assessing, through the presented code, the energy performance of three different building envelopes with several different weather conditions. In particular, dwelling and commercial buildings are analysed. Light and heavyweight envelopes as well as different glazed surfaces areas have been used for every case study. With the achieved results interesting design and operating guidelines can be obtained. Such data have been also compared vs. those calculated by TRNSYS and EnergyPlus. The detected deviation of the obtained results vs. those of such standard tools are almost always lower than 10%
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
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.
Dynamic benchmarking of simulation codes
International Nuclear Information System (INIS)
Henry, R.E.; Paik, C.Y.; Hauser, G.M.
1996-01-01
Computer simulation of nuclear power plant response can be a full-scope control room simulator, an engineering simulator to represent the general behavior of the plant under normal and abnormal conditions, or the modeling of the plant response to conditions that would eventually lead to core damage. In any of these, the underlying foundation for their use in analysing situations, training of vendor/utility personnel, etc. is how well they represent what has been known from industrial experience, large integral experiments and separate effects tests. Typically, simulation codes are benchmarked with some of these; the level of agreement necessary being dependent upon the ultimate use of the simulation tool. However, these analytical models are computer codes, and as a result, the capabilities are continually enhanced, errors are corrected, new situations are imposed on the code that are outside of the original design basis, etc. Consequently, there is a continual need to assure that the benchmarks with important transients are preserved as the computer code evolves. Retention of this benchmarking capability is essential to develop trust in the computer code. Given the evolving world of computer codes, how is this retention of benchmarking capabilities accomplished? For the MAAP4 codes this capability is accomplished through a 'dynamic benchmarking' feature embedded in the source code. In particular, a set of dynamic benchmarks are included in the source code and these are exercised every time the archive codes are upgraded and distributed to the MAAP users. Three different types of dynamic benchmarks are used: plant transients; large integral experiments; and separate effects tests. Each of these is performed in a different manner. The first is accomplished by developing a parameter file for the plant modeled and an input deck to describe the sequence; i.e. the entire MAAP4 code is exercised. The pertinent plant data is included in the source code and the computer
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
PRONTO3D users` instructions: A transient dynamic code for nonlinear structural analysis
Energy Technology Data Exchange (ETDEWEB)
Attaway, S.W.; Mello, F.J.; Heinstein, M.W.; Swegle, J.W.; Ratner, J.A. [Sandia National Labs., Albuquerque, NM (United States); Zadoks, R.I. [Univ. of Texas, El Paso, TX (United States)
1998-06-01
This report provides an updated set of users` instructions for PRONTO3D. PRONTO3D is a three-dimensional, transient, solid dynamics code for analyzing large deformations of highly nonlinear materials subjected to extremely high strain rates. This Lagrangian finite element program uses an explicit time integration operator to integrate the equations of motion. Eight-node, uniform strain, hexahedral elements and four-node, quadrilateral, uniform strain shells are used in the finite element formulation. An adaptive time step control algorithm is used to improve stability and performance in plasticity problems. Hourglass distortions can be eliminated without disturbing the finite element solution using either the Flanagan-Belytschko hourglass control scheme or an assumed strain hourglass control scheme. All constitutive models in PRONTO3D are cast in an unrotated configuration defined using the rotation determined from the polar decomposition of the deformation gradient. A robust contact algorithm allows for the impact and interaction of deforming contact surfaces of quite general geometry. The Smooth Particle Hydrodynamics method has been embedded into PRONTO3D using the contact algorithm to couple it with the finite element method.
Energy Technology Data Exchange (ETDEWEB)
Lee, Chang Hoon; Baek, Sang Yeup; Shin, In Sup; Moon, Shin Myung; Moon, Jae Phil; Koo, Hoon Young; Kim, Ju Shin [Seoul National University, Seoul (Korea, Republic of); Hong, Jung Sik [Seoul National Polytechnology University, Seoul (Korea, Republic of); Lim, Tae Jin [Soongsil University, Seoul (Korea, Republic of)
1996-08-01
The objective of this project is to develop a methodology of the dynamic reliability analysis for NPP. The first year`s research was focused on developing a procedure for analyzing failure data of running components and a simulator for estimating the reliability of series-parallel structures. The second year`s research was concentrated on estimating the lifetime distribution and PM effect of a component from its failure data in various cases, and the lifetime distribution of a system with a particular structure. Computer codes for performing these jobs were also developed. The objectives of the third year`s research is to develop models for analyzing special failure types (CCFs, Standby redundant structure) that were nor considered in the first two years, and to complete a methodology of the dynamic reliability analysis for nuclear power plants. The analysis of failure data of components and related researches for supporting the simulator must be preceded for providing proper input to the simulator. Thus this research is divided into three major parts. 1. Analysis of the time dependent life distribution and the PM effect. 2. Development of a simulator for system reliability analysis. 3. Related researches for supporting the simulator : accelerated simulation analytic approach using PH-type distribution, analysis for dynamic repair effects. 154 refs., 5 tabs., 87 figs. (author)
Analysis code of three dimensional core dynamics for high temperature gas-cooled reactors, COMIC-2
International Nuclear Information System (INIS)
Takano, Makoto
1987-04-01
The code has been improved and modified in order to speedup calculation and to make more effective since it was developed in 1985. This report is written as a user's manual of the latest version of the code (COMIC-2). Speedup of the code is performed by the improvement of program flow and vector programming. The total speedup factor depends on problem, however, is about 10 in the case of a sample ploblem. (author)
Multi-code analysis of scrape-off layer filament dynamics in MAST
DEFF Research Database (Denmark)
Militello, F.; Walkden, N. R.; Farley, T.
2016-01-01
velocities of the order of 1 km s(-1), a perpendicular diameter of around 2-3 cm and a density amplitude 2-3.5 times the background plasma. 3D and 2D numerical codes (the STORM module of BOUT++, GBS, HESEL and TOKAM3X) are used to reproduce the motion of the observed filaments with the purpose of validating...
TRACK The New Beam Dynamics Code
Mustapha, Brahim; Ostroumov, Peter; Schnirman-Lessner, Eliane
2005-01-01
The new ray-tracing code TRACK was developed* to fulfill the special requirements of the RIA accelerator systems. The RIA lattice includes an ECR ion source, a LEBT containing a MHB and a RFQ followed by three SC linac sections separated by two stripping stations with appropriate magnetic transport systems. No available beam dynamics code meet all the necessary requirements for an end-to-end simulation of the RIA driver linac. The latest version of TRACK was used for end-to-end simulations of the RIA driver including errors and beam loss analysis.** In addition to the standard capabilities, the code includes the following new features: i) multiple charge states ii) realistic stripper model; ii) static and dynamic errors iii) automatic steering to correct for misalignments iv) detailed beam-loss analysis; v) parallel computing to perform large scale simulations. Although primarily developed for simulations of the RIA machine, TRACK is a general beam dynamics code. Currently it is being used for the design and ...
Parallelization of quantum molecular dynamics simulation code
International Nuclear Information System (INIS)
Kato, Kaori; Kunugi, Tomoaki; Shibahara, Masahiko; Kotake, Susumu
1998-02-01
A quantum molecular dynamics simulation code has been developed for the analysis of the thermalization of photon energies in the molecule or materials in Kansai Research Establishment. The simulation code is parallelized for both Scalar massively parallel computer (Intel Paragon XP/S75) and Vector parallel computer (Fujitsu VPP300/12). Scalable speed-up has been obtained with a distribution to processor units by division of particle group in both parallel computers. As a result of distribution to processor units not only by particle group but also by the particles calculation that is constructed with fine calculations, highly parallelization performance is achieved in Intel Paragon XP/S75. (author)
Beam-dynamics codes used at DARHT
Energy Technology Data Exchange (ETDEWEB)
Ekdahl, Jr., Carl August [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
2017-02-01
Several beam simulation codes are used to help gain a better understanding of beam dynamics in the DARHT LIAs. The most notable of these fall into the following categories: for beam production – Tricomp Trak orbit tracking code, LSP Particle in cell (PIC) code, for beam transport and acceleration – XTR static envelope and centroid code, LAMDA time-resolved envelope and centroid code, LSP-Slice PIC code, for coasting-beam transport to target – LAMDA time-resolved envelope code, LSP-Slice PIC code. These codes are also being used to inform the design of Scorpius.
SCISEAL: A CFD Code for Analysis of Fluid Dynamic Forces in Seals
Althavale, Mahesh M.; Ho, Yin-Hsing; Przekwas, Andre J.
1996-01-01
A 3D CFD code, SCISEAL, has been developed and validated. Its capabilities include cylindrical seals, and it is employed on labyrinth seals, rim seals, and disc cavities. State-of-the-art numerical methods include colocated grids, high-order differencing, and turbulence models which account for wall roughness. SCISEAL computes efficient solutions for complicated flow geometries and seal-specific capabilities (rotor loads, torques, etc.).
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
International Nuclear Information System (INIS)
2003-11-01
Safety analysis is an important tool for justifying the safety of nuclear power plants. Typically, this type of analysis is performed by means of system computer codes with one dimensional approximation for modelling real plant systems. However, in the nuclear area there are issues for which traditional treatment using one dimensional system codes is considered inadequate for modelling local flow and heat transfer phenomena. There is therefore increasing interest in the application of three dimensional computational fluid dynamics (CFD) codes as a supplement to or in combination with system codes. There are a number of both commercial (general purpose) CFD codes as well as special codes for nuclear safety applications available. With further progress in safety analysis techniques, the increasing use of CFD codes for nuclear applications is expected. At present, the main objective with respect to CFD codes is generally to improve confidence in the available analysis tools and to achieve a more reliable approach to safety relevant issues. An exchange of views and experience can facilitate and speed up progress in the implementation of this objective. Both the International Atomic Energy Agency (IAEA) and the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (OECD/NEA) believed that it would be advantageous to provide a forum for such an exchange. Therefore, within the framework of the Working Group on the Analysis and Management of Accidents of the NEA's Committee on the Safety of Nuclear Installations, the IAEA and the NEA agreed to jointly organize the Technical Meeting on the Use of Computational Fluid Dynamics Codes for Safety Analysis of Reactor Systems, including Containment. The meeting was held in Pisa, Italy, from 11 to 14 November 2002. The publication constitutes the report of the Technical Meeting. It includes short summaries of the presentations that were made and of the discussions as well as conclusions and
High dynamic range coding imaging system
Wu, Renfan; Huang, Yifan; Hou, Guangqi
2014-10-01
We present a high dynamic range (HDR) imaging system design scheme based on coded aperture technique. This scheme can help us obtain HDR images which have extended depth of field. We adopt Sparse coding algorithm to design coded patterns. Then we utilize the sensor unit to acquire coded images under different exposure settings. With the guide of the multiple exposure parameters, a series of low dynamic range (LDR) coded images are reconstructed. We use some existing algorithms to fuse and display a HDR image by those LDR images. We build an optical simulation model and get some simulation images to verify the novel system.
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
International Nuclear Information System (INIS)
Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.
1988-01-01
This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)
International Nuclear Information System (INIS)
Rahatgaonkar, P. S.; Datta, D.; Malhotra, P. K.; Ghadge, S. G.
2012-01-01
Prediction of groundwater movement and contaminant transport in soil is an important problem in many branches of science and engineering. This includes groundwater hydrology, environmental engineering, soil science, agricultural engineering and also nuclear engineering. Specifically, in nuclear engineering it is applicable in the design of spent fuel storage pools and waste management sites in the nuclear power plants. Ground water modeling involves the simulation of flow and contaminant transport by groundwater flow. In the context of contaminated soil and groundwater system, numerical simulations are typically used to demonstrate compliance with regulatory standard. A one-dimensional Computational Fluid Dynamics code GFLOW had been developed based on the Finite Difference Method for simulating groundwater flow and contaminant transport through saturated and unsaturated soil. The code is validated with the analytical model and the benchmarking cases available in the literature. (authors)
International Nuclear Information System (INIS)
Sumner, H.M.
1965-11-01
FIFI 3 is a FORTRAN Code embodying a technique for the analysis of process plant dynamics. As such, it is essentially a tool for the integration of sets of first order ordinary differential equations, either linear or non-linear; special provision is made for the inclusion of time-delayed variables in the mathematical model of the plant. The method of integration is new and is centred on a stable multistep predictor-corrector algorithm devised by the late Mr. F.G. Chapman, of the UKAEA, Winfrith. The theory on which the Code is based and detailed rules for using it are described in Parts I and II respectively. (author)
Computer codes for safety analysis
International Nuclear Information System (INIS)
Holland, D.F.
1986-11-01
Computer codes for fusion safety analysis have been under development in the United States for about a decade. This paper will discuss five codes that are currently under development by the Fusion Safety Program. The purpose and capability of each code will be presented, a sample given, followed by a discussion of the present status and future development plans
Dynamic code block size for JPEG 2000
Tsai, Ping-Sing; LeCornec, Yann
2008-02-01
Since the standardization of the JPEG 2000, it has found its way into many different applications such as DICOM (digital imaging and communication in medicine), satellite photography, military surveillance, digital cinema initiative, professional video cameras, and so on. The unified framework of the JPEG 2000 architecture makes practical high quality real-time compression possible even in video mode, i.e. motion JPEG 2000. In this paper, we present a study of the compression impact using dynamic code block size instead of fixed code block size as specified in the JPEG 2000 standard. The simulation results show that there is no significant impact on compression if dynamic code block sizes are used. In this study, we also unveil the advantages of using dynamic code block sizes.
Validation of the reactor dynamics code HEXTRAN
International Nuclear Information System (INIS)
Kyrki-Rajamaeki, R.
1994-05-01
HEXTRAN is a new three-dimensional, hexagonal reactor dynamics code developed in the Technical Research Centre of Finland (VTT) for VVER type reactors. This report describes the validation work of HEXTRAN. The work has been made with the financing of the Finnish Centre for Radiation and Nuclear Safety (STUK). HEXTRAN is particularly intended for calculation of such accidents, in which radially asymmetric phenomena are included and both good neutron dynamics and two-phase thermal hydraulics are important. HEXTRAN is based on already validated codes. The models of these codes have been shown to function correctly also within the HEXTRAN code. The main new model of HEXTRAN, the spatial neutron kinetics model has been successfully validated against LR-0 test reactor and Loviisa plant measurements. Connected with SMABRE, HEXTRAN can be reliably used for calculation of transients including effects of the whole cooling system of VVERs. Further validation plans are also introduced in the report. (orig.). (23 refs., 16 figs., 2 tabs.)
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
International Nuclear Information System (INIS)
Wang, L.; Bos, I.C. Van den; Hussain, S.M.; Pattynama, P.M.; Vogel, M.W.; Kr estin, G.P.
2008-01-01
The purpose of this article is to explain and illustrate the current status and potential applications of automated and color-coded post-processing techniques for the analysis of dynamic multiphasic gadolinium-enhanced magnetic resonance imaging (MRI) of the liver. Post-processing of these images on dedicated workstations allows the generation of time-intensity curves (TIC) as well as color-coded images, which provides useful information on (neo)-angiogenesis within a liver lesion, if necessary combined with information on enhancement patterns of the surrounding liver parenchyma. Analysis of TIC and color-coded images, which are based on pharmacokinetic modeling, provides an easy-to-interpret schematic presentation of tumor behavior, providing additional characteristics for adequate differential diagnosis. Inclusion of TIC and color-coded images as part of the routine abdominal MRI workup protocol may help to further improve the specificity of MRI findings, but needs to be validated in clinical decision-making situations. In addition, these tools may facilitate the diagnostic workup of disease for detection, characterization, staging, and monitoring of antitumor therapy, and hold incremental value to the widely used tumor response criteria
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
IFR code for secondary particle dynamics
International Nuclear Information System (INIS)
Teague, M.R.; Yu, S.S.
1985-01-01
A numerical simulation has been constructed to obtain a detailed, quantitative estimate of the electromagnetic fields and currents existing in the Advanced Test Accelerator under conditions of laser guiding. The code treats the secondary electrons by particle simulation and the beam dynamics by a time-dependent envelope model. The simulation gives a fully relativistic description of secondary electrons moving in self-consistent electromagnetic fields. The calculations are made using coordinates t, x, y, z for the electrons and t, ct-z, r for the axisymmetric electromagnetic fields and currents. Code results, showing in particular current enhancement effects, will be given
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
ITER Dynamic Tritium Inventory Modeling Code
International Nuclear Information System (INIS)
Cristescu, Ioana-R.; Doerr, L.; Busigin, A.; Murdoch, D.
2005-01-01
A tool for tritium inventory evaluation within each sub-system of the Fuel Cycle of ITER is vital, with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems, however tritium accounting may be achieved by modeling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the systems. To get reliable results, an accurate dynamic modeling of the tritium content in each sub-system is necessary. In order to optimize the configuration and operation of the ITER fuel cycle, a dynamic fuel cycle model was developed progressively in the decade up to 2000-2001. As the design for some sub-systems from the fuel cycle (i.e. Vacuum pumping, Neutral Beam Injectors (NBI)) have substantially progressed meanwhile, a new code developed under a different platform to incorporate these modifications has been developed. The new code is taking over the models and algorithms for some subsystems, such as Isotope Separation System (ISS); where simplified models have been previously considered, more detailed have been introduced, as for the Water Detritiation System (WDS). To reflect all these changes, the new code developed inside EU participating team was nominated TRIMO (Tritium Inventory Modeling), to emphasize the use of the code on assessing the tritium inventory within ITER
International Nuclear Information System (INIS)
Bestion, D.
1989-11-01
The CATHARE code is used to calculate the experiment of the University of Hannover concerning the flooding limit at the fuel element top nozzle area. Some qualitative and quantitativ limit at the fuel element top nozzle area. on both the actual fluid dynamics which is observed in the experiments and on the corresponding code behaviour. Shortcomings of the present models are clearly identified. New developments are proposed which should extend the code capabilities
CINETHICA - Core accident analysis code
International Nuclear Information System (INIS)
Nakata, H.
1989-10-01
A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt
A molecular dynamics simulation code ISIS
International Nuclear Information System (INIS)
Kambayashi, Shaw
1992-06-01
Computer simulation based on the molecular dynamics (MD) method has become an important tool complementary to experiments and theoretical calculations in a wide range of scientific fields such as physics, chemistry, biology, and so on. In the MD method, the Newtonian equations-of-motion of classical particles are integrated numerically to reproduce a phase-space trajectory of the system. In the 1980's, several new techniques have been developed for simulation at constant-temperature and/or constant-pressure in convenient to compare result of computer simulation with experimental results. We first summarize the MD method for both microcanonical and canonical simulations. Then, we present and overview of a newly developed ISIS (Isokinetic Simulation of Soft-spheres) code and its performance on various computers including vector processors. The ISIS code has a capability to make a MD simulation under constant-temperature condition by using the isokinetic constraint method. The equations-of-motion is integrated by a very accurate fifth-order finite differential algorithm. The bookkeeping method is also utilized to reduce the computational time. Furthermore, the ISIS code is well adopted for vector processing: Speedup ratio ranged from 16 to 24 times is obtained on a VP2600/10 vector processor. (author)
Directory of Open Access Journals (Sweden)
Taewan Kim
2012-01-01
Full Text Available In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The results reveal that the prediction by the subchannel code FLICA4 agrees with the experimental data reasonably well in both steady-state and transient conditions. The analyses of single-subchannel experiments by means of the computational fluid dynamic code STAR-CD with the CD-adapco boiling model indicate that the prediction of the void fraction has no significant discrepancy from the experiments. The analyses with TRACE point out the necessity to perform additional assessment of the subcooled boiling model and bulk condensation model of TRACE.
Fire-accident analysis code (FIRAC) verification
International Nuclear Information System (INIS)
Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.
1986-01-01
The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data
International Nuclear Information System (INIS)
Salazar C, J.H.; Ramos P, J.C.; Salazar S, E.; Chavez M, C.
2003-01-01
The present work illustrates the application of the concept of a simulator for analysis, design, instruction and training in a classroom environment associated to a nuclear power station. Emphasis is made on the methodology used to incorporate the best estimate codes RELAP/SCDAP to a prototype under development at the Nuclear Reactor Engineering Analysis Laboratory (NREAL). This methodology is based on a modular structure where multiple processes can be executed in an independent way and where the generated information is stored in shared memory segments and distributed by means of communication routines developed in the C programming language. The utility of the system is demonstrated using highly interactive graphics (mimic diagrams, pictorials and tendency graphs) for the simultaneous dynamic visualization of the most significant variables of a typical transient event (feed water controller failure in a BWR). A fundamental part of the system is its advanced graphic interface. This interface, of the type of direct manipulation, reproduces instruments and controls whose functionality is similar to those found in the current replica simulator for the Laguna Verde Nuclear Power Station. Finally the evaluation process is described. The general behavior of the main variables for the selected transitory event is interpreted, corroborating that they follow the same tendency that those reported for a BWR. The obtained results allow to conclude that the developed system works satisfactorily and that the use of al 1 x 1 real time visualization tools offers important advantages regarding other traditional methods of analysis. (Author)
International Nuclear Information System (INIS)
Abanades, A.; Pena, A.
2009-01-01
A new innovative nuclear installation is under research in the nuclear community for its potential application to nuclear waste management and, above all, for its capability to enhance the sustainability of nuclear energy in the future as component of a new nuclear fuel cycle in which its efficiency in terms of primary Uranium ore profit and radioactive waste generation will be improved. Such new nuclear installations are called accelerator driven system (ADS) and are the result of a profitable symbiosis between accelerator technology, high-energy physics and reactor technology. Many ADS concepts are based on the utilization of heavy liquid metal (HLM) coolants due to its neutronic and thermo-physical properties. Moreover, such coolants permit the operation in free circulation mode, one of the main aims of passive systems. In this paper, such operation regime is analysed in a proposed ADS design applying computational fluid dynamics (CFD)
Bayesian dynamic mediation analysis.
Huang, Jing; Yuan, Ying
2017-12-01
Most existing methods for mediation analysis assume that mediation is a stationary, time-invariant process, which overlooks the inherently dynamic nature of many human psychological processes and behavioral activities. In this article, we consider mediation as a dynamic process that continuously changes over time. We propose Bayesian multilevel time-varying coefficient models to describe and estimate such dynamic mediation effects. By taking the nonparametric penalized spline approach, the proposed method is flexible and able to accommodate any shape of the relationship between time and mediation effects. Simulation studies show that the proposed method works well and faithfully reflects the true nature of the mediation process. By modeling mediation effect nonparametrically as a continuous function of time, our method provides a valuable tool to help researchers obtain a more complete understanding of the dynamic nature of the mediation process underlying psychological and behavioral phenomena. We also briefly discuss an alternative approach of using dynamic autoregressive mediation model to estimate the dynamic mediation effect. The computer code is provided to implement the proposed Bayesian dynamic mediation analysis. (PsycINFO Database Record (c) 2017 APA, all rights reserved).
Light-water reactor safety analysis codes
International Nuclear Information System (INIS)
Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.
1980-01-01
A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented
Vectorization, parallelization and implementation of Quantum molecular dynamics codes (QQQF, MONTEV)
Energy Technology Data Exchange (ETDEWEB)
Kato, Kaori [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Kunugi, Tomoaki; Kotake, Susumu; Shibahara, Masahiko
1998-03-01
This report describes parallelization, vectorization and implementation for two simulation codes, Quantum molecular dynamics simulation code QQQF and Photon montecalro molecular dynamics simulation code MONTEV, that have been developed for the analysis of the thermalization of photon energies in the molecule or materials. QQQF has been vectorized and parallelized on Fujitsu VPP and has been implemented from VPP to Intel Paragon XP/S and parallelized. MONTEV has been implemented from VPP to Paragon and parallelized. (author)
Status of SPACE Safety Analysis Code Development
International Nuclear Information System (INIS)
Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun
2009-01-01
In 2006, the Korean the Korean nuclear industry started developing a thermal-hydraulic analysis code for safety analysis of PWR(Pressurized Water Reactor). The new code is named as SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). The SPACE code can solve two-fluid, three-field governing equations in one dimensional or three dimensional geometry. The SPACE code has many component models required for modeling a PWR, such as reactor coolant pump, safety injection tank, etc. The programming language used in the new code is C++, for new generation of engineers who are more comfortable with C/C++ than old FORTRAN language. This paper describes general characteristics of SPACE code and current status of SPACE code development
Stability analysis by ERATO code
International Nuclear Information System (INIS)
Tsunematsu, Toshihide; Takeda, Tatsuoki; Matsuura, Toshihiko; Azumi, Masafumi; Kurita, Gen-ichi
1979-12-01
Problems in MHD stability calculations by ERATO code are described; which concern convergence property of results, equilibrium codes, and machine optimization of ERATO code. It is concluded that irregularity on a convergence curve is not due to a fault of the ERATO code itself but due to inappropriate choice of the equilibrium calculation meshes. Also described are a code to calculate an equilibrium as a quasi-inverse problem and a code to calculate an equilibrium as a result of a transport process. Optimization of the code with respect to I/O operations reduced both CPU time and I/O time considerably. With the FACOM230-75 APU/CPU multiprocessor system, the performance is about 6 times as high as with the FACOM230-75 CPU, showing the effectiveness of a vector processing computer for the kind of MHD computations. This report is a summary of the material presented at the ERATO workshop 1979(ORNL), supplemented with some details. (author)
Uncertainty analysis of the FRAP code
International Nuclear Information System (INIS)
Peck, S.O.
1978-01-01
A user oriented, automated uncertainty analysis capability has been built into the FRAP code (Fuel Rod Analysis Program) and applied to a PWR fuel rod undergoing a LOCA. The method of uncertainty analysis is the Response Surface Method (RSM). (author)
Binary codes with impulse autocorrelation functions for dynamic experiments
International Nuclear Information System (INIS)
Corran, E.R.; Cummins, J.D.
1962-09-01
A series of binary codes exist which have autocorrelation functions approximating to an impulse function. Signals whose behaviour in time can be expressed by such codes have spectra which are 'whiter' over a limited bandwidth and for a finite time than signals from a white noise generator. These codes are used to determine system dynamic responses using the correlation technique. Programmes have been written to compute codes of arbitrary length and to compute 'cyclic' autocorrelation and cross-correlation functions. Complete listings of these programmes are given, and a code of 1019 bits is presented. (author)
Dynamic Contingency Analysis Tool
Energy Technology Data Exchange (ETDEWEB)
2016-01-14
The Dynamic Contingency Analysis Tool (DCAT) is an open-platform and publicly available methodology to help develop applications that aim to improve the capabilities of power system planning engineers to assess the impact and likelihood of extreme contingencies and potential cascading events across their systems and interconnections. Outputs from the DCAT will help find mitigation solutions to reduce the risk of cascading outages in technically sound and effective ways. The current prototype DCAT implementation has been developed as a Python code that accesses the simulation functions of the Siemens PSS/E planning tool (PSS/E). It has the following features: It uses a hybrid dynamic and steady-state approach to simulating the cascading outage sequences that includes fast dynamic and slower steady-state events. It integrates dynamic models with protection scheme models for generation, transmission, and load. It models special protection systems (SPSs)/remedial action schemes (RASs) and automatic and manual corrective actions. Overall, the DCAT attempts to bridge multiple gaps in cascading-outage analysis in a single, unique prototype tool capable of automatically simulating and analyzing cascading sequences in real systems using multiprocessor computers.While the DCAT has been implemented using PSS/E in Phase I of the study, other commercial software packages with similar capabilities can be used within the DCAT framework.
SASSYS LMFBR systems analysis code
International Nuclear Information System (INIS)
Dunn, F.E.; Prohammer, F.G.
1982-01-01
The SASSYS code provides detailed steady-state and transient thermal-hydraulic analyses of the reactor core, inlet and outlet coolant plenums, primary and intermediate heat-removal systems, steam generators, and emergency shut-down heat removal systems in liquid-metal-cooled fast-breeder reactors (LMFBRs). The main purpose of the code is to analyze the consequences of failures in the shut-down heat-removal system and to determine whether this system can perform its mission adequately even with some of its components inoperable. The code is not plant-specific. It is intended for use with any LMFBR, using either a loop or a pool design, a once-through steam generator or an evaporator-superheater combination, and either a homogeneous core or a heterogeneous core with internal-blanket assemblies
Centrifugal Compressor Aeroelastic Analysis Code
Keith, Theo G., Jr.; Srivastava, Rakesh
2002-01-01
Centrifugal compressors are very widely used in the turbomachine industry where low mass flow rates are required. Gas turbine engines for tanks, rotorcraft and small jets rely extensively on centrifugal compressors for rugged and compact design. These compressors experience problems related with unsteadiness of flowfields, such as stall flutter, separation at the trailing edge over diffuser guide vanes, tip vortex unsteadiness, etc., leading to rotating stall and surge. Considerable interest exists in small gas turbine engine manufacturers to understand and eventually eliminate the problems related to centrifugal compressors. The geometric complexity of centrifugal compressor blades and the twisting of the blade passages makes the linear methods inapplicable. Advanced computational fluid dynamics (CFD) methods are needed for accurate unsteady aerodynamic and aeroelastic analysis of centrifugal compressors. Most of the current day industrial turbomachines and small aircraft engines are designed with a centrifugal compressor. With such a large customer base and NASA Glenn Research Center being, the lead center for turbomachines, it is important that adequate emphasis be placed on this area as well. Currently, this activity is not supported under any project at NASA Glenn.
Evolving a Dynamic Predictive Coding Mechanism for Novelty Detection
Haggett, Simon J.; Chu, Dominique; Marshall, Ian W.
2007-01-01
Novelty detection is a machine learning technique which identifies new or unknown information in data sets. We present our current work on the construction of a new novelty detector based on a dynamical version of predictive coding. We compare three evolutionary algorithms, a simple genetic algorithm, NEAT and FS-NEAT, for the task of optimising the structure of an illustrative dynamic predictive coding neural network to improve its performance over stimuli from a number of artificially gener...
A Case for Dynamic Reverse-code Generation to Debug Non-deterministic Programs
Directory of Open Access Journals (Sweden)
Jooyong Yi
2013-09-01
Full Text Available Backtracking (i.e., reverse execution helps the user of a debugger to naturally think backwards along the execution path of a program, and thinking backwards makes it easy to locate the origin of a bug. So far backtracking has been implemented mostly by state saving or by checkpointing. These implementations, however, inherently do not scale. Meanwhile, a more recent backtracking method based on reverse-code generation seems promising because executing reverse code can restore the previous states of a program without state saving. In the literature, there can be found two methods that generate reverse code: (a static reverse-code generation that pre-generates reverse code through static analysis before starting a debugging session, and (b dynamic reverse-code generation that generates reverse code by applying dynamic analysis on the fly during a debugging session. In particular, we espoused the latter one in our previous work to accommodate non-determinism of a program caused by e.g., multi-threading. To demonstrate the usefulness of our dynamic reverse-code generation, this article presents a case study of various backtracking methods including ours. We compare the memory usage of various backtracking methods in a simple but nontrivial example, a bounded-buffer program. In the case of non-deterministic programs such as this bounded-buffer program, our dynamic reverse-code generation outperforms the existing backtracking methods in terms of memory efficiency.
Parallel processing of structural integrity analysis codes
International Nuclear Information System (INIS)
Swami Prasad, P.; Dutta, B.K.; Kushwaha, H.S.
1996-01-01
Structural integrity analysis forms an important role in assessing and demonstrating the safety of nuclear reactor components. This analysis is performed using analytical tools such as Finite Element Method (FEM) with the help of digital computers. The complexity of the problems involved in nuclear engineering demands high speed computation facilities to obtain solutions in reasonable amount of time. Parallel processing systems such as ANUPAM provide an efficient platform for realising the high speed computation. The development and implementation of software on parallel processing systems is an interesting and challenging task. The data and algorithm structure of the codes plays an important role in exploiting the parallel processing system capabilities. Structural analysis codes based on FEM can be divided into two categories with respect to their implementation on parallel processing systems. The first category codes such as those used for harmonic analysis, mechanistic fuel performance codes need not require the parallelisation of individual modules of the codes. The second category of codes such as conventional FEM codes require parallelisation of individual modules. In this category, parallelisation of equation solution module poses major difficulties. Different solution schemes such as domain decomposition method (DDM), parallel active column solver and substructuring method are currently used on parallel processing systems. Two codes, FAIR and TABS belonging to each of these categories have been implemented on ANUPAM. The implementation details of these codes and the performance of different equation solvers are highlighted. (author). 5 refs., 12 figs., 1 tab
KEWPIE: A dynamical cascade code for decaying exited compound nuclei
Bouriquet, Bertrand; Abe, Yasuhisa; Boilley, David
2004-05-01
A new dynamical cascade code for decaying hot nuclei is proposed and specially adapted to the synthesis of super-heavy nuclei. For such a case, the interesting channel is of the tiny fraction that will decay through particles emission, thus the code avoids classical Monte-Carlo methods and proposes a new numerical scheme. The time dependence is explicitely taken into account in order to cope with the fact that fission decay rate might not be constant. The code allows to evaluate both statistical and dynamical observables. Results are successfully compared to experimental data.
NORTICA—a new code for cyclotron analysis
Gorelov, D.; Johnson, D.; Marti, F.
2001-12-01
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 [1] 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.
1-D hybrid code for FRM dynamics
International Nuclear Information System (INIS)
Stark, R.A.; Miley, G.H.
1985-01-01
A 1-D radial hybrid code has been written to study the start-up of the FRM via neutral-beam injection. This code, named FROST (Field Reversed One-dimensional STart-up), models the plasma as azimuthal symmetric with no axial dependence. A multi-group method in energy and canonical angular momentum describes the large-orbit ions from the beam. This method is designed to be more efficient than those employing particle tracking, since the characteristic timescale of the simulation is the ion slowing down time, rather than the much shorter cyclotron period. A time-differentiated Grad-Shafranov equation couples the ion current to massless fluid equations describing electrons and low energy ions. Flux coordinates are used in this fluid model, in preference to an Eulerian framework, so that coupling of plasma at the two different radii of a closed flux surface may be treated with ease. Since a fluid treatment for electrons is invalid near a field null, a separate model for the electron current has been included for this region, a unique feature. Results of simulation of injection into a 2XIIB-like plasma are discussed. Electron currents are found to retard, but not prevent reversal of the magnetic field at the plasma center
ENSDF analysis codes. IBM version, August 1982
International Nuclear Information System (INIS)
Lorenz, A.
1982-01-01
The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)
ENSDF analysis codes: IBM version. August 1982
International Nuclear Information System (INIS)
Lorenz, A.
1982-09-01
The nuclear structure analysis programme tape consists of physics computer processing codes used in the evaluation of mass-chain structure data. This tape was generated by the National Nuclear Data Centre, Brookhaven National Laboratory in the USA. (author)
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
A modified space charge routine for LINAC beam dynamics codes
International Nuclear Information System (INIS)
Valero, S.; Lapostolle, P.; Lombardi, A.M.; Tanke, E.; Warner, D.
1994-01-01
In 1991 a space charge calculation for bunched beams with three-dimensional ellipsoidal symmetry was proposed for the PARMILA code, replacing the usual SCHEFF routines: it removes the cylindrical symmetry needed for the Fast Fourier Transform method and avoids the point to point interaction computation, where the number of simulation points is limited. This routine has now been improved with the introduction of two (or more) ellipsoids, giving a good representation of actual, pear-shaped bunches (unlike the 3-D ellipsoidal assumption). The ellipsoidal density distributions are computed with a new method, avoiding the difficulty caused by statistical effects, encountered near the centre (the axis in 2-D problems) by the previous method. It also provides a check of the ellipsoidal symmetry for each part of the distribution. Finally, the Fourier analysis reported in 1991 has been replaced by a very convenient Hermite expansion, which gives a simple but accurate representation of practical distributions. Introduced in the new, versatile beam dynamics code, DYNAC, it should provide a good tool for the study of the effects of the various parameters responsible for the halo formation in high intensity linacs. (authors). 11 refs
Dynamic detection technology of malicious code for Android system
Directory of Open Access Journals (Sweden)
Li Boya
2017-02-01
Full Text Available With the increasing popularization of mobile phones,people's dependence on them is rising,the security problems become more and more prominent.According to the calling of the APK file permission and the API function in Android system,this paper proposes a dynamic detecting method based on API interception technology to detect the malicious code.The experimental results show that this method can effectively detect the malicious code in Android system.
KEWPIE: a dynamical cascade code for decaying exited compound nuclei
Bouriquet, Bertrand; Abe, Yasuhisa; Boilley, David
2003-01-01
A new dynamical cascade code for decaying hot nuclei is proposed and specially adapted to the synthesis of super-heavy nuclei. For such a case, the interesting channel is the tiny fraction that will decay through particles emission, thus the code avoids classical Monte-Carlo methods and proposes a new numerical scheme. The time dependence is explicitely taken into account in order to cope with the fact that fission decay rate might not be constant. The code allows to evaluate both statistical...
Porting of serial molecular dynamics code on MIMD platforms
International Nuclear Information System (INIS)
Celino, M.
1995-05-01
A molecular Dynamics (MD) code, utilized for the study of atomistic models of metallic systems has been parallelized for MIMD (Multiple Instructions Multiple Data) parallel platforms by means of the Parallel Virtual Machine (PVM) message passing library. Since the parallelization implies modifications of the sequential algorithms, these are described from the point of view of the Statistical Mechanics theory. Furthermore, techniques and parallelization strategies utilized and the MD parallel code are described in detail. Benchmarks on several MIMD platforms (IBM SP1 and SP2, Cray T3D, Cluster of workstations) allow performances evaluation of the code versus the different characteristics of the parallel platforms
Multitasking the code ARC3D. [for computational fluid dynamics
Barton, John T.; Hsiung, Christopher C.
1986-01-01
The CRAY multitasking system was developed in order to utilize all four processors and sharply reduce the wall clock run time. This paper describes the techniques used to modify the computational fluid dynamics code ARC3D for this run and analyzes the achieved speedup. The ARC3D code solves either the Euler or thin-layer N-S equations using an implicit approximate factorization scheme. Results indicate that multitask processing can be used to achieve wall clock speedup factors of over three times, depending on the nature of the program code being used. Multitasking appears to be particularly advantageous for large-memory problems running on multiple CPU computers.
International Nuclear Information System (INIS)
Chen Zhenpeng; Qi Huiquan
1990-01-01
A comprehensive R-matrix analysis code has been developed. It is based on the multichannel and multilevel R-matrix theory and runs in VAX computer with FORTRAN-77. With this code many kinds of experimental data for one nuclear system can be fitted simultaneously. The comparisions between code RAC and code EDA of LANL are made. The data show both codes produced the same calculation results when one set of R-matrix parameters was used. The differential cross section of 10 B (n, α) 7 Li for E n = 0.4 MeV and the polarization of 16 O (n,n) 16 O for E n = 2.56 MeV are presented
On the Organizational Dynamics of the Genetic Code
Zhang, Zhang
2011-06-07
The organization of the canonical genetic code needs to be thoroughly illuminated. Here we reorder the four nucleotides—adenine, thymine, guanine and cytosine—according to their emergence in evolution, and apply the organizational rules to devising an algebraic representation for the canonical genetic code. Under a framework of the devised code, we quantify codon and amino acid usages from a large collection of 917 prokaryotic genome sequences, and associate the usages with its intrinsic structure and classification schemes as well as amino acid physicochemical properties. Our results show that the algebraic representation of the code is structurally equivalent to a content-centric organization of the code and that codon and amino acid usages under different classification schemes were correlated closely with GC content, implying a set of rules governing composition dynamics across a wide variety of prokaryotic genome sequences. These results also indicate that codons and amino acids are not randomly allocated in the code, where the six-fold degenerate codons and their amino acids have important balancing roles for error minimization. Therefore, the content-centric code is of great usefulness in deciphering its hitherto unknown regularities as well as the dynamics of nucleotide, codon, and amino acid compositions.
On the Organizational Dynamics of the Genetic Code
Zhang, Zhang; Yu, Jun
2011-01-01
The organization of the canonical genetic code needs to be thoroughly illuminated. Here we reorder the four nucleotides—adenine, thymine, guanine and cytosine—according to their emergence in evolution, and apply the organizational rules to devising an algebraic representation for the canonical genetic code. Under a framework of the devised code, we quantify codon and amino acid usages from a large collection of 917 prokaryotic genome sequences, and associate the usages with its intrinsic structure and classification schemes as well as amino acid physicochemical properties. Our results show that the algebraic representation of the code is structurally equivalent to a content-centric organization of the code and that codon and amino acid usages under different classification schemes were correlated closely with GC content, implying a set of rules governing composition dynamics across a wide variety of prokaryotic genome sequences. These results also indicate that codons and amino acids are not randomly allocated in the code, where the six-fold degenerate codons and their amino acids have important balancing roles for error minimization. Therefore, the content-centric code is of great usefulness in deciphering its hitherto unknown regularities as well as the dynamics of nucleotide, codon, and amino acid compositions.
Palo Verde nuclear dynamic analysis (PANDA)
International Nuclear Information System (INIS)
Girjashankar, P.V.; Secker, P.A. Jr.; LeClair, S.J.; Mendoza, J.; Webb, J.R.
1988-01-01
Arizona Nuclear Power Project (ANPP) has initiated the development of a large scale dynamic analysis computer program for the Palo Verde Nuclear Generating Station (PVNGS). This paper presents the decision processes and preliminary development activities that have been pursued related to the code development. The PANDA (Palo Verde Nuclear Dynamic Analysis) code will be used for a variety of applications as described in this paper
The SCEC/USGS dynamic earthquake rupture code verification exercise
Harris, R.A.; Barall, M.; Archuleta, R.; Dunham, E.; Aagaard, Brad T.; Ampuero, J.-P.; Bhat, H.; Cruz-Atienza, Victor M.; Dalguer, L.; Dawson, P.; Day, S.; Duan, B.; Ely, G.; Kaneko, Y.; Kase, Y.; Lapusta, N.; Liu, Yajing; Ma, S.; Oglesby, D.; Olsen, K.; Pitarka, A.; Song, S.; Templeton, E.
2009-01-01
Numerical simulations of earthquake rupture dynamics are now common, yet it has been difficult to test the validity of these simulations because there have been few field observations and no analytic solutions with which to compare the results. This paper describes the Southern California Earthquake Center/U.S. Geological Survey (SCEC/USGS) Dynamic Earthquake Rupture Code Verification Exercise, where codes that simulate spontaneous rupture dynamics in three dimensions are evaluated and the results produced by these codes are compared using Web-based tools. This is the first time that a broad and rigorous examination of numerous spontaneous rupture codes has been performed—a significant advance in this science. The automated process developed to attain this achievement provides for a future where testing of codes is easily accomplished.Scientists who use computer simulations to understand earthquakes utilize a range of techniques. Most of these assume that earthquakes are caused by slip at depth on faults in the Earth, but hereafter the strategies vary. Among the methods used in earthquake mechanics studies are kinematic approaches and dynamic approaches.The kinematic approach uses a computer code that prescribes the spatial and temporal evolution of slip on the causative fault (or faults). These types of simulations are very helpful, especially since they can be used in seismic data inversions to relate the ground motions recorded in the field to slip on the fault(s) at depth. However, these kinematic solutions generally provide no insight into the physics driving the fault slip or information about why the involved fault(s) slipped that much (or that little). In other words, these kinematic solutions may lack information about the physical dynamics of earthquake rupture that will be most helpful in forecasting future events.To help address this issue, some researchers use computer codes to numerically simulate earthquakes and construct dynamic, spontaneous
Building a dynamic code to simulate new reactor concepts
International Nuclear Information System (INIS)
Catsaros, N.; Gaveau, B.; Jaekel, M.-T.; Maillard, J.; Maurel, G.; Savva, P.; Silva, J.; Varvayanni, M.
2012-01-01
Highlights: ► We develop a stochastic neutronic code based on an existing High Energy Physics code. ► The code simulates innovative reactor designs including Accelerator Driven Systems. ► Core materials evolution will be dynamically simulated, including fuel burnup. ► Continuous feedback between the main inter-related parameters will be established. ► A description of the current research development and achievements is also given. - Abstract: Innovative nuclear reactor designs have been proposed, such as the Accelerator Driven Systems (ADSs), the “candle” reactors, etc. These reactor designs introduce computational nuclear technology problems the solution of which necessitates a new, global and dynamic computational approach of the system. A continuous feedback procedure must be established between the main inter-related parameters of the system such as the chemical, physical and isotopic composition of the core, the neutron flux distribution and the temperature field. Furthermore, as far as ADSs are concerned, the ability of the computational tool to simulate the nuclear cascade created from the interaction of accelerated protons with the spallation target as well as the produced neutrons, is also required. The new Monte Carlo code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) is being developed based on the GEANT3 High Energy Physics code, aiming to progressively satisfy all the above requirements. A description of the capabilities and methodologies implemented in the present version of ANET is given here, together with some illustrative applications of the code.
Code Shift: Grid Specifications and Dynamic Wind Turbine Models
DEFF Research Database (Denmark)
Ackermann, Thomas; Ellis, Abraham; Fortmann, Jens
2013-01-01
Grid codes (GCs) and dynamic wind turbine (WT) models are key tools to allow increasing renewable energy penetration without challenging security of supply. In this article, the state of the art and the further development of both tools are discussed, focusing on the European and North American e...
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.)
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.)
CATHENA 4. A thermalhydraulics network analysis code
International Nuclear Information System (INIS)
Aydemir, N.U.; Hanna, B.N.
2009-01-01
Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA) is a one-dimensional, non-equilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. The objective of the present paper is to describe the design, application and future development plans for the CATHENA 4 thermalhydraulics network analysis code, which is a modernized version of the present frozen CATHENA 3 code. The new code is designed in modular form, using the Fortran 95 (F95) programming language. The semi-implicit numerical integration scheme of CATHENA 3 is re-written to implement a fully-implicit methodology using Newton's iterative solution scheme suitable for nonlinear equations. The closure relations, as a first step, have been converted from the existing CATHENA 3 implementation to F95 but modularized to achieve ease of maintenance. The paper presents the field equations, followed by a description of the Newton's scheme used. The finite-difference form of the field equations is given, followed by a discussion of convergence criteria. Two applications of CATHENA 4 are presented to demonstrate the temporal and spatial convergence of the new code for problems with known solutions or available experimental data. (author)
Code system for fast reactor neutronics analysis
International Nuclear Information System (INIS)
Nakagawa, Masayuki; Abe, Junji; Sato, Wakaei.
1983-04-01
A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)
Static Code Analysis with Gitlab-CI
Datko, Szymon Tomasz
2016-01-01
Static Code Analysis is a simple but efficient way to ensure that application’s source code is free from known flaws and security vulnerabilities. Although such analysis tools are often coming with more advanced code editors, there are a lot of people who prefer less complicated environments. The easiest solution would involve education – where to get and how to use the aforementioned tools. However, counting on the manual usage of such tools still does not guarantee their actual usage. On the other hand, reducing the required effort, according to the idea “setup once, use anytime without sweat” seems like a more promising approach. In this paper, the approach to automate code scanning, within the existing CERN’s Gitlab installation, is described. For realization of that project, the Gitlab-CI service (the “CI” stands for "Continuous Integration"), with Docker assistance, was employed to provide a variety of static code analysers for different programming languages. This document covers the gene...
Turbo Pascal Computer Code for PIXE Analysis
International Nuclear Information System (INIS)
Darsono
2002-01-01
To optimal utilization of the 150 kV ion accelerator facilities and to govern the analysis technique using ion accelerator, the research and development of low energy PIXE technology has been done. The R and D for hardware of the low energy PIXE installation in P3TM have been carried on since year 2000. To support the R and D of PIXE accelerator facilities in harmonize with the R and D of the PIXE hardware, the development of PIXE software for analysis is also needed. The development of database of PIXE software for analysis using turbo Pascal computer code is reported in this paper. This computer code computes the ionization cross-section, the fluorescence yield, and the stopping power of elements also it computes the coefficient attenuation of X- rays energy. The computer code is named PIXEDASIS and it is part of big computer code planed for PIXE analysis that will be constructed in the near future. PIXEDASIS is designed to be communicative with the user. It has the input from the keyboard. The output shows in the PC monitor, which also can be printed. The performance test of the PIXEDASIS shows that it can be operated well and it can provide data agreement with data form other literatures. (author)
Uncertainty analysis of the FRAP code
International Nuclear Information System (INIS)
Peck, S.O.
1978-01-01
A user oriented, automated uncertainty analysis capability has been built into the Fuel Rod Analysis Program (FRAP) code and has been applied to a pressurized water reactor (PWR) fuel rod undergoing a loss-of-coolant accident (LOCA). The method of uncertainty analysis is the response surface method. The automated version significantly reduced the time required to complete the analysis and, at the same time, greatly increased the problem scope. Results of the analysis showed a significant difference in the total and relative contributions to the uncertainty of the response parameters between steady state and transient conditions
Development of dynamic simulation code for fuel cycle fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Aoki, Isao; Seki, Yasushi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan); Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1999-02-01
A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
Coding considerations for standalone molecular dynamics simulations of atomistic structures
Ocaya, R. O.; Terblans, J. J.
2017-10-01
The laws of Newtonian mechanics allow ab-initio molecular dynamics to model and simulate particle trajectories in material science by defining a differentiable potential function. This paper discusses some considerations for the coding of ab-initio programs for simulation on a standalone computer and illustrates the approach by C language codes in the context of embedded metallic atoms in the face-centred cubic structure. The algorithms use velocity-time integration to determine particle parameter evolution for up to several thousands of particles in a thermodynamical ensemble. Such functions are reusable and can be placed in a redistributable header library file. While there are both commercial and free packages available, their heuristic nature prevents dissection. In addition, developing own codes has the obvious advantage of teaching techniques applicable to new problems.
Benchmark calculation of subchannel analysis codes
International Nuclear Information System (INIS)
1996-02-01
In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)
Sandia National Laboratories analysis code data base
Energy Technology Data Exchange (ETDEWEB)
Peterson, C.W.
1994-11-01
Sandia National Laboratories, mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The Laboratories` strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia`s technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code ``ownership`` and release status, and references describing the physical models and numerical implementation.
Sandia National Laboratories analysis code data base
Peterson, C. W.
1994-11-01
Sandia National Laboratories' mission is to solve important problems in the areas of national defense, energy security, environmental integrity, and industrial technology. The laboratories' strategy for accomplishing this mission is to conduct research to provide an understanding of the important physical phenomena underlying any problem, and then to construct validated computational models of the phenomena which can be used as tools to solve the problem. In the course of implementing this strategy, Sandia's technical staff has produced a wide variety of numerical problem-solving tools which they use regularly in the design, analysis, performance prediction, and optimization of Sandia components, systems, and manufacturing processes. This report provides the relevant technical and accessibility data on the numerical codes used at Sandia, including information on the technical competency or capability area that each code addresses, code 'ownership' and release status, and references describing the physical models and numerical implementation.
A suite of exercises for verifying dynamic earthquake rupture codes
Harris, Ruth A.; Barall, Michael; Aagaard, Brad T.; Ma, Shuo; Roten, Daniel; Olsen, Kim B.; Duan, Benchun; Liu, Dunyu; Luo, Bin; Bai, Kangchen; Ampuero, Jean-Paul; Kaneko, Yoshihiro; Gabriel, Alice-Agnes; Duru, Kenneth; Ulrich, Thomas; Wollherr, Stephanie; Shi, Zheqiang; Dunham, Eric; Bydlon, Sam; Zhang, Zhenguo; Chen, Xiaofei; Somala, Surendra N.; Pelties, Christian; Tago, Josue; Cruz-Atienza, Victor Manuel; Kozdon, Jeremy; Daub, Eric; Aslam, Khurram; Kase, Yuko; Withers, Kyle; Dalguer, Luis
2018-01-01
We describe a set of benchmark exercises that are designed to test if computer codes that simulate dynamic earthquake rupture are working as intended. These types of computer codes are often used to understand how earthquakes operate, and they produce simulation results that include earthquake size, amounts of fault slip, and the patterns of ground shaking and crustal deformation. The benchmark exercises examine a range of features that scientists incorporate in their dynamic earthquake rupture simulations. These include implementations of simple or complex fault geometry, off‐fault rock response to an earthquake, stress conditions, and a variety of formulations for fault friction. Many of the benchmarks were designed to investigate scientific problems at the forefronts of earthquake physics and strong ground motions research. The exercises are freely available on our website for use by the scientific community.
Development of chemical equilibrium analysis code 'CHEEQ'
International Nuclear Information System (INIS)
Nagai, Shuichiro
2006-08-01
'CHEEQ' code which calculates the partial pressure and the mass of the system consisting of ideal gas and pure condensed phase compounds, was developed. Characteristics of 'CHEEQ' code are as follows. All the chemical equilibrium equations were described by the formation reactions from the mono-atomic gases in order to simplify the code structure and input preparation. Chemical equilibrium conditions, Σν i μ i =0 for the gaseous compounds and precipitated condensed phase compounds and Σν i μ i > 0 for the non-precipitated condensed phase compounds, were applied. Where, ν i and μ i are stoichiometric coefficient and chemical potential of component i. Virtual solid model was introduced to perform the calculation of constant partial pressure condition. 'CHEEQ' was consisted of following 3 parts, (1) analysis code, zc132. f. (2) thermodynamic data base, zmdb01 and (3) input data file, zindb. 'CHEEQ' code can calculate the system which consisted of elements (max.20), condensed phase compounds (max.100) and gaseous compounds. (max.200). Thermodynamic data base, zmdb01 contains about 1000 elements and compounds, and 200 of them were Actinide elements and their compounds. This report describes the basic equations, the outline of the solution procedure and instructions to prepare the input data and to evaluate the calculation results. (author)
Recent Progress on the Marylie/Impact Beam Dynamics Code
International Nuclear Information System (INIS)
Ryne, R.D.; Qiang, J.; Bethel, E.W.; Pogorelov, I.; Shalf, J.; Siegerist, C.; Venturini, M.; Dragt, A.J.; Adelmann, A.; Abell, D.; Amundson, J.; Spentzouris, P.; Neri, F.; Walstrom, P.; Mottershead, C.T.; Samulyak, R.
2006-01-01
MARYLIE/IMPACT (ML/I) is a hybrid code that combines the beam optics capabilities of MARYLIE with the parallel Particle-In-Cell capabilities of IMPACT. In addition to combining the capabilities of these codes, ML/I has a number of powerful features, including a choice of Poisson solvers, a fifth-order rf cavity model, multiple reference particles for rf cavities, a library of soft-edge magnet models, representation of magnet systems in terms of coil stacks with possibly overlapping fields, and wakefield effects. The code allows for map production, map analysis, particle tracking, and 3D envelope tracking, all within a single, coherent user environment. ML/I has a front end that can read both MARYLIE input and MAD lattice descriptions. The code can model beams with or without acceleration, and with or without space charge. Developed under a US DOE Scientific Discovery through Advanced Computing (SciDAC) project, ML/I is well suited to large-scale modeling, simulations having been performed with up to 100M macroparticles. The code inherits the powerful fitting and optimizing capabilities of MARYLIE augmented for the new features of ML/I. The combination of soft-edge magnet models, high-order capability, space charge effects, and fitting/optimization capabilities, make ML/I a powerful code for a wide range of beam optics design problems. This paper provides a description of the code and its unique capabilities
Dynamic Load Balancing for PIC code using Eulerian/Lagrangian partitioning
Sauget, Marc; Latu, Guillaume
2017-01-01
This document presents an analysis of different load balance strategies for a Plasma physics code that models high energy particle beams with PIC method. A comparison of different load balancing algorithms is given: static or dynamic ones. Lagrangian and Eulerian partitioning techniques have been investigated.
GANDALF - Graphical Astrophysics code for N-body Dynamics And Lagrangian Fluids
Hubber, D. A.; Rosotti, G. P.; Booth, R. A.
2018-01-01
GANDALF is a new hydrodynamics and N-body dynamics code designed for investigating planet formation, star formation and star cluster problems. GANDALF is written in C++, parallelized with both OPENMP and MPI and contains a PYTHON library for analysis and visualization. The code has been written with a fully object-oriented approach to easily allow user-defined implementations of physics modules or other algorithms. The code currently contains implementations of smoothed particle hydrodynamics, meshless finite-volume and collisional N-body schemes, but can easily be adapted to include additional particle schemes. We present in this paper the details of its implementation, results from the test suite, serial and parallel performance results and discuss the planned future development. The code is freely available as an open source project on the code-hosting website github at https://github.com/gandalfcode/gandalf and is available under the GPLv2 license.
Development of the criticality accident analysis code, AGNES
International Nuclear Information System (INIS)
Nakajima, Ken
1989-01-01
In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)
A System Structure for a VHTR-SI Process Dynamic Simulation Code
International Nuclear Information System (INIS)
Chang, Jiwoon; Shin, Youngjoon; Kim, Jihwan; Lee, Kiyoung; Lee, Wonjae; Chang, Jonghwa; Youn, Cheung
2008-01-01
The VHTR-SI process dynamic simulation code embedded in a mathematical solution engine is an application software system that simulates the dynamic behavior of the VHTR-SI process. Also, the software system supports a user friendly graphical user interface (GUI) for user input/out. Structured analysis techniques were developed in the late 1970s by Yourdon, DeMarco, Gane and Sarson for applying a systematic approach to a systems analysis. It included the use of data flow diagrams and data modeling and fostered the use of an implementation-independent graphical notation for a documentation. In this paper, we present a system structure for a VHRT-SI process dynamic simulation code by using the methodologies of structured analysis
Code comparison for accelerator design and analysis
International Nuclear Information System (INIS)
Parsa, Z.
1988-01-01
We present a comparison between results obtained from standard accelerator physics codes used for the design and analysis of synchrotrons and storage rings, with programs SYNCH, MAD, HARMON, PATRICIA, PATPET, BETA, DIMAD, MARYLIE and RACE-TRACK. In our analysis we have considered 5 (various size) lattices with large and small angles including AGS Booster (10/degree/ bend), RHIC (2.24/degree/), SXLS, XLS (XUV ring with 45/degree/ bend) and X-RAY rings. The differences in the integration methods used and the treatment of the fringe fields in these codes could lead to different results. The inclusion of nonlinear (e.g., dipole) terms may be necessary in these calculations specially for a small ring. 12 refs., 6 figs., 10 tabs
Electromagnetic field and mechanical stress analysis code
International Nuclear Information System (INIS)
1978-01-01
Analysis TEXMAGST is a two stage linear finite element code for the analysis of static magnetic fields in three dimensional structures and associated mechanical stresses produced by the anti J x anti B forces within these structures. The electromagnetic problem is solved in terms of magnetic vector potential A for a given current density anti J as curl 1/μ curl anti A = anti J considering the magnetic permeability as constant. The Coulombian gauge (div anti A = o) was chosen and was implemented through the use of Lagrange multipliers. The second stage of the problem - the calculation of mechanical stresses in the same three dimensional structure is solved by using the same code with few modifications - through a restart card. Body forces anti J x anti B within each element are calculated from the solution of the first stage run and represent the input to the second stage run which will give the solution for the stress problem
Web interface for plasma analysis codes
Energy Technology Data Exchange (ETDEWEB)
Emoto, M. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)], E-mail: emo@nifs.ac.jp; Murakami, S. [Kyoto University, Yoshida-Honmachi, Sakyo-ku, Kyoto 606-8501 (Japan); Yoshida, M.; Funaba, H.; Nagayama, Y. [National Institute for Fusion Science, 322-6 Oroshi, Toki, Gifu 509-5292 (Japan)
2008-04-15
There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach.
Web interface for plasma analysis codes
International Nuclear Information System (INIS)
Emoto, M.; Murakami, S.; Yoshida, M.; Funaba, H.; Nagayama, Y.
2008-01-01
There are many analysis codes that analyze various aspects of plasma physics. However, most of them are FORTRAN programs that are written to be run in supercomputers. On the other hand, many scientists use GUI (graphical user interface)-based operating systems. For those who are not familiar with supercomputers, it is a difficult task to run analysis codes in supercomputers, and they often hesitate to use these programs to substantiate their ideas. Furthermore, these analysis codes are written for personal use, and the programmers do not expect these programs to be run by other users. In order to make these programs to be widely used by many users, the authors developed user-friendly interfaces using a Web interface. Since the Web browser is one of the most common applications, it is useful for both the users and developers. In order to realize interactive Web interface, AJAX technique is widely used, and the authors also adopted AJAX. To build such an AJAX based Web system, Ruby on Rails plays an important role in this system. Since this application framework, which is written in Ruby, abstracts the Web interfaces necessary to implement AJAX and database functions, it enables the programmers to efficiently develop the Web-based application. In this paper, the authors will introduce the system and demonstrate the usefulness of this approach
High performance computer code for molecular dynamics simulations
International Nuclear Information System (INIS)
Levay, I.; Toekesi, K.
2007-01-01
Complete text of publication follows. Molecular Dynamics (MD) simulation is a widely used technique for modeling complicated physical phenomena. Since 2005 we are developing a MD simulations code for PC computers. The computer code is written in C++ object oriented programming language. The aim of our work is twofold: a) to develop a fast computer code for the study of random walk of guest atoms in Be crystal, b) 3 dimensional (3D) visualization of the particles motion. In this case we mimic the motion of the guest atoms in the crystal (diffusion-type motion), and the motion of atoms in the crystallattice (crystal deformation). Nowadays, it is common to use Graphics Devices in intensive computational problems. There are several ways to use this extreme processing performance, but never before was so easy to programming these devices as now. The CUDA (Compute Unified Device) Architecture introduced by nVidia Corporation in 2007 is a very useful for every processor hungry application. A Unified-architecture GPU include 96-128, or more stream processors, so the raw calculation performance is 576(!) GFLOPS. It is ten times faster, than the fastest dual Core CPU [Fig.1]. Our improved MD simulation software uses this new technology, which speed up our software and the code run 10 times faster in the critical calculation code segment. Although the GPU is a very powerful tool, it has a strongly paralleled structure. It means, that we have to create an algorithm, which works on several processors without deadlock. Our code currently uses 256 threads, shared and constant on-chip memory, instead of global memory, which is 100 times slower than others. It is possible to implement the total algorithm on GPU, therefore we do not need to download and upload the data in every iteration. On behalf of maximal throughput, every thread run with the same instructions
Introduction of thermal-hydraulic analysis code and system analysis code for HTGR
International Nuclear Information System (INIS)
Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi
1984-01-01
Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)
Operational reactor physics analysis codes (ORPAC)
International Nuclear Information System (INIS)
Kumar, Jainendra; Singh, K.P.; Singh, Kanchhi
2007-07-01
For efficient, smooth and safe operation of a nuclear research reactor, many reactor physics evaluations are regularly required. As part of reactor core management the important activities are maintaining core reactivity status, core power distribution, xenon estimations, safety evaluation of in-pile irradiation samples and experimental assemblies and assessment of nuclear safety in fuel handling/storage. In-pile irradiation of samples requires a prior estimation of the reactivity load due to the sample, the heating rate and the activity developed in it during irradiation. For the safety of personnel handling irradiated samples the dose rate at the surface of shielded flask housing the irradiated sample should be less than 200 mR/Hr.Therefore, a proper shielding and radioactive cooling of the irradiated sample are required to meet the said requirement. Knowledge of xenon load variation with time (Startup-curve) helps in estimating Xenon override time. Monitoring of power in individual fuel channels during reactor operation is essential to know any abnormal power distribution to avoid unsafe situations. Complexities in the estimation of above mentioned reactor parameters and their frequent requirement compel one to use computer codes to avoid possible human errors. For efficient and quick evaluation of parameters related to reactor operations such as xenon load, critical moderator height and nuclear heating and reactivity load of isotope samples/experimental assembly, a computer code ORPAC (Operational Reactor Physics Analysis Codes) has been developed. This code is being used for regular assessment of reactor physics parameters in Dhruva and Cirus. The code ORPAC written in Visual Basic 6.0 environment incorporates several important operational reactor physics aspects on a single platform with graphical user interfaces (GUI) to make it more user-friendly and presentable. (author)
Development of disruption thermal analysis code DREAM
Energy Technology Data Exchange (ETDEWEB)
Yamazaki, Seiichiro; Kobayahsi, Takeshi [Kawasaki Heavy Industries Ltd., Kobe (Japan); Seki, Masahiro
1989-07-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author).
Development of disruption thermal analysis code DREAM
International Nuclear Information System (INIS)
Yamazaki, Seiichiro; Kobayahsi, Takeshi; Seki, Masahiro.
1989-01-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing componenets such as first wall and divertor/limiter are subjected to a intensse heat load in a short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs. It causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes and radiation heat loss in required in the design of these components. This paper describes the computer code DREAM, developed to perform the disruption thermal analysis, taking phase changes and radiation into account. (author)
CONTEMPT-DG containment analysis code
International Nuclear Information System (INIS)
Deem, R.E.; Rousseau, K.
1982-01-01
The assessment of hydrogen burning in a containment building during a degraded core event requires a knowledge of various system responses. These system responses (i.e. heat sinks, fan cooler units, sprays, etc.) can have a marked effect on the overall containment integrity results during a hydrogen burn. In an attempt to properly handle the various system responses and still retain the capability to perform sensitivity analysis on various parameters, the CONTEMPT-DG computer code was developed. This paper will address the historical development of the code, its various features, and the rationale for its development. Comparisons between results from the CONTEMPT-DG analyses and results from similar MARCH analyses will also be given
Fluid dynamics and heat transfer methods for the TRAC code
International Nuclear Information System (INIS)
Reed, W.H.; Kirchner, W.L.
1977-01-01
A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)
Fluid dynamics and heat transfer methods for the TRAC code
International Nuclear Information System (INIS)
Reed, W.H.; Kirchner, W.L.
1977-01-01
A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, new highly implicit difference techniques are developed that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained
Aeschliman, D. P.; Oberkampf, W. L.; Blottner, F. G.
Verification, calibration, and validation (VCV) of Computational Fluid Dynamics (CFD) codes is an essential element of the code development process. The exact manner in which code VCV activities are planned and conducted, however, is critically important. It is suggested that the way in which code validation, in particular, is often conducted--by comparison to published experimental data obtained for other purposes--is in general difficult and unsatisfactory, and that a different approach is required. This paper describes a proposed methodology for CFD code VCV that meets the technical requirements and is philosophically consistent with code development needs. The proposed methodology stresses teamwork and cooperation between code developers and experimentalists throughout the VCV process, and takes advantage of certain synergisms between CFD and experiment. A novel approach to uncertainty analysis is described which can both distinguish between and quantify various types of experimental error, and whose attributes are used to help define an appropriate experimental design for code VCV experiments. The methodology is demonstrated with an example of laminar, hypersonic, near perfect gas, 3-dimensional flow over a sliced sphere/cone of varying geometrical complexity.
Extending CANTUP code analysis to probabilistic evaluations
International Nuclear Information System (INIS)
Florea, S.
2001-01-01
The structural analysis with numerical methods based on final element method plays at present a central role in evaluations and predictions of structural systems which require safety and reliable operation in aggressive environmental conditions. This is the case too for the CANDU - 600 fuel channel, where besides the corrosive and thermal aggression upon the Zr97.5Nb2.5 pressure tubes, a lasting irradiation adds which has marked consequences upon the materials properties evolution. This results in an unavoidable spreading in the materials properties in time, affected by high uncertainties. Consequently, the deterministic evaluation with computation codes based on finite element method are supplemented by statistic and probabilistic methods of evaluation of the response of structural components. This paper reports the works on extending the thermo-mechanical evaluation of the fuel channel components in the frame of probabilistic structure mechanics based on statistical methods and developed upon deterministic CANTUP code analyses. CANTUP code was adapted from LAHEY 77 platform onto Microsoft Developer Studio - Fortran Power Station 4.0 platform. To test the statistical evaluation of the creeping behaviour of pressure tube, the value of longitudinal elasticity modulus (Young) was used, as random variable, with a normal distribution around value, as used in deterministic analyses. The influence of the random quantity upon the hog and effective stress developed in the pressure tube for to time values, specific to primary and secondary creep was studied. The results obtained after a five year creep, corresponding to the secondary creep are presented
Learning to Estimate Dynamical State with Probabilistic Population Codes.
Directory of Open Access Journals (Sweden)
Joseph G Makin
2015-11-01
Full Text Available Tracking moving objects, including one's own body, is a fundamental ability of higher organisms, playing a central role in many perceptual and motor tasks. While it is unknown how the brain learns to follow and predict the dynamics of objects, it is known that this process of state estimation can be learned purely from the statistics of noisy observations. When the dynamics are simply linear with additive Gaussian noise, the optimal solution is the well known Kalman filter (KF, the parameters of which can be learned via latent-variable density estimation (the EM algorithm. The brain does not, however, directly manipulate matrices and vectors, but instead appears to represent probability distributions with the firing rates of population of neurons, "probabilistic population codes." We show that a recurrent neural network-a modified form of an exponential family harmonium (EFH-that takes a linear probabilistic population code as input can learn, without supervision, to estimate the state of a linear dynamical system. After observing a series of population responses (spike counts to the position of a moving object, the network learns to represent the velocity of the object and forms nearly optimal predictions about the position at the next time-step. This result builds on our previous work showing that a similar network can learn to perform multisensory integration and coordinate transformations for static stimuli. The receptive fields of the trained network also make qualitative predictions about the developing and learning brain: tuning gradually emerges for higher-order dynamical states not explicitly present in the inputs, appearing as delayed tuning for the lower-order states.
Use of computer codes for system reliability analysis
International Nuclear Information System (INIS)
Sabek, M.; Gaafar, M.; Poucet, A.
1988-01-01
This paper gives a collective summary of the studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRANTIC, FTAP, computer code package RALLY, and BOUNDS codes. Two reference study cases were executed by each code. The results obtained logic/probabilistic analysis as well as computation time are compared
Porting of a serial molecular dynamics code on MIMD platforms
Energy Technology Data Exchange (ETDEWEB)
Celino, M. [ENEA Centro Ricerche Casaccia, S. Maria di Galeria, RM (Italy). HPCN Project
1999-07-01
A molecular dynamics (MD) code, utilized for the study of atomistic models of metallic systems has been parallelized for MIMD (multiple instructions multiple data) parallel platforms by means of the parallel virtual machine (PVM) message passing library. Since the parallelization implies modifications of the sequential algorithms, these are described from the point of view of the statistical mechanical theory. Furthermore, techniques and parallelization strategies utilized and the MD parallel code are described in detail. Benchmarks on several MIMD platforms (IBM SP1, SP2, Cray T3D, cluster of workstations) allow performances evaluation of the code versus the different characteristics of the parallel platforms. [Italian] Un codice seriale di dinamica molecolare (MD) utilizzato per lo studio di modelli atomici di materiali metallici e' stato parallelizzato per piattaforme parallele MIMD (multiple instructions multiple data) utilizzando librerie del parallel virtual machine (PVM). Poiche' l'operazione di parallelizzazione ha implicato la modifica degli algoritmi seriali del codice, questi vengono descritti ripercorrendo i concetti fondamentali della meccanica statistica. Inoltre sono presentate le tecniche e le strategie di parallelizzazione utilizzate descrivendo in dettaglio il codice parallelo di MD: Risultati di benchmark su diverse piattaforme MIMD (IBM SP1, SP2, Cray T3D, cluster of workstations) permettono di analizzare le performances del codice in funzione delle differenti caratteristiche delle piattaforme parallele.
Systemization of burnup sensitivity analysis code
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2004-02-01
To practical use of fact reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoints of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor core 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, development of a analysis code for burnup sensitivity, SAGEP-BURN, has been done and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to user due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functionalities in the existing large system. It is not sufficient to unify each computational component for some reasons; computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For this
Associating schizophrenia, long non-coding RNAs and neurostructural dynamics
Merelo, Veronica; Durand, Dante; Lescallette, Adam R.; Vrana, Kent E.; Hong, L. Elliot; Faghihi, Mohammad Ali; Bellon, Alfredo
2015-01-01
Several lines of evidence indicate that schizophrenia has a strong genetic component. But the exact nature and functional role of this genetic component in the pathophysiology of this mental illness remains a mystery. Long non-coding RNAs (lncRNAs) are a recently discovered family of molecules that regulate gene transcription through a variety of means. Consequently, lncRNAs could help us bring together apparent unrelated findings in schizophrenia; namely, genomic deficiencies on one side and neuroimaging, as well as postmortem results on the other. In fact, the most consistent finding in schizophrenia is decreased brain size together with enlarged ventricles. This anomaly appears to originate from shorter and less ramified dendrites and axons. But a decrease in neuronal arborizations cannot explain the complex pathophysiology of this psychotic disorder; however, dynamic changes in neuronal structure present throughout life could. It is well recognized that the structure of developing neurons is extremely plastic. This structural plasticity was thought to stop with brain development. However, breakthrough discoveries have shown that neuronal structure retains some degree of plasticity throughout life. What the neuroscientific field is still trying to understand is how these dynamic changes are regulated and lncRNAs represent promising candidates to fill this knowledge gap. Here, we present evidence that associates specific lncRNAs with schizophrenia. We then discuss the potential role of lncRNAs in neurostructural dynamics. Finally, we explain how dynamic neurostructural modifications present throughout life could, in theory, reconcile apparent unrelated findings in schizophrenia. PMID:26483630
Associating schizophrenia, long non-coding RNAs and neurostructural dynamics
Directory of Open Access Journals (Sweden)
Veronica eMerelo
2015-09-01
Full Text Available Several lines of evidence indicate that schizophrenia has a strong genetic component. But the exact nature and functional role of this genetic component in the pathophysiology of this mental illness remains a mystery. Long non-coding RNAs (lncRNAs are a recently discovered family of molecules that regulate gene transcription through a variety of means. Consequently, lncRNAs could help us bring together apparent unrelated findings in schizophrenia; namely, genomic deficiencies on one side and neuroimaging, as well as postmortem results on the other. In fact, the most consistent finding in schizophrenia is decreased brain size together with enlarged ventricles. This anomaly appears to originate from shorter and less ramified dendrites and axons. But a decrease in neuronal arborizations cannot explain the complex pathophysiology of this psychotic disorder; however, dynamic changes in neuronal structure present throughout life could. It is well recognized that the structure of developing neurons is extremely plastic. This structural plasticity was thought to stop with brain development. However, breakthrough discoveries have shown that neuronal structure retains some degree of plasticity throughout life. What the neuroscientific field is still trying to understand is how these dynamic changes are regulated and lncRNAs represent promising candidates to fill this knowledge gap. Here, we present evidence that associates specific lncRNAs with schizophrenia. We then discuss the potential role of lncRNAs in neurostructural dynamics. Finally, we explain how dynamic neurostructural modifications present throughout life could, in theory, reconcile apparent unrelated findings in schizophrenia.
Predictive coding of dynamical variables in balanced spiking networks.
Boerlin, Martin; Machens, Christian K; Denève, Sophie
2013-01-01
Two observations about the cortex have puzzled neuroscientists for a long time. First, neural responses are highly variable. Second, the level of excitation and inhibition received by each neuron is tightly balanced at all times. Here, we demonstrate that both properties are necessary consequences of neural networks that represent information efficiently in their spikes. We illustrate this insight with spiking networks that represent dynamical variables. Our approach is based on two assumptions: We assume that information about dynamical variables can be read out linearly from neural spike trains, and we assume that neurons only fire a spike if that improves the representation of the dynamical variables. Based on these assumptions, we derive a network of leaky integrate-and-fire neurons that is able to implement arbitrary linear dynamical systems. We show that the membrane voltage of the neurons is equivalent to a prediction error about a common population-level signal. Among other things, our approach allows us to construct an integrator network of spiking neurons that is robust against many perturbations. Most importantly, neural variability in our networks cannot be equated to noise. Despite exhibiting the same single unit properties as widely used population code models (e.g. tuning curves, Poisson distributed spike trains), balanced networks are orders of magnitudes more reliable. Our approach suggests that spikes do matter when considering how the brain computes, and that the reliability of cortical representations could have been strongly underestimated.
Systemization of burnup sensitivity analysis code. 2
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2005-02-01
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of criticality experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristics is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons; the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion. For
Incoherent scatter studies of upper atmosphere dynamics and coding technique
International Nuclear Information System (INIS)
Haeggstroem, Ingemar.
1990-09-01
Observations by the EISCAT incoherent scatter radar are used to study the dynamics of the auroral upper atmosphere. The study describes some effects of the strong plasma convection occurring at these latitudes and a new coding technique for incoherent scatter radars. A technique to determine the thermospheric neutral wind from incoherent scatter measurements is described. Simultaneous Fabry-Perot interferometer measurements of the wind are compared with those derived from the radar data. F-region electron density depletions in the afternoon/evening sector of the auroral zone, identified as the main ionospheric trough, are investigated. In a statistical study, based on wide latitude scanning experiment made at solar minimum, the trough appearance at a given latitude is compared to the geomagnetic index K p , and an empirical model predicting the latitude of the trough is proposed. Detailed studies, using different experiment modes, show that the equatorward edge of the auroral oval is co-located of up to 1 degree poleward of the trough minimum, which in turn is co-located with the peak convective electric field, with its boundary 1 degree - 2 degree equatorward of the trough minimum. It is shown that the F-region ion composition changes from pure 0 + to molecular ion dominated (NO + /O 2 + ) as the trough moves into the region probed by the radar. In a special case, where a geomagnetic sudden impulse caused an expansion of the plasma convection pattern, the equatorward trough progression is studied together with ionosonde measurements. A new coding technique for incoherent scatter radar measurement is introduced and described. The method, called alternating codes, provides significantly more accurate estimates of the plasma parameters than can be obtained by frequency commutated multipulse measurements. Simple explanations of the method are given as well as a precise definition. Two examples of application of the alternating codes are presented, showing the high
Jellali, Nabiha; Najjar, Monia; Ferchichi, Moez; Rezig, Houria
2017-07-01
In this paper, a new two-dimensional spectral/spatial codes family, named two dimensional dynamic cyclic shift codes (2D-DCS) is introduced. The 2D-DCS codes are derived from the dynamic cyclic shift code for the spectral and spatial coding. The proposed system can fully eliminate the multiple access interference (MAI) by using the MAI cancellation property. The effect of shot noise, phase-induced intensity noise and thermal noise are used to analyze the code performance. In comparison with existing two dimensional (2D) codes, such as 2D perfect difference (2D-PD), 2D Extended Enhanced Double Weight (2D-Extended-EDW) and 2D hybrid (2D-FCC/MDW) codes, the numerical results show that our proposed codes have the best performance. By keeping the same code length and increasing the spatial code, the performance of our 2D-DCS system is enhanced: it provides higher data rates while using lower transmitted power and a smaller spectral width.
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.)
Current lead thermal analysis code 'CURRENT'
International Nuclear Information System (INIS)
Yamaguchi, Masahito; Tada, Eisuke; Shimamoto, Susumu; Hata, Kenichiro.
1985-08-01
Large gas-cooled current lead with the capacity more than 30 kA and 22 kV is required for superconducting toroidal and poloidal coils for fusion application. The current lead is used to carry electrical current from the power supply system at room temperature to the superconducting coil at 4 K. Accordingly, the thermal performance of the current lead is significantly important to determine the heat load requirements of the coil system at 4 K. Japan Atomic Energy Research Institute (JAERI) has being developed the large gas-cooled current leads with the optimum condition in which the heat load is around 1 W per 1 kA at 4 K. In order to design the current lead with the optimum thermal performances, JAERI developed thermal analysis code named as ''CURRENT'' which can theoretically calculate the optimum geometric shape and cooling conditions of the current lead. The basic equations and the instruction manual of the analysis code are described in this report. (author)
the use of the dynamic magnification factor in the dynamic analysis
African Journals Online (AJOL)
Uncle Greg 4 Real
bridges and some country's codes of practice made specifications in respect of the dynamic magnification factor for the analysis and design of various types of structures subjected to ... span. For example the United kingdom code. [10] presented dynamic magnification factors, related ... For SDOF (Single Degree of Freedom).
Performance testing of thermal analysis codes for nuclear fuel casks
International Nuclear Information System (INIS)
Sanchez, L.C.
1987-01-01
In 1982 Sandia National Laboratories held the First Industry/Government Joint Thermal and Structural Codes Information Exchange and presented the initial stages of an investigation of thermal analysis computer codes for use in the design of nuclear fuel shipping casks. The objective of the investigation was to (1) document publicly available computer codes, (2) assess code capabilities as determined from their user's manuals, and (3) assess code performance on cask-like model problems. Computer codes are required to handle the thermal phenomena of conduction, convection and radiation. Several of the available thermal computer codes were tested on a set of model problems to assess performance on cask-like problems. Solutions obtained with the computer codes for steady-state thermal analysis were in good agreement and the solutions for transient thermal analysis differed slightly among the computer codes due to modeling differences
Error-correction coding and decoding bounds, codes, decoders, analysis and applications
Tomlinson, Martin; Ambroze, Marcel A; Ahmed, Mohammed; Jibril, Mubarak
2017-01-01
This book discusses both the theory and practical applications of self-correcting data, commonly known as error-correcting codes. The applications included demonstrate the importance of these codes in a wide range of everyday technologies, from smartphones to secure communications and transactions. Written in a readily understandable style, the book presents the authors’ twenty-five years of research organized into five parts: Part I is concerned with the theoretical performance attainable by using error correcting codes to achieve communications efficiency in digital communications systems. Part II explores the construction of error-correcting codes and explains the different families of codes and how they are designed. Techniques are described for producing the very best codes. Part III addresses the analysis of low-density parity-check (LDPC) codes, primarily to calculate their stopping sets and low-weight codeword spectrum which determines the performance of these codes. Part IV deals with decoders desi...
76 FR 64931 - Building Energy Codes Cost Analysis
2011-10-19
...-0046] Building Energy Codes Cost Analysis AGENCY: Office of Energy Efficiency and Renewable Energy... reopening of the time period for submitting comments on the request for information on Building Energy Codes... the request for information on Building Energy Code Cost Analysis and provide docket number EERE-2011...
Pipe whip analysis using the TEDEL code
International Nuclear Information System (INIS)
Millard, D.; Hoffmann, A.
1985-02-01
In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented
Development of a dynamic coupled hydro-geomechanical code and its application to induced seismicity
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
Step by step parallel programming method for molecular dynamics code
International Nuclear Information System (INIS)
Orii, Shigeo; Ohta, Toshio
1996-07-01
Parallel programming for a numerical simulation program of molecular dynamics is carried out with a step-by-step programming technique using the two phase method. As a result, within the range of a certain computing parameters, it is found to obtain parallel performance by using the level of parallel programming which decomposes the calculation according to indices of do-loops into each processor on the vector parallel computer VPP500 and the scalar parallel computer Paragon. It is also found that VPP500 shows parallel performance in wider range computing parameters. The reason is that the time cost of the program parts, which can not be reduced by the do-loop level of the parallel programming, can be reduced to the negligible level by the vectorization. After that, the time consuming parts of the program are concentrated on less parts that can be accelerated by the do-loop level of the parallel programming. This report shows the step-by-step parallel programming method and the parallel performance of the molecular dynamics code on VPP500 and Paragon. (author)
Analysis of quantum error-correcting codes: Symplectic lattice codes and toric codes
Harrington, James William
Quantum information theory is concerned with identifying how quantum mechanical resources (such as entangled quantum states) can be utilized for a number of information processing tasks, including data storage, computation, communication, and cryptography. Efficient quantum algorithms and protocols have been developed for performing some tasks (e.g. , factoring large numbers, securely communicating over a public channel, and simulating quantum mechanical systems) that appear to be very difficult with just classical resources. In addition to identifying the separation between classical and quantum computational power, much of the theoretical focus in this field over the last decade has been concerned with finding novel ways of encoding quantum information that are robust against errors, which is an important step toward building practical quantum information processing devices. In this thesis I present some results on the quantum error-correcting properties of oscillator codes (also described as symplectic lattice codes) and toric codes. Any harmonic oscillator system (such as a mode of light) can be encoded with quantum information via symplectic lattice codes that are robust against shifts in the system's continuous quantum variables. I show the existence of lattice codes whose achievable rates match the one-shot coherent information over the Gaussian quantum channel. Also, I construct a family of symplectic self-dual lattices and search for optimal encodings of quantum information distributed between several oscillators. Toric codes provide encodings of quantum information into two-dimensional spin lattices that are robust against local clusters of errors and which require only local quantum operations for error correction. Numerical simulations of this system under various error models provide a calculation of the accuracy threshold for quantum memory using toric codes, which can be related to phase transitions in certain condensed matter models. I also present
Kraljić, K.; Strüngmann, L.; Fimmel, E.; Gumbel, M.
2018-01-01
The genetic code is degenerated and it is assumed that redundancy provides error detection and correction mechanisms in the translation process. However, the biological meaning of the code's structure is still under current research. This paper presents a Genetic Code Analysis Toolkit (GCAT) which provides workflows and algorithms for the analysis of the structure of nucleotide sequences. In particular, sets or sequences of codons can be transformed and tested for circularity, comma-freeness, dichotomic partitions and others. GCAT comes with a fertile editor custom-built to work with the genetic code and a batch mode for multi-sequence processing. With the ability to read FASTA files or load sequences from GenBank, the tool can be used for the mathematical and statistical analysis of existing sequence data. GCAT is Java-based and provides a plug-in concept for extensibility. Availability: Open source Homepage:http://www.gcat.bio/
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)
Analytical validation of the CACECO containment analysis code
International Nuclear Information System (INIS)
Peak, R.D.
1979-08-01
The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code
Users' guide to CACECO containment analysis code. [LMFBR
Energy Technology Data Exchange (ETDEWEB)
Peak, R.D.
1979-06-01
The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. The code is included in the National Energy Software Center Library at Argonne National Laboratory as Program No. 762. This users' guide describes the CACECO code and its data input requirements. The code description covers the many mathematical models used and the approximations used in their solution. The descriptions are detailed to the extent that the user can modify the code to suit his unique needs, and, indeed, the reader is urged to consider code modification acceptable.
Verification of the CONPAS (CONtainment Performance Analysis System) code package
International Nuclear Information System (INIS)
Kim, See Darl; Ahn, Kwang Il; Song, Yong Man; Choi, Young; Park, Soo Yong; Kim, Dong Ha; Jin, Young Ho.
1997-09-01
CONPAS is a computer code package to integrate the numerical, graphical, and results-oriented aspects of Level 2 probabilistic safety assessment (PSA) for nuclear power plants under a PC window environment automatically. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules: (1) ET Editor, (2) Computer, (3) Text Editor, and (4) Mechanistic Code Plotter. Compared with other existing computer codes for Level 2 PSA, and CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friendly interface. The computational performance of CONPAS has been verified through a Level 2 PSA to a reference plant. The results of the CONPAS code was compared with an existing level 2 PSA code (NUCAP+) and the comparison proves that CONPAS is appropriate for Level 2 PSA. (author). 9 refs., 8 tabs., 14 figs
Uranium Isotopic Analysis with the FRAM Isotopic Analysis Code
International Nuclear Information System (INIS)
Vo, D.T.; Sampson, T.E.
1999-01-01
FRAM is the acronym for Fixed-Energy Response-Function Analysis with Multiple efficiency. This software was developed at Los Alamos National Laboratory originally for plutonium isotopic analysis. Later, it was adapted for uranium isotopic analysis in addition to plutonium. It is a code based on a self-calibration using several gamma-ray peaks for determining the isotopic ratios. The versatile-parameter database structure governs all facets of the data analysis. User editing of the parameter sets allows great flexibility in handling data with different isotopic distributions, interfering isotopes, and different acquisition parameters such as energy calibration and detector type
Codeword Structure Analysis for LDPC Convolutional Codes
Directory of Open Access Journals (Sweden)
Hua Zhou
2015-12-01
Full Text Available The codewords of a low-density parity-check (LDPC convolutional code (LDPC-CC are characterised into structured and non-structured. The number of the structured codewords is dominated by the size of the polynomial syndrome former matrix H T ( D , while the number of the non-structured ones depends on the particular monomials or polynomials in H T ( D . By evaluating the relationship of the codewords between the mother code and its super codes, the low weight non-structured codewords in the super codes can be eliminated by appropriately choosing the monomials or polynomials in H T ( D , resulting in improved distance spectrum of the mother code.
The analysis of thermal-hydraulic models in MELCOR code
Energy Technology Data Exchange (ETDEWEB)
Kim, M H; Hur, C; Kim, D K; Cho, H J [POhang Univ., of Science and TECHnology, Pohang (Korea, Republic of)
1996-07-15
The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed.
Probabilistic evaluations for CANTUP computer code analysis improvement
International Nuclear Information System (INIS)
Florea, S.; Pavelescu, M.
2004-01-01
Structural analysis with finite element method is today an usual way to evaluate and predict the behavior of structural assemblies subject to hard conditions in order to ensure their safety and reliability during their operation. A CANDU 600 fuel channel is an example of an assembly working in hard conditions, in which, except the corrosive and thermal aggression, long time irradiation, with implicit consequences on material properties evolution, interferes. That leads inevitably to material time-dependent properties scattering, their dynamic evolution being subject to a great degree of uncertainness. These are the reasons for developing, in association with deterministic evaluations with computer codes, the probabilistic and statistical methods in order to predict the structural component response. This work initiates the possibility to extend the deterministic thermomechanical evaluation on fuel channel components to probabilistic structural mechanics approach starting with deterministic analysis performed with CANTUP computer code which is a code developed to predict the long term mechanical behavior of the pressure tube - calandria tube assembly. To this purpose the structure of deterministic calculus CANTUP computer code has been reviewed. The code has been adapted from LAHEY 77 platform to Microsoft Developer Studio - Fortran Power Station platform. In order to perform probabilistic evaluations, it was added a part to the deterministic code which, using a subroutine from IMSL library from Microsoft Developer Studio - Fortran Power Station platform, generates pseudo-random values of a specified value. It was simulated a normal distribution around the deterministic value and 5% standard deviation for Young modulus material property in order to verify the statistical calculus of the creep behavior. The tube deflection and effective stresses were the properties subject to probabilistic evaluation. All the values of these properties obtained for all the values for
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.
2002-01-01
A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)
Computer code for the atomistic simulation of lattice defects and dynamics. [COMENT code
Energy Technology Data Exchange (ETDEWEB)
Schiffgens, J.O.; Graves, N.J.; Oster, C.A.
1980-04-01
This document has been prepared to satisfy the need for a detailed, up-to-date description of a computer code that can be used to simulate phenomena on an atomistic level. COMENT was written in FORTRAN IV and COMPASS (CDC assembly language) to solve the classical equations of motion for a large number of atoms interacting according to a given force law, and to perform the desired ancillary analysis of the resulting data. COMENT is a dual-purpose intended to describe static defect configurations as well as the detailed motion of atoms in a crystal lattice. It can be used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect migration, and defect stability.
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
Dynamic Coding of Signed Quantities in Cortical Feedback Circuits
Directory of Open Access Journals (Sweden)
Dana eBallard
2012-08-01
Full Text Available In the early sensory and motor areas of the cortex, individual neurons transmit information about specific sensory features via a peaked response. This concept has been crystallized as `labeled lines,' to denote that axons communicate the specific properties of their sensory or motor parent cell. Such cells also can be characterized as being polarized, that is, as representing a signed quantity that is either positive or negative. We show in a model simulation that there are two important consequences when learning receptive fields using such signed codings in circuits that subtract different inputs. The first is that, in feedback circuits using labeled lines, such arithmetic operations need to be distributed across multiple distinct pathways. The second consequence is that such pathways must be necessarily dynamic, i.e. that synapses can grow and retract when forming receptive fields. The model monitors the breaking and growing of new circuit connections when their synapses need to change polarities and predicts that the rate of such changes should be inversely correlated with the progress of receptive field formation.
OPR1000 RCP Flow Coastdown Analysis using SPACE Code
Energy Technology Data Exchange (ETDEWEB)
Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)
2016-10-15
The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.
The importance of including dynamic soil-structure interaction into wind turbine simulation codes
DEFF Research Database (Denmark)
Damgaard, Mads; Andersen, Lars Vabbersgaard; Ibsen, Lars Bo
2014-01-01
A rigorous numerical model, describing a wind turbine structure and subsoil, may contain thousands of degrees of freedom, making the approach computationally inefficient for fast time domain analysis. In order to meet the requirements of real-time calculations, the dynamic impedance of the founda......A rigorous numerical model, describing a wind turbine structure and subsoil, may contain thousands of degrees of freedom, making the approach computationally inefficient for fast time domain analysis. In order to meet the requirements of real-time calculations, the dynamic impedance...... of the foundation from a rigorous analysis can be formulated into a so-called lumped-parameter model consisting of a few springs, dashpots and point masses which are easily implemented into aeroelastic codes. In this paper, the quality of consistent lumped-parameter models of rigid surface footings and mono piles...... is examined. The optimal order of the models is determined and implemented into the aeroelastic code HAWC2, where the dynamic response of a 5.0 MW wind turbine is evaluated. In contrast to the fore-aft vibrations, the inclusion of soil-structure interaction is shown to be critical for the side-side vibrations...
Fast neutron analysis code SAD1
International Nuclear Information System (INIS)
Jung, M.; Ott, C.
1985-01-01
A listing and an example of outputs of the M.C. code SAD1 are given here. This code has been used many times to predict responses of fast neutrons in hydrogenic materials (in our case emulsions or plastics) towards the elastic n, p scattering. It can be easily extended to other kinds of such materials and to any kind of incident fast neutron spectrum
Development of time dependent safety analysis code for plasma anomaly events in fusion reactors
International Nuclear Information System (INIS)
Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.
1997-01-01
A safety analysis code SAFALY has been developed to analyze plasma anomaly events in fusion reactors, e.g., a loss of plasma control. The code is a hybrid code comprising a zero-dimensional plasma dynamics and a one-dimensional thermal analysis of in-vessel components. The code evaluates the time evolution of plasma parameters and temperature distributions of in-vessel components. As the plasma-safety interface model, we proposed a robust plasma physics model taking into account updated data for safety assessment. For example, physics safety guidelines for beta limit, density limit and H-L mode confinement transition threshold power, etc. are provided in the model. The model of the in-vessel components are divided into twenty temperature regions in the poloidal direction taking account of radiative heat transfer between each surface of each region. This code can also describe the coolant behavior under hydraulic accidents with the results by hydraulics code and treat vaporization (sublimation) from plasma facing components (PFCs). Furthermore, the code includes the model of impurity transport form PFCs by using a transport probability and a time delay. Quantitative analysis based on the model is possible for a scenario of plasma passive shutdown. We examined the possibility of the code as a safety analysis code for plasma anomaly events in fusion reactors and had a prospect that it would contribute to the safety analysis of the International Thermonuclear Experimental Reactor (ITER). (author)
Challenges on innovations of newly-developed safety analysis codes
International Nuclear Information System (INIS)
Yang, Yanhua; Zhang, Hao
2016-01-01
With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.
Challenges on innovations of newly-developed safety analysis codes
Energy Technology Data Exchange (ETDEWEB)
Yang, Yanhua [Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering; Zhang, Hao [State Nuclear Power Software Development Center, Beijing (China). Beijing Future Science and Technology City
2016-05-15
With the development of safety analysis method, the safety analysis codes meet more challenges. Three challenges are presented in this paper, which are mathematic model, code design and user interface. Combined with the self-reliance safety analysis code named COSINE, the ways of meeting these requirements are suggested, that is to develop multi-phases, multi-fields and multi-dimension models, to adopt object-oriented code design ideal and to improve the way of modeling, calculation control and data post-processing in the user interface.
ZERBERUS - the code for reliability analysis of crack containing structures
International Nuclear Information System (INIS)
Cizelj, L.; Riesch-Oppermann, H.
1992-04-01
Brief description of the First- and Second Order Reliability Methods, being the theoretical background of the code, is given. The code structure is described in detail, with special emphasis to the new application fields. The numerical example investigates failure probability of steam generator tubing affected by stress corrosion cracking. The changes necessary to accommodate this analysis within the ZERBERUS code are explained. Analysis results are compared to different Monte Carlo techniques. (orig./HP) [de
Manometer Behavior Analysis using CATHENA, RELAP and GOTHIC Codes
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Han, Kee Soo; Moon, Bok Ja; Jang, Misuk [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2014-05-15
In this presentation, simple thermal hydraulic behavior is analyzed using three codes to show the possibility of using alternative codes. We established three models of simple u-tube manometer using three different codes. CATHENA (Canadian Algorithm for Thermal hydraulic Network Analysis), RELAP (Reactor Excursion and Leak Analysis Program), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are used for this analysis. CATHENA and RELAP are widely used codes for the analysis of system behavior of CANDU and PWR. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. In this paper, the internal behavior of u-tube manometer was analyzed using 3 codes, CATHENA, RELAP and GOTHIC. The general transient behavior is similar among 3 codes. However, the behavior simulated using GOTHIC shows some different trend compared with the results from the other 2 codes at the end of the transient. It would be resulted from the use of different physical model in GOTHIC, which is specialized for the multi-phase thermal hydraulic behavior analysis of containment system unlike the other two codes.
Content Analysis Coding Schemes for Online Asynchronous Discussion
Weltzer-Ward, Lisa
2011-01-01
Purpose: Researchers commonly utilize coding-based analysis of classroom asynchronous discussion contributions as part of studies of online learning and instruction. However, this analysis is inconsistent from study to study with over 50 coding schemes and procedures applied in the last eight years. The aim of this article is to provide a basis…
A Semantic Analysis Method for Scientific and Engineering Code
Stewart, Mark E. M.
1998-01-01
This paper develops a procedure to statically analyze aspects of the meaning or semantics of scientific and engineering code. The analysis involves adding semantic declarations to a user's code and parsing this semantic knowledge with the original code using multiple expert parsers. These semantic parsers are designed to recognize formulae in different disciplines including physical and mathematical formulae and geometrical position in a numerical scheme. In practice, a user would submit code with semantic declarations of primitive variables to the analysis procedure, and its semantic parsers would automatically recognize and document some static, semantic concepts and locate some program semantic errors. A prototype implementation of this analysis procedure is demonstrated. Further, the relationship between the fundamental algebraic manipulations of equations and the parsing of expressions is explained. This ability to locate some semantic errors and document semantic concepts in scientific and engineering code should reduce the time, risk, and effort of developing and using these codes.
Energy Technology Data Exchange (ETDEWEB)
Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Itakura, Hirofumi
1996-11-01
Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The `ACCORD` code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the `ACCORD` code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)
International Nuclear Information System (INIS)
Takeda, Takeshi; Tachibana, Yukio; Kunitomi, Kazuhiko; Itakura, Hirofumi.
1996-11-01
Safety demonstration test of the High Temperature Engineering Test Reactor will be carried out to demonstrate excellent safety features of a next generation High Temperature Gas-cooled Reactor (HTGR). Analytical code for incore and plant dynamics is necessary to assess the results of the safety demonstration test and to perform a design and safety analysis of the next generation HTGR. Existing analytical code for incore and plant dynamics of the HTGR can analyze behavior of plant system for only several thousand seconds after an event occurrence. Simulator on site can analyze only behavior of specific plant system. The 'ACCORD' code has been, therefore, developed to analyze the incore and plant dynamics of the HTGR. The followings are the major characteristics of this code. (1) Plant system can be analyzed for over several thousand seconds after an event occurrence by modeling the heat capacity of the core. (2) Incore and plant dynamics of any plant system can be analyzed by rearranging packages which simulate plant system components one by one. (3) Thermal hydraulics for each component can be analyzed by separating heat transfer calculation for component from fluid flow calculation for helium and pressurized water systems. The validity of the 'ACCORD' code including models for nuclear calculation, heat transfer and fluid flow calculation, control system and safety protection system, was confirmed through cross checks with other available codes. (author)
Subchannel analysis code development for CANDU fuel channel
International Nuclear Information System (INIS)
Park, J. H.; Suk, H. C.; Jun, J. S.; Oh, D. J.; Hwang, D. H.; Yoo, Y. J.
1998-07-01
Since there are several subchannel codes such as COBRA and TORC codes for a PWR fuel channel but not for a CANDU fuel channel in our country, the subchannel analysis code for a CANDU fuel channel was developed for the prediction of flow conditions on the subchannels, for the accurate assessment of the thermal margin, the effect of appendages, and radial/axial power profile of fuel bundles on flow conditions and CHF and so on. In order to develop the subchannel analysis code for a CANDU fuel channel, subchannel analysis methodology and its applicability/pertinence for a fuel channel were reviewed from the CANDU fuel channel point of view. Several thermalhydraulic and numerical models for the subchannel analysis on a CANDU fuel channel were developed. The experimental data of the CANDU fuel channel were collected, analyzed and used for validation of a subchannel analysis code developed in this work. (author). 11 refs., 3 tabs., 50 figs
Dynamic analysis of CHASNUPP steam generator structure during shipping
International Nuclear Information System (INIS)
Han Liangbi; Xu Jinkang; Zhou Meiwu; He Yinbiao
1998-07-01
The dynamic analysis of CHASNUPP steam generator during shipping is described, including the simplified mathematical model, acceleration power spectrum of ocean wave induced random vibration, the dynamic analysis of steam generator structure under random loading, the applied computer code and calculated results
Sending policies in dynamic wireless mesh using network coding
DEFF Research Database (Denmark)
Pandi, Sreekrishna; Fitzek, Frank; Pihl, Jeppe
2015-01-01
This paper demonstrates the quick prototyping capabilities of the Python-Kodo library for network coding based performance evaluation and investigates the problem of data redundancy in a network coded wireless mesh with opportunistic overhearing. By means of several wireless meshed architectures ...
Evaluation of the DRAGON code for VHTR design analysis.
Energy Technology Data Exchange (ETDEWEB)
Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division
2006-01-12
This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR.
Evaluation of the DRAGON code for VHTR design analysis
International Nuclear Information System (INIS)
Taiwo, T. A.; Kim, T. K.; Nuclear Engineering Division
2006-01-01
This letter report summarizes three activities that were undertaken in FY 2005 to gather information on the DRAGON code and to perform limited evaluations of the code performance when used in the analysis of the Very High Temperature Reactor (VHTR) designs. These activities include: (1) Use of the code to model the fuel elements of the helium-cooled and liquid-salt-cooled VHTR designs. Results were compared to those from another deterministic lattice code (WIMS8) and a Monte Carlo code (MCNP). (2) The preliminary assessment of the nuclear data library currently used with the code and libraries that have been provided by the IAEA WIMS-D4 Library Update Project (WLUP). (3) DRAGON workshop held to discuss the code capabilities for modeling the VHTR
Development and Validation of A Nuclear Fuel Cycle Analysis Tool: A FUTURE Code
Energy Technology Data Exchange (ETDEWEB)
Kim, S. K.; Ko, W. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, Yoon Hee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)
2013-10-15
This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy) code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.
DEVELOPMENT AND VALIDATION OF A NUCLEAR FUEL CYCLE ANALYSIS TOOL: A FUTURE CODE
Directory of Open Access Journals (Sweden)
S.K. KIM
2013-10-01
Full Text Available This paper presents the development and validation methods of the FUTURE (FUel cycle analysis Tool for nUcleaR Energy code, which was developed for a dynamic material flow evaluation and economic analysis of the nuclear fuel cycle. This code enables an evaluation of a nuclear material flow and its economy for diverse nuclear fuel cycles based on a predictable scenario. The most notable virtue of this FUTURE code, which was developed using C# and MICROSOFT SQL DBMS, is that a program user can design a nuclear fuel cycle process easily using a standard process on the canvas screen through a drag-and-drop method. From the user's point of view, this code is very easy to use thanks to its high flexibility. In addition, the new code also enables the maintenance of data integrity by constructing a database environment of the results of the nuclear fuel cycle analyses.
Accelerator-driven transmutation reactor analysis code system (ATRAS)
Energy Technology Data Exchange (ETDEWEB)
Sasa, Toshinobu; Tsujimoto, Kazufumi; Takizuka, Takakazu; Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
1999-03-01
JAERI is proceeding a design study of the hybrid type minor actinide transmutation system which mainly consist of an intense proton accelerator and a fast subcritical core. Neutronics and burnup characteristics of the accelerator-driven system is important from a view point of the maintenance of subcriticality and energy balance during the system operation. To determine those characteristics accurately, it is necessary to involve reactions at high-energy region, which are not treated on ordinary reactor analysis codes. The authors developed a code system named ATRAS to analyze the neutronics and burnup characteristics of accelerator-driven subcritical reactor systems. ATRAS has a function of burnup analysis taking account of the effect of spallation neutron source. ATRAS consists of a spallation analysis code, a neutron transport codes and a burnup analysis code. Utility programs for fuel exchange, pre-processing and post-processing are also incorporated. (author)
Development of safety analysis codes for light water reactor
International Nuclear Information System (INIS)
Akimoto, Masayuki
1985-01-01
An overview is presented of currently used major codes for the prediction of thermohydraulic transients in nuclear power plants. The overview centers on the two-phase fluid dynamics of the coolant system and the assessment of the codes. Some of two-phase phenomena such as phase separation are not still predicted with engineering accuracy. MINCS-PIPE are briefly introduced. The MINCS-PIPE code is to assess constitutive relations and to aid development of various experimental correlations for 1V1T model to 2V2T model. (author)
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1983-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
Status of the CONTAIN computer code for LWR containment analysis
International Nuclear Information System (INIS)
Bergeron, K.D.; Murata, K.K.; Rexroth, P.E.; Clauser, M.J.; Senglaub, M.E.; Sciacca, F.W.; Trebilcock, W.
1982-01-01
The current status of the CONTAIN code for LWR safety analysis is reviewed. Three example calculations are discussed as illustrations of the code's capabilities: (1) a demonstration of the spray model in a realistic PWR problem, and a comparison with CONTEMPT results; (2) a comparison of CONTAIN results for a major aerosol experiment against experimental results and predictions of the HAARM aerosol code; and (3) an LWR sample problem, involving a TMLB' sequence for the Zion reactor containment
Development of the integrated system reliability analysis code MODULE
International Nuclear Information System (INIS)
Han, S.H.; Yoo, K.J.; Kim, T.W.
1987-01-01
The major components in a system reliability analysis are the determination of cut sets, importance measure, and uncertainty analysis. Various computer codes have been used for these purposes. For example, SETS and FTAP are used to determine cut sets; Importance for importance calculations; and Sample, CONINT, and MOCUP for uncertainty analysis. There have been problems when the codes run each other and the input and output are not linked, which could result in errors when preparing input for each code. The code MODULE was developed to carry out the above calculations simultaneously without linking input and outputs to other codes. MODULE can also prepare input for SETS for the case of a large fault tree that cannot be handled by MODULE. The flow diagram of the MODULE code is shown. To verify the MODULE code, two examples are selected and the results and computation times are compared with those of SETS, FTAP, CONINT, and MOCUP on both Cyber 170-875 and IBM PC/AT. Two examples are fault trees of the auxiliary feedwater system (AFWS) of Korea Nuclear Units (KNU)-1 and -2, which have 54 gates and 115 events, 39 gates and 92 events, respectively. The MODULE code has the advantage that it can calculate the cut sets, importances, and uncertainties in a single run with little increase in computing time over other codes and that it can be used in personal computers
Development of the next generation reactor analysis code system, MARBLE
International Nuclear Information System (INIS)
Yokoyama, Kenji; Hazama, Taira; Nagaya, Yasunobu; Chiba, Go; Kugo, Teruhiko; Ishikawa, Makoto; Tatsumi, Masahiro; Hirai, Yasushi; Hyoudou, Hideaki; Numata, Kazuyuki; Iwai, Takehiko; Jin, Tomoyuki
2011-03-01
A next generation reactor analysis code system, MARBLE, has been developed. MARBLE is a successor of the fast reactor neutronics analysis code systems, JOINT-FR and SAGEP-FR (conventional systems), which were developed for so-called JUPITER standard analysis methods. MARBLE has the equivalent analysis capability to the conventional system because MARBLE can utilize sub-codes included in the conventional system without any change. On the other hand, burnup analysis functionality for power reactors is improved compared with the conventional system by introducing models on fuel exchange treatment and control rod operation and so on. In addition, MARBLE has newly developed solvers and some new features of burnup calculation by the Krylov sub-space method and nuclear design accuracy evaluation by the extended bias factor method. In the development of MARBLE, the object oriented technology was adopted from the view-point of improvement of the software quality such as flexibility, expansibility, facilitation of the verification by the modularization and assistance of co-development. And, software structure called the two-layer system consisting of scripting language and system development language was applied. As a result, MARBLE is not an independent analysis code system which simply receives input and returns output, but an assembly of components for building an analysis code system (i.e. framework). Furthermore, MARBLE provides some pre-built analysis code systems such as the fast reactor neutronics analysis code system. SCHEME, which corresponds to the conventional code and the fast reactor burnup analysis code system, ORPHEUS. (author)
Bari, Md. S.; Das, T.
2013-09-01
Tectonic framework of Bangladesh and adjoining areas indicate that Bangladesh lies well within an active seismic zone. The after effect of earthquake is more severe in an underdeveloped and a densely populated country like ours than any other developed countries. Bangladesh National Building Code (BNBC) was first established in 1993 to provide guidelines for design and construction of new structure subject to earthquake ground motions in order to minimize the risk to life for all structures. A revision of BNBC 1993 is undergoing to make this up to date with other international building codes. This paper aims at the comparison of various provisions of seismic analysis as given in building codes of different countries. This comparison will give an idea regarding where our country stands when it comes to safety against earth quake. Primarily, various seismic parameters in BNBC 2010 (draft) have been studied and compared with that of BNBC 1993. Later, both 1993 and 2010 edition of BNBC codes have been compared graphically with building codes of other countries such as National Building Code of India 2005 (NBC-India 2005), American Society of Civil Engineering 7-05 (ASCE 7-05). The base shear/weight ratios have been plotted against the height of the building. The investigation in this paper reveals that BNBC 1993 has the least base shear among all the codes. Factored Base shear values of BNBC 2010 are found to have increased significantly than that of BNBC 1993 for low rise buildings (≤20 m) around the country than its predecessor. Despite revision of the code, BNBC 2010 (draft) still suggests less base shear values when compared to the Indian and American code. Therefore, this increase in factor of safety against the earthquake imposed by the proposed BNBC 2010 code by suggesting higher values of base shear is appreciable.
Development of realistic thermal hydraulic system analysis code
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, B. D; Kim, K. D. [and others
2002-05-01
The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others.
Development of realistic thermal hydraulic system analysis code
International Nuclear Information System (INIS)
Lee, Won Jae; Chung, B. D; Kim, K. D.
2002-05-01
The realistic safety analysis system is essential for nuclear safety research, advanced reactor development, safety analysis in nuclear industry and 'in-house' plant design capability development. In this project, we have developed a best-estimate multi-dimensional thermal-hydraulic system code, MARS, which is based on the integrated version of the RELAP5 and COBRA-TF codes. To improve the realistic analysis capability, we have improved the models for multi-dimensional two-phase flow phenomena and for advanced two-phase flow modeling. In addition, the GUI (Graphic User Interface) feature were developed to enhance the user's convenience. To develop the coupled analysis capability, the MARS code were linked with the three-dimensional reactor kinetics code (MASTER), the core thermal analysis code (COBRA-III/CP), and the best-estimate containment analysis code (CONTEMPT), resulting in MARS/MASTER/COBRA/CONTEMPT. Currently, the MARS code system has been distributed to 18 domestic organizations, including research, industrial, regulatory organizations and universities. The MARS has been being widely used for the safety research of existing PWRs, advanced PWR, CANDU and research reactor, the pre-test analysis of TH experiments, and others
A code for structural analysis of fuel assemblies
International Nuclear Information System (INIS)
Hayashi, I.M.V.; Perrotta, J.A.
1988-08-01
It's presented the code ELCOM for the matrix analysis of tubular structures coupled by rigid spacers, typical of PWR's fuel elements. The code ELCOM makes a static structural analysis, where the displacements and internal forces are obtained for each tubular structure at the joints with the spacers, and also, the natural frequencies and vibrational modes of an equilavent integrated structure are obtained. The ELCOM result is compared to a PWR fuel element structural analysis obtained in published paper. (author) [pt
PAPIRUS - a computer code for FBR fuel performance analysis
International Nuclear Information System (INIS)
Kobayashi, Y.; Tsuboi, Y.; Sogame, M.
1991-01-01
The FBR fuel performance analysis code PAPIRUS has been developed to design fuels for demonstration and future commercial reactors. A pellet structural model was developed to describe the generation, depletion and transport of vacancies and atomic elements in unified fashion. PAPIRUS results in comparison with the power - to - melt test data from HEDL showed validity of the code at the initial reactor startup. (author)
76 FR 57982 - Building Energy Codes Cost Analysis
2011-09-19
... DEPARTMENT OF ENERGY Office of Energy Efficiency and Renewable Energy [Docket No. EERE-2011-BT-BC-0046] Building Energy Codes Cost Analysis Correction In notice document 2011-23236 beginning on page...-23236 Filed 9-16-11; 8:45 am] BILLING CODE 1505-01-P ...
An economic analysis code used for PWR fuel cycle
International Nuclear Information System (INIS)
Liu Dingqin
1989-01-01
An economic analysis code used for PWR fuel cycle is developed. This economic code includes 12 subroutines representing vavious processes for entire PWR fuel cycle, and indicates the influence of the fuel cost on the cost of the electricity generation and the influence of individual process on the sensitivity of the fuel cycle cost
Use of computer codes for system reliability analysis
International Nuclear Information System (INIS)
Sabek, M.; Gaafar, M.; Poucet, A.
1989-01-01
This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author)
Use of computer codes for system reliability analysis
Energy Technology Data Exchange (ETDEWEB)
Sabek, M.; Gaafar, M. (Nuclear Regulatory and Safety Centre, Atomic Energy Authority, Cairo (Egypt)); Poucet, A. (Commission of the European Communities, Ispra (Italy). Joint Research Centre)
1989-01-01
This paper gives a summary of studies performed at the JRC, ISPRA on the use of computer codes for complex systems analysis. The computer codes dealt with are: CAFTS-SALP software package, FRACTIC, FTAP, computer code package RALLY, and BOUNDS. Two reference case studies were executed by each code. The probabilistic results obtained, as well as the computation times are compared. The two cases studied are the auxiliary feedwater system of a 1300 MW PWR reactor and the emergency electrical power supply system. (author).
Global Analysis of Nonlinear Dynamics
Luo, Albert
2012-01-01
Global Analysis of Nonlinear Dynamics collects chapters on recent developments in global analysis of non-linear dynamical systems with a particular emphasis on cell mapping methods developed by Professor C.S. Hsu of the University of California, Berkeley. This collection of contributions prepared by a diverse group of internationally recognized researchers is intended to stimulate interests in global analysis of complex and high-dimensional nonlinear dynamical systems, whose global properties are largely unexplored at this time. This book also: Presents recent developments in global analysis of non-linear dynamical systems Provides in-depth considerations and extensions of cell mapping methods Adopts an inclusive style accessible to non-specialists and graduate students Global Analysis of Nonlinear Dynamics is an ideal reference for the community of nonlinear dynamics in different disciplines including engineering, applied mathematics, meteorology, life science, computational science, and medicine.
User's manual for seismic analysis code 'SONATINA-2V'
International Nuclear Information System (INIS)
Hanawa, Satoshi; Iyoku, Tatsuo
2001-08-01
The seismic analysis code, SONATINA-2V, has been developed to analyze the behavior of the HTTR core graphite components under seismic excitation. The SONATINA-2V code is a two-dimensional computer program capable of analyzing the vertical arrangement of the HTTR graphite components, such as fuel blocks, replaceable reflector blocks, permanent reflector blocks, as well as their restraint structures. In the analytical model, each block is treated as rigid body and is restrained by dowel pins which restrict relative horizontal movement but allow vertical and rocking motions between upper and lower blocks. Moreover, the SONATINA-2V code is capable of analyzing the core vibration behavior under both simultaneous excitations of vertical and horizontal directions. The SONATINA-2V code is composed of the main program, pri-processor for making the input data to SONATINA-2V and post-processor for data processing and making the graphics from analytical results. Though the SONATINA-2V code was developed in order to work in the MSP computer system of Japan Atomic Energy Research Institute (JAERI), the computer system was abolished with the technical progress of computer. Therefore, improvement of this analysis code was carried out in order to operate the code under the UNIX machine, SR8000 computer system, of the JAERI. The users manual for seismic analysis code, SONATINA-2V, including pri- and post-processor is given in the present report. (author)
Kim, Daehee; Kim, Dongwan; An, Sunshin
2016-07-09
Code dissemination in wireless sensor networks (WSNs) is a procedure for distributing a new code image over the air in order to update programs. Due to the fact that WSNs are mostly deployed in unattended and hostile environments, secure code dissemination ensuring authenticity and integrity is essential. Recent works on dynamic packet size control in WSNs allow enhancing the energy efficiency of code dissemination by dynamically changing the packet size on the basis of link quality. However, the authentication tokens attached by the base station become useless in the next hop where the packet size can vary according to the link quality of the next hop. In this paper, we propose three source authentication schemes for code dissemination supporting dynamic packet size. Compared to traditional source authentication schemes such as μTESLA and digital signatures, our schemes provide secure source authentication under the environment, where the packet size changes in each hop, with smaller energy consumption.
Kim, Daehee; Kim, Dongwan; An, Sunshin
2016-01-01
Code dissemination in wireless sensor networks (WSNs) is a procedure for distributing a new code image over the air in order to update programs. Due to the fact that WSNs are mostly deployed in unattended and hostile environments, secure code dissemination ensuring authenticity and integrity is essential. Recent works on dynamic packet size control in WSNs allow enhancing the energy efficiency of code dissemination by dynamically changing the packet size on the basis of link quality. However, the authentication tokens attached by the base station become useless in the next hop where the packet size can vary according to the link quality of the next hop. In this paper, we propose three source authentication schemes for code dissemination supporting dynamic packet size. Compared to traditional source authentication schemes such as μTESLA and digital signatures, our schemes provide secure source authentication under the environment, where the packet size changes in each hop, with smaller energy consumption. PMID:27409616
SLSF loop handling system. Volume III. AISC code evaluations and analysis of critical attachments
International Nuclear Information System (INIS)
Ahmed, H.; Cowie, A.; Malek, R.A.; Rafer, A.; Ma, D.; Tebo, F.
1978-10-01
SLSF loop handling system was analyzed for deadweight and postulated dynamic loading conditions using a linear elastic static equivalent method of stress analysis. Stress computations of Cradle and critical attachments per AISC Code guidelines are presented. HFEF is credited with in-depth review of initial phase of work
DYMEL code for prediction of dynamic stability limits in boilers
International Nuclear Information System (INIS)
Deam, R.T.
1980-01-01
Theoretical and experimental studies of Hydrodynamic Instability in boilers were undertaken to resolve the uncertainties of the existing predictive methods at the time the first Advanced Gas Cooled Reactor (AGR) plant was commissioned. The experiments were conducted on a full scale electrical simulation of an AGR boiler and revealed inadequacies in existing methods. As a result a new computer code called DYMEL was developed based on linearisation and Fourier/Laplace Transformation of the one-dimensional boiler equations in both time and space. Beside giving good agreement with local experimental data, the DYMEL code has since shown agreement with stability data from the plant, sodium heated helical tubes, a gas heated helical tube and an electrically heated U-tube. The code is now used widely within the U.K. (author)
Quasi-3d aerodynamic code for analyzing dynamic flap response
DEFF Research Database (Denmark)
Ramos García, Néstor
A computational model for predicting the aerodynamic behavior of wind turbine airfoil profiles subjected to steady and unsteady motions has been developed. The model is based on a viscous-inviscid interaction technique using strong coupling between the viscous and inviscid parts. The inviscid part...... transition model. Validation of the steady two dimensional version of the code has been carried out against experiments for different airfoil geometries and Reynolds numbers. The unsteady version of the code has been benchmarked against experiments for different airfoil geometries at various reduced...... frequencies and oscillation amplitudes, and generally a good agreement is obtained. The capability of the code to simulate a trailing edge flap under steady or unsteady flow conditions has been proven. A parametric study on rotational effects induced by Coriolis and centrifugal forces in the boundary layer...
Production of analysis code for 'JOYO' dosimetry experiment
International Nuclear Information System (INIS)
Sasaki, Makoto; Nakazawa, Masaharu.
1981-01-01
As part of the measurement and analysis plan for the Dosimetry Experiment at the ''JOYO'' experimental fast reactor, neutron flux spectra analysis is performed using the NEUPAC (Neutron Unfolding Code Package) computer program. The code calculates the neutron flux spectra and other integral quantities from the activation data of the dosimeter foils. The NEUPAC code is based on the J1-type unfolding method, and the estimated neutron flux spectra is obtained as its solution. The program is able to determine the integral quantities and their sensitivities, together with an error estimate of the unfolded spectra and integral quantities. The code also performs a chi-square test of the input/output data, and contains many options for the calculational routines. This report presents the analytic theory, the program algorithms, and a description of the functions and use of the NEUPAC code. (author)
Thermohydraulic analysis of nuclear power plant accidents by computer codes
International Nuclear Information System (INIS)
Petelin, S.; Stritar, A.; Istenic, R.; Gregoric, M.; Jerele, A.; Mavko, B.
1982-01-01
RELAP4/MOD6, BRUCH-D-06, CONTEMPT-LT-28, RELAP5/MOD1 and COBRA-4-1 codes were successful y implemented at the CYBER 172 computer in Ljubljana. Input models of NPP Krsko for the first three codes were prepared. Because of the high computer cost only one analysis of double ended guillotine break of the cold leg of NPP Krsko by RELAP4 code has been done. BRUCH code is easier and cheaper for use. Several analysis have been done. Sensitivity study was performed with CONTEMPT-LT-28 for double ended pump suction break. These codes are intended to be used as a basis for independent safety analyses. (author)
Sensitivity analysis of the RESRAD, a dose assessment code
International Nuclear Information System (INIS)
Yu, C.; Cheng, J.J.; Zielen, A.J.
1991-01-01
The RESRAD code is a pathway analysis code that is designed to calculate radiation doses and derive soil cleanup criteria for the US Department of Energy's environmental restoration and waste management program. the RESRAD code uses various pathway and consumption-rate parameters such as soil properties and food ingestion rates in performing such calculations and derivations. As with any predictive model, the accuracy of the predictions depends on the accuracy of the input parameters. This paper summarizes the results of a sensitivity analysis of RESRAD input parameters. Three methods were used to perform the sensitivity analysis: (1) Gradient Enhanced Software System (GRESS) sensitivity analysis software package developed at oak Ridge National Laboratory; (2) direct perturbation of input parameters; and (3) built-in graphic package that shows parameter sensitivities while the RESRAD code is operational
PACC information management code for common cause failures analysis
International Nuclear Information System (INIS)
Ortega Prieto, P.; Garcia Gay, J.; Mira McWilliams, J.
1987-01-01
The purpose of this paper is to present the PACC code, which, through an adequate data management, makes the task of computerized common-mode failure analysis easier. PACC processes and generates information in order to carry out the corresponding qualitative analysis, by means of the boolean technique of transformation of variables, and the quantitative analysis either using one of several parametric methods or a direct data-base. As far as the qualitative analysis is concerned, the code creates several functional forms for the transformation equations according to the user's choice. These equations are subsequently processed by boolean manipulation codes, such as SETS. The quantitative calculations of the code can be carried out in two different ways: either starting from a common cause data-base, or through parametric methods, such as the Binomial Failure Rate Method, the Basic Parameters Method or the Multiple Greek Letter Method, among others. (orig.)
Three-dimensional reactor dynamics code for VVER type nuclear reactors
Energy Technology Data Exchange (ETDEWEB)
Kyrki-Rajamaeki, R. [VTT Energy, Espoo (Finland)
1995-10-01
A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.).
Three-dimensional reactor dynamics code for VVER type nuclear reactors
International Nuclear Information System (INIS)
Kyrki-Rajamaeki, R.
1995-10-01
A three-dimensional reactor dynamics computer code has been developed, validated and applied for transient and accident analyses of VVER type nuclear reactors. This code, HEXTRAN, is a part of the reactor physics and dynamics calculation system of the Technical Research Centre of Finland, VTT. HEXTRAN models accurately the VVER core with hexagonal fuel assemblies. The code uses advanced mathematical methods in spatial and time discretization of neutronics, heat transfer and the two-phase flow equations of hydraulics. It includes all the experience of VTT from 20 years on the accurate three-dimensional static reactor physics as well as on the one-dimensional reactor dynamics. The dynamic coupling with the thermal hydraulic system code SMABRE also allows the VVER circuit-modelling experience to be included in the analyses. (79 refs.)
CoCoNuT: General relativistic hydrodynamics code with dynamical space-time evolution
Dimmelmeier, Harald; Novak, Jérôme; Cerdá-Durán, Pablo
2012-02-01
CoCoNuT is a general relativistic hydrodynamics code with dynamical space-time evolution. The main aim of this numerical code is the study of several astrophysical scenarios in which general relativity can play an important role, namely the collapse of rapidly rotating stellar cores and the evolution of isolated neutron stars. The code has two flavors: CoCoA, the axisymmetric (2D) magnetized version, and CoCoNuT, the 3D non-magnetized version.
Adaptable Value-Set Analysis for Low-Level Code
Brauer, Jörg; Hansen, René Rydhof; Kowalewski, Stefan; Larsen, Kim G.; Olesen, Mads Chr.
2012-01-01
This paper presents a framework for binary code analysis that uses only SAT-based algorithms. Within the framework, incremental SAT solving is used to perform a form of weakly relational value-set analysis in a novel way, connecting the expressiveness of the value sets to computational complexity. Another key feature of our framework is that it translates the semantics of binary code into an intermediate representation. This allows for a straightforward translation of the program semantics in...
Improvement of QR Code Recognition Based on Pillbox Filter Analysis
Directory of Open Access Journals (Sweden)
Jia-Shing Sheu
2013-04-01
Full Text Available The objective of this paper is to perform the innovation design for improving the recognition of a captured QR code image with blur through the Pillbox filter analysis. QR code images can be captured by digital video cameras. Many factors contribute to QR code decoding failure, such as the low quality of the image. Focus is an important factor that affects the quality of the image. This study discusses the out-of-focus QR code image and aims to improve the recognition of the contents in the QR code image. Many studies have used the pillbox filter (circular averaging filter method to simulate an out-of-focus image. This method is also used in this investigation to improve the recognition of a captured QR code image. A blurred QR code image is separated into nine levels. In the experiment, four different quantitative approaches are used to reconstruct and decode an out-of-focus QR code image. These nine reconstructed QR code images using methods are then compared. The final experimental results indicate improvements in identification.
Establishment of computer code system for nuclear reactor design - analysis
International Nuclear Information System (INIS)
Subki, I.R.; Santoso, B.; Syaukat, A.; Lee, S.M.
1996-01-01
Establishment of computer code system for nuclear reactor design analysis is given in this paper. This establishment is an effort to provide the capability in running various codes from nuclear data to reactor design and promote the capability for nuclear reactor design analysis particularly from neutronics and safety points. This establishment is also an effort to enhance the coordination of nuclear codes application and development existing in various research centre in Indonesia. Very prospective results have been obtained with the help of IAEA technical assistance. (author). 6 refs, 1 fig., 1 tab
Developments of fuel performance analysis codes in KEPCO NF
International Nuclear Information System (INIS)
Han, H. T.; Choi, J. M.; Jung, C. D.; Yoo, J. S.
2012-01-01
The KEPCO NF has developed fuel performance analysis and design code named as ROPER, and utility codes of XGCOL and XDNB in order to perform fuel rod design evaluation for Korean nuclear power plants. The ROPER code intends to cover full range of fuel performance evaluation. The XGCOL code is for the clad flattening evaluation and the XDNB code is for the extensive DNB propagation evaluation. In addition to these, the KEPCO NF is now in the developing stage for 3-dimensional fuel performance analysis code, named as OPER3D, using 3-dimensional FEM for the nest generation within the joint project CANDU ENERGY in order to analyze PCMI behavior and fuel performance under load following operation. Of these, the ROPER code is now in the stage of licensing activities by Korean regulatory body and the other two are almost in the final developing stage. After finishing the developing, licensing activities are to be performed. These activities are intending to acquire competitiveness, originality, vendor-free ownership of fuel performance codes in the KEPCO NF
A compendium of computer codes in fault tree analysis
International Nuclear Information System (INIS)
Lydell, B.
1981-03-01
In the past ten years principles and methods for a unified system reliability and safety analysis have been developed. Fault tree techniques serve as a central feature of unified system analysis, and there exists a specific discipline within system reliability concerned with the theoretical aspects of fault tree evaluation. Ever since the fault tree concept was established, computer codes have been developed for qualitative and quantitative analyses. In particular the presentation of the kinetic tree theory and the PREP-KITT code package has influenced the present use of fault trees and the development of new computer codes. This report is a compilation of some of the better known fault tree codes in use in system reliability. Numerous codes are available and new codes are continuously being developed. The report is designed to address the specific characteristics of each code listed. A review of the theoretical aspects of fault tree evaluation is presented in an introductory chapter, the purpose of which is to give a framework for the validity of the different codes. (Auth.)
Development of a nuclear power plant system analysis code
International Nuclear Information System (INIS)
Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.
1997-07-01
During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs
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
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.
Automated uncertainty analysis methods in the FRAP computer codes
International Nuclear Information System (INIS)
Peck, S.O.
1980-01-01
A user oriented, automated uncertainty analysis capability has been incorporated in the Fuel Rod Analysis Program (FRAP) computer codes. The FRAP codes have been developed for the analysis of Light Water Reactor fuel rod behavior during steady state (FRAPCON) and transient (FRAP-T) conditions as part of the United States Nuclear Regulatory Commission's Water Reactor Safety Research Program. The objective of uncertainty analysis of these codes is to obtain estimates of the uncertainty in computed outputs of the codes is to obtain estimates of the uncertainty in computed outputs of the codes as a function of known uncertainties in input variables. This paper presents the methods used to generate an uncertainty analysis of a large computer code, discusses the assumptions that are made, and shows techniques for testing them. An uncertainty analysis of FRAP-T calculated fuel rod behavior during a hypothetical loss-of-coolant transient is presented as an example and carried through the discussion to illustrate the various concepts
International Nuclear Information System (INIS)
Donea, J.; Fasoli-Stella, P.; Giuliani, S.; Halleux, J.P.; Jones, A.V.
1980-01-01
This report describes the governing equations and the finite element modelling used in the computer code EURDYN - 1 M. The code is a non-linear transient dynamic program for the analysis of coupled fluid-structure systems; It is designed for safety studies on LMFBR components (primary containment and fuel subassemblies)
KEWPIE2: A cascade code for the study of dynamical decay of excited nuclei
Lü, Hongliang; Marchix, Anthony; Abe, Yasuhisa; Boilley, David
2016-03-01
KEWPIE-a cascade code devoted to investigating the dynamical decay of excited nuclei, specially designed for treating very low probability events related to the synthesis of super-heavy nuclei formed in fusion-evaporation reactions-has been improved and rewritten in C++ programming language to become KEWPIE2. The current version of the code comprises various nuclear models concerning the light-particle emission, fission process and statistical properties of excited nuclei. General features of the code, such as the numerical scheme and the main physical ingredients, are described in detail. Some typical calculations having been performed in the present paper clearly show that theoretical predictions are generally in accordance with experimental data. Furthermore, since the values of some input parameters cannot be determined neither theoretically nor experimentally, a sensibility analysis is presented. To this end, we systematically investigate the effects of using different parameter values and reaction models on the final results. As expected, in the case of heavy nuclei, the fission process has the most crucial role to play in theoretical predictions. This work would be essential for numerical modeling of fusion-evaporation reactions.
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
Energy Technology Data Exchange (ETDEWEB)
Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E. [Sandia National Labs., Albuquerque, NM (United States); Tills, J. [J. Tills and Associates, Inc., Sandia Park, NM (United States)
1997-12-01
The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
International Nuclear Information System (INIS)
Murata, K.K.; Williams, D.C.; Griffith, R.O.; Gido, R.G.; Tadios, E.L.; Davis, F.J.; Martinez, G.M.; Washington, K.E.; Tills, J.
1997-12-01
The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of the input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions
Development of three dimensional transient analysis code STTA for SCWR core
International Nuclear Information System (INIS)
Wang, Lianjie; Zhao, Wenbo; Chen, Bingde; Yao, Dong; Yang, Ping
2015-01-01
Highlights: • A coupled three dimensional neutronics/thermal-hydraulics code STTA is developed for SCWR core transient analysis. • The Dynamic Link Libraries method is adopted for coupling computation for SCWR multi-flow core transient analysis. • The NEACRP-L-335 PWR benchmark problems are studied to verify STTA. • The SCWR rod ejection problems are studied to verify STTA. • STTA meets what is expected from a code for SCWR core 3-D transient preliminary analysis. - Abstract: A coupled three dimensional neutronics/thermal-hydraulics code STTA (SCWR Three dimensional Transient Analysis code) is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition (NGFMN-K) is used for solving transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN-K through the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with reference solutions and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation
Multi-dimensional Code Development for Safety Analysis of LMR
International Nuclear Information System (INIS)
Ha, K. S.; Jeong, H. Y.; Kwon, Y. M.; Lee, Y. B.
2006-08-01
A liquid metal reactor loaded a metallic fuel has the inherent safety mechanism due to the several negative reactivity feedback. Although this feature demonstrated through experiments in the EBR-II, any of the computer programs until now did not exactly analyze it because of the complexity of the reactivity feedback mechanism. A multi-dimensional detail program was developed through the International Nuclear Energy Research Initiative(INERI) from 2003 to 2005. This report includes the numerical coupling the multi-dimensional program and SSC-K code which is used to the safety analysis of liquid metal reactors in KAERI. The coupled code has been proved by comparing the analysis results using the code with the results using SAS-SASSYS code of ANL for the UTOP, ULOF, and ULOHS applied to the safety analysis for KALIMER-150
Guide to Using Onionskin Analysis Code (U)
Energy Technology Data Exchange (ETDEWEB)
Fugate, Michael Lynn [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group; Morzinski, Jerome Arthur [Los Alamos National Lab. (LANL), Los Alamos, NM (United States). Statistical Sciences Group
2016-09-15
This document is a guide to using R-code written for the purpose of analyzing onionskin experiments. We expect the user to be very familiar with statistical methods and the R programming language. For more details about onionskin experiments and the statistical methods mentioned in this document see Storlie, Fugate, et al. (2013). Engineers at LANL experiment with detonators and high explosives to assess performance. The experimental unit, called an onionskin, is a hemisphere consisting of a detonator and a booster pellet surrounded by explosive material. When the detonator explodes, a streak camera mounted above the pole of the hemisphere records when the shock wave arrives at the surface. The output from the camera is a two-dimensional image that is transformed into a curve that shows the arrival time as a function of polar angle. The statistical challenge is to characterize a baseline population of arrival time curves and to compare the baseline curves to curves from a new, so-called, test series. The hope is that the new test series of curves is statistically similar to the baseline population.
Operational Dynamic Configuration Analysis
Lai, Chok Fung; Zelinski, Shannon
2010-01-01
Sectors may combine or split within areas of specialization in response to changing traffic patterns. This method of managing capacity and controller workload could be made more flexible by dynamically modifying sector boundaries. Much work has been done on methods for dynamically creating new sector boundaries [1-5]. Many assessments of dynamic configuration methods assume the current day baseline configuration remains fixed [6-7]. A challenging question is how to select a dynamic configuration baseline to assess potential benefits of proposed dynamic configuration concepts. Bloem used operational sector reconfigurations as a baseline [8]. The main difficulty is that operational reconfiguration data is noisy. Reconfigurations often occur frequently to accommodate staff training or breaks, or to complete a more complicated reconfiguration through a rapid sequence of simpler reconfigurations. Gupta quantified a few aspects of airspace boundary changes from this data [9]. Most of these metrics are unique to sector combining operations and not applicable to more flexible dynamic configuration concepts. To better understand what sort of reconfigurations are acceptable or beneficial, more configuration change metrics should be developed and their distribution in current practice should be computed. This paper proposes a method to select a simple sequence of configurations among operational configurations to serve as a dynamic configuration baseline for future dynamic configuration concept assessments. New configuration change metrics are applied to the operational data to establish current day thresholds for these metrics. These thresholds are then corroborated, refined, or dismissed based on airspace practitioner feedback. The dynamic configuration baseline selection method uses a k-means clustering algorithm to select the sequence of configurations and trigger times from a given day of operational sector combination data. The clustering algorithm selects a simplified
Impact testing and analysis for structural code benchmarking
International Nuclear Information System (INIS)
Glass, R.E.
1989-01-01
Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes (''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Cask,'' R.E. Glass, Sandia National Laboratories, 1985; ''Sample Problem Manual for Benchmarking of Cask Analysis Codes,'' R.E. Glass, Sandia National Laboratories, 1988; ''Standard Thermal Problem Set for the Evaluation of Heat Transfer Codes Used in the Assessment of Transportation Packages, R.E. Glass, et al., Sandia National Laboratories, 1988) used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in ''Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks,'' R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem. 6 refs., 5 figs
Experimental transport analysis code system in JT-60
International Nuclear Information System (INIS)
Hirayama, Toshio; Shimizu, Katsuhiro; Tani, Keiji; Shirai, Hiroshi; Kikuchi, Mitsuru
1988-03-01
Transport analysis codes have been developed in order to study confinement properties related to particle and energy balance in ohmically and neutral beam heated plasmas of JT-60. The analysis procedure is divided into three steps as follows: 1) LOOK ; The shape of the plasma boundary is identified with a fast boundary identification code of FBI by using magnetic data, and flux surfaces are calculated with a MHD equilibrium code of SELENE. The diagnostic data are mapped to flux surfaces for neutral beam heating calculation and/or for radial transport analysis. 2) OFMC ; On the basis of transformed data, an orbit following Monte Carlo code of OFMC calculates both profiles of power deposition and particle source of neutral beam injected into a plasma. 3) SCOOP ; In the last stage, a one dimensional transport code of SCOOP solves particle and energy balance for electron and ion, in order to evaluate transport coefficients as well as global parameters such as energy confinement time and the stored energy. The analysis results are provided to a data bank of DARTS that is used to find an overview of important consideration on confinement with a regression analysis code of RAC. (author)
Color Coding of Circuit Quantities in Introductory Circuit Analysis Instruction
Reisslein, Jana; Johnson, Amy M.; Reisslein, Martin
2015-01-01
Learning the analysis of electrical circuits represented by circuit diagrams is often challenging for novice students. An open research question in electrical circuit analysis instruction is whether color coding of the mathematical symbols (variables) that denote electrical quantities can improve circuit analysis learning. The present study…
Energy Technology Data Exchange (ETDEWEB)
Jung, B. D.; Jung, J. J.; Ha, K. S.; Hwang, M. K.; Lee, Y. S.; Lee, W. J. [KAERI, Taejon (Korea, Republic of)
1999-10-01
The readability of MARS code has been enhanced greatly by replacing the bit-packed word with several logical words and integer words and recoding the related subroutines, which have the complicated bit operations and packed words. Functional improvements of code has been achieved through the multiple uses of dynamic link libraries(DLL) for containment analysis module CONTEMPT4 and multidimensional kinetics analysis module MASTER. The establishment of integrated analysis system, MARS/CONTEMPT/MASTER, was validated through the verification calculation for a postulated problem. MARS user-friendly features are also improved by displaying the 2D contour map of 3 D module data on-line. In addition to the on-line-graphics, the MARS windows menus were upgraded to include the on-line-manual, pre-view of input and output, and link to MARS web site. As a result, the readability, applicability, and user-friendly features of MARS code has been greatly enhanced.
International Nuclear Information System (INIS)
Jung, B. D.; Jung, J. J.; Ha, K. S.; Hwang, M. K.; Lee, Y. S.; Lee, W. J.
1999-01-01
The readability of MARS code has been enhanced greatly by replacing the bit-packed word with several logical words and integer words and recoding the related subroutines, which have the complicated bit operations and packed words. Functional improvements of code has been achieved through the multiple uses of dynamic link libraries(DLL) for containment analysis module CONTEMPT4 and multidimensional kinetics analysis module MASTER. The establishment of integrated analysis system, MARS/CONTEMPT/MASTER, was validated through the verification calculation for a postulated problem. MARS user-friendly features are also improved by displaying the 2D contour map of 3 D module data on-line. In addition to the on-line-graphics, the MARS windows menus were upgraded to include the on-line-manual, pre-view of input and output, and link to MARS web site. As a result, the readability, applicability, and user-friendly features of MARS code has been greatly enhanced
Multi-pressure boiler thermodynamics analysis code
International Nuclear Information System (INIS)
Lorenzoni, G.
1992-01-01
A new method and the relative FORTRAN program for the thermodynamics design analysis of a multipressure boiler are reported. This method permits the thermodynamics design optimization with regard to total exergy production and a preliminary costs
Flight Dynamics Analysis Branch 2005 Technical Highlights
2005-01-01
This report summarizes the major activities and accomplishments carried out by the Flight Dynamics Analysis Branch (FDAB), Code 595, in support of flight projects and technology development initiatives in Fiscal Year (FY) 2005. The report is intended to serve as a summary of the type of support carried out by the FDAB, as well as a concise reference of key accomplishments and mission experience derived from the various mission support roles. The primary focus of the FDAB is to provide expertise in the disciplines of flight dynamics including spacecraft navigation (autonomous and ground based); spacecraft trajectory design and maneuver planning; attitude analysis; attitude determination and sensor calibration; and attitude control subsystem (ACS) analysis and design. The FDAB currently provides support for missions and technology development projects involving NASA, other government agencies, academia, and private industry.
Analysis of Iterated Hard Decision Decoding of Product Codes with Reed-Solomon Component Codes
DEFF Research Database (Denmark)
Justesen, Jørn; Høholdt, Tom
2007-01-01
Products of Reed-Solomon codes are important in applications because they offer a combination of large blocks, low decoding complexity, and good performance. A recent result on random graphs can be used to show that with high probability a large number of errors can be corrected by iterating...... minimum distance decoding. We present an analysis related to density evolution which gives the exact asymptotic value of the decoding threshold and also provides a closed form approximation to the distribution of errors in each step of the decoding of finite length codes....
Object-Oriented Parallel Particle-in-Cell Code for Beam Dynamics Simulation in Linear Accelerators
International Nuclear Information System (INIS)
Qiang, J.; Ryne, R.D.; Habib, S.; Decky, V.
1999-01-01
In this paper, we present an object-oriented three-dimensional parallel particle-in-cell code for beam dynamics simulation in linear accelerators. A two-dimensional parallel domain decomposition approach is employed within a message passing programming paradigm along with a dynamic load balancing. Implementing object-oriented software design provides the code with better maintainability, reusability, and extensibility compared with conventional structure based code. This also helps to encapsulate the details of communications syntax. Performance tests on SGI/Cray T3E-900 and SGI Origin 2000 machines show good scalability of the object-oriented code. Some important features of this code also include employing symplectic integration with linear maps of external focusing elements and using z as the independent variable, typical in accelerators. A successful application was done to simulate beam transport through three superconducting sections in the APT linac design
Compendium of computer codes for the safety analysis of LMFBR's
International Nuclear Information System (INIS)
1975-06-01
A high level of mathematical sophistication is required in the safety analysis of LMFBR's to adequately meet the demands for realism and confidence in all areas of accident consequence evaluation. The numerical solution procedures associated with these analyses are generally so complex and time consuming as to necessitate their programming into computer codes. These computer codes have become extremely powerful tools for safety analysis, combining unique advantages in accuracy, speed and cost. The number, diversity and complexity of LMFBR safety codes in the U. S. has grown rapidly in recent years. It is estimated that over 100 such codes exist in various stages of development throughout the country. It is inevitable that such a large assortment of codes will require rigorous cataloguing and abstracting to aid individuals in identifying what is available. It is the purpose of this compendium to provide such a service through the compilation of code summaries which describe and clarify the status of domestic LMFBR safety codes. (U.S.)
Development and validation of sodium fire analysis code ASSCOPS
International Nuclear Information System (INIS)
Ohno, Shuji
2001-01-01
A version 2.1 of the ASSCOPS sodium fire analysis code was developed to evaluate the thermal consequences of a sodium leak and consequent fire in LMFBRs. This report describes the computational models and the validation studies using the code. The ASSCOPS calculates sodium droplet and pool fire, and consequential heat/mass transfer behavior. Analyses of sodium pool or spray fire experiments confirmed that this code and parameters used in the validation studies gave valid results on the thermal consequences of sodium leaks and fires. (author)
A static analysis tool set for assembler code verification
International Nuclear Information System (INIS)
Dhodapkar, S.D.; Bhattacharjee, A.K.; Sen, Gopa
1991-01-01
Software Verification and Validation (V and V) is an important step in assuring reliability and quality of the software. The verification of program source code forms an important part of the overall V and V activity. The static analysis tools described here are useful in verification of assembler code. The tool set consists of static analysers for Intel 8086 and Motorola 68000 assembly language programs. The analysers examine the program source code and generate information about control flow within the program modules, unreachable code, well-formation of modules, call dependency between modules etc. The analysis of loops detects unstructured loops and syntactically infinite loops. Software metrics relating to size and structural complexity are also computed. This report describes the salient features of the design, implementation and the user interface of the tool set. The outputs generated by the analyser are explained using examples taken from some projects analysed by this tool set. (author). 7 refs., 17 figs
Fault tree analysis. Implementation of the WAM-codes
International Nuclear Information System (INIS)
Bento, J.P.; Poern, K.
1979-07-01
The report describes work going on at Studsvik at the implementation of the WAM code package for fault tree analysis. These codes originally developed under EPRI contract by Sciences Applications Inc, allow, in contrast with other fault tree codes, all Boolean operations, thus allowing modeling of ''NOT'' conditions and dependent components. To concretize the implementation of these codes, the auxiliary feed-water system of the Swedish BWR Oskarshamn 2 was chosen for the reliability analysis. For this system, both the mean unavailability and the probability density function of the top event - undesired event - of the system fault tree were calculated, the latter using a Monte-Carlo simulation technique. The present study is the first part of a work performed under contract with the Swedish Nuclear Power Inspectorate. (author)
Code conversion for system design and safety analysis of NSSS
Energy Technology Data Exchange (ETDEWEB)
Lee, Hae Cho; Kim, Young Tae; Choi, Young Gil; Kim, Hee Kyung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1996-01-01
This report describes overall project works related to conversion, installation and validation of computer codes which are used in NSSS design and safety analysis of nuclear power plants. Domain/os computer codes for system safety analysis are installed and validated on Apollo DN10000, and then Apollo version are converted and installed again on HP9000/700 series with appropriate validation. Also, COOLII and COAST which are cyber version computer codes are converted into versions of Apollo DN10000 and HP9000/700, and installed with validation. This report details whole processes of work involved in the computer code conversion and installation, as well as software verification and validation results which are attached to this report. 12 refs., 8 figs. (author)
Parallelization of a beam dynamics code and first large scale radio frequency quadrupole simulations
Directory of Open Access Journals (Sweden)
J. Xu
2007-01-01
Full Text Available The design and operation support of hadron (proton and heavy-ion linear accelerators require substantial use of beam dynamics simulation tools. The beam dynamics code TRACK has been originally developed at Argonne National Laboratory (ANL to fulfill the special requirements of the rare isotope accelerator (RIA accelerator systems. From the beginning, the code has been developed to make it useful in the three stages of a linear accelerator project, namely, the design, commissioning, and operation of the machine. To realize this concept, the code has unique features such as end-to-end simulations from the ion source to the final beam destination and automatic procedures for tuning of a multiple charge state heavy-ion beam. The TRACK code has become a general beam dynamics code for hadron linacs and has found wide applications worldwide. Until recently, the code has remained serial except for a simple parallelization used for the simulation of multiple seeds to study the machine errors. To speed up computation, the TRACK Poisson solver has been parallelized. This paper discusses different parallel models for solving the Poisson equation with the primary goal to extend the scalability of the code onto 1024 and more processors of the new generation of supercomputers known as BlueGene (BG/L. Domain decomposition techniques have been adapted and incorporated into the parallel version of the TRACK code. To demonstrate the new capabilities of the parallelized TRACK code, the dynamics of a 45 mA proton beam represented by 10^{8} particles has been simulated through the 325 MHz radio frequency quadrupole and initial accelerator section of the proposed FNAL proton driver. The results show the benefits and advantages of large-scale parallel computing in beam dynamics simulations.
Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code
Hendricks, Eric S.
2016-01-01
The prediction of turbomachinery performance characteristics is an important part of the conceptual aircraft engine design process. During this phase, the designer must examine the effects of a large number of turbomachinery design parameters to determine their impact on overall engine performance and weight. The lack of detailed design information available in this phase necessitates the use of simpler meanline and streamline methods to determine the turbomachinery geometry characteristics and provide performance estimates prior to more detailed CFD (Computational Fluid Dynamics) analyses. While a number of analysis codes have been developed for this purpose, most are written in outdated software languages and may be difficult or impossible to apply to new, unconventional designs. The Object-Oriented Turbomachinery Analysis Code (OTAC) is currently being developed at NASA Glenn Research Center to provide a flexible meanline and streamline analysis capability in a modern object-oriented language. During the development and validation of OTAC, a limitation was identified in the code's ability to analyze and converge turbines as the flow approached choking. This paper describes a series of changes which can be made to typical OTAC turbine meanline models to enable the assessment of choked flow up to limit load conditions. Results produced with this revised model setup are provided in the form of turbine performance maps and are compared to published maps.
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
Flow analysis of tubular fuel assembly using CFD code
International Nuclear Information System (INIS)
Park, J. H.; Park, C.; Chae, H. T.
2004-01-01
Based on the experiences of HANARO, a new research reactor is under conceptual design preparing for future needs of research reactor. Considering various aspects such as nuclear physics, thermal-hydraulics, mechanical structure and the applicability of HANARO technology, a tubular type fuel has been considered as that of a new research reactor. Tubular type fuel has several circular fuel layers, and each layer consists of 3 curved fuel plates arranged with constant small gap to build up cooling channels. In the thermal-hydraulic point, it is very important to maintain each channel flow velocity be equal as much as possible, because the small gaps between curved thin fuel plates independently forms separate coolant channels, which may cause a thermal-hydraulic problem in certain conditions. In this study, commercial CFD(Computational Fluid Dynamics) code, Fluent, has been used to investigate flow characteristics of tubular type fuel assembly. According to the computation results for the preliminary conceptual design, there is a serious lack of uniformity of average velocity on the each coolant channel. Some changes for initial conceptual design were done to improve the balance of velocity distribution, and analysis was done again, too. The results for the revised design showed that the uniformity of each channel velocity was improved significantly. The influence of outermost channel gap width on the velocity distribution was also examined
Implementation of reactor safety analysis code CATHARE and its use on FACOM M-380
International Nuclear Information System (INIS)
Ishiguro, Misako; Shinozawa, Naohisa; Tomiyama, Mineyoshi; Fujisaki, Masahide
1986-05-01
CATHARE is an advanced safety analysis code developed at the Nuclear Research Center of Grenoble in France. The code simulates thermohydraulic phenomena involved in loss of coolant accidents in pressurized water reactors. The code has been introduced into JAERI as a part of the technical exchange between the JAERI ROSA-IV Program and the French BETHSY-CATHARE Program. The code was delivered in the form of 23 files containing 115,000 statements in total. A large part of CATHARE code has been written in an extended Fortran language 'Esope' which is mainly used for managing dynamic memory allocation. The JAERI version is created from the IBM version which has been used on Amdhal computer at ISPRA. Some modifications are required in order to implement the CATHARE code at JAERI because of difference in softwares. In this report, the overview of the code structure, the JAERI usage, the implementation method, the error correction method, the problems special to install the code in JAERI, and the distribution of computing time are described. (author)
Dynamic ensemble coding of saccades in the monkey superior colliculus.
Goossens, H.H.L.M.; Opstal, A.J. van
2006-01-01
The deeper layers of the midbrain superior colliculus (SC) contain a topographic motor map in which a localized population of cells is recruited for each saccade, but how the brain stem decodes the dynamic SC output is unclear. Here we analyze saccade-related responses in the monkey SC to test a new
Development and improvement of safety analysis code for geological disposal
Energy Technology Data Exchange (ETDEWEB)
NONE
2013-08-15
In order to confirm the long-term safety concerning geological disposal, probabilistic safety assessment code and other analysis codes, which can evaluate possibility of each event and influence on engineered barrier and natural barrier by the event, were introduced. We confirmed basic functions of those codes and studied the relation between those functions and FEP/PID which should be taken into consideration in safety assessment. We are planning to develop 'Nuclide Migration Assessment System' for the purpose of realizing improvement in efficiency of assessment work, human error prevention for analysis, and quality assurance of the analysis environment and analysis work for safety assessment by using it. As the first step, we defined the system requirements and decided the system composition and functions which should be mounted in them based on those requirements. (author)
International Nuclear Information System (INIS)
1977-01-01
The TRICO part of the CEA-SEMT system is concerned with the elasticity or plasticity computation of structures made of thin shells and beams. TRICO uses the finite element method for shells and beams. TRICO also allows the dynamic computing of structures: search for eigenmodes and eigenfrequencies or response to any sinusoidal excitation, response to time dependent loads (direct integration) in elasticity or plasticity. The mechanical structures can offer any shape and be composed of a number of materials. A special effort has been put on data input (read without any format), the data being arranged in optional commands with a precise physical sense corresponding to an order for the program. A dynamic control of the memory allows the size of the program to be adapted to that the problem to be processed. Results are printed on listing, or many be described on a magnetic tape [fr
Two-dimensional disruption thermal analysis code DREAM
International Nuclear Information System (INIS)
Yamazaki, Seiichiro; Kobayashi, Takeshi; Seki, Masahiro.
1988-08-01
When a plasma disruption takes place in a tokamak type fusion reactor, plasma facing components such as first wall and divertor/limiter are subjected to an intense heat load with very high heat flux and short duration. At the surface of the wall, temperature rapidly rises, and melting and evaporation occurs, it causes reduction of wall thickness and crack initiation/propagation. As lifetime of the components is significantly affected by them, the transient analysis in consideration of phase changes (melting/evaporation) and radiation heat loss is required in the design of these components. This paper describes the computer code DREAM developed to perform the two-dimensional transient thermal analysis that takes phase changes and radiation into account. The input and output of the code and a sample analysis on a disruption simulation experiment are also reported. The user's input manual is added as an appendix. The profiles and time variations of temperature, and melting and evaporated thicknesses of the material subjected to intense heat load can be obtained, using this computer code. This code also gives the temperature data for elastoplastic analysis with FEM structural analysis codes (ADINA, MARC, etc.) to evaluate the thermal stress and crack propagation behavior within the wall materials. (author)
Development of the versatile reactor analysis code system, MARBLE2
International Nuclear Information System (INIS)
Yokoyama, Kenji; Jin, Tomoyuki; Hazama, Taira; Hirai, Yasushi
2015-07-01
The second version of the versatile reactor analysis code system, MARBLE2, has been developed. A lot of new functions have been added in MARBLE2 by using the base technology developed in the first version (MARBLE1). Introducing the remaining functions of the conventional code system (JOINT-FR and SAGEP-FR), MARBLE2 enables one to execute almost all analysis functions of the conventional code system with the unified user interfaces of its subsystem, SCHEME. In particular, the sensitivity analysis functionality is available in MARBLE2. On the other hand, new built-in solvers have been developed, and existing ones have been upgraded. Furthermore, some other analysis codes and libraries developed in JAEA have been consolidated and prepared in SCHEME. In addition, several analysis codes developed in the other institutes have been additionally introduced as plug-in solvers. Consequently, gamma-ray transport calculation and heating evaluation become available. As for another subsystem, ORPHEUS, various functionality updates and speed-up techniques have been applied based on user experience of MARBLE1 to enhance its usability. (author)
International Nuclear Information System (INIS)
Zeuch, W.R.; Wang, C.Y.
1985-01-01
This paper presents some of the current capabilities of the three-dimensional piping code SHAPS and demonstrates their usefulness in handling analyses encountered in typical LMFBR studies. Several examples demonstrate the utility of the SHAPS code for problems involving fluid-structure interactions and seismic-related events occurring in three-dimensional piping networks. Results of two studies of pressure wave propagation demonstrate the dynamic coupling of pipes and elbows producing global motion and rigorous treatment of physical quantities such as changes in density, pressure, and strain energy. Results of the seismic analysis demonstrate the capability of SHAPS to handle dynamic structural response within a piping network over an extended transient period of several seconds. Variation in dominant stress frequencies and global translational frequencies were easily handled with the code. 4 refs., 10 figs
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
Integrated framework for dynamic safety analysis
International Nuclear Information System (INIS)
Kim, Tae Wan; Karanki, Durga R.
2012-01-01
In the conventional PSA (Probabilistic Safety Assessment), detailed plant simulations by independent thermal hydraulic (TH) codes are used in the development of accident sequence models. Typical accidents in a NPP involve complex interactions among process, safety systems, and operator actions. As independent TH codes do not have the models of operator actions and full safety systems, they cannot literally simulate the integrated and dynamic interactions of process, safety systems, and operator responses. Offline simulation with pre decided states and time delays may not model the accident sequences properly. Moreover, when stochastic variability in responses of accident models is considered, defining all the combinations for simulations will be cumbersome task. To overcome some of these limitations of conventional safety analysis approach, TH models are coupled with the stochastic models in the dynamic event tree (DET) framework, which provides flexibility to model the integrated response due to better communication as all the accident elements are in the same model. The advantages of this framework also include: Realistic modeling in dynamic scenarios, comprehensive results, integrated approach (both deterministic and probabilistic models), and support for HRA (Human Reliability Analysis)
Methods and computer codes for probabilistic sensitivity and uncertainty analysis
International Nuclear Information System (INIS)
Vaurio, J.K.
1985-01-01
This paper describes the methods and applications experience with two computer codes that are now available from the National Energy Software Center at Argonne National Laboratory. The purpose of the SCREEN code is to identify a group of most important input variables of a code that has many (tens, hundreds) input variables with uncertainties, and do this without relying on judgment or exhaustive sensitivity studies. Purpose of the PROSA-2 code is to propagate uncertainties and calculate the distributions of interesting output variable(s) of a safety analysis code using response surface techniques, based on the same runs used for screening. Several applications are discussed, but the codes are generic, not tailored to any specific safety application code. They are compatible in terms of input/output requirements but also independent of each other, e.g., PROSA-2 can be used without first using SCREEN if a set of important input variables has first been selected by other methods. Also, although SCREEN can select cases to be run (by random sampling), a user can select cases by other methods if he so prefers, and still use the rest of SCREEN for identifying important input variables
COOL: A code for Dynamic Monte Carlo Simulation of molecular dynamics
Barletta, Paolo
2012-02-01
Cool is a program to simulate evaporative and sympathetic cooling for a mixture of two gases co-trapped in an harmonic potential. The collisions involved are assumed to be exclusively elastic, and losses are due to evaporation from the trap. Each particle is followed individually in its trajectory, consequently properties such as spatial densities or energy distributions can be readily evaluated. The code can be used sequentially, by employing one output as input for another run. The code can be easily generalised to describe more complicated processes, such as the inclusion of inelastic collisions, or the possible presence of more than two species in the trap. New version program summaryProgram title: COOL Catalogue identifier: AEHJ_v2_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEHJ_v2_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 1 097 733 No. of bytes in distributed program, including test data, etc.: 18 425 722 Distribution format: tar.gz Programming language: C++ Computer: Desktop Operating system: Linux RAM: 500 Mbytes Classification: 16.7, 23 Catalogue identifier of previous version: AEHJ_v1_0 Journal reference of previous version: Comput. Phys. Comm. 182 (2011) 388 Does the new version supersede the previous version?: Yes Nature of problem: Simulation of the sympathetic process occurring for two molecular gases co-trapped in a deep optical trap. Solution method: The Direct Simulation Monte Carlo method exploits the decoupling, over a short time period, of the inter-particle interaction from the trapping potential. The particle dynamics is thus exclusively driven by the external optical field. The rare inter-particle collisions are considered with an acceptance/rejection mechanism, that is, by comparing a random number to the collisional probability
Dynamic Event Tree Analysis Through RAVEN
Energy Technology Data Exchange (ETDEWEB)
A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. A. Kinoshita; A. Naviglio
2013-09-01
Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics is not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (D-PRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed in a high modular and pluggable way in order to enable easy integration of different programming languages (i.e., C++, Python) and coupling with other application including the ones based on the MOOSE framework, developed by INL as well. RAVEN performs two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, RAVEN also models stochastic events, such as components failures, and performs uncertainty quantification. Such stochastic modeling is employed by using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This paper focuses on the first task and shows how it is possible to perform the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, the Dynamic PRA analysis, using Dynamic Event Tree, of a simplified pressurized water reactor for a Station Black-Out scenario is presented.
PCLOOK: an interactive code for spectral analysis
International Nuclear Information System (INIS)
Macchiavelli, A.O.; Tomasi, D.
1993-01-01
The present work describes an interactive programme for the analysis of spectra developed to run in a PC platform. PCLOOK has a graphic interface that allows the user to get access to different functions using the mouse or directly typing commands. In this way one can switch to a suitable required environment to manage the histograms reassembling in this way a spectrum calculator.The PCLOOK programme was mainly developed to use in nuclear physics applications, but it is also possible to modify it with relative little effort to adapt it to other applications. It was written in Microsoft's BASIC 7.1 installed in a 33MHz 486 Everex PC. For a proper operation an ordinary VGA display and mouse are needed. The memory requirements depend on the size and number of the user defined spectra; for instance, for twenty 2048 channels spectra the available memory space must be 320 KBytes. (author). 5 figs
Development of non-linear vibration analysis code for CANDU fuelling machine
International Nuclear Information System (INIS)
Murakami, Hajime; Hirai, Takeshi; Horikoshi, Kiyomi; Mizukoshi, Kaoru; Takenaka, Yasuo; Suzuki, Norio.
1988-01-01
This paper describes the development of a non-linear, dynamic analysis code for the CANDU 600 fuelling machine (F-M), which includes a number of non-linearities such as gap with or without Coulomb friction, special multi-linear spring connections, etc. The capabilities and features of the code and the mathematical treatment for the non-linearities are explained. The modeling and numerical methodology for the non-linearities employed in the code are verified experimentally. Finally, the simulation analyses for the full-scale F-M vibration testing are carried out, and the applicability of the code to such multi-degree of freedom systems as F-M is demonstrated. (author)
Static Analysis of Dynamic Languages
DEFF Research Database (Denmark)
Madsen, Magnus
Dynamic programming languages are highly popular and widely used. Java- Script is often called the lingua franca of the web and it is the de facto standard for client-side web programming. On the server-side the PHP, Python and Ruby languages are prevalent. What these languages have in common...... with static type systems, such as Java and C# , but the same features are rarely available for dynamic languages such as JavaScript. The aim of this thesis is to investigate techniques for improving the tool- support for dynamic programming languages without imposing any artificial restrictions...... of new dataflow analysis techniques to tackle the nature of dynamic programming languages....
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)
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)
Impact testing and analysis for structural code benchmarking
International Nuclear Information System (INIS)
Glass, R.E.
1989-01-01
Sandia National Laboratories, in cooperation with industry and other national laboratories, has been benchmarking computer codes used to predict the structural, thermal, criticality, and shielding behavior of radioactive materials packages. The first step in the benchmarking of the codes was to develop standard problem sets and to compare the results from several codes and users. This step for structural analysis codes has been completed as described in Structural Code Benchmarking for the Analysis of Impact Response of Nuclear Material Shipping Casks, R.E. Glass, Sandia National Laboratories, 1985. The problem set is shown in Fig. 1. This problem set exercised the ability of the codes to predict the response to end (axisymmetric) and side (plane strain) impacts with both elastic and elastic/plastic materials. The results from these problems showed that there is good agreement in predicting elastic response. Significant differences occurred in predicting strains for the elastic/plastic models. An example of the variation in predicting plastic behavior is given, which shows the hoop strain as a function of time at the impacting end of Model B. These differences in predicting plastic strains demonstrated a need for benchmark data for a cask-like problem
Integrated severe accident containment analysis with the CONTAIN computer code
International Nuclear Information System (INIS)
Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.
1985-12-01
Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite
NACOM - a code for sodium spray fire analysis
International Nuclear Information System (INIS)
Rao, P.M.; Kannan, S.E.
2002-01-01
Full text: In liquid metal fast breeder reactors (LMFBR), leakage of sodium can result in a spray fire. Because of higher burning rates in droplet form combustion of sodium in spray fire, thermal consequences are more severe than that in a sodium pool fire. The code NACOM was developed for the analysis of sodium spray fires in LMFBRs facilities. The code uses the validated model for estimating the falling droplet burning rates in pre-ignition and vapour phase combustion stages. It uses a distribution system to generate the droplet groups of different diameters that represent the spray. The code requires about 20 input parameters like sodium leak rates, sodium temperature, initial cell conditions like oxygen concentration, temperature and dimensions. NACOM is a validated code based on experiments with sodium inventory up to 650 kg in 0 to 21 % O 2 atmospheres. The paper brings out the salient features of the code along with the sensitivity analysis of the main input parameters like spray volume mean diameter, oxygen concentration etc. based on the results obtained. The limitations of the code and the confidence margins applicable to results obtained are also brought out
Analysis of pipe stress using CAESAR II code
International Nuclear Information System (INIS)
Sitandung, Y.B.; Bandriyana, B.
2002-01-01
Analysis of this piping stress with the purpose of knowing stress distribution piping system in order to determine pipe supports configuration. As an example of analysis, Gas Exchanger to Warm Separator Line was chosen with, input data was firstly prepared in a document, i.e. piping analysis specification that its content named as pipe characteristics, material properties, operation conditions, guide equipment's and so on. Analysis result such as stress, load, displacement and the use support type were verified based on requirements in the code, standard, and regularities were suitable with piping system condition analyzed. As the proof that piping system is in safety condition, it can be indicated from analysis results (actual loads) which still under allowable load. From the analysis steps that have been done CAESAR II code fulfill requirements to be used as a tool of piping stress analysis as well as nuclear and non nuclear installation piping system
Demonstration study on shielding safety analysis code (VI)
Energy Technology Data Exchange (ETDEWEB)
Sawamura, Sadashi [Hokkaido Univ., Sapporo (Japan). Faculty of Engineering
1999-03-01
Dose evaluation for direct radiation and skyshine from nuclear fuel facilities is one of the environment evaluation items. This evaluation is carried out by using some shielding calculation codes. Because of extremely few benchmark data of skyshine, the calculation has to be performed very conservatively. Therefore, the benchmark data of skyshine and the well-investigated code for skyshine would be necessary to carry out the rational evaluation of nuclear facilities. The purpose of this steady is to obtain the benchmark data of skyshine and to investigate the calculation code for skyshine. In this fiscal year, the followings are investigated; (1) Construction and improvement of a pulsed radiation measurement system due to the gated counting method. (2) Using the system, carried out the radiation monitoring near and in the facility of 45 MeV Linear accelerator installed at Hokkaido University. (3) Simulation analysis of the photo-neutron production and the transport by using the EGS4 and MCNP code. (author)
Two-phase flow characteristics analysis code: MINCS
International Nuclear Information System (INIS)
Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.
1992-03-01
Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)
A development of containment performance analysis methodology using GOTHIC code
Energy Technology Data Exchange (ETDEWEB)
Lee, B. C.; Yoon, J. I. [Future and Challenge Company, Seoul (Korea, Republic of); Byun, C. S.; Lee, J. Y. [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, J. Y. [Seoul National University, Seoul (Korea, Republic of)
2003-10-01
In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code.
A development of containment performance analysis methodology using GOTHIC code
International Nuclear Information System (INIS)
Lee, B. C.; Yoon, J. I.; Byun, C. S.; Lee, J. Y.; Lee, J. Y.
2003-01-01
In a circumstance that well-established containment pressure/temperature analysis code, CONTEMPT-LT treats the reactor containment as a single volume, this study introduces, as an alternative, the GOTHIC code for an usage on multi-compartmental containment performance analysis. With a developed GOTHIC methodology, its applicability is verified for containment performance analysis for Korean Nuclear Unit 1. The GOTHIC model for this plant is simply composed of 3 compartments including the reactor containment and RWST. In addition, the containment spray system and containment recirculation system are simulated. As a result of GOTHIC calculation, under the same assumptions and conditions as those in CONTEMPT-LT, the GOTHIC prediction shows a very good result; pressure and temperature transients including their peaks are nearly the same. It can be concluded that the GOTHIC could provide reasonable containment pressure and temperature responses if considering the inherent conservatism in CONTEMPT-LT code
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
SIMULAND - A CODE TO ANALYZE DYNAMIC EFFECTS DURING LANDING
Directory of Open Access Journals (Sweden)
Marcel STERE
2010-03-01
Full Text Available The landing gear of an aircraft is part of the aircraft structure. It is the most critical part of the flight mission and also the component that will likely cause the most problems in the aircraft design. The landing gear design combines the best in mechanical, structural and hydraulic design. The designed landing gear should be able to meet the specifications and requirements imposed by the CS23. SIMULAND-01 is a program intended to analyze a reduced model (4-30 DoF of the aircraft under transient dynamic loads during the landing phase (touchdown.
FARO base case post-test analysis by COMETA code
Energy Technology Data Exchange (ETDEWEB)
Annunziato, A.; Addabbo, C. [Joint Research Centre, Ispra (Italy)
1995-09-01
The paper analyzes the COMETA (Core Melt Thermal-Hydraulic Analysis) post test calculations of FARO Test L-11, the so-called Base Case Test. The FARO Facility, located at JRC Ispra, is used to simulate the consequences of Severe Accidents in Nuclear Power Plants under a variety of conditions. The COMETA Code has a 6 equations two phase flow field and a 3 phases corium field: the jet, the droplets and the fused-debris bed. The analysis shown that the code is able to pick-up all the major phenomena occurring during the fuel-coolant interaction pre-mixing phase.
Severe accident analysis code Sampson for impact project
International Nuclear Information System (INIS)
Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh
2001-01-01
Four years of the IMPACT project Phase 1 (1994-1997) had been completed with financial sponsorship from the Japanese government's Ministry of Economy, Trade and Industry. At the end of the phase, demonstration simulations by combinations of up to 11 analysis modules developed for severe accident analysis in the SAMPSON Code were performed and physical models in the code were verified. The SAMPSON prototype was validated by TMI-2 and Phebus-FP test analyses. Many of empirical correlation and conventional models have been replaced by mechanistic models during Phase 2 (1998-2000). New models for Accident Management evaluation have been also developed. (author)
Transient electromagnetic analysis in tokamaks using TYPHOON code
International Nuclear Information System (INIS)
Belov, A.V.; Duke, A.E.; Korolkov, M.D.; Kotov, V.L.; Kukhtin, V.P.; Lamzin, E.A.; Sytchevsky, S.E.
1996-01-01
The transient electromagnetic analysis of conducting structures in tokamaks is presented. This analysis is based on a three-dimensional thin conducting shell model. The finite element method has been used to solve the corresponding integrodifferential equation. The code TYPHOON has been developed to calculate transient processes in tokamaks. Calculation tests and the code verification have been carried out. The calculation results of eddy current and force distibution and a.c. losses for different construction elements for both ITER and TEXTOR tokamaks magnetic systems are presented. (orig.)
Choreographer Pre-Testing Code Analysis and Operational Testing.
Energy Technology Data Exchange (ETDEWEB)
Fritz, David J. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Harrison, Christopher B. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Perr, C. W. [Sandia National Laboratories (SNL-CA), Livermore, CA (United States); Hurd, Steven A [Sandia National Laboratories (SNL-CA), Livermore, CA (United States)
2014-07-01
Choreographer is a "moving target defense system", designed to protect against attacks aimed at IP addresses without corresponding domain name system (DNS) lookups. It coordinates actions between a DNS server and a Network Address Translation (NAT) device to regularly change which publicly available IP addresses' traffic will be routed to the protected device versus routed to a honeypot. More details about how Choreographer operates can be found in Section 2: Introducing Choreographer. Operational considerations for the successful deployment of Choreographer can be found in Section 3. The Testing & Evaluation (T&E) for Choreographer involved 3 phases: Pre-testing, Code Analysis, and Operational Testing. Pre-testing, described in Section 4, involved installing and configuring an instance of Choreographer and verifying it would operate as expected for a simple use case. Our findings were that it was simple and straightforward to prepare a system for a Choreographer installation as well as configure Choreographer to work in a representative environment. Code Analysis, described in Section 5, consisted of running a static code analyzer (HP Fortify) and conducting dynamic analysis tests using the Valgrind instrumentation framework. Choreographer performed well, such that only a few errors that might possibly be problematic in a given operating situation were identified. Operational Testing, described in Section 6, involved operating Choreographer in a representative environment created through Emulytics^{TM} . Depending upon the amount of server resources dedicated to Choreographer vis-á-vis the amount of client traffic handled, Choreographer had varying degrees of operational success. In an environment with a poorly resourced Choreographer server and as few as 50-100 clients, Choreographer failed to properly route traffic over half the time. Yet, with a well-resourced server, Choreographer handled over 1000 clients without missrouting. Choreographer
International Nuclear Information System (INIS)
Taji, Yukichi
1984-06-01
Development has been made of molecular dynamical codes DGR and GGR to simulate radiation damages yielded in the diamond and graphite structure crystals, respectively. Though the usual molecular dynamical codes deal only with the central forces as the mutual interactions between atoms, the present codes can take account of noncentral forces to represent the effect of the covalent bonds characteristic of diamond or graphite crystals. It is shown that lattice defects yielded in these crystals are stable by themselves in the present method without any supports of virtual surface forces set on the crystallite surfaces. By this effect the behavior of lattice defects has become possible to be simulated in a more realistic manner. Some examples of the simulation with these codes are shown. (author)
Development of a code for the isotopic analysis of Uranium
Energy Technology Data Exchange (ETDEWEB)
Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
To strengthen the national nuclear nonproliferation regime by an establishment of nuclear forensic system, the techniques for nuclear material analysis and the categorization of important domestic nuclear materials are being developed. MGAU and FRAM are commercial software for the isotopic analysis of Uranium by using γ-spectroscopy, but the diversity of detection geometry and some effects - self attenuation, coincidence summing, etc. - suggest an analysis tool under continual improvement and modification. Hence, developing another code for HPGe γ- and x-ray spectrum analysis is started in this study. The analysis of the 87-101 keV region of Uranium spectrum is attempted based on the isotopic responses similar to those developed in MGAU. The code for isotopic analysis of Uranium is started from a fitting.
QR Codes in the Library: Are They Worth the Effort? Analysis of a QR Code Pilot Project
Wilson, Andrew M.
2012-01-01
The literature is filled with potential uses for Quick Response (QR) codes in the library. Setting, but few library QR code projects have publicized usage statistics. A pilot project carried out in the Eda Kuhn Loeb Music Library of the Harvard College Library sought to determine whether library patrons actually understand and use QR codes. Results and analysis of the pilot project are provided, attempting to answer the question as to whether QR codes are worth the effort for libraries.
Computer codes for the analysis of flask impact problems
International Nuclear Information System (INIS)
Neilson, A.J.
1984-09-01
This review identifies typical features of the design of transportation flasks and considers some of the analytical tools required for the analysis of impact events. Because of the complexity of the physical problem, it is unlikely that a single code will adequately deal with all the aspects of the impact incident. Candidate codes are identified on the basis of current understanding of their strengths and limitations. It is concluded that the HONDO-II, DYNA3D AND ABAQUS codes which ar already mounted on UKAEA computers will be suitable tools for use in the analysis of experiments conducted in the proposed AEEW programme and of general flask impact problems. Initial attention should be directed at the DYNA3D and ABAQUS codes with HONDO-II being reserved for situations where the three-dimensional elements of DYNA3D may provide uneconomic simulations in planar or axisymmetric geometries. Attention is drawn to the importance of access to suitable mesh generators to create the nodal coordinate and element topology data required by these structural analysis codes. (author)
Java Source Code Analysis for API Migration to Embedded Systems
Energy Technology Data Exchange (ETDEWEB)
Winter, Victor [Univ. of Nebraska, Omaha, NE (United States); McCoy, James A. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Guerrero, Jonathan [Univ. of Nebraska, Omaha, NE (United States); Reinke, Carl Werner [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perry, James Thomas [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)
2015-02-01
Embedded systems form an integral part of our technological infrastructure and oftentimes play a complex and critical role within larger systems. From the perspective of reliability, security, and safety, strong arguments can be made favoring the use of Java over C in such systems. In part, this argument is based on the assumption that suitable subsets of Java’s APIs and extension libraries are available to embedded software developers. In practice, a number of Java-based embedded processors do not support the full features of the JVM. For such processors, source code migration is a mechanism by which key abstractions offered by APIs and extension libraries can made available to embedded software developers. The analysis required for Java source code-level library migration is based on the ability to correctly resolve element references to their corresponding element declarations. A key challenge in this setting is how to perform analysis for incomplete source-code bases (e.g., subsets of libraries) from which types and packages have been omitted. This article formalizes an approach that can be used to extend code bases targeted for migration in such a manner that the threats associated the analysis of incomplete code bases are eliminated.
Development of the vacuum system pressure responce analysis code PRAC
International Nuclear Information System (INIS)
Horie, Tomoyoshi; Kawasaki, Kouzou; Noshiroya, Shyoji; Koizumi, Jun-ichi.
1985-03-01
In this report, we show the method and numerical results of the vacuum system pressure responce analysis code. Since fusion apparatus is made up of many vacuum components, it is required to analyze pressure responce at any points of the system when vacuum system is designed or evaluated. For that purpose evaluating by theoretical solution is insufficient. Numerical analysis procedure such as finite difference method is usefull. In the PRAC code (Pressure Responce Analysis Code), pressure responce is obtained solving derivative equations which is obtained from the equilibrium relation of throughputs and contain the time derivative of pressure. As it considers both molecular and viscous flows, the coefficients of the equation depend on the pressure and the equations become non-linear. This non-linearity is treated as piece-wise linear within each time step. Verification of the code is performed for the simple problems. The agreement between numerical and theoretical solutions is good. To compare with the measured results, complicated model of gas puffing system is analyzed. The agreement is well for practical use. This code will be a useful analytical tool for designing and evaluating vacuum systems such as fusion apparatus. (author)
Issues in computational fluid dynamics code verification and validation
Energy Technology Data Exchange (ETDEWEB)
Oberkampf, W.L.; Blottner, F.G.
1997-09-01
A broad range of mathematical modeling errors of fluid flow physics and numerical approximation errors are addressed in computational fluid dynamics (CFD). It is strongly believed that if CFD is to have a major impact on the design of engineering hardware and flight systems, the level of confidence in complex simulations must substantially improve. To better understand the present limitations of CFD simulations, a wide variety of physical modeling, discretization, and solution errors are identified and discussed. Here, discretization and solution errors refer to all errors caused by conversion of the original partial differential, or integral, conservation equations representing the physical process, to algebraic equations and their solution on a computer. The impact of boundary conditions on the solution of the partial differential equations and their discrete representation will also be discussed. Throughout the article, clear distinctions are made between the analytical mathematical models of fluid dynamics and the numerical models. Lax`s Equivalence Theorem and its frailties in practical CFD solutions are pointed out. Distinctions are also made between the existence and uniqueness of solutions to the partial differential equations as opposed to the discrete equations. Two techniques are briefly discussed for the detection and quantification of certain types of discretization and grid resolution errors.
Spectral history modeling in the reactor dynamics code DYN3D
International Nuclear Information System (INIS)
Bilodid, Yurii
2014-01-01
A new method of treating spectral history effects in reactor core calculations was developed and verified in this dissertation. The nature of history effects is a dependence of fuel properties not only on the burnup, but also on the local spectral conditions during burnup. The basic idea of the proposed method is the use of the plutonium-239 concentration as the spectral history indicator. The method was implemented in the reactor dynamics code DYN3D and provides a correction for nodal cross sections according to the local spectral history. A verification of the new method was performed by single-assembly calculations in comparison with results of the lattice code HELIOS. The application of plutonium-based history correction significantly improves the cross section estimation accuracy both for UOX and MOX fuel, with quadratic and hexagonal geometry. The new method was applied to evaluate the influence of history effects on full-core calculation results. Analysis of a PWR equilibrium fuel cycle has shown a significant effect on the axial power distribution during a whole cycle, which causes axial temperature and burnup redistributions. The observed neutron flux redistribution improves neutron economy, so the fuel cycle is longer than in calculations without history corrections. Analyses of hypothetical control rod ejection accidents have shown a minor influence of history effects on the transient course and safety relevant parameters.
KEWPIE2: A cascade code for the study of dynamical decay of excited nuclei
Lu , H.; Marchix , A.; Abe , Y.; Boilley , D.
2016-01-01
Submitted to Computer Physics Communications; International audience; KEWPIE — a cascade code devoted to investigating the dynamical decay of excited nuclei, specially designed for treating very low probability events related to the synthesis of super-heavy nuclei formed in fusion-evaporation reactions — has been improved and rewritten in C++ programing language to become KEWPIE2. The current version of the code comprises various nuclear models concerning the light-particle emission, fission ...
LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests
International Nuclear Information System (INIS)
Bordner, G.L.
1979-10-01
The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes
Integrated transport code system for a multicomponent plasma in a gas dynamic trap
International Nuclear Information System (INIS)
Anikeev, A.V.; Karpushov, A.N.; Noak, K.; Strogalova, S.L.
2000-01-01
This report is focused on the development of the theoretical and numerical models of multicomponent high-β plasma confinement and transport in the gas-dynamic trap (GDT). In order to simulate the plasma behavior in the GDT as well as that in the GDT-based neutron source the Integrated Transport Code System is developed from existing stand-alone codes calculating the target plasma, the fast ions and the neutral gas in the GDT. The code system considers the full dependence of the transport phenomena on space, time, energy and angle variables as well as the interactions between the particle fields [ru
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
Comparison study of inelastic analysis codes for high temperature structure
Energy Technology Data Exchange (ETDEWEB)
Kim, Jong Bum; Lee, H. Y.; Park, C. K.; Geon, G. P.; Lee, J. H
2004-02-01
LMR high temperature structures subjected to operating and transient loadings may exhibit very complex deformation behaviors due to the use of ductile material such as 316SS and the systematic analysis technology of high temperature structure for reliable safety assessment is essential. In this project, comparative study with developed inelastic analysis program NONSTA and the existing analysis codes was performed applying various types of loading including non-proportional loading. The performance of NONSTA was confirmed and the effect of inelastic constants on the analysis result was analyzed. Also, the applicability of the inelastic analysis was enlarged as a result of applying both the developed program and the existing codes to the analyses of the enhanced creep behavior and the elastic follow-up behavior of high temperature structures and the necessary items for improvements were deduced. Further studies on the improvement of NONSTA program and the decision of the proper values of inelastic constants are necessary.
Easy web interfaces to IDL code for NSTX Data Analysis
International Nuclear Information System (INIS)
Davis, W.M.
2012-01-01
Highlights: ► Web interfaces to IDL code can be developed quickly. ► Dozens of Web Tools are used effectively on NSTX for Data Analysis. ► Web interfaces are easier to use than X-window applications. - Abstract: Reusing code is a well-known Software Engineering practice to substantially increase the efficiency of code production, as well as to reduce errors and debugging time. A variety of “Web Tools” for the analysis and display of raw and analyzed physics data are in use on NSTX [1], and new ones can be produced quickly from existing IDL [2] code. A Web Tool with only a few inputs, and which calls an IDL routine written in the proper style, can be created in less than an hour; more typical Web Tools with dozens of inputs, and the need for some adaptation of existing IDL code, can be working in a day or so. Efficiency is also increased for users of Web Tools because of the familiar interface of the web browser, and not needing X-windows, or accounts and passwords, when used within our firewall. Web Tools were adapted for use by PPPL physicists accessing EAST data stored in MDSplus with only a few man-weeks of effort; adapting to additional sites should now be even easier. An overview of Web Tools in use on NSTX, and a list of the most useful features, is also presented.
[Bioethical analysis of the Brazilian Dentistry Code of Ethics].
Pyrrho, Monique; do Prado, Mauro Machado; Cordón, Jorge; Garrafa, Volnei
2009-01-01
The Brazilian Dentistry Code of Ethics (DCE), Resolution CFO-71 from May 2006, is an instrument created to guide dentists' behavior in relation to the ethical aspects of professional practice. The purpose of the study is to analyze the above mentioned code comparing the deontological and bioethical focuses. In order to do so, an interpretative analysis of the code and of twelve selected texts was made. Six of the texts were about bioethics and six on deontology, and the analysis was made through the methodological classification of the context units, textual paragraphs and items from the code in the following categories: the referentials of bioethical principlism--autonomy, beneficence, nonmaleficence and justice -, technical aspects and moral virtues related to the profession. Together the four principles represented 22.9%, 39.8% and 54.2% of the content of the DCE, of the deontological texts and of the bioethical texts respectively. In the DCE, 42% of the items referred to virtues, 40.2% were associated to technical aspects and just 22.9% referred to principles. The virtues related to the professionals and the technical aspects together amounted to 70.1% of the code. Instead of focusing on the patient as the subject of the process of oral health care, the DCE focuses on the professional, and it is predominantly turned to legalistic and corporate aspects.
Development of a subchannel analysis code MATRA (Ver. α)
International Nuclear Information System (INIS)
Yoo, Y. J.; Hwang, D. H.
1998-04-01
A subchannel analysis code MATRA-α, an interim version of MATRA, has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-IV-I. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and flow distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. MATRA-α has been provided with an improved structure, various functions, and models to give the more convenient user environment and to increase the code accuracy, various functions, and models to give the more convenient user environment and to increase the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the lateral transport models between adjacent subchannels have been improved to increase the accuracy in predicting two-phase flow phenomena. Also included in this report are the detailed instructions for input data preparation and for auxiliary pre-processors to serve as a guide to those who want to use MATRA-α. In addition, we compared the predictions of MATRA-α with the experimental data on the flow and enthalpy distribution in three sample rod-bundle cases to evaluate the performance of MATRA-α. All the results revealed that the prediction of MATRA-α were better than those of COBRA-IV-I. (author). 16 refs., 1 tab., 13 figs
Directory of Open Access Journals (Sweden)
Juan P. Marcolongo
2018-03-01
Full Text Available In this work we present the current advances in the development and the applications of LIO, a lab-made code designed for density functional theory calculations in graphical processing units (GPU, that can be coupled with different classical molecular dynamics engines. This code has been thoroughly optimized to perform efficient molecular dynamics simulations at the QM/MM DFT level, allowing for an exhaustive sampling of the configurational space. Selected examples are presented for the description of chemical reactivity in terms of free energy profiles, and also for the computation of optical properties, such as vibrational and electronic spectra in solvent and protein environments.
Marcolongo, Juan P.; Zeida, Ari; Semelak, Jonathan A.; Foglia, Nicolás O.; Morzan, Uriel N.; Estrin, Dario A.; González Lebrero, Mariano C.; Scherlis, Damián A.
2018-03-01
In this work we present the current advances in the development and the applications of LIO, a lab-made code designed for density functional theory calculations in graphical processing units (GPU), that can be coupled with different classical molecular dynamics engines. This code has been thoroughly optimized to perform efficient molecular dynamics simulations at the QM/MM DFT level, allowing for an exhaustive sampling of the configurational space. Selected examples are presented for the description of chemical reactivity in terms of free energy profiles, and also for the computation of optical properties, such as vibrational and electronic spectra in solvent and protein environments.
Dynamic stimulated Brillouin scattering analysis
DEFF Research Database (Denmark)
Djupsöbacka, A.; Jacobsen, Gunnar; Tromborg, Bjarne
2000-01-01
We present a new simple analysis - including the effect of spontaneous emission - of the (dynamic) influence of SBS on the detected receiver eye diagram. It applies in principle for general types of modulation formats such as the digital formats of ASK, FSK, and PSK. The analysis is formulated fo...
Enabling dynamics in face analysis
Dibeklioğlu, H.
2014-01-01
Most of the approaches in automatic face analysis rely solely on static appearance. However, temporal analysis of expressions reveals interesting patterns. For a better understanding of the human face, this thesis focuses on temporal changes in the face, and dynamic patterns of expressions. In
An Analysis of the Changes in Communication Techniques in the Italian Codes of Medical Deontology.
Conti, Andrea Alberto
2017-04-28
The code of deontology of the Italian National Federation of the Colleges of Physicians, Surgeons and Dentists (FNOMCeO) contains the principles and rules to which the professional medical practitioner must adhere. This work identifies and analyzes the medical-linguistic choices and the expressive techniques present in the different editions of the code, and evaluates their purpose and function, focusing on the first appearance and the subsequent frequency of key terms. Various aspects of the formal and expressive revisions of the eight editions of the Codes of Medical Deontology published after the Second World War (from 1947/48 to 2014) are here presented, starting from a brief comparison with the first edition of 1903. Formal characteristics, choices of medical terminology and the introduction of new concepts and communicative attitudes are here identified and evaluated. This paper, in presenting a quantitative and epistemological analysis of variations, modifications and confirmations in the different editions of the Italian code of medical deontology over the last century, enucleates and demonstrates the dynamic paradigm of changing attitudes in the medical profession. This analysis shows the evolution in medical-scientific communication as embodied in the Italian code of medical deontology. This code, in its adoption, changes and adaptations, as evidenced in its successive editions, bears witness to the expressions and attitudes pertinent to and characteristic of the deontological stance of the medical profession during the twentieth century.
ISODEP, A Fuel Depletion Analysis Code for Predicting Isotopic ...
African Journals Online (AJOL)
The trend of results was found to be consistent with those obtained by analytical and other numerical methods. Discovery and Innovation Vol. 13 no. 3/4 December (2001) pp. 184-195. KEY WORDS: depletion analysis, code, research reactor, simultaneous equations, decay of nuclides, radionuclitides, isotope. Résumé
Recent developments in seismic analysis in the code Aster
International Nuclear Information System (INIS)
Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.
1996-01-01
Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors)
High-Performance Java Codes for Computational Fluid Dynamics
Riley, Christopher; Chatterjee, Siddhartha; Biswas, Rupak; Biegel, Bryan (Technical Monitor)
2001-01-01
The computational science community is reluctant to write large-scale computationally -intensive applications in Java due to concerns over Java's poor performance, despite the claimed software engineering advantages of its object-oriented features. Naive Java implementations of numerical algorithms can perform poorly compared to corresponding Fortran or C implementations. To achieve high performance, Java applications must be designed with good performance as a primary goal. This paper presents the object-oriented design and implementation of two real-world applications from the field of Computational Fluid Dynamics (CFD): a finite-volume fluid flow solver (LAURA, from NASA Langley Research Center), and an unstructured mesh adaptation algorithm (2D_TAG, from NASA Ames Research Center). This work builds on our previous experience with the design of high-performance numerical libraries in Java. We examine the performance of the applications using the currently available Java infrastructure and show that the Java version of the flow solver LAURA performs almost within a factor of 2 of the original procedural version. Our Java version of the mesh adaptation algorithm 2D_TAG performs within a factor of 1.5 of its original procedural version on certain platforms. Our results demonstrate that object-oriented software design principles are not necessarily inimical to high performance.
Dynamic divisive normalization predicts time-varying value coding in decision-related circuits.
Louie, Kenway; LoFaro, Thomas; Webb, Ryan; Glimcher, Paul W
2014-11-26
Normalization is a widespread neural computation, mediating divisive gain control in sensory processing and implementing a context-dependent value code in decision-related frontal and parietal cortices. Although decision-making is a dynamic process with complex temporal characteristics, most models of normalization are time-independent and little is known about the dynamic interaction of normalization and choice. Here, we show that a simple differential equation model of normalization explains the characteristic phasic-sustained pattern of cortical decision activity and predicts specific normalization dynamics: value coding during initial transients, time-varying value modulation, and delayed onset of contextual information. Empirically, we observe these predicted dynamics in saccade-related neurons in monkey lateral intraparietal cortex. Furthermore, such models naturally incorporate a time-weighted average of past activity, implementing an intrinsic reference-dependence in value coding. These results suggest that a single network mechanism can explain both transient and sustained decision activity, emphasizing the importance of a dynamic view of normalization in neural coding. Copyright © 2014 the authors 0270-6474/14/3416046-12$15.00/0.
Energy Technology Data Exchange (ETDEWEB)
James, Scott Carlton; Roberts, Jesse D
2014-03-01
This document describes the marine hydrokinetic (MHK) input file and subroutines for the Sandia National Laboratories Environmental Fluid Dynamics Code (SNL-EFDC), which is a combined hydrodynamic, sediment transport, and water quality model based on the Environmental Fluid Dynamics Code (EFDC) developed by John Hamrick [1], formerly sponsored by the U.S. Environmental Protection Agency, and now maintained by Tetra Tech, Inc. SNL-EFDC has been previously enhanced with the incorporation of the SEDZLJ sediment dynamics model developed by Ziegler, Lick, and Jones [2-4]. SNL-EFDC has also been upgraded to more accurately simulate algae growth with specific application to optimizing biomass in an open-channel raceway for biofuels production [5]. A detailed description of the input file containing data describing the MHK device/array is provided, along with a description of the MHK FORTRAN routine. Both a theoretical description of the MHK dynamics as incorporated into SNL-EFDC and an explanation of the source code are provided. This user manual is meant to be used in conjunction with the original EFDC [6] and sediment dynamics SNL-EFDC manuals [7]. Through this document, the authors provide information for users who wish to model the effects of an MHK device (or array of devices) on a flow system with EFDC and who also seek a clear understanding of the source code, which is available from staff in the Water Power Technologies Department at Sandia National Laboratories, Albuquerque, New Mexico.
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.)
Notes on nuclear reactor core analysis code: CITATION
International Nuclear Information System (INIS)
Cepraga, D.G.
1980-01-01
The method which has evolved over the years for making power reactor calculations is the multigroup diffusion method. The CITATION code is designed to solve multigroup neutronics problems with application of the finite-difference diffusion theory approximation to neutron transport in up to three-dimensional geometry. The first part of this paper presents information about the mathematical equations programmed along with background material and certain displays to convey the nature of some of the formulations. The results obtained with the CITATION code regarding the neutron and burnup core analysis for a typical PWR reactor are presented in the second part of this paper. (author)
FARO and KROTOS code simulation and analysis at JRC Ispra
Energy Technology Data Exchange (ETDEWEB)
Annunziato, A.; Yerkess, A.; Addabbo, C. [European Commission-Joint Research Centre, Inst. for Systems, Informatics and Safety, 21020 Ispra (Italy)
1998-01-01
The paper summarizes relevant results from the pre and post test calculations of fuel coolant interaction and quenching tests performed in the FARO and KROTOS test facilities. The main analytical tools adopted at JRC Ispra are the COMETA and the TEXAS codes. COMETA pre and post test calculations of FARO Test L-20 as well as an application of the code to KROTOS test facility are presented. The analysis provides the need to account for H{sub 2} generation models into the pre-mixing calculations. In addition salient results from the application of TEXAS to FARO and KROTOS tests are shown. (author)
Integrated computer codes for nuclear power plant severe accident analysis
International Nuclear Information System (INIS)
Jordanov, I.; Khristov, Y.
1995-01-01
This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs
Integrated computer codes for nuclear power plant severe accident analysis
Energy Technology Data Exchange (ETDEWEB)
Jordanov, I; Khristov, Y [Bylgarska Akademiya na Naukite, Sofia (Bulgaria). Inst. za Yadrena Izsledvaniya i Yadrena Energetika
1996-12-31
This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs.
Room Heat-Up Analysis with GOTHIC code
International Nuclear Information System (INIS)
Jimenez, G.; Olza, J. M.
2010-01-01
The GOTHIC T M computer code is a state-of-the art program for modeling multiphase, multicomponent fluid flow. GOTHIC is rapidly becoming the industry-standard code for performing both containment design basis accident (DBA) analyses and analyses to support equipment qualification. GOTHIC has a flexible nodding structure that allows both lumped parameter and 3-D modeling capabilities. Multidimensional analysis capabilities greatly enhance the study of noncondensable gases and stratification and permit the calculation of flow field details within any given volume.
RADTRAN 5: A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1991-01-01
RADTRAN 5 is a computer code developed at Sandia National Laboratories (SNL) in Albuquerque, NM, to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI Standard FORTRAN 77 and contains significant advances in the methodology for route-specific analysis first developed by SNL for RADTRAN 4 (Neuhauser and Kanipe, 1992). Like the previous RADTRAN codes, RADTRAN 5 contains two major modules for incident-free and accident risk amlysis, respectively. All commercially important transportation modes may be analyzed with RADTRAN 5: highway by combination truck; highway by light-duty vehicle; rail; barge; ocean-going ship; cargo air; and passenger air
Organization of Risk Analysis Codes for Living Evaluations (ORACLE)
International Nuclear Information System (INIS)
Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.
1987-01-01
ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)
Statistical mechanics analysis of LDPC coding in MIMO Gaussian channels
Energy Technology Data Exchange (ETDEWEB)
Alamino, Roberto C; Saad, David [Neural Computing Research Group, Aston University, Birmingham B4 7ET (United Kingdom)
2007-10-12
Using analytical methods of statistical mechanics, we analyse the typical behaviour of a multiple-input multiple-output (MIMO) Gaussian channel with binary inputs under low-density parity-check (LDPC) network coding and joint decoding. The saddle point equations for the replica symmetric solution are found in particular realizations of this channel, including a small and large number of transmitters and receivers. In particular, we examine the cases of a single transmitter, a single receiver and symmetric and asymmetric interference. Both dynamical and thermodynamical transitions from the ferromagnetic solution of perfect decoding to a non-ferromagnetic solution are identified for the cases considered, marking the practical and theoretical limits of the system under the current coding scheme. Numerical results are provided, showing the typical level of improvement/deterioration achieved with respect to the single transmitter/receiver result, for the various cases.
Statistical mechanics analysis of LDPC coding in MIMO Gaussian channels
International Nuclear Information System (INIS)
Alamino, Roberto C; Saad, David
2007-01-01
Using analytical methods of statistical mechanics, we analyse the typical behaviour of a multiple-input multiple-output (MIMO) Gaussian channel with binary inputs under low-density parity-check (LDPC) network coding and joint decoding. The saddle point equations for the replica symmetric solution are found in particular realizations of this channel, including a small and large number of transmitters and receivers. In particular, we examine the cases of a single transmitter, a single receiver and symmetric and asymmetric interference. Both dynamical and thermodynamical transitions from the ferromagnetic solution of perfect decoding to a non-ferromagnetic solution are identified for the cases considered, marking the practical and theoretical limits of the system under the current coding scheme. Numerical results are provided, showing the typical level of improvement/deterioration achieved with respect to the single transmitter/receiver result, for the various cases
Simulation of IST Turbomachinery Power-Neutral Tests with the ANL Plant Dynamics Code
Energy Technology Data Exchange (ETDEWEB)
Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. J. [Argonne National Lab. (ANL), Argonne, IL (United States)
2016-12-13
The validation of the Plant Dynamics Code (PDC) developed at Argonne National Laboratory (ANL) for the steady-state and transient analysis of supercritical carbon dioxide (sCO2) systems has been continued with new test data from the Naval Nuclear Laboratory (operated by Bechtel Marine Propulsion Corporation) Integrated System Test (IST). Although data from three runs were provided to ANL, only two of the data sets were analyzed and described in this report. The common feature of these tests is the power-neutral operation of the turbine-compressor shaft, where no external power through the alternator was provided during the tests. Instead, the shaft speed was allowed to change dictated by the power balance between the turbine, the compressor, and the power losses in the shaft. The new test data turned out to be important for code validation for several reasons. First, the power-neutral operation of the shaft allows validation of the shaft dynamics equations in asynchronous mode, when the shaft is disconnected from the grid. Second, the shaft speed control with the compressor recirculation (CR) valve not only allows for testing the code control logic itself, but it also serves as a good test for validation of both the compressor surge control and the turbine bypass control actions, since the effect of the CR action on the loop conditions is similar for both of these controls. Third, the varying compressor-inlet temperature change test allows validation of the transient response of the precooler, a shell-and-tube heat exchanger. The first transient simulation of the compressor-inlet temperature variation Test 64661 showed a much slower calculated response of the precooler in the calculations than the test data. Further investigation revealed an error in calculating the heat exchanger tube mass for the PDC dynamic equations that resulted in a slower change in the tube wall temperature than measured. The transient calculations for both tests were done in two steps. The
Probabilistic analysis of crack containing structures with the PARIS code
International Nuclear Information System (INIS)
Brueckner-Foit, A.
1987-10-01
The basic features of the PARIS code which has been developed for the calculation of failure probabilities of crack containing structures are explained. An important issue in the reliability analysis of cracked components is the probabilistic leak-before-break behaviour. Formulae for the leak and break probabilities are derived and it is shown how a leak detection system influences the results. An example taken from nuclear applications illustrates the details of the probabilistic leak-before-break analysis. (orig.) [de
Development and assessment of best estimate integrated safety analysis code
Energy Technology Data Exchange (ETDEWEB)
Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu (and others)
2007-03-15
Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published.
Development and assessment of best estimate integrated safety analysis code
International Nuclear Information System (INIS)
Chung, Bub Dong; Lee, Young Jin; Hwang, Moon Kyu
2007-03-01
Improvement of the integrated safety analysis code MARS3.0 has been carried out and a multi-D safety analysis application system has been established. Iterative matrix solver and parallel processing algorithm have been introduced, and a LINUX version has been generated to enable MARS to run in cluster PCs. MARS variables and sub-routines have been reformed and modularised to simplify code maintenance. Model uncertainty analyses have been performed for THTF, FLECHT, NEPTUN, and LOFT experiments as well as APR1400 plant. Participations in international cooperation research projects such as OECD BEMUSE, SETH, PKL, BFBT, and TMI-2 have been actively pursued as part of code assessment efforts. The assessment, evaluation and experimental data obtained through international cooperation projects have been registered and maintained in the T/H Databank. Multi-D analyses of APR1400 LBLOCA, DVI Break, SLB, and SGTR have been carried out as a part of application efforts in multi-D safety analysis. GUI based 3D input generator has been developed for user convenience. Operation of the MARS Users Group (MUG) was continued and through MUG, the technology has been transferred to 24 organisations. A set of 4 volumes of user manuals has been compiled and the correction reports for the code errors reported during MARS development have been published
Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes
Energy Technology Data Exchange (ETDEWEB)
Seo, Kwi Hyun; Song, Wan Jung [ENERGEO Inc. Sungnam, (Korea, Republic of); Song, Dong Soo; Byun, Choong Sup [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
2006-07-01
The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences.
Benchmarking Analysis between CONTEMPT and COPATTA Containment Codes
International Nuclear Information System (INIS)
Seo, Kwi Hyun; Song, Wan Jung; Song, Dong Soo; Byun, Choong Sup
2006-01-01
The containment is the requirement that the releases of radioactive materials subsequent to an accident do not result in doses in excess of the values specified in 10 CFR 100. The containment must withstand the pressure and temperature of the DBA(Design Basis Accident) including margin without exceeding the design leakage rate. COPATTA as Bechtel's vendor code is used for the containment pressure and temperature prediction in power uprating project for Kori 3,4 and Yonggwang 1,2 nuclear power plants(NPPs). However, CONTEMPTLT/ 028 is used for calculating the containment pressure and temperatures in equipment qualification project for the same NPPs. During benchmarking analysis between two codes, it is known two codes have model differences. This paper show the performance evaluation results because of the main model differences
Analysis of ATLAS Cold Leg SBLOCA Using SPACE Code
International Nuclear Information System (INIS)
Kang, Doo Hyuk; Suh, Jae Seung; Kim, Se Yun
2012-01-01
SPACE Code has been developed to predict the thermal-hydraulic responses of nuclear steam supply system to the anticipated transients and postulated accidents and adopted advanced physical modeling of two-phase flows, mainly two-fluid, three-field models that comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or non-structured meshes. In this paper, a cold-leg SBLOCA which is the experiment, SB-CL-09, of the ATLAS integral effect test facility during the second domestic stand problem (DSP-02) was analyzed. The results were compared with those of MARS-KS code simulations. The SPACE code with a 1.0 version was released by KHNP in 2012. The analysis has been performed in a desktop PC with Windows 7 environment
FAST: An advanced code system for fast reactor transient analysis
International Nuclear Information System (INIS)
Mikityuk, Konstantin; Pelloni, Sandro; Coddington, Paul; Bubelis, Evaldas; Chawla, Rakesh
2005-01-01
One of the main goals of the FAST project at PSI is to establish a unique analytical code capability for the core and safety analysis of advanced critical (and sub-critical) fast-spectrum systems for a wide range of different coolants. Both static and transient core physics, as well as the behaviour and safety of the power plant as a whole, are studied. The paper discusses the structure of the code system, including the organisation of the interfaces and data exchange. Examples of validation and application of the individual programs, as well as of the complete code system, are provided using studies carried out within the context of designs for experimental accelerator-driven, fast-spectrum systems
Analysis of the anisotropy effects with the AZTRAN code
International Nuclear Information System (INIS)
Xolocostli, V.; Vargas, S.; Gomez, A.; Del Valle, E.
2017-09-01
Among the improvements that are made for the deterministic codes with which nuclear reactors are analyzed, is the implementation of the dispersion anisotropic dispersion section, which can obtain better results. With the current computing technology is possible to carry out these implementations, since the computation time is no longer a considerable problem as in the past. In this paper we analyze some effects of anisotropy in the AZTRAN code, a code that solves the Boltzmann transport equation in one, two and three dimensions at steady state, using the multigroup technique, the nodal method RTN-0 and ordered discrete, which is part of the AZTLAN platform for analysis of nuclear reactors, which is currently under development. The implementation of the anisotropy in the AZTRAN code is one of the latest improvements that have been made to the code, leading to different tests and analyzes regarding the anisotropic dispersion, some as a test with homogeneous fuel assemblies. In the case presented here, the benchmark problem of a fuel assembly type BWR is analyzed, which is part of the Benchmark problem suite for reactor physics study of LWR next generation fuels, proposed by the Committee on Reactor Physics organized by the Japan Atomic Energy Research Institute (JAERI). In this problem the behavior of the infinite multiplication factor (k inf ) is analyzed, as well as the behavior of using odd and even anisotropy approximation with respect to the symmetry in the radial power of the assembly. (Author)
Health effects estimation code development for accident consequence analysis
International Nuclear Information System (INIS)
Togawa, O.; Homma, T.
1992-01-01
As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)
Seismic Analysis Code (SAC): Development, porting, and maintenance within a legacy code base
Savage, B.; Snoke, J. A.
2017-12-01
The Seismic Analysis Code (SAC) is the result of toil of many developers over almost a 40-year history. Initially a Fortran-based code, it has undergone major transitions in underlying bit size from 16 to 32, in the 1980s, and 32 to 64 in 2009; as well as a change in language from Fortran to C in the late 1990s. Maintenance of SAC, the program and its associated libraries, have tracked changes in hardware and operating systems including the advent of Linux in the early 1990, the emergence and demise of Sun/Solaris, variants of OSX processors (PowerPC and x86), and Windows (Cygwin). Traces of these systems are still visible in source code and associated comments. A major concern while improving and maintaining a routinely used, legacy code is a fear of introducing bugs or inadvertently removing favorite features of long-time users. Prior to 2004, SAC was maintained and distributed by LLNL (Lawrence Livermore National Lab). In that year, the license was transferred from LLNL to IRIS (Incorporated Research Institutions for Seismology), but the license is not open source. However, there have been thousands of downloads a year of the package, either source code or binaries for specific system. Starting in 2004, the co-authors have maintained the SAC package for IRIS. In our updates, we fixed bugs, incorporated newly introduced seismic analysis procedures (such as EVALRESP), added new, accessible features (plotting and parsing), and improved the documentation (now in HTML and PDF formats). Moreover, we have added modern software engineering practices to the development of SAC including use of recent source control systems, high-level tests, and scripted, virtualized environments for rapid testing and building. Finally, a "sac-help" listserv (administered by IRIS) was setup for SAC-related issues and is the primary avenue for users seeking advice and reporting bugs. Attempts are always made to respond to issues and bugs in a timely fashion. For the past thirty-plus years
Modeling radiation belt dynamics using a 3-D layer method code
Wang, C.; Ma, Q.; Tao, X.; Zhang, Y.; Teng, S.; Albert, J. M.; Chan, A. A.; Li, W.; Ni, B.; Lu, Q.; Wang, S.
2017-08-01
A new 3-D diffusion code using a recently published layer method has been developed to analyze radiation belt electron dynamics. The code guarantees the positivity of the solution even when mixed diffusion terms are included. Unlike most of the previous codes, our 3-D code is developed directly in equatorial pitch angle (α0), momentum (p), and L shell coordinates; this eliminates the need to transform back and forth between (α0,p) coordinates and adiabatic invariant coordinates. Using (α0,p,L) is also convenient for direct comparison with satellite data. The new code has been validated by various numerical tests, and we apply the 3-D code to model the rapid electron flux enhancement following the geomagnetic storm on 17 March 2013, which is one of the Geospace Environment Modeling Focus Group challenge events. An event-specific global chorus wave model, an AL-dependent statistical plasmaspheric hiss wave model, and a recently published radial diffusion coefficient formula from Time History of Events and Macroscale Interactions during Substorms (THEMIS) statistics are used. The simulation results show good agreement with satellite observations, in general, supporting the scenario that the rapid enhancement of radiation belt electron flux for this event results from an increased level of the seed population by radial diffusion, with subsequent acceleration by chorus waves. Our results prove that the layer method can be readily used to model global radiation belt dynamics in three dimensions.
High-dynamic range compressive spectral imaging by grayscale coded aperture adaptive filtering
Directory of Open Access Journals (Sweden)
Nelson Eduardo Diaz
2015-09-01
Full Text Available The coded aperture snapshot spectral imaging system (CASSI is an imaging architecture which senses the three dimensional informa-tion of a scene with two dimensional (2D focal plane array (FPA coded projection measurements. A reconstruction algorithm takes advantage of the compressive measurements sparsity to recover the underlying 3D data cube. Traditionally, CASSI uses block-un-block coded apertures (BCA to spatially modulate the light. In CASSI the quality of the reconstructed images depends on the design of these coded apertures and the FPA dynamic range. This work presents a new CASSI architecture based on grayscaled coded apertu-res (GCA which reduce the FPA saturation and increase the dynamic range of the reconstructed images. The set of GCA is calculated in a real-time adaptive manner exploiting the information from the FPA compressive measurements. Extensive simulations show the attained improvement in the quality of the reconstructed images when GCA are employed. In addition, a comparison between traditional coded apertures and GCA is realized with respect to noise tolerance.
Working research codes into fluid dynamics education: a science gateway approach
Mason, Lachlan; Hetherington, James; O'Reilly, Martin; Yong, May; Jersakova, Radka; Grieve, Stuart; Perez-Suarez, David; Klapaukh, Roman; Craster, Richard V.; Matar, Omar K.
2017-11-01
Research codes are effective for illustrating complex concepts in educational fluid dynamics courses, compared to textbook examples, an interactive three-dimensional visualisation can bring a problem to life! Various barriers, however, prevent the adoption of research codes in teaching: codes are typically created for highly-specific `once-off' calculations and, as such, have no user interface and a steep learning curve. Moreover, a code may require access to high-performance computing resources that are not readily available in the classroom. This project allows academics to rapidly work research codes into their teaching via a minimalist `science gateway' framework. The gateway is a simple, yet flexible, web interface allowing students to construct and run simulations, as well as view and share their output. Behind the scenes, the common operations of job configuration, submission, monitoring and post-processing are customisable at the level of shell scripting. In this talk, we demonstrate the creation of an example teaching gateway connected to the Code BLUE fluid dynamics software. Student simulations can be run via a third-party cloud computing provider or a local high-performance cluster. EPSRC, UK, MEMPHIS program Grant (EP/K003976/1), RAEng Research Chair (OKM).
Performance Analysis of CRC Codes for Systematic and Nonsystematic Polar Codes with List Decoding
Directory of Open Access Journals (Sweden)
Takumi Murata
2018-01-01
Full Text Available Successive cancellation list (SCL decoding of polar codes is an effective approach that can significantly outperform the original successive cancellation (SC decoding, provided that proper cyclic redundancy-check (CRC codes are employed at the stage of candidate selection. Previous studies on CRC-assisted polar codes mostly focus on improvement of the decoding algorithms as well as their implementation, and little attention has been paid to the CRC code structure itself. For the CRC-concatenated polar codes with CRC code as their outer code, the use of longer CRC code leads to reduction of information rate, whereas the use of shorter CRC code may reduce the error detection probability, thus degrading the frame error rate (FER performance. Therefore, CRC codes of proper length should be employed in order to optimize the FER performance for a given signal-to-noise ratio (SNR per information bit. In this paper, we investigate the effect of CRC codes on the FER performance of polar codes with list decoding in terms of the CRC code length as well as its generator polynomials. Both the original nonsystematic and systematic polar codes are considered, and we also demonstrate that different behaviors of CRC codes should be observed depending on whether the inner polar code is systematic or not.
Qualification of ARROTTA code for LWR accident analysis
International Nuclear Information System (INIS)
Huang, P.-H.; Peng, K.Y.; Lin, W.-C.; Wu, J.-Y.
2004-01-01
This paper presents the qualification efforts performed by TPC and INER for the 3-D spatial kinetics code ARROTTA for LWR core transient analysis. TPC and INER started a joint 5 year project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multi-dimensional kinetics calculations such as rod ejection for PWR and rod drop for BWR. To qualify ARROTTA for analysis of FSAR licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicated that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include NEACRP rod ejection problem, 3-D LMW LWR rod withdrawal/insertion problem, and 3-D LRA BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared to other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multi-dimensional core transient for LWRs. (author)
DEFF Research Database (Denmark)
Hundebøll, Martin; Pedersen, Morten Videbæk; Roetter, Daniel Enrique Lucani
2014-01-01
This work studies the potential and impact of the FRANC network coding protocol for delivering high quality Dynamic Adaptive Streaming over HTTP (DASH) in wireless networks. Although DASH aims to tailor the video quality rate based on the available throughput to the destination, it relies...
Code portability and data management considerations in the SAS3D LMFBR accident-analysis code
International Nuclear Information System (INIS)
Dunn, F.E.
1981-01-01
The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available
Development of 3-dimensional neutronics kinetics analysis code for CANDU-PHWR
International Nuclear Information System (INIS)
Kim, M. W.; Kim, C. H.; Hong, I. S.
2005-02-01
The followings are the major contents and scope of the research : development of kinetics power calculation module, formulation of space-dependent neutron transient analysis - implementation of 3-D and 2-G unified nodal method, verification of the kinetics module by benchmark problem - 3-D PHWR kinetics benchmark problem suggested by AECL, reactor trip simulation by shutdown system 1 in Wolsong unit 2. Development of a dynamic linked library code, SCAN D LL, for the coupled calculation with RELAP-CANDU : modeling of shutdown system 1, development of automatic shutdown module - automatic trip module based on rate log power control logic, automatic insertion of shutdown system 1. Development of a link code for coupled calculation - development of SCAN D LL(windows version), verification of coupled code by - 40% reactor inlet header break LOCA power pulse, 100% reactor outlet header break LOCA power pulse, 50% pump suction break LOCA power pulse
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
IllinoisGRMHD: an open-source, user-friendly GRMHD code for dynamical spacetimes
International Nuclear Information System (INIS)
Etienne, Zachariah B; Paschalidis, Vasileios; Haas, Roland; Mösta, Philipp; Shapiro, Stuart L
2015-01-01
In the extreme violence of merger and mass accretion, compact objects like black holes and neutron stars are thought to launch some of the most luminous outbursts of electromagnetic and gravitational wave energy in the Universe. Modeling these systems realistically is a central problem in theoretical astrophysics, but has proven extremely challenging, requiring the development of numerical relativity codes that solve Einstein's equations for the spacetime, coupled to the equations of general relativistic (ideal) magnetohydrodynamics (GRMHD) for the magnetized fluids. Over the past decade, the Illinois numerical relativity (ILNR) group's dynamical spacetime GRMHD code has proven itself as a robust and reliable tool for theoretical modeling of such GRMHD phenomena. However, the code was written ‘by experts and for experts’ of the code, with a steep learning curve that would severely hinder community adoption if it were open-sourced. Here we present IllinoisGRMHD, which is an open-source, highly extensible rewrite of the original closed-source GRMHD code of the ILNR group. Reducing the learning curve was the primary focus of this rewrite, with the goal of facilitating community involvement in the code's use and development, as well as the minimization of human effort in generating new science. IllinoisGRMHD also saves computer time, generating roundoff-precision identical output to the original code on adaptive-mesh grids, but nearly twice as fast at scales of hundreds to thousands of cores. (paper)
A proposed framework for computational fluid dynamics code calibration/validation
International Nuclear Information System (INIS)
Oberkampf, W.L.
1993-01-01
The paper reviews the terminology and methodology that have been introduced during the last several years for building confidence n the predictions from Computational Fluid Dynamics (CID) codes. Code validation terminology developed for nuclear reactor analyses and aerospace applications is reviewed and evaluated. Currently used terminology such as ''calibrated code,'' ''validated code,'' and a ''validation experiment'' is discussed along with the shortcomings and criticisms of these terms. A new framework is proposed for building confidence in CFD code predictions that overcomes some of the difficulties of past procedures and delineates the causes of uncertainty in CFD predictions. Building on previous work, new definitions of code verification and calibration are proposed. These definitions provide more specific requirements for the knowledge level of the flow physics involved and the solution accuracy of the given partial differential equations. As part of the proposed framework, categories are also proposed for flow physics research, flow modeling research, and the application of numerical predictions. The contributions of physical experiments, analytical solutions, and other numerical solutions are discussed, showing that each should be designed to achieve a distinctively separate purpose in building confidence in accuracy of CFD predictions. A number of examples are given for each approach to suggest methods for obtaining the highest value for CFD code quality assurance
Cherubin, S.; Agosta, G.
2018-01-01
We present LIBVERSIONINGCOMPILER, a C++ library designed to support the dynamic generation of multiple versions of the same compute kernel in a HPC scenario. It can be used to provide continuous optimization, code specialization based on the input data or on workload changes, or otherwise to dynamically adjust the application, without the burden of a full dynamic compiler. The library supports multiple underlying compilers but specifically targets the LLVM framework. We also provide examples of use, showing the overhead of the library, and providing guidelines for its efficient use.
THYDE-P2 code: RCS (reactor-coolant system) analysis code
International Nuclear Information System (INIS)
Asahi, Yoshiro; Hirano, Masashi; Sato, Kazuo
1986-12-01
THYDE-P2, being characterized by the new thermal-hydraulic network model, is applicable to analysis of RCS behaviors in response to various disturbances including LB (large break)-LOCA(loss-of-coolant accident). In LB-LOCA analysis, THYDE-P2 is capable of through calculation from its initiation to complete reflooding of the core without an artificial change in the methods and models. The first half of the report is the description of the methods and models for use in the THYDE-P2 code, i.e., (1) the thermal-hydraulic network model, (2) the various RCS components models, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the user's mannual for the THYDE-P2 code (version SV04L08A) containing items; (1) the program control (2) the input requirements, (3) the execution of THYDE-P2 job, (4) the output specifications and (5) the sample problem to demonstrate capability of the thermal-hydraulic network model, among other things. (author)
Installation of aerosol behavior model into multi-dimensional thermal hydraulic analysis code AQUA
International Nuclear Information System (INIS)
Kisohara, Naoyuki; Yamaguchi, Akira
1997-12-01
The safety analysis of FBR plant system for sodium leak phenomena needs to evaluate the deposition of the aerosol particle to the components in the plant, the chemical reaction of aerosol to humidity in the air and the effect of the combustion heat through aerosol to the structural component. For this purpose, ABC-INTG (Aerosol Behavior in Containment-INTeGrated Version) code has been developed and used until now. This code calculates aerosol behavior in the gas area of uniform temperature and pressure by 1 cell-model. Later, however, more detailed calculation of aerosol behavior requires the installation of aerosol model into multi-cell thermal hydraulic analysis code AQUA. AQUA can calculate the carrier gas flow, temperature and the distribution of the aerosol spatial concentration. On the other hand, ABC-INTG can calculate the generation, deposition to the wall and flower, agglomeration of aerosol particle and figure out the distribution of the aerosol particle size. Thus, the combination of these two codes enables to deal with aerosol model coupling the distribution of the aerosol spatial concentration and that of the aerosol particle size. This report describes aerosol behavior model, how to install the aerosol model to AQUA and new subroutine equipped to the code. Furthermore, the test calculations of the simple structural model were executed by this code, appropriate results were obtained. Thus, this code has prospect to predict aerosol behavior by the introduction of coupling analysis with multi-dimensional gas thermo-dynamics for sodium combustion evaluation. (J.P.N.)
Static Analysis for Dynamic XML
DEFF Research Database (Denmark)
Christensen, Aske Simon; Møller, Anders; Schwartzbach, Michael Ignatieff
2002-01-01
We describe the summary graph lattice for dataflow analysis of programs that dynamically construct XML documents. Summary graphs have successfully been used to provide static guarantees in the JWIG language for programming interactive Web services. In particular, the JWIG compiler is able to check...
Single stock dynamics on high-frequency data: from a compressed coding perspective.
Directory of Open Access Journals (Sweden)
Hsieh Fushing
Full Text Available High-frequency return, trading volume and transaction number are digitally coded via a nonparametric computing algorithm, called hierarchical factor segmentation (HFS, and then are coupled together to reveal a single stock dynamics without global state-space structural assumptions. The base-8 digital coding sequence, which is capable of revealing contrasting aggregation against sparsity of extreme events, is further compressed into a shortened sequence of state transitions. This compressed digital code sequence vividly demonstrates that the aggregation of large absolute returns is the primary driving force for stimulating both the aggregations of large trading volumes and transaction numbers. The state of system-wise synchrony is manifested with very frequent recurrence in the stock dynamics. And this data-driven dynamic mechanism is seen to correspondingly vary as the global market transiting in and out of contraction-expansion cycles. These results not only elaborate the stock dynamics of interest to a fuller extent, but also contradict some classical theories in finance. Overall this version of stock dynamics is potentially more coherent and realistic, especially when the current financial market is increasingly powered by high-frequency trading via computer algorithms, rather than by individual investors.
Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa [Japan Atomic Energy Agency, Nuclear Safety Research Center, Tokai, Ibaraki (Japan); Saitou, Hiroaki [ITOCHU Techno-Solutions Corporation, Tokyo (Japan)
2013-10-15
A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)
Input/output manual of light water reactor fuel analysis code FEMAXI-7 and its related codes
International Nuclear Information System (INIS)
Suzuki, Motoe; Udagawa, Yutaka; Nagase, Fumihisa; Saitou, Hiroaki
2013-10-01
A light water reactor fuel analysis code FEMAXI-7 has been developed, as an extended version from the former version FEMAXI-6, for the purpose of analyzing the fuel behavior in normal conditions and in anticipated transient conditions. Numerous functional improvements and extensions have been incorporated in FEMAXI-7, which are fully disclosed in the code model description published in the form of another JAEA-Data/Code report. The present manual, which is the very counterpart of this description document, gives detailed explanations of files and operation method of FEMAXI-7 code and its related codes, methods of input/output, sample Input/Output, methods of source code modification, subroutine structure, and internal variables in a specific manner in order to facilitate users to perform fuel analysis by FEMAXI-7. (author)
Computer code for qualitative analysis of gamma-ray spectra
International Nuclear Information System (INIS)
Yule, H.P.
1979-01-01
Computer code QLN1 provides complete analysis of gamma-ray spectra observed with Ge(Li) detectors and is used at both the National Bureau of Standards and the Environmental Protection Agency. It locates peaks, resolves multiplets, identifies component radioisotopes, and computes quantitative results. The qualitative-analysis (or component identification) algorithms feature thorough, self-correcting steps which provide accurate isotope identification in spite of errors in peak centroids, energy calibration, and other typical problems. The qualitative-analysis algorithm is described in this paper
Carles, Guillem; Ferran, Carme; Carnicer, Artur; Bosch, Salvador
2012-01-01
A computational imaging system based on wavefront coding is presented. Wavefront coding provides an extension of the depth-of-field at the expense of a slight reduction of image quality. This trade-off results from the amount of coding used. By using spatial light modulators, a flexible coding is achieved which permits it to be increased or decreased as needed. In this paper a computational method is proposed for evaluating the output of a wavefront coding imaging system equipped with a spatial light modulator, with the aim of thus making it possible to implement the most suitable coding strength for a given scene. This is achieved in an unsupervised manner, thus the whole system acts as a dynamically selfadaptable imaging system. The program presented here controls the spatial light modulator and the camera, and also processes the images in a synchronised way in order to implement the dynamic system in real time. A prototype of the system was implemented in the laboratory and illustrative examples of the performance are reported in this paper. Program summaryProgram title: DynWFC (Dynamic WaveFront Coding) Catalogue identifier: AEKC_v1_0 Program summary URL:http://cpc.cs.qub.ac.uk/summaries/AEKC_v1_0.html Program obtainable from: CPC Program Library, Queen's University, Belfast, N. Ireland Licensing provisions: Standard CPC licence, http://cpc.cs.qub.ac.uk/licence/licence.html No. of lines in distributed program, including test data, etc.: 10 483 No. of bytes in distributed program, including test data, etc.: 2 437 713 Distribution format: tar.gz Programming language: Labview 8.5 and NI Vision and MinGW C Compiler Computer: Tested on PC Intel ® Pentium ® Operating system: Tested on Windows XP Classification: 18 Nature of problem: The program implements an enhanced wavefront coding imaging system able to adapt the degree of coding to the requirements of a specific scene. The program controls the acquisition by a camera, the display of a spatial light modulator
Moderator 3-D Thermalhydraulic Analysis Using MODTURCCLAS Code
International Nuclear Information System (INIS)
Kim, Hyoung Tae; Yoon, Churl; Park, Joo Hwan
2008-12-01
For the moderator subcooling analysis of the refurbished Wolsong NPP (Nuclear Power Plant) Unit 1, the 3-D moderator thermalhydraulic calculations were preliminarily conducted from September to October in 2008, using the preliminary thermalhydraulic analysis results and the MODTURC C LAS (MODerator TURbulent Circulation Co-Located Advanced Solutions) Ver.2.9-IST, which was developed and validated by OPG (Ontario Power Generation) in Canada. The present report consists of the steady-state calculation and transient calculation. First the grid structure, main input parameters, and boundary conditions needed for the steady-state calculation are produced and the steady-state results are obtained. These steady-state results are used for the initial conditions for the transient analysis during a LOCA. The moderator analysis results during the steady-state calculation show the quasy-steady state behavior, because the thermalhydraulic behavior are fluctuating although all boundary conditions are constant. In the transient calculations, based on the present thermalhydraulic analysis results, 3-D thermalhydraulic behavior and moderator subcooling are predicted for the accident scenarios of reactor inlet header 35% and 40% breaks, outlet header 100% break, and pump suction 80% break, subsequent with loss of Class IV power. In the previous moderator analysis for Wolsong NPP Unit 2,3,4 a PHOENICS code was used, which is different from the MODTURC C LAS code used for the analysis of Wolsong NPP Unit 1. However, the moderator subcooling analysis results by these two codes are qualitatively similar. The minimum subcooling for RIH 40% break of Wolsong NPP Unit 1 is 17 .deg. C which is larger than 13 .deg. C for RIH 35% break of Wolsong NPP Unit 2,3,4. Therefore, it is concluded that the refurbished Wolsong NPP Unit 1 satisfies the channel integrity criteria based on the higher subcooling margin compared with that of Wolsong NPP Unit 2,3,4
Development of thermal hydraulic analysis code for IHX of FBR
International Nuclear Information System (INIS)
Kumagai, Hiromichi; Naohara, Nobuyuki
1991-01-01
In order to obtain flow resistance correlations for thermal-hydrauric analysis code concerned with an intermediate heat exchanger (IHX) of FBR, the hydraulic experiment by air was carried out through a bundle of tubes arranged in an in-line and staggard fashion. The main results are summarized as follows. (1) On pressure loss per unit length of a tube bundle, which is densely a regular triangle arrangement, the in-line fashion is almost the same as the staggard one. (2) In case of 30deg sector model for IHX tube bundle, pressure loss is 1/3 in comparison with the in-line or staggard arrangement. (3) By this experimental data, flow resistance correlations for thermalhydrauric analysis code are obtained. (author)
TRAC-CFD code integration and its application to containment analysis
International Nuclear Information System (INIS)
Tahara, M.; Arai, K.; Oikawa, H.
2004-01-01
Several safety systems utilizing natural driving force have been recently adopted for operating reactors, or applied to next-generation reactor design. Examples of these safety systems are the Passive Containment Cooling System (PCCS) and the Drywell Cooler (DWC) for removing decay heat, and the Passive Auto-catalytic Recombiner (PAR) for removing flammable gas in reactor containment during an accident. DWC is used in almost all Boiling Water Reactors (BWR) in service. PAR has been introduced for some reactors in Europe and will be introduced for Japanese reactors. PCCS is a safety device of next-generation BWR. The functional mechanism of these safety systems is closely related to the transient of the thermal-hydraulic condition of the containment atmosphere. The performance depends on the containment atmospheric condition, which is eventually affected by the mass and energy changes caused by the safety system. Therefore, the thermal fluid dynamics in the containment vessel should be appropriately considered in detail to properly estimate the performance of these systems. A computational fluid dynamics (CFD) code is useful for evaluating detailed thermal hydraulic behavior related to this equipment. However, it also requires a considerable amount of computational resources when it is applied to whole containment system transient analysis. The paper describes the method and structure of the integrated analysis tool, and discusses the results of its application to the start-up behavior analysis of a containment cooling system, a drywell local cooler. The integrated analysis code was developed and applied to estimate the DWC performance during a severe accident. The integrated analysis tool is composed of three codes, TRAC-PCV, CFD-DW and TRAC-CC, and analyzes the interaction of the natural convection and steam condensation of the DWC as well as analyzing the thermal hydraulic transient behavior of the containment vessel during a severe accident in detail. The
Source Code Analysis Laboratory (SCALe) for Energy Delivery Systems
2010-12-01
technical competence for the type of tests and calibrations SCALe undertakes. Testing and calibration laboratories that comply with ISO / IEC 17025 ...and exec t [ ISO / IEC 2005]. f a software system indicates that the SCALe analysis di by a CERT secure coding standard. Successful conforma antees that...to be more secure than non- systems. However, no study has yet been performed to p t ssment in accordance with ISO / IEC 17000: “a demonstr g to a
Validation of OPERA3D PCMI Analysis Code
Energy Technology Data Exchange (ETDEWEB)
Jeun, Ji Hoon; Choi, Jae Myung; Yoo, Jong Sung [KEPCO Nuclear Fuel Co., Daejeon (Korea, Republic of); Cheng, G.; Sim, K. S.; Chassie, Girma [Candu Energy INC.,Ontario (Canada)
2013-10-15
This report will describe introduction of validation of OPERA3D code, and validation results that are directly related with PCMI phenomena. OPERA3D was developed for the PCMI analysis and validated using the in-pile measurement data. Fuel centerline temperature and clad strain calculation results shows close expectations with measurement data. Moreover, 3D FEM fuel model of OPERA3D shows slight hour glassing behavior of fuel pellet in contact case. Further optimization will be conducted for future application of OPERA3D code. Nuclear power plant consists of many complicated systems, and one of the important objects of all the systems is maintaining nuclear fuel integrity. However, it is inevitable to experience PCMI (Pellet Cladding Mechanical Interaction) phenomena at current operating reactors and next generation reactors for advanced safety and economics as well. To evaluate PCMI behavior, many studies are on-going to develop 3-dimensional fuel performance evaluation codes. Moreover, these codes are essential to set the safety limits for the best estimated PCMI phenomena aimed for high burnup fuel.
LACEwING: A New Moving Group Analysis Code
Energy Technology Data Exchange (ETDEWEB)
Riedel, Adric R. [Department of Astronomy, California Institute of Technology, Pasadena, CA 91125 (United States); Blunt, Sarah C.; Faherty, Jacqueline K. [Department of Astrophysics, American Museum of Natural History, New York, NY 10024 (United States); Lambrides, Erini L. [Department of Physics and Astronomy, Johns Hopkins University, Baltimore, MD 21218 (United States); Rice, Emily L. [Department of Engineering Science and Physics, The College of Staten Island, Staten Island, NY 10314 (United States); Cruz, Kelle L., E-mail: arr@astro.caltech.edu [Department of Physics and Astronomy, Hunter College, New York, NY 10065 (United States)
2017-03-01
We present a new nearby young moving group (NYMG) kinematic membership analysis code, LocAting Constituent mEmbers In Nearby Groups (LACEwING), a new Catalog of Suspected Nearby Young Stars, a new list of bona fide members of moving groups, and a kinematic traceback code. LACEwING is a convergence-style algorithm with carefully vetted membership statistics based on a large numerical simulation of the Solar Neighborhood. Given spatial and kinematic information on stars, LACEwING calculates membership probabilities in 13 NYMGs and three open clusters within 100 pc. In addition to describing the inputs, methods, and products of the code, we provide comparisons of LACEwING to other popular kinematic moving group membership identification codes. As a proof of concept, we use LACEwING to reconsider the membership of 930 stellar systems in the Solar Neighborhood (within 100 pc) that have reported measurable lithium equivalent widths. We quantify the evidence in support of a population of young stars not attached to any NYMGs, which is a possible sign of new as-yet-undiscovered groups or of a field population of young stars.
SHEAN (Simplified Human Error Analysis code) and automated THERP
International Nuclear Information System (INIS)
Wilson, J.R.
1993-01-01
One of the most widely used human error analysis tools is THERP (Technique for Human Error Rate Prediction). Unfortunately, this tool has disadvantages. The Nuclear Regulatory Commission, realizing these drawbacks, commissioned Dr. Swain, the author of THERP, to create a simpler, more consistent tool for deriving human error rates. That effort produced the Accident Sequence Evaluation Program Human Reliability Analysis Procedure (ASEP), which is more conservative than THERP, but a valuable screening tool. ASEP involves answering simple questions about the scenario in question, and then looking up the appropriate human error rate in the indicated table (THERP also uses look-up tables, but four times as many). The advantages of ASEP are that human factors expertise is not required, and the training to use the method is minimal. Although not originally envisioned by Dr. Swain, the ASEP approach actually begs to be computerized. That WINCO did, calling the code SHEAN, for Simplified Human Error ANalysis. The code was done in TURBO Basic for IBM or IBM-compatible MS-DOS, for fast execution. WINCO is now in the process of comparing this code against THERP for various scenarios. This report provides a discussion of SHEAN
Essentials of applied dynamic analysis
Jia, Junbo
2014-01-01
This book presents up-to-date knowledge of dynamic analysis in engineering world. To facilitate the understanding of the topics by readers with various backgrounds, general principles are linked to their applications from different angles. Special interesting topics such as statistics of motions and loading, damping modeling and measurement, nonlinear dynamics, fatigue assessment, vibration and buckling under axial loading, structural health monitoring, human body vibrations, and vehicle-structure interactions etc., are also presented. The target readers include industry professionals in civil, marine and mechanical engineering, as well as researchers and students in this area.
Development of the SCHAMBETA code for scoping analysis of HCDA
Energy Technology Data Exchange (ETDEWEB)
Suk, Soo Dong; Hahn, D. H
2000-06-01
A computer code, SCHAMBETA(Scoping Code for HCDA Analysis using Modified Bethe-Tait Method), is developed to investigate the core disassembly process following a meltdown accident in the framework of a mofified Bethe-Tait method as part of the scoping analysis work to demonstrate the inherent safety of conceptual designs of Korea Advanced Liquid Metal Reactor(KALIMER), A 150 Mwe pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. The methodologies adopted in the code ared particularly useful to perform various parametric studies for better understanding of core disassembly process of liquid metal fast reactors as well as to estimate upper-limit values of the energy release resulting from a power excursion. In the SCHAMBETA code, the core kinetics and hydraulic behavior of the KALIMER is followed over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion, starting at the time that the sodium-voided core reaches the melting temperature of the metallic fuels. For this purpose, the equations of state of pressure-energy density relationship are derived for the saturated-vapor as well as the solid liquid of metallic uranium fuel, and implemenmted into the formulations of the disassembly reactivity. Mathematical formulations are then developed, in the framework of Modified Bethe-Tait method, in a form relevant to utilize the improved equations of state as well as to consider Doppler effects, for scoping analysis of the super-prompt-critical power excursions driven by a specified rate of reactivity insertion.
Site-specific Probabilistic Analysis of DCGLs Using RESRAD Code
Energy Technology Data Exchange (ETDEWEB)
Kim, Jeongju; Yoon, Suk Bon; Sohn, Wook [KHNP CRI, Daejeon (Korea, Republic of)
2016-10-15
In general, DCGLs can be conservative (screening DCGL) if they do not take into account site specific factors. Use of such conservative DCGLs can lead to additional remediation that would not be required if the effort was made to develop site-specific DCGLs. Therefore, the objective of this work is to provide an example on the use of the RESRAD 6.0 probabilistic (site-specific) dose analysis to compare with the screening DCGL. Site release regulations state that a site will be considered acceptable for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group of less than the site release criteria, for example 0.25 mSv per year in U.S. Utilities use computer dose modeling codes to establish an acceptable level of contamination, the derived concentration guideline level (DCGL) that will meet this regulatory limit. Since the DCGL value is the principal measure of residual radioactivity, it is critical to understand the technical basis of these dose modeling codes. The objective this work was to provide example on nuclear power plant decommissioning dose analysis in a probabilistic analysis framework. The focus was on the demonstration of regulatory compliance for surface soil contamination using the RESRAD 6.0 code. Both the screening and site-specific probabilistic dose analysis methodologies were examined. Example analyses performed with the screening probabilistic dose analysis confirmed the conservatism of the NRC screening values and indicated the effectiveness of probabilistic dose analysis in reducing the conservatism in DCGL derivation.
International Nuclear Information System (INIS)
Ambrosini, W.; D'Auria, F.; Di Marco, P.; Fantappie, G.; Giot, G.; Emmerechts, D.; Seynhaeve, J.M.; Zhang, J.
1989-11-01
The experimental data of flooding and CCFL in the fuel element top nozzle area collected at the University of Hannover have been analyzed with RELAP5/MOD2 and CATHARE V.1.3 codes. Preliminary sensitivity calculations have been performed to evaluate the influence of various parameters and code options on the results. However, an a priori rational assessment procedure has been performed for those parameters non specific in experimental data (e.g. energy loss coefficients in flow restrictions). This procedure is based on single phase flow pressure drops and no further tuning has been performed to fit experimental data. The reported experimental data and some others demonstrate the complex relation-ship among the involved physical quantities (film thickness, pressure drop etc.) even in a simple geometrical condition with well defined boundary conditions. In the application of the two advanced codes to the selected CCFL experiments it appears that sophisticated models do not simulate satisfactorily the measured phenomena mainly when situations similar to nuclear reactors are dealt with (rod bundles). This result should be evaluated considering that: - dimensional phenomena occurring in flooding experiments are not well reproducible with one dimensional models implemented in the two codes; - a rational and reproducible procedure has been used to fix some boundary conditions (K-tuning); there is the evidence that more tuning can be used to get results closer to the experimental ones in each specific situation; - the uncertainty bands in measured experimental results are not (entirely) specified. The work performed demonstrated that further applications to CCFL experiments of present codes appear to be unuseful. New models should be tested and implemented before any attempt to reproduce CCFL in experimental facilities by system codes
Analysis of transients for NPP with VVER-440 using the code SiTAP
International Nuclear Information System (INIS)
Kalinenko, V.
1994-06-01
The report contains analysis of transients ''Loop connection'' and ''Steam generator tube rupture'' for nuclear power plants (NPP) with VVER-440. To obtain more detailed information about NPP's dynamic characteristics, various variants of initial and boundary conditions are considerd. Calculation of these transients was performed using the SiTAP code developed at the Nuclear Safety Institute of the Russian Research Centre ''Kurchatov Institute''. SiTAP code is a multifunctional computer tool for fast analysis of transient and accidental processes of VVER type reactors for engineers working in the field of NPP dynamics. SiTAP can be used form comparative analysis of several variants of accident scenarios to find out the conditions leading to most serious consequences from a safety point of view. In such cases, additional analyses using best-estimate codes should be carried out. The results of SiTAP for a faulty loop connection leading to a boron dilution accident are intended to be used as boundary conditions for a more detailed anlaysis with the aid of the three-dimensional reactor core model DYN3D, developed in the Research Centre Rossendorf for the simulation of reactivity initiated accidents. (orig.)
Nonlinear analysis of dynamic signature
Rashidi, S.; Fallah, A.; Towhidkhah, F.
2013-12-01
Signature is a long trained motor skill resulting in well combination of segments like strokes and loops. It is a physical manifestation of complex motor processes. The problem, generally stated, is that how relative simplicity in behavior emerges from considerable complexity of perception-action system that produces behavior within an infinitely variable biomechanical and environmental context. To solve this problem, we present evidences which indicate that motor control dynamic in signing process is a chaotic process. This chaotic dynamic may explain a richer array of time series behavior in motor skill of signature. Nonlinear analysis is a powerful approach and suitable tool which seeks for characterizing dynamical systems through concepts such as fractal dimension and Lyapunov exponent. As a result, they can be analyzed in both horizontal and vertical for time series of position and velocity. We observed from the results that noninteger values for the correlation dimension indicates low dimensional deterministic dynamics. This result could be confirmed by using surrogate data tests. We have also used time series to calculate the largest Lyapunov exponent and obtain a positive value. These results constitute significant evidence that signature data are outcome of chaos in a nonlinear dynamical system of motor control.
THYDE-NEU: Nuclear reactor system analysis code
International Nuclear Information System (INIS)
Asahi, Yoshiro
2002-03-01
THYDE-NEU is applicable not only to transient analyses, but also to steady state analyses of nuclear reactor systems (NRSs). In a steady state analysis, the code generates a solution satisfying the transient equations without external disturbances. In a transient analysis, the code calculates temporal NRS behaviors in response to various external disturbances in such a way that mass and energy of the coolant as well as the number of neutrons conserve. The first half of the report is the description of the methods and models for use in the THYDE-NEU code, i.e., (1) the thermal-hydraulic network model, (2) the spatial kinetics model, (3) the heat sources in fuel, (4) the heat transfer correlations, (5) the mechanical behavior of clad and fuel, and (6) the steady state adjustment. The second half of the report is the users' mannual containing the items; (1) the program control, (2) the input requirements, (3) the execution of THYDE-NEU jobs, (4) the output specifications and (5) the sample calculation. (author)
Development of a neutronic analysis code using data from Monju
International Nuclear Information System (INIS)
Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.
2015-01-01
In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)
Comparative study of Thermal Hydraulic Analysis Codes for Pressurized Water Reactors
Energy Technology Data Exchange (ETDEWEB)
Kim, Yang Hoon; Jang, Mi Suk; Han, Kee Soo [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)
2015-05-15
Various codes are used for the thermal hydraulic analysis of nuclear reactors. The use of some codes among these is limited by user and some codes are not even open to general person. Thus, the use of alternative code is considered for some analysis. In this study, simple thermal hydraulic behaviors are analyzed using three codes to show that alternative codes are possible for the analysis of nuclear reactors. We established three models of the simple u-tube manometer using three different codes. RELAP5 (Reactor Excursion and Leak Analysis Program), SPACE (Safety and Performance Analysis CodE for nuclear power Plants), GOTHIC (Generation of Thermal Hydraulic Information for Containments) are selected for this analysis. RELAP5 is widely used codes for the analysis of system behavior of PWRs. SPACE has been developed based on RELAP5 for the analysis of system behavior of PWRs and licensing of the code is in progress. And GOTHIC code also has been widely used for the analysis of thermal hydraulic behavior in the containment system. The internal behavior of u-tube manometer was analyzed by RELAP5, SPACE and GOTHIC codes. The general transient behavior was similar among 3 codes. However, the stabilized status of the transient period analyzed by REPAP5 was different from the other codes. It would be resulted from the different physical models used in the other codes, which is specialized for the multi-phase thermal hydraulic behavior analysis.
TRAWA, a transient analysis code for water reactions
International Nuclear Information System (INIS)
Rajamaeki, M.
1976-06-01
TRAWA is a transient analysis code for water reactors. It solves the two-group neutron diffusion equations simultaneously with the heat conduction equations and the two-phase hydraulic equations for one or more channels. At most one-dimensional submodels are used. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermohydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channel and risers with two-phase flow and of pump lines with incompressible flow. Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. Various transients can be calculated by applying external disturbances. The code is extensively supplied by input and output capabilities. TRAWA is written in FORTRAN V for UNIVAC 1108 computer. (author)
Beam Optics Analysis - An Advanced 3D Trajectory Code
International Nuclear Information System (INIS)
Ives, R. Lawrence; Bui, Thuc; Vogler, William; Neilson, Jeff; Read, Mike; Shephard, Mark; Bauer, Andrew; Datta, Dibyendu; Beal, Mark
2006-01-01
Calabazas Creek Research, Inc. has completed initial development of an advanced, 3D program for modeling electron trajectories in electromagnetic fields. The code is being used to design complex guns and collectors. Beam Optics Analysis (BOA) is a fully relativistic, charged particle code using adaptive, finite element meshing. Geometrical input is imported from CAD programs generating ACIS-formatted files. Parametric data is inputted using an intuitive, graphical user interface (GUI), which also provides control of convergence, accuracy, and post processing. The program includes a magnetic field solver, and magnetic information can be imported from Maxwell 2D/3D and other programs. The program supports thermionic emission and injected beams. Secondary electron emission is also supported, including multiple generations. Work on field emission is in progress as well as implementation of computer optimization of both the geometry and operating parameters. The principle features of the program and its capabilities are presented
Development Of The Computer Code For Comparative Neutron Activation Analysis
International Nuclear Information System (INIS)
Purwadi, Mohammad Dhandhang
2001-01-01
The qualitative and quantitative chemical analysis with Neutron Activation Analysis (NAA) is an importance utilization of a nuclear research reactor, and this should be accelerated and promoted in application and its development to raise the utilization of the reactor. The application of Comparative NAA technique in GA Siwabessy Multi Purpose Reactor (RSG-GAS) needs special (not commercially available yet) soft wares for analyzing the spectrum of multiple elements in the analysis at once. The application carried out using a single spectrum software analyzer, and comparing each result manually. This method really degrades the quality of the analysis significantly. To solve the problem, a computer code was designed and developed for comparative NAA. Spectrum analysis in the code is carried out using a non-linear fitting method. Before the spectrum analyzed, it was passed to the numerical filter which improves the signal to noise ratio to do the deconvolution operation. The software was developed using the G language and named as PASAN-K The testing result of the developed software was benchmark with the IAEA spectrum and well operated with less than 10 % deviation
Energy Technology Data Exchange (ETDEWEB)
Nichols, B. D.; Mueller, C.; Necker, G. A.; Travis, J. R.; Spore, J. W.; Lam, K. L.; Royl, P.; Wilson, T. L.
1998-10-01
Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containment and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included. Volume III
Security Enhancement for Multicast over Internet of Things by Dynamically Constructed Fountain Codes
Directory of Open Access Journals (Sweden)
Qinghe Du
2018-01-01
Full Text Available The Internet of Things (IoT is expected to accommodate every object which exists in this world or likely to exist in the near future. The enormous scale of the objects is challenged by big security concerns, especially for common information dissemination via multicast services, where the reliability assurance for multiple multicast users at the cost of increasing redundancy and/or retransmissions also benefits eavesdroppers in successfully decoding the overheard signals. The objective of this work is to address the security challenge present in IoT multicast applications. Specifically, with the presence of the eavesdropper, an adaptive fountain code design is proposed in this paper to enhance the security for multicast in IoT. The main novel features of the proposed scheme include two folds: (i dynamical encoding scheme which can effectively decrease intercept probability at the eavesdropper; (ii increasing the transmission efficiency compared with the conventional nondynamical design. The analysis and simulation results show that the proposed scheme can effectively enhance information security while achieving higher transmission efficiency with a little accredited complexity, thus facilitating the secured wireless multicast transmissions over IoT.
Dynamic Analysis of a Pendulum Dynamic Automatic Balancer
Directory of Open Access Journals (Sweden)
Jin-Seung Sohn
2007-01-01
Full Text Available The automatic dynamic balancer is a device to reduce the vibration from unbalanced mass of rotors. Instead of considering prevailing ball automatic dynamic balancer, pendulum automatic dynamic balancer is analyzed. For the analysis of dynamic stability and behavior, the nonlinear equations of motion for a system are derived with respect to polar coordinates by the Lagrange's equations. The perturbation method is applied to investigate the dynamic behavior of the system around the equilibrium position. Based on the linearized equations, the dynamic stability of the system around the equilibrium positions is investigated by the eigenvalue analysis.
Impact of dynamic rate coding aspects of mobile phone networks on forensic voice comparison.
Alzqhoul, Esam A S; Nair, Balamurali B T; Guillemin, Bernard J
2015-09-01
Previous studies have shown that landline and mobile phone networks are different in their ways of handling the speech signal, and therefore in their impact on it. But the same is also true of the different networks within the mobile phone arena. There are two major mobile phone technologies currently in use today, namely the global system for mobile communications (GSM) and code division multiple access (CDMA) and these are fundamentally different in their design. For example, the quality of the coded speech in the GSM network is a function of channel quality, whereas in the CDMA network it is determined by channel capacity (i.e., the number of users sharing a cell site). This paper examines the impact on the speech signal of a key feature of these networks, namely dynamic rate coding, and its subsequent impact on the task of likelihood-ratio-based forensic voice comparison (FVC). Surprisingly, both FVC accuracy and precision are found to be better for both GSM- and CDMA-coded speech than for uncoded. Intuitively one expects FVC accuracy to increase with increasing coded speech quality. This trend is shown to occur for the CDMA network, but, surprisingly, not for the GSM network. Further, in respect to comparisons between these two networks, FVC accuracy for CDMA-coded speech is shown to be slightly better than for GSM-coded speech, particularly when the coded-speech quality is high, but in terms of FVC precision the two networks are shown to be very similar. Copyright © 2015 The Chartered Society of Forensic Sciences. Published by Elsevier Ireland Ltd. All rights reserved.
Modeling the reactor core of MNSR to simulate its dynamic behavior using the code PARET
International Nuclear Information System (INIS)
Hainoun, A.; Alhabet, F.
2004-02-01
Using the computer code PARET the core of the MNSR reactor was modelled and the neutronics and thermal hydraulic behaviour of the reactor core for the steady state and selected transients, that deal with step change of reactivity including control rod withdraw starting from steady state at various low power level, were simulated. For this purpose a PARET input model for the core of MNSR reactor has been developed enabling the simulation of neutron kinetic and thermal hydraulic of reactor core including reactivity feedback effects. The neutron kinetic model depends on the point kinetic with 15 groups delayed neutrons including photo neutrons of beryllium reflector. In this regard the effect of photo neutron on the dynamic behaviour has been analysed through two additional calculation. In the first the yield of photo neutrons was neglected completely and in the second its share was added to the sixth group of delayed neutrons. In the thermal hydraulic model the fuel elements with their cooling channels were distributed to 4 different groups with various radial power factors. The pressure lose factors for friction, flow direction change, expansion and contraction were estimated using suitable approaches. The post calculations of the relative neutron flux change and core average temperature were found to be consistent with the experimental measurements. Furthermore, the simulation has indicated the influence of photo neutrons of the Beryllium reflector on the neutron flux behaviour. For the reliability of the results sensitivity analysis was carried out to consider the uncertainty in some important parameters like temperature feedback coefficient and flow velocity. On the other hand the application of PARET in simulation of void formation in the subcooled boiling regime were tested. The calculation indicates the capability of PARET in modelling this phenomenon. However, big discrepancy between calculation results and measurement of axial void distribution were observed
The Analysis of SBWR Critical Power Bundle Using Cobrag Code
Directory of Open Access Journals (Sweden)
Yohannes Sardjono
2013-03-01
Full Text Available The coolant mechanism of SBWR is similar with the Dodewaard Nuclear Power Plant (NPP in the Netherlands that first went critical in 1968. The similarity of both NPP is cooled by natural convection system. These coolant concept is very related with same parameters on fuel bundle design especially fuel bundle length, core pressure drop and core flow rate as well as critical power bundle. The analysis was carried out by using COBRAG computer code. COBRAG computer code is GE Company proprietary. Basically COBRAG computer code is a tool to solve compressible three-dimensional, two fluid, three field equations for two phase flow. The three fields are the vapor field, the continuous liquid field, and the liquid drop field. This code has been applied to analyses model flow and heat transfer within the reactor core. This volume describes the finitevolume equations and the numerical solution methods used to solve these equations. This analysis of same parameters has been done i.e.; inlet sub cooling 20 BTU/lbm and 40 BTU/lbm, 1000 psi pressure and R-factor is 1.038, mass flux are 0.5 Mlb/hr.ft2, 0.75 Mlb/hr.ft2, 1.00 Mlb/hr.ft2 and 1.25 Mlb/hr.ft2. Those conditions based on history operation of some type of the cell fuel bundle line at GE Nuclear Energy. According to the results, it can be concluded that SBWR critical power bundle is 10.5 % less than current BWR critical power bundle with length reduction of 12 ft to 9 ft.
Performance analysis of WS-EWC coded optical CDMA networks with/without LDPC codes
Huang, Chun-Ming; Huang, Jen-Fa; Yang, Chao-Chin
2010-10-01
One extended Welch-Costas (EWC) code family for the wavelength-division-multiplexing/spectral-amplitude coding (WDM/SAC; WS) optical code-division multiple-access (OCDMA) networks is proposed. This system has a superior performance as compared to the previous modified quadratic congruence (MQC) coded OCDMA networks. However, since the performance of such a network is unsatisfactory when the data bit rate is higher, one class of quasi-cyclic low-density parity-check (QC-LDPC) code is adopted to improve that. Simulation results show that the performance of the high-speed WS-EWC coded OCDMA network can be greatly improved by using the LDPC codes.
Directory of Open Access Journals (Sweden)
Wei Cao
2013-01-01
Full Text Available Tumorigenesis is a complex dynamic biological process that includes multiple steps of genetic and epigenetic alterations, aberrant expression of noncoding RNA, and changes in the expression profiles of coding genes. We call the collection of those perturbations in genome space the “cancer initiatome.” Long noncoding RNAs (lncRNAs are pervasively transcribed in the genome and they have key regulatory functions in chromatin remodeling and gene expression. Spatiotemporal variation in the expression of lncRNAs has been observed in development and disease states, including cancer. A few dysregulated lncRNAs have been studied in cancers, but the role of lncRNAs in the cancer initiatome remains largely unknown, especially in esophageal squamous cell carcinoma (ESCC. We conducted a genome-wide screen of the expression of lncRNAs and coding RNAs from ESCC and matched adjacent nonneoplastic normal tissues. We identified differentially expressed lncRNAs and coding RNAs in ESCC relative to their matched normal tissue counterparts and validated the result using polymerase chain reaction analysis. Furthermore, we identified differentially expressed lncRNAs that are co-located and co-expressed with differentially expressed coding RNAs in ESCC and the results point to a potential interaction between lncRNAs and neighboring coding genes that affect ether lipid metabolism, and the interaction may contribute to the development of ESCC. These data provide compelling evidence for a potential novel genomic biomarker of esophageal squamous cell cancer.
Application of an accurate thermal hydraulics solver in VTT's reactor dynamics codes
International Nuclear Information System (INIS)
Rajamaeki, M.; Raety, H.; Kyrki-Rajamaeki, R.; Eskola, M.
1998-01-01
VTT's reactor dynamics codes are developed further and new more detailed models are created for tasks related to increased safety requirements. For thermal hydraulics calculations an accurate general flow model based on a new solution method PLIM has been developed. It has been applied in VTT's one-dimensional TRAB and three-dimensional HEXTRAN codes. Results of a demanding international boron dilution benchmark defined by VTT are given and compared against results of other codes with original or improved boron tracking. The new PLIM method not only allows the accurate modelling of a propagating boron dilution front, but also the tracking of a temperature front, which is missed by the special boron tracking models. (orig.)
Research on verification and validation strategy of detonation fluid dynamics code of LAD2D
Wang, R. L.; Liang, X.; Liu, X. Z.
2017-07-01
The verification and validation (V&V) is an important approach in the software quality assurance of code in complex engineering application. Reasonable and efficient V&V strategy can achieve twice the result with half the effort. This article introduces the software-Lagrangian adaptive hydrodynamics code in 2D space (LAD2D), which is self-developed software in detonation CFD with plastic-elastic structure. The V&V strategy of this detonation CFD code is presented based on the foundation of V&V methodology for scientific software. The basic framework of the module verification and the function validation is proposed, composing the detonation fluid dynamics model V&V strategy of LAD2D.
Development of dynamic simulation code for fuel cycle of fusion reactor
International Nuclear Information System (INIS)
Aoki, Isao; Seki, Yasushi; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1999-02-01
A dynamic simulation code for fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during 2 days pulse operation cycles. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the fuel burn and the function of exhaust, purification, and supply. The processing constants of subsystem for steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using this code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
Energy Technology Data Exchange (ETDEWEB)
Aoki, Isao; Seki, Yasushi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Sasaki, Makoto; Shintani, Kiyonori; Kim, Yeong-Chan
1997-11-01
A dynamic simulation code for the fuel cycle of a fusion experimental reactor has been developed. The code follows the fuel inventory change with time in the plasma chamber and the fuel cycle system during a single pulse operation. The time dependence of the fuel inventory distribution is evaluated considering the fuel burn and exhaust in the plasma chamber, purification and supply functions. For each subsystem of the plasma chamber and the fuel cycle system, the fuel inventory equation is written based on the equation of state considering the function of fuel burn, exhaust, purification, and supply. The processing constants of subsystem for the steady states were taken from the values in the ITER Conceptual Design Activity (CDA) report. Using the code, the time dependence of the fuel supply and inventory depending on the burn state and subsystem processing functions are shown. (author)
The multi-physics, user-friendly gas-dynamics code Visual Tsunami 2.0
International Nuclear Information System (INIS)
Debonnel, C. S.; Trubov, L.; Zeballos, C. A.; Peterson, P. F.
2007-01-01
Since the early 1990's, the series of simulation code known as TSUNAMI has been the main tool employed to explore gas dynamics phenomena in thick-liquid protected inertial fusion target chambers. The applicability and user-friendliness of the code was recently extended through a set of MATLAB pre- and post-processing tools and graphical user interfaces [1]. Geometry, initial, and boundary conditions can be specified from within AutoCAD through a set of in-house AutoLISP graphical user interfaces. A novel MATLAB core was recently developed and tested, and is now routinely used with the user-friendly pre- and post-processors [2]. An overview of Visual Tsunami 2.0, the latest version of the code, is presented here. (authors)
Laminar and Temporal Expression Dynamics of Coding and Noncoding RNAs in the Mouse Neocortex
Directory of Open Access Journals (Sweden)
Sofia Fertuzinhos
2014-03-01
Full Text Available The hallmark of the cerebral neocortex is its organization into six layers, each containing a characteristic set of cell types and synaptic connections. The transcriptional events involved in laminar development and function still remain elusive. Here, we employed deep sequencing of mRNA and small RNA species to gain insights into transcriptional differences among layers and their temporal dynamics during postnatal development of the mouse primary somatosensory neocortex. We identify a number of coding and noncoding transcripts with specific spatiotemporal expression and splicing patterns. We also identify signature trajectories and gene coexpression networks associated with distinct biological processes and transcriptional overlap between these processes. Finally, we provide data that allow the study of potential miRNA and mRNA interactions. Overall, this study provides an integrated view of the laminar and temporal expression dynamics of coding and noncoding transcripts in the mouse neocortex and a resource for studies of neurodevelopment and transcriptome.
Development status of Severe Accident Analysis Code SAMPSON
International Nuclear Information System (INIS)
Iwashita, Tsuyoshi; Ujita, Hiroshi
2000-01-01
The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology' project Phase 1 have been completed. The verification study of Severe Accident Analysis Code SAMPSON prototype developed in Phase 1 was conducted in two steps. First, each analysis module was run independently and analysis results were compared and verified against separate-effect test data with good results. Test data are as follows: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex-Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. Second, with the Simulation Supervisory System, up to 11 analysis modules were executed concurrently in the parallel environment (currently, NUPEC uses IBM-SP2 with 72 process elements), to demonstrate the code capability and integrity. The target plant was Surry as a typical PWR and the initiation events were a 10-inch cold leg failure. The analysis is divided to two cases; one is in-vessel retention analysis when the gap cooling is effective (In-vessel scenario test), the other is analysis of phenomena event is extended to ex-vessel due to the Reactor Pressure Vessel failure when the gap cooling is not sufficient (Ex-vessel scenario test). The system verification test has confirmed that the full scope of the scenarios can be analyzed and phenomena occurred in scenarios can be simulated qualitatively reasonably considering the physical models used for the situation. The Ministry of International Trade and Industry, Japan sponsors this work. (author)
N-MODY: A Code for Collisionless N-body Simulations in Modified Newtonian Dynamics
Londrillo, Pasquale; Nipoti, Carlo
2011-02-01
N-MODY is a parallel particle-mesh code for collisionless N-body simulations in modified Newtonian dynamics (MOND). N-MODY is based on a numerical potential solver in spherical coordinates that solves the non-linear MOND field equation, and is ideally suited to simulate isolated stellar systems. N-MODY can be used also to compute the MOND potential of arbitrary static density distributions. A few applications of N-MODY indicate that some astrophysically relevant dynamical processes are profoundly different in MOND and in Newtonian gravity with dark matter.
Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines
International Nuclear Information System (INIS)
Batygin, Y.
2004-01-01
A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented
Particle-in-Cell Code BEAMPATH for Beam Dynamics Simulations in Linear Accelerators and Beamlines
Energy Technology Data Exchange (ETDEWEB)
Batygin, Y.
2004-10-28
A code library BEAMPATH for 2 - dimensional and 3 - dimensional space charge dominated beam dynamics study in linear particle accelerators and beam transport lines is developed. The program is used for particle-in-cell simulation of axial-symmetric, quadrupole-symmetric and z-uniform beams in a channel containing RF gaps, radio-frequency quadrupoles, multipole lenses, solenoids and bending magnets. The programming method includes hierarchical program design using program-independent modules and a flexible combination of modules to provide the most effective version of the structure for every specific case of simulation. Numerical techniques as well as the results of beam dynamics studies are presented.
ORBIT: A CODE FOR COLLECTIVE BEAM DYNAMICS IN HIGH INTENSITY RINGS
International Nuclear Information System (INIS)
HOLMES, J.A.; DANILOV, V.; GALAMBOS, J.; SHISHLO, A.; COUSINEAU, S.; CHOU, W.; MICHELOTTI, L.; OSTIGUY, J.F.; WEI, J.
2002-01-01
We are developing a computer code, ORBIT, specifically for beam dynamics calculations in high-intensity rings. Our approach allows detailed simulation of realistic accelerator problems. ORBIT is a particle-in-cell tracking code that transports bunches of interacting particles through a series of nodes representing elements, effects, or diagnostics that occur in the accelerator lattice. At present, ORBIT contains detailed models for strip-foil injection, including painting and foil scattering; rf focusing and acceleration; transport through various magnetic elements; longitudinal and transverse impedances; longitudinal, transverse, and three-dimensional space charge forces; collimation and limiting apertures; and the calculation of many useful diagnostic quantities. ORBIT is an object-oriented code, written in C++ and utilizing a scripting interface for the convenience of the user. Ongoing improvements include the addition of a library of accelerator maps, BEAMLINE/MXYZPTLK, the introduction of a treatment of magnet errors and fringe fields; the conversion of the scripting interface to the standard scripting language, Python; and the parallelization of the computations using MPI. The ORBIT code is an open source, powerful, and convenient tool for studying beam dynamics in high-intensity rings
ORBIT: A Code for Collective Beam Dynamics in High-Intensity Rings
Holmes, J. A.; Danilov, V.; Galambos, J.; Shishlo, A.; Cousineau, S.; Chou, W.; Michelotti, L.; Ostiguy, J.-F.; Wei, J.
2002-12-01
We are developing a computer code, ORBIT, specifically for beam dynamics calculations in high-intensity rings. Our approach allows detailed simulation of realistic accelerator problems. ORBIT is a particle-in-cell tracking code that transports bunches of interacting particles through a series of nodes representing elements, effects, or diagnostics that occur in the accelerator lattice. At present, ORBIT contains detailed models for strip-foil injection, including painting and foil scattering; rf focusing and acceleration; transport through various magnetic elements; longitudinal and transverse impedances; longitudinal, transverse, and three-dimensional space charge forces; collimation and limiting apertures; and the calculation of many useful diagnostic quantities. ORBIT is an object-oriented code, written in C++ and utilizing a scripting interface for the convenience of the user. Ongoing improvements include the addition of a library of accelerator maps, BEAMLINE/MXYZPTLK; the introduction of a treatment of magnet errors and fringe fields; the conversion of the scripting interface to the standard scripting language, Python; and the parallelization of the computations using MPI. The ORBIT code is an open source, powerful, and convenient tool for studying beam dynamics in high-intensity rings.
ORBIT: A code for collective beam dynamics in high-intensity rings
International Nuclear Information System (INIS)
Holmes, J.A.; Danilov, V.; Galambos, J.; Shishlo, A.; Cousineau, S.; Chou, W.; Michelotti, L.; Ostiguy, J.-F.; Wei, J.
2002-01-01
We are developing a computer code, ORBIT, specifically for beam dynamics calculations in high-intensity rings. Our approach allows detailed simulation of realistic accelerator problems. ORBIT is a particle-in-cell tracking code that transports bunches of interacting particles through a series of nodes representing elements, effects, or diagnostics that occur in the accelerator lattice. At present, ORBIT contains detailed models for strip-foil injection, including painting and foil scattering; rf focusing and acceleration; transport through various magnetic elements; longitudinal and transverse impedances; longitudinal, transverse, and three-dimensional space charge forces; collimation and limiting apertures; and the calculation of many useful diagnostic quantities. ORBIT is an object-oriented code, written in C++ and utilizing a scripting interface for the convenience of the user. Ongoing improvements include the addition of a library of accelerator maps, BEAMLINE/MXYZPTLK; the introduction of a treatment of magnet errors and fringe fields; the conversion of the scripting interface to the standard scripting language, Python; and the parallelization of the computations using MPI. The ORBIT code is an open source, powerful, and convenient tool for studying beam dynamics in high-intensity rings
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)
A simple in-surge pressure analysis using the SPACE code
International Nuclear Information System (INIS)
Youn, Bum Soo; Kim, Yo Han; Lee, Dong Hyuk; Yang, Chang Keun; Kim, Se Yun; Ha, Sang Jun
2010-01-01
Currently, nuclear safety analysis codes used in Korea are developed by all the overseas. These codes are paying huge fee and permission must be obtained for use in the country. In addition, orders for nuclear power plants must ensure the safety analysis code for independent domestic technology. Therefore, Korea Electric Power Research Institute(KEPRI) is developing the domestic nuclear power safety analysis, SPACE(Safety and Performance Analysis Code for nuclear power plants). To determine the computational power of pressurizer model in development SPACE code, it was compared with existing commercial nuclear power safety analysis code, RETRAN
International Nuclear Information System (INIS)
Lemehov, Sergei; Suzuki, Motoe
2000-01-01
This report summarizes the review on the modeling features of FEMAXI code and proposal of a new theoretical equation model of clad creep on the basis of irradiation-induced microstructure change. It was pointed out that plutonium build-up in fuel matrix and non-uniform radial power profile at high burn-up affect significantly fuel behavior through the interconnected effects with such phenomena as clad irradiation-induced creep, fission gas release, fuel thermal conductivity degradation, rim porous band formation and associated fuel swelling. Therefore, these combined effects should be properly incorporated into the models of the FEMAXI code so that the code can carry out numerical analysis at the level of accuracy and elaboration that modern experimental data obtained in test reactors have. Also, the proposed new mechanistic clad creep model has a general formalism which allows the model to be flexibly applied for clad behavior analysis under normal operation conditions and power transients as well for Zr-based clad materials by the use of established out-of-pile mechanical properties. The model has been tested against experimental data, while further verification is needed with specific emphasis on power ramps and transients. (author)
Nonlinear analysis of pupillary dynamics.
Onorati, Francesco; Mainardi, Luca Tommaso; Sirca, Fabiola; Russo, Vincenzo; Barbieri, Riccardo
2016-02-01
Pupil size reflects autonomic response to different environmental and behavioral stimuli, and its dynamics have been linked to other autonomic correlates such as cardiac and respiratory rhythms. The aim of this study is to assess the nonlinear characteristics of pupil size of 25 normal subjects who participated in a psychophysiological experimental protocol with four experimental conditions, namely “baseline”, “anger”, “joy”, and “sadness”. Nonlinear measures, such as sample entropy, correlation dimension, and largest Lyapunov exponent, were computed on reconstructed signals of spontaneous fluctuations of pupil dilation. Nonparametric statistical tests were performed on surrogate data to verify that the nonlinear measures are an intrinsic characteristic of the signals. We then developed and applied a piecewise linear regression model to detrended fluctuation analysis (DFA). Two joinpoints and three scaling intervals were identified: slope α0, at slow time scales, represents a persistent nonstationary long-range correlation, whereas α1 and α2, at middle and fast time scales, respectively, represent long-range power-law correlations, similarly to DFA applied to heart rate variability signals. Of the computed complexity measures, α0 showed statistically significant differences among experimental conditions (pnonlinear dynamics, (b) three well-defined and distinct long-memory processes exist at different time scales, and (c) autonomic stimulation is partially reflected in nonlinear dynamics. (c) autonomic stimulation is partially reflected in nonlinear dynamics.
Meanline Analysis of Turbines with Choked Flow in the Object-Oriented Turbomachinery Analysis Code
Hendricks, Eric S.
2016-01-01
The Object-Oriented Turbomachinery Analysis Code (OTAC) is a new meanline/streamline turbomachinery modeling tool being developed at NASA GRC. During the development process, a limitation of the code was discovered in relation to the analysis of choked flow in axial turbines. This paper describes the relevant physics for choked flow as well as the changes made to OTAC to enable analysis in this flow regime.
RADTRAN 5 - A computer code for transportation risk analysis
International Nuclear Information System (INIS)
Neuhauser, K.S.; Kanipe, F.L.
1993-01-01
The RADTRAN 5 computer code has been developed to estimate radiological and nonradiological risks of radioactive materials transportation. RADTRAN 5 is written in ANSI standard FORTRAN 77; the code contains significant advances in the methodology first pioneered with the LINK option of RADTRAN 4. A major application of the LINK methodology is route-specific analysis. Another application is comparisons of attributes along the same route segments. Nonradiological risk factors have been incorporated to allow users to estimate nonradiological fatalities and injuries that might occur during the transportation event(s) being analyzed. These fatalities include prompt accidental fatalities from mechanical causes. Values of these risk factors for the United States have been made available in the code as optional defaults. Several new health effects models have been published in the wake of the Hiroshima-Nagasaki dosimetry reassessment, and this has emphasized the need for flexibility in the RADTRAN approach to health-effects calculations. Therefore, the basic set of health-effects conversion equations in RADTRAN have been made user-definable. All parameter values can be changed by the user, but a complete set of default values are available for both the new International Commission on Radiation Protection model (ICRP Publication 60) and the recent model of the U.S. National Research Council's Committee on the Biological Effects of Radiation (BEIR V). The meteorological input data tables have been modified to permit optional entry of maximum downwind distances for each dose isopleth. The expected dose to an individual in each isodose area is also calculated and printed automatically. Examples are given that illustrate the power and flexibility of the RADTRAN 5 computer code. (J.P.N.)
Users manual for CAFE-3D : a computational fluid dynamics fire code
International Nuclear Information System (INIS)
Khalil, Imane; Lopez, Carlos; Suo-Anttila, Ahti Jorma
2005-01-01
The Container Analysis Fire Environment (CAFE) computer code has been developed to model all relevant fire physics for predicting the thermal response of massive objects engulfed in large fires. It provides realistic fire thermal boundary conditions for use in design of radioactive material packages and in risk-based transportation studies. The CAFE code can be coupled to commercial finite-element codes such as MSC PATRAN/THERMAL and ANSYS. This coupled system of codes can be used to determine the internal thermal response of finite element models of packages to a range of fire environments. This document is a user manual describing how to use the three-dimensional version of CAFE, as well as a description of CAFE input and output parameters. Since this is a user manual, only a brief theoretical description of the equations and physical models is included
SRAC: JAERI thermal reactor standard code system for reactor design and analysis
International Nuclear Information System (INIS)
Tsuchihashi, Keichiro; Takano, Hideki; Horikami, Kunihiko; Ishiguro, Yukio; Kaneko, Kunio; Hara, Toshiharu.
1983-01-01
The SRAC (Standard Reactor Analysis Code) is a code system for nuclear reactor analysis and design. It is composed of neutron cross section libraries and auxiliary processing codes, neutron spectrum routines, a variety of transport, 1-, 2- and 3-D diffusion routines, dynamic parameters and cell burn-up routines. By making the best use of the individual code function in the SRAC system, the user can select either the exact method for an accurate estimate of reactor characteristics or the economical method aiming at a shorter computer time, depending on the purpose of study. The user can select cell or core calculation; fixed source or eigenvalue problem; transport (collision probability or Sn) theory or diffusion theory. Moreover, smearing and collapsing of macroscopic cross sections are separately done by the user's selection. And a special attention is paid for double heterogeneity. Various techniques are employed to access the data storage and to optimize the internal data transfer. Benchmark calculations using the SRAC system have been made extensively for the Keff values of various types of critical assemblies (light water, heavy water and graphite moderated systems, and fast reactor systems). The calculated results show good prediction for the experimental Keff values. (author)
Available computer codes and data for radiation transport analysis
International Nuclear Information System (INIS)
Trubey, D.K.; Maskewitz, B.F.; Roussin, R.W.
1975-01-01
The Radiation Shielding Information Center (RSIC), sponsored and supported by the Energy Research and Development Administration (ERDA) and the Defense Nuclear Agency (DNA), is a technical institute serving the radiation transport and shielding community. It acquires, selects, stores, retrieves, evaluates, analyzes, synthesizes, and disseminates information on shielding and ionizing radiation transport. The major activities include: (1) operating a computer-based information system and answering inquiries on radiation analysis, (2) collecting, checking out, packaging, and distributing large computer codes, and evaluated and processed data libraries. The data packages include multigroup coupled neutron-gamma-ray cross sections and kerma coefficients, other nuclear data, and radiation transport benchmark problem results
Development validation and use of computer codes for inelastic analysis
International Nuclear Information System (INIS)
Jobson, D.A.
1983-01-01
A finite element scheme is a system which provides routines so carry out the operations which are common to all finite element programs. The list of items that can be provided as standard by the finite element scheme is surprisingly large and the list provided by the UNCLE finite element scheme is unusually comprehensive. This presentation covers the following: construction of the program, setting up a finite element mesh, generation of coordinates, incorporating boundary and load conditions. Program validation was done by creep calculations performed using CAUSE code. Program use is illustrated by calculating a typical inelastic analysis problem. This includes computer model of the PFR intermediate heat exchanger
Institute of Scientific and Technical Information of China (English)
呼义翔; 雷天时; 吴撼宇; 郭宁; 韩娟娟; 邱爱慈; 王亮平; 黄涛; 丛培天; 张信军; 李岩; 曾正中; 孙铁平
2011-01-01
The transmission-line-circuit model of the Z accelerator, developed originally by W. A. STYGAR, P. A. CORCORAN, et al., is revised. The revised model uses different calculations for the electron loss and flow impedance in the magnetically insulated transmission line system of the Z accelerator before and after magnetic insulation is established. By including electron pressure and zero electric field at the cathode, a closed set of equations is obtained at each time step, and dynamic shunt resistance （used to represent any electron loss to the anode） and flow impedance are solved, which have been incorporated into the transmission line code for simulations of the vacuum section in the Z accelerator. Finally, the results are discussed in comparison with earlier findings to show the effectiveness and limitations of the model.
Code Development and Analysis Program: developmental checkout of the BEACON/MOD2A code
International Nuclear Information System (INIS)
Ramsthaler, J.A.; Lime, J.F.; Sahota, M.S.
1978-12-01
A best-estimate transient containment code, BEACON, is being developed by EG and G Idaho, Inc. for the Nuclear Regulatory Commission's reactor safety research program. This is an advanced, two-dimensional fluid flow code designed to predict temperatures and pressures in a dry PWR containment during a hypothetical loss-of-coolant accident. The most recent version of the code, MOD2A, is presently in the final stages of production prior to being released to the National Energy Software Center. As part of the final code checkout, seven sample problems were selected to be run with BEACON/MOD2A
Numerical analysis and nuclear standard code application to thermal fatigue
International Nuclear Information System (INIS)
Merola, M.
1992-01-01
The present work describes the Joint Research Centre Ispra contribution to the IAEA benchmark exercise 'Lifetime Behaviour of the First Wall of Fusion Machines'. The results of the numerical analysis of the reference thermal fatigue experiment are presented. Then a discussion on the numerical analysis of thermal stress is tackled, pointing out its particular aspects in view of their influence on the stress field evaluation. As far as the design-allowable number of cycles are concerned the American nuclear code ASME and the French code RCC-MR are applied and the reasons for the different results obtained are investigated. As regards a realistic fatigue lifetime evaluation, the main problems to be solved are brought out. This work, is intended as a preliminary basis for a discussion focusing on the main characteristics of the thermal fatigue problem from both a numerical and a lifetime assessment point of view. In fact the present margin of discretion left to the analyst may cause undue discrepancies in the results obtained. A sensitivity analysis of the main parameters involved is desirable and more precise design procedures should be stated
Substitution dynamical systems spectral analysis
Queffélec, Martine
2010-01-01
This volume mainly deals with the dynamics of finitely valued sequences, and more specifically, of sequences generated by substitutions and automata. Those sequences demonstrate fairly simple combinatorical and arithmetical properties and naturally appear in various domains. As the title suggests, the aim of the initial version of this book was the spectral study of the associated dynamical systems: the first chapters consisted in a detailed introduction to the mathematical notions involved, and the description of the spectral invariants followed in the closing chapters. This approach, combined with new material added to the new edition, results in a nearly self-contained book on the subject. New tools - which have also proven helpful in other contexts - had to be developed for this study. Moreover, its findings can be concretely applied, the method providing an algorithm to exhibit the spectral measures and the spectral multiplicity, as is demonstrated in several examples. Beyond this advanced analysis, many...
Analysis of Non-binary Hybrid LDPC Codes
Sassatelli, Lucile; Declercq, David
2008-01-01
In this paper, we analyse asymptotically a new class of LDPC codes called Non-binary Hybrid LDPC codes, which has been recently introduced. We use density evolution techniques to derive a stability condition for hybrid LDPC codes, and prove their threshold behavior. We study this stability condition to conclude on asymptotic advantages of hybrid LDPC codes compared to their non-hybrid counterparts.
Formal analysis of design process dynamics
Bosse, T.; Jonker, C.M.; Treur, J.
2010-01-01
This paper presents a formal analysis of design process dynamics. Such a formal analysis is a prerequisite to come to a formal theory of design and for the development of automated support for the dynamics of design processes. The analysis was geared toward the identification of dynamic design
Formal Analysis of Design Process Dynamics
Bosse, T.; Jonker, C.M.; Treur, J.
2010-01-01
This paper presents a formal analysis of design process dynamics. Such a formal analysis is a prerequisite to come to a formal theory of design and for the development of automated support for the dynamics of design processes. The analysis was geared toward the identification of dynamic design
Robustness Analysis of Dynamic Watermarks
Directory of Open Access Journals (Sweden)
Ivan V. Nechta
2017-06-01
Full Text Available In this paper we consider previously known scheme of dynamic watermarks embedding (Ra- dix-n that is used for preventing illegal use of software. According to the scheme a watermark is dynamic linked data structure (graph, which is created in memory during program execution. Hidden data, such as information about author, can be represented in a different type of graph structure. This data can be extracted and demonstrated in judicial proceedings. This paper declared that the above mentioned scheme was previously one of the most reliable, has a number of features that allows an attacker to detect a stage of watermark construction in the program, and therefore it can be corrupted or deleted. The author of this article shows the weakness of Radix-N scheme, which consists in the fact that we can reveal dynamic data structures of a program by using information received from some API-functions hooker which catches function calls of dynamic memory allocation. One of these data structures is the watermark. Pointers on dynamically created objects (arrays, variables, class items, etc. of a program can be detected by content analysis of computer's RAM. Different dynamic objects in memory interconnected by pointers form dynamic data structures of a program such as lists, stacks, trees and other graphs (including the watermark. Our experiment shows that in the vast majority of cases the amount of data structure in programs is small, which increases probability of a successful attack. Also we present an algorithm for finding connected components of a graph with linear time-consuming in cases where the number of nodes is about 106. On the basis of the experimental findings the new watermarking scheme has been presented, which is resistant to the proposed attack. It is offered to use different graph structure representation of a watermark, where edges are implemented using unique signatures. Our scheme uses content encrypting of graph nodes (except signature
Systemization of burnup sensitivity analysis code (2) (Contract research)
International Nuclear Information System (INIS)
Tatsumi, Masahiro; Hyoudou, Hideaki
2008-08-01
Towards the practical use of fast reactors, it is a very important subject to improve prediction accuracy for neutronic properties in LMFBR cores from the viewpoint of improvements on plant economic efficiency with rationally high performance cores and that on reliability and safety margins. A distinct improvement on accuracy in nuclear core design has been accomplished by the development of adjusted nuclear library using the cross-section adjustment method, in which the results of critical experiments of JUPITER and so on are reflected. In the design of large LMFBR cores, however, it is important to accurately estimate not only neutronic characteristics, for example, reaction rate distribution and control rod worth but also burnup characteristics, for example, burnup reactivity loss, breeding ratio and so on. For this purpose, it is desired to improve prediction accuracy of burnup characteristics using the data widely obtained in actual core such as the experimental fast reactor 'JOYO'. The analysis of burnup characteristic is needed to effectively use burnup characteristics data in the actual cores based on the cross-section adjustment method. So far, a burnup sensitivity analysis code, SAGEP-BURN, has been developed and confirmed its effectiveness. However, there is a problem that analysis sequence become inefficient because of a big burden to users due to complexity of the theory of burnup sensitivity and limitation of the system. It is also desired to rearrange the system for future revision since it is becoming difficult to implement new functions in the existing large system. It is not sufficient to unify each computational component for the following reasons: the computational sequence may be changed for each item being analyzed or for purpose such as interpretation of physical meaning. Therefore, it is needed to systemize the current code for burnup sensitivity analysis with component blocks of functionality that can be divided or constructed on occasion
Natural image sequences constrain dynamic receptive fields and imply a sparse code.
Häusler, Chris; Susemihl, Alex; Nawrot, Martin P
2013-11-06
In their natural environment, animals experience a complex and dynamic visual scenery. Under such natural stimulus conditions, neurons in the visual cortex employ a spatially and temporally sparse code. For the input scenario of natural still images, previous work demonstrated that unsupervised feature learning combined with the constraint of sparse coding can predict physiologically measured receptive fields of simple cells in the primary visual cortex. This convincingly indicated that the mammalian visual system is adapted to the natural spatial input statistics. Here, we extend this approach to the time domain in order to predict dynamic receptive fields that can account for both spatial and temporal sparse activation in biological neurons. We rely on temporal restricted Boltzmann machines and suggest a novel temporal autoencoding training procedure. When tested on a dynamic multi-variate benchmark dataset this method outperformed existing models of this class. Learning features on a large dataset of natural movies allowed us to model spatio-temporal receptive fields for single neurons. They resemble temporally smooth transformations of previously obtained static receptive fields and are thus consistent with existing theories. A neuronal spike response model demonstrates how the dynamic receptive field facilitates temporal and population sparseness. We discuss the potential mechanisms and benefits of a spatially and temporally sparse representation of natural visual input. Copyright © 2013 The Authors. Published by Elsevier B.V. All rights reserved.
Comparative analysis of design codes for timber bridges in Canada, the United States, and Europe
James Wacker; James (Scott) Groenier
2010-01-01
The United States recently completed its transition from the allowable stress design code to the load and resistance factor design (LRFD) reliability-based code for the design of most highway bridges. For an international perspective on the LRFD-based bridge codes, a comparative analysis is presented: a study addressed national codes of the United States, Canada, and...
Spectral analysis of bedform dynamics
DEFF Research Database (Denmark)
Winter, Christian; Ernstsen, Verner Brandbyge; Noormets, Riko
Successive multibeam echo sounder surveys in tidal channels off Esbjerg (Denmark) on the North Sea coast reveal the dynamics of subaquatic compound dunes. Mainly driven by tidal currents, dune structures show complex migration patterns in all temporal and spatial scales. Common methods for the an....... The proposed method overcomes the above mentioned problems of common descriptive analysis as it is an objective and straightforward mathematical process. The spectral decomposition of superimposed dunes allows a detailed description and analysis of dune patterns and migration.......Successive multibeam echo sounder surveys in tidal channels off Esbjerg (Denmark) on the North Sea coast reveal the dynamics of subaquatic compound dunes. Mainly driven by tidal currents, dune structures show complex migration patterns in all temporal and spatial scales. Common methods...... allows the application of a procedure, which has been a standard for the analysis of water waves for long times: The bathymetric signal of a cross-section of subaquatic compound dunes is approximated by the sum of a set of harmonic functions, derived by Fourier transformation. If the wavelength...
Analysis of Isp-42, panda test with the spectra code
International Nuclear Information System (INIS)
Stempniewicz, M.M.
2001-01-01
International Standard Problems (ISP) are organized in order to assess the ability of computer codes to predict the outcome of accidents in Nuclear Power Plants. The ISP-42 test was performed at Paul Scherrer Institute in 1998, as a sequence of six phases, Phase A through F Blind and open calculations of ISP-42 were performed with the computer code SPECTRA for each of the six phases. SPECTRA is a general tool for thermal-hydraulic analyses. Results of blind calculations are in good agreement with experiment. For open calculations several modifications were made in the model. These were mainly corrections of some input errors made in the model used for blind analysis. Some small improvements to the nodalization were made. Results of open calculations are generally closer to the experiment than the blind results. For phase D the containment pressure prediction was somewhat worse in the open calculation. Based on comparisons of blind and open results with experiment several conclusions may be drawn: - use of long ID structures, in contact with pool and atmosphere should be avoided, - PCC units are better represented with larger amount of Control Volumes, - two parallel junctions should be used to represent large openings between vessels, like drywell air line, etc., - careful verification of input decks is needed, - stratification models in SPECTRA are useful for cases with light gas injection; for complex cases a complementary SPECTRA-CFD analysis may be performed. (author)
Investigation of alpha experiment by severe accident analysis code SAMPSON
International Nuclear Information System (INIS)
Baglietto, Emilio; Ninokata, Hisashi; Naitoh, Masanori
2006-01-01
The severe accident analysis code SAMPSON is adopted in this work to evaluate its capability of reproducing the complex gap cooling phenomenon. The ALPHA experiment is adopted for validation, where molten aluminum oxide (Al 2 O 3 ) produced by a thermite reaction is poured into a water filled hemispherical vessel at the ambient pressure of approximately 1.3 MPa. The spreading and cooling of the debris that has relocated into the pressure vessel lower plenum are simulated, including the analysis of the RPV failure. The model included in the core to mimic the water penetration inside the gap is evaluated and improvements are proposed. The importance of the introduction of some mechanistic approach to describe the gap formation and evolution is underlined, where the results show its necessity in order to correctly reproduce the experimental trends. (author)
Optimization and Validation of the Developed Uranium Isotopic Analysis Code
Energy Technology Data Exchange (ETDEWEB)
Kim, J. H.; Kang, M. Y.; Kim, Jinhyeong; Choi, H. D. [Seoul National Univ., Seoul (Korea, Republic of)
2014-10-15
γ-ray spectroscopy is a representative non-destructive assay for nuclear material, and less time-consuming and less expensive than the destructive analysis method. The destructive technique is more precise than NDA technique, however, there is some correction algorithm which can improve the performance of γ-spectroscopy. For this reason, an analysis code for uranium isotopic analysis is developed by Applied Nuclear Physics Group in Seoul National University. Overlapped γ- and x-ray peaks in the 89-101 keV X{sub α}-region are fitted with Gaussian and Lorentzian distribution peak functions, tail and background functions. In this study, optimizations for the full-energy peak efficiency calibration and fitting parameters of peak tail and background are performed, and validated with 24 hour acquisition of CRM uranium samples. The optimization of peak tail and background parameters are performed with the validation by using CRM uranium samples. The analysis performance is improved in HEU samples, but more optimization of fitting parameters is required in LEU sample analysis. In the future, the optimization research about the fitting parameters with various type of uranium samples will be performed. {sup 234}U isotopic analysis algorithms and correction algorithms (coincidence effect, self-attenuation effect) will be developed.
Energy Technology Data Exchange (ETDEWEB)
Di Maio, Francesco, E-mail: francesco.dimaio@polimi.it [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Nicola, Giancarlo [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Zio, Enrico [Energy Department, Politecnico di Milano, Via La Masa 34, 20156 Milano (Italy); Chair on System Science and Energetic Challenge Fondation EDF, Ecole Centrale Paris and Supelec, Paris (France); Yu, Yu [School of Nuclear Science and Engineering, North China Electric Power University, 102206 Beijing (China)
2015-08-15
Highlights: • Uncertainties of TH codes affect the system failure probability quantification. • We present Finite Mixture Models (FMMs) for sensitivity analysis of TH codes. • FMMs approximate the pdf of the output of a TH code with a limited number of simulations. • The approach is tested on a Passive Containment Cooling System of an AP1000 reactor. • The novel approach overcomes the results of a standard variance decomposition method. - Abstract: For safety analysis of Nuclear Power Plants (NPPs), Best Estimate (BE) Thermal Hydraulic (TH) codes are used to predict system response in normal and accidental conditions. The assessment of the uncertainties of TH codes is a critical issue for system failure probability quantification. In this paper, we consider passive safety systems of advanced NPPs and present a novel approach of Sensitivity Analysis (SA). The approach is based on Finite Mixture Models (FMMs) to approximate the probability density function (i.e., the uncertainty) of the output of the passive safety system TH code with a limited number of simulations. We propose a novel Sensitivity Analysis (SA) method for keeping the computational cost low: an Expectation Maximization (EM) algorithm is used to calculate the saliency of the TH code input variables for identifying those that most affect the system functional failure. The novel approach is compared with a standard variance decomposition method on a case study considering a Passive Containment Cooling System (PCCS) of an Advanced Pressurized reactor AP1000.
Analysis of selected Halden overpressure tests using the FALCON code
Energy Technology Data Exchange (ETDEWEB)
Khvostov, G., E-mail: grigori.khvostov@psi.ch [Paul Scherrer Institut, CH 5232 Villigen PSI (Switzerland); Wiesenack, W. [Institute for Energy Technology – OECD Halden Reactor Project, P.O. Box 173, N-1751 Halden (Norway)
2016-12-15
Highlights: • We analyse four Halden overpressure tests. • We determine a critical overpressure value for lift-off in a BWR fuel sample. • We show the role of bonding in over-pressurized rod behaviour. • We analytically quantify the degree of bonding via its impact on cladding elongation. • We hypothesize on an effect of circumferential cracks on thermal fuel response to overpressure. • We estimate a thermal effect of circumferential cracks based on interpretation of the data. - Abstract: Four Halden overpressure (lift-off) tests using samples with uranium dioxide fuel pre-irradiated in power reactors to a burnup of 60 MWd/kgU are analyzed. The FALCON code coupled to a mechanistic model, GRSW-A for fission gas release and gaseous-bubble swelling is used for the calculation. The advanced version of the FALCON code is shown to be applicable to best-estimate predictive analysis of overpressure tests using rods without, or weak pellet-cladding bonding, as well as scoping analysis of tests with fuels where stronger pellet-cladding bonding occurs. Significant effects of bonding and fuel cracking/relocation on the thermal and mechanical behaviour of highly over-pressurized rods are shown. The effect of bonding is particularly pronounced in the tests with the PWR samples. The present findings are basically consistent with an earlier analysis based on a direct interpretation of the experimental data. Additionally, in this paper, the specific effects are quantified based on the comparison of the data with the results of calculation. It is concluded that the identified effects are largely beyond the current traditional fuel-rod licensing analysis methods.
Modeling compositional dynamics based on GC and purine contents of protein-coding sequences
Zhang, Zhang; Yu, Jun
2010-01-01
Background: Understanding the compositional dynamics of genomes and their coding sequences is of great significance in gaining clues into molecular evolution and a large number of publically-available genome sequences have allowed us to quantitatively predict deviations of empirical data from their theoretical counterparts. However, the quantification of theoretical compositional variations for a wide diversity of genomes remains a major challenge.Results: To model the compositional dynamics of protein-coding sequences, we propose two simple models that take into account both mutation and selection effects, which act differently at the three codon positions, and use both GC and purine contents as compositional parameters. The two models concern the theoretical composition of nucleotides, codons, and amino acids, with no prerequisite of homologous sequences or their alignments. We evaluated the two models by quantifying theoretical compositions of a large collection of protein-coding sequences (including 46 of Archaea, 686 of Bacteria, and 826 of Eukarya), yielding consistent theoretical compositions across all the collected sequences.Conclusions: We show that the compositions of nucleotides, codons, and amino acids are largely determined by both GC and purine contents and suggest that deviations of the observed from the expected compositions may reflect compositional signatures that arise from a complex interplay between mutation and selection via DNA replication and repair mechanisms.Reviewers: This article was reviewed by Zhaolei Zhang (nominated by Mark Gerstein), Guruprasad Ananda (nominated by Kateryna Makova), and Daniel Haft. 2010 Zhang and Yu; licensee BioMed Central Ltd.
Modeling compositional dynamics based on GC and purine contents of protein-coding sequences
Zhang, Zhang
2010-11-08
Background: Understanding the compositional dynamics of genomes and their coding sequences is of great significance in gaining clues into molecular evolution and a large number of publically-available genome sequences have allowed us to quantitatively predict deviations of empirical data from their theoretical counterparts. However, the quantification of theoretical compositional variations for a wide diversity of genomes remains a major challenge.Results: To model the compositional dynamics of protein-coding sequences, we propose two simple models that take into account both mutation and selection effects, which act differently at the three codon positions, and use both GC and purine contents as compositional parameters. The two models concern the theoretical composition of nucleotides, codons, and amino acids, with no prerequisite of homologous sequences or their alignments. We evaluated the two models by quantifying theoretical compositions of a large collection of protein-coding sequences (including 46 of Archaea, 686 of Bacteria, and 826 of Eukarya), yielding consistent theoretical compositions across all the collected sequences.Conclusions: We show that the compositions of nucleotides, codons, and amino acids are largely determined by both GC and purine contents and suggest that deviations of the observed from the expected compositions may reflect compositional signatures that arise from a complex interplay between mutation and selection via DNA replication and repair mechanisms.Reviewers: This article was reviewed by Zhaolei Zhang (nominated by Mark Gerstein), Guruprasad Ananda (nominated by Kateryna Makova), and Daniel Haft. 2010 Zhang and Yu; licensee BioMed Central Ltd.
Development of the Monju core safety analysis numerical models by super-COPD code
International Nuclear Information System (INIS)
Yamada, Fumiaki; Minami, Masaki
2010-12-01
Japan Atomic Energy Agency constructed a computational model for safety analysis of Monju reactor core to be built into a modularized plant dynamics analysis code Super-COPD code, for the purpose of heat removal capability evaluation at the in total 21 defined transients in the annex to the construction permit application. The applicability of this model to core heat removal capability evaluation has been estimated by back to back result comparisons of the constituent models with conventionally applied codes and by application of the unified model. The numerical model for core safety analysis has been built based on the best estimate model validated by the actually measured plant behavior up to 40% rated power conditions, taking over safety analysis models of conventionally applied COPD and HARHO-IN codes, to be capable of overall calculations of the entire plant with the safety protection and control systems. Among the constituents of the analytical model, neutronic-thermal model, heat transfer and hydraulic models of PHTS, SHTS, and water/steam system are individually verified by comparisons with the conventional calculations. Comparisons are also made with the actually measured plant behavior up to 40% rated power conditions to confirm the calculation adequacy and conservativeness of the input data. The unified analytical model was applied to analyses of in total 8 anomaly events; reactivity insertion, abnormal power distribution, decrease and increase of coolant flow rate in PHTS, SHTS and water/steam systems. The resulting maximum values and temporal variations of the key parameters in safety evaluation; temperatures of fuel, cladding, in core sodium coolant and RV inlet and outlet coolant have negligible discrepancies against the existing analysis result in the annex to the construction permit application, verifying the unified analytical model. These works have enabled analytical evaluation of Monju core heat removal capability by Super-COPD utilizing the
Interface design of VSOP'94 computer code for safety analysis
International Nuclear Information System (INIS)
Natsir, Khairina; Andiwijayakusuma, D.; Wahanani, Nursinta Adi; Yazid, Putranto Ilham
2014-01-01
Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects
Interface design of VSOP'94 computer code for safety analysis
Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi
2014-09-01
Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.
Dynamic analysis of the BPX machine structure
International Nuclear Information System (INIS)
Dahlgen, F.; Citrolo, J.; Knutson, D.; Kalish, M.
1992-01-01
A preliminary analysis of the response of the BPX machine structure to a seismic input was performed. MSC/NASTRAN 5 , a general purpose XXX element computer code, has been used. The purpose of this paper is to assess the probable range of seismically induced stresses and deflections in the machine substructure which connects the machine to the test cell floor, with particular emphasis on the shear pins which will be used to attach the TF coil modules to the machine substructure (for a more detailed description of the shear pins and structure see ref. 4 in these proceedings). The model was developed with sufficient detail to be used subsequently to investigate the transient response to various dynamic loading conditions imposed on the structure by the PF, TF, and Vacuum Vessel, during normal and off-normal operations. The model does not include the mass and stiffness of the building or the building-soil interaction and as such can only be considered an interim assessment of the dynamic response of the machine to the S.S.E.(this is the Safe Shutdown Earthquake which is also the Design XXX Earthquake for all major structural components)
Dynamic spatial coding within the dorsal frontoparietal network during a visual search task.
Directory of Open Access Journals (Sweden)
Wieland H Sommer
Full Text Available To what extent are the left and right visual hemifields spatially coded in the dorsal frontoparietal attention network? In many experiments with neglect patients, the left hemisphere shows a contralateral hemifield preference, whereas the right hemisphere represents both hemifields. This pattern of spatial coding is often used to explain the right-hemispheric dominance of lesions causing hemispatial neglect. However, pathophysiological mechanisms of hemispatial neglect are controversial because recent experiments on healthy subjects produced conflicting results regarding the spatial coding of visual hemifields. We used an fMRI paradigm that allowed us to distinguish two attentional subprocesses during a visual search task. Either within the left or right hemifield subjects first attended to stationary locations (spatial orienting and then shifted their attentional focus to search for a target line. Dynamic changes in spatial coding of the left and right hemifields were observed within subregions of the dorsal front-parietal network: During stationary spatial orienting, we found the well-known spatial pattern described above, with a bilateral hemifield representation in the right hemisphere and a contralateral preference in the left hemisphere. However, during search, the right hemisphere had a contralateral preference and the left hemisphere equally represented both hemifields. This finding leads to novel perspectives regarding models of visuospatial attention and hemispatial neglect.
Hanhart, Philippe; Řeřábek, Martin; Ebrahimi, Touradj
2015-09-01
This paper reports the details and results of the subjective evaluations conducted at EPFL to evaluate the responses to the Call for Evidence (CfE) for High Dynamic Range (HDR) and Wide Color Gamut (WCG) Video Coding issued by Moving Picture Experts Group (MPEG). The CfE on HDR/WCG Video Coding aims to explore whether the coding efficiency and/or the functionality of the current version of HEVC standard can be signi_cantly improved for HDR and WCG content. In total, nine submissions, five for Category 1 and four for Category 3a, were compared to the HEVC Main 10 Profile based Anchor. More particularly, five HDR video contents, compressed at four bit rates by each proponent responding to the CfE, were used in the subjective evaluations. Further, the side-by-side presentation methodology was used for the subjective experiment to discriminate small differences between the Anchor and proponents. Subjective results shows that the proposals provide evidence that the coding efficiency can be improved in a statistically noticeable way over MPEG CfE Anchors in terms of perceived quality within the investigated content. The paper further benchmarks the selected objective metrics based on their correlations with the subjective ratings. It is shown that PSNR-DE1000, HDRVDP- 2, and PSNR-Lx can reliably detect visible differences between the proposed encoding solutions and current HEVC standard.
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
Dynamic analysis of the condensate feedwater system in boiling water reactor plants
International Nuclear Information System (INIS)
Tanji, J.; Omori, T.
1982-01-01
The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents
Corporate governance codes and their contents : An analysis of Eastern European codes
Hermes, Niels; Postma, Theo J. B. M.; Zivkov, Orestis
2007-01-01
Existing literature suggests that the contents of corporate governance codes are similar due to external forces, such as increased integration of countries in the global economy, the increased role of foreign institutional investors and recommendations on corporate governance practices of
COMTA - a computer code for fuel mechanical and thermal analysis
International Nuclear Information System (INIS)
Basu, S.; Sawhney, S.S.; Anand, A.K.; Anantharaman, K.; Mehta, S.K.
1979-01-01
COMTA is a generalized computer code for integrity analysis of the free standing fuel cladding, with natural UO 2 or mixed oxide fuel pellets. Thermal and Mechanical analysis is done simultaneously for any power history of the fuel pin. For analysis, the fuel cladding is assumed to be axisymmetric and is subjected to axisymmetric load due to contact pressure, gas pressure, coolant pressure and thermal loads. Axial variation of load is neglected and creep and plasticity are assumed to occur at constant volume. The pellet is assumed to be made of concentric annuli. The fission gas release integral is dependent on the temperature and the power produced in each annulus. To calculate the temperature distribution in the fuel pin, the variation of bulk coolant temperature is given as an input to the code. Gap conductance is calculated at every time step, considering fuel densification, fuel relocation and gap closure, filler gas dilution by released fission gas, gap closure by expansion and irradiation swelling. Overall gap conductance is contributed by heat transfer due to the three modes; conduction convection and radiation as per modified Ross and Stoute model. Equilibrium equations, compatibility equations, stress strain relationships (including thermal strains and permanent strains due to creep and plasticity) are used to obtain triaxial stresses and strains. Thermal strain is assumed to be zero at hot zero power conditions. The boundary conditions are obtained for radial stresses at outside and inside surfaces by making these equal to coolant pressure and internal pressure respectively. A multi-mechanism creep model which accounts for thermal and irradiation creep is used to calculate the overall creep rate. Effective plastic strain is a function of effective stress and material constants. (orig.)
Development of a safety analysis code for molten salt reactors
International Nuclear Information System (INIS)
Zhang Dalin; Qiu Suizheng; Su Guanghui
2009-01-01
The molten salt reactor (MSR) well suited to fulfill the criteria defined by the Generation IV International Forum (GIF) is presently revisited all around the world because of different attractive features of current renewed relevance. The MSRs are characterized by using the fluid-fuel, so that their technologies are fundamentally different from those used in the conventional solid-fuel reactors. In this work, in particular, the attention is focused on the safety characteristic analysis of the MSRs, in which a point kinetic model considering the flow effects of the fuel salt is established for the MSRs and calculated by developing a microcomputer code coupling with a simplified heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the molten salt actinide recycler and transmuter system (MOSART) by simulating three types of basic transient conditions including the unprotected loss of flow, unprotected overcooling accident and unprotected transient overpower. Some reasonable results are obtained for the MOSART, which show that the MOSART conceptual design is an inherently stable reactor design. The present study provides some valuable information for the research and design of the new generation MSRs.
Light curves for ''bump Cepheids'' computed with a dynamically zoned pulsation code
International Nuclear Information System (INIS)
Adams, T.F.; Castor, J.E.; Davis, C.G.
1978-01-01
The dynamically zoned pulsation code developed by Castor, Davis, and Davison has been used to recalculate the Goddard model and to calculate three other Cepheid models with the same period (9.8 days). This family of models shows how the bumps and other features of the light and velocity curves change as the mass is varied at constant period. This study, with a code that is capable of producing reliable light curves, shows again that the light and velocity curves for 9.8-day Cepheid models with standard homogeneous compositions do not show bumps like those that are observed unless the mass is significantly lower than the ''evolutionary mass.'' The light and velocity curves for the Goddard model presented here are similar to those computed independently by Fischel, Sparks, and Karp. They should be useful as standards for future investigators
When are network coding based dynamic multi-homing techniques beneficial?
DEFF Research Database (Denmark)
Pereira, Carlos; Aguiar, Ana; Roetter, Daniel Enrique Lucani
2016-01-01
Mechanisms that can cope with unreliable wireless channels in an efficient manner are required due to the increasing number of resource constrained devices. Concurrent use of multiple communications technologies can be instrumental towards improving services to mobile devices in heterogeneous...... networks. In our previous work, we developed an optimization framework to generate channel-aware transmission policies for multi-homed devices under different cost criteria. Our formulation considers network coding as a key technique that simplifies load allocation across multiple channels and provides...... high resiliency under time-varying channel conditions. This paper seeks to explore the parameter space and identify the operating regions where dynamic coded policies offer most improvement over static ones in terms of energy consumption and channel utilization. We leverage meta-heuristics to find...
A wide-range model of two-group gross sections in the dynamics code HEXTRAN
International Nuclear Information System (INIS)
Kaloinen, E.; Peltonen, J.
2002-01-01
In dynamic analyses the thermal hydraulic conditions within the reactor core may have a large variation, which sets a special requirement on the modeling of cross sections. The standard model in the dynamics code HEXTRAN is the same as in the static design code HEXBU-3D/MODS. It is based on a linear and second order fitting of two-group cross sections on fuel and moderator temperature, moderator density and boron density. A new, wide-range model of cross sections developed in Fortum Nuclear Services for HEXBU-3D/MOD6 has been included as an option into HEXTRAN. In this model the nodal cross sections are constructed from seven state variables in a polynomial of more than 40 terms. Coefficients of the polynomial are created by a least squares fitting to the results of a large number of fuel assembly calculations. Depending on the choice of state variables for the spectrum calculations, the new cross section model is capable to cover local conditions from cold zero power to boiling at full power. The 5. dynamic benchmark problem of AER is analyzed with the new option and results are compared to calculations with the standard model of cross sections in HEXTRAN (Authors)
Analysis of visual coding variables on CRT generated displays
International Nuclear Information System (INIS)
Blackman, H.S.; Gilmore, W.E.
1985-01-01
Cathode ray tube generated safety parameter display systems in a nuclear power plant control room situation have been found to be improved in effectiveness when color coding is employed. Research has indicated strong support for graphic coding techniques particularly in redundant coding schemes. In addition, findings on pictographs, as applied in coding schemes, indicate the need for careful application and for further research in the development of a standardized set of symbols
Fuel analysis code FAIR and its high burnup modelling capabilities
International Nuclear Information System (INIS)
Prasad, P.S.; Dutta, B.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.
1995-01-01
A computer code FAIR has been developed for analysing performance of water cooled reactor fuel pins. It is capable of analysing high burnup fuels. This code has recently been used for analysing ten high burnup fuel rods irradiated at Halden reactor. In the present paper, the code FAIR and its various high burnup models are described. The performance of code FAIR in analysing high burnup fuels and its other applications are highlighted. (author). 21 refs., 12 figs
Ethical Code Effectiveness in Football Clubs: A Longitudinal Analysis
Constandt, Bram; De Waegeneer, Els; Willem, Annick
2017-01-01
As football (soccer) clubs are facing different ethical challenges, many clubs are turning to ethical codes to counteract unethical behaviour. However, both in- and outside the sport field, uncertainty remains about the effectiveness of these ethical codes. For the first time, a longitudinal study design was adopted to evaluate code effectiveness. Specifically, a sample of non-professional football clubs formed the subject of our inquiry. Ethical code effectiveness was...
International Nuclear Information System (INIS)
Wang Rubin; Yu Wei
2005-01-01
In this paper, we investigate how the population of neuronal oscillators deals with information and the dynamic evolution of neural coding when the external stimulation acts on it. Numerically computing method is used to describe the evolution process of neural coding in three-dimensioned space. The numerical result proves that only the suitable stimulation can change the coupling structure and plasticity of neurons
Dynamic quality of service differentiation using fixed code weight in optical CDMA networks
Kakaee, Majid H.; Essa, Shawnim I.; Abd, Thanaa H.; Seyedzadeh, Saleh
2015-11-01
The emergence of network-driven applications, such as internet, video conferencing, and online gaming, brings in the need for a network the environments with capability of providing diverse Quality of Services (QoS). In this paper, a new code family of novel spreading sequences, called a Multi-Service (MS) code, has been constructed to support multiple services in Optical- Code Division Multiple Access (CDMA) system. The proposed method uses fixed weight for all services, however reducing the interfering codewords for the users requiring higher QoS. The performance of the proposed code is demonstrated using mathematical analysis. It shown that the total number of served users with satisfactory BER of 10-9 using NB=2 is 82, while they are only 36 and 10 when NB=3 and 4 respectively. The developed MS code is compared with variable-weight codes such as Variable Weight-Khazani Syed (VW-KS) and Multi-Weight-Random Diagonal (MW-RD). Different numbers of basic users (NB) are used to support triple-play services (audio, data and video) with different QoS requirements. Furthermore, reference to the BER of 10-12, 10-9, and 10-3 for video, data and audio, respectively, the system can support up to 45 total users. Hence, results show that the technique can clearly provide a relative QoS differentiation with lower value of basic users can support larger number of subscribers as well as better performance in terms of acceptable BER of 10-9 at fixed code weight.
International Nuclear Information System (INIS)
Tamitani, Masashi; Maruyama, Hiromi; Ishii, Kazuya; Izutsu, Sadayuki; Yamaguchi, Masao
2000-01-01
Critical experiments of UO 2 and full mixed oxide (MOX) fuel cores conducted at the Tank-type Critical Assembly (TCA) were analyzed using BWR design-purpose codes HINES and CERES with ENDF/B files and Monte Carlo fine analysis codes VMONT and MVP with the JENDL-3.2 library. The averaged values of the multiplication factors calculated with HINES/CERES, VMONT and MVP agreed with those of experiments within 0.3%Δk. The values by the design-purpose codes showed a small difference of 0.1%Δk between UO 2 and MOX cores. Monte Carlo code results showed that the JENDL-3.2 library had a tendency to overestimate the multiplication factors of UO 2 cores by about 0.3%Δk compared with those values of MOX cores. The root mean square errors of calculated power distributions were less than 1% for HINES/CERES and VMONT. These results showed that (1) the accuracy of these codes when applied to full MOX cores was almost the same as their accuracy for UO 2 cores, which confirmed the accuracy of present core design codes for full MOX cores; and (2) the accuracy of the 190-energy-group Monte Carlo calculation code VMONT was almost the same as that of the continuous-energy Monte Carlo calculation code MVP. (author)
Safety analysis of MOX fuels by fuel performance code
Energy Technology Data Exchange (ETDEWEB)
Suzuki, Motoe [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment
2002-12-01
Performance of plutonium rick mixed oxide fuels specified for the Reduced-Moderation Water Reactor (RMWR) has been analysed by modified fuel performance code. Thermodynamic properties of these fuels up to 120 GWd/t burnup have not been measured and estimated using existing uranium fuel models. Fission product release, pressure rise inside fuel rods and mechanical loads of fuel cans due to internal pressure have been preliminarily assessed based on assumed axial power distribution history, which show the integrity of fuel performance. Detailed evaluation of fuel-cladding interactions due to thermal expansion or swelling of fuel pellets due to high burnup will be required for safety analysis of mixed oxide fuels. Thermal conductivity and swelling of plutonium rich mixed oxide fuels shall be taken into consideration. (T. Tanaka)
Severe accident analysis using MARCH 1.0 code
International Nuclear Information System (INIS)
Guimaraes, A.C.F.
1987-09-01
The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt
DYNAVAC: a transient-vacuum-network analysis code
International Nuclear Information System (INIS)
Deis, G.A.
1980-01-01
This report discusses the structure and use of the program DYNAVAC, a new transient-vacuum-network analysis code implemented on the NMFECC CDC-7600 computer. DYNAVAC solves for the transient pressures in a network of up to twenty lumped volumes, interconnected in any configuration by specified conductances. Each volume can have an internal gas source, a pumping speed, and any initial pressure. The gas-source rates can vary with time in any piecewise-linear manner, and up to twenty different time variations can be included in a single problem. In addition, the pumping speed in each volume can vary with the total gas pumped in the volume, thus simulating the saturation of surface pumping. This report is intended to be both a general description and a user's manual for DYNAVAC
Energy Technology Data Exchange (ETDEWEB)
Salazar C, J.H.; Ramos P, J.C.; Salazar S, E.; Chavez M, C. [UNAM, Fac. de Ingenieria, Laboratorio de Analisis en Ingenieria de Reactores Nucleares, DEPFI Campus Morelos, Cuernavaca (Mexico)]. e-mail: cchavez2@cableonline.com.mx
2003-07-01
The present work illustrates the application of the concept of a simulator for analysis, design, instruction and training in a classroom environment associated to a nuclear power station. Emphasis is made on the methodology used to incorporate the best estimate codes RELAP/SCDAP to a prototype under development at the Nuclear Reactor Engineering Analysis Laboratory (NREAL). This methodology is based on a modular structure where multiple processes can be executed in an independent way and where the generated information is stored in shared memory segments and distributed by means of communication routines developed in the C programming language. The utility of the system is demonstrated using highly interactive graphics (mimic diagrams, pictorials and tendency graphs) for the simultaneous dynamic visualization of the most significant variables of a typical transient event (feed water controller failure in a BWR). A fundamental part of the system is its advanced graphic interface. This interface, of the type of direct manipulation, reproduces instruments and controls whose functionality is similar to those found in the current replica simulator for the Laguna Verde Nuclear Power Station. Finally the evaluation process is described. The general behavior of the main variables for the selected transitory event is interpreted, corroborating that they follow the same tendency that those reported for a BWR. The obtained results allow to conclude that the developed system works satisfactorily and that the use of al 1 x 1 real time visualization tools offers important advantages regarding other traditional methods of analysis. (Author)
Practical Analysis of the Dynamic Characteristics of JavaScript
Wei, Shiyi
2015-01-01
JavaScript is a dynamic object-oriented programming language, which is designed with flexible programming mechanisms. JavaScript is widely used in developing sophisticated software systems, especially web applications. Despite of its popularity, there is a lack of software tools that support JavaScript for software engineering clients. Dataflow analysis approximates software behavior by analyzing the program code; it is the foundation for many software tools. However, several unique features...
International Nuclear Information System (INIS)
Shi, Chengbin; Cheng, Maosong; Liu, Guimin
2016-01-01
Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center
Energy Technology Data Exchange (ETDEWEB)
Shi, Chengbin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Sciences, Beijing 100049 (China); Cheng, Maosong, E-mail: mscheng@sinap.ac.cn [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Liu, Guimin [Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)
2016-08-15
Highlights: • New point kinetics and thermo-hydraulics models as well as a numerical method are added into RELAP5 code to be suitable for liquid fueled molten salt reactor. • The extended REALP5 code is verified by the experimental benchmarks of MSRE. • The different transient scenarios of the MSBR are simulated to evaluate performance during the transients. - Abstract: The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes as inherent safety, economical efficiency, natural resource protection, sustainable development and nuclear non-proliferation. It is important to make system safety analysis for nuclear power plant of MSR. In this paper, in order to developing a system analysis code suitable for liquid fueled molten salt reactors, the point kinetics and thermo-hydraulic models as well as the numerical method in thermal–hydraulic transient code Reactor Excursion and Leak Analysis Program (RELAP5) developed at the Idaho National Engineering Laboratory (INEL) for the U.S. Nuclear Regulatory Commission (NRC) are extended and verified by Molten Salt Reactor Experiment (MSRE) experimental benchmarks. And then, four transient scenarios including the load demand change, the primary flow transient, the secondary flow transient and the reactivity transient of the Molten Salt Breeder Reactor (MSBR) are modeled and simulated so as to evaluate the performance of the reactor during the anticipated transient events using the extended RELAP5 code. The results indicate the extended RELAP5 code is effective and well suited to the liquid fueled molten salt reactor, and the MSBR has strong inherent safety characteristics because of its large negative reactivity coefficient. In the future, the extended RELAP5 code will be used to perform transient safety analysis for a liquid fueled thorium molten salt reactor named TMSR-LF developed by the Center
Introduction of the bubble rise dynamic model into the ALMOD 3 code pressurizer
International Nuclear Information System (INIS)
Madeira, A.A.; Camargo, C.T.M.
1985-01-01
A new evaporation model for the ALMOD 3 code pressurizer is implemented in order to estimate more accurately the water level behaviour and its influence in the pressure transient for very fast depressurization cases. For the inclusion of the bubble rise dynamic model it was necessary to consider a two-phase mixture in the water volume. The modifications don't require additional input data and virtually had not modified the processing time. The results and processing time for the original and the new models are presented. (F.E.) [pt
Directory of Open Access Journals (Sweden)
Y. M. Tumanov
2010-03-01
Full Text Available The article offers the method of malicious JavaScript code detection, based on behavior analysis. Conceptions of program behavior, program state, an algorithm of malicious code detection are described.
Design and Analysis of Adaptive Message Coding on LDPC Decoder with Faulty Storage
Directory of Open Access Journals (Sweden)
Guangjun Ge
2018-01-01
Full Text Available Unreliable message storage severely degrades the performance of LDPC decoders. This paper discusses the impacts of message errors on LDPC decoders and schemes improving the robustness. Firstly, we develop a discrete density evolution analysis for faulty LDPC decoders, which indicates that protecting the sign bits of messages is effective enough for finite-precision LDPC decoders. Secondly, we analyze the effects of quantization precision loss for static sign bit protection and propose an embedded dynamic coding scheme by adaptively employing the least significant bits (LSBs to protect the sign bits. Thirdly, we give a construction of Hamming product code for the adaptive coding and present low complexity decoding algorithms. Theoretic analysis indicates that the proposed scheme outperforms traditional triple modular redundancy (TMR scheme in decoding both threshold and residual errors, while Monte Carlo simulations show that the performance loss is less than 0.2 dB when the storage error probability varies from 10-3 to 10-4.
Development of integrated computer code for analysis of risk reduction strategy
International Nuclear Information System (INIS)
Kim, Dong Ha; Kim, See Darl; Kim, Hee Dong
2002-05-01
The development of the MIDAS/TH integrated severe accident code was performed in three main areas: 1) addition of new models derived from the national experimental programs and models for APR-1400 Korea next generation reactor, 2) improvement of the existing models using the recently available results, and 3) code restructuring for user friendliness. The unique MIDAS/TH models include: 1) a kinetics module for core power calculation during ATWS, 2) a gap cooling module between the molten corium pool and the reactor vessel wall, 3) a penetration tube failure module, 4) a PAR analysis module, and 5) a look-up table for the pressure and dynamic load during steam explosion. The improved models include: 1) a debris dispersal module considering the cavity geometry during DCH, 2) hydrogen burn and deflagration-to-detonation transition criteria, 3) a peak pressure estimation module for hydrogen detonation, and 4) the heat transfer module between the molten corium pool and the overlying water. The sparger and the ex-vessel heat transfer module were assessed. To enhance user friendliness, code restructuring was performed. In addition, a sample of severe accident analysis results was organized under the preliminary database structure
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
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
Turbine Airfoil Optimization Using Quasi-3D Analysis Codes
Directory of Open Access Journals (Sweden)
Sanjay Goel
2009-01-01
Full Text Available A new approach to optimize the geometry of a turbine airfoil by simultaneously designing multiple 2D sections of the airfoil is presented in this paper. The complexity of 3D geometry modeling is circumvented by generating multiple 2D airfoil sections and constraining their geometry in the radial direction using first- and second-order polynomials that ensure smoothness in the radial direction. The flow fields of candidate geometries obtained during optimization are evaluated using a quasi-3D, inviscid, CFD analysis code. An inviscid flow solver is used to reduce the execution time of the analysis. Multiple evaluation criteria based on the Mach number profile obtained from the analysis of each airfoil cross-section are used for computing a quality metric. A key contribution of the paper is the development of metrics that emulate the perception of the human designer in visually evaluating the Mach Number distribution. A mathematical representation of the evaluation criteria coupled with a parametric geometry generator enables the use of formal optimization techniques in the design. The proposed approach is implemented in the optimal design of a low-pressure turbine nozzle.
Evaluation of MOSTAS computer code for predicting dynamic loads in two bladed wind turbines
Kaza, K. R. V.; Janetzke, D. C.; Sullivan, T. L.
1979-01-01
Calculated dynamic blade loads were compared with measured loads over a range of yaw stiffnesses of the DOE/NASA Mod-O wind turbine to evaluate the performance of two versions of the MOSTAS computer code. The first version uses a time-averaged coefficient approximation in conjunction with a multi-blade coordinate transformation for two bladed rotors to solve the equations of motion by standard eigenanalysis. The second version accounts for periodic coefficients while solving the equations by a time history integration. A hypothetical three-degree of freedom dynamic model was investigated. The exact equations of motion of this model were solved using the Floquet-Lipunov method. The equations with time-averaged coefficients were solved by standard eigenanalysis.
International Nuclear Information System (INIS)
Chabard, J.P.; Viollet, P.L.
1991-08-01
Most of the computational fluid dynamics applications which are encountered at the Research and Development Division of EDF (RDD) are dealing with thermal exchanges. The development of numerical tools for the simulation of flows, devoted to this class of application, has been under way for 15 years. At the beginning this work was mainly concerned with a good simulation of the dynamics of the flow. Now these tools can be used to compute flows with thermal exchanges. The presentation will be limited to incompressible and one phase flows. First the softwares developed at RDD will be presented. Then some applications of these tools to flows with thermal exchanges will be discussed. To conclude, the paper will treat be general case of the CFD codes. The challenges for the next years will be detailed in order to make these tools available for users involved in complex physical modeling
International Nuclear Information System (INIS)
1984-01-01
Appropriate working rules, safety features and monitoring requirements for general X-ray analysis units or equipment are laid down in this Code which is intended for users of such equipment. The Code advises that establishments draw up their own working procedures based on appropriate legislation and on the recommendations contained in this Code. The Code also describes the requirements for X-ray analysis equipment necessary to ensure safety. (NEA) [fr
Anas, Siti Barirah Ahmad; Seyedzadeh, Saleh; Mokhtar, Makhfudzah; Sahbudin, Ratna Kalos Zakiah
2016-10-01
Future Internet consists of a wide spectrum of applications with different bit rates and quality of service (QoS) requirements. Prioritizing the services is essential to ensure that the delivery of information is at its best. Existing technologies have demonstrated how service differentiation techniques can be implemented in optical networks using data link and network layer operations. However, a physical layer approach can further improve system performance at a prescribed received signal quality by applying control at the bit level. This paper proposes a coding algorithm to support optical domain service differentiation using spectral amplitude coding techniques within an optical code division multiple access (OCDMA) scenario. A particular user or service has a varying weight applied to obtain the desired signal quality. The properties of the new code are compared with other OCDMA codes proposed for service differentiation. In addition, a mathematical model is developed for performance evaluation of the proposed code using two different detection techniques, namely direct decoding and complementary subtraction.
Generator dynamics in aeroelastic analysis and simulations
Energy Technology Data Exchange (ETDEWEB)
Larsen, T.J.; Hansen, M.H.; Iov, F.
2003-05-01
This report contains a description of a dynamic model for a doubly-fed induction generator implemented in the aeroelastic code HAWC. The model has physical input parameters (resistance, reactance etc.) and input variables (stator and rotor voltage and rotor speed). The model can be used to simulate the generator torque as well as the rotor and stator currents, active and reactive power. A perturbation method has been used to reduce the original generator model equations to a set of equations which can be solved with the same time steps as a typical aeroelastic code. The method is used to separate the fast transients of the model from the slow variations and deduce a reduced order expression for the slow part. Dynamic effects of the first order terms in the model as well as the influence on drive train eigenfrequencies and damping has been investigated. Load response during time simulation of wind turbine response have been compared to simulations with a linear static generator model originally implemented i HAWC. A 2 MW turbine has been modelled in the aeroelastic code HAWC. When using the new dynamic generator model there is an interesting coupling between the generator dynamics and a global turbine vibration mode at 4.5 Hz, which only occurs when a dynamic formulation of the generator equations is applied. This frequency can especially be seen in the electrical power of the generator and the rotational speed of the generator, but also as torque variations in the drive train. (au)
Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code
Energy Technology Data Exchange (ETDEWEB)
Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1997-12-31
The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.
Sensitivity analysis on various parameters for lattice analysis of DUPIC fuel with WIMS-AECL code
Energy Technology Data Exchange (ETDEWEB)
Roh, Gyu Hong; Choi, Hang Bok; Park, Jee Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1998-12-31
The code WIMS-AECL has been used for the lattice analysis of DUPIC fuel. The lattice parameters calculated by the code is sensitive to the choice of number of parameters, such as the number of tracking lines, number of condensed groups, mesh spacing in the moderator region, other parameters vital to the calculation of probabilities and burnup analysis. We have studied this sensitivity with respect to these parameters and recommend their proper values which are necessary for carrying out the lattice analysis of DUPIC fuel.
High-performance computational fluid dynamics: a custom-code approach
International Nuclear Information System (INIS)
Fannon, James; Náraigh, Lennon Ó; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain
2016-01-01
We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier–Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing. (paper)
High-performance computational fluid dynamics: a custom-code approach
Fannon, James; Loiseau, Jean-Christophe; Valluri, Prashant; Bethune, Iain; Náraigh, Lennon Ó.
2016-07-01
We introduce a modified and simplified version of the pre-existing fully parallelized three-dimensional Navier-Stokes flow solver known as TPLS. We demonstrate how the simplified version can be used as a pedagogical tool for the study of computational fluid dynamics (CFDs) and parallel computing. TPLS is at its heart a two-phase flow solver, and uses calls to a range of external libraries to accelerate its performance. However, in the present context we narrow the focus of the study to basic hydrodynamics and parallel computing techniques, and the code is therefore simplified and modified to simulate pressure-driven single-phase flow in a channel, using only relatively simple Fortran 90 code with MPI parallelization, but no calls to any other external libraries. The modified code is analysed in order to both validate its accuracy and investigate its scalability up to 1000 CPU cores. Simulations are performed for several benchmark cases in pressure-driven channel flow, including a turbulent simulation, wherein the turbulence is incorporated via the large-eddy simulation technique. The work may be of use to advanced undergraduate and graduate students as an introductory study in CFDs, while also providing insight for those interested in more general aspects of high-performance computing.
Nonlinear dynamics of laser systems with elements of a chaos: Advanced computational code
Buyadzhi, V. V.; Glushkov, A. V.; Khetselius, O. Yu; Kuznetsova, A. A.; Buyadzhi, A. A.; Prepelitsa, G. P.; Ternovsky, V. B.
2017-10-01
A general, uniform chaos-geometric computational approach to analysis, modelling and prediction of the non-linear dynamics of quantum and laser systems (laser and quantum generators system etc) with elements of the deterministic chaos is briefly presented. The approach is based on using the advanced generalized techniques such as the wavelet analysis, multi-fractal formalism, mutual information approach, correlation integral analysis, false nearest neighbour algorithm, the Lyapunov’s exponents analysis, and surrogate data method, prediction models etc There are firstly presented the numerical data on the topological and dynamical invariants (in particular, the correlation, embedding, Kaplan-York dimensions, the Lyapunov’s exponents, Kolmogorov’s entropy and other parameters) for laser system (the semiconductor GaAs/GaAlAs laser with a retarded feedback) dynamics in a chaotic and hyperchaotic regimes.
A dynamical systems approach to characterizing the contribution of neurogenesis to neural coding
Directory of Open Access Journals (Sweden)
Merav Stern
2014-03-01
index reaches a maximum when young neurons having hyper-excitability ratio of ~4 comprised ~2% of the population. This agrees with experimental estimates (Cameron and McKay, 2001; Deng et al., 2010; Spalding et al., 2013; Tashiro et al., 2007 without any adjustable parameters in the model. Figure 2. Computational analysis of distributed coding in heterogeneous networks. Networks can be efficiently trained only in the regime where prior to training to reproduce target input patterns they exhibit chaotic dynamics (Sussillo and Abbott, 2009. For a model network based on one type of rate neurons (Sompolinsky et al., 1988, the transition to chaotic dynamics occurs when at least some modes in the network respond to perturbation with exponents (eigenvalues that have real parts > 1 (purple line. Imaginary parts indicate oscillatory dynamics along the respective modes, and are not relevant indicators of chaotic dynamics. Our analytic estimate for the limits of exponents (blue circle matches the numerical simulation (small open circles, each circle is a separate mode. In contrast, predictions based on average synaptic weights (red circle are not accurate. This example network is in the chaotic regime prior to training, because some modes have exponents with real parts > 1. The corresponding neural responses over time from different types of neurons (group 1 and group 2 are shown on the right.
International Nuclear Information System (INIS)
Nichols, B.D.; Mueller, C.; Necker, G.A.; Travis, J.R.; Spore, J.W.; Lam, K.L.; Royl, P.; Redlinger, R.; Wilson, T.L.
1998-01-01
Los Alamos National Laboratory (LANL) and Forschungszentrum Karlsruhe (FzK) are developing GASFLOW, a three-dimensional (3D) fluid dynamics field code as a best-estimate tool to characterize local phenomena within a flow field. Examples of 3D phenomena include circulation patterns; flow stratification; hydrogen distribution mixing and stratification; combustion and flame propagation; effects of noncondensable gas distribution on local condensation and evaporation; and aerosol entrainment, transport, and deposition. An analysis with GASFLOW will result in a prediction of the gas composition and discrete particle distribution in space and time throughout the facility and the resulting pressure and temperature loadings on the walls and internal structures with or without combustion. A major application of GASFLOW is for predicting the transport, mixing, and combustion of hydrogen and other gases in nuclear reactor containments and other facilities. It has been applied to situations involving transporting and distributing combustible gas mixtures. It has been used to study gas dynamic behavior (1) in low-speed, buoyancy-driven flows, as well as sonic flows or diffusion dominated flows; and (2) during chemically reacting flows, including deflagrations. The effects of controlling such mixtures by safety systems can be analyzed. The code version described in this manual is designated GASFLOW 2.1, which combines previous versions of the United States Nuclear Regulatory Commission code HMS (for Hydrogen Mixing Studies) and the Department of Energy and FzK versions of GASFLOW. The code was written in standard Fortran 90. This manual comprises three volumes. Volume I describes the governing physical equations and computational model. Volume II describes how to use the code to set up a model geometry, specify gas species and material properties, define initial and boundary conditions, and specify different outputs, especially graphical displays. Sample problems are included
Vehicle systems: coupled and interactive dynamics analysis
Vantsevich, Vladimir V.
2014-11-01
This article formulates a new direction in vehicle dynamics, described as coupled and interactive vehicle system dynamics. Formalised procedures and analysis of case studies are presented. An analytical consideration, which explains the physics of coupled system dynamics and its consequences for dynamics of a vehicle, is given for several sets of systems including: (i) driveline and suspension of a 6×6 truck, (ii) a brake mechanism and a limited slip differential of a drive axle and (iii) a 4×4 vehicle steering system and driveline system. The article introduces a formal procedure to turn coupled system dynamics into interactive dynamics of systems. A new research direction in interactive dynamics of an active steering and a hybrid-electric power transmitting unit is presented and analysed to control power distribution between the drive axles of a 4×4 vehicle. A control strategy integrates energy efficiency and lateral dynamics by decoupling dynamics of the two systems thus forming their interactive dynamics.
Energy Technology Data Exchange (ETDEWEB)
Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)
2011-06-01
This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the
Uncertainty and sensitivity analysis using probabilistic system assessment code. 1
International Nuclear Information System (INIS)
Honma, Toshimitsu; Sasahara, Takashi.
1993-10-01
This report presents the results obtained when applying the probabilistic system assessment code under development to the PSACOIN Level 0 intercomparison exercise organized by the Probabilistic System Assessment Code User Group in the Nuclear Energy Agency (NEA) of OECD. This exercise is one of a series designed to compare and verify probabilistic codes in the performance assessment of geological radioactive waste disposal facilities. The computations were performed using the Monte Carlo sampling code PREP and post-processor code USAMO. The submodels in the waste disposal system were described and coded with the specification of the exercise. Besides the results required for the exercise, further additional uncertainty and sensitivity analyses were performed and the details of these are also included. (author)
Dynamical analysis of highly excited molecular spectra
Energy Technology Data Exchange (ETDEWEB)
Kellman, M.E. [Univ. of Oregon, Eugene (United States)
1993-12-01
The goal of this program is new methods for analysis of spectra and dynamics of highly excited vibrational states of molecules. In these systems, strong mode coupling and anharmonicity give rise to complicated classical dynamics, and make the simple normal modes analysis unsatisfactory. New methods of spectral analysis, pattern recognition, and assignment are sought using techniques of nonlinear dynamics including bifurcation theory, phase space classification, and quantization of phase space structures. The emphasis is chaotic systems and systems with many degrees of freedom.
Gouge, Michael F.
2011-01-01
Hypervelocity impact tests on test satellites are performed by members of the orbital debris scientific community in order to understand and typify the on-orbit collision breakup process. By analysis of these test satellite fragments, the fragment size and mass distributions are derived and incorporated into various orbital debris models. These same fragments are currently being put to new use using emerging technologies. Digital models of these fragments are created using a laser scanner. A group of computer programs referred to as the Fragment Rotation Analysis and Lightcurve code uses these digital representations in a multitude of ways that describe, measure, and model on-orbit fragments and fragment behavior. The Dynamic Rotation subroutine generates all of the possible reflected intensities from a scanned fragment as if it were observed to rotate dynamically while in orbit about the Earth. This calls an additional subroutine that graphically displays the intensities and the resulting frequency of those intensities as a range of solar phase angles in a Probability Density Function plot. This document reports the additions and modifications to the subset of the Fragment Rotation Analysis and Lightcurve concerned with the Dynamic Rotation and Probability Density Function plotting subroutines.
Transient analysis of ABWR reactor using a best estimate code
International Nuclear Information System (INIS)
Mizokami, S.; Kitamura, H.; Mototani, A.; Ono, H.
2004-01-01
Since the recirculation pumps are mounted internally within the ABWR, core flow will decrease rapidly in the event of a loss of their driving force. A rapid reduction in core flow may cause the onset of boiling transition (BT). Therefore, in order to prevent the onset of BT, a motor-generator (MG) set is added to the power supply system of the reactor internal pump (RIP). Recent studies, however, have shown that dryout within a fuel assembly over a short time period will result in only a small rise in fuel cladding temperature and thus does not pose a threat to fuel integrity. In response to this finding, the standards committee of the Atomic Energy Society of Japan (AESJ) has proposed a post-BT standard which incorporates a cladding temperature criterion. If it is assumed that the MG-set is not added to the RIP power supply system, the result of the safety analysis shows the onset of BT with a subsequent rise in fuel cladding temperature. Although BT occurs under the conservative assumptions of this safety analysis, a possibility exists that BT will not occur under actual operating conditions. The best estimate code TRACG was used to show that BT does not occur and that fuel integrity can be sufficiently maintained under actual conditions. (author)
Analysis of fuel-coolant interaction with VAPEX code
International Nuclear Information System (INIS)
Melikhov, O.I.; Melikhov, V.I.; Sokolin, A.V.; Yakush, S.E.
2004-01-01
The analysis of the FARO L-33 test has been carried out with the VAPEX code in which a submodel for hydrogen release and transport was implemented. The FARO test was aimed at studying the premixing and quenching processes for large (about 100 kg) masses of corium. The specific features of the FARO L-33 test are: high subcooling (124 K), low pressure (4.1 bar), presence of non-condensable gas (argon) and triggered vapor explosion when melt reached the bottom of the vessel. A numerical simulation of FARO L-33 test was carried out using 2-D nodalization. The fragmentation model is based on the Saito correlation. The model for hydrogen release assumes direct proportionality between the total hydrogen mass release rate and the total fragmentation rate of the melt jet. The proportionality constant was taken from the experimental estimates for test conditions. Calculation of the premixing stage gave some delay in the pressure growth, which is most probably connected with inadequacy of the fragmentation model at the initial stage of melt jet-water interaction. The calculated pressurization rate, however, agrees reasonably with the measured one. Modeling of vapor explosion, which occurred in the test, yielded reasonable correlation with the test data when hydrogen formation was taken into account. Thus, VAPEX analysis of the FARO L-33 test has shown reasonable agreement between the experimental and calculated data. (author)
Analysis of Memory Codes and Cumulative Rehearsal in Observational Learning
Bandura, Albert; And Others
1974-01-01
The present study examined the influence of memory codes varying in meaningfulness and retrievability and cumulative rehearsal on retention of observationally learned responses over increasing temporal intervals. (Editor)
TRANSURANUS: A fuel rod analysis code ready for use
Energy Technology Data Exchange (ETDEWEB)
Lassmann, K; O` Carroll, C; Van de Laar, J [Commission of the European Communities, Karlsruhe (Germany). European Inst. for Transuranium Elements; Ott, C [Paul Scherrer Inst. (PSI), Villigen (Switzerland)
1994-12-31
The basic concepts of fuel rod performance codes are discussed. The TRANSURANUS code developed at the Institute for Transuranium Elements, Karlsruhe (GE) is presented. It is a quasi two-dimensional (1{sub 1/2}-D) code designed for treatment of a whole fuel rod for any type of reactor and any situation. The fuel rods found in the majority of test- or power reactors can be analyzed for very different situations (normal, off-normal and accidental). The time scale of the problems to be treated may range from milliseconds to years. The TRANSURANUS code consists of a clearly defined mechanical/mathematical framework into which physical models can easily be incorporated. This framework has been extensively tested and the programming very clearly reflects this structure. The code is well structured and easy to understand. It has a comprehensive material data bank for different fuels, claddings, coolants and their properties. The code can be employed in a deterministic and a statistical version. It is written in standard FORTRAN 77. The code system includes: 2 preprocessor programs (MAKROH and AXORDER) for setting up new data cases; the post-processor URPLOT for plotting all important quantities as a function of the radius, the axial coordinate or the time; the post-processor URSTART evaluating statistical analyses. The TRANSURANUS code exhibits short running times. A new WINDOWS-based interactive interface is under development. The code is now in use in various European institutions and is available to all interested parties. 7 figs., 15 refs.
Energy Technology Data Exchange (ETDEWEB)
Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)
1995-08-01
The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)
Code development for eigenvalue total sensitivity analysis and total uncertainty analysis
International Nuclear Information System (INIS)
Wan, Chenghui; Cao, Liangzhi; Wu, Hongchun; Zu, Tiejun; Shen, Wei
2015-01-01
Highlights: • We develop a new code for total sensitivity and uncertainty analysis. • The implicit effects of cross sections can be considered. • The results of our code agree well with TSUNAMI-1D. • Detailed analysis for origins of implicit effects is performed. - Abstract: The uncertainties of multigroup cross sections notably impact eigenvalue of neutron-transport equation. We report on a total sensitivity analysis and total uncertainty analysis code named UNICORN that has been developed by applying the direct numerical perturbation method and statistical sampling method. In order to consider the contributions of various basic cross sections and the implicit effects which are indirect results of multigroup cross sections through resonance self-shielding calculation, an improved multigroup cross-section perturbation model is developed. The DRAGON 4.0 code, with application of WIMSD-4 format library, is used by UNICORN to carry out the resonance self-shielding and neutron-transport calculations. In addition, the bootstrap technique has been applied to the statistical sampling method in UNICORN to obtain much steadier and more reliable uncertainty results. The UNICORN code has been verified against TSUNAMI-1D by analyzing the case of TMI-1 pin-cell. The numerical results show that the total uncertainty of eigenvalue caused by cross sections can reach up to be about 0.72%. Therefore the contributions of the basic cross sections and their implicit effects are not negligible
Design of an Object-Oriented Turbomachinery Analysis Code: Initial Results
Jones, Scott M.
2015-01-01
Performance prediction of turbomachines is a significant part of aircraft propulsion design. In the conceptual design stage, there is an important need to quantify compressor and turbine aerodynamic performance and develop initial geometry parameters at the 2-D level prior to more extensive Computational Fluid Dynamics (CFD) analyses. The Object-oriented Turbomachinery Analysis Code (OTAC) is being developed to perform 2-D meridional flowthrough analysis of turbomachines using an implicit formulation of the governing equations to solve for the conditions at the exit of each blade row. OTAC is designed to perform meanline or streamline calculations; for streamline analyses simple radial equilibrium is used as a governing equation to solve for spanwise property variations. While the goal for OTAC is to allow simulation of physical effects and architectural features unavailable in other existing codes, it must first prove capable of performing calculations for conventional turbomachines. OTAC is being developed using the interpreted language features available in the Numerical Propulsion System Simulation (NPSS) code described by Claus et al (1991). Using the NPSS framework came with several distinct advantages, including access to the pre-existing NPSS thermodynamic property packages and the NPSS Newton-Raphson solver. The remaining objects necessary for OTAC were written in the NPSS framework interpreted language. These new objects form the core of OTAC and are the BladeRow, BladeSegment, TransitionSection, Expander, Reducer, and OTACstart Elements. The BladeRow and BladeSegment consumed the initial bulk of the development effort and required determining the equations applicable to flow through turbomachinery blade rows given specific assumptions about the nature of that flow. Once these objects were completed, OTAC was tested and found to agree with existing solutions from other codes; these tests included various meanline and streamline comparisons of axial
Energy Technology Data Exchange (ETDEWEB)
Guihot, P.; Devesa, G.; Dumond, A.; Panet, M.; Waeckel, F.
1996-12-31
Progress in the field of seismic qualification and design methods made these last few years allows physical phenomena actually in play to be better considered, while cutting down the conservatism associated with some simplified design methods. So following the change in methods and developing the most advantageous ones among them contributes to the process of the seismic margins assessment and the preparation of new design tools for future series. In this paper, the main developments and improvements in methods which have been made these last two years in the Code Aster, in order to improve seismic calculation methods and seismic margin assessment are presented. The first development relates to making the MISS3D soil structure interaction code available, thanks to an interface made with the Code Aster. The second relates to the possibility of making modal basis time calculations on multi-supported structures by considering local non linearities like impact, friction or squeeze fluid forces. Recent developments in random dynamics and postprocessing devoted to earthquake designs are then mentioned. Three applications of these developments are then ut forward. The first application relates to a test case for soil structure interaction design using MISS3D-Aster coupling. The second is a test case for a multi-supported structure. The last application, more for manufacturing, refers to seismic qualification of Main Live Steam stop valves. First results of the independent validation of the Code Aster seismic design functionalities, which provide and improve the quality of software, are also recalled. (authors). 11 refs.
SWAAM-LT: The long-term, sodium/water reaction analysis method computer code
International Nuclear Information System (INIS)
Shin, Y.W.; Chung, H.H.; Wiedermann, A.H.; Tanabe, H.
1993-01-01
The SWAAM-LT Code, developed for analysis of long-term effects of sodium/water reactions, is discussed. The theoretical formulation of the code is described, including the introduction of system matrices for ease of computer programming as a general system code. Also, some typical results of the code predictions for available large scale tests are presented. Test data for the steam generator design with the cover-gas feature and without the cover-gas feature are available and analyzed. The capabilities and limitations of the code are then discussed in light of the comparison between the code prediction and the test data
International Nuclear Information System (INIS)
Cevolani, S.
1995-01-01
The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of such a code was made possible by two facts: firstly, the increase, in the computing power of the desk machines; secondly, the fact that several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes. (author)
ANTEO: An optimised PC computer code for the steady state thermal hydraulic analysis of rod bundles
International Nuclear Information System (INIS)
Cevolani, S.
1996-07-01
The paper deals with the description of a Personal Computer oriented subchannel code, devoted to the steady state thermal hydraulic analysis of nuclear reactor fuel bundles. The development of a such code was made possible by two facts: first, the increase the computing power of the desk machines; secondly, the fact several years of experience into operate subchannels codes have shown how to simplify many of the physical models without a sensible loss of accuracy. For sake of validation, the developed code was compared with a traditional subchannel code, the COBRA one. The results of the comparison show a very good agreement between the two codes
Performance Analysis of Optical Code Division Multiplex System
Kaur, Sandeep; Bhatia, Kamaljit Singh
2013-12-01
This paper presents the Pseudo-Orthogonal Code generator for Optical Code Division Multiple Access (OCDMA) system which helps to reduce the need of bandwidth expansion and improve spectral efficiency. In this paper we investigate the performance of multi-user OCDMA system to achieve data rate more than 1 Tbit/s.
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.
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.
Nonlinear dynamic analysis of nuclear reactor primary coolant systems
International Nuclear Information System (INIS)
Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.
1979-01-01
The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development
Sample problem manual for benchmarking of cask analysis codes
International Nuclear Information System (INIS)
Glass, R.E.
1988-02-01
A series of problems have been defined to evaluate structural and thermal codes. These problems were designed to simulate the hypothetical accident conditions given in Title 10 of the Code of Federal Regulation, Part 71 (10CFR71) while retaining simple geometries. This produced a problem set that exercises the ability of the codes to model pertinent physical phenomena without requiring extensive use of computer resources. The solutions that are presented are consensus solutions based on computer analyses done by both national laboratories and industry in the United States, United Kingdom, France, Italy, Sweden, and Japan. The intent of this manual is to provide code users with a set of standard structural and thermal problems and solutions which can be used to evaluate individual codes. 19 refs., 19 figs., 14 tabs
ANDREA: Advanced nodal diffusion code for reactor analysis
International Nuclear Information System (INIS)
Belac, J.; Josek, R.; Klecka, L.; Stary, V.; Vocka, R.
2005-01-01
A new macro code is being developed at NRI which will allow coupling of the advanced thermal-hydraulics model with neutronics calculations as well as efficient use in core loading pattern optimization process. This paper describes the current stage of the macro code development. The core simulator is based on the nodal expansion method, Helios lattice code is used for few group libraries preparation. Standard features such as pin wise power reconstruction and feedback iterations on critical control rod position, boron concentration and reactor power are implemented. A special attention is paid to the system and code modularity in order to enable flexible and easy implementation of new features in future. Precision of the methods used in the macro code has been verified on available benchmarks. Testing against Temelin PWR operational data is under way (Authors)
MMA, A Computer Code for Multi-Model Analysis
Poeter, Eileen P.; Hill, Mary C.
2007-01-01
This report documents the Multi-Model Analysis (MMA) computer code. MMA can be used to evaluate results from alternative models of a single system using the same set of observations for all models. As long as the observations, the observation weighting, and system being represented are the same, the models can differ in nearly any way imaginable. For example, they may include different processes, different simulation software, different temporal definitions (for example, steady-state and transient models could be considered), and so on. The multiple models need to be calibrated by nonlinear regression. Calibration of the individual models needs to be completed before application of MMA. MMA can be used to rank models and calculate posterior model probabilities. These can be used to (1) determine the relative importance of the characteristics embodied in the alternative models, (2) calculate model-averaged parameter estimates and predictions, and (3) quantify the uncertainty of parameter estimates and predictions in a way that integrates the variations represented by the alternative models. There is a lack of consensus on what model analysis methods are best, so MMA provides four default methods. Two are based on Kullback-Leibler information, and use the AIC (Akaike Information Criterion) or AICc (second-order-bias-corrected AIC) model discrimination criteria. The other two default methods are the BIC (Bayesian Information Criterion) and the KIC (Kashyap Information Criterion) model discrimination criteria. Use of the KIC criterion is equivalent to using the maximum-likelihood Bayesian model averaging (MLBMA) method. AIC, AICc, and BIC can be derived from Frequentist or Bayesian arguments. The default methods based on Kullback-Leibler information have a number of theoretical advantages, including that they tend to favor more complicated models as more data become available than do the other methods, which makes sense in many situations. Many applications of MMA will
Structural dynamics in fast reactor accident analysis
International Nuclear Information System (INIS)
Fistedis, S.H.
1975-01-01
Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code
Dynamic analysis of embedded structures
International Nuclear Information System (INIS)
Kausel, E.; Whitman, R.V.; Morray, J.P.
1977-01-01
The paper presents simplified rules to account for embeddment and soil layering in the soil-structure interaction problem, to be used in dynamic analysis. The relationship between the spring method, and a direct solution (in which both soil and structure are modeled with finite elements and linear members) is first presented. It is shown that for consistency of the results with the two solution methods the spring method should be performed in the following three steps: 1. Determination of the motion of the massless foundation (having the same shape as the actual one) when subjected to the same input motion as the direct solution. 2. Determination of the frequency dependent subgrade stiffness for the relevant degrees of freedom. 3. Computations of the response of the real structure supported on frequency dependent soil springs and subjected at the base of these springs to the motion computed in step 1. The first two steps require, in general, finite element methods, which would make the procedure not attractive. It is shown in the paper, however, that excellent approximations can be obtained, on the basis of 1-dimensional wave propagation theory for the solution of step 1, and correction factors modifying for embeddment the corresponding springs of a surface footing on a layered stratum, for the solution of step 2. (Auth.)
International Nuclear Information System (INIS)
Ceuca, S.C.; Herb, J.; Schoeffel, P.J.; Hollands, T.; Austregesilo, H.; Hristov, H.V.
2017-01-01
The realistic numerical prediction of transient fluid-dynamic scenarios including the complex, three-dimensional flow mixing phenomena occurring in the reactor pressure vessel (RPV) both in normal or abnormal operation are an important issue in today's reactor safety assessment studies. Both Computational Fluid Dynamics (CFD) tools as well as fluid-dynamic system analysis codes, each with its advantages and drawbacks, are commonly used to model such transients. Simulation results obtained with the open-source CFD tool-box OpenFOAM and the German thermal-hydraulic system code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients), the later developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) for the analysis of the whole spectrum of operational transients, design-basis accidents and beyond design basis accidents anticipated for nuclear energy facilities, are compared against experimental data from the ROssendorf Coolant Mixing (ROCOM) test facility. In the case of the OpenFOAM CFD simulations the influence of various turbulence models and numerical schemes has been assessed while in the case of the system analysis code ATHLET a multidimensional nodalization recommended for real power plant applications has been employed. The simulation results show a good agreement with the experimental data, indicating that both OpenFOAM and ATHLET can capture the key flow features of the mixing processes in the Reactor Pressure Vessel (RPV). (author)
Safety analysis and code development for nuclear fuel cycle facilities
Energy Technology Data Exchange (ETDEWEB)
NONE
2013-08-15
We are estimating that the debris containing fuel are piled in the containment and the pressure vessel bottoms of Fukushima-Daiichi NPPs. A radioactive Xe concentration discharged in recriticality is being monitored by utilizing the gas management system set up in NPPs unit 1-3. For this reason, we can confirm the recriticality might not be broken out. However, the debris conditions distributed in the containment vessel and the pressure vessel bottoms are not clear. The internal and external surrounding changes will make recriticality become possible. According to TEPCO's roadmap, TEPCO will launch extracting task within 10 years. Even in the case that the fuel condition changes due to debris relocation and mixture, subcriticality must be secured. Criticality safety analysis with non-uniform effect should therefore be essential for the molten debris. For above reasons, we studies the optimum distributions with some parameters that have a large reactivity change were assessed with OPT-DANT code. Finally, the boron concentration was estimated in order to keep subcriticality. (author)
Kuosheng BWR/6 containment safety analysis with gothic code
International Nuclear Information System (INIS)
Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan
2011-01-01
Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)
Application of best estimate thermalhydraulic codes for the safety analysis of research reactors
International Nuclear Information System (INIS)
Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.
2006-01-01
An established international expertise in relation to computational tools, procedures for their application including Best Estimate (BE) methods supported by uncertainty evaluation, and comprehensive experimental database exists within the safety technology of Nuclear Power Plant (NPP). The importance of transferring NPP safety technology tools and methods to RR safety technology has been noted in recent IAEA activities. However, the ranges of parameters of interest to RR are different from those for NPP: this is namely true for fuel composition, system pressure, adopted materials and overall system geometric configuration. The large variety of research reactors prevented so far the achievement of systematic and detailed lists of initiating events based upon qualified Probabilistic Safety Assessment (PSA) studies with results endorsed by the international community. However, bounding and generalized lists of events are available from IAEA documents and can be considered for deeper studies in the area. In the area of acceptance criteria, established standards accepted by the international community are available. Therefore no major effort is needed, but an effort appears worthwhile to check that those standards are adopted and that the related thresholds are fulfilled. The importance of suitable experimental assessment is recognized. A large amount of data exists as the kinetic dynamic core behaviour form SPERT reactors tests. However, not all data are accessible to all institutions and the relationship between the range of parameters of experiments and the range of parameters relevant to RR technology is not always established. However, code-assessment through relevant set of experimental data are recorded and properly stored. An established technology exists for development, qualification and application of system thermal-hydraulics codes suitable to be adopted for accident analysis in research reactors. This derives from NPP technology. The applicability of
Computer code for the atomistic simulation of lattice defects and dynamics
International Nuclear Information System (INIS)
Schiffgens, J.O.; Graves, N.J.; Oster, C.A.
1980-04-01
The computer code COMENT used to simulate the effect of temperature, impurities, and pre-existing defects on radiation-induced defect production mechanisms, defect properties, defect migration, and defect stability. This report documents Version IV of COMENT (models, methods, and implementation) and defines current code options. Version IV of COMENT generates only face-centered-cubic (fcc) crystal lattices. However, an effort was made to structure COMENT to allow addition of new options with a minimum of change in the existing version of the code. This document describes the calling program and thirty-two user-defined subroutines. Fourteen subroutines (ALOYORD, DASPKA, DFCT, DSLOAN, DSLOIN, EXPAND, POT1, POT2, POT3, POT4, POT5, POT6, POT7, and THRMAL) are associated with the selection of program options; only a few of these are used in any given analysis. Seven of the other subroutines (CRYSTL, IEAF, INCBOX, LABLE, MINILAT, SPEFORS, and SQUEZ are used to establish a variety of arrays and conditions required for each analysis; most of them are used once in a given calculation. The remaining eleven subroutines (DAMP, DIRECT, IDDEF, NEAF, INBIN, FILBIN, FTBIN, PAC3, UNPAC3, PACF, and UNPACF) are used many times in each calculation; the last eight of these are used many times in each time step during the integration and, therefore, are written in COMPASS (CDC assembly language). The COMPASS subroutines are described in sufficient detail to permit easy conversion to some other assembly language or to FORTRAN
Energy Technology Data Exchange (ETDEWEB)
Lee, Won Jae; Chung, Bub Dong; Jeong, Jae Jun; Ha, Kwi Seok [Korea Atomic Energy Research Institute, Taejon (Korea)
1998-06-01
A multi-dimensional realistic thermal-hydraulic system analysis code, MARS version 1.3 has been developed. Main purpose of MARS 1.3 development is to have the realistic analysis capability of transient two-phase thermal-hydraulics of Pressurized Water Reactors (PWRs) especially during Large Break Loss of Coolant Accidents (LBLOCAs) where the multi-dimensional phenomena domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, domain the transients. MARS code is a unified version of USNRC developed COBRA-TF, three-dimensional (3D) reactor vessel analysis code, and RELAP5/MOD3.2.1.2, one-dimensional (1D) reactor system analysis code., Developmental requirements for MARS are chosen not only to best utilize the existing capability of the codes but also to have the enhanced capability in code maintenance, user accessibility, user friendliness, code portability, code readability, and code flexibility. For the maintenance of existing codes capability and the enhancement of code maintenance capability, user accessibility and user friendliness, MARS has been unified to be a single code consisting of 1D module (RELAP5) and 3D module (COBRA-TF). This is realized by implicitly integrating the system pressure matrix equations of hydrodynamic models and solving them simultaneously, by modifying the 1D/3D calculation sequence operable under a single Central Processor Unit (CPU) and by unifying the input structure and the light water property routines of both modules. In addition, the code structure of 1D module is completely restructured using the modular data structure of standard FORTRAN 90, which greatly improves the code maintenance capability, readability and portability. For the code flexibility, a dynamic memory management scheme is applied in both modules. MARS 1.3 now runs on PC/Windows and HP/UNIX platforms having a single CPU, and users have the options to select the 3D module to model the 3D thermal-hydraulics in the reactor vessel or other
Meyn, Larry A.
2018-01-01
One of the goals of NASA's Revolutionary Vertical Lift Technology Project (RVLT) is to provide validated tools for multidisciplinary design, analysis and optimization (MDAO) of vertical lift vehicles. As part of this effort, the software package, RotorCraft Optimization Tools (RCOTOOLS), is being developed to facilitate incorporating key rotorcraft conceptual design codes into optimizations using the OpenMDAO multi-disciplinary optimization framework written in Python. RCOTOOLS, also written in Python, currently supports the incorporation of the NASA Design and Analysis of RotorCraft (NDARC) vehicle sizing tool and the Comprehensive Analytical Model of Rotorcraft Aerodynamics and Dynamics II (CAMRAD II) analysis tool into OpenMDAO-driven optimizations. Both of these tools use detailed, file-based inputs and outputs, so RCOTOOLS provides software wrappers to update input files with new design variable values, execute these codes and then extract specific response variable values from the file outputs. These wrappers are designed to be flexible and easy to use. RCOTOOLS also provides several utilities to aid in optimization model development, including Graphical User Interface (GUI) tools for browsing input and output files in order to identify text strings that are used to identify specific variables as optimization input and response variables. This paper provides an overview of RCOTOOLS and its use
Performance Analysis of Faulty Gallager-B Decoding of QC-LDPC Codes with Applications
Directory of Open Access Journals (Sweden)
O. Al Rasheed
2014-06-01
Full Text Available In this paper we evaluate the performance of Gallager-B algorithm, used for decoding low-density parity-check (LDPC codes, under unreliable message computation. Our analysis is restricted to LDPC codes constructed from circular matrices (QC-LDPC codes. Using Monte Carlo simulation we investigate the effects of different code parameters on coding system performance, under a binary symmetric communication channel and independent transient faults model. One possible application of the presented analysis in designing memory architecture with unreliable components is considered.
Normalized value coding explains dynamic adaptation in the human valuation process.
Khaw, Mel W; Glimcher, Paul W; Louie, Kenway
2017-11-28
The notion of subjective value is central to choice theories in ecology, economics, and psychology, serving as an integrated decision variable by which options are compared. Subjective value is often assumed to be an absolute quantity, determined in a static manner by the properties of an individual option. Recent neurobiological studies, however, have shown that neural value coding dynamically adapts to the statistics of the recent reward environment, introducing an intrinsic temporal context dependence into the neural representation of value. Whether valuation exhibits this kind of dynamic adaptation at the behavioral level is unknown. Here, we show that the valuation process in human subjects adapts to the history of previous values, with current valuations varying inversely with the average value of recently observed items. The dynamics of this adaptive valuation are captured by divisive normalization, linking these temporal context effects to spatial context effects in decision making as well as spatial and temporal context effects in perception. These findings suggest that adaptation is a universal feature of neural information processing and offer a unifying explanation for contextual phenomena in fields ranging from visual psychophysics to economic choice.
CASKETSS: a computer code system for thermal and structural analysis of nuclear fuel shipping casks
International Nuclear Information System (INIS)
Ikushima, Takeshi
1989-02-01
A computer program CASKETSS has been developed for the purpose of thermal and structural analysis of nuclear fuel shipping casks. CASKETSS measn a modular code system for CASK Evaluation code system Thermal and Structural Safety. Main features of CASKETSS are as follow; (1) Thermal and structural analysis computer programs for one-, two-, three-dimensional geometries are contained in the code system. (2) Some of the computer programs in the code system has been programmed to provide near optimal speed on vector processing computers. (3) Data libralies fro thermal and structural analysis are provided in the code system. (4) Input data generator is provided in the code system. (5) Graphic computer program is provided in the code system. In the paper, brief illustration of calculation method, input data and sample calculations are presented. (author)
National Research Council Canada - National Science Library
Ong, Choon
1998-01-01
The performance analysis of a differential phase shift keyed (DPSK) communications system, operating in a Rayleigh fading environment, employing convolutional coding and diversity processing is presented...
Code development of total sensitivity and uncertainty analysis for reactor physics calculations
International Nuclear Information System (INIS)
Wan, C.; Cao, L.; Wu, H.; Zu, T.; Shen, W.
2015-01-01
Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)
Code development of total sensitivity and uncertainty analysis for reactor physics calculations
Energy Technology Data Exchange (ETDEWEB)
Wan, C.; Cao, L.; Wu, H.; Zu, T., E-mail: chenghuiwan@stu.xjtu.edu.cn, E-mail: caolz@mail.xjtu.edu.cn, E-mail: hongchun@mail.xjtu.edu.cn, E-mail: tiejun@mail.xjtu.edu.cn [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Shen, W., E-mail: Wei.Shen@cnsc-ccsn.gc.ca [Xi' an Jiaotong Univ., School of Nuclear Science and Technology, Xi' an (China); Canadian Nuclear Safety Commission, Ottawa, ON (Canada)
2015-07-01
Sensitivity and uncertainty analysis are essential parts for reactor system to perform risk and policy analysis. In this study, total sensitivity and corresponding uncertainty analysis for responses of neutronics calculations have been accomplished and developed the S&U analysis code named UNICORN. The UNICORN code can consider the implicit effects of multigroup cross sections on the responses. The UNICORN code has been applied to typical pin-cell case in this paper, and can be proved correct by comparison the results with those of the TSUNAMI-1D code. (author)
Systems-Dynamic Analysis for Neighborhood Study
Systems-dynamic analysis (or system dynamics (SD)) helps planners identify interrelated impacts of transportation and land-use policies on neighborhood-scale economic outcomes for households and businesses, among other applications. This form of analysis can show benefits and tr...
Development and validation of a fuel performance analysis code
International Nuclear Information System (INIS)
Majalee, Aaditya V.; Chaturvedi, S.
2015-01-01
CAD has been developing a computer code 'FRAVIZ' for calculation of steady-state thermomechanical behaviour of nuclear reactor fuel rods. It contains four major modules viz., Thermal module, Fission Gas Release module, Material Properties module and Mechanical module. All these four modules are coupled to each other and feedback from each module is fed back to others to get a self-consistent evolution in time. The computer code has been checked against two FUMEX benchmarks. Modelling fuel performance in Advance Heavy Water Reactor would require additional inputs related to the fuel and some modification in the code.(author)
Performance analysis of LDPC codes on OOK terahertz wireless channels
International Nuclear Information System (INIS)
Liu Chun; Wang Chang; Cao Jun-Cheng
2016-01-01
Atmospheric absorption, scattering, and scintillation are the major causes to deteriorate the transmission quality of terahertz (THz) wireless communications. An error control coding scheme based on low density parity check (LDPC) codes with soft decision decoding algorithm is proposed to improve the bit-error-rate (BER) performance of an on-off keying (OOK) modulated THz signal through atmospheric channel. The THz wave propagation characteristics and channel model in atmosphere is set up. Numerical simulations validate the great performance of LDPC codes against the atmospheric fading and demonstrate the huge potential in future ultra-high speed beyond Gbps THz communications. (paper)
Exposure calculation code module for reactor core analysis: BURNER
Energy Technology Data Exchange (ETDEWEB)
Vondy, D.R.; Cunningham, G.W.
1979-02-01
The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.
A comparison of LOCA analysis using SMOKIN and CERBERUS codes
Energy Technology Data Exchange (ETDEWEB)
Younis, M H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Gaboury, G [Ontario Hydro, Toronto, ON (Canada)
1996-12-31
This paper presents the results of a comparison of the analyses of a postulated Loss of Coolant Accident (LOCA) in Pickering NGS A reactors using the two neutron kinetics codes SMOKIN and CERBERUS. Both codes have been used to simulate the space-time neutronic behaviour of CANDU-PHWR reactors. The main objective of the present study is to evaluate the accuracy with which SMOKIN can predict power transients compared to CERBERUS. The comparison shows that the two codes produce similar bulk power and reactivity transients. However, SMOKIN was found to overestimate the power transient (relative to CERBERUS) in some regions of the core, which is indicative of the spatial differences between the two codes. It was demonstrated that part of this overestimate is due to the use of reaction-rate averaged fuel properties in SMOKIN, compared to instantaneous fuel properties in CERBERUS. (author). 5 refs., 3 tabs., 6 figs.
Vehicle Codes and Standards: Overview and Gap Analysis
Energy Technology Data Exchange (ETDEWEB)
Blake, C.; Buttner, W.; Rivkin, C.
2010-02-01
This report identifies gaps in vehicle codes and standards and recommends ways to fill the gaps, focusing on six alternative fuels: biodiesel, natural gas, electricity, ethanol, hydrogen, and propane.